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Radiological environmental impact analysis of a 2-MW thorium molten salt reactor during an accident

NUCLEAR ENERGY SCIENCE AND ENGINEERING

Radiological environmental impact analysis of a 2-MW thorium molten salt reactor during an accident

Chang-Qi Chen，
Xiao-Bin Xia，
Zhi-Hong Zhang，
Jun Cai，
Chang-Yuan Li
Nuclear Science and TechniquesVol.30, No.5Article number 78Published in print 01 May 2019Available online 11 Apr 2019
DOI：10.1007/s41365-019-0605-3
48600

The thorium molten salt reactor–liquid fuel (TMSR-LF1) has inherent safety features. The accident occurrence possibility and their consequences are much lower for the TMSR-LF1 than that of traditional reactors. Based on accident analysis, the maximum credible accident and the radioactive source terms of the TMSR-LF1 were first estimated. Then, the total effective dose of the maximum credible accident was calculated. Based on calculations, the cover gas flow rate can significantly affect the radiation consequences of the maximum credible accident when it changes from 0 to 10 L/min. If no cover gas is flowing, a site area emergency would be required within the range of 50–73 m from the reactor. In the case of cover gas flow, only an abnormal notification and an alert two emergency class would be required within the range of 50 m.

TMSR-LF1Accident classification and descriptionMaximum credible accidentEmergency class

1 INTRODUCTION

The molten salt reactor (MSR) is among the six advanced reactor types proposed in the Generation IV International Forum [1]. The Chinese Academy of Sciences launched a thorium molten salt reactor–liquid fuel (TMSR-LF1), a type of MSR, to develop the fourth generation advanced nuclear energy technology [2]-[5]. From a safety perspective, the TMSR-LF1 has several characteristics: a negative feedback mechanism, no risk of fuel melt, low pressure operation, cooling without water, an online cover-gas purge system, and a redundant passive air-cooled system [6]-[9].

The accident occurrence possibility and their consequences are much lower for the TMSR-LF1 than that of traditional reactors. For increased safety, a systematic and complete radiological environmental impact should be analyzed to support nuclear emergency preparedness and response for reactors.

A number of mature studies of radiological environmental impact have been conducted on nuclear power plants [10]-[13]. Studies of the radiological environmental impact of research reactors are fewer than those of nuclear power plants because of the complex reactor types. Sutton’s formula was used in radiological environmental impact analysis for the Molten Salt Reactor Experiment (MSRE) [14]. As time has passed, Sutton’s formula has rarely been used in radiological environmental impact analysis. A systematic restudy of the radiological environmental impact of MSRs is necessary.

This study attempted to determine the maximum credible accident and its consequence among all possible nuclear accidents of the TMSR-LF1, assess the radioactive severity of the consequences of the maximum credible accident, and discuss the corresponding emergency preparedness for the TMSR-LF1.

2 INVENTORY OF RADIONUCLIDES IN THE TMSR-LF1

2.1 Radionuclides produced in core

Radioactive materials in the fuel molten salt of the TMSR-LF1 mainly originate from three sources as follows:

1) Fission products from the core fuel.

2) Neutron activation products from the molten salt and impurities in it.

3) Neutron activation of corrosion products from pipe and heat exchanger structural materials.

The TMSR-LF1 parameters are shown in Table 1. Considering the aforementioned three sources and cover gas purges, the inventory of radionuclides in the core fuel of the TMSR-LF1 after 60 days continuing full power operation can be calculated using SCALE6.1[15] as shown in Table 2.

Table 1.
TMSR-LF1 parameters
Parameters Value
Thermal power (MW) 2
Lifetime (years) 10
Effective full power days (days/year) 60
Fuel composition LiF-BeF2-ZrF4-UF4
Fuel salt load (m3) 1.68
7Li abundance (at%) >99.95
235U enrichment (wt%) 19.75
Fuel inlet temperature (℃) 630
Fuel outlet temperature (℃) 650
Volume
Fuel salt volume (m3) 1.68
Cover gas volume (m3) 1.6
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Table 2.
Inventory of radionuclides in the fuel salt of TMSR-LF1 after 60 days of full power operation (Bq)
Nuclide Activity Nuclide Activity Nuclide Activity Nuclide Activity
Fission products
88Rb 2.21×1015 106Rh 2.88×1013 132I 2.44×1015 142La 3.67×1015
89Sr 1.67×1015 106Ru 2.71×1013 132Te 2.70×1015 143Pr 3.42×1015
90Sr 1.43×1013 110mAg 1.14×108 133I 3.79×1015 144Ce 4.69×1014
90Y 1.36×1013 115mCd 2.58×1011 133Te 2.33×1015 144Pr 4.69×1014
91Sr 3.64×1015 123Sn 1.81×1011 134Cs 3.92×1010 147Nd 1.38×1015
91Y 1.84×1015 125Sb 8.12×1011 134I 4.42×1015 147Pm 4.46×1013
92Sr 3.73×1015 125Sn 6.20×1012 135I 3.55×1015 148Pm 1.17×1012
95Nb 8.86×1014 125mTe 5.26×1010 136Cs 2.80×1012 148mPm 3.71×1011
95mNb 1.99×1013 127Sb 9.88×1013 137mBa 1.39×1013 151Sm 3.19×1011
95Zr 1.97×1015 127Sn 6.01×1013 137Cs 1.45×1013 154Eu 1.40×109
97Zr 3.83×1015 127Te 8.78×1013 138Cs 4.22×1015 155Eu 4.65×1011
99Mo 3.85×1015 129Te 3.13×1014 139Ba 4.02×1015 156Eu 9.63×1012
99mTc 3.39×1015 129mTe 4.05×1013 140Ba 3.75×1015 160Tb 8.60×107
103mRh 1.23×1015 131I 1.60×1015 140La 3.73×1015 239Am 1.75×10-3
103Ru 1.25×1015 131Te 1.63×1015 141Ce 2.64×1015 240Am 3.06×10-1
242Am 1.20×106 131mTe 2.47×1014 141La 3.67×1015 241Am 4.48×104
242mAm 2.23×101 243Cm 9.41×100 239Pu 5.01×1010 233U 1.76×102
243Am 3.07×10-1 244Cm 1.52×101 240Pu 5.07×108 234U 1.70×105
244Am 2.48×100 245Cm 1.83×10-5 241Pu 7.08×108 235U 3.78×109
244mAm 3.89×101 246Cm 1.86×10-8 242Pu 2.17×101 236U 6.53×107
245Am 3.96×10-7 237Np 2.11×105 243Pu 9.16×104 237U 6.34×1012
241Cm 1.09×10-5 239Np 1.11×1016 232U 1.18×104 238U 2.51×109
242Cm 4.67×104 236Pu 6.57×101 240Pu 5.07×108 239U 1.12×1016
83mKr 3.38×1014 238Pu 3.47×106 135Xe 3.69×1015 133I 4.21×1015
85Kr 1.97×1012 88Kr 2.19×1015 135mXe 7.63×1014 134I 4.91×1015
85mKr 7.98×1014 131mXe 1.80×1013 131I 1.78×1015 135I 3.94×1015
87Kr 1.63×1015 133Xe 3.99×1015 132I 2.71×1015 133mXe 1.11×1014
Neutron activation of corrosion products
51Cr 2.27×1012 58Co 1.89×1010 54Mn 1.60×109 65Ni 7.81×109
56Mn 4.38×1012 55Fe 3.84×1010 59Ni 1.36×107
60Co 1.42×1011 59Fe 1.61×107 63Ni 1.66×109
Neutron activation of impurities
8Li 1.29×1015 16N 5.28×1014 3H 1.57×1014 20F 9.31×1014
6He 3.23×1014 19O 3.69×1013 14C 6.14×108
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2.2 Inventory of radionuclides in cover gas

Experimental data on the MSRE show that, under normal conditions, most of the noble gases via fission in the molten salt would diffuse into the cover gas. A small amount (less than 1%) of iodine would diffuse into the cover gas from the fuel molten salt [16]. In addition, approximately 50% of the tritium would diffuse into the cover gas [17].

Figure 1 shows the major radionuclide migration path in the TMSR-LF1. Under the normal operational condition of the TMSR-LF1, fuel mainly originates from fission and neutron activation products. All the noble gas and a portion of the iodine produced by the fission reaction would diffuse from the fuel salt into the cover gas system. The TMSR-LF1 uses argon gas as a cover gas. The cover gas also can be neutron activated. During reactor operation, cover gas would flow through the fuel salt into the exhaust system. Finally, the treated radioactive gas is discharged into the environment through the chimney.

Fig 1.
Major radionuclide migration path in the TMSR-LF1
pic

According to the design of the TMSR-LF1, the cover gas flows at a constant rate. It is assumed that 100% of the noble gas and 10% of the iodine would enter the cover gas from the fuel salt. In addition, among all the neutron activation products listed in Table 2, only 14C and tritium would release from the molten salt into the cover gas. Conserve assume that all 14C would enter the cover gas in the form of CO2.

Tritium release should be a concern as a large amount of tritium is generated through the neutron adsorption of lithium [18]. Tritium is absorbed and desorbed by graphite and permeates through structural materials [19]. In the TMSR-LF1, most tritium permeates through the metal walls to the surroundings at high temperatures [20]. Except for the part absorbed by graphite, tritium does not concentrate somewhere in the reactor for very long. Approximately 91.76% of the tritium continuously permeates out to the environment during the whole operational period via the TMSR-TTAC [21] code. Therefore, radiological environmental impact of tritium would be assessed under normal operation and is beyond the scope of this article.

Radioactive materials pass from the molten salt fuel to the cover gas and decrease with the cover gas flowing. The atomic concentration Ci(t) (atomic number/L) of a nuclide i in the cover gas is as follows:

dCi(t)dt=qiV−gVCi(t)−λiCi(t), (1)

where qi (atomic number/s) is the average rate of nuclide i passing from the fuel salt into the cover gas; V (L)is the volume of the cover gas; λi (s-1) is the decay constant of the nuclide i; and g (L/s) is the cover gas flow rate.

The equation can be solved as follows:

Ci(t)=qig+λiV(1−e−(λi+gV)t). (2)

The equilibrium concentration of radionuclide i in the cover gas is as follows:

Ci(t)=qig+λiV. (3)

The time the nuclides pass from the fuel salt into the cover gas and the deposition of their decay daughters in the gas pipeline are ignored in the calculation.

The cumulative number of radioactive materials in the cover gas is related to the cover gas flow rate. Assuming that the cover gas flow rate is 4 L/min, 10 L/min, 50 L/min, 100 L/min, and 0 L/min, respectively, the inventory of radionuclides in the cover gas of the TMSR-LF1 after 60 days continuing full power operation was calculated and is shown in Table 3.

Table 3.
Inventory of radionuclides in the cover gas of the TMSR-LF1 after 60 days of full power operation (Bq)
Nuclide Flow rate (L/min)
0 4 10 50 100
14C 6.14×108 2.72×105 1.09×105 2.18×104 1.09×104
37Ar 1.15×1010 9.08×107 3.65×107 7.30×106 3.65×106
39Ar 2.62×105 1.22×103 4.87×102 9.74×101 4.87×101
41Ar 6.62×1011 4.75×1011 3.33×1011 1.11×1011 6.08×1010
83mKr 2.00×108 1.43×108 9.98×107 3.32×107 1.81×107
85Kr 1.89×1011 8.75×108 3.50×108 7.00×107 3.50×107
85mKr 5.36×1011 2.72×1011 1.57×1011 4.09×1010 2.12×1010
87Kr 3.98×1013 3.13×1013 2.36×1013 8.98×1012 5.06×1012
88Kr 1.49×1014 9.29×1013 5.93×1013 1.74×1013 9.24×1012
131mXe 3.44×108 3.28×105 1.00×105 1.66×104 8.06×103
133Xe 7.74×1011 6.52×109 1.64×109 2.00×108 9.10×107
133mXe 1.81×1011 1.32×1010 5.48×109 1.12×109 5.59×108
135Xe 2.78×1013 5.13×1012 1.88×1012 3.00×1011 1.42×1011
135mXe 1.92×1013 1.58×1013 1.40×1013 8.94×1012 6.28×1012
131I 3.33×1010 7.84×108 3.18×108 6.41×107 3.21×107
132I 7.79×1010 5.22×1010 3.49×1010 1.09×1010 5.85×109
133I 7.39×1011 1.33×1011 5.98×1010 1.28×1010 6.45×109
134I 4.29×1012 3.59×1012 2.89×1012 1.26×1012 7.39×1011
135I 2.54×1013 1.04×1013 5.51×1012 1.33×1012 6.84×1011
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3 ACCIDENT CLASSIFICATION AND DESCRIPTION OF THE TMSR-LF1

The main reactor structures of the TMSR-LF1 are shown in Fig. 2. All the fuel salt is contained in the reactor vessel. Outside of the reactor vessel are a safety vessel and concrete containment. The cavity is divided into an upper and lower cavity. The entire cavity is belowground and the reactor vessel is in the lower cavity.

Fig 2.
Schematic of main reactor structures in the TMSR-LF1
pic
3.1 Accident Classification of the TMSR-LF1

All possible TMSR-LF1 accidents are listed in Table 4. According to the previous analysis in chapter 2, TMSR-LF1 radioactivity is concentrated in the fuel salt and cover gas. Thus, the integrity of the fuel and cover gas boundaries determines the possibility of a radioactive release occurring. From this perspective, fuel salt and cover gas leaks undoubtedly need attention. In addition, unexpected reactivity initiation and a decrease in the core heat removal are both suspected to overheat the reactor core and may damage the core boundary. The four accident types are described in detail in the next section. As for unexpected natural disasters such as earthquakes, the possibility of occurrence is too low to be considered in this study.

Table 4.
Accident classification and initiating conditions of the TMSR-LF1 [22]
Accident Classification Initiating Condition
Reactivity Initiated Accident An uncontrolled rod withdrawal under sub criticality or low power, an uncontrolled rod withdrawal under power operation, Accidental criticality during the process of charging
Core Heat Removal Decrease Accident Molten salt pump rotor seizure accident, Station blackout, Loss of heat sink, Air cooling fan failure
Core Heat Removal Increase Accident Increase in cooling salt flow, Decrease in cooling salt temperature
Facility Leakage Fuel salt leak, Cover gas leak
Anticipated Transient Without Scram (ATWS) Loss of outer power without scram, Uncontrolled rod withdrawal without scram
Hazards and Other Conditions Affecting Reactor Earthquake, Flooding, Explosion, Fire
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3.2 Reactivity Initiated Accident and Core Heat Removal Decrease Accident

A reactivity initiated accident is an important part of reactor safety analysis because the consequence of initiating reactivity transient affects reactor safety. The reactivity coefficients of the TMSR-LF1, such as the fuel temperature reactivity, reactivity power, and graphite reactivity temperature coefficients, are negative[9], [23]. The negative feedback mechanism enables the TMSR-LF1 to maintain a steady state. During all the transient conditions, the maximum temperature of the fuel salt and graphite are far below the temperature limit of the material. Therefore, a reactivity initiated accident would not damage the core boundary and release radioactivity.

An extended and verified RELAP5 code was used to perform transient safety analysis for the TMSR-LF1 [8]. The results show that the temperature of the TMSR-LF1 would not exceed the safety limits given a decrease in core heat removal. For example, when a station blackout accident occurred, the increased core temperature would be 4℃. Therefore, core heat removal decrease accident would not damage the core boundary and release radioactivity.

3.3 Fuel salt leak

In the MSRE, it is considered credible that a rupture of pipelines and facilities can lead to a release of molten salt fuel from the fuel circulation system into the reactor containment. Based on the safety analysis of the MSREMSRE, its maximum credible accident is a simultaneous molten salt and water leak. High-temperature molten salt and water would form mixed steam and maximize the secondary container pressure. No more than 4 L of molten salt would finally leak during the maximum credible accident [14]. After molten salt fuel leakage from the fuel circle system, most would be contained in containment. Only gaseous nuclides can pass through containment and release into the building, including 100% of the noble gas, 10% of the iodine with a 50% plate out on the secondary container surfaces, and 10% of the fissile metals will disperse from the molten salt into the air. The radioactive release from the container to the building would take 4 h [14].

Assume the fuel molten salt would leak from a rupture of equipment or weld seams and the maximum leakage of the fuel salt is 4 L. There is no water in the TMSR-LF1 so a mixed steam of molten salt and water would not appear following a fuel salt leak. Therefore, only a molten salt leak would not result in too much pressure on the reactor or safety vessels [14].

Because of the influence of gravitational potential energy, molten salt running from the fuel cycle will flow downward after the accident occurs. Except for extreme cases such as earthquakes, run-off molten salt can hardly erupt upward. The run-off molten salt would remain in the reactor vessel with high probability, not affecting the safety vessel and outside environment. In consideration that there is no historical record of destructive earthquakes at the plant site of the TMSR-LF1 and surrounding area, particularly no digital record from instruments after 1970, this study did not consider the occurrence of earthquakes [24].

Another scenario is that leaking molten salt has the rare chance to break through the physical barrier of the reactor vessel into the lower cavity. The melting point of the fuel salt of the TMSR-LF1 is approximately 450℃ and the lower cavity temperature is lower than this melting point; leaking molten salt would coagulate as the temperature decreases. The heat dissipating capacity of the cavity is 40 Kw [24], assuming that 90% of the heat dissipating capacity is used in core residual heat discharge and 10% is used to discharge heat from the leaking molten salt. The leaking fuel salt would coagulate from 650℃ to 450℃ in approximately 15 min. Radioactive nuclides stored in the run-off fuel salt would have difficultly releasing in 15 min as well as residual fission products and decay products. Outside of the lower cavity are a safety vessel and concrete containment to stop radioactive materials. Therefore, a fuel salt leak accident can hardly result in a radiation impact on the building and external environment.

3.4 Cover-gas leak

In the TMSR-LF1, an online cover-gas purge system can remove radioactive nuclides from the molten salt fuel to the cover gas during reactor operation. A mass cumulated number of radioactive nuclides may exist in the cover gas. Thus, the complete release of radioactive materials in the cover gas should be carefully considered and analyzed [10], [22].

Suppose all the cover gas leaks from a rupture of equipment or weld seams into the safety vessel at maximum fuel burnup. When the low-pressure signal of the cover gas system reaches the reactor trip set points, reactor control rods will fall for scram in 30 s (or be manually shutdown by operators according to the monitoring display). At the same time, the residual heat removal system starts and the residual heat of the core will be derived.

During this accident, new radioactivity would be produced during the 30 s scram process and then enter in the cover gas from the molten salt in the same ratio as that mentioned in section 2.2. These radionuclides passing into the cover gas are listed in Table 5.

Table 5.
Radioactivity produced during 30-s scram process (Bq)
Nuclide Activity Nuclide Activity Nuclide Activity
83mKr 5.90×107 131mXe 4.44×103 131I 8.77×106
85Kr 1.02×106 133Xe 1.98×107 132I 1.45×109
85mKr 2.47×1010 133mXe 1.30×108 133I 2.76×109
87Kr 2.27×1012 135Xe 3.26×1010 134I 2.25×1011
88Kr 3.25×1012 135mXe 2.51×1012 135I 2.34×1011
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All the radioactive materials stored in the cover gas system and those produced during the scramming process are first released into the safety vessel. The leakage rate of the safety vessel to the building hall is 5 vol%/d; all the gaseous radioactive materials will release into building following the accident within 24 h. Because the number of radioactive materials in cover gas is related to the gas flow rate, the source terms of the cover gas leak accident are also closely related to the flow rate. In addition, as xenon could be produced by the decay of iodine, radioactive decay was considered in the calculation. The source terms of the cover gas leak releasing into the building at each flow rate are listed in Table 6. The main source terms are krypton, xenon, and iodine.

Table 6.
Source terms of the cover gas leak of the TMSR-LF1
Nuclide Flow rate (L/min)
0 4 10 50 100
14C 3.07×107 1.32×104 5.44×103 1.09×103 5.44×102
37Ar 5.72×108 4.52×106 1.82×106 3.64×105 1.82×105
39Ar 1.31×104 6.11×101 2.43×101 4.88×100 2.43×100
41Ar 7.71×109 5.53×109 3.88×109 1.30×109 7.07×108
83mKr 2.75×106 2.07×106 1.57×106 7.85×105 6.06×105
85Kr 9.45×109 4.38×107 1.75×107 3.55×106 1.80×106
85mKr 1.26×1010 6.60×109 3.94×109 1.29×109 8.32×108
87Kr 3.95×1011 3.18×1011 2.52×1011 1.23×1011 8.86×1010
88Kr 2.48×1012 1.55×1012 1.00×1012 3.17×1011 1.83×1011
131mXe 1.72×107 2.23×104 7.55×103 1.51×103 8.46×102
133Xe 3.87×1010 4.96×108 1.61×108 2.77×107 1.39×107
133mXe 8.47×109 6.24×108 2.63×108 5.80×107 3.18×107
135Xe 1.18×1012 2.70×1011 1.16×1011 2.34×1010 1.21×1010
135mXe 1.50×1011 1.17×1011 8.81×1010 4.91×1010 3.51×1010
131I 1.63×109 3.88×107 1.60×107 3.54×106 1.97×106
132I 1.13×109 7.38×108 4.97×108 1.63×108 9.32×107
133I 3.07×1010 5.60×109 2.57×109 6.23×108 3.60×108
134I 2.98×1010 2.55×1010 2.07×1010 9.48×109 5.92×109
135I 7.35×1011 3.05×1011 1.63×1011 4.28×1010 2.42×1010
Total 5.08×1012 2.61×1012 1.65×1012 5.68×1011 3.51×1011
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4. MAXIMUM CREDIBLE ACCIDENT

According to the aforementioned accident analysis of the TMSR-LF1, among all the possible accident types, only the cover gas leak may lead to serious radioactive release. Therefore, the cover gas leak was chosen as the maximum credible accident.

The 24-h cumulative dose at the site boundary of an individual without any protection measures after an accident occurs should be considered when developing an emergency plan for a research reactor [25]. The 24-h total effective dose is considered as a sum of the external exposure from plume immersion and internal exposure from inhalation. Calculated formulas are as follows [26]:

Hp=∑iCi⋅(χ/Q)⋅t⋅DCFp,i⋅Sp, (4) Hb=∑iCi⋅(χ/Q)⋅t⋅DCFb,i⋅B⋅Sb, (5)

where H (Sv) is the cumulative radiation dose; Ci (Bq) is the release concentration of a nuclide i; χ/Q (s·m-3) is the atmospheric diffusion factor; t (s) is the diffusion time; DCFp,i (Sv·(Bq·s·m−3)−1) is the dose coefficient for air submersion; DCFb,i (Sv·Bq−1) is the dose coefficient for inhalation; B (m3·s−1) is the respiration rate; and S (l)is the shielding factor.

The aforementioned dose coefficients of each radionuclide for two exposure paths that are used in calculation were taken from standards and guidance [27]-[29]. Adult respiratory rates were derived from Regulatory Guide 1.195 from the U.S [30].

The ARCON96 code was chosen to calculate atmospheric diffusion factors [31],[32]. The ARCON96 code implements a straight-line Gaussian model with building wake and low wind speed corrections. These two corrections can effectively correct the dispersion consequence of the Gaussian model in power plant area [31],[33].

In the site plan, the distance from the reactor to the nearest site boundary is 73 m and the farthest distance is 942 m, thus the calculation range was from 50 m to 1000 m. Some site limit atmosphere diffusion factors are shown in Table 7 using meteorological data of the plan site.

Table 7.
Atmospheric diffusion factors of the plan site (s·m-3)
Distance (m) 0-2 h 2-8 h 8-24 h
50 1.31×10-2 1.11×10-2 5.48×10-3
73 5.93×10-3 5.20×10-3 2.55×10-3
100 3.54×10-3 2.96×10-3 1.48×10-3
150 2.28×10-3 1.95×10-3 9.66×10-4
200 1.10×10-3 9.66×10-4 4.78×10-4
250 6.27×10-4 5.27×10-4 2.65×10-4
300 2.96×10-4 2.48×10-4 1.25×10-4
400 1.75×10-4 1.49×10-4 7.47×10-5
500 1.20×10-4 9.97×10-5 5.05×10-5
1000 8.78×10-5 7.39×10-5 3.73×10-5
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Calculation doses at the nearest site boundary (d=73 m) for each cover gas flow rate are shown in Table 8.

Table 8.
24-h total effective dose of the maximum credible accident at the nearest site boundary (mSv)
Flow rate（L/min） Immersion Inhalation Total
0 1.53 0.958 2.49
4 0.910 0.361 1.27
10 0.590 0.194 0.78
50 0.197 0.052 0.250
100 0.119 0.030 0.149
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From Table 8, the plume immersion exposure is the most influential exposure path. Its proportion is greater than 70%. The second highest proportion is inhalation exposure at greater than 20%.

Some concerning calculation doses in Table 8 are shown during different time periods in Table 9. The statistics of effective doses produced during three time periods following the maximum credible accident show their contributions to the 24-h total effective dose are approximately 29%, 51%, and 19%, respectively. As a simple average, the suffered dose per hour from 0 to 2 h phase accounted for 14.5% of the 24-h cumulative dose, the suffered dose per hour from 2 to 8 h phase accounted for 8.5%, and the suffered dose per hour from 8 to 24 h phase accounted for approximately 1.2%.

Table 9.
Effective dose of the maximum credible accident at the nearest site boundary (d=73 m) for different times (mSv)
Flow rate（L/min） 0-2 h 2-8 h 8-24 h 0-24 h
0 0.731 1.28 0.475 2.49
4 0.399 0.65 0.218 1.27
10 0.255 0.401 0.128 0.78
50 8.67×10-2 1.27×10-1 3.60×10-2 0.250
100 5.29×10-2 7.65×10-2 1.98×10-2 0.149
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5. DISCUSSION

5.1 Influence of the cover gas flow rate

Different cover gas flow rates would lead to different source terms for the maximum credible accident and then produce an effect on the maximum credible accident consequence, thus the influence of cover gas flow rate is worth discussing. Figure 3 (a) shows the proportions of some radionuclide inventories in the cover gas at each flow rate. Whether conducting cover gas flow purging or not has an obvious influence on the proportions of Kr and Xe; cover gas flow purging can increase the proportion of Kr and decrease the proportion of Xe. When the cover gas flow continues purging, the proportions of Kr and I both decrease as the cover gas flow rate increases, while the proportion of Xe decreases in an opposite trend. This is because the radioactivity concentration is connected to the decay constant of radionuclides. Long-lived radionuclides can be more easily purged by gas flow and short-lived radionuclides can be more easily stored in the cover gas. For Xe, the 135Xe and 135mXe isotopes both are short-lived radionuclides thus their concentrations would more slowly decrease with a flow rate increase than other Xe isotopes. This leads to the proportion of Xe increasing as the cover gas flow rate increases.

Fig 3.
Influence of cover gas flow rate: (a) Proportions of nuclide inventories in the cover gas; (b) Proportions of various nuclides in the release source terms
pic

Figure 3 (b) shows that the proportions of various nuclides in the release source terms are somewhat different from the proportions of the nuclide inventories in the cover gas. The basic trend of the lines in the two images is largely the same. The Kr and Xe proportion lines have inflection points at 10 L/min, the probable reason being that radioactive decay is considered in the source calculation.

Figure 4 (a) shows the effect of the cover gas flow rate on the contribution proportions of different types of radionuclides. Comparing Fig. 3 (b) to Fig. 4 (a), the variation tendency of the dose contribution proportions of the radionuclides mostly accords with the trend of the nuclide proportions in the source terms except krypton. The variation trend of the dose contribution proportions of the krypton is opposite to that of its proportions in the source terms. This is because of all the isotopes of krypton 87Kr and 88Kr are the most common and their dose conversion factors are several orders of magnitude higher than those of the other isotopes. Their contributions to the effective dose rank in the first position and the third position, respectively. Thus, the range of the dose contribution growth of krypton is greater than the reduction of the range of the proportion in the source term, and the contribution of krypton to the effective dose would increase as cover gas flow rate increases.

Fig 4.
Contribution Proportions of 24-h total effective dose: (a) Proportions of radionuclide various types; (b) Proportions of exposure paths
pic

Figure 4 also shows the effect of the cover gas flow rate on the contribution proportions of exposure paths in image (b). With the increase in flow rate, the contribution of plume immersion irradiation increases from approximately 71% to 79% and the proportion of inhalation radiation gradually decreases from approximately 29% to 21%.

The influence of the cover gas flow rate on 24-h total effective dose is shown in Figure 5. Obviously, the higher the gas flow rate, the smaller the total effective dose. The trend line smoothly slopes during the final stage showing that the influence of the cover gas leak accident on the maximum credible accident is increasingly less. This is because the radionuclide inventory in the cover gas would decrease as the cover gas flow rate increases.

Fig 5.
Influence of cover gas flow rate at 73 m
pic

The point of inflection on this curve appears at approximately 10 L/min, and above 100 L/min the curve is straight which means the influence of the cover gas flow rate is slight. When the cover gas flow rate is changing from 0–10 L/min, the rate has a significant effect on the 24-h total effective dose of the maximum credible accident. The cover gas flow rate also affects the 24-h total effective dose from 50–100 L/min.

5.2 Influence of radioactivity produced during the scram process

In the source term calculation of the maximum credible accident, radioactive products originating from the scramming process (30 s) were considered. Therefore, these radionuclides affect the dose calculation. The 24-h total effective dose of the source terms generated during the 30-s scram process are provided in Table 10.

Table 10.
Doses during the scram process
Distance (m) Dose (mSv) Distance (m) Dose (mSv)
50 7.28×10-2 250 3.47×10-3
73 3.37×10-2 300 1.64×10-3
100 1.95×10-2 400 9.77×10-4
150 1.27×10-2 500 6.60×10-4
200 6.25×10-3 1000 4.87×10-4
Show more

Comparing Table 10 to Table 8, some dose contributions of the source terms produced by scramming at the nearest site boundary are provided in Table 11.

Table 11.
Dose contributions of source terms produced during the scram process at each cover gas flow rate
Flow rate (L/min) Contribution Flow rate (L/min) Contribution
0 1.35% 50 13.5%
4 2.65% 100 22.6%
10 4.32%
Show more

As can be seen from Tables 10 and 11, the numerical value of the doses caused by scramming, whose order of magnitude is no greater than 10-2 mSv, is extremely small under the power level of the TMSR-LF1. However, the source terms produced during the scramming process contribute more than 1% of the 24-h total effective dose of the maximum credible accident. In the case the scram is a failure and manual shutdown is needed, the shutdown time will increase and its dose contribution will also increase.

Dose contributions of source terms produced during the scramming process are also related to the cover gas flow rate. The higher the flow rate, the greater the dose contribution. This means scram duration would have a greater impact on the consequences of the maximum credible accident as flow rate increases. Therefore, an excessive cover gas flow rate is not necessary because it may reduce the reliability of the safety system.

5.3 Guidance for emergency preparedness

After dose analysis of the maximum credible accident, an emergency class designation of the accident could be the next step. Emergency class is among a minimum set of names or titles for grouping abnormal nuclear reactor conditions according to their relative radiological seriousness; onsite and offsite radiological emergency preparedness actions are necessary to respond to such conditions [27]. The existing radiological emergency classes, in descending order of seriousness, are the emergency classes and corresponding emergency action levels in the Chinese national guidance as shown in Table 12 [25].

Table 12.
Emergency action level for a research reactor
Emergency classification Emergency action level
Abnormal notification Release of gaseous or liquid radioactivity resulting in a site boundary of a 24-h total effective dose greater than 0.15 mSv
Alert Release of gaseous or liquid radioactivity resulting in a site boundary of a 24-h total effective dose greater than 0.75 mSv
Site-area emergency Release of gaseous or liquid radioactivity resulting in a site boundary of a 24-h total effective dose greater than 3.75 mSv
General emergency Release of gaseous or liquid radioactivity resulting in site boundary of a 24-h total effective dose greater than 10 mSv
Show more

Emergency classes for the maximum credible accident under different cover gas flow rate are shown in Fig. 6.

Fig 6.
Emergency classes for the maximum credible accident under different cover gas flow rates
pic

In Fig. 6, at the same location, the emergency class will be lower if the cover gas flow rate is higher. When the cover gas flow rate is greater than 50 L/min, only an abnormal notification would be needed. When the cover gas flow rate is 4–10 L/min, an alert would be needed. When there is no flow rate, a site-area emergency would be needed.

Therefore, when the cover gas flow rate is no less than 4 L/min, a site-area emergency classification can be abolished outside 50 m from reactor in the emergency plan for the TMSR-LF1. Under the same emergency class, if a closer distance is desired, the cover gas flow rate needs to be increased. For this same reason, if a lower flow rate is desired, the distance from the reactor to the building boundary needs to increased.

6. CONCLUSION

All the possible accidents of the TMSR-LF1 were analyzed and show that only a cover gas leak would lead to a radiological release to the environment. A cover gas leak was chosen as the maximum credible accident of the TMSR-LF1.

Based on the effective dose calculation of the maximum credible accident, when the cover gas flow rate is 0 L/min, the 24-h total effective dose is 0.04–5.30 mSv. When the cover gas flow rate is 4 L/min, the 24-h total effective dose is 0.02–2.49 mSv. When cover gas flow rate is 10 L/min, the 24-h total effective dose is 0.01–1.63 mSv. When the cover gas flow rate is 50 L/min, the 24-h total effective dose is 3.14E-03–0.47 mSv. When the cover gas flow rate is 100 L/min, the 24-h total effective dose is 1.68E-03–0.25 mSv.

The cover gas flow rate can affect the radiation consequence of the maximum credible accident. When the cover gas flow rate changes from 0 to 10 L/min, the cover gas flow rate has a significant effect on the 24-h total effective dose of the maximum credible accident. The cover gas flow rates are also related to the dose contributions of the source terms produced during the scramming process. When the flow rate is 10 L/min, the dose contribution of the source terms produced during the scramming process is approximately 4.32%. Therefore, 10 L/min would be an adaptive cover gas flow rate.

Comparison of the 24-h total effective dose of the maximum credible accident and national guidance for a research reactor emergency, without cover gas flow, indicates a site-area emergency class would be needed within the 50–73-m range from the reactor. In the case of cover gas flow, only an abnormal notification and an alert two emergency class would be required more than 50 m from the reactor.

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Recommended Articles

1 INTRODUCTION

2 INVENTORY OF RADIONUCLIDES IN THE TMSR-LF1

2.1 Radionuclides produced in core

2.2 Inventory of radionuclides in cover gas

3 ACCIDENT CLASSIFICATION AND DESCRIPTION OF THE TMSR-LF1

3.1 Accident Classification of the TMSR-LF1

3.2 Reactivity Initiated Accident and Core Heat Removal Decrease Accident

3.3 Fuel salt leak

3.4 Cover-gas leak

4. MAXIMUM CREDIBLE ACCIDENT

5. DISCUSSION

5.1 Influence of the cover gas flow rate

5.2 Influence of radioactivity produced during the scram process

5.3 Guidance for emergency preparedness

6. CONCLUSION

References

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Nuclear Science and Techniques
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