Keywords

1 Introduction

The steam generator tube rupture (SGTR) is one of the design basis accidents with high frequency and great impact during the life of the pressurized water reactor nuclear power plant, and it is also the focus of domestic nuclear safety supervision and review agencies. After SGTR, the water and steam in secondary coolant were radioactively polluted due to the release of primary coolant containing radionuclides to the secondary coolant through the break in the tube of steam generator. The radioactively polluted secondary coolant steam is released to the environment through the condenser extraction or the safety valve of the steam generator, resulting in radioactive contamination of the environment [1]. Therefore, it is great significance to carry out the SGTR source term analysis conservatively and reasonably.

This study fully investigates the SGTR source term analysis methods of M310 and Hualong 1 (ACP1000). By comparing the design differences of M310, ACP1000 and ACP100, according to the calculation assumptions and parameters of M310 and ACP1000, combined with the design characteristics of ACP100, proposed the SGTR source term analysis method of ACP100, and an ACP100 nuclear power unit SGTR source term analysis and radiological consequence evaluation are carried out using this method.

2 Calculation Assumptions and Parameters

In this study, the calculation assumptions and parameters of M310, ACP1000 and ACP100 regarding the primary coolant activity and the iodine carryover coefficient in the steam generator were compared and analyzed.

2.1 Primary Coolant Activity

In the accident condition, the primary coolant activity will have an activity peak phenomenon. At this time, the primary coolant activity is in a transient value, and transient assumptions is related with steady state value before the accident.

2.1.1 Primary Coolant Activity in Steady State

SGTR source term analysis of M310, the primary coolant activity in steady state is normalized at 4.44 GBq/t in I-131 equivalent, it is the maximum value of 200 reactor-year operations in a French nuclear power plant [2]. This value is not conservative enough to be lower than the operating limit of primary coolant activity [3]. SGTR source term analysis of ACP1000, the primary coolant activity in steady state is normalized at 37 GBq/t DE I-131, it is assumed that the fuel element cladding failure rate is 0.25%. This value corresponds to the technical specification the maximum radioactivity limit condition is conservative enough [4]. Therefore, ACP100 refers to assumption of ACP1000, and the primary coolant activity in steady stat is assumed that the fuel element cladding failure rate is 0.25%, and the calculation results are normalized at 7.4 GBq/t DE I-131, as shown in Table 1.

2.2 Primary Coolant Activity in Transient State

2.2.1 Primary Coolant Noble-Gas Activity in Transient State

Primary coolant noble-gas activity in transient state of M310 and ACP1000 are the steady state value multiplied by the crest factor, the crest factor is taken from the operating experience of the French nuclear power station. The main factors affecting the crest factor are the reactor power and operating pressure. M310, ACP1000 and ACP100 have different reactor power, but the fuel assemblies used are of the same type, the mechanism of fission products released from damaged fuel rods to the primary coolant is the same, and the primary coolant system pressure is similar, so the primary coolant noble-gas activity in transient state calculation of ACP100 can use the same crest factor as that of M310 and ACP1000. The crest factor is shown in Table 1.

2.2.2 Primary Coolant Iodine Activity in Transient State

For M310, the primary coolant iodine activity in transient state is the steady state value multiplied by the crest factor. ACP1000 considers two case of preaccident iodine spike and concurrent iodine spike, referring to the assumptions in RG1.183 [5]. For preaccident iodine spike case, primary coolant iodine concentration to the maximum value (typically 60 μCi/gm DE I-131) peimitted by the technical specification. For concurrent iodine spike, the increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 μCi/gm DE I-131) specified in technical specifications, the assumed iodine spike duration should be 8 h. Through comparative analysis, the iodine peak release phenomenon in the SGTR source term analysis of M310 considers the transient state before the accident, which is similar to the preaccident iodine spike case of ACP1000, but the M310 does not consider the peak release phenomenon similar to concurrent iodine spike case of ACP1000.

According to the stipulations in Analysis criterion of the design basis accident source terms for pressurized water reactor nuclear power plant(NB/T 20444-2017RK) [6], the SGTR source term analysis in new pressurized water reactor nuclear power plants should consider the preaccident iodine spike case and concurrent iodine spike case, SGTR preaccident iodine spike case accident is a limit accident, and SGTR concurrent iodine spike case accident is a rare accident. Therefore, the ACP100 SGTR accident source term analysis intends to consider preaccident iodine spike case and concurrent iodine spike case.

  1. 1)

    Primary coolant activity in preaccident iodine spike

    If ACP100 directly refers to the assumption of RG1.183, it is obviously too conservative to increase primary coolant activity in preaccident iodine spike to 2220 GBq/t DE I-131. Therefore, for the SGTR preaccident iodine spike, the primary coolant activity in preaccident iodine spike calculation method is proposed in this paper of ACP100.

Referring to ACP1000, the iodine peak assumes that the primary coolant activity in preaccident iodine spike increases from 37 GBq/t DE I-131 to 2220 GBq/t DE I-131, that is, the primary coolant activity in preaccident iodine spike increases to 60 times steady state value. Therefore, assuming SGTR preaccident iodine spike of ACP100, the primary coolant activity in preaccident iodine spike also increases to 60 times the steady state value(7.4 GBq/t DE I-131), that is, 444 GBq/t DE I-131. Under this assumption, primary coolant activity in preaccident iodine spike of ACP100 is shown in Table 1.

  1. 2)

    Primary coolant activity in concurrent iodine spike

For SGTR concurrent iodine spike accident analysis of ACP100, referring to the assumption of RG1.183, assuming that the iodine release rate from the fuel rods to the primary coolant increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (7.4 GBq/t DE I-131) specified in technical specifications, that is, the leakage rate from the fuel element into the primary coolant is 335 times the normal leakage rate, the assumed iodine spike duration should be 8 h.

The leakage rate of iodine from the fuel element into the primary coolant in transient state:

$$ L_{i} = \frac{{C \cdot R_{i} }}{{I_{i} \cdot f_{i} \cdot \eta }} $$
(1)

  • \(L_{i}\) is the leakage rate of the nuclide from the fuel element into the primary coolant in transient state, 1/s;

  • \(C\) is the iodine concentration increase times in transient state;

  • \(R_{i}\) is the equilibrium iodine release rate of nuclide, Bq/s;

  • \(I_{i}\) is the inventory of the nuclide in core, Bq;

  • \(f_{i}\) is the nuclide fraction in the fuel pellet-cladding gap;

  • \(\eta\) is the fuel cladding damage fraction, 0.25%.

Table 1. Primary coolant activity
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2.3 Iodine Carrying Coefficient in Steam Generator

The design of the once-through steam generator used in ACP100 is quite different from the U-tube steam generator used in M310 and ACP1000 [7]. The U-tube steam generator is a saturated steam generator. The heat transfer tube of the steam generator is completely immersed in water. The iodine of primary coolant leaking to the secondary coolant through the break of the steam generator tube enters the secondary coolant. The iodine of liquid phase is carried into the gas phase on the secondary coolant with the steam, and then released to the environment through the safety valve of the steam generator.

For the once-through steam generator, the secondary coolant is divided into a preheating section, an evaporation section and a superheating section according to the characteristics of the boiling phase change of the secondary coolant. In the superheat section, the iodine in the liquid flashes quickly into the vapor phase and is released to the environment through the steam generator safety valve. Compared with the U-tube steam generator, the iodine in the once-through steam generator is more easily carried by the vapor from the liquid phase to the vapor phase. Therefore, the carry coefficient is obviously not conservative enough if ACP100 directly refers to the iodine of M310 or ACP1000 on the secondary coolant of the steam generator. Since the liquid on the wall of the heat transfer tube will evaporate to dryness in the superheating section of the once-through steam generator, during this process, the iodine in the secondary coolant liquid phase will flash into the secondary coolant vapor phase. So it can be conservatively assumed that iodine carryover factor on the secondary coolant of steam generators is 1.

3 SGTR Source Term Analysis and Radiological Consequence Evaluation of ACP100

3.1 Source Term Analysis Results

Using the calculation assumptions and parameters in Sect. 2, the cumulative source term to the environment after SGTR accident calculated using the ASTA program are shown in Table 2. ASTA is a program independently developed by nuclear power institute of china for calculating the release of radionuclides to the environment under accident conditions.

Table 2. Activities released to the environment after SGTR accident
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3.2 Analysis Results of Radiological Consequences

The individual dose limits in the “Principles for Safety Review of Modular Small Pressurized Water Reactor Nuclear Power Plant Demonstration Projects (Trial version) in Chinese” are respectively determined as: The effective dose that the public individual (adult) may receive in each rare accident should be controlled below 5 mSv, and the thyroid equivalent dose should be controlled below 50 mSv; In each extreme accident, the effective dose that the public (adult) may receive should be controlled below 10 mSv, and the thyroid equivalent dose should be controlled below 100 mSv.

Using the site meteorological data of an ACP100 nuclear power plant, the radioactive consequences outside the factory were calculated for the source term of preaccident iodine spike case and concurrent iodine spike case are shown in Table 3, meet the acceptance criteria.

Table 3. Consequence of SGTR accident
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4 Conclusions

In this paper, on the basis of fully learning from the existing nuclear power projects, combined with the actual design characteristics of ACP100, a set of source term analysis methods for SGTR preaccident iodine spike case and concurrent iodine spike case suitable for ACP100 are proposed, and this method is used to analyze the SGTR accident source term of an ACP100 nuclear power plant has been identified, and its radiological consequences also meet the radiological consequences acceptance criteria in “Principles for Safety Review of Small Modular Pressurized Water Reactor Nuclear Power Plant Demonstration Projects (trial version) in Chinese”.