Transient analysis of IPWR

The integral pressurized water reactor (IPWR) is a small-to-medium-sized nuclear reactor design (SMR) in which most of the primary loop components are housed in a single reactor pressure vessel (RPV). These reactors are designed to enhance safety and reliability by utilizing inherent safety features and passive safety systems, particularly as in marine-type nuclear reactors [1].

In this paper a study has been carried out on the full power load percentages of IPWR under steady-state and transient conditions, such as a loss of feed water (LOFW) accident scenario, by using a thermal hydraulic system code (Relap5/Mod3.4).

For these purposes, the normal operational state of the reactor is first obtained at 100% full power, and then the steady-state results at different full powers are obtained. Thirdly we consider a transient condition, such as loss of feedwater accident [5], a category II nuclear reactor safety analysis, that is, one of moderate frequency [12]. More research on the reactor’s vulnerability to different accidents [4] such as loss of coolant accidents (LOCA) [6,7], pressurizer breakdown [8], the need to mitigate residual heat from the core, station blackout accidents, anticipated transient without scram (ATWS) [9,10] and steam generator tube rupture accidents (SGTR) [11], and others, can be investigated at a later date.

In order to simulate the steady state as well as transient situation, a Relap5 nodalization diagram of the reactor has been constructed and used.

In this paper, research on the Integral Pressurized Water Reactor that has a 100 MW power output [13] has been performed as shown in Figure 1. In this reactor, once-through steam generators (OTSGs) are placed in the annular space between core barrel and the reactor pressure vessel. Since the reactor is designed to contain four loops, 20 OTSGs have been modeled. The pressurizer is placed at the top of the reactor vessel in a dome.

Feedwater enters the reactor and collects in a small water tank in the annular space between the core barrel and the reactor pressure vessel, from where it moves into the secondary tube of the OTSG, where it takes heat from the primary side of the OTSG. As steam, it travels to and from the secondary loop as shown in Figure 1.

The key feature of this reactor is that the water flows without the use of a pump; it works on the principle of natural circulation. Primary water flows downward and collects in the downcomer, from which it enters the core. There, the temperature of the water increases, and it rises upward through the riser above the core towards the pressurizer, until it turns and goes back down towards the OTSG. In this reactor, a passive residual heat removal system (PRHRS) composed of a water tank and the heat exchanger enables residual heat to be compensated easily in case of accidents (see also Figure 1) [14].

The thermal hydraulic system code Relap5/Mod3.4 is used for the simulation of LWR steady-states and transients during postulated accidents. The code was developed at the Idaho National Laboratory for the U.S. Nuclear Regularity Commission (NRC). It is highly generic code and can be used for the simulation of a wide range of thermal hydraulic phenomena, including both nuclear and non-nuclear systems such as steam, water or non-condensable gases [16]. The reactor has been modeled by a nodalization diagram that breaks down pipes into multiple units.

In the passive residual heat removal systems (PRHRS), trip valves open in case of a reactor transient, diverting steam into a PRHRS heat exchanger submerged inside a water tank (Figure 1). The steam condenses inside this heat exchanger, which flows back to the reactor. Thus there are two natural circulation loops: one inside the reactor and one in the PRHRS outside the reactor [17].

Since the reactor under study works on the principle of natural circulation, we use a mathematical model based on Boussinesq assumptions for the steady-state analysis under load variations. (Equations have been omitted for space but are available on www.neimagazine.com/ipwrreferences).

First, full-power steady-state parameters were acquired (Table 1). Then the variation of parameters with load changes was studied. We conclude that in a certain loop structure, the power of the reactor can determine the steady-state characteristics of parameters that govern single-phase natural circulation (such as mass flow rate, temperature differences of the core inlet and outlet, and so on). Changes in power create two significantly different types of steady state characteristics: linear flow and turbulent flow. Between these values is an a transient flow state. Only turbulent flow is considered in this paper. The distinction between those characteristics are determined by the eigenvalue ‘m’, which is included in the steady state mathematical equations.

The buoyancy generated by the density differences between hot and cold fluid is the only driving force for natural circulation. This effect ultimately changes the power level, since the steady-state characteristics of single-phase natural circulation are affected by flow patterns created by the driving forces created by different power levels. The relationship between temperature difference and load is linear (Figure 2). The relationship between mass flow and load is approximately linear (Figure 3). The figures also show that the Relap5 calculations match the numerical calculations well.

Figures 4 to 6 show the relationship between power level and steady state characteristics (temperature distribution of the steam, water and the void fraction in the secondary tubes) of the OTSG. It is shown that the OTSG tube section can be divided into three parts: the preheat section (containing supercooled water), evaporation section (containing saturated water) and superheated section (containing superheated steam). Supercooled water enters the preheat section, which changes the fluid temperature and enthalpy gradually. The fluid temperature rises to saturation value, and then continues to be heated to reach the evaporation stage. Finally, superheated steam is generated. Table 2 summarises the results of this study using graphs produced by the Relap5 code.

Transient analysis

In the light of the above studies, a transient analysis has been simulated by considering a loss of feedwater accident. A LOFW accident can be caused by a feedwater pipe line break, the closure of feedwater control isolation valves, the failure of feedwater pumps, the startup of the auxiliary feedwater system, or other factors. It results in the increase of primary coolant temperature and pressure conditions with the decline of nuclear power suddenly due to the reactivity feedback phenomenon. In the current research, simulated with the Relap5/Mod3.4 code, up to t=2000sec the reactor is in normal operational state. The feedwater pipe is closed at t=2000sec, resulting in increased reactivity. The reactor is in an unstable state, so at about t=2010sec, the reactor is SCRAMmed. At the same time, the pressure in the pressuriser increases, which is followed by the closure of the coolant supply. At t=2015, the steam line is considered to be closed. At t=2020, the PRHRS opens and the reactor core is flooded by natural circulation a few seconds afterwards. Reactor stability is achieved, as can be observed from the graphs (Figures 7-10).

It is clear that in LOFW accidents the major components of the reactor have different trends (Figures 7-10). Figure 7 shows how after the initialization of accident the reactor power declines from its normal operational value. As the PRHRS operates and removes the decay heat, the reactor power normalises. In this period also the temperature of the steam generator declines to a low value. Figure 8 clearly shows how at 2000 sec the mass flow rate decreases enormously and reaches a low level. But after 20 seconds, with the opening of the PRHRS, the water level in the reactor increases, thus compensating for the transient. It is to be noted that there is some oscillation of the mass flow rate due to the establishment of natural circulation with the PRHRS. Figure 9 shows how there is a sudden increase in temperature due to feedwater loss and the reactor SCRAM, but soon after, natural circulation is established and the difference between core inlet and outlet temperature again reaches a stable value after the initialization of the PRHRS. Figure 10 shows how pressure increases sharply after the accident. Due to the loss of secondary water, the heat exchange between primary and secondary loops becomes unstable, but gradually declines down to a stable level with the intake of water from the PRHRS, until the pressure reaches 0.1 MPa, which is atmospheric pressure.

The purpose of this study is to investigate whether the concerned reactor can be operated well in different power levels, and to investigate the circumstances or the problems that occur during load variations and in different accidents. Since the simulation results are in good agreement with numerical and design values, this study verifies that the reactor has met the safety criteria.

Author Info:

Salah Ud-Din Khan and Minjun Peng Nuclear Power Simulation and Research Centre, College of Nuclear Science and Technology, Harbin Engineering University, 145-Nantong Street, Nangang District, Heilongjiang Province, Harbin City, 150001, P.R. China.

The first author is very grateful to his co-author for his keen and valuable research advice. He thanks the College of Nuclear Science and Technology, Harbin Engineering University, China, which funded the research.

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