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Journal of Nuclear Science and Technology

ISSN: 0022-3131 (Print) 1881-1248 (Online) Journal homepage: https://www.tandfonline.com/loi/tnst20

Experiments on the Heat Transfer and Natural

Circulation Characteristics of the Passive Residual

Heat Removal System for an Advanced Integral

Type Reactor

Hyun-Sik PARK , Ki-Yong CHOI , Seok CHO , Choon-Kyung PARK , Sung-Jae YI ,

Chul-Hwa SONG & Moon-Ki CHUNG

To cite this article: Hyun-Sik PARK , Ki-Yong CHOI , Seok CHO , Choon-Kyung PARK ,

Sung-Jae YI , Chul-Hwa SONG & Moon-Ki CHUNG (2007) Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor, Journal of Nuclear Science and Technology, 44:5, 703-713, DOI: 10.1080/18811248.2007.9711859

To link to this article: https://doi.org/10.1080/18811248.2007.9711859

Published online: 05 Jan 2012.

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Experiments on the Heat Transfer and Natural Circulation Characteristics

of the Passive Residual Heat Removal System for an Advanced

Integral Type Reactor

Hyun-Sik PARK1; , Ki-Yong CHOI1 , Seok CHO1 , Choon-Kyung PARK1 , Sung-Jae YI1 , Chul-Hwa SONG1

and Moon-Ki CHUNG1

1

Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-600, Korea (Received July 12, 2006 and accepted in revised form January 29, 2007)

Experiments on the heat transfer characteristics and natural circulation performance of the passive re-sidual heat removal system (PRHRS) for the SMART-P have been performed by using the high temper-ature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.

KEYWORDS: SMART, integral type reactor, PRHRS, two-phase, heat transfer, natural circulation

I. Introduction

The SMART1–3) is an advanced modular integral type

pressurized water reactor with a power of 330 MWt and sev-eral enhanced safety features. It contains the major RCS components, such as the main coolant pumps, steam gener-ators, and pressurizers, within a reactor vessel to avoid the occurrence of a large break LOCA. The basic design of the SMART was completed in 2002 by KAERI. A pilot plant of the SMART, SMART-P, will be constructed in Korea and it will have a rated power of 65 MWt. The SMART and the SMART-P are designed both for a forced convection core cooling during start-up and normal operating conditions and for a natural circulation core cooling during accidental conditions.

Among the systems in the SMART-P, the passive residual heat removal system (PRHRS) is the system installed to pre-vent an over-heating and over-pressurization of the reactor coolant system during accidental conditions. The PRHRS re-moves the core decay heat and sensible heat by a natural cir-culation under emergency conditions when normal

feedwa-ter supply and steam extraction are unavailable.

Existing reactor designs and new concepts rely, to varying degrees, on residual heat removal processes driven by natu-ral circulation as a potentially important design feature or ul-timate heat removal mechanism.4) Table 1 summarizes the

innovative small- and medium-sized reactors and their resid-ual heat removal systems.5)The listed thirteen reactors could be classified by the connection location of the RHRS and the type of natural circulation. The RHRSs are connected to ei-ther a primary circuit for core cooling or a secondary circuit for steam generator cooling, and the coolant is naturally cir-culated either in single-phase or in two-phase conditions.

There are very limited experimental data on the perform-ance of the PRHRS. A passive steam condensing channel has been developed for emergency cooling of an integral reactor and its feasibility and efficiency are validated by the exper-imental data acquired by using a thermo-physical test facili-ty.6)Therefore, it is necessary to perform safety-related

ex-periments to validate the performance capability of the PRHRS.

A high temperature/high pressure thermal-hydraulic test facility, VISTA (Experimental Verification by Integral Sim-ulation of Transients and Accidents),7)has been constructed to simulate the SMART-P. The VISTA facility is an integral effect test facility and its scaled ratios are 1/1 in height and 1/96 in volume with respect to the SMART-P. Its design



This article was received and accepted as ‘‘Original Paper’’.

Ó

Atomic Energy Society of Japan



Corresponding author, E-mail: hspark@kaeri.re.kr

ARTICLE

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pressure and temperature are set to simulate the steady-state and transient conditions of the SMART-P. The PRHRS of the VISTA facility is designed to simulate the operating con-ditions and system characteristics of the reference system, namely SMART-P.

Researches on the thermal-hydraulic behaviors of the SMART PRHRS and the VISTA facility have been per-formed previously. Chung et al.8)investigated the

thermal-hydraulic characteristics for the PRHRS in the SMART by using the MARS code and showed that the PRHRS fulfilled its functions in removing the heat transferred from the pri-mary side in the steam generator when the heat exchanger is submerged in the emergency cooldown tank. Preliminary performance tests by using the VISTA facility were reported by Choi et al.9)They included steady states and power step/ ramp changing tests, primary natural circulation tests, and preliminary PRHRS transient tests. Transient operations for an integral type reactor, the SMART-P, have also been

experimentally investigated by using the VISTA facility in order to verify the system design and performance of the SMART-P.10) Park et al.11) performed preliminary

experi-ments on the performance of the PRHRS for the SMART-P by using the VISTA facility.

In this paper, the experimental investigations of the heat transfer and the natural circulation characteristics of the PRHRS of the VISTA facility are reported in detail. The pri-mary objective of these PRHRS transient tests is to investi-gate the characteristics of a natural circulation and pressure drop in the circulation loop to better understand the natural circulation performance of the PRHRS in the SMART-P. The second objective is to investigate the heat transfer char-acteristics of the PRHRS heat exchangers and the emergency cooldown tank. The last objective is to investigate the over-all thermal-hydraulic behavior in the primary system during a PRHRS operation.

Table 1 Summary of the innovative small and medium sized reactors and their residual heat removal systems

Reactor Residual heat removal system Connection location of RHRS Type of natural Circulation SMART (System-integrated

modular advanced reactor)

PRHRS (Passive

residu-al heat removresidu-al system) Secondary circuit Two-phase IRIS (International reactor

inno-vative and secure)

EHRS (Emergency heat

removal system) Secondary circuit Two-phase CAREM (Central Argentina de

Elementos Modulares)

RHRS (Residual heat

removal system) Primary circuit Two-phase MARS (Multipurpose advanced

reactor, inherently safe)

SCCS (Safety core

cool-ing system) Primary circuit

Single-phase and two-phase

SCOR (Simple compact reactor)

RRP (Residual heat removal system on the primary circuit)

Primary circuit Two-phase

IMR (Integrated modular water reactor)

SDHS (Stand-alone direct heat removal system)

Secondary circuit Two-phase

VBER-300 (Water cooled mod-ular power reactor)

EHRS (Emergency heat

removal system) Secondary circuit Two-phase VK-300 (Water cooled and

mod-erated natural circulation boiling water reactor)

RHRS (Residual heat

removal system) Primary circuit

Single-phase or two-phase

CCR (Compact containment

boiling water reactor) IC (Isolation condenser) Primary circuit Two-phase RMWR (Reduced moderation

water reactor of 300 MW(e)) IC (Isolation condenser) Primary circuit Two-phase AHWR (Advanced heavy

water reactor) IC (Isolation condenser) Primary circuit

Single-phase or two-phase RUTA-70 (Reactor facility for

district heating with atmospheric pressure in the primary circuit)

ASEC (Air system for

emergency cooldown) Secondary circuit Single-phase

KAMADO (Concept of a passive-safety reactor)

WPCS (Water pool

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II. Description of the Test Facility

1. Configuration of the VISTA Facility

The schematic diagram of the VISTA facility for the SMART-P is shown in Fig. 1. The VISTA facility is an in-tegral test facility to simulate the primary system and secon-dary system as well as the major safety-related systems of the SMART-P. Its scaled ratios are 1/1 in height and 1/96 in volume with respect to the SMART-P.

The reactor core is simulated by electric heaters with a ca-pacity of 818.75 kW, which is about 120% of the scaled power. Unlike the integrated arrangements of the SMART-P, the VISTA primary components including a reactor ves-sel, a main coolant pump, a steam generator, and a pressur-izer are connected to each other by pipes for an easy instal-lation of the instrumentation and simple maintenance. The secondary system with a single train is simply designed to remove the primary heat source. Besides these major sys-tems, a make-up water system and a chilled water system are installed to control the feedwater supply and its temper-ature.

2. Scaling of the PRHRS

The scaling ratio of the number of heat exchanger tubes and trains of the PRHRS is defined as follows:

NR¼Nm=Np¼ 1=96 ð1Þ

TR¼Tm=Tp¼ 1=4: ð2Þ

As the working fluid and the operating pressure of the model are same as those of the prototype, all the physical properties are the same. The relationship between each

de-sign parameter, which is determined by the volume scaling methodology, can be expressed as follows:

AHX;R ¼NRTRÿ1¼ 1=24 ð3Þ VPRHRS;R¼NRTRÿ1¼ 1=24 ð4Þ f  l dþ K   s;R ¼NRÿ2TR2ds;R4 ð5Þ f  l dþ K   w;R ¼NRÿ2TR2dw;R4 ; ð6Þ

where A, V, s, w, and d mean the heat transfer area, volume, steam, water, and pipe diameter, respectively. The technical specifications of the PRHRS of the VISTA facility are shown in Table 2.

3. Configuration of the VISTA PRHRS

The schematic diagram of the PRHRS of the VISTA fa-cility is shown in Fig. 2. The PRHRS of the VISTA fafa-cility is composed of a train for the cooling subsystem, which in-cludes an emergency cooldown tank (ECT), a heat exchang-er (HX), a compensating tank (CT), sevexchang-eral valves, and relat-ed piping. The PRHRS of the VISTA facility should have the capability to simulate both the passive and active cooling of the reference system. It is connected to both the feedwater and the steam lines of the secondary system to provide a flow path for the natural circulation. Enough cooling water should be supplied to the ECT from the component cooling water system (CCWS) to remove the heat from the internal HX ef-ficiently, and also, enough water and nitrogen gas should be supplied to the CT from the makeup water system (MWS) and the nitrogen supply system (NSS), respectively. Pressure Vessel Steam Generator Pressurizers Gas Cylinder Emergency Cooldown Tank Compensation Tank Silencer MCP Feedwater Supply Tank Bypass Valve Isolation Valves

Pressure Control Valve

Feedwater Control Valve FWP Isolation Valves Bypass Valve Electrical Heater

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It is designed to have the same pressure drop and heat transfer characteristics and it is arranged to have the same el-evation and position as those of the reference system. Also the diameter, thickness, pitch, and orientation of the heat ex-changer tubes of the VISTA facility are the same as those of the reference system. Only one train of the cooling

subsys-tem is installed in the VISTA facility. The PRHRS can be isolated from the secondary system using bypass valves in normal operating conditions. Also the location of the compo-nents, pressure drop characteristics, and the heat exchanging capabilities are properly calculated and reflected to have the same natural circulation capability as the reference system. During the PRHRS operation, the superheated steam gen-erated from the steam generator secondary side is injected into and condensed in the PRHRS heat exchangers by a nat-ural circulation. The condensed water is drained through the PRHRS condensate line and returned to the feedwater line of the secondary system. The condensing heat is transferred to the emergency cooldown tank, the heat of which is removed by the component cooling water.

Figures 3 and 4 show a schematic diagram and a cross-sectional view of the PRHRS heat exchanger, respectively. The PRHRS heat exchanger is installed vertically in the emergency cooldown tank and it is composed of 6 ing tubes, and the upper and lower headers. All the condens-ing tubes are made from Inconel-600, and their length, inner diameter, and outer diameter are 1,200, 13.0, and 15.0 mm, respectively. K-type thermocouples are attached to the sur-face of the condensing tube to measure the sursur-face temper-ature, and three measuring ports are also installed to measure the pressure and temperature of the upper and lower headers. The measuring locations of the flow rate, pressure, tem-perature, and differential pressure in the natural circulation loop are shown in Fig. 5. The flow rate is measured by using

Table 2 Technical specifications of the PRHRS of the VISTA fa-cility

Parameters Unit Prototype (P) VISTA (M) No. of trains EA 4 1

Operating pressure MPa 3.5 3.5 Operating temperature C 242.5 242.5

No. of HX tubes/train EA 141 6

HX: tube length m 1.2, 0.013, 1.2, 0.013, ID: thickness 0.0025 0.0025 HX: tube material — Titanium alloy Inconel-600 CT: volume/tank m3 0.35 0.009 CT: ID, height m 0.55, 0.0873, 1.5 1.5 ECT: volume/tank m3 NA 0.25

ECT: ID, height m NA 0.4, 2.0

Emergency

Cooldown Tank

Steam Line

MWS NSS

Compensation

Tank

Isolation Valve P T T

Feed Water Line

Steam

Generator

Steam

Generator

P P T T 3/4" sch.160s CCWS CCWS 1/2" tube DP DP TC–1 TC–2a,b,c TC–3 TC–5 TC–6 TC–8 TC–7 T P PT–1 DP–2 DP–3 PT–2 PT–3

RHRS Pump

LT–3 PT–5 PT–4

HXs

DP T Chilled water T TC–4a,b,c T P 1/2" tube 3/8" tube 3/8" tube Orifice–1 Orifice–2 OV–SND–02 OV–SND–01 OV–SND–05 OV–SND–04 OV–RHS–01 MV–RHS–01 MV–RHS–02 OV–RHS–04 VC–RHS–01 VC–RHS–02 OV–RHS–02 SV–RHS–01 OV–RHS–03 MV–RHS–03 MV–MWS–03 T T T T DP DP–1

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the Coriolis-type mass flowmeter with an accuracy of 0.25% of the reading. The temperature is measured by using a K-type thermocouple with an accuracy of 0.4% of the reading. Its maximum error is 1.1C. The pressure and the differential

pressure are measured by using smart-type sensors with an accuracy of 0.075% of their full spans.

The steam from the steam line of the secondary system is cooled and condensed in the heat exchanger submerged in the emergency cooldown tank, and the condensed water is re-circulated into the feedwater line of the secondary system and it passes through the secondary side of the steam gener-ator to cool the primary system. The single-phase or two-phase fluid circulates naturally in the loop of the secondary side of the steam generator, the steam line of the secondary system, the PRHRS steam line, the condensing heat ex-changer, the PRHRS condensate line, and the feedwater line of the secondary system.

4. Data Acquisition System

The data acquisition system provides the data collection functions for the VISTA facility. This system consists of two parts, the computer and display terminal, and the VXI

C-size mainframe and terminal panels residing in the control room. These two parts are connected via an industry stand-ard IEEE 1394 (Firewire) serial control and data interface. The data acquisition system is isolated from the control sys-tem by using several types of signal distributors.

The computer collects and saves the data from the various instruments that measure the pressure including the differen-tial pressure, water level, flow rate, pump speed, and the temperature. The instrumentation part of this system consists of an industry standard VXI mainframe from Hewlett-Pack-ard (E8403A), and a firewire controller interface cHewlett-Pack-ard (E8491B with option 001), and several (currently four) state-of-the-art data acquisition A/D cards (E1413C) with several types of functional signal conditioning plug-ins (SCPs). The maximum A/D conversion rate on each E1413C card is normally 100,000 Sample/sec, but it is ad-justable to the user’s requirements. The normal data-scan-ning rate is set to 100 Hz, but the data saving rate is 2 Hz with the mean values of 50 data points.

5. Control System

The VISTA facility is designed to be operated by a com-bination of a manual and automatic operation. Several initial operations such as an inventory filling, startup, and cool-down are operated manually by an operator. Once the major thermal-hydraulic parameters reach a steady state condition, they are switched over to be controlled automatically by the PID feedback control logics to maintain the achieved steady state condition. For an automatic control, several PID control blocks were installed into a programmable logic controller (PLC) in the control system.

The controlled components include the electrical heating rod, the main coolant pump, the feedwater control valve, the steam pressure control valve, the feedwater storage tank heater, and the makeup pump. During a normal operation, the electrical heater power is automatically controlled to pro-vide a constant core exit temperature, e.g., 300C. A special

146 1200 102 86 42 30 30 140 140 520 5 2 0 130.5 TC-RHS-2a, 2d, 4a TC-RHS-2b, 2e, 4b TC-RHS-2c, 2f, 4c ( Unit: mm ) 29.5

Fig. 3 The schematic of the PRHRS heat exchanger

18 12 TC–RHS–2d, e, f TC–RHS–2a, b, c 18 HX Tubes (6 EA) Upper & Lower

Plenums (140 mm, OD)

TC–RHS–4a, b, c

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combination of three heater groups is used to prevent the maximum heater surface temperature from exceeding the safety limit. During the PRHRS operation, the predeter-mined power from the programmed ANS-73 decay curve is given to the core simulating heater.

III. Test Procedure and Test Matrix

Several PRHRS tests have been performed after the steady state conditions were achieved for the given powers. After reaching a steady state condition for a given power, the

Compens ation Tank Co o ld o w n P u m p EL. 2.764 m N4 N5 N1 N6 N2 N3 N5-2 N5-1

Feed Water Line

Main Steam Line EL. 4.930 m EL. 2.830 m EL. 4.015 m EL. 3.319 m EL. 2.963 m EL. 4.533 m EL. 6.380 m EL. 7.580 m EL. 4.467 m EL. 4.533 m EL. 3.630 m EL. 1.7543 m EL. 2.0543 m EL. 2.7543 m EL. 3.4543 m EL. 3.7543 m to Boiler from Boiler ECT H/X from CWS to CWS DP-RHS-03 DP-RHS-02 DP-RHS-01 DP-RHS-01L PT-RHS-04 PT-RHS-02 PT-RHS-03 PT-RHS-01 PT-RHS-05 DP-SG2-04 DP-SND-01 DP-SND-0 2 PT-SND-04 PT–SND–01 FT-SND-02 FT-SND-01 TC-SND-04 PT-SG2-04 TC-SG2-21 PT–SG2–06 TC–SG2–22 PT-SND-01 TC-SND-01

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PRHRS is immediately triggered both to start opening the bypass valves which connect the PRHRS to the secondary system and to start closing the secondary system isolation valves which isolate the secondary system from the feedwa-ter supply tank and the silencer. The test matrix for the PRHRS operation is summarized in Table 3.

Nine tests in total have been carried out so far. When a PRHRS start button is pressed, the electrical heater and the main coolant pump are switched off immediately, and the valve control signals are generated with a certain time delay. In the present study, a compensation tank was not used. For the former six tests, the electrical power was tripped off at the start of the test without the simulation of the decay pow-er, and the valve opening time interval between the secon-dary and the PRHRS valves was set to about 0.5 s. For the latter three tests, the electrical power was controlled in con-formity with the ANS-73 decay curve in order to simulate the decay power in the core when a reactor is tripped. In these cases, the isolation valves and bypass valves were closed and opened simultaneously. The overall trend is sim-ilar to each other among the tests performed except for mi-nor changes due to the variations of the boundary conditions. In the steady state conditions, before the PRHRS was oper-ated, the initial powers supplied from the core simulating heater were 10, 25, 36, 50, 75, and 100% of the rated power, and the initial flow rates of the primary loop were 36, 50, and 100% of the rated flow rate.

The scaled 100% flow rate of the primary system was 19.6 m3

/h at the pressure and temperature of 147 bar and 310C, respectively, and the scaled 100% flow rates of the

secondary system were 0.25 kg/s. With an operation of the PRHR system, the subcooled water is injected into the steam generator to be heated by the energy transfer from the pri-mary loop and it is finally changed to superheated steam.

IV. Results and Discussions

The test results of the transient PRHRS operation are de-scribed in this Chapter. H-P36-Q36-D-PRHR test simulated the core decay heat and it provided typical test results. Its

in-itial core power and inin-itial feedwater flow rate were 36% of the rated values and the initial primary flow rate was 36% of the rated flow rate. The test results showed that the initial core power, the initial feedwater flow rate, and the initial pri-mary flow rate had little effect on the thermal-hydraulic be-havior in the PRHRS loop.

Figure 6 shows a comparison of the measured decay power with the programmed power. The results show that the agreement was excellent between them. Figure 7 shows the opening and closing intervals of the four bypass and iso-lation valves used during the PRHRS operation. The PRHRS is triggered to start with the PRHRS actuation signals, which are generated under the conditions of a low flow rate of the main feedwater line, a low pressure of the main steam line, loss of power, etc. The opening and closing characteristics of these four isolation valves play important roles in the PRHRS operation. The opening intervals of the PRHRS by-pass valves (OV-RHS-01, OV-RHS-03) are about 2.0 s, and the closing intervals of the secondary system isolation valves (OV-SND-02, OV-SND-04) are about 4.0 s.

1. Analysis of the PRHRS Natural Circulation Perform-ance

Figure 8 shows a typical feedwater flow rate variation during a PRHRS operation for four tests. The initial feedwa-ter flow rate ranged between 0.09 and 0.25 kg/s and it de-creased rapidly to about 0.03 kg/s, which is about 12% of the scaled secondary flow rate. A natural circulation loop was formed within a few seconds and its flow rate decreased gradually. The trend of the flow rate during the natural cir-culation was similar regardless of the initial feedwater flow rate when the decay heat is transferred to the primary sys-tem. However, the natural circulation flow rate of H-P100-Q100-PRHR without the simulated decay heat was smaller than that with the simulated decay heat. It is due to the fact that an additional heat source was not supplied to the pri-mary loop and the cooldown of the pripri-mary loop is quicker in the case of H-P100-Q100-PRHR when compared with the case of H-P36-Q36-D-PRHR, thus the primary loop

gradual-Table 3 Test matrix for the PRHRS operation

Test ID Initial power (%) Initial flow (%) Decay heat Delay time (sec) H-P10-Q36-PRHR 10 36 NA 0.5 H-P25-Q50-PRHR 25 50 NA 0.5 H-P36-Q100-PRHR 36 100 NA 0.5 H-P50-Q100-PRHR 50 100 NA 0.5 H-P75-Q100-PRHR 75 100 NA 0.5 H-P100-Q100-PRHR 100 100 NA 0.5 H-P36-Q36-D-PRHR 36 36 ANS73 0.0 H-P50-Q100-D-PRHR 50 100 ANS73 0.0 H-P75-Q100-D-PRHR 75 100 ANS73 0.0 0 20 40 60 80 100 120 0 10 20 30 40 50 60 70 80 90 100

- Decay curve: ANS-73 - Heat loss: 10.25 kW - Full power: 682.3 kW

Decay Power (kW)

Time (sec)

Programmed Power (Input) Measured Power (Output)

Fig. 6 Comparison of the measured decay power to the program-med power

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ly loses the potential to increase the driving force in the PRHRS loop.

Figure 9 shows a variation of the PRHRS pressure during a PRHRs operation. The PRHRS pressure increased rapidly after the isolation valves were opened and it reached its peak before the superheated steam was condensed in the heat ex-changer. As the steam is condensed in the heat exchanger, its energy is transferred to the water flowing through the emer-gency cooldown tank (ECT) and the system pressure is de-creased. As the heat is removed, the primary coolant temper-ature is decreased. Accordingly, the steam tempertemper-ature in the steam generator decreased and the PRHRS pressure decreas-ed rapidly.

Figure 10 shows a typical temperature variation in the PRHRS during a PRHRS operation. The fluid in the upper plenum of the heat exchanger was slightly superheated, and the variation of the steam temperature showed a similar trend to that of the system pressure. The water in the lower plenum of the heat exchanger, in the heat exchanger outlet

line, and in the PRHRS condensate line was subcooled, which indicates that the decay heat could be easily removed in the heat exchanger tubes.

Figure 11 shows a typical differential pressure variation during a PRHRS operation. The differential pressure meas-ured in the PRHRS steam line was negligibly small. The dif-ferential pressure measured in the PRHRS heat exchanger increased rapidly to 4 kPa in the early stages and, thereafter, it gradually increased. As the frictional pressure drop is neg-ligibly small due to the low flow rate of the naturally-circu-lated condensate in the PRHRS loop, it can be assumed that the measured differential pressure is mainly due to the con-densate level in the PRHRS heat exchanger during the PRHRS operation. It could be estimated that the PRHRS loop was completely filled with water in the initial state, that the water level in the PRHRS heat exchanger dropped to 40 cm, and that it gradually decreased during the PRHRS op-eration. The differential pressure measured in the PRHRS condensate line increased rapidly to 18 kPa in the early

0 1 2 3 4 5 0 10 20 30 40 50 60 70 80 90 100 Valve Opening (%) Time (second) OV-SND-02 OV-SND-04 OV-RHS-01 OV-RHS-03

Fig. 7 Opening and closing intervals of the four isolation valves during the PRHRS operation

0 500 1000 1500 2000 2500 3000 3500 0.00 0.05 0.10 0.15 0.20 0.25 Flow rate (kg/s) Time (sec) H-P36-Q36-D-PRHR H-P50-Q100-D-PRHR H-P75-Q100-D-PRHR H-P100-Q100-PRHR

Fig. 8 Variation of the feedwater flow rate during the PRHRS operation 0 500 1000 1500 2000 2500 3000 3500 0 10 20 30 40 50 60 70 Pressure (bar) Time (sec) PRHRS, Steam Line PRHRS, Lower Plenum PRHRS, HEX Outlet Line PRHRS, Condensate Line

Fig. 9 Variation of the PRHRS pressure during the PRHRS oper-ation 0 500 1000 1500 2000 2500 3000 3500 0 50 100 150 200 250 300 Temperature ( o C) Time (sec)

PRHRS, HEX Upper Plenum PRHRS, HEX Lower Plenum PRHRS, HEX Outlet Line PRHRS, Condensate Line

Fig. 10 Typical temperature variation in the PRHRS during the PRHRS operation

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stages and, thereafter, it gradually decreased in the latter stages during the PRHRS operation. The water which was contained in the PRHRS condensate line was drained rapidly to a level of 180 cm and it continued to refill during the PRHRS operation.

The variation of the degree of superheat in the secondary system during a PRHRS operation is given in Fig. 12. Be-fore the initiation of the PRHRS operation, the steam from the steam generator was in a superheated condition. When a natural circulation was started by the initiation of the PRHRS operation, the steam temperature started to fall due to the energy transfer through the heat exchanger in the emergency cooldown tank (ECT) and the inflow of the cold condensate water. The decreasing rate of the steam tem-perature was very large and the steam temtem-perature suddenly dropped to below a saturated condition. Except for the initial stages, the steam from the SG was slightly superheated dur-ing the PRHRS operation.

2. Analysis of the Heat Transfer Characteristics

Figure 13 shows a variation of the surface temperature of the heat exchanger during a PRHRS operation. The meas-ured surface temperature in the upper part of the heat ex-changer was always above 100C. It means that a boiling

oc-curs in the upper part of the heat exchanger, which was also observed through the installed observation glass.

Figure 14 shows a variation of the temperature distribu-tion in the ECT during a PRHRS operadistribu-tion. The temperature of the component cooling water was low in the intake area of the lower region, but it was maintained at around 60C in the

discharge area of the upper region. The difference of the flu-id temperature in the mflu-iddle lower parts was very small be-cause of a complete condensation of the steam in the upper region.

Figure 15 shows a variation of the inlet and outlet peratures in the ECT during a PRHRS operation. The tem-perature difference between the ECT inlet and outlet increas-0 500 1000 1500 2000 2500 3000 3500 -5 0 5 10 15 20 25 30 35 40

Differential Pressure (kPa)

Time (sec)

PRHRS, Condensate Line PRHRS, Heat Exchanger PRHRS, Steam Line

Fig. 11 Typical differential pressure variation during the PRHRS operation 0 500 1000 1500 2000 2500 3000 3500 0 20 40 60 Degree of Superheat ( o C) Time(sec) SG, Steam Outlet

Fig. 12 Variation of the degree of superheat in the secondary sys-tem during the PRHRS operation

0 500 1000 1500 2000 2500 3000 3500 0 50 100 150 200 Temperature ( o C) Time (sec)

PRHRS, HEX Tube Top (4a) PRHRS, HEX Tube Center (4b) PRHRS, HEX Tube Bottom (4c)

Fig. 13 Variation of the surface temperature of the heat exchang-er during the PRHRS opexchang-eration

0 500 1000 1500 2000 2500 3000 3500 0 20 40 60 80 100 120 140 160 Temperature ( o C) Time (sec)

PRHRS, ECT Top (2a) PRHRS, ECT Center (2b) PRHRS, ECT Bottom (2c) PRHRS, ECT Top (2d) PRHRS, ECT Center (2e) PRHRS, ECT Bottom (2f)

Fig. 14 Variation of the temperature distribution in the ECT dur-ing the PRHRS operation

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ed rapidly in the early stages and it decreased gradually thereafter due to a decrease in the heat transfer.

The heat transfer rates can be calculated either from the two-phase mixture enthalpy change inside the heat exchang-er tubes or from the cooling watexchang-er tempexchang-erature change out-side the tubes. The heat transfer rate from the enthalpy change of the steam-water mixture can be obtained as fol-lows:

QHXðtÞ ¼ _mmHX ðhinÿhoutÞ: ð7Þ

The heat transfer rate from the temperature change of the cooling water can be obtained as follows:

QECTðtÞ ¼ _mmECTcp ðhoutÿhinÞ: ð8Þ

Figure 16 shows a variation of the heat removed during a PRHRS operation. The heat transfer rate calculated from the enthalpy change shows a rapid change of the heat transfer while the heat transfer rate calculated from the cooling water temperature change shows a slow heat transfer response.

3. Overall Thermal-Hydraulic Behavior of the Primary System

Figure 17 shows a variation of the primary system sures during a PRHRS operation. The primary system pres-sures decreased rapidly due to a rapid increase of the heat transfer through the steam generator in the early stages of the PRHRS operation. The reduction rate of the system pres-sures was lowered with a decrease in the natural circulation flow rate.

Figure 18 shows a variation of the SG primary side tem-peratures during a PRHRS operation. The tendency of the temperature variation was similar to that of the pressure var-iation The temperatures decrease gradually and the temper-ature difference between the inlet and outlet of the SG pri-mary side is almost constant, which shows that the decay heat is sufficiently removed during a PRHRS operation.

Figure 19 shows a variation of the primary system flow rate during a PRHRS operation. The primary coolant started to circulate naturally with the operation of the PRHRS, and 0 500 1000 1500 2000 2500 3000 3500 0 20 40 60 80 100 Temperature ( o C) Time (sec) PRHRS, ECT Inlet PRHRS, ECT Outlet

Fig. 15 Variation of the inlet and outlet temperatures in the ECT during the PRHRS operation

0 500 1000 1500 2000 2500 3000 3500 0 20 40 60 80

Power Exchanged - HX & ECT (kW)

Time (sec)

PRHRS, HEX Inside PRHRS, ECT

Fig. 16 Variation of the heat removed during the PRHRS operation

0 500 1000 1500 2000 2500 3000 3500 0 25 50 75 100 125 150 175 200 Pressure (bar) Time (sec)

PS, Boiler Pressure Vessel PS, Main Coolant Pump

Fig. 17 Variation of the primary system pressures during the PRHRS operation 0 500 1000 1500 2000 2500 3000 3500 0 50 100 150 200 250 300 350 400 Temperature ( o C) Time (sec)

S/G Primary Side, Inlet S/G Primary Side, Outlet

Fig. 18 Variation of the SG primary side temperatures during the PRHRS operation

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its flow rate reached a peak value in the early stages and it decreased gradually thereafter.

V. Conclusions

The heat transfer characteristics and the natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been experimentally inves-tigated in the high temperature/high pressure thermal-hy-draulic test facility (VISTA). In the PRHRS transient tests, the natural circulation flow rate through the PRHRS loop reached around 12 percent of the rated value in the early stages of the PRHRS operation and it decreased gradually thereafter. The PRHRS pressure increased rapidly after the PRHRS operation and it decreased slowly as the steam con-densed in the PRHRS heat exchanger tubes. Also the PRHRS temperature showed the same trend. The heat trans-fer between the PRHRS heat exchanger and the ECT is large enough for the coolant to maintain a natural circulation in the PRHRS loop. The pressure and temperature of the pri-mary system decreased rapidly due to a rapid increase in the heat transfer through the steam generator in the early stages of the PRHRS operation and they decreased gradually in the latter stages. The experimental results show that the core decay heat can be easily removed by a PRHRS opera-tion. Nomenclature A: Area cp: Specific heat d: Pipe diameter f: Friction factor h: Enthalpy ID: Inner diameter

K: Form loss coefficient l: Length

_ m

m: Flow rate

N: Number of heater rod Q: Heat transfer rate

t: Time

T: Number of train V: Volume (Subscripts)

ECT: Emergency cooldown tank HX: Heat exchanger in: Inlet m: Model out: Outlet p: Prototype R: Ratio s: Steam w: Water References

1) M. H. Chang, J. W. Yeo, S. Q. Zee et al., Basic Design Report of SMART, KAERI/TR-2142/2002, KAERI, Daejeon, Korea (2002).

2) G. H. Lee, S. H. Yang, S. W. Lee et al., Performance Assess-ment for SMART Basic Design, KAERI/TR-2171/2002, KAERI, Daejeon, Korea (2002).

3) S. H. Yang, H. C. Kim, Y. J. Chung et al., Safety Analysis Re-port for SMART Basic Design, KAERI/TR-2173/2002, KAERI, Daejeon, Korea (2002).

4) International Atomic Energy Agency, Status of Advanced Light Water Reactor Designs 2004, IAEA-TECDOC-1391, IAEA, Vienna, Austria (2004).

5) International Atomic Energy Agency, Status of Innovative Small and Medium Sized Reactor Designs 2005–Reactors with Conventional Refueling Schemes, IAEA-TECDOC-1485, IAEA, Vienna, Austria (2006).

6) O. B. Samoilov, V. S. Kuul, V. A. Malamud et al., ‘‘Integral nuclear power reactor with natural coolant circulation. Inves-tigation of passive RHR system,’’ Nucl. Eng. Des., 165, 259–264 (1996).

7) S. J. Yi, K. Y. Choi, H. S. Park et al., Basic Design of the High Temperature/High Pressure Thermal-Hydraulic Test Facility, 10394-TE-RR840-02, Rev. 0 (Internal Report), KAERI (2001).

8) Y. J. Chung, S. H. Yang, H. C. Kim et al., ‘‘Study on the ther-mal hydraulic characteristics of a residual heat removal system for the SMART plant,’’ The 10th Int. Topical Meeting on Nu-clear Reactor Thermal Hydraulics (NURETH-10), October 5– 9, Seoul, Korea (2003).

9) K. Y. Choi, H. S. Park, S. Cho et al., ‘‘VISTA: Thermal-hy-draulic integral test facility for the SMART reactor,’’ The 10th Int. Topical Meeting on Nuclear Reactor Thermal Hy-draulics (NURETH-10), October 5–9, Seoul, Korea (2003). 10) K. Y. Choi, H. S. Park, S. Cho et al., ‘‘Parametric studies on

thermal hydraulic characteristics for transient operations of an integral type reactor,’’ Nucl. Eng. Technol., 38[2], 185– 194 (2006).

11) H. S. Park, K. Y. Choi, S. Cho et al., Experiments for Heat Transfer Characteristics and Natural Circulation Perform-ance of PRHRS of the High Temperature/High Pressure Ther-mal-Hydraulic Test Facility (VISTA), KAERI/TR-2656/2004, KAERI, Daejeon, Korea (2004).

0 500 1000 1500 2000 2500 3000 3500 0 1 2 3 4 Flowrate (kg/s) Time (sec) PS, Downcomer

Fig. 19 Variation of the primary system flow rate during the PRHRS operation

Figure

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References