Passive Cooldown Performance of Integral Pressurized Water Reactor

doi:10.4236/epe.2013.54B097

Paper Menu >>

Journal Menu >>

Energy and Power Engineering, 2013, 5, 505-509

doi:10.4236/epe.2013.54B097 Published Online July 2013 (http://www.scirp.org/journal/epe)

Passive Cooldown Performance of Integral Pressurized

Water Reactor

Shoubao Dai, Chunnan Jin, Jingfu Wang, Yuxiang Chen

NO.703 Research Institute of China Shipbuilding Industry Corporation, Harbin, China

Email: [email protected]

Received January, 2013

ABSTRACT

The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor

by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of

passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of

the more formidable and dangerous operation environments of them. This paper presents results of marine black out

accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the

reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation,

and the shutdown cooling syste m (SCS) is designed to cool the primary system from a hot shutdown condition to a re-

fueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4

code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident

analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core

decay heat by a natural circulation.

Keywords: An Integral Pressurized Water Reactor (IPWR); Passive Safety System; Styling; Natural Circulation

1. Introduction

Most of the primary circuit components of an integral

pressurized water reactor (IPWR) are housed within a

single reactor pressure vessel (RPV). This layout can

provide the optimum configuration for adopting inherent

and passive safety design. So they are suitable for the

medium/small size nuclear power plants and marine nu-

clear systems. However, the characteristics of passive

safety systems for the marine reactors are quiet different

from those for the land nuclear power plant because of

the more formidable and dangerous run environments of

them. Consequently, in orde r to ensure the reliability and

the security of the marine-used IPWR, it is of great im-

portance to perform detailed investigation on the opera-

tion characteristics of their passive safety systems [1,2].

The IPWR with 100 MW capab ility is designed in this

paper. Fundamental concept and general arrangement of

the IPWR in this paper is based on the same principle of

the Inherent Safe UZrHx Power Reactor [3] and the Rus-

sian ABV-6M IPWR [4]. This 100 MW integral reactor

adopts the arc plate fuels in core, casing once-through

steam generator (OTSG), inherent safety improving fea-

tures such as a large volume of the reactor coolant, a

large negative moderator temperature coefficient, a low

core power density, and a passive residual heat removal

system (PRHRS). The major design parameters are

summarized in Table 1. The Figure 1 shows a schematic

diagram of the safety systems for the IPWR design.

These safety systems are designed to meet the relevant

redundancy and independency design requirements to

ensure a high reliability and safety.

Table 1. Initial parameters for the system.

System parameters Values Unites

Thermal power 100 MW

Reactor pressure vessel

Operating pressu re

Coolant mass flow

Core inlet temperature

Core outlet temperature

Steam generator

Main steam flow rate

Steam temperature

Feedwater temperature

Steam pressure

Tube outer diameter

Tube inner diameter

Tube length

1500

235

286

160

263.84

150

2.943

8×1

10×1.5

1.5

MPa

t/h

℃

t/h

℃

MPa

PRHRS

Coolant initial temperature

Coolant initial pressure

Tube inner diameter

Tube length

Water temperature in tank

Water pressure in tank

0.1013

16×1.2

1.6

0.1013

℃

MPa

℃

MPa

S. B. DAI ET AL.

506

Figure 1. Schematic diagram of the safety system of IPWR.

1.1. Passive Residual Heat Removal System

(PRHRS)

The PRHRS passively removes a core decay heat and a

sensible heat by a natural circulation in case of an emer-

gency condition such as unavailability of feed-water

supply or marine black out. Besides, the PRHRS may

also be used in case of long-term cooling for repair [5 ].

The IPWR can cool down the primary coolant from a

normal operation to a hot shutdown using a passive safe-

ty system when the reactor is tripped while a loop-type

commercial PWR cools down the co olant using an activ e

system such as an auxiliary feed-water system. The ac-

tive system can maintain a co ns tant coo ldown rate for the

transient, however, the passive system is relatively diffi-

cult to maintain a constant value because a driving force

reduces when a density difference between a hot and cold

sides is small at the end of transient. Therefore, a cool-

down analysis should be accomplished according to a

well-defined methodology.

1.2. Shutdown Cooling System (SCS)

The shutdown cooling system is designed to remove a

normal heat and a sensible heat in the reactor vessel from

a hot shutdown to a refueling condition. When the reac-

tor coolant system reaches 208℃ of a coolant tempera-

ture and 2.3 MPa of its pressure, the shutdown cooling

system cools the reactor coolant system to a refueling

condition. The SCS include s a shutdown cooling pump, a

heat exchanger, a water storage tank and valves. The

SCS sucks up a coolant at the main coolant pump (MCP)

suction duct, and then it discharges the coolant to the

main coolant pump discharge region [6].

2. Analysis Results

To evaluate the passive cooldo wn cap ability of the IPWR

in the case of marine black out accident, an analysis is

performed by using the system analysis code, RE-

LAP5/MOD3.4[7]. The nodalization for a passive cool-

down analysis of the IPWR is shown in Figure 2.

The results of the cooldown analysis are presented.

The marine black out accident begins after operating at

full power for 100 s. In the case of marine black out tran-

sient, the reactor is rapidly tripped by a trip signal, then

the PRHRS isolation valve is opened and the MFIV/

MSIV are closed .

The mass flow rate in the primary lo op drops immedi-

ately due to the main coolant pump coasting down on

reactor trip. The natural circulation flow is well estab-

lished after 8s as noticed in Figure 3. Due to su b-cooled

water of residual heat exchanger (RHE) tubes entering

OTSG, fluid temperature at the core inlet initially de-

creases, but reduced heat removal by the secondary sys-

tem makes the core inlet fluid temperature increase. After

50s the core inlet fluid temperature begins to decrease

gradually with decreased power. The core outlet tem-

perature has the same trend as the core inlet temperature

as shown in Figure 4. It is noted in the Figures 5 and 6

that the primary pressure and pressurizer level continues

to decrease with reactor trip. The secondary pressure

increases rapidly on reactor trip. It is because that ini-

tially tubes of the PRHRS is full of sub-cooled water, and

PRHRS have no enough heat transfer area to condense

the steam which is from OTSG, so the mass and energy

Figure 2. The nodalization for integral reactor.

S. B. DAI ET AL. 507

Figure 3. Mass flow rate in the primary loop.

Figure 4. Fluid temperature at the core inlet, outlet.

Figure 5. The pressure and water level in the pressurizer.

are increasing at the outlet of OTSG in the secondary

loop. The secondary pressures thus continue to increase

for a short period until the natural circulation establishes

in the PRHRS loops. Figure 7 shows the heat generated

in the core, the heat removed from the primary side to the

steam generator and from the steam generator to the ul-

timate heat sink by the PRHRS during the accident. As

shown in Figure 7, the heat transferred to the secondary

and third sides exceeds the decay heat generated in the

core after the PRHRS comes into operation, thus proving

the capacity of the PRHRS in mitigating the marine black

out accident.

Figure 8 shows the coolant temperature at the core in-

let and outlet, which continues to decrease with de-

creased power after the natural circulation flow estab-

lished well in the PRHRS loops (Reactor trip at 0 s).

When the coolant temperature reaches the SCS entry

condition, which is 208℃ of the coolant temperature at

core outlet, the saturation temperature decreases rapidly

due to a decreasing system pressure as shown in Figure

9, which is depressurized by an operator’s action using

an operation of the reactor coolant gas vent system

(RCGVS).The pressurizer pressure begins to decrease

Figure 6. Outlet pressure of the OTSG secondary side.

Figure 7. Comparison among the powers of the three loops.

S. B. DAI ET AL.

508

rapidly with the transient and then, it stabilizes at 9MPa.

The system condition needs to be lower than 208℃ and

2.3 MPa in order to connect the SCS. The reactor coolant

gas vent system is operated to achieve these conditions

when the coolant temperature reaches 208℃. Then, the

system pressure decreases rapidly again to 2.3 MPa as

shown in Figure 9.

A sensitivity study is performed to establish the cool-

down capability of the 100 MW integral reactor for vari-

ous system conditions such as natural and forced circula-

tion conditions, and the number of PRHRS. The results

are shown in Figure 1 0.

The pn, pf, sn and the sf denote a natural circulation at

the primary system, a forced circulation at the primary

system, a natural circulation at the secondary and a

forced circulation at the secondary system, respectively.

The forced circulation at the primary system is a case

where the main coolant pump is available after a reactor

is tripped and the forced circulation at the secondary

system is a case where 10% of the normal feed-water

flow rate is supplied by a feed-water pump. For the pri-

mary forced circulation condi tion, the coolant temperature

Figure 8. Primary system coolant temperature.

Figure 9. Pressurizer pressure.

Figure 10. Coolant temperature for the sensitivity study.

of the primary system reaches the primary system entry

condition nearly at the same time as the reference calcu-

lation. The cooldown capability of the IPWR is depend

on the primary system condition, that is, a h eat transfer at

the steam generator primary side. Also, the coolant tem-

perature can be cooled to 208℃ by only one t PRHRS.

However, it takes about 1100 s to reach the SCS entry

condition.

3. Summary and Conclusions

The passive cooldown characteristics and the natural

circulation performance of IPWR with a PRHRS have

been investigated in the case of the marine black out ac-

cident. It is shown that the reactor coolant system and the

PRHRS adequately remove the core decay heat, and as-

sure to remove appropriately a heat capacity in the reac-

tor vessel by natural circulation. For the design bases

accident conditions, IPWR cools down the coolant to the

SCS entry condition within 1100 S for all the possible

boundary conditions. Furthermore, the system can reach

the SCS entry condition by using only one PRHRS. Also,

the safety systems function properly and thus they can

secure the reactor to a safe condition for any accident.

4. Acknowledgements

Dai Shoubao would like to take this chance to express

my sincere gratitude to my supervisor, Jin Chunnan and

Chen Yuxiang, who provide the kindly assistance and

valuable suggestions during the process of this paper.

This gratitud e also extends to the professor Peng Min-

jun who taught him during his undergraduate years for

his kind encouragement and patient instructions.

REFERENCES

[1] Y.-J. Chung, H.-C. Kim and B.-D. Chung, “Two Phase

Natural Circulation and Heat Transfer in the Passive

S. B. DAI ET AL.

509

Residual Heat Removal System of an Integral Type Re-

actor,” Annals of Nuclear Energy, Vol. 33, No. 3, 2006,

pp. 262-270. doi:10.1016/j.anucene.2005.09.009

[2] M. H. Chang, S. K. Sim, D. J. Lee, “SMART Behavior

under Over-Pressurizing Accident Conditions,” Nuclear

Engineering and Design, Vol. 199, No. 1-2, 2000, pp.

187-196. doi:10.1016/S0029-5493(99)00068-0

[3] S. L. Cheng and G. J. Leng, “Inherent Safe UZrHx Power

Reactor INSURE-100 Preliminary Research,” Nuclear

Power Engineering, Vol. 15, No. 4, Aug. 1994.

[4] J. K. Liu and C. B. Tang, “Summarisation on Russian

ABV-6M Integral Pressurized Water Reactor,” Nuclear

Power Engineering, Vol. 18, No. 3,1997.

[5] Y. J. Chung, S. H. Kim and H. C. Kim, “Thermal Hy-

draulic Analysis of SMART for Heat Removal Transients

by a Secondary System,” Nuclear Engineering and De-

sign, 2003, pp. 225-257.

[6] Y.-J. Chung and S. W. Lee, “Passive Cooldown Per-

formance of a 65 MW Integral Reactor,” Nuclear Engi-

neering and Design, Vol. 238, 2008, pp. 1681-1689.

doi:10.1016/j.nucengdes.2007.12.008

[7] C. D. Fletcher and R. R. Schultz. RELAP5/MOD3 Code

Manual Volume V: USER’ S GUIDELINES[Z], 1995,

38-42, 130-133, 139-158, 173-180.