Studies on Key Effect Factors of Natural Circulation

Research ArticleStudies on Key Effect Factors of Natural CirculationCharacteristics for Advanced PWR Reactor CavityFlooding System

Dekui Zhan 1 Xinhai Zhao1 Shaoxiong Xia1 Peng Chen1 and Huandong Chen2

1China Nuclear Power Technology Research Institute Shenzhen 518000 China2Sino-French Institute of Nuclear Engineering and Technology Sun Yat-sen University Zhuhai 519000 China

Correspondence should be addressed to Dekui Zhan zhandekuicgnpccomcn

Received 24 December 2019 Revised 26 March 2020 Accepted 29 June 2020 Published 1 September 2020

Academic Editor Han Zhang

Copyright copy 2020 Dekui Zhan et al is is an open access article distributed under the Creative Commons Attribution Licensewhich permits unrestricted use distribution and reproduction in any medium provided the original work is properly cited

In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR) different kinds of severeaccident mitigation strategies have been proposed In-Vessel Retention (IVR) is one of the important severe accident managementmeans by External Reactor Vessel Cooling Reactor cavity would be submerged to cool the molten corium when a severe accidenthappens e success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than thelocal critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall canmaintain the integritye residual thickness of RPV is determined by the heat flux transfer from the corium pool and the coolingcapability of outer wall of the RPV ere are various factors which would influence the CHF and the cooling capability of outerwall of the RPV In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outsidethe actual reactor vessel a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built A large number ofsensitivity tests were carried out to study how these sensitivity factors affect CHF value and natural circulation Based on the testresults the structure of the test section flow channel has an obvious effect on the CHF distribution e flow channel optimizedcan effectively enhance the CHF value especially to enhance the CHF value near the ldquoheat focusrdquo region of the molten pool ewater level in the reactor pit has also a great impact on the natural circulation flow Although natural circulation can bemaintainedwith a low water level it will lead to a decrease of the cooling capacity Meanwhile some noteworthy test phenomena have beenfound which are also essential for the design of the reactor pit flooding system

1 Introduction

e probability of severe accidents in nuclear power plantsis much low but once a severe accident occurs it willcause the nuclear reactor core to melt e core coriumrelocates to the lower head forming a molten pool andtransferring heat to the RPV wall If the molten poolcannot be effectively cooled down the pressure vessel maybe melted through due to excessive thermal load whichgreatly increases the possibility of radioactive materialrelease to the environment en how to maintain theintegrity of RPV and limit the corium in the lower headbecome the focus of research on mitigation measures forsevere accidents worldwide

External Reactor Vessel Cooling (ERVC) to achieve In-Vessel Retention (IVR) is an important mitigation measurefor severe accidents Due to its advantage of low con-struction difficulty and good economy IVR-ERVC is widelyused in the 3rd generation million-kilowatt nuclear powerplants Chinese advanced pressurized water reactor mainlyadopts this strategy to mitigate severe accidents

In the design of IVR-ERVC the reactor pit floodingsystem is one of the most important dedicated safety sys-tems In order to verify the effectiveness of IVR-ERVC it isnecessary to carry out related research to evaluate the factorsthat affect the cooling capability of the reactor pit floodingsystem e factors include the design of RPV flow channelthe water level in the reactor pit and so forth In this paper

HindawiScience and Technology of Nuclear InstallationsVolume 2020 Article ID 4765046 11 pageshttpsdoiorg10115520204765046

the description of a new reactor pit flooding system and theREVECT (Reactor Vessel External Cooling Test) facility areintroduced and the influence factors of cooling capability arestudied Based on the REVECT tests a series of problemswhich have confused the designers in long-time are solvede research results have been applied to the engineeringdesign of reactor pit flooding system and the establishmentof management guidelines for severe accidents in new nu-clear power plants

2 Research Status

After the occurrence of a postulated severe accident coriumwill collapse into lower head of RPV e disposal mode ofcorium becomes a common concern Currently there aremainly two treatment ways for corium cooling (1) Ex-VesselRetention (EVR) this method is mainly used in EPR especific method is setting a core catcher at the bottom ofcontainment which has functions of isolating molten debrisand concrete as well as providing long-term cooling EVRstrategy is also adopted in Tianwan nuclear power plant(NPP) in China (2) In-Vessel Retention (IVR) IVR strategyis implemented to remove decay heat from the RPV and tokeep the corium in the RPV [1] is strategy has also beenused in VVER440 AP1000 HPR1000 and APR1400 [2]

In order to verify the effectiveness of the IVR strategysome different test facilities which are used to simulate theflow and heat transfer characteristics have been built such asULPL test facility and HERMES-HALF test facility econclusions of the tests are as follows (1) An aged coppersurface exhibits a similar coolability performance as theexternal surface of the RPV wall which is verified by ULPUand BETA-NC test [3ndash5] (2) Optimized flow channel canenhance the CHF of the external surface of lower head (3)e chemical properties of some coolants can affect theCHF For instance the coolant with boric acid can decreasethe CHF to some extent

In China different types of CHF tests and related an-alyses have been performed in recent years Based on a large-scale test facility Yang [6] investigated the CHF charac-teristics of chemical solution boiling on a downward facingcurved surface It was found that the CHF of mixed solutionof H3BO5 and Na3PO4 increases firstly and then decreaseswith the increase of Na3PO4 concentration rough a two-dimensional full-scale facility FIRMWei [7] discovered thatthe aging effect could enhance the CHF of SA508 owing tothe generation of Fe3O4 oxide To study the CHF margin ofExternal Reactor Vessel Cooling (ERVC) for ChineseAP1400 the FIRM subcooled flow boiling facility conductedby State Nuclear Power Technology Research amp Develop-ment Center (SNPTRD) was built [8] It was also shown thatthe concentration of Na3PO4 has an important effect on theCHF behavior Based on the experimental data a CHFcorrelation for IVR-ERVC was developed by Mei et al [9]with considering the effects of surface orientation thermaleffusively and corrosion Besides Tan and Kuang [10]preliminarily determined nominal values of reactor vesselinsulation design parameters according to ERVC relatedfunctional reliability criteria and related statistical analysis

Guo et al [11] proposed a newmethod to study the transientfeasibility of IVR-ERVC in which a theoretical CHF modelwas developed for the outer surface of the lower head Jinet al [12] performed the study on in-vessel and ex-vesselcoupled analysis of IVR-ERVC phenomena for large-scalePWR by using MELCOR Cheng et al [13] used a CDF codeFluent coupled with a boiling model by UDF (User-DefinedFunction) to investigate the CHF of ERVC which wasvalidated by experimental CHF values obtained by SNPTRD

Even though a large number of tests and analyses haveconducted there are little test results about the effect of thekey factors on the CHF distribution such as the geometry offlow channel and the water level in the reactor cavity eabove key factors are also essential for the reactor floodingsystem design

3 Design andManagement of Reactor FloodingSystem for Advanced PWR

31 Reactor Pit Flooding System e reactor pit floodingsystem is a dedicated severe accident mitigation system toachieve the IVR-ERVC Generally the reactor pit floodingsystem is composed of a reactor pit injection system and anatural circulation system When a severe accident happenscooling water can be injected into the reactor cavity tosubmerge the RPV via passive injection mode or activeinjection mode e decay heat is finally removed throughthe natural circulation in the reactor cavity to maintain theintegrity of RPV

e passive reactor pit injection subsystem mainly con-sists of the reactor pit flooding tank located on a place higherthan the main coolant pipes of Reactor Coolant System (RCP)in the containment pipelines and valves During a severeaccident after entering the severe accident managementguideline (SAMG) the valves on the passive injection pipesare opened by the operator and the water in the reactor pitflooding tank flows to the reactor pit by gravity with a largeflow rate When the water level in the pit flooding tank dropsto a certain threshold the injection is switched to small flowrate to compensate for the water loss due to evaporationensuring the reactor cavity to be kept flooded

e active reactor pit injection is provided by activewater injection pipelines which connect to the in-contain-ment refuelling water storage tank (IRWST) When thereactor pit flooding tank is depleted the cooling water istaken by pump from the IRWST and is injected into thereactor pit after cooled by Component Cooling WaterSystem or Extra Cooling System e schematic diagram ofpassive reactor pit injection subsystem and active reactor pitinjection subsystem is illustrated in Figure 1

Since the core outlet temperature reaches 650degC which isa significant temperature alarm for the NPP to carry out thesevere accident management guideline (SAMG) the oper-ator opens the valves on the passive injection pipelines tosubmerge the reactor pit with a large flow rate en thewater injection flow is switched to small flow rate phase tomake up the evaporated water Before the IVR tank is emptythe operator starts up the pump to inject water from the in-containment refuelling water storage tank (IRWST)

2 Science and Technology of Nuclear Installations

32 Key Phenomena of Natural Circulation in the ReactorCavity e main phenomena and the cooling process re-lated to the implementation of the reactor pit floodingsystem after severe accidents are as follows

When the cooling water is injected into the reactor pitunder severe accident conditions the water inlets andsteam outlets can be automatically opened And the coolingwater can enter the flow channel around the RPV to cool itdirectly

During a severe accident the molten corium will dropdown and fall into the water pool in the lower plenum ecorium pool will heat up the internal wall of the lowerplenum And many bubbles on the external surface of RPVwill produce under the heating of the lower head Naturalcirculation will form due to the density difference in thecirculation loop e water-steam two-phase flow will flowupward to the steam vent ports e water-steam separationwill occur at the top of the cylinder space around the supportring Finally the water will flow back into the reactor pit bythe recirculation pipes e structural drawing of reactor pitis shown in Figure 2

Natural circulation forms in the flow channel after thereactor cavity being flooded e specific cooling processafter the severe accidents in the reactor pit is as follows

Containment

IVRtank

M

MM

Reactor pit

IRWST

RPV

Figure 1 Design scheme of reactor pit flooding system

Outlet

Shield wall

Backflowchannel

Reactor pit

RPV support ring

RPV

Flow channel

Inlet

Figure 2 e natural circulation in the reactor pit

Science and Technology of Nuclear Installations 3

(1) e cooling water is heated by RPV outside wall andmany bubbles produce

(2) e water-steam flows upward along the gap be-tween RPV outside wall and insulation and thenvents out through the steam vent port of insulation

(3) e water flows back into the reactor cavity bybackflow channel after steam-water separation

33 REVECT Test Facility

331 Presentation of IVR Test Facility To study theinfluencing factors which determine the CHF distributionalong the outer surface of the RPV and natural circulationflow the REVECT (Reactor Vessel External Cooling Test)facility test facility for the advanced generation PWR hasbeen built e flowchart and the overview diagram of theREVECT test facility are shown in Figures 3 and 4 e testfacility consists of electrical system cooling system powercontrol system primary loop and instrumentation andcontrol (IampC) system and so forth e REVECTfacility isa two-dimensional facility with a full-height circulationloop and a 1 1 radial scaled slice-type test section e testsection is the key component of the facility composed of acopper heating section and stainless steel flow channel asthe simulator of the RPV lower head and the flow pathbetween the RPV and its insulation respectively Hun-dreds of cartridge heaters are inserted into the copperheating section to simulate the decay heat e heatingsection is divided into more than 20 heating zones torealize independent heating power control of differentzones e stainless steel flow channel is bounded byseveral baffles which are used to simulate the insulationstructure of the RPV lower heade schematic diagram ofthe test section is shown in Figure 4 An integrated watertank is located under the test section in connection withthe stainless steel flow channel to simulate the reactor pitspace under the RPV insulation e upper water tank islocated at the top of the facility to simulate the annularregion around the RPV support ring in the prototype emain parameters of the REVECT facility are listed inTable 1

e heating section is used to simulate the lower head ofRPV after a large amount of molten corium relocated in ite material of the heating block is copper e width ofheating block is constant in the test section e slice-typeheating section is divided into more than 20 independentpower control zones to realize the simulation of the decaypower distribution of the prototype along the lower headree baffles are installed in the flow channel to simulatethe structure and the gap size of the RPV lower headinsulation flow channel In each heating zone there arepairs of thermocouples installed in two concentric circleswithin the heating block to obtain the temperature data sothat the heat flux could be calculated according to themeasured temperature under the Fourier heat conductionequation e structure of the test section is shown inFigure 5 e main parameters of the test section are listedin Table 2

332 Key Phenomena in the REVECT Facility Loop eREVECT test facility can simulate the structure of circula-tion loops in the prototype For example the geometry of thetest section is the same as that of the flow channel outside theRPV It can simulate the flow and the heat transfer char-acteristic which is similar to the actual phenomena in theflow channel outside the lower head of the RPV e heightof the upward flow pipe is the same as that in the prototypeand the resistance characteristic of the upward pipe and thesteam exhaust ports are considered e upward pipe andthe steam exhaust ports can simulate the two-phase flow inthe flow channel outside the RPV cylinder part and thesteam vent ports e integrated tank can simulate thecylinder space around the support ring of the RPV ephenomenon of water-steam separation takes place in theintegrated tank which also happens above the cylinder spacearound the support ring of the RPV e resistance char-acteristic of the downward pipe is also modelled It cansimulate the recirculation pipes in the reactor pit shield wall

34 Test Research on the Key Techniques for Reactor PitFlooding System In the design of the reactor pit floodingsystem for new type nuclear power plants the researchfocuses on the natural circulation flow rate and CHF dis-tribution e reasons are the following

(1) CHF distribution the design of the flow channel hasa great relation with the CHF value on the outer wallof the lower head of RPV e specific influencingfactors are as follows the flow channel structure ofRPV metal insulation the area and the inlet positionof RPV metal insulation and the area of the outlet ofRPV metal insulation for steam venting

(2) Natural circulation natural circulation flow rate isalso closely related to the CHF value on the outerwall of the lower head e larger the natural cir-culation flow rate is the higher the value of CHF ise factors which affect the natural circulation flowinclude the design of the backwater channel waterlevel the makeup amount of water e vibrationfrequency and amplitude of the two-phase flowwhich is important for the design of RPV metalinsulation are also necessary to be considered

341 Input Power e power control strategy used in thisexperiment was proposed by Professor eofanous et al [1]at the University of California USA e similarity criterionthat makes the flow of test section is similar to that of theprototype e similarity criterion is successfully applied toIVR of AP600 AP1000 and APR1400 In the test it mainlyincludes the following two aspects

(1) e superficial vapor velocities of downstream po-sition match up with the prototype for all θgt θm

(2) e vapor flow rates build up gradually so as tosmoothly approach the value required at θ le θmwhile allowing a ldquonaturalrdquo development of boundarylayer in all of the upstream region


eupstream θ power distribution calculation formula is

qe(θ) qp(θ)sin θsin θm

for θle θm (1)

For downstream locations of the test area the powerdistribution calculation formula is the following

1113946θ

θm

qe θprime( 1113857dθprime Jp θm( 1113857 +1

sin θ1113946θ

θm

qp θprime( 1113857sin θprime( 1113857dθprime minus Je θm( 1113857

qp 1 minus cos θm( 11138571

sin θminus

1sin θm

1113888 1113889 +1

sin θ1113946θ

θm

qp θprime( 1113857sin θprime( 1113857dθprime for θgt θm

(2)

DV01

DV03

DV02

DV05 DV04

FV01

DV06

DV07

M102

M101

V101

M103 DV11

DV10

F03

F04

Glass pipe

V103Cooling

tower

F01 F02

V102P101

Filtrator

Check valve

4th platform

3rd platform

2nd platform

1st platform

Steamexhaust port

Water inlet of test section

Water outlet of test section

Figure 3 e flowchart of REVECT test facility

Figure 4 Overview of REVECT test facility


where qp is an average flux over 0lt θle θm defined by

qp sin θm

1 minus cos θm

1113946θm

0qp θprime( 1113857sin θprime( 1113857dθprime (3)

e theoretical and actual input heat flux density dis-tribution in the test heating zone 23 is shown in Figure 6

4 Sensitivity Study on theDifferent Structure ofthe Flow Channel

41TestConditions Nine test conditions (FC01-FC09) withdifferent flow channel distances and different bafflestructures were performed e specific test conditions are

shown in Table 3e flow channel consists of three straightbaffles and heating surfaces of the test section e bafflesmainly include a vertical section (CD) a middle sectionbaffle (BC) and an inclined baffle (AB) e sensitivity testswere divided into three stages In the first stage theminimum distance from the middle section baffle (BC) tothe heating surface is 250mm then adjusting the inclinebaffle (AB) to different three positions So three differentflow channel structures are formed (FC01-FC03) Similarlyin the second and third stages the minimum distancesfrom the middle baffle to the heat section surface are set as200mm and 150mm respectively then adjusting the in-cline baffle to different positions to perform the sensibilityanalysis

42 Test Results of Sensitivity Study on the Different FlowChannels Figure 7 shows the CHF distribution under threedifferent test conditions (FC01-FC03)

e CHF value of FC01 changes in a ldquowave shaperdquo withthe increase of the azimuth angle and the CHF value isrelatively small e CHF values of FC02 and FC03 show aldquoW-shaperdquo change with the azimuth angle change from 35degto 85deg

Figure 8 gives the CHF test data in the second stage(FC04-FC06) e CHF values are similar to the test resultsof FC01-FC03e CHF value of FC04-FC06 also changes ina ldquoW-shaperdquo from 35deg to 90deg

Figure 9 shows the CHF test data in the third stage(FC07-FC09) e CHF values of these three tests (FC07-FC09) are higher than those in the former two stages andthe CHF values of most areas are greater than 14MWm2With the increase of the azimuth angle from 35deg to 90deg theCHF value in the third stage increases firstly and thendecreases until the position of 68deg and then increases eCHF values of the FC07 and FC08 change relativelysmoothly and the value is relatively large e CHF valuesof most areas are larger than 15MWm2 Besides the bafflewith a minimum distance of 100mm (FC09) can increase

Table 1 Key parameters of REVECT facility

Parameters Valuedefinition RemarksTest pressure Atmosphere Pressure in upper water tankWater temperature atinlet of test section sim100degC e saturation temperature corresponds to the

pressure of the upper water tankTarget water level of theupper water tank 7300mm e elevation of the bottom of the heating section is

selected as the 0m level the same as the prototypeElevation of steamexhaust port 8000mm Relative to the bottom of the heating section

Circulation type Natural circulation e same as the prototypeWorking fluid Deionized waterSteamwater outlet areaof the upflow pipe sim001m2 About 1100 of the prototype RPV insulation steam

invent port areaCooling water inlet areaof the test section sim001m2 About 1100 of the prototype RPV insulation

cooling water inlet port area

Structure of baffles 3-part multilateral structure With the same structure and ldquogap sizerdquo distributionas the prototype

Circulation resistance Scaling as the prototype to ensure the consistence of thecomprehensive resistance of facility and the prototype

Based on the geometry and resistance distributionanalysis of the prototype flow path

Water inlet

Heating section

2200

Incline baffle

Outlet

Flow channelMiddle section

baffle

θ

L1

L2

D

C

B

A

2500

Figure 5 Schematic of the test section


the CHF value at the position between 52deg and 64deg andinhibit the CHF value in the upstream at the positionbetween 68deg and 72deg

e nine test results show that when the flow channeldistance ranges from 250mm to 200mm and then furtherdecreases to 150mm CHF value tends to increase while thedistance of flow channel continued to decrease to 100mm(FC09) the CHF value begins to decrease at the upstreamheating zone It can be considered that the CHF value tends

to increase with the space of flow channel decreases but thevalue of CHF will be inhibited when the distance of flowchannel decreases to a certain level

43 Optimization of the Flow Channel Structure Outside theRPV e above results show that different flow channelconfigurations lead to different CHF distributions Accordingto the test results the optimal design of the structure for the

Table 2 Key parameters of REVECT test section

Parameters Valuedefinition Remarks

Circumferential scale of the test section to prototype 1 100

Radial scale of the heating section to the prototype RPV 1 1 With the same outer diameter as the as the prototype RPVlower head

Height scale of the natural circulation loop to the prototypeRPF system 1 1

Number of independent heating zones 23 Cover the angular range of 0degndash90degDesigned maximum surface heat flux sim25MWm2 e actual heat flux is lower than 14MWm2

Width of flow channel 150ndash250mm Effective widthOuter diameter of the heating section 4800mm e same as the prototype RPV lower head

0 20 40 60 80 100000102030405060708091011121314

Theoretical heat flux distributionActual input heat flux distribution

The d

istrib

utio

n of

hea

t flux

(MW

m2 )

Azimuth angle θ (deg)

Figure 6 eoretical and actual input heat flux distribution in zone 23

Table 3 Test conditions for the flow channel optimization

No Test types Minimum distance of flow channel L1L2 Temperature of the inlet Pressure1 Flow channel optimization test FC01 250250mm sim100degC 1 bar2 Flow channel optimization test FC02 250210mm sim100degC 1 bar3 Flow channel optimization test FC03 250140mm sim100degC 1 bar4 Flow channel optimization test FC04 200200mm sim100degC 1 bar5 Flow channel optimization test FC05 200180mm sim100degC 1 bar6 Flow channel optimization test FC06 200130mm sim100degC 1 bar7 Flow channel optimization test FC07 150150mm sim100degC 1 bar8 Flow channel optimization test FC08 150145mm sim100degC 1 bar9 Flow channel optimization test FC09 150100mm sim100degC 1 bar


insulation is obtained and the verification test under theoptimized structure is carried out e test results show thateven though the CHF date at the position of 52deg is lower thanthat before being optimized the CHF data with the optimizedflow channel at higher angles (67degndash90deg) are obviously largerthan the experimental data in the case that the flow channel isnot optimized e average increase amplitude is about 20e schematic of optimized flow channel is shown inFigure 10 at means that there is larger margin for theheat focus effect owing to the optimized flow channel ecomparison of CHF values between optimized and non-optimized flow channels (FC08) is shown in Figure 11

e mechanism could be that when the distance changesfrom 250mm to 140mm the velocity of the two-phase flowchanges to higher the film boiling is more unlikely tohappen While the distance further decreases to 100mm the

two-phase flow with many large bubbles is likely to block theflow channel so the film boiling at the position of theminimum distance is more likely to happen

5 Natural Circulation Flow andEvaporated Water

51 Test Condition Tests of different water levels in thereactor cavity are also conducted to study the effect on thenatural circulation characteristic e details of test condi-tion are shown in Table 4

52 Test Results

521 6e Flow Rate of the Nature Circulation and theEvaporation e natural circulation flow exceeds 40m3hDue to 1 100 circumferential scale of the test section toprototype the natural circulation flow rate is about 4000m3hfor the actual nuclear power plants e test result shows thatthe natural circulation flow is very large once the steadynatural circulation forms so the backflow channel is essentialBesides the test results also show that there is about 05m3

water evaporated per hour so small flow rate (gt50m3h) isneeded to inject into reactor cavity to make up the water andkeep the level of water in the reactor cavity at the higherheight

522 Water Level Effect According to the test results thenatural circulation can be maintained if the water level isbetween 8m and 5m but the circulation flow rate is obviouslydecreased with lower water level When the water level dropsto 65m especially the intermittent discharge occurs causingthe pressure in the flow channel to fluctuate correspondinglye results are shown in Figures 12 and 13 Due to thephenomenon of intermittent discharge in the flow channeland the adverse cooling condition the temperature on theouter surface of the lower head of RPV may rise at the same

00

02

04

06

08

10

12

14

16

18

Azimuth angle θ (deg)30 40 50 60 70 80 90 100

The CHF test data of FC01The CHF test data of FC02The CHF test data of FC03

q CH

F (M

Wm

2 )

Figure 7 CHF distribution of the FC01-FC03 test conditions

30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )



time which is not conducive to the cooling of the lower headerefore in the design of reactor cavity flooding system aclear request is made for the lowest water level of the reactorcavity after water injection e water level of the reactorcavity must be kept as high as possible to ensure that thenatural circulation can be maintained with larger flow rateand take away more decay heat from the corium pool

523 Vibration Frequency and Amplitude To obtain thevariation of the vibration frequency of the two-phase flow inthe flow channel with different water levels in the reactorcavity the fluctuation pressure and frequency of the two-phase flow in the flow channel are considered

When the liquid level decreases from 8m to 7m thefluctuation pressure is within plusmn10 kPaWhen the water level is

130

390

2200

1210

43deg

170

250

D

C

B

A67deg

Figure 10 e schematic diagram with the optimized flow channel of the test section

30 40 50 60 70 80 90 10000

05

10

15

20

CHF value before non-optimized structure of insulationCHF value after optimized structure of insulation

Hea

t flux

(MW

m2 )

Angle (deg)

Figure 11 Comparison of CHF values between optimized and nonoptimized flow channels

Table 4 Test conditions for the natural circulation characteristics

No Test types Water level Temperature of the inlet Pressure1 Natural circulation characteristics tests NC01 8m-5m sim100degC 1 bar


high the pressure fluctuation is regular and the frequency isabout 14Hz With the decrease of water level the pressurefluctuation begins to appear irregular but the fluctuationpressure is still in the range of plusmn10 kPa So it can be found thatthe pressure fluctuation range shows a decreasing trend withthe decrease of water level and the fluctuation frequency isabout 11Hz at a relatively regular fluctuation as shown inFigures 14 and 15

Based on test results in order to prevent the occurrenceof resonance phenomenon after reactor pit injection systemis implemented which may lead to the destruction of theinsulation the inherent frequency of the insulation is re-quired to avoid 11Hzndash14Hz and the antivibration pressureof insulation needs to be more than plusmn10 kPa at the sametime

6 Conclusions

Based on the above experimental results the natural cir-culation characteristics of two-phase flow are investigated indetail when implementing the reactor pit flooding systemafter severe accidents and the following design requirementsare proposed for the reactor pit flooding system for new typenuclear power plants e specific requirements are asfollows

(1) e flow channel structure outside the RPV has asignificant effect on the CHF value e flow channeloptimized can effectively enhance the CHF valueespecially to enhance the CHF value near the ldquoheatfocusrdquo region of the molten pool erefore the

0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)

Natural circulation flow rateVariation of pressure (P1 pressure sensor)Variation of pressure (P2 pressure sensor)

05101520253035404550556065707580859095100

Figure 12 Fluctuation pressure is stable with water level of 8m

05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Figure 13 Fluctuation pressure is changing unstably with waterlevel decreases

500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)

Vibration frequency and amplitude (pressure sensor HP01)Vibration frequency and amplitude (pressure sensor HP02)

Pres

sure

(kPa

)

Figure 14 Vibration frequency and amplitude with water level of8m

0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)


Figure 15 Change of vibration frequency and amplitude withwater level decreases


effectiveness of the IVR-ERVC strategy can also befurther enhanced

(2) Natural circulation flow can exceed 4000m3h afterreactor pit flooding system is implemented ere-fore multiple backwater flow channels must be set toensure that the water discharged from the outlet ofinsulation can flow back into the reactor pit again

(3) e water level in the reactor pit has a great impacton the natural circulation flow Although naturalcirculation can be maintained with low water levelsit will lead to a decrease in the cooling capacityerefore the water level must be maintained asmuch as possible at higher position

(4) e inherent frequency of RPV insulation is requiredto avoid 11Hzndash14Hz and the antivibration pres-sure of insulation needs to be more than plusmn10 kPa atthe same time

Data Availability

e data used in the study have been included in the article

Conflicts of Interest

e authors declare that there are no conflicts of interestregarding the publication of this article

Acknowledgments

is work was supported by National Key RampD Program ofChina (Grant no 2018YFB1900100)

References

[1] T G eofanous C Liu S Addition S AngeliniO Kymaelaeinen and T Salmassi ldquoIn-vessel coolability andretention of a core meltrdquo vol 1 Office of Scientific andTechnical Information Oak Ridge TN USA 1996

[2] T G eofanous S J Oh and J H Scobel ldquoIn-vessel re-tention technology develoopment and use for advanced PWRdesigns in the USA and Koreardquo vol 14 Office of Scientific andTechnical Information Oak Ridge TN USA 2004

[3] T N Dinh J P Tu and T Geofanous ldquoTwo phase naturalcirculation flow in AP-1000 in-vessel retention-related ULPU-V facility experimentsrdquo in Proceedings of the ICAPPrsquo04Pittsburge PA USA June 2004

[4] Y H Jeong W P Baek and S H Chang ldquoCHF experimentsfor IVR-RVC using 2-D slice test sectionrdquo in Proceedings ofthe Korean Nuclear Society Spring Meeting Kwangju SouthKorea May 2002

[5] T G eofanous J P Tu A T Dinh and T N Dinh ldquoeboiling crisis phenomenon part I nucleation and nucleateboiling heat transferrdquo Experimental 6ermal and Fluid Sci-ence vol 26 no 6-7 pp 775ndash792 2002

[6] S Yang ldquoExperimental study on critical heat flux of chemicalwater boiling on a downward facing curved surface for IVR-ERVC strategyrdquo Nuclear Power Engineering vol 6 pp 23ndash272016

[7] W Tao ldquoExperimental research on influence of real surfacematerial and aging effect on ERVC-CHF of RPVrdquo AtomicEnergy Science and Technology vol 50 p 1786 2016

[8] H Chang T Hu W Lu S Yang and X Zhang ldquoExperi-mental study on CHF using a full scale 2-D curved test sectionwith additives and SA508 heater for IVR-ERVC strategyrdquoExperimental 6ermal amp Fluid Science vol 84 pp 1ndash9 2017

[9] YMei Y Shao S Gong Y Zhu andH Gu ldquoEffects of surfaceorientation and heater material on heat transfer coefficientand critical heat flux of nucleate boilingrdquo InternationalJournal of Heat amp Mass Transfer vol 121 pp 632ndash640 2018

[10] G Tan and B Kuang ldquoGeometric optimization and reliabilityassessment of reactor vessel insulation for IVR-ERVC basedon passive system functional reliability conceptrdquo AtomicEnergy Science and Technology vol 3 pp 54ndash60 2011

[11] R Guo W Xu Z Cao X Liu and X Cheng ldquoA new methodto study the transient feasibility of IVR-ERVC strategyrdquoProgress in Nuclear Energy vol 87 pp 47ndash53 2016

[12] Y Jin W Xu and X Liu ldquoIn- and ex-vessel coupled analysisof IVR-ERVC phenomenon for large scale PWRrdquo Annals ofNuclear Energy vol 8 pp 322ndash337 2015

[13] X Cheng T Hu D Chen Y Zhong and H Gao ldquoCFDsimulation on critical heat flux of flow boiling in IVR-ERVCof a nuclear reactorrdquo Nuclear Engineering amp Design vol 304pp 70ndash79 2016


the description of a new reactor pit flooding system and theREVECT (Reactor Vessel External Cooling Test) facility areintroduced and the influence factors of cooling capability arestudied Based on the REVECT tests a series of problemswhich have confused the designers in long-time are solvede research results have been applied to the engineeringdesign of reactor pit flooding system and the establishmentof management guidelines for severe accidents in new nu-clear power plants

2 Research Status

After the occurrence of a postulated severe accident coriumwill collapse into lower head of RPV e disposal mode ofcorium becomes a common concern Currently there aremainly two treatment ways for corium cooling (1) Ex-VesselRetention (EVR) this method is mainly used in EPR especific method is setting a core catcher at the bottom ofcontainment which has functions of isolating molten debrisand concrete as well as providing long-term cooling EVRstrategy is also adopted in Tianwan nuclear power plant(NPP) in China (2) In-Vessel Retention (IVR) IVR strategyis implemented to remove decay heat from the RPV and tokeep the corium in the RPV [1] is strategy has also beenused in VVER440 AP1000 HPR1000 and APR1400 [2]

In order to verify the effectiveness of the IVR strategysome different test facilities which are used to simulate theflow and heat transfer characteristics have been built such asULPL test facility and HERMES-HALF test facility econclusions of the tests are as follows (1) An aged coppersurface exhibits a similar coolability performance as theexternal surface of the RPV wall which is verified by ULPUand BETA-NC test [3ndash5] (2) Optimized flow channel canenhance the CHF of the external surface of lower head (3)e chemical properties of some coolants can affect theCHF For instance the coolant with boric acid can decreasethe CHF to some extent

In China different types of CHF tests and related an-alyses have been performed in recent years Based on a large-scale test facility Yang [6] investigated the CHF charac-teristics of chemical solution boiling on a downward facingcurved surface It was found that the CHF of mixed solutionof H3BO5 and Na3PO4 increases firstly and then decreaseswith the increase of Na3PO4 concentration rough a two-dimensional full-scale facility FIRMWei [7] discovered thatthe aging effect could enhance the CHF of SA508 owing tothe generation of Fe3O4 oxide To study the CHF margin ofExternal Reactor Vessel Cooling (ERVC) for ChineseAP1400 the FIRM subcooled flow boiling facility conductedby State Nuclear Power Technology Research amp Develop-ment Center (SNPTRD) was built [8] It was also shown thatthe concentration of Na3PO4 has an important effect on theCHF behavior Based on the experimental data a CHFcorrelation for IVR-ERVC was developed by Mei et al [9]with considering the effects of surface orientation thermaleffusively and corrosion Besides Tan and Kuang [10]preliminarily determined nominal values of reactor vesselinsulation design parameters according to ERVC relatedfunctional reliability criteria and related statistical analysis

Guo et al [11] proposed a newmethod to study the transientfeasibility of IVR-ERVC in which a theoretical CHF modelwas developed for the outer surface of the lower head Jinet al [12] performed the study on in-vessel and ex-vesselcoupled analysis of IVR-ERVC phenomena for large-scalePWR by using MELCOR Cheng et al [13] used a CDF codeFluent coupled with a boiling model by UDF (User-DefinedFunction) to investigate the CHF of ERVC which wasvalidated by experimental CHF values obtained by SNPTRD

Even though a large number of tests and analyses haveconducted there are little test results about the effect of thekey factors on the CHF distribution such as the geometry offlow channel and the water level in the reactor cavity eabove key factors are also essential for the reactor floodingsystem design

3 Design andManagement of Reactor FloodingSystem for Advanced PWR

31 Reactor Pit Flooding System e reactor pit floodingsystem is a dedicated severe accident mitigation system toachieve the IVR-ERVC Generally the reactor pit floodingsystem is composed of a reactor pit injection system and anatural circulation system When a severe accident happenscooling water can be injected into the reactor cavity tosubmerge the RPV via passive injection mode or activeinjection mode e decay heat is finally removed throughthe natural circulation in the reactor cavity to maintain theintegrity of RPV

e passive reactor pit injection subsystem mainly con-sists of the reactor pit flooding tank located on a place higherthan the main coolant pipes of Reactor Coolant System (RCP)in the containment pipelines and valves During a severeaccident after entering the severe accident managementguideline (SAMG) the valves on the passive injection pipesare opened by the operator and the water in the reactor pitflooding tank flows to the reactor pit by gravity with a largeflow rate When the water level in the pit flooding tank dropsto a certain threshold the injection is switched to small flowrate to compensate for the water loss due to evaporationensuring the reactor cavity to be kept flooded

e active reactor pit injection is provided by activewater injection pipelines which connect to the in-contain-ment refuelling water storage tank (IRWST) When thereactor pit flooding tank is depleted the cooling water istaken by pump from the IRWST and is injected into thereactor pit after cooled by Component Cooling WaterSystem or Extra Cooling System e schematic diagram ofpassive reactor pit injection subsystem and active reactor pitinjection subsystem is illustrated in Figure 1

Since the core outlet temperature reaches 650degC which isa significant temperature alarm for the NPP to carry out thesevere accident management guideline (SAMG) the oper-ator opens the valves on the passive injection pipelines tosubmerge the reactor pit with a large flow rate en thewater injection flow is switched to small flow rate phase tomake up the evaporated water Before the IVR tank is emptythe operator starts up the pump to inject water from the in-containment refuelling water storage tank (IRWST)






Containment

IVRtank

M

MM

Reactor pit

IRWST

RPV


Outlet

Shield wall

Backflowchannel

Reactor pit

RPV support ring

RPV

Flow channel

Inlet



















for θle θm (1)


1113946θ

θm


sin θ1113946θ

θm



sin θminus

1sin θm

1113888 1113889 +1

sin θ1113946θ

θm


(2)

DV01

DV03

DV02

DV05 DV04

FV01

DV06

DV07

M102

M101

V101

M103 DV11

DV10

F03

F04

Glass pipe

V103Cooling

tower

F01 F02

V102P101

Filtrator

Check valve

4th platform

3rd platform

2nd platform

1st platform

Steamexhaust port







qp sin θm

1 minus cos θm

1113946θm




















Water inlet

Heating section

2200

Incline baffle

Outlet


baffle

θ

L1

L2

D

C

B

A

2500














0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References



















Containment

IVRtank

M

MM

Reactor pit

IRWST

RPV


Outlet

Shield wall

Backflowchannel

Reactor pit

RPV support ring

RPV

Flow channel

Inlet



















for θle θm (1)


1113946θ

θm


sin θ1113946θ

θm



sin θminus

1sin θm

1113888 1113889 +1

sin θ1113946θ

θm


(2)

DV01

DV03

DV02

DV05 DV04

FV01

DV06

DV07

M102

M101

V101

M103 DV11

DV10

F03

F04

Glass pipe

V103Cooling

tower

F01 F02

V102P101

Filtrator

Check valve

4th platform

3rd platform

2nd platform

1st platform

Steamexhaust port







qp sin θm

1 minus cos θm

1113946θm




















Water inlet

Heating section

2200

Incline baffle

Outlet


baffle

θ

L1

L2

D

C

B

A

2500














0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References































for θle θm (1)


1113946θ

θm


sin θ1113946θ

θm



sin θminus

1sin θm

1113888 1113889 +1

sin θ1113946θ

θm


(2)

DV01

DV03

DV02

DV05 DV04

FV01

DV06

DV07

M102

M101

V101

M103 DV11

DV10

F03

F04

Glass pipe

V103Cooling

tower

F01 F02

V102P101

Filtrator

Check valve

4th platform

3rd platform

2nd platform

1st platform

Steamexhaust port







qp sin θm

1 minus cos θm

1113946θm




















Water inlet

Heating section

2200

Incline baffle

Outlet


baffle

θ

L1

L2

D

C

B

A

2500














0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References

















for θle θm (1)


1113946θ

θm


sin θ1113946θ

θm



sin θminus

1sin θm

1113888 1113889 +1

sin θ1113946θ

θm


(2)

DV01

DV03

DV02

DV05 DV04

FV01

DV06

DV07

M102

M101

V101

M103 DV11

DV10

F03

F04

Glass pipe

V103Cooling

tower

F01 F02

V102P101

Filtrator

Check valve

4th platform

3rd platform

2nd platform

1st platform

Steamexhaust port







qp sin θm

1 minus cos θm

1113946θm




















Water inlet

Heating section

2200

Incline baffle

Outlet


baffle

θ

L1

L2

D

C

B

A

2500














0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References
















qp sin θm

1 minus cos θm

1113946θm




















Water inlet

Heating section

2200

Incline baffle

Outlet


baffle

θ

L1

L2

D

C

B

A

2500














0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References


























0 20 40 60 80 100000102030405060708091011121314


The d

istrib

utio

n of

hea

t flux

(MW

m2 )











52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References




















52 Test Results




00

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18



q CH

F (M

Wm

2 )


30 40 50 60 70 80 90 10000

02

04

06

08

10

12

14

16

18

20



q CH

F (M

Wm

2 )






130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References


















130

390

2200

1210

43deg

170

250

D

C

B

A67deg


30 40 50 60 70 80 90 10000

05

10

15

20


Hea

t flux

(MW

m2 )

Angle (deg)







6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References

















6 Conclusions



0 50 100 150 200 250 300 350 400 450 500 550 600 65005

101520253035404550556065707580859095

100

Flow

rate

(m3 h

)

Pres

sure

(kPa

)

Time (s)


05101520253035404550556065707580859095100


05

1015202530354045505560657075808590

051015202530354045505560657075808590

0 50 100 150 200 250 300 350 400 450 500 550 600 650

Time (s)


Flow

rate

(m3 h

)

Pres

sure

(kPa

)


500 550 600 650 700 750 800 850 900 950 1000 1050175

180

185

190

195

200

Time (116s)


Pres

sure

(kPa

)


0 50 100 150 200 250 300 350 400 450

178

180

182

184

186

188

190

192

194

196

198

Pres

sure

(kPa

)

Time (116s)








Data Availability




Acknowledgments


References



















Data Availability




Acknowledgments


References















Documents

Studies on Key Effect Factors of Natural Circulation