research article temperature response of the htr-10 during

Research ArticleTemperature Response of the HTR-10 duringthe Power Ascension Test

Fubing Chen, Yujie Dong, and Zuoyi Zhang

Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced NuclearEnergy Technology, Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Beijing 100084, China

Correspondence should be addressed to Fubing Chen; [email protected]

Received 5 June 2015; Revised 25 August 2015; Accepted 26 August 2015

Academic Editor: Alejandro Clausse

Copyright © 2015 Fubing Chen et al. This is an open access article distributed under the Creative Commons Attribution License,which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

The 10MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor inChina. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned andperformed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. Inthis study, the power ascension process is preliminarily simulated using the THERMIX code.The code satisfactorily reproduces thereactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature.Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples.THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreementbetween the calculated temperatures and themeasured ones. Based on the comparison results, the THERMIX simulation capabilityfor the HTR-10 dynamic characteristics during the power ascension process can be demonstrated.With respect to the reactor safetyfeatures, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower thanthe fuel temperature limit of 1620∘C.

1. Introduction

The 10MW High Temperature Gas-Cooled Reactor-TestModule (HTR-10), located at the Institute of Nuclear andNew Energy Technology (INET) of Tsinghua University, isthe first High Temperature Gas-Cooled Reactor (HTGR) inChina. As a pebble bed modular HTGR, the HTR-10 utilizesspherical fuel elements containing ceramic coated particles,whilst it adopts helium as coolant and graphite as moderator.Attaining the first criticality in December 2000, the HTR-10experienced the power ascension process in January 2003,with the purpose of raising the reactor power from 30%to 100% rated power (RP). After the successful completionof this power ascension test, the HTR-10 achieved the 72 hfull power operationwith satisfactory technical specificationsmeeting the design requirements very well [1].

The HTR-10 itself is a complicated system characterizedby multi-inputs, multi-outputs, and strong coupling, conse-quently making the power ascension a somewhat complexprocess which needs the close coordination of different power

regulationmethods and encounters the large-range change ofoperation parameters in both the primary and the secondarycircuits. In light of the HTR-10 operation procedure, thestrategy of proportionally increasing the secondary feedwater mass flow rate, the primary helium mass flow rate,and the reactor power was adopted to complete the powerascension test. Through the test results, the practicability andvalidity of the HTR-10 power regulation means were fullydemonstrated. Besides, transient test data of the reactor coreand other components were obtained for the validation ofcodes and models employed in the design process.

Based on the actual test conditions, the HTR-10 powerascension process is preliminarily simulated using the THER-MIX code. The simulation puts the emphasis on the pri-mary circuit and reproduces the reactor transients verywell. Important operation parameters, such as the reactorpower, the primary pressure, and the internals temperatures,are compared with their measured values. Good to fairagreement between the calculation results and the test ones

Hindawi Publishing CorporationScience and Technology of Nuclear InstallationsVolume 2015, Article ID 302648, 13 pageshttp://dx.doi.org/10.1155/2015/302648

2 Science and Technology of Nuclear Installations

Helium blower

Steam generator

Reactor

Hot gas duct

Figure 1: Primary system of the HTR-10.

Hot helium temperatureHelium mass flowHelium pressure

Helium mass flowHelium pressure

Reactivity Feed water temperatureFeed water mass flowFeed water pressure

Steam temperatureSteam mass flowSteam pressure

Cold helium temperature

Reac

tor c

ore

Stea

m g

ener

ator

Figure 2: Operation parameters of the HTR-10.

0 500 1000 1500 2000 2500 30000

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Figure 3: Measured reactivity worth of a single control rod.

Science and Technology of Nuclear Installations 3

TR6

TR5

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SR3

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BR3

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CB8

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CB6

TR1

SR12

SR10

SR8

FD5

CB7

CB5

Figure 4: Thermocouples in the reactor internals.

Table 1: Main design parameters of the HTR-10.

Parameter Unit ValueReactor thermal power MW 10Average power density MW/m3 2.0Primary helium pressure MPa 3.0Primary helium inlet/outlet temperature ∘C 250/700Secondary water inlet/steam outlet temperature ∘C 104/440Feed water mass flow rate kg/s 3.49Steam outlet pressure MPa 4.0

proves the THERMIX simulation capability for the HTR-10dynamic characteristics during the power ascension process.

2. Power Ascension Process

The HTR-10 primary system in side-by-side arrangementessentially consists of a reactor, a steam generator, a hot gasduct, and a helium circulator, as shown in Figure 1, wherethe primary coolant flow direction is also illustrated. Someimportant design parameters are listed in Table 1 and thedetailed design information is referred to elsewhere [2].

Since theHTR-10 steamgenerator is of once-through typeand runs at a medium secondary pressure, water-steam two-phase flow instability may occur in the secondary circuit,especially when the feed water mass flow rate is less than 30%of its rated value [3]. Accordingly, two operation stages areconsidered in the HTR-10 operation procedure.

(1) Start-Up Operation Stage. At this stage, the reactor poweris at 0–30%RP,while the primary heliummass flow rate keepsat 30% of its rated value, and so does the secondary feed watermass flow rate. At this time, the outlet steam cannot meet therequirements of the steam turbine. Alternatively, the steam isbypassed to provide process heat.

(2) Power Operation Stage. At this stage, the reactor isoperated with a power level of 30%–100% RP. Meanwhile, thehelium mass flow rate and the feed water mass flow rate areproportional to the reactor power:

𝑃

𝑃0

=𝑀𝐻

𝑀𝐻0

=𝑀𝑊

𝑀𝑊0

, (1)

where 𝑃, 𝑀𝐻, and 𝑀

𝑊are the reactor power, the helium

mass flow rate, and the feed watermass flow rate, respectively,while the subscript 0 denotes the rated value.When theHTR-10 is operated in this stage, it can be used for electricitygeneration because the outlet steam is of quality to drive thesteam turbine.

At the HTR-10 power operation stage, main operationparameters as well as the relationship among them are brieflypresented in Figure 2.

Based on theHTR-10 dynamic characteristics, three kindsof regulation methods are mainly adopted to change theoperation parameters in both the primary and the secondarycircuits, as listed below [4].

(1) Introducing Reactivity by Moving the Control Rod. As thefirst reactivity control system, 10 control rods are designed forthe HTR-10. These control rods are symmetrically placed intheir corresponding channels in the side reflector and eachof them has the same amount of neutron absorber.Therefore,reactivity worth of every control rod can be considered iden-tical. In the HTR-10 commissioning phase, reactivity worthmeasurement tests were carried out several times for thecontrol rod system [5]. In the helium atmosphere, the integralreactivity worth of a single control rod is about 14× 10−3 Δk/k,and the linear segment of the integral curve is in the rodposition of 600–1500mm, as shown in Figure 3. Through thecontrol rod movement, reactivity can be introduced into thecore, thus influencing the reactor power directly.


0. 6.5

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Figure 5: Gas convection model of the HTR-10. (1) Reactor core; (2, 3) flow channel in the bottom reflector; (4) hot helium plenum; (5) topcavity of the core; (6) non-flow region; (7) bottom cavity of the RPV; (8, 9) bottom coolant channels; (10) flow channel in the side reflector;(11, 12) throttle plate; (13) control rod channel; (14) cold helium plenum; (15) small plenum in the bottom reflector; (16) inlet cavity of theRPV; (17) annular space of the RPV; (18) flow channel in the top reflector; (19) leak flow region.

(2) Changing the Helium Mass Flow Rate by Adjusting theHelium Circulator Rotation Speed. The HTR-10 helium circu-lator, integrated into the upper part of the steam generatorpressure vessel, is a vertical single-stage centrifugal compres-sor driven by an electric motor. Under normal operation,the helium coolant is pumped by the helium circulator andcompletes its circulation in the primary circuit after takingheat from the reactor core to the steam generator. Connectedto the electricmotor by a coupling shaft, the helium circulatoris powered by a transducer. The rotation speed of the heliumcirculator is proportional to the output frequency of thetransducer, and 1Hz variation in the output frequency resultsin 60 rpm change in the rotation speed. Regarding the heliummass flow rate, it is proportional to the helium circulatorrotation speed. The latter can be precisely controlled from10% to 100% of its rated speed, so the helium mass flow

rate can be adjusted to satisfy different operation conditionswithout any doubt.

(3) Changing the Feed Water Mass Flow Rate by Adjustingthe Feed Water Pump Rotation Speed. In the HTR-10 sec-ondary circuit, the feed water is driven by its pump to thesteam generator, wheremedium-pressure, superheated steamis produced after the heat exchange between the primaryhelium and the secondary water. The power supply of thefeed water pump is also controlled by a transducer, whoseoutput frequency can be adjusted. Hence, the watermass flowrate can be proportionally regulated via the feed water pumprotation speed that is proportional to the transducer outputfrequency.

The objective of the power ascension test is to raise thereactor power from 30% RP to 100% RP via the coordination


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Figure 6: Helium mass flow rate and helium inlet temperatureduring the test.

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Figure 7: Reactor power during the test.

of different regulation methods, while the outlet steamtemperature should be kept unchanged as far as possible forthe sake of turbine generator stable operation. In addition,other parameters are not allowed to exceed their operationlimits during the power ascension process. According to theHTR-10 dynamic characteristics, a specific test procedure wasformulated for the power ascension process. And the mainpoints of the procedure are depicted as follows:

(1) Thewater and the heliummass flow rate are increasedsuccessively. Due to the internal negative temperaturefeedbackmechanism, the reactor power will automat-ically rise to a higher level.

(2) If the power increase is not sufficient to maintainthe outlet steam temperature, then the control rod

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Figure 8: Helium pressure during the test.

0 1 2 3 4 5 6 7 8 9 10 11 12500

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Time (h)

TestTHERMIX

Tem

pera

ture

(∘C)

Figure 9: Helium outlet temperature during the test.

should be withdrawn for an appropriate intervalto compensate the downtrend of the outlet steamtemperature.

(3) The above-mentioned actions should be repeateduntil the reactor power achieves 10MW power leveland the helium outlet temperature reaches 700∘C.

Before the power ascension test, the HTR-10 was undernormal operation with a power level of 30% RP, a primaryhelium pressure of 2.8MPa, a helium inlet temperature of212∘C, and a heliumoutlet temperature of 675∘C. In the powerascension process, the reactor power, the helium mass flowrate, and the water mass flow rate went up proportionally,and the outlet steam temperature always stayed at about420∘C. During the test, the reactor power is monitored by


30% rated power

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−1939−219−65

03672

108144180195210

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Tem

pera

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(∘C)

Figure 10: Temperature fields at different power levels.

the nuclear measurement system [6]. This system is com-posed of three kinds of ex-core neutron flux instrumentationfacilities covering the source range, the intermediate range,and the power range. Other key parameters are recorded bythe thermal measurement system [7]. Due to the internalstructure limitation, there is no proper flow passage accom-modating the common flowmeters, for example, the orificeflowmeter. Thus, the helium mass flow rate is calculated bya specific formula as a function of the helium pressure, thecold helium temperature, and some related parameters of thehelium circulator, such as themotor power, the pressure head,and the rotation speed. The cold and the hot helium tem-peratures are measured by class 1E thermocouples installedat the outlet and the inlet sections of the steam generator,respectively.Themeasuring points of hot helium temperatureare in the downstream location of the hot helium plenum,so the coolant here has been adequately homogenized withhigh mixing degree [8]. Figure 4 shows the distribution ofthe temperature measuring points arranged in the reactorinternals [9]. TR1∼TR6 represent two symmetrical columnsof thermocouples in the top internals. SR1∼SR6 and SR7∼SR12 are two rows of thermocouples in the side internalsat two different heights. In the bottom reflector, BR1 andBR3, and BR2 and BR4 are installed in pairwise symmetricalpositions. Around the fuel discharging tube there are also twosymmetrical columns of measuring points labeled as FD1∼FD6. Similarly, CB1∼CB8 are located symmetrically in thebottom carbon brick.

3. Analysis Methods

The system analysis code THERMIX is applied for thesimulation of the HTR-10 power ascension test. This codecan analyze the thermal-hydraulic performance of pebble bed

HTGRs under normal operation and accident conditions. Asa modular software package, THERMIX mainly comprisesanalysis modules for neutron kinetics, solid heat conductionin reactor, gas convection in reactor, and fluid flow in primarycircuit. Brief descriptions of these modules are presented asfollows [10, 11].

3.1. Neutron Kinetics Module. In this module, nuclear char-acteristics are evaluated by a conventional point kineticsmodel with six groups of delayed neutrons. Fission power iscalculated by a balance of feedback reactivity and externalreactivity. The former results from the variation of fuel,moderator, and reflector temperatures as well as the change ofxenon concentration, while the latter is caused by the move-ment of control rods. In addition, decay heat is estimated bykinetic equations of fission products.

3.2. Solid Heat ConductionModule. Thismodule consists of atwo-dimensional transient temperaturemodel for solidmate-rials and a one-dimensional transient temperature model forspherical fuel elements. It solves the time-dependent generalheat conduction equationwith temperature-dependentmate-rial properties. Main components of the HTR-10, such as thepebble bed core, the graphite reflectors, the carbonbricks, andthe reactor pressure vessel, can be represented using differentmaterial compositions.

3.3. Gas Convection Module. In this module, a quasi-stationary gas flow model is used to simulate the complexflow conditions in a pebble bedHTGR. Coolant flow throughthe fuel elements is regarded as the flow in the homogeneousmedia. Coupling with a given time-dependent temperatureprofile of the solid structures, this module solves steady-state continuity, momentum, and energy equations of gas in


0 1 2 3 4 5 6 7 8 9 10 11 12Time (h)

TR1TR4THERMIX

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TR2TR5THERMIX

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TR3TR6THERMIX

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mpe

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(∘C)

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Figure 11: Temperature transient of the top internals.

the reactor. The HTR-10 gas convection model, which isa two-dimensional axisymmetrical one in (r, z) geometry,contains 19 different flow regions divided into 18 radial and38 axial mesh points, as shown in Figure 5. The calculatingmodel covers the main flow passages in the HTR-10, forexample, the reactor core, the cold helium channels, the coldand the hot helium plena, and the control rod channels.

3.4. Primary Circuit Module. This module calculates pres-sure, temperature, and mass flow rate of coolant in theprimary circuit. Using a quasi-stationary model composedof steady-state continuity, momentum, and energy equationsof fluid, this module can model different components in theHTR-10 primary circuit, involving the hot gas duct, the steamgenerator, the helium circulator, and so forth.

Besides the past validation work, the THERMIX code isnow further checked in INET against the test data from theHTR-10 with the aim of verifying its simulation capability fordifferent scenarios of pebble bed HTGRs [12].

4. Simulation Results

Based on the initial operation parameters listed in Section 2and the calculating models established in Section 3, thepower ascension process of the HTR-10 is simulated usingthe THERMIX code. The preliminary simulation puts theemphasis on the primary circuit so that it decouples thesteam generator model. As a result, the following inputsobtained from the test are adopted by the computation: (1)the primary helium mass flow rate; (2) the primary helium


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Figure 12: Temperature transient of the side internals (upper part).


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inlet temperature which is affected directly by the secondaryoperation parameters, for example, the water mass flow rate;(3) the control rod position history which is converted intopositive reactivity introduced into the reactor core accordingto the reactivity worth curve given in Figure 3. The heliummass flow rate and the helium inlet temperature recordedduring the test are given in Figure 6.

It was confirmed that the HTR-10 had achieved itsequilibrium state before the power ascension test. After theinitiation of the test, the reactor power was gradually raisedup to 100% RP in accordance with the test procedure. Whenthe reactor came to 50% RP and 75% RP, it stayed at thosepower levels for some time. The whole simulation lasts 12 h,including 3 h for the steady-state operation under 30%RPand2.6 h for the full power operation.

As illustrated in Figure 7, the THERMIX code accuratelyreproduces the reactor power transient during the test pro-cess. Following the power ascension from 30% RP to 100%RP, the primary helium pressure concomitantly rises from2.8MPa to 2.9MPa, as depicted in Figure 8, from which itcan be seen that the calculated pressure corresponds very wellwith the test one. Moreover, the test result shows that theprimary helium outlet temperature reaches 700∘C when thereactor is operated under 100% RP, as shown in Figure 9. Adifference of about 10∘C exists between the calculation andthe test. At present, such deviation is preliminarily attributedto the mass flow rate used as an input condition by thesimulation, because the test value was derived from theindirect calculation rather than the direct measurement.

Temperature fields of the reactor under 30%RP and 100%RP are shown in Figure 10, with the broken lines indicatingthe representative configuration of the pebble bed core. Forthe two different power levels, some similar phenomena canbe qualitatively observed from their corresponding temper-ature fields: (1) both the radial and the axial temperature

gradients are greatly intensive in the core; (2) heat transportin the top and the bottom ceramic internals is mainlydetermined by the longitudinal heat transfer, while the onein the side ceramic internals is primarily dependent on thelateral heat transfer; (3) with regard to the axial temperatureprofiles in the pebble bed, the core temperature increasesalong the axial direction at first and then descends afterachieving a maximum value; (4) at the core inlet, the radialtemperature distribution is nearly flat, but in the lower part ofthe core the central zone is much hotter than the peripheralzone.

However, temperature values of the HTR-10 main com-ponents are raised up to different extent along with the powerascension. For example, temperatures of the core region, theside internals, and the bottom internals are in the ranges of200–900, 200–600, and 200–800∘C, respectively, at 30% RP.For the 100% RP case, the three temperature ranges are 300–1000, 200–700, and 200–900∘C. Due to the combination ofthe axial power distribution and the helium flow direction,the hottest spots at the two power levels both appear at thepoints of 𝑅 = 0 cm and 𝑍 = 180 cm, which means theintersection of the core centerline and the bottom surface ofthe cylindrical part of the pebble bed. And themaximumcoretemperatures are 875 and 953∘C, respectively.

Asmentioned in Section 2, there are a number of thermo-couples installed in the reactor internals. In this study, tem-peratures of the measuring points in the following compo-nents are calculated and compared with the measured values:the top internals, the side internals, the bottom reflector, andthe fuel discharging tube. Since the two-dimensional axisym-metrical calculating models are used by the THERMIX codeand the azimuthal temperature distribution cannot be takeninto account at present, one calculated temperature will becompared with twomeasured values for the top internals, thebottom reflector, and fuel discharging tube where every two


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Figure 15: Temperature transient of the fuel discharging tube.

symmetrical thermocouples are in different circumferentialpositions but have the same radius and height.

Figure 11 shows the temperature transition of the topinternals. During the test process, TR1 and TR4 both expe-rience a temperature rise. THERMIX basically predicts suchvariation tendency, although it overestimates the temperaturevalues of these two measuring points. For TR2 and TR5,the calculated temperature which almost keeps unchangedis higher than their test values from which a temperatureincrease can be observed. Likewise, measured temperaturesof TR3 and TR6 show a higher rise than the correspondingcode result. Based on the comparison, the maximum calcu-lation deviation in the top internals is found to be 36∘C thatoccurs at the measuring position of TR1.

Figure 12 presents the temperature transient in the upperpart of the side internals. In this part, the calculated tem-perature of SR1 agrees very well with the test one. Asregards the other five thermocouples, obviously the calcu-lated temperature curves are all above the test ones. However,THERMIX still reproduces their general behaviors with alargest deviation of 46∘Cwhich appears at thermocouple SR5.

Comparatively speaking, more accurate simulationresults are obtained at the measuring points of the lower part,as shown in Figure 13. Except for SR11, all the thermocouplesget satisfactory prediction temperatures which are in goodaccordance with the measured ones. At the position ofSR11, THERMIX generates the correct temperature changetendency and the largest discrepancy is 48∘C.


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Figure 16: Calculated maximum fuel temperature during the test.

Figure 14 gives the analysis and the test temperaturetransients of the bottom reflector. For the first pair ofthermocouples, the calculated temperature curve is locatedbetween the test curve of BR1 and the one of BR3, so it isconsidered that the code result is pretty reasonable. In thebottom reflector, thermocouples are located below the coreoutlet channels through which hot heliumwith different hightemperatures flows to the hot helium plenum for uniformmixing. Compared with BR1 and BR3, the location of BR2and BR4 is closer to the core centerline and that makes themaffected by the hotter helium from the core outlet. Thus, anunderestimation can be seen from the comparison betweenthe analysis and the test temperatures of BR2 and BR4 andthe maximum deviation is 141∘C. However, the calculatedcurve still reflects the actual temperature change tendencyand becomes closer and closer to themeasured curves duringthe test process.

The temperature variation of the fuel discharging tubeis presented in Figure 15. Whether for FD1 and FD4 or forFD3 and FD6, sufficient agreement between the calculatedtemperature and the test ones is obtained. Underestimatingthe measured temperatures of FD2 and FD5, THERMIXsimulates the general trend and gets a better result whenthe reactor is at 100% RP. At the beginning of the test, themaximum calculation deviation of FD2 and FD5 reaches97∘C.

With respect to the safety features of the HTR-10, themaximum fuel center temperature rises from 900 to 1020∘C,and it does not exceed the limit value 1620∘C all through thetest time, as shown in Figure 16.

5. Conclusions

The power ascension test was conducted on the HTR-10 withthe purpose of raising the reactor power from 30% to 100%rated power. The test results prove the practicability and

validity of the three power regulation means of the HTR-10and lead to the following main findings:

(1) The strategy of proportionally raising the secondaryfeed water mass flow rate, the primary helium massflow rate, and the reactor power facilitates the goal ofthe power ascension test.

(2) Main operation parameters, such as the reactorpower, the helium inlet/outlet temperatures, and thehelium pressure, change smoothly during the wholetest process. In the meantime, the outlet steam tem-perature keeps stable. As a consequence, none of theoperation limits is exceeded.

The THERMIX code is used to simulate the HTR-10 powerascension process. The simulation reproduces the reactortransients very well, for example, the reactor power, thehelium pressure, and the helium outlet temperature. Addi-tionally, the reactor internals temperatures are calculated andcompared with actual values recorded by the thermocouples.Generally speaking, THERMIX can simulate the generaltemperature change trend of different measuring points andthe calculated temperatures show good to fair agreementwiththe test ones. Some calculation differences are larger (141∘Cfor BR2 and BR4, 97∘C for FD2 and FD5) in some smallhot zones beyond the range of the model symmetry at thebottom of the reactor internals. Considering the measuredtemperatures that exceed 750∘C, the relative deviations arenot so great. On the basis of the above-mentioned compar-isons, the THERMIX simulation capability for the HTR-10dynamic characteristics during the power ascension processcan be demonstrated.

With respect to the reactor safety features, it is of utmostimportance that the maximum fuel center temperature dur-ing the test process is always much lower than 1620∘C, whichis the limit value of the HTR-10 fuel.

Conflict of Interests

The authors declare that there is no conflict of interestsregarding the publication of this paper.

Acknowledgment

This work has been supported by the Chinese National S&TMajor Project (Grant no. ZX069).

References

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