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Nuclear Engineering and Design 265 (2013) 244– 253
Contents lists available at ScienceDirect
Nuclear Engineering and Design
j ourna l h om epa ge: www.elsev ier .com/ locate /nucengdes
hermal hydraulic investigations of an extended station blackoutvent in FBTR
. Natesan ∗, K. Velusamy, P. Selvaraj, P. Chellapandi, S. Varatharajanndira Gandhi Centre for Atomic Research, Kalpakkam 603 102, India
i g h l i g h t s
Plant dynamic modelling of a loop type sodium cooled fast breeder reactor.Demonstration of core safety under natural convection heat removal conditions.Investigation of possibility of sodium freezing under extended station blackout.
r t i c l e i n f o
rticle history:eceived 20 December 2012eceived in revised form 15 July 2013ccepted 16 July 2013
a b s t r a c t
Thermal hydraulic investigation into the capability of Fast Breeder Test Reactor (FBTR) in handling anextended station blackout event was carried out. The main concerns during this event are decay heatremoval and prevention of sodium freezing in circuits. The means of removing decay heat in FBTR are (i)by natural convection of air in steam generator casing and (ii) heat loss through long sodium piping ofthe loop type reactor. The study is focussed towards the current operating power. The plant dynamicscode DYNAM developed specific to this reactor system and validated against commissioning tests has
been utilized for this purpose. It has been established that core cooling is not a concern even when thesystem cooling through steam generator casing is not activated. Piping heat loss is adequate to removethe decay heat. Cooling through SG casing is essential only for preventing heat up of cold leg sodiumsystems. Concern on sodium freezing arises when the decay power reduces to very low level below thenatural heat loss through pipe lines. This situation arises only after a year for the current power level ofthe plant.. Introduction
The Fast Breeder Test Reactor (FBTR) is a 40 MWt, loop typeodium cooled fast reactor, currently under successful operationt Kalpakkam (Srinivasan et al., 2006). This is the first fast reac-or established in India in the year 1985 as a fore-runner to theecond stage of Indian nuclear power programme. Apart from pri-ary and secondary sodium systems, the reactor has steam–water
ystem and turbo-generator to produce electricity. It uses uniqueranium–Plutonium carbide as the fuel material. Currently FBTR
s operating at 22.1 MWt conditions. The operating experience ofBTR has provided sufficient feed-back and confidence to launchhe 500 MWe Prototype Fast Breeder Reactor (PFBR) currently
nder advanced stage of construction in India.Following the Fukushima accident, which originated from atation blackout, countries across the world pursuing nuclear
∗ Corresponding author. Tel.: +91 44 27495215; fax: +91 44 27480104.E-mail addresses: [email protected], [email protected] (K. Natesan).
029-5493/$ – see front matter © 2013 Elsevier B.V. All rights reserved.ttp://dx.doi.org/10.1016/j.nucengdes.2013.07.022
© 2013 Elsevier B.V. All rights reserved.
power programme critically reviewed the safety status of powerplants in operation (Andrija and Andrej, 2011; Lauren, 2011;Swedish National Report, 2011). Prior to this event, safe man-agement of station black out event in a nuclear power plant wasrequired to be demonstrated only for a few hours. Possibility of suchan event was also conceived to be very rare. After the Fukushimaevent, there is greater requirement to demonstrate plant safety fora large duration extended over several days. Towards this, thermalhydraulic analysis of FBTR has been carried out to study the plantbehaviour under an extended station blackout (SBO) condition. Theliquid sodium cooled fast reactors have several advantages com-pared to light water reactors. One of them is that the main coolantsystem during operation is not in pressurized state because of itshigh boiling point (as high as 880 ◦C at atmospheric pressure). Dueto this, the coolant system remains in single phase liquid conditionsfor long time during events where the heat removal from the core
is affected. The decay heat produced in the core can be effectivelyremoved by the coolant if natural convection flow is established.Since air is used as ultimate heat sink in FBTR, its non-availabilityis totally ruled out. If an effective natural convection flow path is
K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253 245
inivas
etofTf
chatetSafltbpootadtotdsagas
Fig. 1. (a) Schematic flow sheet of FBTR (Sr
stablished between the reactor core and the ultimate heat sink,hen SBO event of any duration can be managed in the plant. Thenly hindrance to maintain natural convection flow in the plantor long duration is the possibility of sodium freezing in the circuit.ime at which this crisis can evolve in the plant should be estimatedor charting out the emergency management procedures.
Fig. 1a shows the schematic flow sheet of heat transport cir-uits of FBTR (Srinivasan et al., 2006). FBTR is a loop type reactoraving two primary and two secondary sodium loops. Heat gener-ted in the reactor core is removed by primary sodium loops, and isransferred to secondary sodium loops through intermediate heatxchangers (IHX). Each secondary sodium loop is provided withwo once through steam generators (SG). Steam from all the fourG modules is fed to a common steam water circuit comprising of
turbine-generator and a 100% dump condenser. Primary sodiumow through the core is established by two primary sodium pumpshrough a common header. From the primary vessel, sodium flowsy gravity to IHX and then back to the pump suction. In the IHX,rimary sodium flows on the shell side and secondary sodium flowsn the tube side. Secondary sodium is pumped through IHX by sec-ndary sodium pumps. After removing heat from primary sodium,he secondary sodium enters SG. A surge tank located between IHXnd SG acts as a buffer against pressure wave transmission to IHXuring sodium–water reaction possible in steam generators dueo water leaks. There are no active valves in the primary and sec-ndary sodium main loops. Flow control is achieved by controllinghe speeds of sodium pumps. Flywheels are mounted on the pumprives to provide sufficient inertia to achieve slow coast down ofodium flow during a power failure event. The primary pump drives
re provided with emergency power supply from the station dieselenerators and battery backup (for a limited period). SG modulesre once-through, counter flow type, with sodium flowing in thehell side from top to bottom and water flowing in the tube side.an et al., 2006). (b) Schematic of SG casing.
The steam generator modules are not provided without any insula-tion on the outer surface of their shells. All the modules are housedinside a casing (Fig. 1b). The outer surface of casing is insulated. Atall chimney is located above the casing. By opening the trap doorsat the bottom and top of the casing, natural convection of air will beestablished through the casing. There will be transfer of heat fromsodium contained inside the shells of SG modules to air flowingthrough the casing. By this way, decay heat can be removed fromthe reactor by natural convection to ambient.
Several studies have been carried out by computational eval-uation of natural convection flow in reactor systems and itseffectiveness in core cooling. These studies highlight the impor-tance of experimental validation to establish natural convectioncharacteristics of reactor systems. There is added complexity infast breeder reactor systems due to the use of liquid metal asthe coolant. Eguchi et al. (1997) have proposed an evaluationmethod through water tests for establishing natural circulationcharacteristics in a liquid metal cooled reactor. Natural circulationtests are essential to validate plant dynamics codes. Passive decayheat removal system of a pool type fast reactor has been testedin water/air in the KIWA facility (Costa and Grand, 1989) whichincludes simulation of air heat exchangers also. Water simulation ofpassive heat removal system under steady and transient conditionsof SNR-2 reactor has been carried out in RAMONA (Hoffmann et al.,1995) and NEPTUN (Rust et al., 1995) facilities. Similarly, water sim-ulation of natural convection heat removal of SPX-2 reactor hasbeen carried out in HIPPO (Betts and Ashton, 1991) and GODOM2(Azarian et al., 1990) facilities. Experiments in a sodium loop withseven electrically heated subassemblies have been carried out by
Nishimura et al. (2000) to understand inter subassembly and inter-wrapper heat transfer in core during natural convection conditions.There are also experimental studies reported on lead–bismuth sys-tems (Ma et al., 2007) on flow instability in natural convection
246 K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253
schem
s‘ePui1tc
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Fig. 2. Nodalization
ystems. Several tests on passive safety of experimental fast reactorJoyo’ have been carried out by Takamatsu et al. (2007). COLTEMPxperiment was conducted in the pool type fast breeder reactorhenix at 6 MW power to demonstrate decay heat removal by nat-ral convection in the primary circuit (Valentin et al., 1990). JANUS
s a scaled down sodium model of SPX reactor (Tenchine and Grand,989), used to understand the natural circulation flow pattern andhermal stratification in piping and to validate a computationalodes.
The RAMONA test results have been used basically to vali-ate the DYANA/ATTICA coupled code system (Ertel and Reinders,986; Betts and Ashton, 1991), the TRIO code (Grand et al., 1991),he ASTEC code (Georgeoura and Keeton, 1992) and FLUTAN codeBorgwaldt et al., 1992). The SSC-L (Super System Code-Loop) haseen validated against natural convection tests in the experimentalast reactor, JOYO (Yamaguchi et al., 1989; Ohshima et al., 1993).ETFLOW++ is a code developed by Mochizuki (2010), and the nat-ral convection prediction capability of the code has been validatedgainst experimental results of Nishimura et al. (2000). Super-OPD code developed for sodium cooled fast reactors (Yamadat al., 2009) has been validated against start up tests carried outn MONJU reactor. Mochizuki (2012) has carried out the analysisf extended station blackout event in MONJU reactor using NET-LOW++ code. This study demonstrates the core safety under anBO event when at least one loop is intact. The severity of SBO eventappening during the post shutdown phase of the reactor is demon-trated in these studies. All these studies bring out the importancef natural convection decay heat removal and various reactor spe-ific aspects relevant to natural convection heat removal. Furtherhese studies focus only on short term core cooling. On the otherand, the focus of the present study is long term evolution of SBOvent in FBTR. This paper presents the details of the analysis ofxtended SBO event carried out using the plant dynamics codeYANAM (Vaidyanathan et al., 2010).
. Plant dynamic modelling of FBTR
Mathematical models in the DYNAM code are based on thosesed for most of the fast reactors (Agarwal and Khatib Rahbar,
e of DYNAM code.
1980). Detailed descriptions of various models adopted in the codeare discussed by Vaidyanathan et al. (2010). Nodalization schemeof the code is shown schematically in Fig. 2. Important featuresof the transient one dimensional model are as follows. Sodiumflow through various components of the plant is assumed to beincompressible and single phase through out. In all the placeswhere mixing and recirculating flow patterns exist, a perfect mix-ing assumption is made in the thermal model and incoming kinetichead of flow is assumed to be fully converted into static pres-sure head in hydraulic calculation. Significant stratification effectsare possible during natural convection conditions in the hot pool.However, the reactor being of loop type and design power ratingbeing small the pool volume is very small compared to that of pooltype reactors of commercial power rating. Hence, the stratificationeffects in the pool are considered to be negligible and ignored in thecalculation. This assumption has been found to be valid while com-paring the code predictions against natural convection test results.Axial heat conduction in coolant, pipe material, heat exchangertubes, fuel pellet and clad walls are neglected in comparison toradial heat transfer. The hybrid core FBTR core (consisting of car-bide and oxide subassemblies) is modelled as several radial andaxial regions.
The power generated in the core is assumed to be separa-ble in space and time. Normalized spatial distribution of poweris obtained from diffusion calculations and assumed to remainunchanged throughout the transient. Time variation of power iscalculated by solving the space averaged one energy group kinet-ics equation known as point kinetics approximation. For the smallcore size of FBTR, this approximation is quite adequate (Walterand Reynolds, 1981). The total reactivity at any time is the sumof the applied external reactivity by control rod movement andvarious reactivity feedback contributions. The feedback reactivityconsists of contributions due to (i) fuel expansion, (ii) structuralexpansion, (iii) coolant expansion, (iv) spacer pad expansion and (v)grid plate expansion. Doppler reactivity is negligible for FBTR core
and hence is neglected in the analysis. FBTR uses carbide fuel andhence the neutron spectrum is harder than that in oxide core. Withharder spectrum, the Doppler reactivity effect becomes insignif-icant. Numerical solution of the kinetic equations is obtained by
ering a
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K. Natesan et al. / Nuclear Engine
rompt jump approximation for neutronic power. From this, theotal thermal power is obtained after adding the power due to decayf fission products.
For the purpose of thermal hydraulic analysis, one subassem-ly from each core zone is modelled. Through the application ofnergy balance for each of the axial zones coupled ordinary dif-erential equations for fuel, clad and coolant are obtained. Solvinghem simultaneously in a sequential manner from the bottom toop of the subassembly gives the axial distribution of sodium cladnd fuel temperatures. Hot channel factors are used to account forncertainties in the evaluation of various core temperatures. Forxample, the hotspot temperature of sodium (TNa) and clad (TCl)re obtained by multiplying the nominal values of various tempera-ure drops estimated by the code, viz., average central subassemblyemperature rise (�TCSA), sodium film drop (�Tflm) and clad mid-le to outer surface drop (�Tcl) with appropriate hotspot factorsamely FSA, fch, ffilm and fcl as follows:
Na = TRI + FSA �TCSA
Cl = TNa + fch �TCSA + fflm �Tflm + fcl �Tcl
Various uncertainties can be classified into direct and random.he uncertainties relating to dimensions and other fabrication vari-bles are random in nature. Uncertainties relating to correlationf linear heat rating, thermo physical properties of fuel, clad &oolant and heat transfer coefficient of sodium are systematic orirect in nature. Hotspot factors due to direct uncertainties multi-ly the nominal values directly in a cumulative manner. Randomncertainties in thermal hydraulics follow a normal distribution. 3�
evel confidence has been considered for the estimation of hotspotactors due to random uncertainties and these factors are com-ined together statistically. This approach is similar to that followedor the fast reactor designs, viz., FFTF (Calamai et al., 1974) andRBRP (Carelli and Markley, 1975). Under transient conditions,ot all uncertainties considered for nominal conditions may bealid. Additional uncertainties may creep in under natural convec-ion conditions. Same hot channel factors estimated for nominalonditions have been used for the transient evaluations also. Addi-ional uncertainties under natural convection conditions have beenreated through conservative boundary condition, viz., conserva-ive hydraulic resistance values and conservative heat loss models.
A two zone radially lumped one dimensional model with a singleemperature for the primary sodium and secondary sodium is usedor modelling the heat transfer in the IHX. These assumptions leado a set of two coupled hyperbolic partial differential equations.
eighted nodal heat balance scheme is applied for the derivationf heat balance equations.
The governing equations for the primary sodium circuitydraulics are obtained by making momentum balances betweenarious segments of the circuit, by mass balances for the hot poolnd IHX plenum and by torque balance on the primary pumphaft. Similarly, the governing equations for secondary sodiumircuit hydraulics are obtained by making momentum balancesetween various segments of the circuit, mass balance for surgeank and pump tank and by torque balance on secondary pumphaft. Head developed and impeller-consumed torque of the pumpre obtained from homologous pump characteristics (Streeter andylie, 1967). The sodium pipelines delay the thermal propagation
f sodium temperature changes from one end to other end of piping.his delay depends on the mass velocities and thus they are capa-
le of wide variation during flow transients. Apart from this, thehermal model should also account for heat losses from sodium inpite of the insulation provided. For modelling these effects, nodaleat balance scheme is adopted.nd Design 265 (2013) 244– 253 247
For evaluating the steady state and transient temperatureprofiles of the SG, a radially lumped model with separate represen-tation of shell, sodium, tubes and water/steam in tubes is adopted.For accurate representation of the steam/water side, a coupledsolution of the continuity, energy and momentum equations hasto be carried out. It is apparent that the two-phase compressibleflow analysis is more complicated than its single-phase counter-part. In order to simplify the modelling, only the energy equation issolved in every axial node and the effects due to momentum equa-tion and continuity equations are neglected. Single energy equationmodel is applicable for steam water side when the pressure dropsuffered by water/steam flow in the steam generator is less com-pared to the system pressure. Therefore, energy equation is solvedwith constant pressure assumption in the entire SG. This model isadequate for simulation of transients originating from the primaryand secondary sodium systems. During these transients the pres-sure control system in place (governor) controls the main steamheader pressure very smoothly. However, during transients origi-nating from the steam water system such as rupture of steam line,spurious opening of steam dump valve, etc. the constant pressureassumption cannot simulate the heat transfer in the SG in realis-tic manner. Detailed models with the consideration momentumbalance equation should be adopted for these studies. Neverthe-less, these transients are enveloped by instantaneous loss of steamwater system in a conservative manner for evaluating the conse-quences in the nuclear steam supply system. Hence, from the pointof view of safety evaluation, constant pressure assumption for thesteam/water side of SG is adequate when it is coupled with conser-vative boundary conditions. It may be noted that station blackoutevent is analysed by considering instantaneous loss of water flowthrough SG. Numerical comparisons of the momentum integral andsingle-mass flow rate models have demonstrated that a generalagreement exists (Meyer, 1961). The waterside pressure drop is asmall fraction of the operating pressure and hence the effect of spa-tial and temporal pressure distributions on water/steam propertiescan be neglected. On the water/steam side, specific enthalpy is usedas the state variable and the energy balance equation is formulatedbased on it. DYANM code has been validated through various testscarried out in FBTR including pump coast down test (Vaidyanathanet al., 1994).
3. Station blackout scenario in FBTR
Currently, FBTR is operating with core configuration producing22.1 MW of thermal power. In order to achieve higher temperatureconditions in primary sodium circuit at low power conditions rela-tive to the design power of the plant, the plant is currently operatingwith three out of 7 tubes plugged in each steam generator mod-ule. Under SBO, normal heat removal system will not be availableand decay heat removal during this condition is by natural con-vection through the SG casing. When the trap doors located at SGcasing are opened, natural convection of air is established throughthe casing which enables heat removal from the secondary sodiumflowing through the steam generators. The heat removal capacity ofSG casing as a function of secondary sodium temperature obtainedthrough plant measurements is shown in Fig. 3. In the analysis,heat removal by the SG trap door system is estimated as a func-tion of average temperature of sodium in the SG. This heat sink isdistributed uniformly all along the length of SG. Apart from this,there will be heat loss from sodium flowing in the pipe lines to theambient air. This heat loss is estimated based on local temperature
of sodium in the piping. Piping is also divided into several controlvolumes for accurate estimate of this parameter. Under a prolongedSBO event, electrical line heaters, provided for maintaining sodiumin liquid state without allowing the system temperature to reduce
248 K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253
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Fig. 3. Measured heat removal rate through SG casing.
o freezing point of sodium will not be available. Therefore, whenhe sodium systems get cooled down by natural means, there maye a concern of sodium freezing during a prolonged SBO situation.
. Results and discussion
.1. Safety parameters
SBO event is simulated by considering the trip of all primarynd secondary sodium pumps along with instantaneous loss of feedater flow through SG. The emergency battery power backup pro-
ided for the operation of primary pumps at a low speed (150 rpm)or a short duration of 1/2 h is not considered conservatively. Inhis case, off site power failure leads to SCRAM on LOR (Loweringf Rods) ineffective parameter. Several LOR parameters, viz., (i) Cir-ulating water pump trip, (ii) main boiler feed pump trip, (iii) lowow in feed water system, (iv) trip of drive systems of primary andecondary sodium pumps, (v) low flows in primary and secondaryodium loops, (vi) high mean core outlet temperature and (vii) highean temperature rise across core get triggered during this event.
here will also be SCRAM signals from (i) high mean core outletemperature at 4 s, (ii) high mean temperature rise across core at
s and (iii) low sodium flow through the core (<80%) at 12 s. Thisvent is analysed by considering reactor SCRAM at 6 s by the highore outlet temperature parameter. This is by considering a con-ervative assumed delay of 2 s for the measurement and trip circuitctuation.
.2. Power and flow evolutions
Predicted evolution of reactor power is shown in Fig. 4. Reactorower falls to decay power level within 7 s and reduces subse-uently. Decay power values after 1 day and 10 days are 135 kW and0 kW, respectively. SG trap doors are manually opened at 1800 sdelay for manual action). They are considered to be closed whenhe sodium temperature at the outlet of SG (TSGO) reaches 200 ◦Cnd they are again considered to be opened when TSGO reaches60 ◦C. Short term and long term evolutions of clad hotspot tem-erature are shown in Fig. 5a and b, respectively. Two peaks arebserved in the clad hotspot temperature with the first peak if04 ◦C occurring close to the SCRAM instant and the second peak of75 ◦C occurring during the evolution of natural convection. Whenhe trap door is closed at ∼1 h after its initial opening, small peak-
ng in the clad temperature happens as evident in Fig. 5b. Fig. 6and b shows short and long term evolutions of sodium flows (peroop) in the primary and secondary sodium circuits. Oscillationsith smaller period are observed during the initial duration of the
Time, h
Fig. 4. Decay power evolution in the core.
transient (up to 4 h) before a reasonably stable natural convectionflow is established. It can be observed that whenever the trap doorsare opened, secondary sodium flow oscillates before stabilization.This is due to the presence surge tank and pump tank at the samehigh elevation in the hot and cold leg piping, respectively. Thesetanks delay the propagation of relatively faster thermal transientoccurring in SG (due to trap door opening) to IHX side. Therefore,
10001001010.1Time, h
Fig. 5. (a) Short term evolution of clad hotspot temperature. (b) Long term evolutionof clad hotspot temperature.
K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253 249
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ig. 6. (a) Short term evolution of sodium flows. (b) Long term evolution of sodiumows.
uring natural convection is 2% of its nominal value. When the ther-al transient in the secondary system is slow (after 20 days when
he trap doors are no more opened) and is comparable with the mix-ng time constant of pump and surge tanks, the secondary sodiumow evolves smoothly without any oscillations. Initially at about.3 h, secondary sodium flow increases before a reduction. The flow
ncrease is due to the propagation of cold sodium originating fromhe IHX outlet reaching the steam generator and the downwardravel of cold front through SG enhances buoyancy head. When thisold front reaches the vertical piping towards pump tank, adverseuoyancy develops and flow reduces.
.3. Transient in the reactor vessel
Short term and long term evolutions of hot pool and reactornlet temperatures are shown in Fig. 7a and b, respectively. Theres an increase in the reactor inlet temperature immediately afterhe occurrence of the initiating event (Fig. 6a) due to faster coastown of secondary sodium flow in comparison to primary sodiumow resulting in reduced heat transfer from primary to secondary.fter 1 h, reactor inlet temperature starts increasing in response to
he increase in the secondary sodium temperature at the inlet ofHX. The transient behaviour of this temperature is in response tohat of primary sodium temperature at the outlet of IHX as shownn Fig. 8a and b. When the primary sodium temperature at the IHX
xit decreases and increases in response to the trap door manoeu-ring, the same is communicated to the reactor inlet also. Hot poolemperature reduces initially immediately after the occurrence ofFig. 7. (a) Short term evolution of reactor vessel temperature. (b) Long term evolu-tion of reactor vessel temperature.
the initiating event due to SCRAM and then follows the variation ofreactor inlet temperature.
4.4. Transient in the IHX
It can be observed from Fig. 8a and b that primary sodium tem-perature at the outlet of IHX follows the secondary sodium inlettemperature. Secondary sodium temperature at the outlet of IHXfollows primary sodium inlet temperature very closely up to 120 s.During the initial period (up to 3 h) of establishing natural convec-tion in secondary sodium circuit, oscillations can be observed in allthe IHX end temperatures except in the primary inlet temperature.This is due to the oscillations in the primary and secondary sodiumflows. Reduction in the primary sodium flow through IHX to verylow levels at ∼150 s causes IHX secondary sodium outlet tempera-ture to increase above primary inlet temperature for some time. Atthis low primary sodium flow through IHX, there is negligible heattransfer from primary to secondary (heat transfer is even reversedfor some time in the top region of IHX). It is the oscillation in sec-ondary sodium flow that causes IHX secondary outlet temperatureto reduce to a minimum and then increase at about 0.3 h. However,secondary sodium flow oscillations caused by SG trap door openingdo not cause oscillations in IHX end temperatures. This is due to thefact that these thermal oscillations are mitigated by the pump tank
and there is negligible flow oscillation in the primary side. Rise andfall of IHX end temperatures in response to trap door manoeuvringcan be observed in these figures.
250 K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253
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ig. 8. (a) Short term evolution of IHX end temperatures. (b) Long term evolutionf IHX end temperatures.
.5. Transient in the primary piping
Evolutions of temperature in the primary piping between IHXutlet and reactor inlet are shown in Fig. 9a and b. It can be observedhat temperature of the piping follows the evolution of temper-ture at IHX outlet. Significant difference in the temperature atarious points in the primary piping at a particular instant can bebserved up to about 3 h. This is due to the delay in propagationf thermal transient originating at IHX outlet to reactor inlet. Thehermal transient during this period is faster due to larger temper-ture difference prevailing along the piping than the time constantf propagation along the piping. Subsequently, when the temper-ture difference along piping is reduced beyond 3 h, temperaturesf various points in the piping evolve with negligible difference.
.6. Transient in the SG
Evolutions of sodium temperature at the inlet and outlet ofG are shown in Fig. 10a and b. During the initial period, theres an increase in sodium temperature at SG outlet (TSGO) due tooss of heat sink. Also, there is reduction in the inlet temperatureue to reduction in heat transfer through IHX. Subsequently, whenhe secondary sodium flow reduces below the primary flow, theecondary sodium temperature at IHX outlet increases and thisncrease is communicated to SG inlet after a delay. Temperature
eduction of secondary sodium at the outlet of IHX gets communi-ated to SG inlet at about 0.25 h. This causes SG outlet temperatureo start reducing from 0.35 h and this continues even after 0.5 h dueo the opening of trap doors. When the SG trap doors are openedFig. 9. (a) Short term evolution of primary circuit temperatures. (b) Long termevolution of primary circuit temperatures.
after half an hour, temperature of sodium at the outlet of SG reducesand reaches 200 ◦C at 1 h. The trap doors have to be closed at thistime. In case they are not closed, then sodium temperature canreduce to very low levels. Once the trap doors are closed, TSGOincreases again and reaches 250 ◦C at 2 h. SG trap doors are consid-ered to be opened when TSGO reaches 260 ◦C at 2.2 h. TSGO continuesto increase for some time due to the increase in the inlet tempera-ture and reduces subsequently. It reaches 200 ◦C at 15.2 h and trapdoors are considered to be closed at this time. Secondary sodiumtemperature is maintained between 200 ◦C and 260 ◦C by adoptingthe following strategy for opening and closing of trap doors.
• Open the trap doors when sodium temperature increases above260 ◦C.
• Close the trap doors when sodium temperature reduces below200 ◦C.
This mode of trap door manoeuvring will have to be performedup to 17 days and during this period, hot pool temperature changesbetween 240 ◦C and 280 ◦C. When the trap doors are closed at 1 hsecondary flow oscillates which causes SG sodium inlet and outlettemperatures to oscillate. Whenever the trap door is opened thereis increase in secondary sodium flow which causes the hot front totravel faster and temperature increase results at SG outlet beforeit starts reducing subsequently. Trap door (open/close) position forthis mode of operation is shown in Fig. 11. After this period, the
natural heat loss through pipe lines (at a sodium temperature of260 ◦C) matches with the decay power produced in the core. Sub-sequent, reduction in sodium temperature is governed by decaypower evolution in the core as the heat removal is by natural heat
K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253 251
150
200
250
300
350
400
450
500a
b
600540480420360300240180120600Time, s
Tem
pera
ture
, deg
C
SG in letSG outle t
150
200
250
300
350
400
450
500
10001001010.1Time, h
Tem
pera
ture
, deg
C
SG in letSG outle t
Fo
l2hhsFspbd
0
20
40
60
80
100
120
140
160
400350300250200150100500
Dec
ay p
ower
, kW
0
20
40
60
80
100
120
140
160500450400350300250200150100
System temper ature, deg C
Hea
t los
s, k
W
Decay power Vs time
Heat loss Vs system te mperatur e
In order to bring out the need for cooling the sodium systemsthrough SG casing, analysis has been carried out for SBO eventwithout considering the opening of SG trap doors throughout the
250
300
350
400
450
500
550
600
650
700
750a
Cla
d ho
tspo
t tem
pera
ture
, deg
C
ig. 10. (a) Short term evolution of SG end temperatures. (b) Long term evolutionf SG end temperatures.
osses thorough the sodium piping. That is, sodium temperature of00 ◦C will be reached when the core decay power matches witheat loss in the piping at 200 ◦C sodium temperature. Variation ofeat loss through the primary and secondary piping as function ofodium temperature is shown in Fig. 12. System temperature inig. 12 refers to the average temperature of primary and secondaryodium piping which dictates the heat loss from the circuit. Decay
ower evolution in the core is also shown in the same figure. It cane observed from this figure that heat loss at 200 ◦C matches withecay power in the core after 100 days. Thus, sodium temperatureFig. 11. Status of SG trap door position.
b
Time after shutdown, days
Fig. 12. Decay power evolution and behaviour heat loss through sodium piping.
reaches 200 ◦C only after 100 days. Similarly, the time required forthe sodium temperature to reach 150 ◦C is 360 days. Thus, the con-cern on freezing of sodium in the pipe lines arises only after oneyear.
4.7. Need for cooling through SG casing
200600540480420360300240180120600
Time, s
200
250
300
350
400
450
500
550
600
650
700
750
10001001010.1Time, h
Cla
d ho
tspo
t tem
pera
ture
, deg
C
Fig. 13. (a) Short term evolution of clad hotspot temperature (SG trap doors notopened at all). (b) Long term evolution of clad hotspot temperature (SG trap doorsnot opened at all).
252 K. Natesan et al. / Nuclear Engineering and Design 265 (2013) 244– 253
-8
-6
-4
-2
0
2
4
6
8
00010010111.0
Flow
, kg/
s
Primary flowSecondary flow
tltiotate
iiftaroiwlMp4atoip
F
200
250
300
350
400
450
500
10001001010.1Time, h
Tem
pera
ture
, deg
C
IHX primary inle tIHX primary outletIHX sec ondary inle tIHX secondary ou tlet
Fig. 16. Evolution of IHX end temperatures (SG trap doors not opened at all).
200
250
300
350
400
450
500
00010010111.0
Tem
pera
ture
, deg
C
IHX outletPump inPump ou tCoulouteReactor inle t
Time, h
Fig. 14. Evolution of sodium flows (SG trap doors not opened at all).
ransient. In this case, the decay heat removal is only by natural heatosses through sodium piping. Predicted evolutions of clad hotspotemperature for short and long terms during the event are shownn Fig. 13a and b, respectively. In this case also, only two peaks arebserved during the initial period similar to that in the case with SGrap door opening considered. Further, the peak values reached arelso the same. The value of third peak that occurs at ∼5 h is lowerhan the initial two peaks. Therefore, SG trap door opening is notssential to ensure core safety during this event.
Predicted evolutions of primary and secondary flows are shownn Fig. 14. It can be observed that stable natural convection flows established in primary and secondary sodium circuits after aew minor oscillations at ∼5 h. The minor oscillations are due tohe special layout of secondary circuit in which both pump tanknd surge tank are located at the same high elevation. From theeactor vessel temperature evolution depicted in Fig. 15, it can bebserved that due to the non-opening of SG trap doors, reactornlet temperature shows an increasing trend up to 24 h and after
hich it starts reducing gradually. Similar trend is seen in the coldeg piping of primary and secondary sodium circuits (Figs. 16–18).
aximum temperatures of sodium reached at reactor inlet, hotool, IHX primary outlet and IHX secondary inlet are 466 ◦C, 492 ◦C,69 ◦C and 483 ◦C, respectively. These temperatures are reachedt 24 h and then reduce subsequently. This implies that at 24 h
he heat loss through sodium piping matches with the decay heatf 135 kW. At 24 h, the average temperature of sodium systemss ∼480 ◦C. During this period (i.e., first 24 h), the additional heatroduced in the core after subtracting the heat losses though the200
250
300
350
400
450
500
10001001010.1Time, h
Tem
pera
ture
, deg
C
Reactor inletHot pool
ig. 15. Evolution of reactor vessel temperatures (SG trap doors not opened at all).
Time, h
Fig. 17. Evolution of primary piping temperatures (SG trap doors not opened at all).
piping is utilized for increasing the temperature of cold leg systemsfrom 400 ◦C to 480 ◦C. After 24 h, temperatures of sodium systemsreduce in accordance with the reduction in decay power. Times atwhich sodium system temperatures reach 200 ◦C and 150 ◦C are100 days and 360 days, respectively which are same as that in the
case with trap door opening considered. After the trap doors areclosed, the temperature evolution of sodium systems is decided bythe decay power produced in the core which is independent of the200
250
300
350
400
450
500
10001001010.1Time, h
Tem
pera
ture
, deg
C
SG inl etSG outlet
Fig. 18. Evolution of SG end temperatures (SG trap doors not opened at all).
ering a
odc
taoTo
5
irtdtttdaehtds
R
A
A
A
B
B
C
C
C
E
E
Fuel Cycles (FR09), Japan.
K. Natesan et al. / Nuclear Engine
perating conditions after reactor SCRAM. Therefore, there is noifference between the time of sodium freezing crisis in the twoases with and without trap door opening.
Due to the non-opening of SG trap doors, the cold leg systems inhe primary and secondary sodium circuits get heated up to 485 ◦Cnd they are cooled below 400 ◦C only after 5 days. This heating upf cold leg systems can be prevented by opening the SG trap doors.hus, opening of SG trap doors help in preventing excessive heatingf cold leg systems and early cooling of sodium systems to ∼250 ◦C.
. Conclusions
Thermal hydraulic analysis of extended station black out eventn FBTR for the current 22.1 MWt core configuration has been car-ied out using the plant dynamics code DYNAM. It is establishedhat core safety is not affected by this event even if the SG trapoors are not opened for removing the decay heat. Being a loopype reactor, the heat loss through sodium piping significantly con-ributes to decay heat removal. By suitably manoeuvring the SGrap doors, primary and secondary sodium systems can be cooledown and maintained in the temperature range between 200 ◦Cnd 280 ◦C up to 20 days. After 20 days, trap door opening is notssential and decay heat removal can be managed through naturaleat losses through sodium piping. Times at which sodium sys-em temperatures reach 200 ◦C and 150 ◦C are 100 days and 360ays, respectively demonstrating safety of plant against any risk ofodium freezing.
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