thermal hydraulics system for SCWR

Annals of Nuclear Energy 45 (2012) 37–45

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Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR Khurrum Saleem Chaudri a,b, Yali Su a, Ronghua Chen a, Wenxi Tian a, Guanghui Su a, Suizheng Qiu a,⇑ a b

School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, China Pakistan Institute of Engineering and Applied Sciences (PIEAS), Nilore 45650, Islamabad, Pakistan

a r t i c l e

i n f o

Article history: Received 28 September 2011 Received in revised form 16 January 2012 Accepted 19 February 2012 Available online 22 March 2012 Keywords: Sub-channel HPLWR SCWR Coupled analysis

a b s t r a c t Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in the coupled calculations. This difference is due to the use of different values of fuel temperature to include Doppler feedback in our and reference coupled systems. This also lays emphasis on the use of true representative values of critical parameters in the design calculations to get the real picture of conditions rather than over or under estimated values. Ó 2012 Elsevier Ltd. All rights reserved.

1. Introduction Supercritical Water Reactor (SCWR) is an emerging 4th generation reactor. Higher thermal efficiency so cheap electricity, simplified system due to one phase coolant/moderator flow so low capital cost, absence of boiling crisis during normal operation due to no change of phase, low flow rate as compared to Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) due to large specific heat around pseudo critical point, developed secondary side due to experience with supercritical water fossil fueled power plants and small containment size are few of the advantages that SCWR can bring to the table. Thermal (Liu and Cheng, 2009; Oka et al., 1992; Squarer et al., 2003), fast (Oka and Koshizuka, 1998) and mixed (Liu and Cheng, 2007) spectrum kind of reactor concepts are being studied for SCWR. Pressure vessel (Liu and Cheng, 2009; Oka et al., 1992; Squarer et al., 2003) and pressure ⇑ Corresponding author. Tel.: +86 29 82668648; fax: +86 29 82667802. E-mail address: [email protected] (S. Qiu). 0306-4549/$ - see front matter Ó 2012 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2012.02.014

tube (Shan et al., 2009) kind of reactor designs are under study. Major work about SCWR is being done in Japan, Europe, China, Canada and Korea. European design is known as High Performance Light Water Reactor (HPLWR) (Squarer et al., 2003). In supercritical phase of liquids, thermal hydraulic behavior of the system is quite different from normal PWR or BWR system. For example, due to operation under supercritical pressure, there will be no Onset of Nucleate Boiling (ONB) or Departure from Nucleate Boiling (DNB) in SCWR. Instead, the phenomena of Onset of Heat Transfer Deterioration or Heat Transfer Deterioration (HTD) are encountered. Heat transfer correlations are different from subcritical water reactor. Due to large and abrupt variation of moderator and coolant density, both thermal hydraulics and neutronics properties vary drastically over the active length of the fuel. The number of thermal hydraulics codes being used in nuclear industry for research regarding supercritical water reactor are very few e.g. (Shan et al., 2009; Yoo et al., 1999) Conventionally used thermal hydraulics codes like COBRA (Basile et al., 1987) are being modified to work with supercritical water reactors but still are in developing

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

Nomenclature m v s Dz A g f h k /2 w G p

axial mass flow rate (kg/s) cross flow rate (kg/s) length faced into neighboring channel (m) length of control volume (m) flow area (m2) gravitational acceleration (m/s2) friction coefficient specific enthalpy (J/kg) pressure drop coefficient of grid spacer two-phase friction multiplier coolant velocity (m/s) mass flux (kg/m2 s) pressure (Pa)

phases (Ammirabile, 2010). Hence the scarcity of computational codes for thermal hydraulic analyses of supercritical water systems presents a potential challenge and opportunity for research. The recent advancements in the field of computer sciences have opened new horizons for scientific community. Calculations which took days or weeks can now be done in a matter of minutes or hours. Availability of high computational capability systems have started many new activities focusing on the better understanding and inclusion of interaction between neutron kinetics and thermal hydraulics in the field of nuclear reactor design and safety. A project named CRISSUE-S was launched by Europe in 2002–2003. The main emphasis of the program was to re-evaluate critical issues in nuclear reactor design and safety by implementing 3D neutronics/ thermal hydraulics coupled systems. The reports presenting the findings of this project have strongly suggested the use of coupled calculations for safety evaluation of LWRs (Partners, 2004a,b,c). Design studies of the supercritical water fuel assemblies need coupled analysis using 3D neutronics and thermal hydraulics calculations. This is due to the fact that density of coolant can vary from a value of above 700 kg/m3 to less than 100 kg/m3 over the active fuel length. Conventional design studies use the approach of constant coolant/moderator density throughout the fuel assembly (Oka et al., 1992). Looking at the magnitude of density variation of SCWR, the best approach is to perform the coupled neutronics/ thermal hydraulics analyses. Studies conducted have shown the difference between coupled and non-coupled systems is very large (Waata, 2006). Design and safety studies for all kinds of SCWR are adopting the coupled calculations methodology. For HPLWR, various studies have been performed on the recent 3-pass core design (Schulenberg et al., 2008). Analyses about finding the equilibrium core (Maráczy et al., 2011; Monti, 2009), doing the burn-up calculations (Reiss et al., 2008) or performing the safety analyses (Maráczy et al., 2010) are being performed using coupled calculations. A new design for HPLWR is also proposed by the name of Simplified Super Critical Water Reactor (SSCWR) by using zirconium hydride to compensate for under moderation present in SCWR (Reiss et al., 2010). The account of Japanese SCWR (thermal and fast) design methodology, design calculations and safety studies can be found in this reference (Oka et al., 2010). Studies about pressure tube type SCWR design using coupled analyses can be found in this work (Shan et al., 2009). Design of a new fuel assembly using coupled calculations for thermal SCWR is proposed by Chinese researchers (Liu and Cheng, 2009). A new kind of SCWR reactor is also proposed by the same team known as mixed spectrum SCWR (Liu and Cheng, 2010). All these research activities emphasize the need and importance of coupled analyses for SCWR systems.

D l Kg q t

q a b

u i, j, k l v

fuel rod outer diameter (m) turbulence length (m) frictional pressure drop coefficient for transverse flow linear power density (W/m) temperature (K) density (kg/m3) void fraction turbulence factor fuel rod fraction included in the channel sub-channel subscripts liquid phase vapor phase

The selection of MCNP code in coupled calculation system offers certain clear advantages to perform neutron kinetics calculations. A very detailed 3D geometry modeling is one of them. We do not have to use approximations to model the geometry as is the practice with diffusion theory and transport theory codes (Monti, 2009). Quite complex cross section modeling techniques are to be used for coupled calculations with diffusion and transport theory codes (Watson and Ivanov, 2002). For MCNP higher temperature cross section modeling, we just need to prepare new cross section libraries. To avoid generating a large number of libraries at each possible temperature, mixture technique can be used to interpolate the cross section value between two available libraries (Bernnat et al., 2000). In this paper, first the development of SACoS code is explained briefly. Then, a coupled system using Monte Carlo N-Particle (MCNP) code (Briesmeister, 2000) version 4c and SACoS is introduced which can perform coupled design calculations of SCWR fuel assembly.

2. Sub-channel Analysis Code of Supercritical reactor (SACoS) Using the basic theory of sub-channel analyses, Sub-channel Analysis Code of SCWR (SACoS) has been developed. Fig. 1 represents the sub-channel nomenclature used for basic conserved

gap k channel n=l+l'-i

channel i

j+1

m ij

j+1

P ij

j

q"ij Δz

h ij α

ij

ρ

ij

T ij

v kj axial interval j v' kj

m ij-1 P ij-1 Fig. 1. Basic quantities sub-channel nomenclature.

axial level j-1

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

quantities. A coolant centered sub-channel scheme is used. The basic parameters are initialized in the main code while the geometric parameters of the assembly, to be thermal hydraulically evaluated, along with initial and boundary conditions are supplied in the input files. The basic equations used in SACoS code to get the solution for thermal hydraulic parameters are given here. Mass conservation equation:

mijþ1  mij ¼ Dz

ni X

Begin Read input data Axial momentum Update dP

v ijþ1;k slk qik;j

ð1Þ

k¼1

N

Energy conservation equation: mi q þq mijþ1 hijþ1  mij hij X uik kjþ1 kj  ¼ Dz 2 k¼1    ni  X hij qij þ hkj qkj sik ðt ij  t kj Þ bGik sik ðhij  hkj Þ þ kcij þ v ijþ1;k sik 2 lik k¼1

ð2Þ

Mass balance Y

Lateral momentum Update dP Update

Mass balance Y

ni X   mijþ1 wijþ1  mij wij ¼ Ai Pijþ1  Pij  Ai qij g Dz  Dzsik v ijþ1;k Gik;j k¼1 2

Crossflow rate

N

Axial momentum balance equation:

ni X

Coolant velocity

Update

1 Dz/ k 1 1 f b Dzsik Gik;j ðwij  wkj Þ  þ þ  2 D q q q q h l ij ijþ1 ij k¼1

Energy balance

!!  2 mij Ai

Enthalpy

Update

N

Convergence

ð3Þ

2 P P 1 Kg Dzðqij þ qkj Þðv ij;k Þ v ijþ1;k Gik;jþ1  v ij;k Gik;j ¼ Dz ij kj  2 lik 2lik

Mass flow rate, temperature distribution

Y

Lateral momentum balance equation:

! ð4Þ

Eqs. (1)–(4) are solved numerically to get the basic thermal hydraulics parameters. The calculation flow chart for SACoS is shown in Fig. 2. This figure illustrates the calculation methodology of the code. Axial momentum and lateral momentum equations are solved first and with each of them, the mass conservation is checked. After the conservation of basic quantities (mass and momentum), the energy balance is performed. This process goes on until the convergence is achieved. After the convergence, velocity, mass flow rate and temperature are printed out in separate output files. Specific channels needed for thermal hydraulic calculation of SCWR i.e. sub-channels adjacent to water box along with conventional sub-channels can be modeled in this code. Heat transfer of coolant with moderator can also be calculated. Both co-current and counter-current flow modes can be simulated using SACoS code. Different heat transfer models and flow friction correlations used for different flow regimes used in SACoS code are given in Tables 1 and 2 respectively. In the case of supercritical water regime, heat transfer is calculated by Bishop Correlation (Bishop et al., 1965) as suggested by research review and application to HPLWR work carried out by researchers (Cheng and Schulenberg, 2001). To check SACoS performance, geometry shown in Fig. 3 was solved with both COBRA and SACoS. The description of the problem being referred to in Fig. 3 is a 17  17 PWR fuel rod bundle with 25 unheated rods. Here, unheated rods refer to thimble guide tubes which can be used for control rod and instrumentation. Data for this assembly is given in Table 3. Due to symmetry present, only 1/8th assembly is modeled in COBRA and SACoS. The steady state calculations are performed at 50% increase of nominal power. The comparison of results obtained from both the codes is shown in Figs. 4–6. Fig. 4 shows the void fraction along the active height in sub-channel 10 and 26 calculated by COBRA and SACoS. Fig. 5 shows the coolant temperature profile along axial length cal-

Maximum fuel Temperature, DNBR Output End Fig. 2. SACoS calculation flowchart.

Table 1 Heat transfer models used in SACoS. Flow/heat transfer regime Single phase flow Re < 2500 Re < 2500 Sub-cooled boiling Saturated nucleate boiling Super critical water

Heat transfer correlation Collier correlation Dittus–Boelter correlation Jens–Lottes correlation Chen correlation Bishop correlation Dittus–Boelter correlation Watts and Chou correlation

culated by both codes for sub-channel 9, 10 and 45. Fig. 6 shows centerline fuel temperature calculated by the two codes for fuel rods 10, 32 and 45. The results shown in Figs. 4–6 justify the statement that SACoS and COBRA are in excellent agreement with each other and that the SACoS can be used safely with other systems also. 3. Coupled MCNP and SACoS system The large density variation axially in a SCWR makes it almost compulsory to carry out coupled calculations for design and safety calculations. The recent advancement in computational capability has enabled us to perform coupled calculations even for computationally expensive systems i.e. Monte Carlo and Computational

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

Table 2 Flow friction correlations used for different flow regimes. Flow regimes

Darcy friction coefficient correlations

Re < 2000

f = C/Re different values for C for various geometrical channels f = 0.048 Blausius correlation or McAdams correlation Blausius correlation

2000 < Re < 3000 Re > 3000 Super critical pressure

Fig. 4. Void fraction comparison between COBRA and SACoS.

Fig. 3. 1/8th Fuel assembly solved by COBRA and SACoS.

Fig. 5. Coolant temperature profile along active height.

Table 3 PWR fuel assembly data for comparing COBRA and SACoS. Parameter

Unit

Value

Assembly side Vertical length Fuel rod array Number of fuel rods Number of water rods Rod diameter

m m

0.215 4.267 17  17 264 25 0.95  103

m

Fluid Dynamics (CFD) systems (Seker et al., 2007). In this paper, we will present the coupled MCNP and SACoS system used for design calculations of HPLWR. An external coupling is established between MCNP4c and SACoS code. A master module controls the coupling and transfer of relevant data between neutronics and thermal hydraulics codes. Small modules perform the duty of data transfer in successive iterations. The coupling procedure carries on until a reasonably converged power profile is achieved. An under-relaxation factor of 0.2 is applied to dampen out the fluctuations and get a converged profile more quickly. A flow chart depicting the coupling process is shown in Fig. 7.

except that a fully square assembly is used rather than using one with round corners. The main design parameters of the assembly are given in Table 4. Cross sectional view of the HPLWR fuel assembly being used for calculations is shown in Fig. 8a. Due to the symmetry present, only 1/8th fuel assembly is modeled in MCNP4c and SACoS. Fig. 8b shows the 1/8th assembly modeled in MCNP. The boundary conditions used for thermal hydraulic calculations in SACoS code are described in Table 5 and boundary conditions used for neutronics calculations in MCNP4c are given in Table 6. Along with these boundary conditions, the coupled calculation is run for 10,000 particles and 700 cycles from which first 50 cycles are discarded. Cross section data for higher temperatures was used from ENDF-B VII library. For simulation of fuel temperatures between the available libraries, mixing technique (Bernnat et al., 2000) was used. Mixing technique provides a mean for a linear interpolation for the cross section of fuel to simulate the Doppler broadening effect in fuel. The active length (4.20 m) of fuel is divided into 21 equal parts. Fuel and coolant/moderator properties like density and temperature are considered constant over one axial part during one iteration.

3.1. Reference system

4. Results and discussion

The HPLWR assembly used for testing the performance of coupled system is the same as designed by Hofmeister et al. (2007)

Using the coupled system of MCNP4c and SACoS, calculations were performed for aforementioned HPLWR reference system.

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45 Table 4 Geometric parameters for HPLWR assembly (Hofmeister et al., 2007).

Fig. 6. Centerline fuel temperature along the active fuel height.

Start

Run MCNP

Linear Power Density

Run SACoS

Coolant/Moderator Density & Temperature, Fuel Temperature

Power Convergence ?

No

Yes Stop Fig. 7. Flowchart describing the coupling procedure.

The basic parameter to look out for was k-effective value for the system. The results of k-effective value are shown in Table 7. The difference in k-effective values can be explained as difference of cross section libraries being used. For the reference work (Waata, 2006), JEFF 2.2 library was used for the calculation while we used ENDF-B VII library. The second reason can be the absence of inactive part of 255 mm on top and bottom of active fuel length in the reference case which was not modeled in our calculation. Also the author has mentioned that all results are preliminary and they include an error of 7%. The value of multiplication factor in our calculation is showing the correct trend i.e. for partially coupled system the value is less than that of uncoupled system which is due to reduction in density of water. The value used for uncoupled case is 0.6 g/cm3 for coolant and moderator as indicated in Table 6. In case of partially coupled system, the density of coolant from converged solution comes out to be 0.3 g/cm3. In case of fully

Parameter

Unit

Value

Diameter of fuel pellet Inner cladding diameter Outer cladding diameter Cladding thickness Active height Pitch/Diameter (P/D) Gap between fuel rod and box wall 1/2 gap around one fuel assembly

mm mm mm mm mm mm mm

6.9 7 8.0 0.5 4200 1.15 1.0 5.0

Fuel assembly box Inner Side Length Wall thickness Outer side length

mm mm mm

65.2 1 67.2

Moderator box Outer side length Wall thickness Inner side length

mm mm mm

26.8 0.3 26.2

coupled system, the value of multiplication factor is greater than that of partially coupled system but less than that of uncoupled system. This is due to smaller fuel temperature as compared to uncoupled or partially coupled system. For uncoupled and partially coupled system, the fuel temperature used is 1500 K. For fully coupled system, the average fuel temperature value comes out to be 1000 K which explains the increase in multiplication factor. This also indicates that the Doppler coefficient of reactivity is negative. The variation of linear power density averaged over fuel rods along the active fuel height is shown in Fig. 9. This figure also lays emphasis on the use of coupled analyses for a supercritical water reactor. The linear power density varies drastically for uncoupled and coupled cases. The results for linear power density for uncoupled, partially coupled and fully coupled cases are compared with the reference values in Figs. 10–12. Fig. 10 shows the comparison of linear power density for uncoupled case for calculated and reference values. Both curves match quite well. The little difference can be explained as the absence of top and bottom reflector in our calculations. The inclusion of feedback effect from coolant/moderator temperature and density changes the shape of linear power density quite dramatically. Fig. 11 depicts the linear power density comparison of coupled case without Doppler feedback effect. The difference in values is due to absence of top and bottom reflector. Same reason for similar results has been quoted by Reiss et al. (2008). The reason for two peaks is the counter current flow of coolant and moderator i.e. higher moderator density gives a peak at the top of fuel rod and higher coolant density gives a peak near the bottom of fuel rod. The result which shows the most deviation from reference case is shown in Fig. 12. When we include the Doppler feedback effect in our calculations, we seem to get a result which show quite flat profile as compared to the reference case. In our case, we have used the average temperature for the fuel pellet as fuel pellet is modeled as a single lump. The formula used to calculate the average temperature of the fuel is Tavg = 4/9Tcenterline + 5/9Tsurface (Rowlands, 1962). This formula has been shown to work very well in case of both low and high power cases (Greifenkamp et al., 2008). The variation of fuel temperature (averaged over the seven rods) along the active height of fuel rod is presented in Fig. 13. This figure clearly shows that the maximum temperature along the height of fuel is close to 1100 K whereas, in the uncoupled case or coupled case without Doppler feedback effect, the constant temperature used is 1500 K. Hence due to negative nature of Doppler feedback effect i.e. higher temperature introduces negative reactivity and lower

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

Fig. 8. (a) Reference fuel assembly (Waata, 2006). (b) 1/8th Assembly modeled in MCNP.

Table 5 Boundary conditions for SACOS code (1/8th assembly) (Waata, 2006). Parameters

Unit

Value

Total power of the 1/8th assembly Inlet pressure of the moderator channels Inlet temperature of the moderator channels Total coolant mass flow rate Mass flow rate in the moderator tube

kW MPa °C

327.5 25 280

kg/s kg/s

Mass flow rate for inter assembly gap

kg/s

0.167 0.0278 (16.65% of total flow) 0.0139 (8.32% of total flow)

Table 6 Boundary conditions used in MCNP. Parameters

Density (kg/m3)

Temperature of the cross section library (K)

Fuel – UO2 (5% enrichment and 4% in the corner rod) Cladding-Alloy 316 Moderator Coolant

10600

1500

7450 769 769

800 600 600

Fig. 9. Linear power density variation along active fuel height.

Table 7 Comparison of k-effective values.

Reference case Calculated

Uncoupled

Coupled without Doppler feedback

Coupled with Doppler feedback

1.16619 ± 0.00022

1.16365 ± 0.00022

1.17112 ± 0.00023

1.12019 ± 0.00022

1.07043 ± 0.00022

1.09101 ± 0.00023

temperature introduces positive reactivity, we are having a higher value of linear power density due to smaller temperature values. Our calculations show a much flatter fuel temperature along active height leading to a much flatter linear power density profile. Fig. 14 gives the comparison of fuel temperature for corner fuel rod between calculated and reference values. Reference value for fuel temperature is showing two peaks which is the reason for two peaks in the power profile and hence supports our conclusion. Monti (2009) has used the same formula to calculate the average fuel temperature and has shown that the variation of power along active height for an HPLWR exhibits a similar trend as our calculations near the center of the core.

The average density variation for coolant and moderator along active height is shown in Fig. 15. Coolant density is averaged over the fuel rods and moderator density is averaged over water box and inter-assembly gap value. As shown in Table 5, 25% of total flow at the inlet of pressure vessel goes through moderator channels (water boxes and gap between assemblies) in the downward direction. At the lower plenum, it mixes with the rest of 75% of total flow and then travels upward acting as coolant. The mixing with 75% water, which is still at inlet conditions i.e. 280 Celsius at the lower plenum, tends to increase the density of moderator a little bit. This increase in moderator density is visible in Fig. 15. The density profile is for the fully coupled system i.e. including Doppler feedback effect. A similar profile is exhibited in the case of partially coupled system i.e. without Doppler feedback effect which gives rise to two peaks as shown in Fig. 10. Fuel centerline temperature distribution for the seven fuel rods is shown in Fig. 16 for the converged solution. We can see that fuel rod 1 showing less temperature as compared to other rods. This is due to the fact that fuel rod 1 is a corner rod and so we are using an enrichment of 4% for that to avoid hotspot factor. The maximum temperature reached is 1440 K in fuel rod 6 which is well within safe operating limits of UO2 fuel. Coolant temperature distribution in sub-channels is shown in Fig. 17. We can see that, similar to the profile for fuel centerline

K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

43

Fig. 13. Variation of linear power density and fuel temperature along active height. Fig. 10. Linear power density comparison for uncoupled case.

Fig. 14. Comparison of fuel temperature used as feedback for corner fuel rod. Fig. 11. Linear power density comparison of coupled case without Doppler feedback.

Fig. 15. Average density profile for coolant and moderator. Fig. 12. Linear power density comparison of coupled case with Doppler feedback.

temperature, sub-channel 1 shows the minimum value for temperature among all the sub-channels. The reason is 4% enrichment for corner fuel rod. The maximum coolant temperature is 778 K for sub-channel 8 at exit.

Clad surface temperature value is a very important design and safety criteria which must be satisfied in order to avoid the phenomenon of Heat Transfer Deterioration (HTD) in supercritical water systems. For nickel based alloys, the value for Maximum Clad Surface Temperature (MCST) must be limited below 650 Celsius. Fig. 18 shows the average clad surface temperature for fuel

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K.S. Chaudri et al. / Annals of Nuclear Energy 45 (2012) 37–45

showing a slightly higher value as compared to partially coupled system. This is due to higher value of linear power density in the upper part of fuel rod shown by fully coupled system. However, in both the cases, the MCST criterion is satisfied. 5. Conclusions

Fig. 16. Fuel centerline temperature for fuel rods.

The aim of the study was to develop a coupled analyses tool for SCWR. To do that, we need a thermal hydraulics code capable of calculating thermal hydraulic parameters for supercritical water systems. A sub-channel analyses code called Sub-channel Analysis Code of SCWR (SACoS) has been developed and verified. Using SACoS and MCNP4c, a coupled system for calculation have been developed. To check the validity of our coupled system, the results are compared with that of HPLWR calculations. The results show quite good agreement. The striking difference between the results was seen in the case of coupled system with Doppler feedback effect. The reason for the difference is use of different fuel temperature as feedback in calculation. The results obtained lay emphasis on use of correct simulation parameters as variation of parameters can lead to large difference in the simulation results obtained. On one hand, overestimation can lead to a safer system but can hurt the prospects of cheap power. On the other hand, underestimation can increase the danger of nuclear catastrophe. So a true picture of the actual phenomena is quite necessary to built a safe and yet economically feasible system. Acknowledgments I would like to thank Prof. Liangzhi Cao (Xi’an Jiaotong University), Dr. T. Reiss (Budapest University of Technology and Economics) and my friend Mr. Zeeshan Anjum for their guidance and encouragement during this work. References

Fig. 17. Coolant temperature distribution in sub-channels.

Fig. 18. Average clad surface temperature comparison for fully and partially coupled systems.

rods in case of fully coupled and partially coupled systems. Due to more uniform axial power distribution, the value of clad surface temperature will rise more uniformly for fully coupled system as compared to the partially coupled system. Near the top of the fuel rod, the average clad temperature for fully coupled system is

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