A PWR HOT-ROD MODEL: RELAP5/MOD3.2.2Y AS A SUBCHANNEL CODE I.C. KIRSTEN (1), G.R. KIMBER (2), R. PAGE (3), J.R. JONES (1) ABSTRACT

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1 FR lh International Conference on Nuclear Engineering, ICONE April 2001, Nice, France A PWR HOT-ROD MODEL: RELAP5/MOD3.2.2Y AS A SUBCHANNEL CODE I.C. KIRSTEN (1), G.R. KIMBER (2), R. PAGE (3), J.R. JONES (1) (1) British Energy, Barnwood, Gloucester, GL4 3RS, UK (2) AEA Technology, Winfrith, Dorset, DT2 8EZ, UK (3) NNC, Booths Hall, Knutsford, Cheshire, WA16 8QZ, UK Key words: Subchannel Thermal-Hydraulics, RELAP5/MOD3.2.2Y ABSTRACT The use of the PWR transient analysis code RELAP5 for detailed assessment of Departure from Nucleate Boiling (DNB) has previously implied coupling it in some way to a subchannel code, either by direct code-to-code coupling or by transferring core boundary conditions to the subchannel code. This paper shows an alternative by using a group of subchannels modelled in RELAP5 to represent a hot rod. The model consists of three parallel channels, each more refined than its neighbour: The first channel represents a quadrant of the core; the second a quadrant on a fuel assembly and the final channel represents a passage adjacent to a single fuel pin. The model is intended for use as part of point kinetics assessments and each channel is assigned a radial form factor designed to conservatively represent the hottest fuel pins in the reactor core. The main outputs from the model are minimum Departure from Nucleate Boiling Ratio (DNBR) and clad oxidation for the hot rod (lead pin). The DNBR results from the hot-rod model are benchmarked against the subchannel code COBRA 3-CP and the results are presented in this paper. Some of the modelling problems that needed to be resolved are also highlighted. For transient calculations, several hot-rod models have been "embedded" into the core of a RELAP5 plant deck, each having a different radial form factor. Postprocessing of the output then allows statistical analysis and an estimate of a wholecore census of the dependent variables (e.g. DNBR and clad oxidation) to be made if required. The time penalty of running RELAP5 with several hot rods has been quantified. It is concluded that this RELAP model provides an efficient way of assessing minimum DNBR and clad oxidation for PWR fault transient calculations. 3 3/11

2 OGOO INTRODUCTION Safety cases for PWRs often rely on predicting the minimum margin to the Critical Heat Flux (CHF) for the hottest fuel rod in the core in fault transients. A reactor-system modelling code is usually used for predicting the core boundary conditions in such a fault transient. These boundary conditions may then be used in a subchannel code to model the hot assembly and predict the minimum margin to the safety limit (expressed as a DNB Ratio). That approach necessitates coupling between the reactor systems code (here RELAP5) and a subchannel code. The approach outlined in this paper is to introduce a model of the hot rod directly into RELAP5 with the advantage of having to use only one code. METHODOLOGY Thermal Hydraulic Model RELAP5 solves one-dimensional two-phase flow and heat transfer in an arbitrary network of channels connected by user-defined junctions. It is not designed for the modelling of 3D flow distributions because of simplifications in its solution of the momentum equation. However, the effect of these simplifications can be overcome by careful choice of nodalisation as discussed later. The following channels were added to a standard RELAP5 four-quadrant model of the core to represent a hot-rod model, and each channel is connected to its neighbour at all axial levels to permit cross flow, as shown in Figure 1.: a 'Viewfinder' channel representing approximately a quarter of the hot assembly; a 'Typical' subchannel (representing the passage between 4 fuel rods); a 'Thimble' subchannel (surrounded by 3 fuel rods and 1 thimble tube). Typical channel: F AH =1-65 Viewfinder channel: F AH =1.5 Thimble channel: FAH=1-65 Figure 1: Subchannel layout in the reactor core part of the RELAP5 model. Dashed lines signify crossflow connections at all axial levels. The 4 th quadrant is adjusted (reduced) in power and flow area to account for the additional channels. In the axial direction the model consists of 24 nodes. This axial refinement was found to be sufficient to resolve cross-flows in core loadings of more than one fuel type. The model is intended for use in combination with point-kinetics calculations. Consequently the radial form factors for the various hot rods are selected to conservatively represent conditions likely in any real core. Representative values are shown in Figure 1. Moreover, user specified axial rating shapes are employed as

3 oooo appropriate to the analysis. For the purposes of this paper a chopped cosine axial power profile was chosen. However, if RELAP5 is linked to an appropriate reactor physics code, radial and axial form factor data could be supplied to the hot rod for better estimate assessment. The TALINK code provides such a linkage capability (Ref. 1). There is the option to include more than one hot rod model with different form factors for radiological release assessment. The default CHF calculation in RELAP5/MOD3.2.2y uses the 1986 CHF tables of Groeneveld and these were employed for the current work. Safety limits for the Groeneveld tables were derived in COBRA by comparison of predictions against bundle CHF data. Fuel Rod Model The fuel rods that are connected to the typical and thimble channel are modelled with 9 radial nodes (6 in the pellet, 1 in the gap and 2 in the cladding). The RELAP5 dynamic gap conductance model was employed. The data is chosen such that it bounds the fuel temperature predicted by the British Energy fuel performance code ENIGMA as a function of linear pin rating and burnup. The hot-rod model enables a detailed assessment of the lead-pin clad oxidation and fuel temperature for post-dryout conditions. MODEL ASSESSMENT The model results have been compared with steady-state predictions of the subchannel code COBRA 3-CP, which has previously been qualified for licensing calculations in the UK (Ref. 2). The comparison was carried out for a sequence of 32 statepoints, covering a broad range of regimes that could result in DNB in a PWR fault transient, including: high power, low flow, low pressure and high inlet temperature. The conditions that were studied are summarised in Figure 2. In each case, the limiting CHF conditions have been predicted using COBRA 3-CP (minimum DNBR = 1) and the margin to CHF reassessed in RELAP5. The DNBR in RELAP5 therefore represents the difference between the two codes, see Figure 2, and the predictions show no trend with any of the parameters.

4 CHF ratio COBRA: RELAP CHF ratio COBRA : RELAP + I Typical Thimble 0.2 Typical Thimble Flow(%) Pressure (bar) CHF ratio COBRA: RELAP CHF ratio COBRA: RELAP Typical -Thimble * Typical Thimble Power (%) C Figure 2: Range of model assessment. The main thermal-hydraulic parameters of interest are local flow, void fraction, quality, temperature and DNBR in the hot channels. It can be seen from Figure 3 that the minimum DNBR and other thermal-hydraulic parameters are generally consistently predicted. The main feature in both sets of predictions is that the mass flux reduces higher up in the hot channel, due to cross flow caused by the higher power in these channels than their neighbours. Small differences between the predictions of the codes are evident. Partly these are due to the neglect of turbulent mixing between subchannels in RELAP5, but RELAP5 is a two-fluid code, whereas COBRA 3-CP has a mixture model, so that some of the difference is caused by slip. Also there is a difference in the choice of modelling of the grid loss coefficients, which are smeared out in the RELAP5 model, whereas they are modelled at the appropriate axial node in the COBRA 3-CP model. Based on these results and since the COBRA 3-CP model has been qualified against fuel CHF data, it can be concluded that the RELAP5 model can also be used for assessment, given suitable safety analysis limits.

5 oooo ratio n Comparison of RELAP5 and COBRA3-CP (thimble channel) ^P i \ A A RELAP5 COBRA \ l\ l\ J vj V A \. /\ / & 1 Comparison of RELAP5 and COBRA 3-CP (typical channel),... i. i i i i i Distance up fuel rod Comparison of mass fluxes Distance up fuel rod Comparison of mass fluxes ' RELAP5 >-COBRA i. g I/I " s RELAP5 COBRA Comparison of quality and void fraction Comparison of quality and void fraction Figure 3: Comparison of RELAP5 hot-rod model against COBRA 3-CP predictions, for two extreme fault statepoints: a) 50% flow (results for thimble channel), and b) low pressure (results for typical channel). (Note that RELAP5 here sets the DNB ratio to 0 if it is significantly above 4, hence the apparent discrepancy in the short distance up the fuel rod.) CROSSFLOW MODELLING The modelling of the cross flow in RELAP5 has to be exercised carefully. The reason is that the momentum equation in RELAP5 considers only the component of momentum flux normal to the surface of a junction. Cross-flow junctions in RELAP5 essentially represent horizontal pipes joining the flow passages, rather than a gap between two fuel rods. This approximation has two effects: - The flow in the junction has no axial (i.e. vertical) momentum and therefore may become horizontally stratified. This could lead to inappropriate modelling of

6 OGO interphase drag. A relatively simple option to overcome this is to force the cross flow to be homogeneous. The flow through cross-flow junctions is calculated employing the same momentum equation as used for the main flow direction. However, the onedimensional nature of this equation means that the cross-product terms are neglected. The vertical momentum removed from the donor channel, but not transferred to the recipient channel. The coded equation in finite difference form is given as follows: ( a *P*)" K + ' " v l) x i + k <<W "[ K> "-- < V P K] At + VISCOUS TERMS ADDED MASS + MASS TRANSFER MOMENTUM + STRATIFIED PRESSURE GRADIENT EFFECT Conventional notation is used for the fluid physical properties. The term FWG is a wall drag coefficient, FIG the interface drag coefficient, HLOSSG the form (frictional) loss and B the body force. Details of the terms noted in this equation are given in the code manual [ref. 3] The loss of momentum could have a significant effect on the modelling of the recipient channel, but it is acceptable for the model proposed in this paper. The main reasons for this relate to the direction of the cross flow and the dimensions of the channels. The flow areas of adjacent channels each differ by about two orders of magnitude, so that the larger channel is insensitive to conditions in its smaller neighbour. Liquid and vapour flow outwards from both the typical and thimble channels because of their higher power density. The additional axial momentum that would appear, in practice, in the Viewfinder channel is not significant because of the relatively large size of the Viewfinder. Thus the error in the RELAP5 calculation of lateral convection of axial momentum will also be small. In the model, the Viewfinder channel is linked to the rest of the quadrant. The area ratio is again very large and the above argument is equally applicable. The model presented is therefore appropriate for the proposed use. MULTIPLE HOT ROD MODELS

7 oooo Assessment of fault progression can require calculation of either the fraction of the core that exceeds the CHF, or the quantity of hydrogen generated by cladding oxidation. This requires modelling of rods at various radial form factors and has been achieved by employing multiple copies of the hot-rod model. The addition of four hotrod models had the effect of doubling the CPU time for a pump coast-down fault. No convergence problems were encountered due to the addition of the extra channels. CONCLUSIONS It has been demonstrated that RELAP5/MOD3.2.2Y can be used with some confidence for PWR reactor core subchannel analysis under certain limited conditions. No new code development is required to achieve this. Agreement between the model predictions and those of conventional subchannel analysis is good and the code can potentially be qualified for DNB margin assessment. RELAP5 has a tendency to select an inappropriate flow pattern to represent the transverse flow and this must be prevented by the selection of suitable modelling options. The limitations of the RELAP5 momentum model can be overcome by suitable nodalisation. The hot rod model might also be suitable for use when running RELAP5 coupled to a suitable 3D neutronics code (such as PANTHER). REFERENCES 1. R Page & JR Jones, "Development of an Integrated Thermal-Hydraulics Capability Incorporating RELAP5 and PANTHER Neutronics Code", OECD/CSNI workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements, November IC Kirsten et al, "Changing Fuel Vendor: A Utility (British Energy) and a Fuel Vendor (Siemens) Perspective of the Thermal-Hydraulic Implications", NURETH-9 conference, October RELAP5/MOD3 Code Manual Vol. VI. The RELAP5 Code Development Team NUREG/CR June 1995 ACKNOWLEDGEMENTS We would like to acknowledge the many useful discussions held with JP Rippon, PAW Bratby, SC Bubb and MG Woodhill. We would also like to thank M El-Shanawany for his support in placing this project under the HSE/NII nuclear safety research programme.