Relap 5 for Research Reactors

- 1 – Copyright © 2005 by CNS

13th International Conference on Nuclear Engineering Beijing, China, May 16-20, 2005

ICONE13-50124

VALIDATION OF RELAP/SCDAPSIM/MOD3.4 FOR RESEARCH REACTOR APPLICATIONS

DRAFT

Anhar R. Antariksawan National Atomic Energy Agency, Reactor

Safety Technology Research Center, Kawasan Puspiptek Serpong, Tangerang

15310 INDONESIA.

Md. Quamrul. Huda Institute of Nuclear Science & Technology,

Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-

1000, BANGLADESH.

Tiancai Liu China Institute of Atomic Energy, P.O.

Box 275(33), Beijing, CHINA.

Jelena Zmitkova Nuclear Research Institute Rez plc, 250 68

Rez, CZECH REPUBLIC.

*Chris M. Allison* Innovative Systems Software, LLC, 1284

South Woodruff, Idaho Falls, Idaho, 83404, USA.

Phone: 01-208-523-4200, Fax: 01-208-523-4649 [email protected]

Judith K. Hohorst

Innovative Systems Software, LLC, 1284 South Woodruff, Idaho Falls, Idaho,

83404, USA.

Keywords: RELAP5, RELAP/SCDAPSIM, Validation, Research Reactor ABSTRACT

The RELAP/SCDAPSIM/MOD3.4 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD3.4 uses publicly available RELAP/MOD3.3 and SCDAP/RELAP5/MOD3.2 models, developed by the US Nuclear Regulatory Commission, in combination with (a) new models for fission product transport and deposition, fuel assembly behavior, and in-vessel melt retention, (b) advanced programming and numerical techniques, and (c) integrated graphics displays.

The RELAP5 and SCDAP models have been validated for a wide range of accident conditions using a variety of experiments and plant data. However, the validation of the models for research reactor applications has been much more limited. As a result, a group within SDTP, including the Nuclear Research Institute, Rez (Czech Republic), National Nuclear Energy Agency of Indonesia, China Institute of Atomic Energy, and Bangladesh Atomic Energy Research Establishment, has started work to validate the code for a variety of research reactor designs including TRIGA and other unique research reactors.

This paper describes the reactor designs currently being considered in the study, the development and qualification of input models for the facilities, and the transients being analyzed as part of this effort.

1. INTRODUCTION RELAP/SCDAPSIM1,2,3, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed at Innovative Systems Software (ISS) as part of the international SCDAP Development and Training Program (SDTP).4 RELAP/SCDAPSIM uses the publicly available SCDAP/RELAP55 models developed by the US Nuclear Regulatory Commission in combination with proprietary (a) advanced programming and numerical methods, (b) user options, and (c) models developed by ISS and other members of the SDTP. RELAP/SCDAPSIM/MOD3.4 includes (a) new models for fission product transport and deposition, fuel assembly behavior, and in-vessel melt retention, (b) advanced programming and numerical techniques, and (c) integrated graphics displays. Section 2.0 provides a brief description of the capabilities of the code. The RELAP5 and SCDAP models and RELAP/SCDAPSIM have been validated for a wide


range of accident conditions using a variety of experiments and plant data including TMI-26-11. However, the validation of the models for research reactor applications has been much more limited. Examples include References 12 through 14. As a result, a special group within SDTP has been formed to validate the code for a variety of research reactor designs, and make any modeling improvements that might be necessary. The special group members currently include Nuclear Research Institute, Rez (Czech Republic), National Nuclear Energy Agency of Indonesia, China Institute of Atomic Energy, and Bangladesh Atomic Energy Research Establishment. These group members are currently focused on the validation of the code for TRIGA and other unique research reactor designs.

This paper describes the reactor designs currently being considered in the study, the development and qualification of input models for the facilities, and the transients being analyzed as part of this effort. Section 3 briefly describes the different types of reactors. Section 4 provides a general description of the approach recommended for the development and validation of the input models along with representative RELAP5 nodalization diagrams used for each of the reactor designs. Section 5 provides a brief list of the type of transients being considered. Section 6 briefly summarizes the preliminary conclusions from the study thus far.

2. RELAP/SCDAPSIM RELAP/SCDAPSIM/MOD3.4 is designed to

describe the overall reactor coolant system (RCS) thermal hydraulic response and core behavior under normal operating conditions or under design basis or severe accident conditions. The RELAP5 models calculate the overall RCS thermal hydraulic response, control system behavior, reactor kinetics, and the behavior of special reactor system components such as valves and pumps. The SCDAP models calculate the behavior of the core and vessel structures under normal and accident conditions. The SCDAP portion of the code includes user-selectable reactor component models for LWR fuel rods, Ag-In-Cd and B4C control rods, BWR control blade/channel boxes, and general core and vessel structures. The SCDAP portion of the code also includes models to treat the later stages of a severe accident including debris and molten pool formation, debris/vessel interactions, and the structural failure (creep rupture) of vessel structures. The latter models are automatically invoked by the code as the damage in the core and vessel progresses.

3. RESEARCH REACTOR DESIGNS CURRENTLY INCLUDED

The research reactors currently being considered in the study include (a) the LVR-15 reactor located at

the Nuclear Research Institute in Rez, Czech Republic, (b) the CARR reactor being built in Beijing, China by the China Institute of Atomic Energy, and (c) TRIGA reactors located at the Atomic Energy Research Establishment in Dhaka, Bangadesh and National Nuclear Energy Agency in Bandung Indonesia. Other reactor designs, including the MPR (National Nuclear Energy Agency) in Serpong Indonesia, will be added to the study in the near future but are not discussed in this paper.

LVR-15 is a light-water moderated and cooled pool type reactor with a nominal thermal power of 15 MW. The pool operates at atmospheric pressure with an average coolant temperature in the core of 320 K. The reactor design is shown in Figure 3.1 with a schematic of the coolant loop in Figure 3.2.

Figure 3.1. LVR-15 reactor design

The coolant in the reactor core flows downward during normal operation. The heat generated in the core is removed by demineralized water circulating in the closed primary circuit through two heat exchangers. Circulation of water in the primary circuit is maintained by six circulating pumps. Five of them are mutually interchangeable and work as the main circulation pumps. The sixth pump is part of the emergency cooling system. Three pumps maintain flow of purified water in the closed secondary circuit pump. The reactor also has closed high pressure/temperature loops suitable for testing of materials under PWR and BWR conditions. The reactor core is composed of several IRT-2M type fuel assemblies. A single IRT-2M fuel assembly consists of four or three square concentric tubes. The fuel is an Al-U alloy(37% U). The average enrichment is 80% (U-235). The fuel element has a total thickness of 2 mm with 0.4mm of fuel sandwiched between 0.4 mm of Al alloy cladding.


Figure 3.2. LVR-15 primary coolant flow circuit

CARR, China Advanced Research Reactor, is a

tank-in-pool design, cooled and moderated by light water and reflected by heavy water. The rated power is 60 MW. The core consists of 21 plate-type fuel assemblies. The fuel elements consist of U3Si2-ALx dispersion with 6061 aluminum alloy cladding. The reactor has a small core with high heat flux and high coolant velocity. The reactor core and associated structures are submerged in light water in the reactor pool. The approximately 700m3 of water in the reactor pool also acts as a temporary heat sink during some accidents. The reactor coolant system piping is connected to the reactor pool through two opening strainers that are located on the decay tank. In this way, the reactor pool can be used as a pressurizer. The reactor coolant system consists of four pumps, four heat exchangers and related pipes and valves. The ECC, emergency core cooling, pipe is connected to the reactor inlet pipe by a valve. Under normal conditions, the coolant will go through the flow guiding tank, reactor core, decay tank, main pumps, heat exchangers and back to the flow guiding tank. The ECC pumps are put into operation with the main pumps simultaneously. When the pressure in the inlet pipe is lower than the pressure limit, the ECC pump will suck water from the reactor pool. In this case, the pool water will go through the core, the decay tank and back to the pool. If the pressure difference between the inside and the outside of the flapping valve is lower than the limit, the passive flapper valve will open and the natural circulation will be established. There is an inverse flow when the forced circulation turns to the natural circulation. Figure 3.3 shows the reactor system with the expected flow direction under different conditions.

TRIGA-2000, located in Indonesia, is a pool type reactor with 2 MW thermal power. The reactor tank, made of anodized aluminum, is covered by a concrete shielding, 60-90 cm thick. The core is placed at the bottom of the tank with approximately 5 m of water above for vertical shielding. A 28 cm thick graphite ring surrounds the core. Figure 3.4 shows vertical cut-views of the reactor.

Figure 3.3. CARR primary circuit flow

Figure 3.4. Vertical cut-view of TRIGA-2000 design.

The reactor core is composed of an array of solid

fuel rods, arranged in a hexagonal array. The fuel is made of ZrH, mixed homogeneously with the enriched uranium. The U-ZrH fuel and graphite rod end pieces placed at the top and bottom of the fuel are inserted in a 304-stainless steel tube. The tube is 0.5 mm thick and 75.2 cm long.

The primary cooling system is diagrammed in Figure 3.5. The reactor core is cooled by water in either forced or natural circulation, depending upon the conditions. During normal operation, one centrifugal pump is used to feed the water to the reactor tank. The inlet pipe is located at the bottom of the core and the outlet pipe installed closed to the water surface. The inlet pipe (cold leg) discharges cold coolant into the reactor tank and the outlet pipe (hot leg) sucks the hot coolant out from the reactor tank into to the heat exchanger. A plate type heat


exchanger is used to cool the hot coolant by transferring the heat to the secondary cooling system.

Figure 3.5. Primary system of TRIGA-2000. TRIGA-3000, a 3MW TRIGA MARK II research

reactor located near Dhaka, Bangladesh, was commissioned in late 1986. The reactor is a light water cooled, graphite-reflected one, designed for continuous operation at a steady-state power level of 3000 kW (thermal), and for routine pulsing with reactivity insertions of up to 1.4% δk/k ($ 2.00). The reactor and experimental facilities are surrounded by a concrete shield structure similar to that of TRIGA-2000. The reactor core and reflector assembly are located at the bottom of a 2 m diameter aluminum tank 8.2 m deep. Approximately 6.4 m of water above the core provides vertical shielding. The TRIGA core consists of 100 fuel elements arranged in a concentric hexagonal array within the core shroud. The fuel is a solid, homogeneous mixture of Er-U-ZrH alloy containing 20% by weight of uranium enriched to 19.7% 235U and contains a pure zirconium rod of diameter 0.635 cm at the center.

The reactor can be operated at power levels up to 500 kW with natural convection cooling of the core. For higher power, the forced flow mode (downward) of operation is required to transfer the reactor heat to the cooling tower. The primary cooling system is designed to pump 1.32 m3 of de-mineralized water per minute through the reactor core. The heat generated through fission in the fuel material is conducted through the fuel, through the fuel-cladding interface, and through the cladding to the coolant. The cooling system of the TRIGA reactor shown in Figure 3.10 provides the following four water systems: 1. Primary cooling system which transfers the heat

generated in the reactor core to a water-to-water heat exchanger.

2. Secondary cooling system which transfers the reactor core heat from the heat exchanger to a cooling tower.

3. Purification system which maintains the purity of the primary coolant.

4. Emergency core cooling system which supplies water to cool the shutdown reactor in the event of a tank accident.

Fig. 3.10. Cooling system of the TRIGA-3000 MARK II Research reactor.

4. DESCRIPTION OF THE RELAP/SCDAPSIM INPUT MODELS

The RELAP/SCDAPSIM input models include all of the major components of each reactor system including reactor tank, reactor core and associated structures, and reactor cooling system including pumps, valves, and heat exchangers. The secondary sides of the heat exchanger(s) are also modeled where appropriate. 4.1 Development and Qualification of the Input Models

Because each of the research reactors being analyzed have unique features, it has been necessary for each of the participants to develop and qualify their input model independently. However, all of the participants have received RELAP/SCDAPSIM user training for research reactor applications from the training staff at one of the SDTP regional training centers. Although the training received by the participants has varied from two weeks up to a maximum of 3 months, each of the participants received representative research reactor and other input models and performed a variety of transients using the models. In the case where more extensive training was possible, the participants worked together with the training staff to develop preliminary models for their own reactor designs. The development of a preliminary TRIGA-3000 model for


the Bangladesh reactor by one of the authors, Md. Q. Huda, is a good example of the process that was used.

The development of the preliminary input model was started by Dr. Huda under a 3 month training fellowship supported by the International Atomic Energy Agency. Technical support, training, and technical review of the model were performed by the ISS staff at the SDTP regional training center in Idaho. During the training period, the input model was built and tested in stages. Individual parts of the system were built and tested with simple boundary conditions. The physical arrangement was verified using the integrated 3D orthographic displays available in RELAP/SCDAPSIM. Figure 4.1.1 shows an example of the 3D display of the input for the TRIGA-3000 pool and reactor.

Figure 4.1.1. RELAP/SCDAPSIM 3D

orthographic display of TRIGA-3000 pool and core. As seen in the figure, it was possible to verify

that the physical arrangement specified in the input model was consistent with the actual reactor design. The 3D image is drawn to scale by the code based on the reactor physical dimensions and orientations used in the input model. The display can also be interactively rotated and scaled during the simulation to view different features of the model. In addition, the display can show the hydrodynamic volumes and junction numbers from the input model as well as any calculated results. Once the physical arrangement of the system model was verified, the model was then run using time dependent boundary conditions at the entrance and exit of the system to verify that the flow and other expected conditions were adequately reproduced. This approach was used for each of the major features of the reactor system. Then after each part of the system is tested and verified separately, the complete model was then put together as shown in Figure 4.5.1. Once the complete model was assembled, this complete model was then used for comparisons with plant steady state data. This comparison helped verify that the flow resistances and other model input assumptions were set properly. In

addition, where available, reactor startup data or other transient data was used to verify the thermal capacitances of the system, pump coast down characteristics, and other transient characteristics of the input model. During this process the results were reviewed and any flaws in the steady state input models are corrected. Examples of the steady state results from representative calculations are included in the Appendix.

4.2 LVR-15 Input Model

The RELAP/SCDAPSIM nodalization for the primary cooling system is shown in Figure 4.2.1. Reactor core flow is described using 8 parallel channels representing the flow in the core with one additional channel representing the bypass flow in the pool. The loop nodalization includes the two heat exchangers and all six pumps to allow a variety of different transients to be considered. The fuel assemblies and other structures in the reactor system including the reactor tank and piping are described using one-dimensional RELAP5 heat structures.

Figure 4.2.1. RELAP/SCDAPSIM nodalization for LVR-15

4.3 CARR Input Model

Figure 4.3.1 showS the nodalization of the reactor system. All of the fuel assemblies are modeled by hydrodynamic components and heat structures. The reactor coolant system is also modeled according to the actual design. In order to describe the flow in the reactor pool, the pool is separated into two parts and there are multiple junctions between these two parts. 4.4 TRIGA-2000 (Indonesia) Input Model The development of the input model(s) for TRIGA-2000 is still in the early stages of the process. The nodalization used in the initial input model is shown in Figure 4.4.1. In this case, the model only included the primary system. The secondary system is defined as model boundary conditions. The model includes the reactor tank and the primary coolant piping system, including a plate type heat exchanger.


Figure 4.3.1. Nodalization of reactor system for CARR.

The reactor tank model includes several volumes

to represent the different flow regions inside the tank, as shown in Figure 4.4.2. The coolant flow from inlet pipe is distributed into the core and core by-pass flow regions through lower plenum. The flows come together in the mixing volume before going to the outlet pipe. The core is divided into three channels: hot, average and cold channels. The hot channel represents a single channel containing the hottest fuel. The average channel models other 99 fuel rods with average flux and the cold channel represents the rest of the core.

Figure 4.4.1. TRIGA-2000 primary coolant

system nodalization.

4.5 TRIGA-3000 (Bangladesh) Input Model The reactor can be operated in either a forced

downflow mode at power level up to 3000 kW or a natural circulation mode at power levels up to 500 kW. The heat generated in the reactor core is transferred to a secondary water cooling system by a water-to-water heat exchanger. The major components of the forced downflow cooling system are the reactor core and reactor tank, the N16 decay tank, the primary pumps and the heat exchanger, arranged as shown in Fig. 3.10. A RELAP5 model consists of a system of control volumes which are connected by flow junctions and use of 3D display reactor which is shown in Fig. 4.5.1.

Figure 4.4.2. TRIGA-2000 reactor tank and

core nodalization.

Fig. 4.5.1. Nodalization scheme of TRIGA-3000 primary coolant system.

In the core region where the axial distribution is important, pipe components with 15 axial distributed volumes were used for the average and hot channel regions. The LEU feedback coefficients for Doppler, coolant temperature and coolant void/density and other kinetics parameters were chosen to match those determined by General Atomics and used in the PARET code in the earlier study15.

5. TRANSIENTS BEING ANALYZED A variety of transients are being analyzed for the different reactor designs. Many of the results are still too preliminary to discuss in any detail. In some cases, work is still focusing on the validation of the input models using plant steady state data or other operational plant data. However, preliminary transient calculations thus far show significant


margins between the calculated core conditions and the safety limits for the reactors. In those cases where previous safety analysis results from the original reactor vendor are available, the preliminary transient results are consistent with the original vendor results. In the follow subsections, representative transients were selected for discussion and examples of the results are provided. 5.1 TRIGA-2000 (INDONESIA) The input model development is still underway with some work on the steady state validation of the model completed. Several sample transients have been performed. The example selected for this paper is a loss-of-flow transient. In this transient, the primary system flow is assumed begin to decrease at time = 0 s reaching zero at 10 s. Based on the TRIGA-2000 operational limit and conditions, the reactor will be scrammed when the coolant temperature in reactor tank achieves 322 K. The calculation results show that the natural circulation is established after the primary flow stops. As can be seen in Figure 5.1.1, the flow in core by-pass region reverses from an upward flow during steady state condition to a downward flow after the primary flow reduced to zero. The new circulation flow path is from the core to the chimney, mixing volume, core by-pass, lower plenum and back to the core. Due to the loss of the heat sink to the secondary side, the coolant temperature in the reactor tank begins to increase. When the mixing pool temperature reaches 322 K at approximately 200 sec. after the event initiation, the reactor is scrammed. The coolant and fuel temperatures increase only slightly prior to scram and then decrease as the power decays.

Figure 5.1.1. Transient coolant flow rates in different region of reactor tank.

5.2 TRIGA (BANGLADESH)

The development of the input model is still continuing, however, the validation of the model against steady state data and the Safety Analysis Report is well underway. A variety of transients based upon IAEA – TECDOC-64316 including (a)

pulse mode operation, (b) fast and slow loss of flow, (c) reactivity initiated transients, and (d) small and large loss of coolant transients have been performed. Figure 5.2.1 shows an example of the calculated power and reactivity for a pulse mode operation transient. It was observed that pulsing of the reactor by inserting an excess reactivity of $ 2.0 shoots the reactor power level to 865 MW compared to an experimental value of 852MW; the maximum fuel temperature corresponding to this peak power was found to be 505oC.

Figure 5.2.1. Calculated power and reactivity for pulse mode operation in TRIGA-3000.

6. CONCLUSIONS A systematic approach to develop and validate

RELAP/SCDAPSIM/MOD3.4 for research reactor applications is well underway. Participants in the activities have gone through training at the SDTP regional training centers plus continue to receive technical support from the ISS staff and other SDTP members. In addition, several of the participants have received additional training and support from the International Atomic Energy Agency. Although the individual participants are responsible for the development and validation of their input models for research reactor facilities, they have benefited through the exchange of training materials, input models, and more recently the results of their calculations. The benefits of technical exchange on similar TRIGA reactor designs and analysis in Bangladesh and Indonesia are the most obvious, however, participants looking at other reactor designs have been benefited as well. For example, lessons learned about nodalization strategies, the influence of the proper modeling of the fuel-cladding gap, and the type of transients that should be considered have helped all of the participants.

The development of a systematic approach to build and validate the input models has also been an important factor in the activities. As shown by the example for the Bangladesh TRIGA reactor, the RELAP/SCDAPSIM interactive 3D displays are very


helpful to verify that the input model is physically consistent with the reactor system. The development and testing of the individual components of the system model prior to the assembly of the full model also proved to be an important factor in the process. The development of a base TRIGA input model and the analysis of both steady state and a wide range of transients over a 3 month period by an analyst that had not previously used RELAP5 before is a strong testament for such an approach.

The use of plant steady state data, and other plant data or vendor safety analysis where available, has also been an important factor in the development and validation activities. Although many of the input models are still incomplete, the analysts have been able to match the reactor data and vendor safety analysis results quite well. This has helped with the validation of the input model and has given the analysts valuable experience.

Finally the performance of RELAP/SCDAPSIM/MOD3.4 has been important factor in the effort. Even though some of the conditions for the simulations have been very challenging due to the presence of non-condensible gases and sub-atmospheric pressures, the code has run reliably and quickly. Because most of the calculations run significantly faster than real time on inexpensive Windows PCs, the time required to run a large number of transients is no longer of significance

REFERENCES

1. Allison, C. M and Wagner, R. J. Dec. 2004, RELAP/SCDAPSIM/MOD3.4 Supplemental, Supplemental, Innovative Systems Software, LLC.

2. www.relap.com 3. Chavez-Mercado, C. , Hohorst, J. K.,

Allison, C. M., October 4-8, 2004, National Autonomous University Of Mexico RELAP/SCDAPSIM-Based Plant Simulation and Training Applications to the Laguna Verde NPP, The 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6), Nara, Japan.

4. www.sdtp.org 5. July, 1998, SCDAP/RELAP5/MOD3.2

Code Manual, Vol. 1-5, NUREG/CR-6150, INEL-96/0422.

6. December 2001, RELAP5/MOD3.3 Code Manual, Developmental Assessment Problems, Vol. 3, NUREG/CR-5535.

7. July, 1998, SCDAP/RELAP5/MOD3.2 Code Manual, Developmental

Assessment, Vol. 5, NUREG/CR-6150, INEL-96/0422.

8. J. K. Hohorst and C. M. Allison, Lessons Learned from ISP-46 Using RELAP/SCDAPSIM, ICAPP-03, Córdoba, Spain, May 2003.

9. C. M. Allison and J. K. Hohorst, An Assessment of RELAP/SCDAPSIM/MOD3.2 Using Bundle Heating and Melting Experiments with Irradiated Fuel, 11th International Conference on Nuclear Engineering, Tokyo, JAPAN, April 20-23, 2003.

10. E. Honaiser and S. Anghaie, Analysis of RELAP/SCDAPSIM/MOD3.2 Computer Code using QUENCH Experiments, ICAPP-04, Pittsburgh, PA USA, June 13-17, 2004.

11. J. K. Hohorst and C. M. Allison, An Assessment of RELAP/SCDAPSIM/MOD3.4 Using the Phebus FPT-2 Bundle Heating and Melting Experiment, ICAPP-05, Seoul, Korea, May 15-19, 2005.

12. IAEA Regional Workshop on Thermal Hydraulics Safety Analysis from 29 November to 10 December 2004, at the Centre for Research and Development of Advanced Technology in Bandung, Indonesia.

13. Hari, S., Hassan, Y. A., Tu, J., June 2000, Analysis of Transient Events Without Scram in a Research Reactor Using the RELAP5/MOD3.2 Computer Code, Nuclear Technology, Volume 130 · Number 3· pp 296-309.

14. Fletcher, C. D., et. al., “Conceptual Design Station Blackout and Loss-of-Flow Accident Analyses for the Advanced Neutron Source Reactor,” Nucl. Technol., 106, 31 (April 1994).

15. M.Q. Huda, S.I. Bhuiyan, T.K. Chakrobortty, M.M. Sarker and M.A.W. Mondal, “Thermal Hydraulic Analysis of the 3 MW TRIGA MARK II Research Reactor”, Nuclear Technology, 135(1): 51-66, July 2001.

16. “Research Reactor Core Conversion Guidebook-Volume 3: Analytical Verification,” IAEA-TECDOC-643, April 1992.

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Relap 5 for Research Reactors