Fuel Thermo-Mechanical Performance during Transient Events in Laguna Verde Nuclear Power Plant with

International Journal of Nuclear Energy Science and Engineering Volume 4 Issue 3, September 2014 doi: 10.14355/ijnese.2014.0403.01

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Fuel Thermo-Mechanical Performance during Transient Events in Laguna Verde Nuclear Power Plant with FETMA Code Hector Hernandez-Lopez*1 Nuclear Systems Department, National Institute for Nuclear Research, Carretera Mexico-Toluca s/n, La marquesa, Ocoyocac, Mexico, Mexico hector.hernandez@inin.gob.mx

*1

Abstract In the National Institute for Nuclear Research of Mexico, the Fuel Management System (FMS) has been used for a long time to simulate the operation a boiling water reactor of nuclear power plant at steady state as well as transient events. To evaluate the thermo-mechanical performance of fuel rod during transient events, changes were made that have been developed and implemented within the Institute to Fuel Element Thermo-Mechanical Analysis code (FETMA). The results of fuel rods thermo-mechanical behavior in hot channel for simulation of transient events of a boiling water reactor are shown. The transient events considered for this job are a load rejection and failure of feedwater control, which are some of the major events that can occur in a BWR. The results showed that the conditions that lead to failure of the fuel rod at no time appeared for both events. Furthermore, we show that the transient load rejection is more demanding in terms of safety than the failure of feedwater controller.

Keywords Fuel Rod; Transient Events; BWR Reactor

Introduction In the National Nuclear Research Institute (Instituto Nacional de Investigaciones Nucleares, ININ) in Mexico, a thermo-mechanical analysis code of fuel elements has been developed for some time. The code is called Fuel Element Thermo-Mechanical Analysis (FETMA) (2012, Hernandez), in which temperature profiles can be calculated in the axial and radial directions as well as the mechanical performance of the materials constituting the fuel rods. FETMA is currently used to study the behavior of fuel elements loaded in both boiling water reactors (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) by the ININ. Several codes have been developed by research and industry laboratories. However, most codes have been

developed to fuel pressurized water reactors. While boiling water reactors only have the following: FRAPCON-2 code developed by Department of Energy of United States of America (DOE) (1981, Berna, et al), the FEMAXI-VI code system (2006, Suzuki) and RODBURN (1993, Uchida & Saito) are a development for the Atomic Energy Agency of Japan (JAEA), the Nuclear Energy Agency (NEA) is developing the TRANSURANIUM code (1992, Lassmann) (2011, Lassmann, et al) and in the case of each manufacturer, they have developed their own tools and their use is restricted.

Of the three codes mentioned above, ININ only has access to FRAPCON and FEMAXI-VI & RODBURN codes; these codes do not have the ability to perform analysis of the behavior during the evolution of transient events. For this reason it requires tools which they can verify independent assessments with, and confirm what is indicated by the manufacturer. Thus, it is necessary to have a code to evaluate the performance of the fuel rods with.

For several years of Scandpower, ININ has also used fuel management system codes (FMS) for the simulation of BWR reactors, especially LVNPP reactors. The FMS is composed of three main codes: HELIOS (1991, Casal, et al), for the generation of nuclear databanks; CM-PRESTO (1996, OĂąa) which is used for 3D simulation of the reactor core in steady state, and RAMONA (1995, Ferri) which simulates the behavior of the reactor core during transient events.

For this work we have developed and implemented some changes FETMA in order to perform thermomechanical analysis of fuel elements during the evolution of transient events, from the data obtained from simulation of the event with RAMONA code.

In a previous work, we have shown the results of

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International Journal of Nuclear Energy Science and Engineering Volume 4 Issue 3, September 2014

thermo-mechanical analysis of fuel elements for normal operation in steady state for LVNPP using FETMA (2012, Hernandez). However, to have a more complete fuel performance, it is necessary to make some studies of transient events that could occur during the operation of the Nuclear Power Plant. In this document the fuel rods behavior during transient events within hot channel is evaluated. Transient events analyzed are load rejections and feedwater controller failure; these events are considered among the most important to the safety of a BWR reactor.

Code Description FETMA code is a computer program which is divided into three modules: a) Neutronic, b) Thermal and c) Mechanical. The thermo-mechanical behavior code evaluates the fuel rods under normal operating conditions with the reactor in steady-state. Currently, changes are being made to the code in order to be employed in the analysis during the evolution of transient events.

In the neutronic module, the code solves the diffusion equation with six energy groups in order to have a better estimate with regards to neutron flux energies above 0.821 MeV and below 0.625 eV which are responsible for the embrittlement of the cladding and thermal power, respectively. The four remaining energy groups correspond to the resonances of both 239Pu and 238U. In the case of analysis during transient events this module is optional and the fast neutron flux calculated by the RAMONA code is taken into account. In this module, the kinetic point for the next versions will be incorporated. On thermal module, from the calculated power distribution on the neutronic module radial and axial distributions of temperature, thermal stresses both in the fuel and cladding are calculated, as well as other phenomena associated with the temperature distribution. In the mechanical module, the strain caused by thermal stress and by the pressure changes caused by the amount of fission gases released from the fuel to the gap are determined, as well as the pressure from fuel to cladding, as consequence of the swelling in the pellet, and the strain cladding produced by the neutron flux with energy greater than 0821 MeV. Methodology Firstly, it is necessary to perform the simulation of reactor operating at steady state, from the beginning of the cycle to the point where transient events begin

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with the CM-PRESTO code, using actual operating conditions of a nuclear power plant, from data obtained for average power, burned and thermal neutron flux of each fuel element.

Secondly, the RAMONA code is used to perform the simulation of transient events, using the obtained data as initial data in steady-state simulation. This will determine the hot channel for the event. The data obtained and subsequently used by FETMA code, is the moderator temperature, the heat transfer coefficient from the moderator and the average heat flux that is obtained in the hot channel.

When the nuclear databank is generated with HELIOS, the power factors are also obtained per rod in the lattices that make up the fuel assembly; thus, it is determined which rod is the hottest. With the above parameters, the input data is constituted with regards to the FETMA code; thus obtaining the radial temperature distribution in the rod, as well as stress and strain distributions in both the pellet and cladding.

Procedure Steady State The nuclear power plant has two reactors at General Electric boiling water reactor (BWR-5) MARK II reactor core is composed of 444 fuel assemblies and 190 control rods, both units have a nominal power of 1931 MWt, electric power 644 MWe, coolant flow 7749 kg/s. Currently both reactors operate with cycles of 18 months and a plant factor of 90%.

FIGURE 1. BEHAVIOR REACTOR CORE BEFORE TRANSIENT EVENTS.

Figure 1, shows the behavior of the thermal power and coolant flow during the steady stage prior to the simulation of transient events. Operation during this cycle was considered an increase in a rated power of

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105%, without altering the flow of coolant.

Load Rejection

While starting the reactor transient events, it exhibited a power of 99.60%, a flow rate of 97.9%, and an inlet pressure of 99.88% in relation to the values of the power increase. Values for reactor operating in the starting transient events are shown in Table I.

The transient occurred because of a fast closing of the control valves of the turbine, caused by a charge drop in the generator originated by electrical disturbances in the network to which the power plant was connected.

Figure 2 shows the maximum values (1/4 core) burn up level, thermal neutron flux once the hot channel has determined at the start of the transient events, while figure 3 shows the initial conditions of power and flow of hot channel.

The fast closing of the control valves of the turbine, besides activating the reactor SCRAM, tripping the recirculation pumps and opening of the bypass valves signals, causes a reduction of the steam volume creating an increase in the pressure in the steam lines, generating a compression wave that propagates throughout the steam lines until reaching the reactor core where the voids in the core are collapsed.

TABLE I. NOMINAL PARAMETERS FOR OPERATING THE REACTOR LVNPP WHEN STARTING TRANSIENT EVENTS

Parameter Thermal Power [MW] Coolant Flow [Kg/s] Inlet Pressure [Bar]

Nominal

Uprate

Operation

1931

2027

2019

7749

7749

7586

70.0

70.0

69.92

The reduction in the void fraction causes an insertion of positive reactivity to the core; the reactivity increase causes a power rise, which again leads to a new increase of the pressure until reaching a set point, which opens the relief valves, and thus reducing the pressure to the setting point at which the relief valves close again. Depending on the initial condition, the process of opening and closing these valves will repeat periodically until the pressure of the reactor is alleviated and it reaches a stable condition.

Feedwater Control Failure

FIGURE 2. BURNING MEAN VALUES AND RELATIVE POWER IN THE CORE AT THE START OF TRANSIENT EVENTS.

A feedwater control failure results in an increase of the pumps speed and a consequent increase in the reactor water level. To compensate this increase, the other pump is decelerated at minimum level of flow, and the failed pump is not switched to manual control, in order to control the reactor level. To take the reactor into a safe operational zone, the recirculation flow was varied and movement in the control rods pattern was performed, to avoid a SCRAM. Results From the simulation results with the code RAMONA transient events it can be observed, in both events, an increase in the reactor power, caused by the positive reactivity insertion during the development of transient events is higher in the case of load rejection.

Figure 4, shows the behavior of maximum relative power of fuel rod with higher heat flow in the hot channel during the 90 sec. taken into account for transient events.

FIGURE 3. THE INITIAL CONDITIONS IN THE HOT CHANNEL

Due to the increased power by positive reactivity insertion, it results in an increase in the average maximum temperature that reaches the hot channel,

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which is shown in Figure 5; also observing that the same behavior occurs for power, reaching once again the maximum during the event of load rejection.

Fuel rod temperature profile determined by FETMA. Figure 6 shows the behavior of maximum temperature for the events analyzed, observing that in the case of load rejection if the maximum temperature is 2300°C, whereas the failure to control feedwater temperature maximum is 2200°C. Both values are still far from the value of the melting temperature of UO2.

FIGURE 4. RELATIVE POWER IN HOT CHANNEL FOR BOTH TRANSIENT EVENTS.

FIGURE 5. AVERAGE MAXIMUM TEMPERATURE IS REACHED INSIDE THE HOT CHANNEL DURING TRANSIENT EVENTS.

FIGURE 6. TEMPERATURE OF THE FUEL ROD IN HOT CHANNEL EVALUATED IN BOTH TRANSIENT EVENTS.

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Furthermore, the linear heat generation in the fuel rod for both events tested is shown in Figure 7. In both cases, the ratio of linear heat generation is about 410 W/cm, which is set as a value that could lead to failure of the fuel rod.

Figures 8 and 9 show both radial strains at the inner surface and outer cladding respectively stressing the strains that are less than 1%, so it can be said that no failure occurs in the fuel.

FIGURE 7. LINEAR HEAT GENERATION RATE IN THE FUEL ROD DURING TRANSIENT EVENTS.

FIGURE 8. MAXIMUM RADIAL STRAIN ON THE INNER SURFACE OF THE CLADDING, DURING THE EVOLUTION OF TRANSIENT EVENTS.

FIGURE 9. MAXIMUM RADIAL STRAIN ON THE OUTER SURFACE OF THE CLADDING, DURING THE EVOLUTION OF TRANSIENT EVENTS.

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Conclusions

for the Calculation of Steady State Thermal-Mechanical

According to the results obtained with FETMA code, based on the simulation of transient events with the RAMONA code, indications are not observed to determine the fuel precursor fault, since in no case the established values are reached as limits. The results shown for RAMONA should be considered; these values correspond to the average values of the entire assembly while FETMA represents the values of a single rod.

Behavior of Oxide Fuel Rods," NUREG/CR-1845 R3.

Unfortunately, a comparison parameter is nonexistent with other experimental values as mentioned above; access to codes is not possible with the ability to evaluate the thermo-mechanical behavior during transient events. However, this is a first iteration in the development of our own analysis tool.

REFERENCES

Casal J J, et al, (1991) HELIOS: Geometric capabilities of a new fuel assembly program, ITMAMCRP Pittsburgh, Pennsylvania, April28-May 2, 1991, 2, 10.2.1 1-13.

Ferri A., (1995) RAMONA-3 User Manual, Scandpower.

G.A. Berna, M.P. Bohn, W.N. Rausch, R.E. Williford, and

Hernandez-Lopez H.; (2012)FETMA: A Simple Code for Thermo-Mechanical Analysis on BWR Fuel Rods; IJNESE Vol. 2 No. 3 2012; www.ijnese.org, World Academic Publishing.

Lassmann K (1992). TRANSURANUS: a fuel rod analysis code ready for use, Journal of Nuclear Materials, Vol. 188, pp. 295-302.

Lassmann K., Schubert A., Van Uffelen P., Győri C., van de Laar J. (2011). TRANSURANUS Handbook, Copyright ©1975-2011,

Institute

for

Transuranium

Elements,

Karlsruhe, Germany.

Masaaki Uchida and Hiroaki Saito, (1993), “RODBURN: A Code for Calculating Power Distribution in Fuel Rods,” JAERI-M 93-108.

Motoe SUZUKI and Hiroaki SAITOU: (2006), Light Water Reactor Fuel Analysis Code FEMAXI-6 (Ver.1) -Detailed Structure and User's Manual-, JAEA-Data/Code 2005003.

Oña C, (1996) CM-PRESTO User Manual, Scandpower.

D.D. Lanning, (1981), "FRAPCON-2: A Computer Code

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