F ACOLTÀ DI I NGEGNERIA
RELAZIONE PER IL CONSEGUIMENTO DELLA LAUREA SPECIALISTICA IN INGEGNERIA NUCLEARE E
DELLA SICUREZZA INDUSTRIALE
“Studio dell’ incidente di espulsione di barra di controllo in reattori PWR tramite l’utilizzo di codici accoppiati di
Neutronica 3D e Termoidraulica”
MASTER DEGREE THESIS IN NUCLEAR ENGINEERING
“Coupled 3D Neutron-Kinetic and Thermal-Hydraulic codes applied to control rod ejection accident in Pressurized Water Reactors”
RELATORI: CANDIDATO:
Prof. Francesco D’Auria Osvaldo Monteiro de Carvalho Junior
Prof. Roberto Bovalini
Pisa, Dicembre 2005
ABSTRACT
The present thesis deals with a 3D analysis of core behaviour during a fast control rod ejection accident in a Pressurized Water Reactor.
The rod ejection accident (REA) is a design basis accident for a PWR and is characterised by a rapid reactivity insertion together with a power burst and an adverse core power distribution. If this event were to happen, a fuel rod thermal transient which could cause DNB may occur together with limited fuel damage.
In order to study such a particular accident, the coupled 3D Neutron Kinetic and Thermal-Hydraulic (3D-NK-TH) codes technique was implemented to simulate a multi-dimensional core behaviour. Coupled RELAP5/3.3-PARCS and RELAP5/3D-NESTLE codes were used in the development of this work.
The 3D method for TH and NK calculations of a reactivity-initiated accident is at present time worldwide spread and tends to replace the former 2D or 1D evaluations. The motivation for its extensive use is the possibility of gaining margins by predicting a more realistic core behaviour during the accident than the former calculation tools. Recent computer developments, resulting in the availability of powerful computation capabilities at reasonable costs, made it possible to perform detailed dynamic TH system analysis together with coupled 3D-NK core simulation even on a standard commercial PC system.
The main objectives of the present work are the following:
• Study of the coupled codes methodology, identification of its potentialities and fields of application;
• Appraisal of computational tools and acquisition of capability to perform 3D-NK-TH coupling studies;
• Comparison between predictions of different codes in best-estimate transient simulations;
• Verification of the capability of these system codes to analyse complex transients with coupled core- plant interactions;
• Evaluation of safety margins for the reference plant.
The study confirms the consistency between the methods used for 3D transient calculations and their strong capability in predicting core response during a REA in a PWR.
REFERENCE PLANT DESCRIPTION & MODELLING
The NPP that has been chosen for simulating a Rod Ejection Accident (REA) was the Three Mile Island Unit 1, a PWR-type plant, supplied by Babcock & Wilcox, with nominal power of 2772 MW t (850 MW e ), still in operation in Pennsylvania (USA), which was object of an OECD benchmark for the study of a Main Steam Line Break (MSLB) accident.
In order to more accurately predict transient response, different combined 3D neutronic and thermal- hydraulic nodalizations were implemented for each reference case studied and depending on the coupled codes applied to it. In the present work, two different coupled codes were used to model the plant in order to simulate a control rod ejection accident:
• RELAP 5/3.3-PARCS – this coupling requires Parallel Virtual Machine (PVM) to be operated.
• RELAP 5/3D-NESTLE – in this case the neutronic code is embedded in the TH code.
The radial geometry of the core is presented in Fig. 1, in which it is divided into cells 21.811 cm wide, each corresponding to fuel assemblies and reflector assemblies (shaded area). It corresponds to an incomplete 17 x 17 matrix, totalizing 241 assemblies (177 FA + 64 reflector assemblies).
Fig. 1 – Core radial geometry Fig. 2 – one-eight core symmetry with FA
THERMAL-HYDRAULIC MODELLING
The nodalization used for transient calculation was a modification of the previous one developed by the University of Pisa for the TMI-1 MSLB benchmark. The sketch of the nodalization is showed in the Fig. 3. It models the whole primary side (RPV, Hot Legs (2), Cold Legs (4), OTSG (2), Main Circulation Pumps (4) and ECCS) and the secondary side of the plant until the turbine. The nodalization was qualified as far as possible at the University of Pisa.
The main modifications of the core nodalization are here briefly summarized :
• addition of channel S (pipe 179), where rod ejection is supposed to occur, connected to branches 115 (LP) and 127 (UP), as can be seen in Fig. 4;
• creation of heat structure related to the FA where rod ejection is supposed to occur;
• new assignment of the mass flow, flow area, etc. for modified core channels;
• new assignment of the heat exchange surfaces, relative power coefficients, etc. for modified core structures;
• generation of spatial mapping file (MAPTAB) for NK-TH nodes association;
• new assignment of neutronic parameters in NK input deck for PARCS and RELAP5/3D;
• creation of new TH and NK input decks for HZP and HFP for all reference cases studied and different simulations performed, in steady-state and transient calculations.
Fig.3 – TMI-1 Nodalization
The final nodalization described the core through 21 independent thermal-hydraulic channels, 19 describing the FA channels and 2 describing the core by-pass (Vol. 125 and 126) as indicated in Fig. 4 and 5.
115 116 117 118
127 128 129 130
126 125
A G M N S A G M N
A G M N S B C H I O B C H I B C H I O P J D E K P J D E P J D E K A G M N S A G M N A G M N S P J D E K
P J D E
P J D E K A G M N S A G M N A G M N S B C H I O
B C H I B C H I O A G M N S
A G M N A G M N S
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 1
2 14 14 14 14 14
3 15 14 14 14 14 14 14 14 13
4 15 15 9 8 8 8 8 8 7 13 13
5 15 15 15 9 8 8 8 8 8 7 19 13 13
6 15 9 9 9 3 2 2 2 1 7 7 7 13
7 15 15 9 9 3 3 2 2 2 1 1 7 7 13 13 8 15 15 9 9 3 3 3 2 1 1 1 7 7 13 13 9 16 15 9 10 4 3 4 2 6 1 6 12 7 13 18 10 16 16 10 10 4 4 4 5 6 6 6 12 12 18 18 11 16 16 10 10 4 4 5 5 5 6 6 12 12 18 18 12 16 10 10 10 4 5 5 5 6 12 12 12 18 13 16 16 16 10 11 11 11 11 11 12 18 18 18 14 16 16 10 11 11 11 11 11 12 18 18 15 16 17 17 17 17 17 17 17 18
16 17 17 17 17 17
17
Fig. 4 – Detailed nodalization of core channels Fig. 5 – Radial configuration of TH channels
Limitations in the maximum number of junctions belonging to a single branch (RELAP 5 component) imposed the need to split the lower and upper plena in four parts, at least in the zones of connection between the LP and the UP with the core itself. The downcomer has been split in four parts, corresponding to the four Cold Legs of the plant. The coolant flowing in one part does not mix with the others. The UP has been divided in two parts, corresponding to the number of Hot Legs.
3D NEUTRON KINETIC MODELLING
The reactor core was divided in two regions: active core and the reflector region. The model consists in 241 assemblies, among these 177 fuel assemblies and 64 reflector assemblies. Sixty-one of the 177 FA were considered ‘rodded’ FA, in order to model their capacity of hosting a CR cluster.
Fuels assemblies with different U 235 enrichment and varying number of burnable absorber rods are present in the core. Thirty FA types are contained within the core, as indicated in Fig.2. There are 438 unrodded and 195 rodded compositions and each composition is defined by material properties and burn-up. A complete set of diffusion coefficients and macroscopic cross-sections for scattering, absorption, and fission as a function of the moderator density and fuel temperature is defined for each composition. All cross-section data was supplied by the benchmark.
A tough task connected to the NK nodalization is related to the study of the NK codes adopted, in order to comprehend the methodology and parameterization used to model the reactor core in 3D (e.g. node numbering). Understanding the way codes model control rod data was also fundamental in order to enable simulation of the ejection of a CR.
CONTROL ROD WORTH CALCULATION
In order to identify the control rod assembly whose ejection from the core causes the highest insertion of reactivity, the neutronic code PARCS was performed in a stand-alone mode.
Two states of the plant were analyzed in this procedure: the Hot Full Power (HFP) and the Hot Zero Power (HZP), to provide calculation in two different power conditions.
There are 61 Control Assemblies (CA), divided in 7 groups, as indicated in Fig. 6. These rods contain a strong neutron absorber over a length that spans most of the active core region. The total CA length, which coincides with the absorber length, is 342.7055 cm. Measured in units of steps, complete insertion and withdrawal of a CA correspond to 0 and 971 steps, respectively. Each step is 0.3531 cm.
The HZP and HFP with the nuclear fuel at EOC were analyzed, assuming β = 5.2 x 10 -3 . The control rod was always assumed to be ejected in 0.1 seconds from the fully inserted to the completely withdrawn position, in the HZP case. In the HFP condition, at EOC, all the control rod banks are supposed to be withdrawn from the reactor core; therefore a CA ejection was idealized from a 40% to 100% extracted position.
As a result of steady-state calculation, it was observed that CR Bank n° 7 possesses the highest rod worth for both HZP and HFP conditions. After further calculation, among those rods belonging to Bank n° 7, the CA in the position (13,5), as indicated in Fig. 6, was identified as the CR having the highest worth in the core.
Values of the reactivity inserted are presented in Table 1.
1 6 1
3 5 5 3
7 7 7
3 5 4 4 5 3
1 6 2 6 1
5 4 2 2 4 5
6 7 2 7 2 7 6
5 4 2 2 4 5
1 6 2 6 1
3 5 4 4 5 3
7 7 7
3 5 5 3
1 6 1
POWER CONDITION CORE POWER LEVEL
EJECTED ROD WORTH ($)
HZP 1.00E-03 0.74
HFP 1.00E+00 0.19
Fig. 6 – CA configuration for rod worth calculation Table 1 - CA worth for HZP and HFP
DISCUSSION OF ACHIEVED RESULTS
For all the simulations performed, the following procedure was executed:
In RELAP-PARCS coupling, the analyses have been carried out with the following steps:
• Steady-state calculation using RELAP5/3.3 in a stand-alone mode, with 0-D neutron kinetic calculation, in order to verify the correctness of the TH nodalization;
• Null transient calculation using RELAP5/3.3 in a stand-alone, with a duration of 100 seconds. This serves to verify if there is any perturbation or problems in the input deck;
• Steady-state calculation with coupled codes RELAP-PARCS, using the restart file obtained by RELAP5/3.3;
• Transient calculation with PARCS alone, in order to have indication of power excursion and reactivity trend, before TH coupling;
• Transient calculation with coupled codes RELAP-PARCS, using their both restart files generated in the previous steady-state calculation;
• General data processing, using Fortran programs and commercial software.
In RELAP5/3D-NESTLE coupling, the analyses have been carried out with the following steps:
• Steady-state calculation launching RELAP5/3D in order to verify the correctness of the nodalization;
• Transient calculation launching RELAP5/3D, using restart file generated in the previous steady-state calculation;
• General data processing, using Fortran programs and commercial software.
Calculation were performed in order to obtain the following results:
• Reactivity and power excursion during the control rod ejection accident;
• Spatial power distribution (axially and radially) during transient;
• Energy released to the fuel during REA;
• Time trend of the clad and fuel centerline temperatures for the FA most affected by the transient;
• time trend of the main relevant plant parameters.
To perform analyses in different power conditions of the reactor, with a wide range of reactivity insertion, three reference cases were idealized, in which a rod ejection accident was postulated, as indicated in Table 2.
REF. N° REFERENCE CASE CONTROL ROD CONFIGURATION
1 HFP_40% Bank 1-7 ARO - CA# 8 is 40% withdrawn
2 HZP_1rod Bank 1 - 7 ARI - CA# 8 is 100% inserted
3 HZP_2rods Bank 1 & 3 ARI; Bank 2, 4 & 5-7 ARO; Bank 8 is 100% inserted Table 2 - Reference cases studied
STEADY-STATE CALCULATION
Before transient implementation, calculation methods are adjusted based on a series of stationary configurations. Regarding the HFP condition, a synthesis of the main relevant parameters calculated by coupled codes and reference values are presented in Table 3.
PARAMETER DESIGN VALUE RELAP5/3.3-PARCS RELAP5/3D- NESTLE
Total core power output (MWt) 2772.0 2772.0 2772.0
RCS cold leg temperature (°C) 291.0 290.2 290.8
RCS hot leg temperature (°C) 318.0 339.8 320.5
RCS pressure (MPa) 14.96 14.99 14.95
Lower plenum pressure (MPa) 15.36 15.31 15.31
Upper plenum pressure (MPa) 15.17 15.13 15.13
Total RCS flow rate (kg/s) 17602.2 17100.6 16917.7
Feedwater flow per OTSG 761.59 761.59 761.59
OTSG inlet temperature (°C) 299.5 292.5 295.8
OTSG outlet pressure (MPa) 6.41 6.41 6.41
Table 3 - Main steady-state parameters (HFP)
The most important parameter in steady-state calculation is the spatial power distribution. Axial power distributions are presented in Fig. 7 and 8; a confrontation with the reference data provided by the benchmark is also presented. As it can be seen, stationary study results confirm the consistency between the codes and reference data.
Radial power distribution is showed in Fig. 9 and 10, where a three-dimensional view of reactor core is provided. As can be observed from the figures, there is an homogeneous power distribution inside the reactor core.
Fuel centreline and moderator temperature distribution at HFP are presented in Fig. 11 and 12.
Axial power distribution in HZP
0 0.5 1 1.5 2 2.5 3
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 2021 2223 24 25 26
Axial plane number
Relative power
Parcs-Relap Relap3D
Axial Power Distribution - HFP
0 0.2 0.4 0.6 0.8 1 1.2 1.4
0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26
Axial plane number
Relative power
Relap 3D Relap-Parcs Reference
Fig. 7 - Axial power distribution (HZP) Fig. 8 - Axial power distribution (HFP)
X 5
10
15 Y
5
10
15
RelativeAssemblyPower
0 1
1.5 1.4 1.3 1.2 1.1 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 Frame 001 15 Dec 2005 Radial Power Distribution - HZP
X 5
10 15
Y 5 10 15 RelativeAssemblyPower
0 0.5 1
1.3 1.2 1.1 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 Frame 001 26 Nov 2005 Radial Power Distribution - HFP
Fig. 9 - Radial power distribution (HZP) Fig. 10 - Radial power distribution (HFP)
X 5 10 15 Y
5 10
15
Temperature
0 100 200 300 400 500
Temperature 500 450 400 350 300 250 200 150 100 50 Frame 001 28 Nov 2005 Assemblywise Averaged Fuel Centerline Temp.Distribution - HZP
X 5
10 15
Y 5 10 15
Temperature
0 50 100 150 200 250 300
Temperature 300 280 260 240 220 200 180 160 140 120 100 80 60 40 20 Frame 001 28 Nov 2005 Assemblywise Moderator Temp. Distribution - HFP
Fig. 11 – Fuel centerline temperature distribution (HZP) Fig. 12 - Moderator temperature distribution (HFP)
TRANSIENT CALCULATION
Transient calculations were performed for each reference case listed in Table 2. The basic scenario envisaged is supposed to be originated by a fast ejection of one control assembly, while the reactor is operating at EOC. The fuel burn-up rates of the various assemblies differ from 23 to 58 GWd/t, an average radial burn-up rate being considered for each assembly.
Transient duration of 8 seconds was established for all reference cases. The CA ejection duration is 100 ms, and it takes place at t ejection = 1sec, after transient starts.
Regarding processing time and computational resources used, it should be noted that each transient implied a running time of roughly 30 minutes on a notebook equipped with a 1.6 GHz Intel™ processor and a Windows™ XP™ operating system.
HZP – 1 CONTROL ROD EJECTION
In this first reference case, accident is initiated from a critical core at HZP and is characterized by the ejection of the CA n° 8, which has the highest worth value (0.74 $), as calculated previously. Fig. 13 outlines the control bank configuration used for this transient. The rods of the regulation banks (5, 6 and 7) and safety banks (1, 2, 3 and 4) are fully inserted in the core.
1 6 1
3 5 5 3
7 7 8
3 5 4 4 5 3
1 6 2 6 1
5 4 2 2 4 5
6 7 2 7 2 7 6
5 4 2 2 4 5
1 6 2 6 1
3 5 4 4 5 3
7 7 7
3 5 5 3
1 6 1
Fig. 13 - Control rod configuration for HZP CA#8 ejection
Initially, the core power is 0.1 % P N (with P N = 2772 MW), the pressure is 151.68 bar, the average moderator temperature is 551 K, the water density is 0.7695 g/cc, the fuel temperature is 551 K, and the boron mass concentration is 5 ppm.
The transient is characterised by a prompt increase in the total reactivity in the core, as seen in Fig. 14.
The power excursion follows the trend of the reactivity inserted, resulting in a prompt increase of approximately 4 times the initial power; then, power continues to increase till the new power condition (11 % of P N ) be reached (Fig.18).
This behaviour confirms that the reactor in object, in the history condition evaluated at HZP, reacts very well to the ejection of the most reactive CA. Increase in fuel or clad temperatures do not present any risk concerning reactor safety. Calculations performed with coupled codes RELAP-PARCS and RELAP5/3D-Nestle are presented in Fig.18. They show a good agreement, despite power increase in the latter occurs more slowly.
However, both tend asymptotically to same new power condition, after feedback reactivity compensation.
The different contributions to the reactivity versus time obtained by PARCS during the transient are plotted in Fig.14. The global reactivity, which grows for 100 ms until reaching the maximum inserted reactivity (0.74 $), start falling at around 2 s, when the Doppler feedback becomes significant. This decrease is accelerated in the following seconds by the moderator effect.
The maximum power, after the rod ejection, is reached in the assembly (12,5), adjacent to the one which contains the ejected rod (13,5), as indicated in Fig. 16. This can also be seen in Fig. 17, where a view from the top of the 3D surface generated is presented.
Calculated energy released to the fuel during transient is showed in Fig. 20 and 21.
Reactivity excursion - HZP rod ejection
-0.4 -0.2 0 0.2 0.4 0.6 0.8 1
0 1 2 3 4 5 6 7 8 9
Time (s)
Reactivity ($)
sumrho tfrho dmrho crrho
Power excursion - HZP rod ejection
0.0%
0.1%
0.2%
0.3%
0.4%
0.5%
1 1.05 1.1 1.15 1.2 1.25 1.3
Time (s)
Power (%)
Power Level
Fig. 14 - Different contributes in reactivity excursion calculated with PARCS during REA in HZP
Fig. 15 - Power excursion calculated with RELAP5 during REA in HZP
X 5
10 15
Y 5 10 15
Power
0 1 2 3
Power 3.4 3.2 3 2.8 2.6 2.4 2.2 2 1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 Frame 001 28 Nov 2005 Power Excursion for rod ejection at HZP
X
Y
5 10 15
5 10
15
Power3.4 3.2 3 2.8 2.6 2.4 2.2 2 1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 Frame 001 28 Nov 2005 Power Excursion for rod ejection at HZP
Fig. 16 - Radial power distribution at t = 1.51 s during REA (HZP) Fig. 17 - Radial power distribution during REA (HZP) – Top view
Power excursion - HZP rod ejection
0.0E+00 5.0E+07 1.0E+08 1.5E+08 2.0E+08 2.5E+08 3.0E+08 3.5E+08
0.0 10.0 20.0 30.0 40.0 50.0 60.0
time (s)
Power (W)
Relap 3D Relap-Parcs
Fuel CL Temperature Variation - HZP rod ejection
400 500 600 700 800 900 1000
0.0 10.0 20.0 30.0 40.0 50.0 60.0
Time (s)
Temperature (K)
Relap-Parcs Relap 3D httemp122702301
Fig. 18 - Power excursion calculated with coupled codes during REA at HZP
Fig. 19 - Fuel CL Temperature variation during transient
Energy Released to Fuel during REA
0.0 10.0 20.0 30.0 40.0 50.0 60.0
0 1 2 3 4 5 6 7 8 9
Time (s)
Energy (KJ/kg) Plane 01
Plane 05 Plane 10 Plane 15 Plane 18
Energy Released to Fuel during REA
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0
0 1 2 3 4 5 6 7 8 9
Time (s)
Energy (KJ/kg)
Plane 20 Plane 21 Plane 22 Plane 23 Plane 24
Fig 20 - Energy released to fuel, on axial planes 1, 5, 10, 15 & 18 inside channel S, during REA at HZP
Fig. 21 - Energy released to fuel, on axial planes 20, 21, 22, 23 & 24
inside channel S, during REA at HZP
HZP – COLLAPSED CONTROL ROD EJECTION
This reference case was created to idealize an accident with reactivity insertion beyond prompt critical, in order to verify core behaviour and evaluate parameters related to safety, although it is known that the probability of ejection of two different adjacent control assemblies is negligible for safety analysis. This was done because, in HZP condition, the ejection of the control assembly with the higher rod worth value do not have a large impact on core behaviour and could not cause the reactor to go beyond prompt critical. So, the concept was to consider an ejection of two adjacent control assemblies collapsed and inserted in the same thermal-hydraulic channel.
Basically, the accident is initiated from a critical core at HZP and is characterized by the ejection of a collapsed control assembly, indicated as CA n° 8 in Fig. 22. In the TH nodalization, these two FA belong to the same TH channel. The rods of the regulation banks (5, 6 and 7) and safety banks (2 and 4) are fully inserted in the core.
1 6 1
3 5 5 3
7 7 8
3 5 4 4 8 3
1 6 2 6 1
5 4 2 2 4 5
6 7 2 7 2 7 6
5 4 2 2 4 5
1 6 2 6 1
3 5 4 4 5 3
7 7 7
3 5 5 3
1 6 1
PARAMETER PARCS ALONE RELAP5/3.3- PARCS
RELAP5/3D - NESTLE
Control rod worth ($) 1.44 1.44 1.44
Peak power (% Nom) 530.8 949.2 994.8
Time of peak (s) 1.185 1.178 1.175
Pulse width (ms) 42 30 29
Fig. 22 - Control rod configuration Table 4 - Power pulse data and transient parameters
Core behaviour during this transient was the typical one encountered in REA with high reactivity insertion; a power spike turned around by Doppler effect, followed by a period of low power decay. In this calculation, reactor trip was delayed to simplify the comparison between methodologies used to simulate the problem. Normally, the shutdown banks would enter the core and their effect would be felt at around 2 seconds after transient starts.
Different contributions to the reactivity versus time obtained by PARCS during this transient are plotted in Fig. 24. The global reactivity grows rapidly and at t = 1.13 s reaches the total inserted reactivity (1.44 $), and starts to fall promptly because of the Doppler feedback. As can be drawn from this figure, the moderator feedback is as strong in magnitude as the Doppler feedback, but it starts being felt with some delay.
This magnitude of reactivity insertion caused a sharp power spike, as can be seen in Fig. 23 and 25. A power peak of more than 900% P N was predicted by both coupled codes; PARCS code in a stand-alone mode (meaning that limited TH feedback was used) calculated a peak of 500% P N . Table 4 presents information about the main important transient parameters and power pulse data.
Power Excursion - HZP collpsed rod ejection
0%
200%
400%
600%
800%
1000%
1200%
0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5
Time (s)
Power (%)
Relap 3D Relap-Parcs
Reactivity excursion - HZP collapsed rod ejection
-1.0 -0.5 0.0 0.5 1.0 1.5 2.0
0 1 2 3 4 5 6 7 8 9
Time (s)
Reactivity ($)
sumrho tfrho dmrho crrho
Fig. 23 – Power excursion calculated by coupled codes during collapsed REA in HZP
Fig. 24 - Different contributes in reactivity excursion calculated
with coupled RELAP-PARCS during collapsed REA in HZP
The maximum power, after the rod ejection, is reached in the closest less irradiated assembly (12,5) to the one which contains the ejected rod (13,5). This can also be seen in Fig. 26, where a view from the top of the 3D surface generated is presented.
On the other hand, axial power distribution, during rod ejection, remains practically uninfluenced, as showed in Fig. 27. It can be noticed a modest decrease in the relative power in the core upper zones, where most of producing power is concentrated. Energy released to the fuel during transient is presented in Fig. 28.
X 5
10 15
Y 5 10 15
Power
0 1 2 3 4
Power 4.2 4 3.8 3.6 3.4 3.2 3 2.8 2.6 2.4 2.2 2 1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 Frame 001 27 Nov 2005 Power Excursion
X
Y
5 10 15
5 10
15
Power4.2 4 3.8 3.6 3.4 3.2 3 2.8 2.6 2.4 2.2 2 1.8 1.6 1.4 1.2 1 0.8 0.6 0.4 0.2 Frame 001 27 Nov 2005 Power Excursion
Fig. 25 - Radial power distribution at t = 1.178 s during collapsed REA (HZP)
Fig. 26 - Radial power dist. at t = 1.178 s during collapsed REA (HZP) – Top view
Axial power distribution during transient
0 0.5 1 1.5 2 2.5 3
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26
Axial plane number
Relative power
Initial peak
Energy Released to Fuel during REA
0.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0 160.0 180.0 200.0
0 1 2 3 4 5 6 7 8 9
Time (s)
Energy (KJ/kg)
Plane 20 Plane 21 Plane 22 Plane 23 Plane 24
Fig. 27 - Axial power distribution at t=0 and t = 1.178 s, corresponding to peak power during collapsed REA in HZP
Fig. 28 - Energy released to fuel, on axial planes 20, 21, 22, 23 & 24 inside channel S, during collapsed REA in HZP
2 3 4 5 6 7 8 9 10 11 12 13 14 15 16
2 0.506 0.628 0.670 1.033 1.092
3 0.423 0.754 1.335 1.553 1.127 2.665 3.155 2.413 1.819
4 0.423 0.852 0.928 1.448 1.098 1.825 1.958 3.812 3.482 4.493 2.555
5 0.296 0.684 0.501 1.025 0.985 1.292 0.902 2.430 2.894 4.590 4.277 4.468 1.801
6 0.422 0.586 0.777 0.593 0.980 0.771 1.362 1.447 3.043 3.215 4.556 3.433 2.364
7 0.211 0.600 0.723 0.576 0.717 0.534 0.888 0.819 1.598 1.572 3.016 2.841 3.716 3.061 1.056 8 0.212 0.546 0.411 0.549 0.393 0.595 0.470 0.838 0.810 1.579 1.418 2.363 1.892 2.568 0.994 9 0.153 0.261 0.436 0.236 0.425 0.335 0.495 0.309 0.823 0.798 1.317 0.869 1.748 1.077 0.640 10 0.163 0.412 0.304 0.391 0.272 0.396 0.293 0.486 0.453 0.851 0.735 1.229 1.042 1.470 0.595 11 0.142 0.387 0.443 0.331 0.380 0.258 0.389 0.322 0.564 0.503 0.920 0.926 1.362 1.256 0.477
12 0.238 0.309 0.376 0.254 0.374 0.264 0.405 0.368 0.666 0.549 0.952 0.863 0.704
13 0.145 0.312 0.205 0.370 0.322 0.375 0.222 0.508 0.530 0.713 0.460 0.784 0.391
14 0.170 0.308 0.302 0.427 0.288 0.406 0.377 0.659 0.533 0.623 0.387
15 0.143 0.230 0.371 0.390 0.241 0.499 0.545 0.383 0.269
16 0.136 0.154 0.142 0.193 0.192
0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0
Time (s) 540.0
550.0 560.0 570.0 580.0 590.0 600.0
Temp (K)
WinGraf 4.1 - 11-23-2005
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JJJ tmi_HZP2_TRno_3D.r httemp122700510
JJ J J J JJ JJ J JJ JJ J J J J J
Fig. 29 - Averaged radial assembly power distribution at t = 1.178 s, corresponding to peak power 949,25 % during REA at HZP
Fig. 30 - Clad Temperature trend inside channel S on axial planes 5,
10,15, 20 & 23 calculated with RELAP5/3D during REA at HZP
HFP – CONTROL ROD 40% EJECTION
In this reference case, accident is initiated from a critical core at full power (HFP) and is characterized by the ejection of the CA n° 8 (see Fig. 13), initially 40 % withdrawn (2.016 m inserted from the top of the core). Control rod regulating banks (5, 6 and 7) and safety banks (1, 2, 3 and 4) are fully withdrawn.
The transient is completed in 8 seconds. Simulations activating reactor trip at 114% and without it were performed. If reactor trip were not to act, power would have reached 115% of P N , still not presenting problems to reactor safety, as accurately predicted by both coupled codes. Power excursion is shown in Fig. 31 and 35.
Radial power distribution at peak power is presented in Fig. 33 and 34.
Different contributions to the reactivity versus time obtained with PARCS code during this transient are plotted in Fig. 32. The total reactivity grows rapidly and at t = 1.10 s reaches the total inserted reactivity (0.11
$), and starts to fall because of the Doppler and moderator feedbacks.
Energy released to fuel during transient is presented in Fig. 36.
Power excursion with reactor trip - Rod ejection HFP
0 0.2 0.4 0.6 0.8 1 1.2 1.4
0 1 2 3 4 5 6 7 8
Time (s)
Relative Power
Relap-Parcs Relap 3D
Reactivity excursion with reactor trip- rod ejection in HFP
-0.3 -0.2 -0.1 0 0.1 0.2
0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5
Time (s)
Reactivity ($)
sumrho tfrho dmrho crrho
Fig.31 - Power excursion during REA in HFP with reactor trip Fig. 32 - Reactivity excursion during REA in HFP with reactor trip
X
5 10
15
Y 5
10 15
Power
0 0.5 1
Power 1.3 1.2 1.1 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1
Frame 001 28 Nov 2005 Power Excursion for rod ejection at HFPX
Y
5 10 15
5 10
15
Power1.3 1.2 1.1 1 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 Frame 001 28 Nov 2005 Power Excursion for rod ejection at HFP
Fig. 33 - Radial power distribution at t = 1.10 s during REA (HFP) Fig. 34 - Radial power distribution at t = 1.10 s during REA (HFP) – Top view
0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0
Time (s) .2
.4 .6 .8 1 1.2 1.4 x 10 6
Power (W)
WinGraf 4.1 - 11-28-2005
XXX tmi_HFP_TRno_3D.r rkozntpw522
X XX
XXXXXXX XXXXXXXX
X X
X YYY tmi_HFP_TRno_3D.r rkozntpw520
Y Y Y
YYYYYYYY YYYYYYY
Y
Y
ZZZ tmi_HFP_TRno_3D.r rkozntpw518
Z Z
ZZZZZZZZZ ZZZZZZ
Z Z
VVV tmi_HFP_TRno_3D.r rkozntpw515
V VVVVVVVVVVVVVVV
V V
V JJJ tmi_HFP_TRno_3D.r rkozntpw510
JJJJJJJJJJJJJJJ J
J J
Energy Released to Fuel during REA
0.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0
0 1 2 3 4 5 6 7 8 9
Time (s)
Energy (KJ/kg)
Plane 20 Plane 21 Plane 22 Plane 23 Plane 24
