NUREG/IA-0465 

International 

 Agreement Report

Fuel Rod Performance Uncertainty Analysis 

During Overpressurization Transient for 

Kuosheng Nuclear Power Plant with TRACE/

FRAPTRAN/ DAKOTA Codes in SNAP Interface 

Prepared by: 

Chunkuan Shih*, Hao-Chun Chang*, Jong-Rong Wang*, Shao-Wen Chen*, Hao-Tzu Lin, Show-Chyuan

Chiang**, Tzu-Yao Yu**

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New Energy

Education and Research Foundation 

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

**Department of Nuclear Safety, Taiwan Power Company

242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan

K. Tien, NRC Project Manager

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2015

Date Published: March 2016

Prepared as part of 

The technical exchange and cooperation between the U.S. Nuclear Regulatory Commission and the Task Force for

Safety Critical Software in the Field of Nuclear Safety Research

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0465 

International 

 Agreement Report

Fuel Rod Performance Uncertainty Analysis 

During Overpressurization Transient for 

Kuosheng Nuclear Power Plant with TRACE/

FRAPTRAN/ DAKOTA Codes in SNAP Interface 

Prepared by: 

Chunkuan Shih*, Hao-Chun Chang*, Jong-Rong Wang*, Shao-Wen Chen*, Hao-Tzu Lin, Show-Chyuan

Chiang**, Tzu-Yao Yu**

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New Energy

Education and Research Foundation 

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

**Department of Nuclear Safety, Taiwan Power Company

242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan

K. Tien, NRC Project Manager

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2015

Date Published: March 2016

Prepared as part of 

The technical exchange and cooperation between the U.S. Nuclear Regulatory Commission and the Task Force for

Safety Critical Software in the Field of Nuclear Safety Research

Published by

U.S. Nuclear Regulatory Commission
 
iii 

ABSTRACT 

After the MUR (measurement uncertainty recapture) power uprates, Kuosheng Nuclear Power 

Plant has uprated the power to 2943 MWt. Recently, Taiwan Power Company is concerned in 

SPU (stretch power uprated) plan and uprates the power to 3030 MWt, which is 104.7% of the 

designed power. Before the stretch power uprates, several transient analysis should be done for 

ensuring that the power plant could maintain stability in high power operating conditions. The 

advanced thermal hydraulic analysis code TRACE which is conducted by U.S. NRC was applied 

for overpressurization transient including main steam line isolation valves closure, turbine stop 

valves closure and turbine control valves closure. This closure of valves increased the dome 

pressure; as a result, the void fraction inside the reactor core decreased. The decline of the void 

fraction will increase the reactivity feedback; hence, the power increased rapidly until the reactor 

scram. Further, with safety relief valves open, the dome pressure would not exceed the criteria 

regulated by ASME. In addition, to cover the insufficiency of thermal hydraulic code, fuel rod 

transient analysis code FRAPTRAN was applied. To perform fuel rod transient analysis, thermal 

information from TRACE code would be entered into FRAPTRAN with power history. In addition, 

DAKOTA code was applied to concern the influence of fuel rod manufacturing tolerance. With 

uncertainty bands from DAKOTA analysis, the criteria could be judged with more confidence. 

This research successfully established a procedure of thermal hydraulic and fuel rod property 

analysis. Further, with SNAP interface, the FRAPTRAN and DAKOTA were combined 

successfully to perform the uncertainty analysis. 
iv 

 
v 

FOREWORD 

The U.S. NRC (United States Nuclear Regulatory Commission) is developing an advanced 

thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development 

of TRACE is based on TRAC, integrating RELAP5 and other programs. NRC has determined 

that in the future, TRACE will be the main code used in thermal hydraulic safety analysis, and no 

further development of other thermal hydraulic codes such as RELAP5 and TRAC will be 

continued. A graphic user interface program, SNAP (Symbolic Nuclear Analysis Program) which 

processes inputs and outputs for TRACE is also under development. One of the features of 

TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more 

accurate and detailed safety analysis of nuclear power plants. TRACE usually used in the 

nuclear power plants analysis. This report showed TRACE can also do the calculation of small 

system such as dry-storage cask. 

Taiwan and the United States have signed an agreement on CAMP (Code Applications and 

Maintenance Program) which includes the development and maintenance of TRACE. INER 

(Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) is the organization in 

Taiwan responsible for the application of TRACE in thermal hydraulic safety analysis, for 

recording user’s experiences of it, and providing suggestions for its development. To meet this 

responsibility, this research established a procedure of thermal hydraulic, fuel rod property and 

uncertainty analysis by using TRACE/FRAPTRAN/DAKOTA/SNAP for Kuosheng Nuclear Power 

Plant. 
vi 

 

 
vii 

CONTENTS 

Page

ABSTRACT ............................................................................................................................... iii 

FOREWORD .............................................................................................................................. v 

CONTENTS .............................................................................................................................. vii 

FIGURES ................................................................................................................................... ix 

TABLES .................................................................................................................................... xi 

EXECUTIVE SUMMARY ......................................................................................................... xiii 

ABBREVIATIONS .................................................................................................................... xv 

1. INTRODUCTION ................................................................................................................1-1 

2. METHODOLOGY ...............................................................................................................2-1 

3. MODEL ESTABLISHMENT ...............................................................................................3-1 

3.1 TRACE Thermal Hydraulic Model ..............................................................................3-1 

3.2 FRAPTRAN Fuel Rod Model ................................................................................... 3-14 

3.3 Animation Model ...................................................................................................... 3-18 

3.4 DAKOTA Uncertainty Analysis Model ...................................................................... 3-19 

4. RESULTS AND DISCUSSION ...........................................................................................4-1 

4.1 Main Steam Line Isolation Valves Closure (MSIVC) ..................................................4-1 

4.2 Turbine Stop Valves Closure (TSVC) ...................................................................... 4-14 

4.3 Turbine Control Valves Closure (TCVC) .................................................................. 4-24 

4.4 Transient Comparison ............................................................................................. 4-34 

5. CONCLUSIONS .................................................................................................................5-1 

6. REFERENCES ...................................................................................................................6-1 
viii 

 

 
ix 

FIGURES 

Page 

Figure 1 Flow Chart of the Uncertainty Analysis ................................................................. 2-2

Figure 2 TRACE Model of Kuosheng NPP .......................................................................... 3-4

Figure 3 Vessel Geometry of Kuosheng NPP in the TRACE Model .................................... 3-5

Figure 4 Fuel Rods Arrangement of Kuosheng NPP in TRACE Model................................ 3-6

Figure 5 Feedwater Control System of Kuosheng NPP in TRACE Model ........................... 3-7

Figure 6 General Setting of Power Component of Kuosheng NPP in TRACE Model .......... 3-8

Figure 7 Decay Heat Setting of Kuosheng NPP in TRACE Model ....................................... 3-9

Figure 8 Main Steam Pipe Lines Properties of Kuosheng NPP in TRACE Model .............. 3-10

Figure 9 Properties of Main Steam Line Isolation Valve of Kuosheng NPP ....................... 3-11

Figure 10 Feedwater Pumps Settings of the Kuosheng NPP in TRACE Model ................... 3-12

Figure 11 Recirculation Pump Properties of Kuosheng NPP in TRACE Model ................... 3-13

Figure 12 Setting of Radial Nodes in FRAPTRAN Model .................................................... 3-16

Figure 13 Setting of Axial Nodes in FRAPTRAN Model ...................................................... 3-17

Figure 14 Animation Model of Kuosheng NPP .................................................................... 3-18

Figure 15 Numeric Variable Setting of DAKOTA Analysis in SNAP Interface ..................... 3-20

Figure 16 Connection of the Numeric Variables and the Rod Properties ............................ 3-21

Figure 17 Uncertainty Job Stream Setting in SNAP Interface ............................................. 3-22

Figure 18 Variable Setting of DAKOTA Analysis in SNAP Interface .................................... 3-23

Figure 19 Variation of Dome Pressure and Steam Flow Rate in MSIVC Transient ............... 4-4

Figure 20 Variation of Dome Pressure and Core Power in MSIVC Transient ........................ 4-5

Figure 21 Relationship between Recirculation Flow Rate and Dome Pressure ..................... 4-6

Figure 22 Relationship between Dome Pressure and Safety Valves Flow Rates .................. 4-7

Figure 23 Cladding Temperature of FRAPTRAN Results in MSIVC Transient ...................... 4-9

Figure 24 Cladding Hoop Strain of FRAPTRAN Results in MSIVC Transient ....................... 4-9

Figure 25 Fuel Enthalpy of FRAPTRAN Results in MSIVC Transient ................................. 4-10

Figure 26 Oxidation Thickness of FRAPTRAN Results in MSIVC Transient ....................... 4-10

Figure 27 Gap Thickness of FRAPTRAN Results in MSIVC Transient ............................... 4-11

Figure 28 Uncertainty Band of Cladding Temperature at Node 5 in MSIVC Transient ........ 4-11

Figure 29 Uncertainty Band of Cladding Hoop Strain at Node 3 in MSIVC Transient .......... 4-12

Figure 30 Uncertainty Band of Gap Thickness at Node 12 in MSIVC Transient .................. 4-12

Figure 31 Uncertainty Band of Fuel Enthalpy at Node 7 in MSIVC Transient ...................... 4-13

Figure 32 Variation of Dome Pressure and Steam Flow Rate in TSVC Transient ............... 4-16

Figure 33 Variation of Core Power in TSVC Transient ........................................................ 4-16

Figure 34 Relationship between Dome Pressure and Rerlief Valves Flow Rate ................. 4-17

Figure 35 Cladding Temperature of FRAPTRAN Results in TSVC Transient...................... 4-19

Figure 36 Cladding Hoop Strain of FRAPTRAN Results in TSVC Transient ....................... 4-19

Figure 37 Fuel Enthalpy of FRAPTRAN Results in TSVC Transient ................................... 4-20

Figure 38 Oxidation Thickness of FRAPTRAN Results in TSVC Transient ......................... 4-20

Figure 39 Gap Thickness of FRAPTRAN Results in TSVC Transient ................................. 4-21

Figure 40 Uncertainty Band of Cladding Temperature at Node 8 in TSVC Transient .......... 4-21

Figure 41 Uncertainty Band of Cladding Hoop Strain at Node 8 in TSVC Transient............ 4-22

Figure 42 Uncertainty Band of Gap Thickness at Node 10 in TSVC Transient .................... 4-22

Figure 43 Uncertainty Band of Fuel Enthalpy at Node 8 in TSVC Transient ....................... 4-23

Figure 44 Variation of Dome Presssure and Steam Flow Rate in TCVC Transient ............. 4-26

Figure 45 Variation of Core Power in TCVC Transient ........................................................ 4-26

Figure 46 Variation of Feedwater Flow Rate in TCVC Transient ......................................... 4-27

Figure 47 Cladding Temperature of FRAPTRAN Results in TCVC Transient ..................... 4-29

Figure 48 Cladding Hoop Strain of FRAPTRAN Results in TCVC Transient ....................... 4-29

Figure 49 Gap Thickness of FRAPTRAN Results in TCVC Transient ................................. 4-30

Figure 50 Fuel Enthalpy of FRAPTRAN Results in TCVC Transient ................................... 4-30

Figure 51 Oxidation Thickness of FRAPTRAN Results in TCVC Transient ......................... 4-31

Figure 52 Uncertainty Band of Cladding Temperature at Node 4 in TCVC Transient .......... 4-31

Figure 53 Uncertainty Band of Cladding Hoop Strain at Node 5 in TCVC Transient ........... 4-32
x 

Figure 54 Uncertainty Band of Gap Thickness at Node 12 in TCVC Transient ................... 4-32

Figure 55 Uncertainty Band of Fuel Enthalpy at Node 8 in TCVC Transient ....................... 4-33

Figure 56 Comparisons of Peak Value Time....................................................................... 4-36
xi 

TABLES 

Page

Table 1 Comparison of the Initial Conditions of FSAR and TRACE ...................................... 3-3

Table 2 Fuel Rod Geometry ............................................................................................... 3-15

Table 3 Uncertainty Parameters Setting ............................................................................. 3-19

Table 4 Event Sequence of MSIVC Transient ....................................................................... 4-2

Table 5 Event Sequence of TSVC Transient ...................................................................... 4-14

Table 6 Event Sequence of TCVC Transient ...................................................................... 4-24

Table 7 Comparison of the Important Parameters for Three Hypothetical Accident ............ 4-35

 
xii 

 
xiii 

EXECUTIVE SUMMARY 

According to the user manual, TRACE is the product of a long term effort to combine the 

capabilities of the U.S. NRC’s four main systems codes (TRAC-P, TRAC-B, RELAP5 and 

RAMONA) into one modernized computational tool. U.S. NRC has ensured that TRACE will be 

the main code used in thermal hydraulic safety analysis in the future without further development 

of other thermal hydraulic codes, such as RELAP5 and TRAC. Besides, the 3-D geometry model 

of reactor vessel, which is one of the representative features of TRACE, can support a more 

accurate and detailed safety analysis of NPPs. TRACE also can provide greater simulation 

capability than the previous codes, especially for events like LOCA. 

FRAPTRAN (Fuel Rod Analysis Program Transient) is a light water reactor fuel rods simulation 

code which is conducted by U.S. NRC and developed by Pacific Northwest National Laboratory 

(PNNL) since 1997. FRAPTRAN is developed to calculate the single fuel rod behavior during 

transients and hypothetical accidents at burnup level up to 62 gigawatt-days per metric ton of 

Uranium (GWd/MTU). In this research, FRAPTRAN version 1.4 was applied. With FRAPTRAN 

1.4, the important criteria such as cladding temperature, cladding hoop strain, fuel enthalpy and 

radial gap can be determined. 

DAKOTA (Design Analysis Kit for Optimization and Terascale Applications) is a tool kit which is 

developed by Sandia National Laboratory (SNL). It supplies a simple setting interface for 

researchers to perform different iteration method with the analysis code. From the DAKOTA 

manual, the uncertainty quantification can be determined by several simulations with different 

input parameter sets. That is, with the comparison and computation of the input parameter 

distribution and output data variation, the uncertainty can be determined. 

SNAP (Symbolic Nuclear Analysis Package) is an interface of NPP analysis codes which 

developed by U.S. NRC and Applied Programming Technology, Inc. Different from the traditional 

input deck in text files, the graphical control blocks and thermal hydraulic connections make 

researches comprehend the whole power plant and control system more easily. Further, without 

any patch files, the SNAP interface allow researchers connect the DAKOTA uncertainty 

parameters with the analysis codes directly. It is a powerful tool to do uncertainty calculations 

more efficiently. 

Kuosheng NPP is located on the northern coast of Taiwan. Its nuclear steam supply system is a 

type of BWR/6 designed and built by General Electric on a twin unit concept. Each unit includes 

two loops of recirculation piping and four main steam lines, with the thermal rated power of 

2894MWt. After the project of MUR for Kuosheng NPP, the operating power is 2943 MWt. Unit 1 

will start SPU from Cycle 24 and Unit 2 will start SPU from Cycle 23. The operating power will be 

3030 MWt. In order to uprate the power, the assessments for the fuel rods mechanical properties 

under the SPU conditions are required. This research established a procedure of thermal 

hydraulic, fuel rod property and uncertainty analysis by using 

TRACE/FRAPTRAN/DAKOTA/SNAP. The overpressurization hypothetical accidents were 

analyzed in this study. 

From these analysis data, the Kuosheng NPP will be safe during the overpressurization transient 

after SPU. In spite that turbine bypass valve failed, the dome pressure would not exceed the 

criteria regulated by ASME if certain amounts of safety/relief valves opened. Further, the closure 

time of valves in this research were shorter than that in reality. The hypothetical transient 

conditions were more severe. However, the reactor scram setting and other safety mechanism 

could still maintain the integrity of the NPP and the fuel rods. On the other hand, for the 

uncertainty analysis, the manufacturing tolerance might cause 1.7% of cladding temperature 

deviation, 14% of cladding hoop strain deviation and 10% of fuel enthalpy deviation. 

Nevertheless, even concerning the uncertainty deviation, the fuel rod properties would not 

exceed the criteria. 
xiv 

 

 
xv 

ABBREVIATIONS 

ASME American Society of Mechanical Engineers 

BWR Boiling-Water Reactor 

cal/g calorie(s) per gram 

DAKOTA Design Analysis Kit for Optimization and Terascale Applications 

FRAPTRAN Fuel Rod Analysis Program Transient 

GWd/MTU Gigawatt-days per metric ton of uranium 

g gram(s) 

INER Institute of Nuclear Energy Research Atomic Energy Council, R.O.C. 

K Kelvin 

kg kilogram(s) 

kW kilowatt(s) 

LOCA Loss of Coolant Accidents 

MPa Megapascal(s) 

MSIVC Main Steam Isolation Valves Closure 

MUR Measurement Uncertainty Recapture 

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission 

Pa Pascal(s) 

PCT Peak Cladding Temperature 

PCMI Pellet Cladding Mechanical Interaction 

PDFs Probability Distributions Functions 

PNNL Pacific Northwest National Laboratory 

SNAP Symbolic Nuclear Analysis Program 

SNL Sandia National Laboratory 

SPU Stretch Power Uprated 

SRV Safety/Relief Valves 

TCVC Turbine Control Valves Closure 

TRACE TRAC/RELAP Advanced Computational Engine

TSVC Turbine Stop Valves Closure 

W Watt(s) 

US United States 

μm micrometer(s) 

 
xvi 

 

 
1-1 

1. INTRODUCTION 

Kuosheng NPP is located on the northern coast of Taiwan. Its nuclear steam supply system is a 

type of BWR/6 designed and built by General Electric on a twin unit concept. Each unit includes 

two loops of recirculation piping and four main steam lines, with the thermal rated power of 

2894MWt. After the project of MUR for Kuosheng NPP, the operating power is 2943 MWt [1, 2]. 

Unit 1 will start SPU from Cycle 24 and Unit 2 will start SPU from Cycle 23. The operating power 

will be 3030 MWt. In order to uprate the power, the assessments for the fuel rods mechanical 

properties under the SPU conditions are required. This research established a procedure of 

thermal hydraulic, fuel rod property and uncertainty analysis by using 

TRACE/FRAPTRAN/DAKOTA/SNAP. The overpressurization hypothetical accidents were 

analyzed in this study. 

TRACE (TRAC/RELAP Advanced Computational Engine) is a thermal hydraulic analysis code 

which is developed by Apt Inc. and conducted by U.S. NRC. It can perform the Loss of Coolant 

Accidents (LOCAs), operational transients or other accident scenario with best-estimate 

analyses [3]. Different from other traditional thermal hydraulic code, in the TRACE code, the 

component Vessel which can automatically considered the three dimensional heat transfer is 

applied to simulate reactor vessel. With this component, researchers can easily assign how 

many sections, radial rings and height level are needed. 

FRAPTRAN (Fuel Rod Analysis Program Transient) is a light water reactor fuel rods simulation 

code which is conducted by U.S. NRC and developed by Pacific Northwest National Laboratory 

(PNNL) since 1997. FRAPTRAN is developed to calculate the single fuel rod behavior during 

transients and hypothetical accidents at burnup level up to 62 gigawatt-days per metric ton of 

Uranium (GWd/MTU) [4]. In this research, FRAPTRAN version 1.4 was applied. With 

FRAPTRAN 1.4, the important criteria such as cladding temperature, cladding hoop strain, fuel 

enthalpy and radial gap can be determined. However, some long term properties such as fuel 

densification, swelling, radiation growth and cladding creep are not concluded in FRAPTRAN 

code. Hence, the FRAPCON-3 code is also applied in this research.

FRAPCON-3 is also conducted by U.S. NRC and developed by PNNL. Different from 

FRAPTRAN code, FRAPCON-3 analyzes the fuel rod behavior when the power and coolant 

conditions change slowly. The calculation results supply several fuel rods properties and burnup 

information after a long-term operation [5]. The restart file from FRAPCON-3 analysis can be the 

initial conditions of FRAPTRAN analysis. With this restart file, some long-term operating 

properties can be concerned in FRAPTRAN transient analysis. In this research, FRAPTRAN 

code referenced the restart file of 18-month fuel operation which was calculated by FRAPCON-3 

code. 

DAKOTA (Design Analysis Kit for Optimization and Terascale Applications) is a tool kit which is 

developed by Sandia National Laboratory (SNL). It supplies a simple setting interface for 

researchers to perform different iteration method with the analysis code. From the DAKOTA 

manual, the uncertainty quantification can be determined by several simulations with different 

input parameter sets. That is, with the comparison and computation of the input parameter 

distribution and output data variation, the uncertainty can be determined [6]. 

SNAP (Symbolic Nuclear Analysis Package) is an interface of NPP analysis codes which 

developed by U.S. NRC and Applied Programming Technology, Inc. Different from the traditional 

input deck in text files, the graphical control blocks and thermal hydraulic connections make 

researches comprehend the whole power plant and control system more easily. Further, without 

any patch files, the SNAP interface allow researchers connect the DAKOTA uncertainty 

parameters with the analysis codes directly. It is a powerful tool to do uncertainty calculations 

more efficiently. 
1-2 

 

 
2-1 

2. METHODOLOGY 

For the fuel rod transient analysis in FRAPTRAN code, the input parameters required can be 

basically divided into three parts including materials of rods, geometry of rods and thermal 

hydraulic conditions. The previous two parts can be obtained from the FSAR and Taiwan Power 

Company training documents. The thermal hydraulic conditions, however, are hard to be 

determined especially in hypothetical accidents. Typically, conditions of some similar plant 

transients would be used. Nevertheless, it is hard to find the transient conditions quite the same 

as the hypothetical accident. To overcome this difficulty, in this research, the thermal hydraulic 

code TRACE is performed for gathering thermal conditions, which would be closer to the 

hypothetical accidents [7]. Then, the thermal conditions would be input into the FRAPTRAN code 

to do the further fuel rods analysis. In addition, the manufacturing tolerance of fuel rods had been 

concerned. Together with the DAKOTA uncertainty job stream in SNAP interface, the output data 

results of FRAPTRAN code would not be a single curve but an uncertainty band which varied 

with time. Due to the DAOKOTA uncertainty iterations, there would be several FRAPTRAN 

computations performed and generate a large amounts of data results. As a result, a data 

extraction program, which can extract a specific interesting parameters from all data results in a 

short time, was developed by our group [8]. To conclude, the flowchart of this methodology is 

shown in Figure 1. 
2-2 

Figure 1 Flow Chart of the Uncertainty Analysis

 
3-1 

3. MODEL ESTABLISHMENT 

3.1 TRACE Thermal Hydraulic Model 

TRACE model of Kuosheng NPP was built up according to the FSAR, design documents and 

TRACE manual. Further, only the NSSS was considered in this TRACE model. As a result, the 

turbine was set as boundary conditions in this research. The power plants system outside the 

turbine would not be considered in this model, as shown in Figure 2. From the comparison of 

TRACE steady state data results and the operating conditions as shown in Table 1, it is noticed 

that the TRACE model suitable for the Kuosheng NPP. Details of the pressure vessel, reactor 

core, feedwater controlling system, water level computation and power simulation of this analysis 

model are described as follows: 

Vessel 

The pressure vessel is simulated by the 3-D component Vessel which includes down comers, 

fuel assemblies, upper and lower plenum. This component are divided into 2 sections, 4 radial 

rings and 11 height levels. Details of the vessel component is shown in Figure 3. 

Channel 

The fuel assemblies of Kuosheng NPP are simulated by six 1-D components “Channel”, which 

were applied for simulating the flow inside and besides the fuel assemblies. Each Channels was 

divided into 27 axial nodes, 6 radial nodes. There are 96 to 120 fuel assemblies were set up in 

these six channels respectively. Further, inside the fuel assemblies, there are three kinds of fuel 

rods including full-length, half-length and water rods. Details of the rods arrangement and 

properties were set up according to the FSAR and shown in Figure 4. 

Feedwater Control System 

Three feedback signals including water level, steam flow rate and feedwater flow rate are set up 

in this TRACE model. As these three signals deviate the normal operating conditions, the 

controlling system will adjust the valves open of the feedwater. Moreover, although the water 

level can be measured directly through the control blocks in the TRACE model, the NRWL is also 

built up referred to the reality in the TRACE model, as shown in Figure 5. 

Power 

In the TRACE code, there are three neutron feedback modes including Power Table, Point 

kinetics or combined with PARCS. The point kinetics mode with reactivity feedback table is 

chosen for Kuosheng NPP in this TRACE model. In addition to the reactivity feedback, decay 

heat should also be concerned. There are also several decay heat groups that have been built in. 

In this model, the ANS-73 was chosen. Details of the power components setting are shown in 

Figure 6 and Figure 7. 

Steam Lines 

According to the configuration of Kuosheng NPP, there are four main steam pipe lines in the 

TRACE model. On the main steam pipe lines, there are main steam line isolation valves, turbine 

stop valves and turbine control valves. Further, 16 safety/relief valves are also set up on the main 

steam pipe lines. At the ends of the pipe lines, the Break components are set to simulate the 

turbines, which are assumed to be the boundaries of this model. In addition, the turbine bypass 

pipe line and valve are also built in this model as shown in Figure 2. All the geometry and thermal 

hydraulic conditions of these components mentioned above are referred to the FSAR, training 

materials and configuration of Kuosheng NPP. Details of the properties are shown in Figure 8 

and Figure 9. 
3-2 

Pumps 

Basically, there are three kinds of pumps including feedwater pump, recirculation pump and jet 

pump to maintain the circulation of the reactor core. In this model, due the assumption of the 

boundaries, two feedwater pumps in this model are simulated by the components Fills. These 

Fills are connected to valves which flow area could be changed to control the feedwater flow rate. 

These valves then would be connected to the vessel component at the position radial ring 4, 

section 1 and level 7. 

There are 2 recirculation pumps shown in Figure 2. In front of the outlet of each recirculation 

pump, there is a control valve which can be used for controlling the recirculation flow rate. 

Further, a jet pump is connected on each recirculation loops to inject recirculation flow inside the 

reactor vessel at position radial ring 4, section 1 and level 6. Details of these pumps are shown in 

Figure 10 to Figure 12. 

 
3-3 

Table 1 Comparison of the Initial Conditions of FSAR and TRACE 

Parameters Unit FSAR TRACE model 

Power 

Dome Pressure 

Feedwater Flow 

Steam Flow 

Core inlet flow 

Recirculation flow (Single loop) 

Core exit pressure 

NWRL 

(MWt) 

(MPa) 

(kg/sec) 

(kg/sec) 

(kg/sec) 

 (kg/sec) 

(MPa) 

(m) 

3030 

7.17 

1645.3 

1645.3 

10645 

1549.5 

7.23 

0.934 

3030 

7.17 

1641 

1641 

10704.8 

1538.6 

7.3 

0.934 

 
3-4 

Figure 2 TRACE Model of Kuosheng NPP
3-5 

Figure 3 Vessel Geometry of Kuosheng NPP in the TRACE Model 
3-6 

Figure 4 Fuel Rods Arrangement of Kuosheng NPP in TRACE Model 

 
3-7 

Figure 5 Feedwater Control System of Kuosheng NPP in TRACE Model
3-8 

Figure 6 General Setting of Power Component of Kuosheng NPP in TRACE Model 
3-9 

Figure 7 Decay Heat Setting of Kuosheng NPP in TRACE Model 

 
3-10 

Figure 8 Main Steam Pipe Lines Properties of Kuosheng NPP in TRACE Model 
3-11 

Figure 9 Properties of Main Steam Line Isolation Valve of Kuosheng NPP 
3-12 

Figure 10 Feedwater Pumps Settings of the Kuosheng NPP in TRACE Model 
3-13 

Figure 11 Recirculation Pump Properties of Kuosheng NPP in TRACE Model 

 
3-14 

3.2 FRAPTRAN Fuel Rod Model 

Fuel Rod Geometry and Convergence Settings 

The fuel rod geometric information was shown in Table 2. Some steady state fuel rod information 

such as burn-up value and fission gas is calculated by FRAPCON-3. The long-term information 

of FRAPCON-3 data results will be combined into the restart file, which can be read and 

calculated by FRAPTRAN code. As mentioned before, the boundary conditions required would 

be performed by TRACE and transferred to FRAPTRAN code manually. 

In the FRAPTRAN analysis, the time interval was 0.01 second. The analysis time was set up 

according to the hypothetical accidents respectively. The gap pressure and temperature 

convergence are both 0.001. The maximum temperature variation between two iterations is 6K. 

After setting up the convergence index, the limitation of iteration numbers is also required to 

prevent the computer crashing if the calculations could not meet the convergence index. In this 

research, both the limitations of iterations in transient and steady state are set as 100. 

Fuel Rod Node 

In this research, the fuel rod nodes were divided into radial and axial directions. For the radial 

directions, it can be further divided into two settings including nodes inside the fuel pellet and 

nodes inside the cladding. As shown in Figure 12 and Figure 13, there are 13 radial nodes with 

equal interval started from the center of the fuel pellet and ended at the surface of the fuel pellet.

Besides, same in the radial directions, there are two nodes inside and outside the cladding. 

However, it should be noticed that the radials nodes would be only calculated in the code 

iterations. The output data would normalized the radial nodes information. For the axial direction, 

in this research, the fuel rod is divided into 12 nodes with equal interval. The output information 

of temperature, pressure, strain, stress and so on will be recorded and plotted mainly according 

to these 12 axial nodes. 

 
3-15 

Table 2 Fuel Rod Geometry

Parameters Input codes value Unit Notes 

Length RodLength 12.45 Feet 

Diameter RodDiameter 0.03608 Feet 

Gap thickness gapthk 0.00031 Feet Replaced with FRAPCON data

Upper spring coils ncs 59 N/A 

Upper spring height spl 0.7513 Feet 

Upper spring diameter scd 0.02583 Feet 

Upper spring wire diameter swd 0.00425 Feet 

Cold works coldw 0.5 N/A 

Cladding roughness roughc 1 um 

Helium fraction gfrac(1) 1 N/A 

Fill pressure gappr0 74.7 psia 

Pellet diameter FuelPelDiam 0.0305 Feet Replaced with FRAPCON data

Pellet height pelh 0.0366 Feet Replaced with FRAPCON data

Fuel surface roughness roughf 1 um Replaced with FRAPCON data

Pellet density frden 0.945 N/A 

Fuel sintering temperature tsntrk 1773 K 

Fuel grain size fgrns 10 um 
3-16 

Figure 12 Setting of Radial Nodes in FRAPTRAN Model 

 
3-17 

Figure 13 Setting of Axial Nodes in FRAPTRAN Model 
3-18 

3.3 Animation Model 

Because of developing analysis model with SNAP interface, analysis data results could be 

transferred into animations which could illustrate the situation of NPP during transients more 

easily and clearly. As shown in Fig.14, the water level of reactor vessel and the main steam pipe 

lines, which color depth would be changed to illustrate variation of void fraction during transients, 

were connected to the TRACE data results. Inside the reactor vessel, there are three bars 

divided into 12 blocks, which connected to FRAPTRAN data results, to display different kind of 

fuel rod properties including cladding temperature, cladding hoop strain and fuel enthalpy. On 

the both side of this figure, there are 5 bars which can show the specific value of important 

criteria. Further, in the middle of this figure, there are initial conditions of Kuosheng NPP and 

events sequence of the specific transient. From the animation model, interactions among 

different parts of the NPP could be observed more easily. Moreover, combining data results of 

TRACE and FRAPTRAN codes is also a new concepts for NPP safety analysis. 

Figure 14 Animation Model of Kuosheng NPP 

 
3-19 

3.4 DAKOTA Uncertainty Analysis Model 

In order to simplified analysis and reduce computational time, the influence of sub-model was not 

concerned in this research. Nonetheless, for conservative reason, all the sub-model such as 

Ballooning, BJK model were turned on. According to NUREG/CR-7001 report [9], which was 

prepared in 2009 by the FRPATRNA developing group, 6 manufacturing tolerances including 

cladding diameter, fuel pellet diameter, fuel pellet density, gap fill pressure, cladding roughness 

and fuel pellet surface roughness were concerned in this research. Table 3 lists the setting 

details of these uncertainty parameters. This table was set through the SNAP interface which is 

enable to connect DAKOTA variables with the FRAPTRAN analysis. 

Table 3 Uncertainty Parameters Setting 

Uncertainty parameters Mean Std. Distribution

Cladding outer diameter (mm) 11.22 0.04 Normal 

Cladding roughness (um) 1.00 0.30 Normal 

Fuel pellet outer diameter (mm) 0.98 0.01 Normal 

Fuel pellet density (% TD)* 94.50 0.91 Normal 

Fuel pellet roughness (um) 1.00 0.3 Normal 

Rod fill pressure (Pa) 515038.37 68999.97 Normal 

After built up the FRAPTRAN model with the SNAP interface, the variables which could be 

connected with the DAKOTA code should be set up. On the left column of SNAP interface, as 

shown in Figure 15, there is a Numerics list which could define several types of variables such as 

integers, reals, Booleans, strings and so on. According to the parameters shown in Table 3, 

there are six variables, “real”, being defined in this research. After setting up the variables, as 

shown in Figure 16, there will be connection buttons next to the input box of the rod geometry 

and fuel pellet setting columns. After connecting the variables respectively, the ranges of 

variables should also be defined. As shown in Figure17, the Job stream types should be set as 

“DAKOTA Uncertainty” mode. Then, clicking the parameter properties button in the “Job stream” 

column as shown in Figure 18, the uncertainty parameters could be defined respectively 

according to Table 3. 

In this research, all the uncertainty parameters were calculated at the same time. This method 

would be closer to the reality because the interaction of each parameters would be concerned. 

However, a large amount of data results would be generated; as a result, the well-established 

statistical concepts and data extraction method should be applied. In order to process the data 

files, a data extraction program was developed by our group. 

 
3-20 

Figure 15 Numeric Variable Setting of DAKOTA Analysis in SNAP Interface 
3-21 

Figure 16 Connection of the Numeric Variables and the Rod Properties 
3-22 

Figure 17 Uncertainty Job Stream Setting in SNAP Interface 
3-23 

Figure 18 Variable Setting of DAKOTA Analysis in SNAP Interface 
3-24 

 
4-1

4. RESULTS AND DISCUSSION 

According to the regulation of American Society of Mechanical Engineers (ASME), transients 

including main steam line isolation valve closure (MSIVC), turbine stop valve closure (TSVC) and 

turbine control valve closure (TCVC) have the potential to cause maximum peak system 

pressure. For conservative reason, turbine bypass valve in hypothetical accidents mentioned 

above would not open; further, setpoints of recirculation trip and safety/relief valves open would 

be higher than the real operating conditions in the NPP. 

In the past works, the hypothetical scenario would be applied in the Kuosheng NPP TRACE 

model. Data results calculated by TRACE model could be compared with the criteria. According 

to the Boiler and Pressure Vessel Code by ASME, pressure of reactor vessel should not exceed 

110% of the design value during transients, that is: 

Safety criteria of dome pressure = 1.10 x 8.62 MPa = 9.58 MPa 

Moreover, because FRAPTRAN code was also applied, criteria of the fuel rods would be more 

important in this research. Due to the variation of the reactivity and increasing of the power 

during the transient, fuel rod criteria of Reactivity Induced Accident (RIA) were referred in this 

research. According to the NUREG 0800, during the RIA transient, the cladding temperature 

should not exceed 1088K; the cladding hoop strain should not exceed 0.01; the fuel enthalpy 

should not exceed 710600 J/kg (170 cal/g) and the oxidation layer increasing should not be 

larger 17% of the initial oxidation thickness. Further, the gap thickness between fuel pellets and 

fuel cladding should also be notice to prevent the Pellet Cladding Mechanical Interaction (PCMI). 

Based on these criteria and regulations, safety of the Kuosheng NPP during the 

overpressurization transients could be determined. 

4.1 Main Steam Line Isolation Valves Closure (MSIVC) 

Transient Assumption

In this hypothetical accident, a 210-second steady state was performed for ensuring that all the 

parameters matched the operating conditions. At time point 210 second, the main steam line 

isolation valves (MSIVs) closed in three seconds. In order to promote the difficulties of this 

transient to the plant system, the closure time in this case was shorter than that described in 

FSAR. According to the FSAR description, reactor scram signal in this case might come from two 

ways including neutron flux exceeding 122% or dome pressure exceeding 7.66 MPa. Further, as 

the dome pressure reached to 7.82 MPa, recirculation pump trip signal would be sent out with 

delayed time 0.14 second. For the conservation reason, only 11 safety valves works even though 

that there are 16 safety/relief valves on the main steam pipelines. All these 11 safety valves were 

still divided into three groups with different set-points including 8.38 MPa for two valves, 8.48 

MPa for five valves and 8.55 MPa for four valves. Besides, after the closure signal sent out, the 

safety valves would fully close in 0.15 second with an electronic delayed time 0.4 second. Table 

4 lists the MSIVC events sequence and important set-points. From the data results, the reactor 

scrammed due to neutron flux exceeding 122% in this transient. Further, the safety valves will 

fully open in 0.15 second with delayed time 0.4 second. Once the safety valves open, the dome 

pressure was under controlled and the NPP was back to the safe situation. 

 
4-2

Table 4 Event Sequence of MSIVC Transient 

Time (sec) Events Notes 

0~210 Steady state 

Power 3030 MWt 

Feedwater flow rate 1641 kg/sec 

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 kg/sec 

Dome pressure 7.17MPa 

210 MSIVs closure Fully closure time 3 seconds 

213.06 MSIVs fully closed 

213.14 Reactor scram signal 

initiated Power reached to 122% 

213.23 Reactor scram Delayed time 0.09 second 

213.46 Recirculation pumps 

tripped Dome pressure 7.82MPa 

214.53

Safety valves group 1 

opened (8.38MPa,Group 

1)

Delayed 0.4 second 

214.65 

Safety valves group 2 

opened (8.48MPa,Group 

2) 

Delayed 0.4 second 

214.73 

Safety valves group 3 

opened (8.55MPa,Group 

3) 

Delayed 0.4 second 

220 End of the analysis 

 
4-3

Analysis Data Results 

The MSIVs started to close at 210 second and fully closed at 213 second. As a result, Figure 19 

shows that the steam flow rate greatly decreased 2 second later the transient started. Further, at 

213 second, the MSIVs fully closed and the dome pressure increased greatly. As a result, the 

void fraction inside the reactor core decreased which increased the power greatly as shown in 

Figure 20. This figure also plotted two scram conditions including the 122% reactor power and 

dome pressure 7.66 MPa. From this figure, it is known that the reactor scrammed because of the 

excessive power. 

Despite the fact that the reactor had scrammed at 213.23 second, the decay heat still heated the 

reactor vessel and generated steam. As a result, the dome pressure kept increasing. As the 

dome pressure reached to 7.82 MPa, the recirculation pumps tripped as shown in Figure 21. At 

214.53 second, the dome pressure was higher enough to meet the setpoint of group 1 safety 

valves. However, 3 opened safety valves discharged the steam insufficiently. The dome 

pressure kept increasing until the group 2 and group 3 safety valves opened at 214.65 and 

214.73 second respectively as shown in Figure 22. Further, because the dome pressure varied 

near the setpoints, the safety valves opened and closed for three times. 

From the TRACE data, it is known that the core power could be controlled by the scram system. 

In addition, as long as the safety valves opened, the dome pressure would not exceed the ASME 

criteria 9.58 MPa. Kuosheng NPP was safe in the MSIVC transient. 
4-4

210 212 214 216 218 220

7.0

7.2

7.4

7.6

7.8

8.0

8.2

8.4

8.6

8.8

9.0

 Dome Pressure Steam Flow

Time (Second)

Dome Pressure (MPa)

-500

0

500

1000

1500

2000

 Steam Flow (kg/s)

Figure 19 Variation of Dome Pressure and Steam Flow Rate in MSIVC Transient 
4-5

210 212 214 216 218 220

7.2

7.4

7.6

7.8

8.0

8.2

8.4

8.6

8.8

9.0

 Dome Pressure Core Power

Time (second)

Dome Pressure (MPa)

7.66 MPa

122% Neutron flux

0

1000

2000

3000

4000

5000

6000

7000

 Core Power (MW)

Figure 20 Variation of Dome Pressure and Core Power in MSIVC Transient
4-6

210 212 214 216 218 220

7.0

7.2

7.4

7.6

7.8

8.0

8.2

8.4

8.6

8.8

9.0

Recirculation Pump Flow Rate & Dome Pressure (MSIVC)

 Dome Pressure

 Recirculation Flow

 Setpoint of Recirculation Pump Trip (7.82 MPa)

Time (Second)

Dome Pressure (MPa)

-200

0

200

400

600

800

1000

1200

1400

1600

Mass Flow Rate (Kg/s)

Figure 21 Relationship between Recirculation Flow Rate and Dome Pressure 

 
4-7

212 214 216 218 220

7.0

7.2

7.4

7.6

7.8

8.0

8.2

8.4

8.6

8.8

Setpoint:

Group 3, 8.55 MPa

Group 2, 8.48 MPa

 Dome Pressure

 Valve 2010 (Group 1)

 Valve 2011 (Group 2)

 Valve 2013 (Group 3)

Time (Second)

Dome Pressure (MPa)

Group 1, 8.38 MPa

-20

0

20

40

60

80

100

120

140

Mass Flow Rate (Kg/s)

Figure 22 Relationship between Dome Pressure and Safety Valves Flow Rates 
4-8

So far, only the thermal hydraulic properties were discussed. Details of the fuel rods such as 

burnup information and mechanical properties could not be calculated by the TRACE code. 

Figure 23 shows the average cladding temperature variation during the MSIVC transient. During 

the MSIVC transient, the cladding temperature increased because of the increasing power. 

However, the cladding temperature of Node 5 to Node 8 appeared higher value because of the 

power distribution of the fuel rods. Further, cladding temperature of Node 1 was much lower than 

that of other Nodes because of the good heat transfer conditions. 

Figure 24 shows the average cladding hoop strain variation during the MSIVC transient. Once 

the power and temperature increased, the gap gas would inflate and the fuel cladding would 

expand with heat. On the other hand, the increasing dome pressure would squeeze the fuel 

cladding, especially at the lower position such as Node 1. To conclude, the hoop strain vibrated 

during the transient because the thermal hoop strain and elastic hoop strain opposed to each 

other. From Figure 24, it is obvious that the cladding hoop strain would be at most 0.00154 which 

is much lower than the criteria because of the high coolant pressure. 

In addition to the cladding hoop strain, the fuel pellet enthalpy should also be noticed in 

overpressurization transient. Figure 25 shows the enthalpy variation during the transient. The 

peak value is about 165 kJ/kg (39.47 cal/g), which is much lower than the criteria 170 cal/g. From 

those three criteria mentioned above, it is known that the fuel rods would keep good integrity 

during the MSIVC transient. However, some further parameters should be examined to ensure 

whether the water-metal reaction and the Pellet Cladding Mechanical Interaction (PCMI) 

happened or not. Figure 26 shows oxide thickness of the inside cladding. From this figure, it is 

noticed that there was no oxidation growth during the transient. On the other hand, from Figure 

27, it is noticed that gap distance between cladding and fuel pellet was always larger than 

0.064mm, which means that there is no PCMI. 

In addition to the general FRAPTRAN analysis, uncertainty analysis of fuel rod properties are 

also concerned in this research. To make the uncertainty analysis results simple, there is only 

one uncertainty band with the highest value will be plotted in the figure. According to Figure23, 

the Node 5 was chosen as the representable in Figure 28. With the manufacturing tolerance 

uncertainties, the peak cladding temperature might increase up to 607K during the MSIVC 

transient. However, it is still much lower than the criteria 1088K even with the manufacturing 

tolerance concern. For the cladding hoop strain, as shown in Figure 29, the peak value of the 

uncertainty band is about 0.00154, which does not exceed the criteria 0.01. However, it should 

be noticed that the lower bound of the hoop strain uncertainty band is lower than zero, which 

means that the cladding might be squeezed. As a result, the gap distance between cladding and 

fuel pellet should be examined. According to the results in Figure 27, Figure 30 shows the 

uncertainty band of gap distance at Node 12. From this figure, it can be noticed that the gap 

distance is always larger than 0.52mm. The PCMI effect would not happen in MSIVC transient. 

Figure 31 shows the uncertainty band of fuel enthalpy at Node 7. With manufacturing 

uncertainties, the peak value is about 154000 J/kg, which is much lower than the criteria 170 

cal/g. From these general and uncertainty analysis results of FRAPTRAN code mentioned above, 

the fuel rod still kept good integrity in MSIVC transient. 
4-9

210 212 214 216 218 220

575

580

585

590

595

600

Cladding Temperature (K)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Cladding Temperature (MSIVC)

Figure 23 Cladding Temperature of FRAPTRAN Results in MSIVC Transient 

210 212 214 216 218 220

0.00134

0.00136

0.00138

0.00140

0.00142

0.00144

0.00146

0.00148

0.00150

0.00152

0.00154

Cladding Hoop Strain

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Hoop Strain (MSIVC)

Figure 24 Cladding Hoop Strain of FRAPTRAN Results in MSIVC Transient 
4-10

210 212 214 216 218 220

100000

110000

120000

130000

140000

150000

160000

170000

Avetage Fuel Enthalpy (J/kg)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Fuel Enthalpy (MSIVC)

Figure 25 Fuel Enthalpy of FRAPTRAN Results in MSIVC Transient 

210 212 214 216 218 220

0.0000

0.0002

0.0004

0.0006

0.0008

0.0010

Oxide thickness (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Inside Oxidation (MSIVC)

Figure 26 Oxidation Thickness of FRAPTRAN Results in MSIVC Transient 
4-11

210 212 214 216 218 220

0.064

0.066

0.068

0.070

0.072

0.074

0.076

0.078

0.080

0.082

0.084

Radial gap (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Pellet-Cladding Gap (MSIVC)

Figure 27 Gap Thickness of FRAPTRAN Results in MSIVC Transient 

210 212 214 216 218 220

570

575

580

585

590

595

600

605

610

Temperature (K)

Time (Second)

 AVE MAX MIN

Cladding Temperature Uncertainty at Node 5

Figure 28 Uncertainty Band of Cladding Temperature at Node 5 in MSIVC Transient 

 
4-12

210 212 214 216 218 220

-0.0005

0.0000

0.0005

0.0010

0.0015

0.0020

Hoop Strain

Time (Second)

 AVE MAX MIN

Cladding Hoop Strain Uncertainty at Node 3

Figure 29 Uncertainty Band of Cladding Hoop Strain at Node 3 in MSIVC Transient 

210 212 214 216 218 220

0.050

0.055

0.060

0.065

0.070

0.075

Gap (mm)

Time (Second)

 AVE MAX MIN

Structual Radial Gap Uncetainty at Node 12

Figure 30 Uncertainty Band of Gap Thickness at Node 12 in MSIVC Transient 
4-13

210 212 214 216 218 220

100000

110000

120000

130000

140000

150000

160000

170000

180000

Enthalpy (J/kg)

Time (Second)

 AVE MAX MIN

Average Fuel Enthalpy Uncertainty at Node 7

Figure 31 Uncertainty Band of Fuel Enthalpy at Node 7 in MSIVC Transient 
4-14

4.2 Turbine Stop Valves Closure (TSVC) 

Transient Assumption 

In the hypothetical accident of TSVC, a 500 second steady state was performed to ensure all the 

parameters matched the operation conditions. At 500 second, the turbine tripped. As a result, the 

turbine stop valves started to close with closure time 0.1 second. As the TSVs reached to 90% 

open, the reactor scram signal was initiated. According to the FSAR, there would be 0.08 second 

delayed time. After the TSVs closure, the dome pressure increased and the void fraction inside 

the reactor core decreased. Hence, a positive reactivity feedback was performed and the power 

increased. 

For conservative reason, on 6 safety relief valves opened with delayed time 0.4 second as the 

dome pressure reached to the setpoints 7.94 MPa. Further, the safety relief valves would fully 

opened in 0.15 second. Once the safety relief valves opened, the dome pressure and the power 

plants would be under controlled. Table 5 listed the event sequences and important setpoints of 

the TRACE model during TSVC transient. 

Table 5 Event Sequence of TSVC Transient 

Time (sec) Events Notes 

0~499 Steady state 

Power 3030 MWt 

Feedwater flow rate 1641 kg/sec 

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 kg/sec 

Dome pressure 7.17MPa 

500 TSVs start to close Fully closure time 0.1 second 

500.01 Reactor scram signal 

initiated Signal initiated at TSVs 90% open 

500.09 Reactor scram Delayed time 0.08 second 

500.1 TSVs fully closed 

501.77 Relief valves opened Open at dome pressure 7.94MPa and 

close at dome pressure 7.63MPa 

502.27 Reactor dome pressure 

peak value Dome pressure 8.20 MPa 

505 End of the analysis 

 
4-15

Analysis Data Results 

The TSVs started to close at 500 second and fully closed at 500.1 second. Figure 32 shows that 

at the beginning of the transient, the steam flow dropped rapidly. The steam inside the reactor 

vessel was trapped; hence, the dome pressure increased. Due to the increasing dome pressure, 

the void fraction of the reactor core would decrease which caused a positive reactivity feedback; 

as a result, the power increased as shown in Figure 32. Once the TSVs reached to 90% open, 

reactor scram signal was sent out with a delayed time 0.08 second. That is, the negative scram 

feedback would function at 500.09 second. However, the negative reactivity feedback would be 

decreasing with time. Hence, as shown in Figure 33, the core power kept increasing until the 

scram feedback dominated the power variation at 500.14 second. 

As the dome pressure reached to 7.94 MPa, the relief valves open signal was sent out with 

delayed time 0.4 second. Hence, as shown in Figure 34, the relief valves functioned when the 

dome pressure reached to 8.1 MPa. The steam flow rate increased again. However, only six 

relief valves functioned in this hypothetical accident, the steam flow was only 0.3 times of the 

normal operating flow rates. 

From the analysis results above, it is noticed that the reactor scram system can successfully 

inhibit positive reactivity which comes from the decreasing void fraction. In addition, even though 

there are only six relief valves functioned, the dome pressure can still be reduced sufficiently. 

 
4-16

500 501 502 503 504 505

7.0

7.2

7.4

7.6

7.8

8.0

8.2

 Dome Pressure Steam Flow

Time (Second)

Dome Pressure (MPa)

-200

0

200

400

600

800

1000

1200

1400

1600

1800

 Steam Flow (kg/s)

Figure 32 Variation of Dome Pressure and Steam Flow Rate in TSVC Transient 

500 501 502 503 504 505

0

500

1000

1500

2000

2500

3000

3500

4000

Power (MW)

Time (Second)

Core Power (TSVC)

Figure 33 Variation of Core Power in TSVC Transient 

 
4-17

500 501 502 503 504 505

7.0

7.2

7.4

7.6

7.8

8.0

8.2

 Dome Pressure SRV Flow

Time (Second)

Dome Pressure (MPa)

0

20

40

60

80

100

120

Mass Flow Rate (kg/s)

Figure 34 Relationship between Dome Pressure and Rerlief Valves Flow Rate 

 
4-18

With the thermal hydraulic data results, the fuel rods properties of the TSVC transient can be 

calculated by FRAPTRAN code. As shown in Figure 35, the cladding temperature increased with 

power increasing. Due to the increasing temperature of the fuel rod, the gas inside would inflate 

and expand the cladding. On the other hand, the increasing dome pressure would squeeze the 

fuel cladding. As shown in Figure 36, at the beginning of the TSVC transient, the temperature 

increased greatly; as a result, the gap gas pressure inside cladding dominated the cladding hoop 

strain variation. After the reactor scram, the power dropped rapidly and the temperature 

decreased. The dome pressure outside cladding dominated the hoop strain variation. In addition 

to the cladding temperature and hoop strain, fuel pellet enthalpy is also an important criteria 

during the transient. Figure 37 shows that the maximum enthalpy in the TSVC transient was 

about 223512 J/kg, which was much lower than the criteria 710600 K/kg. Same as the MSIVC 

transient, the gap distance and the oxide thickness should also be checked in TSVC transient. 

Figure 38 shows that the oxide layer did not grow during the transient. There was no water-metal 

reaction in TSVC transient. Moreover, the fuel pellet surface expanded because of the increasing 

temperature at the beginning of the transient; as a result, the gap distance reduced before 

reactor scram. Once the reactor scrammed, fuel pellet shrank due to the decreasing temperature. 

The gap distance extended as shown in Figure39. Overall, the gap distance of all the positions 

were always larger than 0.06 mm, which means that there is no PCMI effect during the TSVC 

transient. 

After the general analysis, uncertainty analysis data results of fuel rod properties are shown 

below. According to the Figure 35, position Node 8 was chosen the uncertainty band of cladding 

temperature in Figure 40. This figure shows that, with the manufacturing tolerance concern, the 

cladding temperature might reach to 627K during the power increasing transient. Further, 

according to Figure 36, Figure 41 shows that the uncertainty band of cladding hoop strain at 

Node 8 would increase up to 0.002, which was much lower than the criteria 0.01. Different with 

the variation in MSIVC transient, the cladding hoop strain in TSVC transient was always larger 

than 0. The cladding was not compressed. Nonetheless, the geometric uncertainty might 

influenced the deformation mechanism. As a result, Figure 42 shows the uncertainty band of gap 

distance at Node 8. From this figure, the gap distance was at least larger than 0.048 mm. The 

PCMI effect did not happen in this transient. Figure 43 shows the uncertainty of fuel enthalpy at 

Node 8. The peak value of the fuel enthalpy was about 25400 J/kg, which was much lower than 

the criteria. From the TRACE and FRAPTRAN data results mentioned above, it can be confirmed 

that both the reactor vessel and fuel rods kept good integrity during the TSVC transient. 

 
4-19

500 501 502 503 504 505

570

575

580

585

590

595

600

605

610

Temperature (K)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Cladding Temperature (TSVC)

Figure 35 Cladding Temperature of FRAPTRAN Results in TSVC Transient 

500 501 502 503 504 505

0.00135

0.00140

0.00145

0.00150

0.00155

0.00160

0.00165

Strain

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Hoop Strain (TSVC)

Figure 36 Cladding Hoop Strain of FRAPTRAN Results in TSVC Transient 

 
4-20

500 501 502 503 504 505

80000

100000

120000

140000

160000

180000

200000

220000

Enthalpy (J/kg)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Fuel Enthalpy (TSVC)

Figure 37 Fuel Enthalpy of FRAPTRAN Results in TSVC Transient 

500 501 502 503 504 505

0.0000

0.0002

0.0004

0.0006

0.0008

0.0010

Oxide Thickness (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Inside Oxidation (TSVC)

Figure 38 Oxidation Thickness of FRAPTRAN Results in TSVC Transient 

 
4-21

500 501 502 503 504 505

0.060

0.062

0.064

0.066

0.068

0.070

0.072

0.074

0.076

0.078

Radial Gap (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Pellet-Cladding Gap (TSVC)

Figure 39 Gap Thickness of FRAPTRAN Results in TSVC Transient 

500 501 502 503 504 505

585

590

595

600

605

610

615

620

625

630

Temperature (K)

Time (Second)

 AVE MAX MIN

Cladding Temperature Uncertainty at Node 8

Figure 40 Uncertainty Band of Cladding Temperature at Node 8 in TSVC Transient 

 
4-22

500 501 502 503 504 505

0.0008

0.0010

0.0012

0.0014

0.0016

0.0018

0.0020

Strain

Time (Second)

 AVE MAX MIN

Cladding Hoop Strain Uncertainty at Node 8

Figure 41 Uncertainty Band of Cladding Hoop Strain at Node 8 in TSVC Transient 

500 501 502 503 504 505

0.050

0.055

0.060

0.065

0.070

0.075

Gap (mm)

Time (Second)

 AVE MAX MIN

Structual Radial Gap Uncertainty at Node 10

Figure 42 Uncertainty Band of Gap Thickness at Node 10 in TSVC Transient 

 
4-23

500 501 502 503 504 505

140000

160000

180000

200000

220000

240000

260000

Enthalpy (J/kg)

Time (Sencond)

 AVE MAX MIN

Average Fuel Enthalpy Uncertainty at Node 8

Figure 43 Uncertainty Band of Fuel Enthalpy at Node 8 in TSVC Transient 
4-24

4.3 Turbine Control Valves Closure (TCVC) 

Transient Assumption 

Once the turbine control valves closed, the main steam line flow decreased immediately which 

increased the dome pressure. The positive void fraction reactivity feedback performed and the 

power increased. Different from the previous two cases, the recirculation pumps trip signal and 

the reactor scram signal were sent out immediately. However, due to the signal delayed time, the 

reactor would scram 0.07 second later and the recirculation pumps would be out of service 0.14 

second later. For conservative reason, there were only 7 safety valves and 6 relief valves 

functioned in this case with signal delayed time 0.4 second. Same as previous two cases, once 

the safety/relief valves open, the dome pressure decreased and the NPP was under controlled. 

Table 6 lists the important setpoints and transient events of the TCVC hypothetical accident. 

Table 6 Event Sequence of TCVC Transient 

Time 

(sec) 

Events Notes 

0~209 Steady state Power 3030 MWt 

Feedwater flow rate 1641 kg/sec 

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 kg/sec 

Dome pressure 7.17MPa 

210 Turbine control valves 

closed 

Start of the transient; valve closure 0.15 second 

210.07 Reactor scram Signal initiated at TCVs close with delayed 0.07 

second 

211.4 Recirculation pumps trip Signal initiated at TCVs close with delayed 0.14 

second 

211.76 Safety/relief valves open 7 safety valves and 6 relief valves functioned in this 

model with delayed time 0.4 second 

215 End of analysis 

 
4-25

Analysis Data Results 

Figure 44 shows that once the turbine control valves closed, the steam flow rate dropped and the 

dome pressure increased immediately. As a result, the void fraction inside the reactor core 

decreased which caused a positive reactivity feedback. The power increased as shown in Figure 

45. Different from previous two cases, the reactor scram signal was generated at once as the 

turbine control valves closed. The power would decrease at the beginning of the transient. After 

210.5 second, the power increased because the positive void fraction reactivity feedback 

dominated the power. However, at 211.2 second, the negative scram reactivity feedback 

dominated the power again. The power dropped rapidly. After 211.76 second, the safety/relief 

valves opened and released the steam inside the reactor vessel. The pressure was under 

controlled and the steam flow rate was back to a stable value. On the other hand, the 

recirculation pump and feedwater pump trip signal would be sent out to avoid the void fraction 

increased after reactor scram. As a result, Figure 46 shows that the feedeater flow rate and the 

recirculation flow rate decreased. After 212 seconds, the feedwater flow rate kept constant 

because the water level of the reactor vessel was stable at this value. 

From those data results mentioned above, the dome pressure did not exceed the criteria 

9.58Mpa which was regulated by ASME. Further, all the inside and outside flow rates became 

stable. Kuosheng NPP was under control in the TCVC transient. 

 

 
4-26

210 211 212 213 214 215

7.2

7.4

7.6

7.8

8.0

8.2

 Dome Pressure Steam Flow

Time (Second)

Dome Pressure (MPa)

-400

-200

0

200

400

600

800

1000

1200

1400

1600

1800

2000

 Steam Flow (kg/s)

Figure 44 Variation of Dome Presssure and Steam Flow Rate in TCVC Transient 

210 211 212 213 214 215

0

500

1000

1500

2000

2500

3000

3500

4000

Power (MW)

Time (Second)

Core Power (TCVC)

Figure 45 Variation of Core Power in TCVC Transient 

 
4-27

210 211 212 213 214 215 216 217

900

1000

1100

1200

1300

1400

1500

1600

1700

Feedwater Flow (kg/s)

Time (Second)

 Feedwater Flow

Feedwater Flow (TCVC)

Figure 46 Variation of Feedwater Flow Rate in TCVC Transient 

 
4-28

With the thermal hydraulic data results, the fuel rods properties of the TCVC transient can be 

calculated by FRAPTRAN code. As shown in Figure 47, once the TCVs started to close, the 

power increased and as a result, the cladding temperature of all positions raised. However, the 

reactor scrammed immediately as the TCVs start to close. The period of power increasing was 

shorter than that of the previous two cases. Hence, the peak cladding temperature would be 

about 588K which was lower than that of previous two cases. On the other hand, because the 

cladding temperature did not increase a lot, the cladding hoop strain shrank since the beginning 

of the transient as shown in Figure 48. This variation indicates that the cladding would not 

expand and rupture but the PCMI effect might happen. As a result, the gap distance variation 

between fuel pellet and cladding should be examined. Figure 49 shows that at the beginning of 

the transient, the gap distance decreased because of the increasing power and decreasing hoop 

strain. After 211.5 second, the power dropped rapidly and hence the fuel pellet would no longer 

expand. Further, with the opening of the safety/relief valves after 211.5 second, the dome 

pressure decreased. Based on the above two points, the gap distance increased. Figure 50 

shows the variation of fuel enthalpy during the TCVC transient. Due to the power increasing 

being controlled rapidly, the fuel enthalpy did not increase a lot. The peak value of enthalpy is 

about 15500 J/kg, which is much lower than the criteria regulated by U.S. NRC. In addition, 

Figure 51 shows the variation of the oxide thickness. From this figure, it is obviously that the 

oxidation layer had not been produced. 

After the general analysis, uncertainty analysis data results of fuel rod properties are shown 

below. According to the results of Figure 47, the highest peak cladding temperature was 

appeared at position Node 4. As a result, Figure 52 shows the uncertainty band of cladding 

temperature at Node 4. With the manufacturing tolerance, the peak cladding temperature might 

reached to 598K. After the reactor scram, the power decreased and hence the deviation of the 

uncertainty band reduced. According to the results of Figure 48, Figure 53 shows the uncertainty 

band of cladding hoop strain at Node 5. From this figure, the peak cladding hoop strain would 

reach to 0.0018, which was still much lower than the criteria 0.01. Further, the negative deviation 

of hoop strain was larger than 0, which means that the cladding would not be squeezed. 

However, the fabrication uncertainty might change the variation of the fuel swelling. The PCMI 

effect would still possibly happen. As a result, according to Figure 49, Figure 54 shows the 

uncertainty band of gap distance at Node 12. From this figure, the gap distance was always 

larger than 0.056 mm. The PCMI effect would not happen in this transient. Figure 55 shows the 

uncertainty band of fuel enthalpy at position Node 8. The variation of enthalpy was very similar to 

that of cladding temperature. The larger deviation was appeared in power increasing period. 

After the reactor scram, the deviation would reduce. In spite of considering manufacturing 

tolerance, the peak fuel enthalpy did not exceed the criteria 170 cal/g. Based on these data 

results mentioned above, the structure of the power plant was safe and the fuel rods still 

maintained good integrity. 
4-29

210 211 212 213 214 215

572

574

576

578

580

582

584

586

588

590

Temperature (K)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Cladding Temperature (TCVC)

Figure 47 Cladding Temperature of FRAPTRAN Results in TCVC Transient 

210 211 212 213 214 215

0.00140

0.00142

0.00144

0.00146

0.00148

0.00150

0.00152

0.00154

Strain

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Hoop Strain (TCVC)

Figure 48 Cladding Hoop Strain of FRAPTRAN Results in TCVC Transient 

 
4-30

210 211 212 213 214 215

0.066

0.068

0.070

0.072

0.074

0.076

0.078

0.080

0.082

0.084

Radial Gap (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Pellet-Cladding Gap (TCVC)

Figure 49 Gap Thickness of FRAPTRAN Results in TCVC Transient 

210 211 212 213 214 215

110000

120000

130000

140000

150000

160000

Enthalpy (J/kg)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Average Fuel Enthalpy (TCVC)

Figure 50 Fuel Enthalpy of FRAPTRAN Results in TCVC Transient 

 
4-31

210 211 212 213 214 215

0.0000

0.0002

0.0004

0.0006

0.0008

0.0010

Oxide Thickness (mm)

Time (Second)

 Node 1

 Node 2

 Node 3

 Node 4

 Node 5

 Node 6

 Node 7

 Node 8

 Node 9

 Node 10

 Node 11

 Node 12

Cladding Inside Oxidation (TCVC)

Figure 51 Oxidation Thickness of FRAPTRAN Results in TCVC Transient 

210 212 214 216

570

575

580

585

590

595

600

Temperature (K)

Time (Second)

 AVE MAX MIN

Cladding Temperature Uncertainty at Node 4

Figure 52 Uncertainty Band of Cladding Temperature at Node 4 in TCVC Transient 

 
4-32

210 212 214 216

0.0000

0.0002

0.0004

0.0006

0.0008

0.0010

0.0012

0.0014

0.0016

0.0018

Strain

Time

 AVE MAX MIN

Cladding Hoop Strain Uncertainty at Node 5

Figure 53 Uncertainty Band of Cladding Hoop Strain at Node 5 in TCVC Transient 

210 212 214 216

0.056

0.058

0.060

0.062

0.064

0.066

0.068

0.070

0.072

0.074

Gap (mm)

Time (Second)

 AVE MAX MIN

Structual Radial Gap Uncertainty at Node 12

Figure 54 Uncertainty Band of Gap Thickness at Node 12 in TCVC Transient 

 
4-33

210 212 214 216

110000

120000

130000

140000

150000

160000

170000

Enthalpy (J/kg)

Time (Second)

 AVE MAX MIN

Average Fuel Enthalpy Uncertainty at Node 8

Figure 55 Uncertainty Band of Fuel Enthalpy at Node 8 in TCVC Transient 
4-34

4.4 Transient Comparison 

From these data results, it is obvious that the variation of power, dome pressure and fuel rods 

properties are familiar for these three overpressurization transient. However, due to the 

difference of the valves closure time, peak values of dome pressure and other fuel rods 

properties are quite different. Table 7 lists the comparison of the important parameters for these 

three hypothetical accident. From this table, due to the longer power increasing period, the range 

of power increasing in MSIVC is much larger than that in the other two cases. Further, in the 

MSIVC transient, it was assumed that the relief valves failed. There were only 11 safety valves 

functioned. As a result, the discharging of steam was less and the dome pressure of the MSIVC 

transient was larger than the other two cases. In addition, from Table 7, it should be noticed that 

the time points of peak dome pressure were later than that of the safety/relief valves opening. It 

means that the discharging rates of the steam through these valves were less than the 

generating rates of the steam inside the reactor vessel. 

For the fuel rod properties, Figure 56 shows the comparisons of peak value time of the three 

criteria. From this figure, it should be noticed that the timepoints of peak fuel properties would 

delay for few seconds after that of peak power. For example, in the TSVC and TCVC transients, 

the fuel enthalpy which was influenced mainly by the fuel temperature would reach to the peak 

value first because of the increasing reactivity and power. After that, the increasing power heated 

the gas inside the gap of the fuel rods. The gas inflated and expanded the cladding. As a result, 

the cladding hoop strain increased. Then, the gas transferred the heat to the cladding. Hence, 

the cladding temperature reached to the peak value the last. However, in the MSIVC transient, 

the dome pressure increased a lot. The gas inflation inside the gap was countervailed by the 

outside coolant pressure at the beginning of the transient. Hence, the cladding hoop strain 

reached to the peak value the last, which was different from the variation of TSVC and TCVC 

transient. 

 
4-35

Table 7 Comparison of the Important Parameters for Three Hypothetical Accident 

Events MSIVC TSVC TCVC 

Valves closure time (Sec) 3 0.243 0.16 

Maximum power (MWt) 6933 3900 3608 

Maximum power time (Sec) 3.62 1.29 1.10 

Peak value of dome pressure (MPa) 8.82 8.20 8.20 

Maximum dome pressure time (Sec) 4.77 2.27 1.90 

Reactor scram time (Sec) 3.25 0.24 0.08 

Peak cladding temperature (K) 606.70 625.88 597.43 

Maximum cladding temperature time (Sec) 4.67 1.41 1.69 

Maximum cladding hoop strain 0.00178 0.00197 0.00177 

Maximum cladding hoop strain time (Sec) 5.41 1.4 1.51 

Maximum fuel enthalpy (J/kg) 1.70E+05 2.47E+05 1.64E+05

Maximum fuel enthalpy time (Sec) 4.38 1.30 1.40 
4-36

Figure 56 Comparisons of Peak Value Time 

0 1 2 3 4 5 6

MSIVC

TSV

TCV

Comparisons of peak value time

Valve closure time (Sec) Maximum power time (Sec)

Maximum dome pressure time (Sec) Reactor scram time (Sec)
5-1 

5. CONCLUSIONS 

From these analysis data, Kuosheng NPP will be safe during the overpressurization transient 

after SPU. In spite that turbine bypass valve failed, the dome pressure would not exceed the 

criteria regulated by ASME if certain amounts of safety/relief valves opened. Further, the closure 

time of valves in this research were shorter than that in reality. The hypothetical transient 

conditions were more severe. However, the reactor scram setting and other safety mechanism 

could still maintain the integrity of the NPP and the fuel rods. On the other hand, for the 

uncertainty analysis, the manufacturing tolerance might cause 1.7% of cladding temperature 

deviation, 14% of cladding hoop strain deviation and 10% of fuel enthalpy deviation. 

Nevertheless, even concerning the uncertainty deviation, the fuel rod properties would not 

exceed the criteria. 

In this research, a method, which combines thermal hydraulic and fuel rod properties analysis, 

was successfully built up. The researchers can not only obtain the thermal hydraulic data of the 

NPP in transient, but also can observe the fuel rods properties through these thermal hydraulic 

results. Further, with the SNAP interface, power plant models of TRACE and FRAPTRAN can be 

built up more efficiently. The visible control blocks and thermal hydraulic components help 

researchers comprehend the control logic and the power plant model more easily. However, due 

to the restriction of the program, data transfer from TRACE to FRAPTRAN was not automatically 

in this research. It needs a lot of human resources to transfer thermal hydraulic data results to 

the FRAPTRAN input deck. Hope that in the near future, a program which can transfer data 

results from TRACE to FRAPTRAN codes will be developed. 

With the advantages of DAKOTA code setting in SNAP interface, an uncertainty analysis 

methodology was also built up in this research. With this method, the manufacturing tolerance 

influence for the fuel rods in transient could be concerned. However, due to the restriction of data 

collection, there were only six fuel rod geometric parameters discussed in this research. Hope 

that in the researches to come, the fabrication uncertainty information can be collected more 

completely. Even more, to gather a more convincing data results, the uncertainty of thermal 

hydraulic components should be collected and concerned in the future researches. 

 
5-1 

 
6-1 

6. REFERENCES 

1. Taiwan Power Company, Final Safety Analysis Report for Kuosheng Nuclear Power Station 

Units 1&2 (FSAR), Taiwan Power Company, Republic of China (Taiwan), 2001. 

2. Taiwan Power Company, BWR-6 MARK-III Containment of Kuosheng NPS, Taiwan Power 

Company, Republic of China (Taiwan), 1995. 

3. Ken Jones, TRACE V5.0 user’s manual, U.S. NRC, Washington DC, 2010 

4. K.J. Geelhood, W.G.L., C.E. Beyer, FRAPTRAN 1.4: Integral Assessment, Pacific 

Northwest National Laboratory, Richland, 2011. 

5. K.J. Geelhood, W.G.L., C.E. Beyer, FRAPCON-3.4: Integral Assessment, Pacific Northwest 

National Laboratory, Richland, 2011. 

6. Brian M. Adams et al, Dakota, A Multilevel Parallel Object-Oriented Framework for Design 

Optimization, Parameter Estimation, Uncertainty Quantification, and Sensitivity Analysis: 

Version 6.0 User's Manual, Sandia National Laboratories, Albuquerque, 2014. 

7. H.C. Chang, W.K. Lin, W.Y. Lin, C. Shih, J.R. Wang, H.T. Lin, The Turbine Trip without 

Bypass Analysis of Kuosheng BWR/6 Using TRACE/FRAPCON/FRPATRAN, ANS 2014 

annual meeting, Reno, Neveda, 2014. 

8. S.W. Chen, H.C. Chang, W.Y. Li, W.K. Lin, C. Shih, J. R. Wang, H.T. Lin, Fuel Rod 

Uncertainty Analysis of Kuosheng NPP with TRACE/FRAPTRAN/DAKOTA codes in SNAP 

interface, ICAPP 2015, Nice, France, 2015. 

9. K.J. Geelhood, W.G. Luscher, C.E. Beyer, D.J. Senor, and M.E. Cunningham, D.D. Lanning, 

H.E. Adkins, Predictive Bias and Sensitivity in NRC Fuel Performance Codes, 

NUREG/CR-7001, 2009. 







NUREG/IA-0465

Transient for Kuosheng Power Plant with TRACE/FRAPTRAN/DAKOTA 

Fuel Rod Performance Uncertainty Analysis During Overpressurization 

Codes in SNAP Interface