1.
SNERDI Internship Final Report
NERS 499 ­ Professor Lumin Wang
Sebastian Clavijo
8/14/2015

2. 1.0 Introduction
The object of the 6­week internship was to learn more about the thermal hydraulic and
safety aspects of PWRs, more specifically the AP1000. This work was conducted under the
supervision of Ms. Zhen Xu from Shanghai Nuclear Engineering Research and Design Institute
and Professor Wang from the University of Michigan.
The following is the typed schedule for the internship, assigned and proposed by Ms.Zhen Xu:
Date Job content
July.1​st​
–July.10​th
Read and understand the section 1.1、1.2、4.4、and chapter 15 of DCD of
AP1000.
July.10​st​
–July.19​th
Read and understand the section 7 in chapter 4 of volume III and section 5 in
Chapter 5 of volume II in document of URD, summarize the requirements
relate to core design and accident analysis.
July.20​st​
–July.24​th
Gather the latest international laws and regulations related with the thermal
hydraulic analysis of PWR. Then, search for the document about the EPR, and
finish a report comparing the characteristic of ESF of EPR and AP1000.
July.25​st​
­Aug.6​th
Calculate the cases of “Malfunction of feedwater system which cause the
decrease of feedwater temperature” and “RCS boron concentration balance
design transient”.
Aug.7​th​
­Aug.12​th
Accomplish a report which summarizes the work and results during the
internship of this two months.
This report is an all encompassing report that combines the previous reports and the
recent calculations that were finished on the 12​th​
of August. Upon completion, this report will be
turned as my semester assignment for NERS 499 under Professor Wang.
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3. 2.0 AP1000
2.1 Vol. 3 Sec.7 Chpt. 4 – Reactor Systems
2.1.1 PWR Core
The basic core design for a PWR must fulfill proper design expectations while trying to be
economically feasible, as well as being within engineering safety limits. A core design consists
of fuel assemblies, fuel rods, control rods, reactivity control system, neutron system and core
instrumentation.
The major functions that the PWR core accomplish are:
● Fuel systems that provides adequate power and cycle energy production.
● Fission products and fuel are contained within cladding.
● The control of excess reactivity in the core.
◦ Control rods for short term reactivity adjustment.
◦ Dissolved poison concentration in coolant for long term reactivity adjustment.
▪ e.i. Fuel depletion and Reactor cool­down
Assembly layout and loading pattern are determined by obtaining a realistic minimum level of
neutron leakage at the perimeter of the assembly, and for a negative moderator temperature
coefficient at critical level operation. Control rods should be easily inserted, and capable of
properly initiating scram. Lastly, the core should be designed to not undergo bulk boiling in the
hot channel under standard operating conditions. Overall, the material used in fuel assembly
construction should be corrosive resistant to prevent fretting damage and damage to the stainless
steel springs. Therefore, the material used must be mechanically sound to hold the weight of the
assembly and not corrode in the core environment.
2.1.2. PWR Material
The fuel system in a PWR core should maintain a high neutron economy, but should be
mechanically sound and corrosive resistant. A common cladding material is Zircaloy due to its
high corrosive resistance. Unfortunately, it has a higher tendency of hydrogen formation causing
a plethora of other problems. Ultimately, standard designs limit hydrogen formation to less than
2ppm on a Uranium weight basis. For control rods, the material composition changes depending
on the exposure. For example, in the AP1000 design, Wet Annular Burnable Absorbers
(Ag­In­Cd) and Integrate Fuel Burnable Absorbers ( ) are used for high and low exposureCB4
settings respectively. IFBAs coat the surface of the fuel rods, while the WABAs take up open
spaces in the guide tubes.
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4. 2.2. Vol 2. Sec. 5. Chpt. 5 – PWR Core Damage Prevention Requirements
2.2.2. General Requirements
Analysis of core coolant inventory control, and decay heat removal system using the following
core safety features:
● Residual Heat Removal
● Emergency Feedwater
● Safety Injection System
● Safety Depressurization and Vent System
Further analysis of the safety system will include the use of motor driven and turbine driven
pumps, as well as the heat exchanger used in the RHR and valve systems. All these system
should be readily available for control and monitoring during start­up, steady state and cool
down operations.
2.2.3. Residual Heat Removal System (RHR)
To begin, the RHR system does a variety of different functions from decay heat removal
along with RCS heat removal to transferring water from refueling cavity, removing heat from the
IRWST during feed and bleed, lower the temperature during overpressure conditions and as a
potential back up to the CSS during a long term LOCA.
The RHR system in a PWR takes water from one or two RCS hot legs, cools it, and
pumps it back to the cold legs or core flooding tank nozzles. The suction and discharge lines for
the RHR pumps have valving to provide reasonable assurance that the low­pressure RHR system
is isolated from the RCS when the RCS pressure is greater than the RHR system design pressure.
Relief valves are provided to protect the RHR system from an overpressure condition, although
the relief capability is not sufficient to protect the RHR system from an overpressure condition if
isolation valves are open when the RCS pressure is significantly greater than the RHR design
pressure.
In PWRs, the RHR system is also used to fill, drain, and remove heat from the refueling
water cavity during refueling operations, to circulate coolant through the core during plant
startup before RCS pump operation, and in some to provide an auxiliary pressurizer spray.
Like many safety features, redundancy is implemented to prevent complete failure of
RHR function. Therefore, the RHR has two independent systems as a precaution. Furthermore,
in the case of low RCS water levels, the RCS is actually designed to never go below the
minimum water level for RHR operation.
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5. 2.2.4. Emergency Feedwater System (EFW)
In the case of loss of feedwater, the EFW will be triggered and will begin inserting
feedwater at a systematic flow rate into the Steam Generator. The flow rate will depend on the
following:
● Potential spills in Steam Generator
● Status of Reactor Coolant Pumps
● Single Failure
There is a risk of overcooling RCS, overfilling the Steam Generator, increasing the
pressure in the containment and ultimately causing an unnecessary diesel generator capacity.
It is important to note, that the EFW can also be initialized manually in the case of SG
levels being deemed too low. The inserted water is standard secondary water that is pumped
through either electric or steam turbine pump. The electric pump allows for more reliability
whereas the steam turbine driven pump allows for decay heat removal during station blackout.
For the steam turbine drive pumps, the pipes are designed to prevent the condensation of the
steam along the inside walls of the pipe. Additionally, it should be noted that there is standard
testing on pumps, valves and instrumentation should be regularly monitored as to prevent any
abnormal incidents.
2.2.5. Safety Injection System (SIS)
The overall function of the SIS, is to maintain acceptable boron concentration in the
coolant. The SIS injects boron during cold core shutdown and during main steam line breaks.
Furthermore, it will also used to adjust excessive boron levels seen in RCS levels in cold leg
LOCAs, to prevent possible boron precipitation. During safe cool shutdowns, the injection of
boron will taken from the IRWST to maintain safe shut down by keeping the reactor at
subcritical levels. In the case of a main steam line break, the SIS will take water from the
refueling storage tank and inject the coolant to maintain an dose limit within 10 CFR 20 limits. It
should be noted, that in emergencies, the SIS is capable of providing cooling during a bleed and
feed operations.
2.2.6. Safety Depressurization and Vent System
The SDVS is the system comprised of the valves and piping that connects and allows for
the flow from the pressurized steam space and reactor upper head to the IWRST. The SDVS is a
general system that is used for the depressurization and venting of some high pressure systems. It
is designed to be a safe and ultimately useful tool to eliminate high pressure in the RCS. In the
case of a transient loss of feedwater (TLOFW) the SDVS will trigger and will allow for core
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6. cooling using a single bleed path in a 2/4 pump system and preventing immediate fuel
uncovering.
The SIS is designed to operate in any condition, excluding start up. It is used mainly in
standard operation and in upset or accident scenarios. Furthermore, the valves in the SDVS
should be able to operate in a blackout, allowing for bleeding. Ultimately, the SDVS is not
automatically actuated, but rather manually actuated from the control system.
3.0 Evolutionary Power Reactor
3.1 Introduction
The EPR is a Franco­German designed reactor from the companies of AREVA, EDF and
Siemens. With multiple ERPs being built in many countries, and with many other countries
verifying the design, the EPR is one of the most popular modern reactor designs in the world
today. The EPR is one of the biggest reactors under construction, with an approximate 1650
MWe and a four loop safety system making it a very interesting and feasible reactor under
construction today.
3.2 Design Benefits
Arguably the greatest advantage of the EPR design is the four loop safety system. The concept
behind this design was to create a safety system with vast amounts of redundancy. This concept
ultimately brought about the idea of have four independent emergency cooling systems. Each of
the loops provides adequate decay heat removal for up to 1­3 years after reactor shutdown.
Compared to many PWR, it has reduced fuel consumption allowing for easier system
maintenance. Additionally, there is a slightly higher thermal efficiency with means a reduction in
water usage, but considering the four cooling loops, the amount of water used is still relatively
similar to other PWRs. A very useful and distinct benefit to other PWRs is the use of hydrogen
re­combiners that will minimize the possibility of hydrogen explosions by a drastic margin.
3.2.1 Comparison to the AP1000
The EPR boasts some very impressive safety features, but does have some very important
differences compared to the AP1000. In a safety standpoint, the core accident frequency of the
EPR is approximately 3.7 *10^­7 (Rx­yr) while the AP1000 boasts a core accident frequency of
~1*10^­6 (Rx­yr). Although both designs boast very low frequencies, it should be noted that the
NRC’s frequency threshold is set at 10^­4 (Rx­yr)
The AP1000 operates with two loops, along with a PRHR heat exchanger which allows for
natural circulation heat removal. The Passive Safety Injection Systems includes Core Makeup
Tanks, Accumulators, IRWST injection system and a means of automatic RCS depressurization.
In the case of core meltdown, the EPR uses the concept of a ‘Core Catcher’ while the AP1000
keeps the corium in In­vessel retention.
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7.
4.0 Case 1: Malfunction in Feedwater System
The reduction of the feedwater temperature causes the decrease of the RCS temperature which
cause reactivity insertion into the core. The case can occur at either at Hot Zero Power (HZP) or
during Hot Full Power (HFP). The HZP case is bound by the HFP, because the load and
feedwater flow is lower and the variety rate of energy is less at HZP. Therefore, it is only
necessary to consider the reduction of the feedwater temperature at HFP.A simple case was
generated to show the difference in enthalpy to analyze the excessive load increase in secondary
steam flow.
The reduction of feedwater enthalpy could be compared with the case of excessive load
increase in secondary steam flow. If the reduction of feedwater enthalpy in the case of
malfuction of feedwater system(case A) is less than the reduction of feedwater enthalpy in the
case of excessive load increase in secondary steam flow(case B), The case A is bounded by the
case B. The reduction of feedwater enthalpy was calculated in the case of excessive load increase
in secondary steam flow(case B) using the initial conditions in Appendix A.1.
The following equation was used to determine Heq:
Eq. 4.1Sec. Steam Enthalpy ec.Feedwater Enth.) .1 Sec.Steam Enth. eq) .0 ( − S * 1 = ( − H * 1
The calculated enthalpy was 342 btu/lbm, demonstrating a 78 btu/lbm enthalpy reduction in the
feedwater.
5.0 Case 2: RCS Boron Concentration Design Transient
In the case of a varied boron concentration in the RCS, the spray is tripped and begins to
process to balance out the boron concentration between the RCS loop and the Pressurizer.
During this scenario, the standby heater causes an increase in pressure inside the Pressurizer. The
pressure then reaches a set minimum pressure level in which the spray is tripped therefore
maintaining the pressure inside the Pressurizer at a level proportional to the spray rate. After
continued operation of the spray, the boron concentration in the RCS loop and the Pressurizer
should level off. This scenario can apply to lower or higher levels of boron concentration.
For case 2, the assumption that the secondary side was not affected by the transient was
made. The RCS pressure will vary with the pressure in the Pressurizer as the coolant flowing
from the Pressurizer heats the nozzle of the surge line connected to the RCS hot leg.
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8. ρ h )P = V˙ * * ( f − hcl Eq 5.1
Where P is power, is volumetric flow rate, ρ is the density of the fluid in the spray, is the V˙ hcl
enthalpy of the fluid in the cold leg and is the unknown variable.hf
Next, the second equation was used to determine the flow rate with the determined pressure of
the Pressurizer. Knowing that the pressurizer flow rate is proportional to the maximum and
minimum pressure values seen in Appendix B.1. The equation simply becomes the following:
​ ​Eq 5.2P −Pf l
V −V˙
f
˙
l
= P −Pm l
V −V˙ m
˙
l
Where stands for the volumetric flow rate of the fluid relative to the pressure . The ‘m’ V˙
f Pf
and ‘l’ subscripts indicate the maximum and minimum, for the volumetric flow rate and the
pressure.
This part of the calculation is where I had problems to resolve the correct values due to lack of
resources provided by SNERDI. Based on rough estimations from the NIST tables, and the
conservation of energy and mass equations, I was able to determine an estimated saturation of
pressure of 15.719 MPa with a volumetric flow rate of 13.844 with the heater on /hr m3
(1230kW). From this, the maximum temperature reached in the Pressurizer was approximately
345 C based on the saturation pressure of 15.719 MPa. Furthermore, it was determined that the
estimated saturated enthalpy was ~1230 kJ/kg. The length of time of the entire transient case
should take ~120 minutes based on the total volume of the Pressurizer and the flow rate out of it
(which is the same as the flow rate into the Pressurizer). That’s the length of time before the
water in the Pressurizer is completely replaced and the RCS’s boron concentration levels are
back at nominal conditions.
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9.
6.0 Appendix
A.1 Initial Conditions for the Feedwater System Transiet
● The secondary side steam enthalpy is H​steam​= 1200.0 Btu/lbm after the accident;
● The secondary side feedwater enthalpy is H​feed​=420.0 Btu/lbm after the accident;
● The normalized secondary side steam flow is R=1.1 after the accident;
● The secondary side steam enthalpy maintains the same before and after the accident;
● The feedwater flow maintains the same before and after the accident.
B.1 The initial conditions for the Boron Conectration RCS Transient:
● Operating Power of Pressurizer standby heater: 1230 kW
● Initial spray rate pressure: 15.686 MPa
● Maximum spray rate pressure: 16.065 MPa
● Maximum spray rate: 159 m​3​
/h
● RCS cold leg temperature: 279.4 C
● RCS hot leg temperature: 322.3 C
● Pressurizer norminal water volume: 28.32 m​3
● Setpoint value of low temperature alarm of spray pipe: 262.8 C
● Total volume of spray line: 0.350 m​3
● Norminal Pressurizer pressure: 15.5MPa
The main assumptions for the case were as follows:
1） The parameters on the secondary side are not affected by the transient. The RCS
pressure vary with the pressurizer pressure, as the fluid flow from the pressurizer, the
heat impact may be caused to the nozzle of pressurizer surge line connect to the RCS
hot leg;
2） The pressurizer water level is maintained stably, the fluid flow in the surge line equal to
the spray flow.
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10. ​ C.1 Resources
● “1.1, 1.2, 4.4, 15." ​AP1000 Design Control Documents​. Rev. 19 ed. 6. Print.
● "Design Certification Application Review ­ U.S. EPR." ​NRC:​. Web. 17 Aug. 2015.
● "International Atomic Energy Agency | Atoms For Peace." Web. 17 Aug. 2015.
● "NRC: Home Page." Web. 17 Aug. 2015.
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