final paper rev1

Activation Concern for I2S-LWR In-Vessel Primary Heat Exchangers

William Bryans, Adnan Hashim, Joshua McCann, Ayuko Morikawa, Syfuddin Rashid

Nuclear and Radiological EngineeringGeorgia Institute of Technology

770 State St, Boggs Bldg.Atlanta, GA 30332-0745

The I2S-LWR reactor incorporates integrated heat exchangers and other secondary components built within the pressure vessel to minimize total volume and increase the ease of transportability. However, due to this integrated heat exchanger design, the possibility of heat exchanger material activation may be a concern. Thus the objective of this investigation is to alter heat exchanger height to evaluate a possible method of reducing total heat exchanger activation, which in modern industry could relate to reducing worker radiation exposure and future unforeseeable costs. Using MCNP, the study analyzes typical neutron flux values generated from the reactor core that strike the heat exchanger surfaces. Once an exact value is calculated, it becomes possible to utilize this model as a foundation to evaluate other variables within the pressure vessel.

Key Words: Heat Exchanger, Neutron Activation, MCNP Model, I2S-LWR

1. INTRODUCTION

1.1 Literature Review

1.1.1. Material Research

Numerous articles were researched and studied to learn more information on the design topic. The articles found gave a better understanding on materials, shielding, MCNP coding, and fouling/plating. The article that was most helpful for material selections was “Experimental neutron attenuation measurements in possible fast reactor shield materials” by Mawutorli Nyarku, Ramanthapura S. Keshavamurthy, Venkata D. Subrmanian, Adish Haridas, and Eric T. Glover. This article was very helpful because it discussed the best shielding materials for neutron attenuation. The authors researched whether Ferro-tungsten or mild steels would be better at attenuating epithermal and fast neutrons. Results showed that even though mild steels work exceptionally well at attenuating neutrons, Ferro-tungsten was found to be a more effective neutron attenuator. Another article that was beneficial with shielding and materials was “Shielding experiments for optimization of shield materials in fast reactor using SSNTDs” by R V Kolekar, R Kumar, and DN Sharma. By using activation techniques and solid state nuclear track detectors the authors researched neutron transport through several shield materials. Stainless steel-316, borated graphite, and sodium were the three single shield materials used during the experiment. Results showed stainless steel-316 caused more backscattering than both the borated graphite and sodium.

NRE 4232 Nuclear and Radiological Engineering p. 1/29 Spring 2015

X1. Heat Exchanger Activation

1.1.2. Fouling and Plating Concerns

After brief discussion with Dr. Petrovic, the possibility of having the possibility of fouling and plating within the heat exchangers became a concern in the overall analysis. Fouling typically is formed from residual micro-particle deposition within the heat exchangers whereas plating occurs when chemicals are inserted into the system to remove said microparticles. These chemicals eventually lead to carbonate based buildup, examples being compounds such as calcium carbonate and boron carbonate (Cite “alaquainc” link). After a short period of investigation into this variable of concern and acquiring a better understanding, the issue was presented to the experts at the Waterford 3 plant. Though the conversation that took place over the topic was relatively brief, the answer provided by Waterford 3 was considered a reliable source. A more detailed description entailing the conversation follows below.

1.1.3. Discussion with Waterford 3 Nuclear Power Plant, Kenner, Louisiana

In order to acquire a better understanding of the general structure, function, operation, and composition of the primary coolant systems that directly worked in tandem with the reactor, the Waterford 3 nuclear plant was contacted. With the assistance of Pamela Hernandez, Waterford 3’s Senior Nuclear Engineer Manager, and Keith Kunkel, Waterford 3’s Heat Systems Manager, significant documentation involving operation and construction of the reactor systems as well as general explanations were acquired. Direct copies of the emails as well as the documents provided by Waterford 3 are available for inspection in the appendices.

The documents displayed values including typical heat exchanger, pressure vessel, and biological shield dimensions and specifications. The documents also gave general systems operation and flow systems found within the primary coolant loop. Both Keith as well as Pamela directly answered concerns involving buildup and fouling. As quoted by Mr. Kunkel, “We get a buildup of corrosion products on the secondary side of the steam generator. The primary side is all stainless steel and remains clean.” These statements alongside the documents provided remain to be the most significant contributions given by Waterford 3 Staff.

1.1.4. Fiscal Correlation to Possible Radiation Worker Dose

The overall scope of the investigation is to ascertain the total activation in the primary heat exchangers

from the core via neutron flux. The application of this investigation in the end is to utilize this activation value as an accurate foundation to acquire possible future dose deposition to radiation workers that may interact with the heat exchangers throughout the timespan of its operation. This activation value may also be used as a fundamental value for other investigative analysis’ involving heat exchanger impacts outside the field of dose study. However, in order to accurately relate the heat exchangers’ activation, possible deposition, and its possible risk magnitude to workers, a brief analysis must be performed on the direct correlation between the U.S dollar and the unit of dose utilized by plants for cumulative dose deposition: the person-rem. The person-rem unit accounts for the total dose in rem acquired by all workers working within a similar range of operations in a set period of time and area of operation. Finding this correlation deemed to be a relatively difficult task as the underlying philosophy of correlating life to a monetary material value is one of much debate and discord. However, documents provided by Dr. Bojan Petrovic, an advisor provided the short and to the point answers that were sought. A specific document titled “Reassessment of NRC’s Dollar Per Person-Rem Conversion Factor Policy” created by the Division of Regulatory Applications, US Regulatory Commission provided the official value allotted to financing the compensation rate for injury or suffering accrued by personnel exposed to radiation doses. The value set in 1995 was 2000 U.S dollars. Forward inflated this value can be related to



approximately 3080 U.S dollars present day of this paper (April 18 th, 2015) (Division of Regulatory Applications, 1995).

1.1.5. Similar MCNP Project

One of the articles that influenced some of the design decisions of this project was the “PWR Facility Dose Modeling Using MCNP5 and the CADIS/ADVANTG Variance-Reduction Methodology” performed at Oak Ridge National Laboratory by E. Blakeman et al. In this study, the dose from a PWR core was measured across the plant facilities at different reactor conditions. Several ideas were taken from this model such as using a watt fission spectrum for the source neutrons, homogenizing the core, using a mesh tally, and variance reduction techniques.

1.1.6. CAD to MCNP via STEP File

Given the intricate nature of the I2S-LWR design, considerable thought was given in finding an efficient yet effective model to accurately represent the neutron activation experienced by the PHXRs. Given the primary language to code these types of activations and reactions was through MCNP, it became vital to begin coding a proper INP file to represent the reactor. Unfortunately, the MCNP language created by Los Alamos National Laboratory in 1957 has become fairly antiquated, even given the major improvements made upon it. This has led to the INP files being fairly complicated to understand and often error prone as well. Given the small base of MCNP users, finding outside resources for help would also prove to be difficult. Thus a natural cross road was reached, does one learn MCNP code and create the INP file through trial and error, does one find an intermediary form of coding/design and then convert to MCNP, or does one disregard MCNP entirely.

The third option was quickly disregarded for the reasons mentioned above in that MCNP was the foremost language in describing neutron particle interactions. The second option was given considerable thought based off research done in converting CAD files to MCNP files through intermediary STEP files (Zhou 2014). The reason for the intermediary comes from the research community’s desire to utilize a CAD model data exchange that was leveraged upon neutral files. With that consensus reached, Zhou and his team began developing complex algorithms to convert these intermediary files into well-written and clean MCNP code. It is important to remember, while much information can be gleaned from the CAD models, the data specifications in block 3 of standard MCNP cannot. Zhou’s team ultimately ended their research with improved algorithms in two dimensions from their previous research, but still look to improve their models in the future.

Given the promise that these types of algorithms were showing, more research was done into software and code that could convert more manageable programming languages into MCNP INP files. It was quickly discovered that these models outstripped the budget by magnitudes of over 3 and 4. Given that time was also a limited resource, the option to find conversion codes was quickly canned, and writing the code by hand was settled upon.

1.1.7. MCNP Lattice Structure

The majority of the component research was done using the resources given by Dr. Petrovic. This consisted of a series of reports regarding the I2S-LWR reactor written for the Interinstitutional Committee for Academic Program Planning (ICAPP). In particular, a report titled “Integral Inherently Safe Light Water Reactor (I2S-LWR) Concept: Integral Vessel Layout” written by Matthew J Memmott of Westinghouse Electric Company, Matthew Marchese, and Bojan Petrovic was heavily used to gain knowledge of other components outside of the primary heat exchanger that would be included in the MCNP model. This report outlines the



configuration of the primary system and describes the design and function of all relevant components within it. Additionally, it mentions alternative configurations for the I2S-LWR primary system. The physical dimensions for components such as the primary heat exchanger, pressure vessel, core barrel, and alignment plates used in the MCNP model was found in this report and played a crucial role in creating a model that was scaled accurately with the actual reactor design. Some component dimensions,such as upper barrel diameter, were subject to change as the reactor design was updated. These dimensions were modified per discussion with Dr. Petrovic.

2. THEORY

2.1. Material Activation Concern in PHXR

The core of the investigation revolves around ascertaining what materials within the heat exchangers are

activated. Once neutron flux is deduced, neutrons traveling with a spectrum of energies strike the heat exchanger surface with the possibility of activating the material struck. Depending on the material, the type of radiation emitted can vary, with some reactions being more likely than others. This also means that the irradiated material decays and forms different isotopes, which in turn, could also decay and release radiation. An isotope of concern in this case is Iron-58. Iron, when irradiated by neutrons, will form Cobalt-59, which eventually, will also absorb another neutron and form Cobalt-60m, a metastable isotope that quickly decays to Cobalt-60. Cobalt-60 is also radioactive, but has a longer half-life than Cobalt-60m. However, Cobalt-60 emits a form of radiation that is of concern for workers that may interact with the material irradiated. Thus it was imperative to closely study the decay chains that this iron isotope can create. Other isotopes as of now do not seem to be a significant source of hazardous radiation due to extremely short half-lives. Below is a diagram depicting the decay chain of Cobalt-60m to its base form of Nickel (Cite: Wiki decay page and image author for this and image).

Table I. Decay Products of Concern


Fe-58 Co-59 Co-60m Co-60

Gamma or Beta E (MeV) - - .05859 Beta .31, 1.48 Betas

Half-life - - 10.467 mins 5.272 years

Probability σ= 1.3 barns σ = 21.2 barns 100% 98.88%, .12%

Decay or Neutron absorbed N N Decay Decay


Figure 1. Decay Chain Diagram of Daughter Isotopes of Neutron Activation of Iron-58

2.2. Flux Calculation

The current MCNP code utilizes a simplified fuel core that has been homogenized by weight percent of

materials found within it. However, due to the decision to not utilize a fission k-code to emulate an actual core, it was a necessity to ascertain the average fluence exiting the core. Using values provided from previous I2S-LWR groups, and some basic reactor physics, the average flux value was found. First, the total thermal output of 3000 MWth was converted to joules per second, which in turn was converted to MeV per second. It is also known that typical Uranium-235 fission reactions produce approximately 200 MeV/fission. Each fission is induced by a neutron thus a value of fissions per second is found by dividing the MeV/s power by the MeV/fissions. However, each fission neutron creates on average 2.76 neutrons per reaction. Thus by multiplying the fissions per second value by the 2.76 neutrons per fission, the final average neutron fluence can be calculated (Cite:Team G2, 2014 Burnable Absorbers). Below is the step-by-step calculation.

1 W=1 Js (1)

1.609 x10−13 JMeV

∧200 MeVfission (2)

Thus: 3000 x 106 J

s

1.609 x 10−13 JMeV

∙ 1

200 MeVFission

∙2.77 neutronsfission (3)

¿2.57303 x1020 neutronss (In simulated core) (4)



2.3. Heat Exchanger Design

The primary heat exchangers that are used in the I2S-LWR reactor design are manufactured by Heatric and are called printed circuit heat exchangers (PCHEs). These heat exchangers are mainly composed of Stainless Steel 304, and utilize a series of stacked cross flow plates that have coolant flowing across them in tiny channels. The primary and secondary coolants flow in alternating plates throughout the length of the heat exchanger. Each plate is 2 mm thick with a 350 mm x 600 mm base, and has 150 small channels etched parallel to the width of the plate. The channels are 1 mm wide by 1 mm deep and have a pitch of 20 mm. In the plates containing the primary coolant, the fluid enters the plate on its longer side and travels across the 350 mm width of the plate. On the plates containing the secondary coolant, the fluid enters on the short side of the plate through a 100 mm cut in the fins, then flows parallel to the width of the plate as well. The plates are stacked vertically on top of each other to form blocks that are 600 mm in height. Ten of these blocks are then stacked vertically to form the full heat exchanger, which is 600 cm in length and has a 100 cm by 50 cm base. There are a total of 8 primary heat exchangers that are positioned in a circular pattern above the reactor core (Memmott, Marchese, Petrovic 2014).

The primary coolant enters the heat exchanger assembly through a rectangular channel cut into the upper core alignment plate. This channel feeds into a vertical duct that is 980 mm by 186 mm at its highest point and becomes smaller towards the bottom of the duct. The secondary coolant enters the heat exchangers through a single 150 mm diameter pipe that feeds into two primary heat exchangers. This flow then exits into a 190 mm pipe before it recombines into a 150 mm diameter pipe under the heat exchanger (Memmott, Marchese, Petrovic 2014).



Figure 2. Outer View of Printed Circuit Heat Exchanger

3. MATERIAL CONSTRAINTS

3.1. Material of Heat Exchanger/Reflector

The first and perhaps the most important aspect in designing any component inside a nuclear reactor is the type of material used. Various material properties must be considered before choosing a material to be implemented inside a reactor. The material must be corrosion resistant and absolutely cannot be prone to activation. Through extensive literature research on materials used in nuclear reactors a list of approximately ten materials was created for possible materials to use for the PCHE and the reflector. Then further research on the lists material properties was conducted until the list was narrowed to three materials. The table below shows the three materials along with important material properties.

Table II. Material Properties



SS-304 Boron Carbide Ferro-Tungsten

Cost ($/kg) 2.84 15 34.83

Melting T (C) 1425 2489.85 1650-2100

Tensile Strength (MPa) 505 261-569 690-3000

Density (kg/m3) 8000 2300 19300

Corrosion Resistance Yes Yes Yes

Boron carbide and Ferro-tungsten both have better melting temperatures and tensile strength than stainless steel-304. However, the costs of boron carbide and Ferro-tungsten are significantly higher in comparison to stainless steel-304. Additionally, Ferro-tungsten and boron carbide are both currently not approved by the NRC, whereas stainless steel-304 is approved by the NRC. Due to stainless steel-304 being the only material of the three approved by the NRC, it was concluded the PCHE and the reflector would be made out of stainless steel-304. It is corrosion resistant, has a high melting temperature and tensile strength, and is relatively low in price.

3.2. Incorporation of a Standalone Shielding Structure

During the preliminary analysis stage of the project, the option of including a standalone shield structure between the PHXR’s and the core was considered. Theoretically this would have been the most effective method of reducing activation to the PHXR by reducing neutron flux directly. A plethora of design options were considered, ranging from a diagonally slanted “wall” between the PHXR and the core to reflect neutrons, to “plating” the PHXR in carbonate based compound to absorb neutrons. Materials considered included Ferro Tungsten, SS-304, Boron Carbide, and a Silicon Dioxide based experimental compound. The experimental compound was intended to act like a neutron absorber by being plated on the surface of the PHXR’s. Ferro-Tungsten and Boron Carbide would have been plated on top of a steel wall to serve as the standalone shield. Below is a table of the materials aforementioned and some others not mentioned alongside a few characteristics analyzed to select the most effective shield:

Table III. Properties of Possible Materials for Standalone Shield

SS-304 Boron Carbide Ferro-Tungsten Mn SiO2 Cu

Melting Temp (K) 1425 2489.25 2100 1246 1600 1085

Cost ($/kg) 2.84 15.00 2.15 34.83 1950.0 5.77

Tensile Strength (MPa @ 300C) 505 269-569 690-3000 7.21 - 220

Corrosion (Y/N) Yes Yes Yes Yes NA No

Density (kg/ 8000 2300 19300 8650 2650 8960



However, after acquiring a better understanding of the internals of the reactor vessel and its plethora of components, it was concluded that the best option in reality was to not include a standalone shield or plating of any sort. Due to the costs of acquiring and constructing the aforementioned materials into a usable format, a lack of true free volume within the pressure vessel to place the shield, and the multitude of unknown safety issues that are nearly uncorrectable, it was concluded to forego the standalone shield option in the final design for the project.

4. COST ANALYSIS

4.1. Material Costs The cost analysis of increasing the height of the reflector and axial location of the primary heat exchanger units began with the cost of possible materials. The list of researched materials was narrowed down to three following examination of material properties and cost of material. The three materials were Boron Carbide, Ferro-Tungsten, and Stainless Steel 304. Table 2 below shows materials and their associated cost per kilogram. Ferro-Tungsten is 15 dollars per kilogram, Boron Carbide is 34.83 dollars per kilogram, and Stainless Steel 304 is 2.84 dollars per kilogram. The price of the material is subject to changing over time due market prices and availability. Numerically the price for Stainless Steel 304 appears to be significantly lower than Boron Carbide and Ferro-Tungsten. However, relative to the overall cost of manufacturing the reflector and primary heat exchangers, the material is essentially negligible.

Table IV: Cost of Materials

Material Cost per Kilogram

Boron Carbide 34.83

Ferro-Tungsten 15

Stainless Steel 304 2.84

4.2. Cost of Reflector The reflector was decided to be composed of 100% Stainless Steel 304. Previously knowing the I2S-LWR’s reflector height being 4.058 meters and the thickness being .165 meters the volume of the reflector was calculated to be 6,384,171.65 cm3. Having the volume and knowing the density of Stainless Steel 304 being 7.94 g/cm3, the weight the reflector was calculated to be 50.69 tons. The price of Stainless Steel 304 per ton is 2,978. Therefore the total cost of I2S-LWR’s reflector is 150, 954.82 dollars. Every additional cm added on the height of the reflector would cost approximately 372 dollars.

5. MODELING

5.1. Introduction



The primary expenditure of time in this senior design experiment came from creating an accurate model in MCNP. To accomplish this, the method of approach decided upon was an incremental addition method. The initial thought had been to create the fully-fledged model immediately and leave as much time for testing as possible. However, given the complexity of MCNP and how error prone the code can become, the incremental method was selected as it allowed for testing throughout the entire process.

In the following analysis an initial model was created and will be explained in detail. From there, components such as the reflector, pressure vessel, core barrel, etc. are added and explained in detailed with special attention given to composition, assumptions made to create simplifications in the code, and justification for dimensions utilized. A brief disclaimer: the numbers used are the most accurate to date, however, as this is a working model, dimensions can and will change. During our examination period, a few dimensions were changed in the reactor and these had to be accounted for in the final design.

5.2. Initial Model

The model initially created only contained three components, which included the core itself, the coolant around the core, and a single heat exchanger. The initial core was modeled as a cylindrical volumetric source that had an isotropic distribution of neutrons. The material card used in the core was simply a mixture of Uranium Oxide. The core was placed with its base at the origin of the MCNP model extending upwards with an active fuel length of 3.66 meters and a radius of 1.485 meters. The core consisted of an inner fuel region, a stainless steel reflector, and a stainless steel lower barrel with outer radii of 1.485, 1.60, and 1.65 m respectively. The coolant was supposed to be modeled as light water, however upon later model revision it was discovered that heavy water had been used instead. This mistake will be brought up again later when looking at the initial results. Finally, the heat exchanger was modeled as a rectangular prism with a height of 6 meters, a thickness of 1 meter, and a width of 54 cm approximately. The material card used for the heat exchanger was simply the elemental mixture of stainless steel 304.

The heat exchanger was placed approximately 3 meters above the core offset from the center. Both the heat exchanger and core were placed inside of a sphere of water, which in turn was placed inside a graveyard vacuum sphere to notify MCNP to no longer track particles. A diagram of the initial model can be seen below:



Figure 3. Initial Model in MCNP

The models derived from this model can be seen below when the MCNP code was run for 500,000 particles. As can be seen from this fairly simplistic model, an exponential decay in flux can be seen as the heat exchanger moves up the pressure vessel.

Figure 4. Flux in Heat Exchanger



5.3. Addition of Components

As mentioned earlier, the initial model also consisted of heavy water instead of light water, which is a much better moderator of neutrons, which in turn reduced the flux seen by the heat exchanger. From this point, the next additions made to the model were the additions of the upper barrel, vessel, biological shield, control rod plates, and the remaining printed circuit heat exchangers. The fuel and reflector were then modified from cylindrical shapes to a more accurate shape based on fuel assembly location. Finally code was added to account for changing boron concentration in the water.

The lower barrel was modeled as a cylinder composed of Stainless Steel 304. A thickness of 5 cm was used, making the outer radius of the lower barrel 1.65 m. A length of 4.358 m was used as the lower barrel length. For the upper barrel, a decreased outer radius of 1.5 m was used. The upper barrel was also 5 cm thick. A 25 cm thick stainless steel ring, positioned on top of the lower barrel, was used to connect the upper and lower barrels. The ring was hollow with the same inner diameter of the upper barrel and outer diameter of the lower barrel. The upper barrel had a length of 10.43 m, with the total length of the combined barrels being 15.04 m.

The pressure vessel was modeled as a cylinder capped by two half spheres and composed of carbon steel. The vessel had an inner diameter of 2.45 m and a wall thickness of 27.5 cm. The two vessel domes had a thickness of 13.75 cm. The cylindrical wall of the vessel was modeled to have a length of 15.04 m. In the actual design the vessel is approximately 10 % longer. The assumption was made that the shorter dimension would not affect the flux in the heat exchanger caused by backscattering effects. This assumption is believed to be valid because the flux was found to be zero far below this position from the core. The difference in vessel length was for modeling purposes only and was not intended to be a design suggestion. The area outside of the pressure vessel and between the bioshield was modeled as room temperature air.

The bioshield was modeled as a cylindrical wall with a connecting lower half sphere. The vessel was composed of concrete and had an inner diameter of 3.03 m and a wall thickness of 30.5 cm. The bioshield was not modeled to be dimensionally accurate but to simulate any backscattering of neutrons. A thickness was chosen that would effectively account for the scattered neutrons. Similarly the shield length was chosen to be equal to that of the vessel in order to conservatively account for the most neutron scatters.

The primary heat exchangers were modeled as rectangular boxes composed of a homogenous mixture of stainless steel 304 and water. The heat exchangers were modeled to be 100 cm wide with a depth of 54 cm. This model does not include the primary coolant header. The base of the heat exchanger was positioned at a vertical location of 6.65 meters away from the base of the core. It was modeled to have a length of 6 meters; however, in the most recent design the heat exchanger is 6.6 meters in length. This difference is assumed to have negligible effects because the flux above several meters of the heat exchanger was observed to be zero. The model was designed to allow the vertical position of the heat exchanger to be adjusted relatively easily. The weight percent of water was determined by taking into consideration the volume of the micro channels in the heat exchanger. It was determined that the heat exchanger was 7.3% water and 92.7% steel by weight. The model includes the eight primary heat exchangers. The secondary decay heat exchangers were not included in this model.

The control rod drive mechanism and alignment plates were modeled as 10 cm thick cylindrical disks composed of stainless steel. The upper alignment plate was modeled to be located directly on top of the fuel. The lower and upper CRDMs were positioned at a height 8.43 and 12.9 m in the vertical direction. The radius of these plates matched the inner radius of the barrel.



Figure 5. 2D Vised Representation of Final MCNP Model (NOT to Scale)

The next step in improving the model was to redefine the fuel region from a cylindrical shape to a more accurate lattice structure. The fuel core was modeled as a homogenous region with the shape of the combined 121 assemblies. The dimensions for this region were determined from the assembly pitch of 23 cm and number of assemblies per row in the lattice structure. The weight percentage of the elements that compose the fuel pellets, cladding, and the coolant were included in this fuel region. The length of the fuel was taken to be the active length of 3.658 meters.

Figure 6. Final Model for Reactor Core(Homogenized fuel shown as purple, reflector and lower barrel shown as light blue)



The change of the fuel shape also affected the reflector dimensions. The reflector was modeled as 100% Stainless Steel 304. The reflectors outer radius and length remained the same as the initial model with values of 1.6 and 4.058 m respectively. However the inside of the reflector was now defined by the shape of the lattice structure fuel instead of the cylindrical shape. 5.4. Light Water and Borated Water

One of the most integral parts of the MCNP model is the water incorporated around the different reactor components. It serves as the primary moderator in the system and reduces the flux seen by the heat exchanger by orders of magnitude. When the model was initially created, the only line representing water was a single material card with the atomic percentages of hydrogen and oxygen. However, as the model has advanced, so has the material card for water.

The first advancement made upon the material card was to incorporate the correct cross sectional data. By taking a conservative estimate of the water to be 600K at any time, the cross sections of .71c and .53c were used for were used for Hydrogen and Oxygen respectively.

One important addition made was that of a scattering kernel, or in more formal terms an MT card. The importance of this card comes into play when considering hydrogen molecularly bound in either water or some other constituent particle. This binding affects the slower neutrons making collisions, and in turn dictates how they scatter. Therefore, this MT card takes into account special cross-section data treatments for binding effects of Hydrogen. It’s important to remember, without the presence of the MT card, Hydrogen would simply be treated as a monatomic gas, with complete disregard to any binding effects.

The final advancement made upon the material card for water is the sensitivity analysis of heat exchanger flux with varying concentrations of boric acid. In most typical light water reactors the concentration of boron decreases with the life of the reactor. In making the model as accurate as possible, the presence of boron would be added through homogenizing the water with boric acid. The range decided upon would be from 0 ppm to 1000 ppm, broken into 200 ppm segments as specified by Dr. Petrovic. These varying material cards for different ppm have been included in the code with comment cards as can be seen below:

Figure 7. Material Card for Water in MCNP

By simply removing a comment card indicator ‘c’, and by commenting out the previous moderator, one can go through and augment the ppm of boron at will.

5.5. Air



The importance of modeling air in the system is realized when adding the biological shield. In the model air occupies any space outside of the pressure vessel and not inside the biological shield. It fills the cylinder that designates the graveyard entirely and is composed of a standard composition for air.

This standard composition consists of approximately 75% nitrogen, 23% oxygen and the rest a medley of carbon and argon. These elements all utilize room temperature cross sections which is designated by not adding a cross sectional table at the end of the ZAID number. While it is understood that the temperature of air around the pressure vessel will be more than likely higher than room temperature, the bins MCNP typically creates for their cross sectional tables are separated by 300 degrees. The importance of including air may not be paramount, however, it does contribute to certain neutron interactions and thus cannot be substituted with a vacuum. 5.6. Source Definition

The source definition modeling will be broken into three components to best explain not only the chronological process taken but allow for readers to easily comprehend the difficulty in defining a correct source definition.

5.6.1. Homogenization

The very first step in defining the source was to find a way to represent the material in the core. This could be done through two different methodologies. The first being separating the core into its component parts, such as the Uranium Silicide (U3Si2) fuel, the water, the cladding, etc. In turn, this would mean a different material card for each component in the reactor core. However, modeling the intricacies of the complex lattice structure seen inside the I2S-LWR and the figure below is not only time consuming in writing the code, but also time consuming in running the code. By attempting to model the lattice structure exactly, the amount of cells and surface cards would also exponentially increase.

Figure 8. Core Lattice Structure



The question then becomes, what accuracy of results are needed that would warrant the use of such a complex code. After consultation with Dr. Petrovic, the idea of completing such a complex structure was dismissed for the reasons of limited time and code complexity.

The second option is then to homogenize the material inside the core and treat the source as a single material. The homogenization process is explained through the following table:

Table V. Material Homogenization

First, the weight percentages are found for each individual component of the reactor core. In the model the core consists of water, cladding made from Oxide Dispersion Strengthened (ODS) steel, and U 3Si2. Boron rods are neglected in this calculation as the highest estimate for flux is desired in the heat exchanger. As boron rods act as neutron poison/absorbers they would reduce the overall flux in the core, and ultimately the flux affecting the heat exchangers.

Returning to homogenization, the weight percentages are taken, and multiplied by the density of the individual component the weight percentages were pulled from. In this case the density of U3Si2 is 11.3 g/cm3. This density is multiplied by the individual elements that make up the U3Si2 which are approximately 7% U-235, 85% U-238, and 7% Silicon. These values are then multiplied by the volume of the individual component, which yields the mass of the individual elements in a component. For example in the U 3Si2 core fuel here is approximately 143.0277 g of U-235. Finally this number is divided by the total mass of the core to yield the homogenized weight percentage. This process is repeated for each component of the core with the final sum of the weight percentages equaling one.

With these values and the ZAID codes, a material card is created for the homogenized material and can be seen below:



Figure 9. Material Card for Homogenized Fuel

Note the cross section tables used in the core vary from any other tables used to date. As peak temperatures in the core are much higher than values found in the water surround the core, the .72c tables are utilized which correspond to 900K. From this point the question becomes how should the source definition be modeled, as KCODE or as a Fixed Source?

5.6.2. KCODE

Initially, the design of the source code was modeled as a fixed source. The issue that continually arose with the original code was an unencumbered neutron multiplication. While this was not thought to be a cause for concern, it was quickly discovered that in multiplying the neutrons in the source volume as they were, MCNP would quickly resign from tracking the sheer number of neutrons produced after only a few thousand particles were run, as each one could produce additional fissions.

In an attempt to curb this rampant neutron multiplication, the original fixed source definition was quickly replaced with a KCODE definition, which can be seen below:



Figure 10. KCODE for Source Model

The beauty in using the KCODE comes from being able to initially define the neutron multiplication constant for the reactor otherwise known as the eigenvalue, and then maintaining the criticality within the range of the specified parameter. All that is required from the user are a few initial run parameters and fission points.

In the first line after the word kcode come the parameters for number of particles per cycle, initial eigenvalue, the number of particles to skip before averaging the eigenvalue, and finally the total number of cycles to run. The line below it beginning with ksrc defines the initial spatial distribution of the fission points. In the model the points form a 3d Cartesian axis directly through the very center of the cylindrical source.



5.6.3. Fixed Source

As this KCODE model was unveiled, Dr. Petrovic strongly suggested the use of a fixed source instead of the kcode citing the rampant multiplication of neutrons could be reduced through disabling any fission from occurring within the source. This can be seen below in the current source definition, which incorporates a NONU card, which treats all fission events as capture events instead.

Figure 11. Fixed Source Code with NONU Card

The source definition itself is created by defining some variables within the card. The first variable ERG, defines the energy spectrum utilized by the source, in the model this is a Watt Fission Spectrum, exactly the same as would be seen in a kcode. The second variable Cel defines the cells which are being treated as the source. In the model, these represent the different components of the lattice structure. The rest of the variables help define the cell within which the homogenized volumetric source resides. It is important to remember that the cell defined here must encompass the entire source cell.

5.7. Heat Exchanger Partitions

In performing the flux to activation calculations, an important factor to take into account is how the heat exchanger will have a higher activation near the bottom as opposed to the top. While the top of the heat exchanger is expected to receive some flux, and in turn activate to the extent to which this happens will be magnitudes less than what is experienced by the bottom. This is due in part to differences in distance from the core and neutron flux being reduced exponentially as neutrons travel through the steel. To account for this and to have the best possible tally information, the primary heat exchangers have been partitioned into graduated length segments. While this has no effect on the functionality of the heat exchanger within the model, it does allow for a better understanding of the flux distribution through the height of the heat exchanger.

The partitions vary from 2 cm segments to 50 cm and the planes defining the partitions can be seen in the code below:



Figure 12. Code for Heat Exchanger Partitioning

In the current model the flux tallies are only being calculated through the first 13 cells, allowing the reduction of run time. Once more time can be dedicated to large, extensive runs, it is recommended that flux tallies be performed across all partitions.

5.8. Mesh Tally

In understanding the flux distribution and to verify the neutrons are behaving as they should in the model, mesh tallies were utilized to track the progress of neutrons through the reactor.

The code for the mesh tally can be found below:

Figure 13. Code for Mesh Tally Analysis



Where the origin describes the starting point for the mesh tally, and the I, J, and K mesh terms describe how far to take the mesh tally, discretized by cm units designated by the INTS terms. If one were to imagine this mesh, it would be a full rectangular block that encompasses the entirety of the pressure vessel, not the bioshield. However, it has a more defined resolution along the z axis rather than the x or y. These mesh tallies became important later on when understanding how adding additional neutrons to each run would impact the distance which flux could reach. 5.9. Variables to Change

In the final analysis model three independent variables were chosen, these included the vertical location of the primary heat exchangers, the height of the reflector surrounding the core, and the boron concentration in the water. As mentioned previously, extensive research was done in materials as well as shielding, ultimately these were decided against and the three options above were selected.

In the final MCNP model, the vertical location of the heat exchangers is modified simply through augmenting the z dimension of each box cell created in MCNP. However, the challenge comes when attempting to push up the 50 planes that divide up the heat exchanger into tally segments. Fortunately through Microsoft Excel, the text can be augmented off of a lynch pin number and each segment adjusted accordingly.

The proposed variations in vertical location are approximately 20 cm movements both above and below the original location of the heat exchanger. With the 20 cm movements approximately 7 tally points should be found for the heat exchanger. In terms of the reflector variation, the alteration of code becomes slightly trickier. Because of the position of the core barrel connector ring, the reflector is immediately limited in the height it can obtain. However, through advisement from Dr. Petrovic, the lower barrel can be extended, which in turn pushes up the connector ring.

Figure 14. MCNP Model Showing Upper and Lower Barrel



It is important to recognize, however, when the lower barrel is extended, if pressure vessel dimensions are to be retained, the upper barrel length must be shortened. Fortunately with the current components included in the MCNP model, augmenting the barrels imposes no problems upon any other components, the changes simply involve changing cylinder heights and plane locations. These changes would be enacted upon cell block 1 and 2.

The final variable to change is the concentration of boric acid in the water card. Fortunately, this is easily done through another homogenization process and well-placed comments. The varying concentrations for borated water are commented out for now but, by quickly removing the comment designation symbol and commenting out another water material card, the concentration is quickly and easily varied.

6. DESIGN CONSTRAINTS

As one of the design tasks of this project, the sensitivity of the activation of the primary heat exchangers as a function of their axial distance to the core was analyzed, in addition to the effects of altering the height of the radial reflector. For varying the axial distance of the heat exchangers, an increase and decrease in height of 2 feet was decided. This range was recommended by Dr. Petrovic and would theoretically be enough to produce a sufficient sensitivity analysis of the activation of the heat exchanger. With increasing or decreasing the axial distance of these components, there is also a concern for altering the locations of other components surrounding them. In the model created in MCNP, the primary heat exchangers are situated approximately 2.3 meters or 7.54 feet above the lower barrel of the core. The top of the vessel in the model is located about 2.4 meters or 7.87 feet above the top of the heat exchangers. However, there are pipes located both immediately above and below the heat exchangers that act as inlets/outlets for the primary and secondary coolants. In addition, there are a total of four pipes between every set of two heat exchangers that act as the decay heat removal system. These pipes are 700 mm in diameter and 7,300 mm in height and are longer than the primary heat exchangers. Because these pipes are located in close vicinity of the heat exchangers, the range of ±2 feet is a realistic approximation as to how much physical space is available for the axial movement of the heat exchangers.



Figure 15. Decay Heat Removal System

In addition, the radial reflector will be extended in height to analyze its effect on heat exchanger activation. As mentioned previously, the reflector spans the length of the core and is 405.8 cm tall in the model. The reflector is then tightly surrounded by the lower section of the barrel. This introduces a minor obstacle in increasing the height of the reflector, because the upper barrel is significantly smaller in diameter and is positioned directly above both the reflector and lower barrel. Thus, when the height of the reflector and lower barrel are increased, the length of the upper barrel will be decreased an equal amount. The height of the reflector will be increased up until it reaches the bottom of the primary heat exchangers at approximately 665 cm.

7. FINAL DESIGN ANALYSIS AND EVALUATION

7.1. Final Model

As the semester progressed it was made readily evident that the amount of time to run each code was increasing exponentially. With that knowledge in mind, the question was asked, does one compromise the accuracy of the model for run efficiency or create an accurate model but forego run efficiency? Given the senior design group is transient but the work itself is transitional, accuracy was decided as the focal point for the model.

With this in mind, to date, zero flux has been recorded across any of the heat exchangers. The reasoning for this is understood simply from the complexity of the model. Given neutrons along the Watt Spectrum have to travel through almost three meters of water and layers of steel it’s understandable and expected that zero flux would be recorded with such a low number of particle sampling. Low being a relative term, as most would consider multibillion particle runs to be fairly significant, when compared to the activity magnitude in the core of almost 1018, almost 9 orders of magnitude greater than the number of particles run. However, given the



tenacious nature of most Georgia Tech students, settling for zero results wasn’t enough. In an attempt to remedy this, a few solutions were proposed and implemented.

The first solution to reduce run time and increase particle sampling was through having energy cutoffs in different portions of the model. Specifically for the water regions outside of the core, anytime a neutron enters with below 2 MeV, MCNP would immediately stop tracking the particle. This assumption was made given that any neutron below 2 MeV would struggle to travel through the number of mean free paths entailed in 3 meters of water. By not accounting for these particles, the run time was reduced significantly, yet zero flux was still being seen for the heat exchangers, even with running 2 billion particles!

The next solution to garner results was to reduce the scope of the discretized mesh tally. Initially, the mesh tally was designed to surround the entire pressure vessel to get a visual representation of where in the reactor was receiving flux. In sample models this mesh only need apply to the heat exchanger as that is the area of significance in the scope of this project. However, even with this reduction in run time, zero flux was still being received across the board even with some of the longer runs taking longer than 6 full days. Perhaps with access to highly powerful computers that could run multiple parallel processes or servers that would allow for month long uninterrupted runs, results could be found, but as of now with the limited computing resources at hand it seems improbable any flux will be found. However, this leads into a perfect opportunity for any groups in the future who wish to use the code and advance the research done to date.

Figure 16. 3D View of Final Model

7.2. TecPlot Models

Given the lack of flux found after running the MCNP, multiple trials have been run to see how expanding the number of neutrons expands area within which flux is accounted for. Tecplot 360 models have



been created to demonstrate this. Alongside each Tecplot model, a few numbers are calculated as well such as the furthest x, y, and z points from the center of the core base. And also the furthest point where flux is seen from the center based off total distance, found through the Pythagorean Theorem expanded into three dimensions, with the accompanying flux at that location. These have been created for MCNP codes run with nps of 1 million, 5 million, 25 million, 125 million, 500 million, 1 billion, 2 billion, 5 billion, and perhaps 10 billion.

1 MillionX Position (cm) 214.375Y Position (cm) 238.875Z Position (cm) 425.25Furthest Point (6.125, 202.125,

425.25)Distance from Center (cm) 470.8817Flux Tally (n/cm2) 1.18 x 10-9

5 MillionX Position 226.625Y Position 238.875Z Position 546.75Furthest Point (79.625, 67.375,

546.75)Distance from Center 556.6104Flux at that Location 1.56 x 10-11





1 BillionX Position 238.875Y Position 238.875Z Position 546.75Furthest Point (-128.625, -238.875,




2 BillionX Position 238.875Y Position 238.875Z Position 546.75Furthest Point (-226.625, -214.375,


10 BillionX Position 238.875Y Position 238.875Z Position 1032.75Furthest Point (238.875, -

177.625, 1032.75

Distance from Center 1074.795Flux at that Location 4.03 x 10-14

As can be seen above, with incremental increases to the nps, fluxes further and further from the bottom of the core are achieved. With this idea in mind the question was asked, how many particles would have to be run for particles to potentially reach the bottom of the heat exchanger, approximately 686.5095 cm. The assumption is made that there is some causal relation between number of particles run and furthest point flux can reach in the model. Based off of the numbers from the mesh tallies the following graph has been created to find an exponential relationship between nps and distance, which can be found below:



400 450 500 550 600 650 7001.00E+06

1.00E+07

1.00E+08

1.00E+09

1.00E+10

1.00E+11

1.00E+12

f(x) = 1.34485180056902E-05 exp( 0.0516390789501701 x )

Distance from Center (cm)

Nu

mb

er o

f Neu

tron

s

Figure X: Furthest Flux Distance from Core Base verse Particles Tracked

Extrapolating from the relationship found between the data, it is predicted that approximately 25 billion particles would yield some sort of flux at the very bottom edge of the heat exchanger. Given the current capacity this run is estimated to take 5 days. However, given a full distribution is desired throughout the heat exchanger, much more than 25 billion particles would have to be run.

7.3. Future Design Plans

Given the complexity of the model, the majority of the research time has been primarily spent building the model. Unfortunately, with such complex models the time required to run even just a billion particles can take upwards of 6 days, based on the computer being utilized. Given the limited resources Georgia Tech students have in terms of computer power, the primary resource being the Citrix servers provided by Georgia Tech. This circumstance being understood, a few suggestions have been made on what could be done in place of a robust MCNP model if time permitted.

The first suggestion would be to utilize particle splitting in regions of relatively high importance, simply utilized as a variance reduction technique. Out of all the methodologies this one would most likely be easiest to implement.

The second suggestion would be to create a deterministic model instead of using a Monte Carlo radiation transport method such as with MCNP. Software suggestions along this vein include SCALE and also CASMO & HELIOS. The reasoning for using deterministic software versus MCNP comes from the problems to date. Due to model complexity, running a statistically significant number of particles is extremely time-consuming. However, if these statistically significant numbers are not reached, no flux is shown in any of the heat exchangers. With deterministic models even though the run times are typically longer than MCNP models, an exact solution is guaranteed everywhere along the discretized mesh.

A third suggestion comes from attempting hand calculations. However, even in the most simplistic solution of a 2-region/2-group problem, with an incoming current equivalent to that of the top of the core, can take pages upon pages of calculations. Given the level of advanced mathematics required to solve these types of problems, this solution would be recommended to those needing an initial calculation but also having a strong math background. The other option along this branch would be to use a digital differential equation solver.



The fourth and final alternative that can be taken is to produce an approximation of a two dimensional slice of the pressure vessel. If looking at the reactor from above the situation would be similar to taking a sliver of a pie chart.

Figure 14. 3D Representation of Slice Model

The justification for this approach comes from the assumption neutron flux is isotropic radially. With this assumption in mind a simplistic model can be created by simply modeling a slice of the heat exchanger, reactor vessel, barrel, etc. Any portion outside of the slice would be treated as a reflective surface to account for any backscattering from other components. With this model in mind, a simple mesh tally can then be taken over the volume or “area” of the heat exchanger. Unfortunately with the simplifications that all neutrons backscatter and the flux is isotropic radially, the flux experienced by the heat exchanger would be much greater than what would actually occur.

7.4. Suggestion for Future Work

This investigation on PHXR activation was performed over the course of 15 weeks. Due to this time constraint, the objective was only to acquire the PHXR activation rate. Had there been more time allotted to this project, an end application such as worker dose analysis could have been studied as well. However, with the results as they are, the data presented can act as a foundation for future uses. For example, if one desired to acquire the total dose a worker would receive standing directly atop the PHXR, a simple MCNP model utilizing a cylindrical macro body of water alongside the PHXR model provided would suffice. The PHXR would also need to be cut into multiple slices, to effectively model a correct activation distribution across the geometry of the PHXR structure. This combined with a tally of choice would easily provide a general idea of the possible dose deposition rates or values. Companies would benefit significantly by using the provided model to ascertain the dose values that are possible during operation and thus in turn could assist reducing risk oriented costs for its workers.



8. ACKNOWLEDGEMENTS

This project would not have been possible without the help of Christopher Edgar, particularly with modeling in MCNP. Chris met with our group for countless hours and gave us valuable insight on different ways to approach our problem through MCNP.



[References]

Flat Product Stainless Steel Grade Sheet. (2014). Retrieved February 5, 2015, from http://www.northamericanstainless.com/wp-content/uploads/2010/10/Grade-316-316L.pdf

Bell, T. (2013). Metal Profile: Manganese. Retrieved February 5, 2015, from http://metals.about.com/od/properties/a/Metal-Profile-Manganese.htm

Diehl, P. (1999). Neutron Activation Calculator. Retrieved January 25, 2015, from http://www.wise-uranium.org/rnac.html

Memmott, M., Petrovic, B., & Marchese, M. (2014). Integral Inherently Safe Light Water Reactor (I2S-LWR) Concept: Integral Vessel Layout. Proceedings of the 2014 International Congress on Advances in Nuclear Power Plants. Bell, J., Charry, C., Dingman, N., DiMascio, P., Hanley, B., & Powell, C. (2013). Microchannel Heat Exchanger for Integral Pressurized Water Reactor. Georgia Tech Nuclear and Radiological Engineering Design.

"Cobalt-60m-decay" by Tubas-en - Own work. Licensed under Public Domain via Wikimedia Commons - http://commons.wikimedia.org/wiki/File:Cobalt-60m-decay.svg#/media/File:Cobalt-60m-decay.svg

"Cobalt | Radiation Protection | US EPA 2012

Zhou, Q., Yang, J., Wu, J., Tian, Y., Wang, J., Jiang, H., & Li, K. (2014). An improved algorithm to convert CAD model to MCNP geometry model based on STEP file. Annals of Nuclear Energy, 78, 81-88. Retrieved from ELSEVIER.

Nyarku, Mawutorli, Ramanthapura S. Keshavamurthy, Venkata D. Subrmanian, Adish Haridas, and Eric T. Glover. "Experimental Neutron Attenuation Measurements in Possible Fast Reactor Shield Materials." Georgia Tech Library. Science Direct, 26 Nov. 2012. Web. R V Kolekar, R Kumar, and DN Sharma. “Shielding experiments for optimization of shield materials in fast reactor using SSNTDs” Georgia Tech Library. Science Direct, 3 Feb. 2013. Web.

I 2 S-LWR Burnable Absorbers Design Team G2 Andrew Conant, Casey McArthur, Gage Richert, Angelo Spinetta, Aaron Tumulak 2014

Autodesk Inventor Modeling of Designed and Major Reactor Components Design Team 3D Brian Barron, Matthew Marchese, Sterling Olson, Paul Rose, Michael Saunders, Brian Schwartz 2013

http://alaquainc.com/Heat_Exchangers.aspx#Calcium_carbonate

Appendices[Waterford 3 emails]



Pamela,My son, Adnan, is requesting me to collect some info for his Nuclear Engineering Project at GA Tech. I will appreciate if you or other nuclear engineers could answer to his questions noted below.

Thanks,

Joel M. HashimSenior Engineer/ Fire Protection EngineeringEngineering Programs & ComponentsEntergy, Waterford 3 Plant(504) 739 - 6446

-----Original Message-----From: Adnan Hashim [mailto:[email protected]] Sent: Tuesday, March 10, 2015 5:32 PMTo: Hashim, Joel; Joel HashimSubject: Senior design questions about Primary Heat Exchangers in LWR

EXTERNAL SENDER. DO NOT click links if sender is unknown. DO NOT provide your user ID or password.

Hey Dad,

My group and I would greatly appreciate it if you could help us find resources or exact values/documentation about the Primary Heat Exchanger in your PWR at Waterford 3.\ We are working on a research design on an experimental reactor known as the I2S-LWR (integral inherently safe light water reactor) for the univerisity. These are some basic questions that imply more detailed questions following them.

Questions:

1) Estimated costs of producing one heat exchanger unit.2) Company that produces the heat exchanger.3) Information on printed circuit heat exchangers from a company known as Heatric4) Estimated dose rate or activation rate of Waterford 3 heat exchangers from neutron leakage from the core (neutron activation)5) Any personnel or faculty that can provide more information about these questions.6) Material composition of heat exchanger.7) -Possible cost analysis of core reflector material and amount of material.

Sincerely grateful,Adnan Hashim2:

Adnan,

Are you asking about the steam generator? Or secondary side? That would make a big difference in the questions you're asking. I'm not familiar with the I2S-LWR design.



I've forwarded your email on to a system engineer who works with heat exchangers and he may be able to help figure out what it is you're looking for.

As for the dose rate and activation questions, I'm not sure what you're looking for there. No heat exchanger is directly impacted by neutrons to the point that we worry about activation or embrittlement. Dose rate issues are mostly caused by crud buildup in the tubes if on the primary side (steam generator) and are not a concern in the secondary system.

For the core reflector, the carbon steel vessel is lined with stainless steel and inside the primary shield wall that is ~5.5' thick concrete. It is not actually credited in the design as a reflector though. Our core design is such that the perimeter fuel assemblies are high burnup and act as our reflector leading to a center peaked power profile, limiting neutron fluence on the vessel to preserve vessel life.

I've attached some of our system descriptions. The information they contain is not marked as proprietary or security related, but I ask that you not widely distribute it or post it anywhere online.

3:

Pamela,I greatly appreciate your response and also appreciate the attachments. I apologize for the confusion on the heat exchangers. In our study case the I2S LWR being designed by a few universities as a hypothetical SMR has the primary and decay heat exchangers integrated within the reactor vessel right outside the core barrel for easy transport. The objective of the reactor is to be able to be transported and replaced easily. I misconstrued the question forgetting that standard PWRs do not incorporate primary coolant heat exchangers within the reactor vessel. I apologize for the confusion.The info you have provided is very useful and again I appreciate your time.Sincerely,Adnan Hashim

4: Kunkel

This is the information that I know. 1) Estimated costs of producing one heat exchanger unit. $175M2) Company that produces the heat exchanger. Westinghouse Nuclear3) Information on printed circuit heat exchangers from a company known as Heatric. Do not know4) Estimated dose rate or activation rate of Waterford 3 heat exchangers from neutron leakage from the core (neutron activation). We cannot access the generators at power due to dose. Based on known neutron dose rates I estimate 10R and the lower portion of the generator. 5) Any personnel or faculty that can provide more information about these questions.6) Material composition of heat exchanger. Inconel Alloy 690 tubes.7) -Possible cost analysis of core reflector material and amount of material. We use thick (3-5 feet) concrete walls for shielding.



Keith Kunkel 5:

Mr. Kunkel,

I apologize for the late response but I’d like to to thank you for your time and the valuable information you have provided. If you don’t mind, I’d also like to inquire the composition and quantity of “gunk” buildup in heat exchangers mentioned to me by Pamela. Could you possibly expound on that?

Sincerely,Adnan HashimGeorgia Tech NRE class of 2016

6: Kunkel

We get a buildup of corrosion products on the secondary side of the steam generator. The primary side is all stainless steel and remains clean.

7:

Mr. Kunkel,

Looking inside the primary coolant side steam generators/ heat exchangers, is there any activated steel or metal particle build up? If so do you happen to know the quantity or their radiation effects?

Thanks,Adnan Hashim

8: Kunkel

There is no buildup on the primary side. The primary water is kept extremely clean.


final paper rev1

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