The document describes a mixed-spectrum supercritical water reactor (MSWR) design. Key points: 1) The MSWR was scaled down from a reference US supercritical water reactor to a power output of around 800 MWth. It uses a mixed spectrum of neutron energies and 8% enriched uranium dioxide fuel. 2) Fuel cycle analysis found 8% enriched UO2 to be the most cost effective fuel choice, corresponding to a three cycle rotation of the fuel assemblies. 3) Thermal-hydraulic analysis determined coolant temperature and pressure drop profiles. Coupling neutronics and thermal hydraulics flattened the axial power distribution due to negative moderator temperature coefficients. 4) Safety analysis

1. NERS 442
Final Report
Mixed-spectrum Supercritical Water Reactor
WINTER 2013
Team 3 - Yuan Gao, Douglas Kripke, Nishant Patel, and Eric Welch
Department of Nuclear Engineering and Radiological Sciences
University of Michigan
Ann Arbor, Michigan

2. 2
TABLE OF CONTENTS
I. ABSTRACT..................................................................................................................................3
II. PROJECT DEFINITION AND REFERENCE CORE BENCHMARK ........................................................3
1. Background on SCWR concept
2. US reference design
3. Introduction to our unique reactor concept and design specifications
III. FUEL SELECTION AND ANALYSIS ................................................................................................8
1. Fuel and enrichment selection
2. Fuel Cycle Analysis
i. k-infinity
ii. Isotopic depletion analysis
IV. THERMAL-HYDRAULIC COUPLED NEUTRONICS AND BEGINNING OF CYCLE ANALYSIS............11
1. Coolant and moderator channel temperature profiles and feedback
2. Coolant and moderator pressure drop calculations
3. Radial and axial power distribution
V. SAFETY ANALYSIS ....................................................................................................................14
VI. PROLIFERATION RISK .............................................................................................................. 15
VII. ECONOMIC ANALYSIS............................................................................................................. 15
VIII. SUMMARY ............................................................................................................................. 16
IX. REFERENCES ............................................................................................................................ 17

3. 3
I. Abstract
The mixed-spectrum supercritical water reactor (MSWR) is designed to meet the current
demands of the nuclear energy industry. It is modular to reduce capital costs and streamlined to
include passive safety features. As name implies, it utilizes supercritical water as the coolant to
achieve a thermal efficiency of 44% (compared to 33% in typical LWRs) as well as the full
spectrum of neutron energies. The MSWR was benchmarked using the US supercritical water
cooled reactor (SWR). From the US SWR design, the MSWR was scaled down from a power
output of 3400 MWth to around 800 MWth. The macroscopic design is most similar to the
economic simplified BWR design; but the utilization of high pressure (25 MPa) eliminates the
need for a steam separator, steam dryer, and recirculation pump due to the single phase flow of
the coolant, thus greatly reducing capital costs. Fuel cycle analysis indicated 8% enriched
uranium dioxide (UO2) as the most cost effective fuel choice corresponding to a three cycle
rotation of the fuel assemblies. With this configuration, the busbar cost of MSWR will be
around 67 mill/kWhr compared to 75 mill/kWhr of the AP100. Favorable reactivity coefficients
demonstrate the safety of MSWR design, and isotopic analysis show an increased proliferation
risk.
II. PROJECT DEFINITION AND REFERENCE CORE BENCHMARK
The super-critical water-cooled reactor (SCWR) is a generation IV conceptual design
employing super-critical water as both the coolant and moderator. In order to achieve this, the
system must be kept above water’s critical point of 22 MPa and 374°C. Utilizing super-critical
water allows the reactor to share in the same thermal efficiency experienced by super-critical
coal fired power-plants. With a core inlet/outlet temperature difference of 220°C (compared to
30°C for standard LWR systems), thermal efficiency is improved from 35% to nearly 45%. This
corresponds to the core outlet temperature reaching around 500°C. Once the water has passed its
critical point, properties such as density, specific heat, and enthalpy undergo drastic changes as
seen in Fig. 1. The rapidly increasing enthalpy allows one to achieve heat removal comparable
to current LWR systems while using a coolant flow rate that is up to an order magnitude lower.
Moreover the single-phase property of super-critical water eliminates the need for a steam
separator, a dryer, and recirculation pumps. These design simplifications will make the SCWR
cheaper to manufacture by reducing capital costs.

4. 4
.
Figure 1: Macroscopic view of the reactor (left) and thermo-physical properties of water (right)
[Hu08]
The US reference SCWR design was chosen as a starting point to benchmark to before
beginning any work on the MSWR. We initially hoped to use the mixed spectrum design as our
reference core, however, the literature on this reactor concept proved insufficient. As a result,
we switched to using the more established US design. Figure 2 displays the planned core and
fuel assembly of the US reference design. This design features square fuel elements as opposed
to the hexagonal array of the MSWR design. Furthermore, the US design uses a thermal
spectrum and low enriched UO2 fuel compared with a mixed spectrum and various potential
Figure 2: Cross-section of fuel assembly (left) and water rod (right) for the US reference SCWR
design [Hu08]
Moderator rod
Coolant Channel
Fuel Rod

5. 5
Table I: Parameters of the US reference SCWR design [Hu08]
combinations of UO2 and thorium mixed-oxide (TMOX) fuel for the MSWR design. Table 1
quantifies all of the design details necessary for benchmarking the MSWR design. The water
rods are a unique feature of the US reference design, which are shown with greater detail in
figure 2. In the blue region, the coolant water goes down with a higher density of about 0.7
g/cm3
. The higher density supercritical water can be used as a moderator. In the red region,
coolant water go up with a lower density of about 0.15 g/cm3
. With a lower density and higher
temperature, the water losses its moderating function. The green region represents the fuel rod.
As coolant enters the core, about 90% of the feed-water goes into the upper plenum of the core,

6. 6
and then flows downward in the water rods. At the lower plenum it mixes with the rest of feed-
water, and the mixture then flows upward as coolant.
In meeting the benchmark, the thermal-hydraulics was not coupled with the neutronics
due to time restriction. To benchmark our results with the reference design, a core radial power
distribution was created with some control rods partially inserted. The results showed good
agreement with the reference design. Additionally, the assembly level k-infinite value was
modeled as a function of burn-up. The results were very close to that of the reference paper,
with error of about 1%. At this point, the benchmarking was considered satisfactory enough to
move forward with determining the parameters of an ideal and modular MSWR design.
The size of the reference design was reduced in order to make the MSWR design
modular. As a result, the MSWR has a power output of 355 MWe (the reference design was
scaled down from 3400 MWth to 807 MWth). This was done by reducing the size of the
assemblies to 17x17 and reducing the number of assemblies in the core to 61. The layout of the
core is shown in figure 3. Apart from the dimensions, the overall design of the MSWR remained
unchanged from the reference design shown in figure 1.
As the name implies, the MSWR operates using a mixed spectrum of neutron energies.
Unlike the US SCWR, the MSWR does not have any water rods for extra moderation. Instead,
the flow of the coolant was redesigned such that each assembly has the coolant either flowing
upward or downward. The core has three types of assemblies fresh burnt assemblies are the only
assemblies that have coolant flowing down through them. The fresh fuel assemblies and the
once burnt assemblies have coolant flowing upward. The coolant from inlet initially flows down
Figure 3: Cross-section of core (left) and fuel assembly (right) for the MSWR design

7. 7
through the twice burnt assemblies, and then flows upward through the fresh fuel and the once
burnt assemblies. The water becomes supercritical towards the bottom of the core, and therefore
it has a high enough density to be a good moderator as it is flowing through the twice burnt
assemblies. Thus, the twice burnt assemblies primarily capture thermal neutrons. When the
water is flowing upward, it is supercritical and has low density with poor moderating
capabilities. Therefore, the fresh fuel and the once burnt assemblies are the fast assemblies. The
assembly design for the MSWR is shown in figure 3. The cladding material was also changed to
silicon carbide from stainless steel in the reference design because SiC has a lower neutron
absorption cross section (the value of kinf increased to 1.45 with SiC up from 1.16, which was
previously determined by altering the number of control rods, burnable absorbers, and the fuel
enrichment). Thus SiC was chosen as the cladding material. Table II displays various properties
of the MSWR core.
Table II: Summary of the MSWR core properties

8. 8
III. FUEL SELECTION AND ANALYSIS
At the process of fuel selection, three types of nuclear fuel were considered to be the fuel
of our core, MOX, (Th+Pu+U)O2, UO2. The MOX was considered to be a fuel for the light water
reactor, and the (Th+Pu+U)O2 was chosen as a mixed spectrum reactor fuel. UO2 has a relative
high melting point, which goes to 2865°C. UO2 has cheaper enrichment price, which leads to a
much cheaper fuel cost. The uranium dioxide has less weapon grade products compare to the
MOX and (Th+Pu+U)O2. So the UO2 was chosen as our reactor fuel.
The fuel enrichment is a big issue in our design. Two factors were considered in the
selection of the fuel enrichment, one is the k-infinity and the fuel enrichment cost. From figure 4,
we can see that the k-infinity of 7% enrichment uranium dioxide goes down to 0.89876 smaller
than 0.9 at the 50 (GWD/UT), which does not meet our design requirement. The 8% and 10%
enriched fuel has a k-infinity higher than 0.9, which meet our design requirement. If we only
consider the k-infinity factor, both 8% and 10% enrichment fuel can satisfy our design
requirement. But the enrichment cost is another important factor that we have to consider.
Figure 4 is a plot of the fuel cost versus fuel tails enrichment. We can see that the price of the 10%
enrichment fuel is 25% more expensive than the 8% enrichment fuel. So the 8% enrichment
UO2 was chosen as our fuel.
Figure 4: Cost versus tails enrichment for MSWR design
0
500
1000
1500
2000
2500
3000
3500
4000
4500
0 0.001 0.002 0.003 0.004 0.005 0.006
Cost($/kgU)
Tails Enrichment (%)
7% Fuel enrichment
8% Fuel enrichment
10% Fuel enrichment

9. 9
Figure 5: k-infinity versus fuel discharge burn-up from SCALE for MSWR design
Figure 6: Fuel burn-up analysis for MSWR design from SCALE
MSWR uses standard UO2 fuel pellets, but with an initial enrichment of 8 wt% U-235.
The fuel will be shuffled in the core in two year cycles until being discharged after the third
0.8
0.85
0.9
0.95
1
1.05
1.1
1.15
1.2
0 10 20 30 40 50
K-Infinity
Fuel discharge burnup (MWd/kgU)

10. 10
cycle when the average urnup is near 60 MWd/kgU. Depletion calculations at the assembly
level were carried out with SCALE and the resulting kinf values are shown in figure 6. The
important thing to notice is that the fuel spends two cycles in a fast spectrum region (where
neutron moderation is limited). The fuel then moves to a thermal spectrum region for its third
cycle and receives a boost in reactivity from increased neutron moderation. At this transition
period, the fuel composition is still mostly U-238, but with 4.5% U-235 and 3% fissile
Plutonium. When the fuel is discharged after the third cycle, it contains 3.2% U-235.
Full core depletion was done with PARCS without any thermal hydraulic feedback. The
value of keff starts at 1.1 and drops to just below 1 after the third cycle. To help control the extra
reactivity at BOC, the MSWR would utilize control rods with small insertions. We were unable
to complete a full TH-coupled depletion analysis due to our hand calculation method and a lack
of extended time.
IV. THERMAL-HYDRAULIC COUPLED NEUTRONICS AND BOC Analysis
The thermal hydraulic code packages that were available to us for this project cannot
correctly model supercritical water without modifying their source code. Unfortunately, we did
not have this option, so we moved forward with hand calculations to find coolant temperature
profiles for the thermal and fast assemblies. This was done by finding the change in enthalpy of
a core-average channel:
( )
( ) (1)
where W is the mass flow rate (see table III) and M is the wetted perimeter. We split the core
into 30 axial sections and used a discretized version of (1) to find the enthalpy of the nth
section:
∑ (2)
This initial step assumed a cosine shaped axial power distribution. From the enthalpy
distribution, we looked up the corresponding temperature T(z) and density ρ(z) profiles. We
then remodeled our core with SCALE and PARCS using the new ρ(z) for six axial regions. The
resultant axial power distribution is labeled “1st
” in figure 7. This entire process was repeated
once to get a neutron-thermal-hydraulic coupled BOC axial distribution, labeled “2nd
” in figure

11. 11
7. Figure 8 shows the coolant temperature profile difference between the neutron-coupled and
non-coupled cases. Coupling the thermal hydraulics to the neutronics in the MSWR flattens
the axial power distribution because of the effects of the largely negative MTC. That is, a region
that receives extra power is heated more, which lowers the reactivity of that region, thus
Table III: Summary of the MSWR thermo-hydraulic parameters
lowering its power. The inverse effect happens to regions receiving too little power, so they get
a power boost.
Figure 7: Radial and axial power distribution for MSWR design from PARCS

12. 12
Figure 8: Hand calculated coolant temperature distribution of MSWR design
The fuel temperature can be calculated by the axial power density and the coolant
temperature. The relation can be shown by the following equation:
( ) (3)
where q is the axial power density, Tb is the coolant bulk temperature, Tc is the fuel centerline
temperature, and Uc is the overall heat transfer coefficient
(4)
where a is the fuel OR, hg is the gap conductance, kc is the cladding thermal conductivity, h is the
general conductivity coefficient of water, tg is the thickness of the gap, and tc is the thickness of
the cladding. The fuel temperature profile was divided into 30 data points. The fuel temperature
distribution is shown in figure 9. Here, we can see that the temperature on fast assembly is
higher than the thermal assembly. The shape of fuel temperature distribution is quite similar to
that in reference paper.
Our core has two coolant channels, and there is turn around point at the bottom of the
core, so the pressure drop is mainly determined by three factors. The frictional drop,
gravitational pressure drop and the form pressure drop happen on the turnaround point. The
frictional pressure drop can be calculated by the following equation:
0
0.5
1
1.5
2
2.5
3
250 300 350 400 450 500 550
Coreheight(m)
Temperature(°C) Moderator with feedback Coolant with feedback
Moderator without feedback Coolant without feedback

13. 13
(5)
where H is the height, f is the friction factor of the coolant, v is the mass velocity. The
gravitational pressure drop is determined by the density and relative height, and can be
Figure 9: Coupled temperature profile from PARCS for MSWR design
calculated by the following equation:
(6)
The form pressure drop is determined by multiple factors, but the most important one is the bend
angle, and the form pressure drop can be estimated by the following equation:
(7)
The coolant goes down from top to bottom in the moderator assembly. So the total pressure
should be the following equation:
(8)
So the total pressure drop of coolant is tiny in the moderator assembly. The result is 745 Pa. The
coolant goes up from bottom to top in the coolant assembly. So the total pressure come to the
following equation:
(9)
From (9), we can see that the pressure drop is much bigger than the previous one. The result is
13604 Pa, which is not a too much number compare to the 25 MPa core pressure. Thus, the
0
0.5
1
1.5
2
2.5
3
0 200 400 600 800 1000 1200 1400
Coreheight(m)
Fuel Temperature (°C)
Thermal assembly
Fast assembly

14. 14
pressure drop will maintain the water in a supercritical state as it moves through the core as
required by our design.
V. SAFETY ANALYSIS
In designing a reactor, it is important to have a negative power coefficient of reactivity
(PCR) for safety purposes. This means that an increase in power from the reactor results in a
decrease of reactivity, which pushes the power back down. For our MSWR design, it is
important to consider a PCR for the thermal assemblies as well as one for the fast assemblies.
The PCR αP is found by combining the fuel temperature coefficient (FTC) αF with the moderator
temperature coefficient (MTC) αM:
(10)
(11)
(12)
Table IV: Calculated reactivity coefficients for MSWR design from SCALE
Estimates for all of the reactivity coefficients were calculated using outputs from SCALE and are
tabulated in table IV. The two PCRs calculated for the MSWR are both negative, ensuring that
the reactor will self-regulate in the case of a power increase.
VI. PROLIFERATION RISK
At the end of fuel cycle, we have 3.2% transuranium elements, in which 99.5% is plutonium. So
the proliferation resistance is an important issue that we have to think about. The advantage of
Thermal Assembly Fast Assembly
αF (pcm/K) -2.8 -3.4
αM (pcm/K) -80 -60
αP {(%Δk/k)/(% power)} -0.04 -0.13

15. 15
our core design in nonproliferation part is that we have a much longer discharge burn-up, is 63
GWd/Mt. We have less mass fuel used in our core, which leads to less mass weapon grade
products. As we have a higher concentration of plutonium in the transuranic elements, we can
reprocess the fuel wastes to MOX fuel. The high concentration of plutonium can help the
reprocessing to MOX fuel, and MOX fuel can be used as fuel of some light water reactors. The
longer burn-up time and the fuel waste reprocessing to MOX fuel can help on the proliferation
resistance work. We still need to control this reactor customer. The background and objective of
the customer should be check. We may not sell the reactor to the country without nuclear
weapon right now, or they may use the wastes to develop nuclear weapons.
VII. ECONOMIC ANALYSIS
The MSWR is designed to be economical. The higher thermal efficiency of the reactor
helps lower the total cost significantly. The principal amount was estimated to be $4B/GWe plus
a 20% increase for making it modular—that is $1.3B for MSWR. Over the Construction period
of four years, the interested accrued is $256M assuming a 10% interest rate per year. This yields
a capital cost of $1.5B. This is to be paid over the period of 40 years. The decommissioning
cost is estimated to be 20% of the capital cost. Management and operations (M&O) cost was
estimated to be 9 mill/kWhr. The busbar cost of the reactor comes out to be 67 mill/kWhr,
which is 10% lower than the busbar cost of AP 1000. Moreover since it only has one coolant
loop and no steam separator, steam dryer or recirculation pump the cost is further reduced (this
however was not taken into account). The breakdown of generation cost is shown in figure 10.
The depleted fuel turns out to be 3% enriched, which is the enrichment used in BWRs.
Therefore, the depleted fuel can be sold to BWR lowering their cost of fuel and our cost of fuel
below 4 mill/kWhr.

16. 16
Figure 10: Breakdown of the generation costs (in mill/kWhr) for MSWR design
VIII. SUMMARY
The initial interest of the group was a mixed-spectrum design. The goal to make the reactor
smaller was achieved with an output of 800 MWth. Thermal-hydraulic analysis showed the
viability of using supercritical water to take advantage of a mixed spectrum configuration. A
fuel choice of 8% enriched UO2 along with high burn-up will yield reduced operating costs of
around 67 mill/kWhr. Favorable reactivity coefficients demonstrate the safety of MSWR design,
and isotopic analysis show an increased proliferation risk. Overall cost analysis indicate that the
MSWR design is a viable reactor design that is capable of helping our country meet the
challenges of its increasing energy demands.
IX. REFERENCES
[Hu08] P. Hu, “Coupled Neutronics/Thermal-Hydraulics Analyses of SuperCritical Water
Reactor,” PhD Thesis, University of Wisconsin - Madison (2008).
[INE05] “Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power
Production,” INEEL/EXT-04-02530, Idaho National Engineering and Environmental
54
4
0.24
9
Construction
Fuel
Decomission
M&O

17. 17
Laboratory, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric
Company Award Number DE-FG07-02SF22533 (2005).