1. Travis A. Chapman
ENNU 808 I
Final Assignment
Encapsulated Nuclear Heat Source and mPower Generation: Comparisons with General
Design Criteria, the Super Critical Carbon Dioxide Cycle, and the Future
In the decades ahead the United States will likely see the advent of a new generation of
reactors. With the growing realization that carbon dioxide emissions must be controlled, coupled
with a growing population, new and unique demands for energy, and a growing emphasis on
safety, the market for clean affordable energy will continue to rise. The niche uses for small
modular reactors (SMRs) will grow into a larger share of the energy market, but current hurdles
before full realization must be overcome.
This report will compare two very different designs. The University of California-
Berkley, through Lawrence Livermore National Laboratory and Argonne National Laboratory
(along with other national laboratory support) have advanced a concept known as the
encapsulated nuclear heat source – heat pipe (ENHS-HP) [Ref. 1]. This SMR design is clearly
Generation IV, with novel components for almost every function. Given its maturity (between
feasibility and conceptual design stages), some of the details for this report will be assumed from
the literature. In contrast, Babcock and Wilcox (B&W) has partnered mPower Generation to
fully develop the mPower, an advanced, integral light water reactor (LWR) [Ref. 2]. Up until
2014 the design was likely the front-runner of SMRs to receive licensing attention from the U.S.
Nuclear Regulatory Commission (NRC). While many features appear novel, this design is an
evolution from previous LWRs and utilizes many of the same general components. A brief
description of each will be made.
While focusing on the ENHS-HP, practical licensing issues will be discussed and
compared with the mPower design. A selection of the NRC’s General Design Criteria (GDC)
[Ref. 3] will form the basis to analyze the realism in achieving one particular licensing
requirement. A brief analysis of the supercritical carbon dioxide (S-CO2) cycle will be made,
given its likely incorporation into the ENHS-HP design. Lastly, some conclusions and thoughts
for future work will be discussed.

2. Figure 1: ENHS-HP schematic [1]
Design Description - ENHS-HP: The ENHS-
HP is a small modular reactor combining several novel
features in an effort to increase availability, extend the
useful core life, improve operational safety through the
use of passive design features, and provide a
proliferation-resistant system that could be utilized by
both developed and developing nations. At its current
design stage, the ENHS-HP proposes using uranium
nitride fuel (although several other options are possible)
operating in the fast neutron spectrum. Fuel rods are
contained in a solid molybdenum alloy block, or
monolith, in an array alongside numerous heat pipes,
both oriented horizontally in the reactor vessel base. The
heat pipes act as the effective primary coolant for the
system. These heat pipes are designed to use sodium as
their working fluid, transferring heat from the core to an intermediate coolant. Liquid salt, a
blend of lithium-fluoride and beryllium-fluoride (F-Li-Be, known as flibe), is used as the
intermediate coolant. Once heat is transferred from the heat pipe heat exchange region the hot
salt is buoyed up through a central riser section by the significant density change, where it passes
through a tertiary heat exchanger before falling back through an outer annulus. The newest
ENHS-HP iteration proposes using S-CO2 as the tertiary working fluid to drive a direct-cycle
power generation cycle; some diagrams contained here reference the original steam generator
Rankine cycle proposal.
Key design characteristics of the ENHS-HP are listed below:
Rated core power 125 MWth
Nominal operating life 20 years between refueling
Linear Heat Generation Rate 9 kW/m
Vessel height / diameter 9m / 4m
Core height / length /width 1.767m / 1.767m / 1.5m
Primary coolant (heat pipe) Sodium
Intermediate/secondary coolant Flibe

3. Figure 2: mPower Schematic [2]
Tertiary coolant S-CO2
# Heat pipes (P/D) 7408 (2.0)
# Fuel rods (P/D) 11111 (1.0)
B&W’s mPower is one of the more mature Generation III+ designs. From the company’s
literature, the mPower is an integral LWR capable of producing 180 MWe per module, using a
standard Westinghouse low-enriched uranium
oxide fuel design, with a four year refueling
cycle. mPower incorporates an integral system
design: primary coolant and steam generation
components are all integrated in a single vessel.
The concept is to construct the major
components, including large forgings,
domestically for follow-on transport by rail or
barge. A site would incorporate two units
driving individual turbines with some common
infrastructure shared between reactors, including
spent fuel pools.
Key design characteristics of the mPower are listed below:
Rated core power 530 MWth
Nominal operating life 4 years between refueling
Linear Heat Generation Rate 11.5 kW/m
Coolant Volume 92 m3
Fuel (# assemblies, enrichment) 69 - 17x17 assemblies, <5 wt%
Primary coolant Water
Secondary coolant Water
Licensing and Regulatory Concerns: Practical licensing issues exist for both designs.
The ENHS-HP is a less mature technology with more challenges to overcome. Technically
defensible information related to many of the components does not yet exist, although testing
and experimental data are under development. In particular the performance of components in

4. abnormal or accident conditions will require further analysis to fully support a licensing effort.
Since many of the materials are both novel individually and in their integrated systems, there will
be significant effort needed to verify performance during normal, abnormal, and accident
conditions, as well as to substantiate the ability to continue safe operations through twenty years
or longer. The fuel cycle proposed for ENHS-HP will also require additional design, analysis,
and supporting information to achieve licensing. Such unique fuel blends have been produced
for one-off designs, such as early space and demonstration reactors, but the commercial ability to
produce nitride fuel using uranium and other actinide blends does not exist. Such fuel will
require characterization studies before acceptance by the NRC. The designers readily
acknowledge such limitations and provide suggested targets to focus near-term efforts on [Ref.
4].
The mPower design appears to have fewer technical challenges to overcome. From the
conceptual design forward the designers deliberately chose design characteristics and parameters
that would reduce the regulatory burden of certification. While some components are unique,
most have their genesis in early LWR designs and so are familiar to the NRC and generally fit
within the regulatory framework that exists today. The fuel is a previously certified
Westinghouse design that has extensive operating experience and supporting data; it is well
characterized in normal, abnormal, and accident operating regimes. Current testing efforts are
focused on gathering operational data on how integrated systems will work together,
benchmarking codes used in the development of the systems, and better understanding a handful
of unique components such as the integral control rod drive mechanisms [Ref. 5].
From a policy perspective the ENHS-HP is more dependent on changes to current
regulations that the mPower. Both design concepts are limited by currently perceived limitations
in the regulatory framework that do not scale with the difference between large LWRs and new
SMR designs. Both designers can claim that a reduced emergency planning zone is warranted,
and provide technical data to support such claims, but the existing regulations effectively
mandate 10 to 50 mile planning zones irrespective of power output. Regulations on source term,
security requirements, emergency planning, and common infrastructure are understood to be
policy issues of concern by both the industry, the regulator, the promoter (the Department of
Energy, DOE), and Congress. Efforts are in progress to address these issues, with input from all
stakeholders and some general horizons provided by the NRC, DOE, and Congress. It is likely

5. that integrated LWR designs like mPower will be the first to navigate a new regulatory structure,
or learn just how difficult seeking regulatory relief through technical justification will be. This is
not necessarily a bad thing, as Westinghouse and Southern Nuclear Company benefited from
voluntarily tackling design certification and combined licensing challenges with the NRC ahead
of other Generation III+ designs. Being first to market may have made new reactor viable, as the
AP1000 design is the most sought after for large LWR new construction in the U.S. [Ref. 6].
A challenge recently faced by mPower was the decision to pursue licensing under the
Code of Federal Regulations (CFR), Title 10, Part 52 regulations instead of their original concept
of a dual-track licensing action. At the time the company believed that licensing under 10 CFR
Part 50, which would have involved issuing a construction permit and later, after adequate
testing and verification was complete, an operating license, was a preferred choice. This
structure would allow them to effectively design and build at the same time, learning their
lessons in situ as construction advanced. Rather than try and answer potentially unlimited
questions by the regulator before shovel touched earth, they would be able to advance the design
to a reasonable point but maintain the flexibility of making small changes as challenges emerged
during construction. Once the best construction method and design changes were captured
through the construction permit stage, the applicant (in this case Tennessee Valley Authority,
TVA) would apply for their operating license while mPower submitted a more completed design
certification document in parallel. The decision to continue this course of action was in jeopardy
through 2014.
mPower ultimately felt confident that a sole 10 CFR Part 52 track was viable, and
potential applicants like the ENHS-HP will face similar decisions. Given the number of novel
components and limited construction experience of many systems, the idea of dual-track
licensing seems to make more sense. Unique components may be fabricated without the concern
of regulatory approvals from minor design changes; only ensuring such changes are adequately
justified, quality assured, and communicated in a timely manner to the NRC. Given the
likelihood of schedule delays due to developing so many unique things, the overall production
timeline would be more flexible for the licensee and regulator. Negative perception of
construction or fabrication problems may also be partly alleviated. When issues arose at the
Vogtle site during pre-combined license inspections of engineered soil and later in rebar
placement the perception became “Problems with a certified design” instead of the more truthful

6. “Challenges with doing first-of-a-kind projects” [Ref. 7]. In practice the construction permit
process accommodates such events easier, has been proven through several decades of large
LWR projects, and would likely be enhanced by modern project management tool and practices,
better quality assurance techniques, and many years of lessons learned. Guiding the regulator to
agree may be the most significant challenge of all.
Lastly, the base knowledge of the NRC and others in the nuclear enterprise is still at the
earliest stages for many Generation IV technologies. A design such as mPower contains almost
all well-known systems and component types, so the nomenclature and concepts are easily
understood. The processes by which the reactor operates are analogous to large LWR designs,
and so regulatory guides and supporting direction to reviewers are relatively easy to transfer into
the SMR design philosophy. However, the knowledge gap of the average NRC reviewer in the
area of metallic coolants, supercritical gas power cycles, heat pipe design and operation, and
non-uranium oxide fuels is likely large. Without design-specific review guides, the NRC would
likely be challenged to provide a review plan with confidence in the timeliness of their activities.
This is not a criticism of the agency, but a hurdle they must overcome to provide the meaningful
and effective review that a design certification deserves. Increasing the agency’s level of
knowledge while maintaining their current expertise in LWR designs and continuing to oversee
operating reactors is a recognized challenge [Ref. 8]
The subject of NRC interactions raises the question of cost for the licensee and designer.
The NRC operates under a fee-recovery framework, where applicants and licensees pay a
standard fee per hour and agree to pay penalties determined by the agency’s Office of
Enforcement when regulations are not adequately met. This can include the quality and
timeliness of information, and so even a new applicant may end up facing financial penalties
during the design certification process. The designer must balance the level of design maturity
with the NRC’s estimates of review time, factoring in the cost of new design or testing work
which may result from an early NRC review, as well as the NRC’s own desire to begin such
engagements as early as possible in order to get ahead of problems. For a small designer with
potentially limited means, this may prove too challenging a bar to reach. It is not surprising to
see a large engineering firm like B&W become the backer of the mPower design; they had a
better chance of weathering the steep financial curve of initial licensing. Regardless, even such a
power house was not immune to pocketbook pain, as the lack of a clear customer base ultimately

7. led them to throttle licensing activities despite an influx of funding from the U.S. government
[Ref. 9]. The decision of when and how to engage with the regulator is a pivotal point for any
design, and potential Generation IV applicants like the ENHS-HP must recognize that more risk
exists for them than their LWR counterparts.
General Design Criteria: As part of the application for a design certification the
applicants will be required to develop principal design criteria and show how their design meets
them. The GDC are a series of criteria previously developed by the NRC and industry to support
LWR design reviews. The specific purpose of the GDC are to “establish the necessary design,
fabrication, construction, testing, and performance requirements for structures, systems, and
components important to safety; that is, structures, systems, and components that provide
reasonable assurance that the facility can be operated without undue risk to the health and safety
of the public” [Ref. 3]. As written, the current GDCs must be addressed in an application. An
applicant must provide the basis and justification for any GDC that cannot be met, although there
is risk involved in this: the NRC determines, without the help of regulatory guides, what
constitutes sufficient justification. The NRC also clearly states that the GDC are imperfect, even
for LWR designs. From Reference 3, front matter:
The development of these General Design Criteria is not yet complete. For example,
some of the definitions need further amplification. Also, some of the specific design
requirements for structures, systems, and components important to safety have not as yet
been suitably defined. Their omission does not relieve any applicant from considering
these matters in the design of a specific facility and satisfying the necessary safety
requirements.
Some of the matters not yet fully defined include how systems important to safety (not
safety-class, but supporting the safety-class systems) are treated with respect to single-failure
criteria, redundancy, and diversity for passive components. Also of concern is how to
characterize possible loss of coolant accident initiation locations, given the varying types, sizes,
and orientations of possible breaks in a system. Lastly, addressing how systematic, nonrandom,
concurrent failures may affect redundant portions of the protection and reactivity control systems

8. remains incomplete. With the incorporation of more digital instrumentation and control systems
into future designs this last factor will certainly remain a concern.
In the case of Generation IV designs, the ability to translate novel design features into the
basis for meeting the GDC is difficult. While the final requirement of producing principle design
criteria is relatively easily met through an interactive design approach, the NRC’s specific
direction to address the GDC remains. The ENHS-HP designers admit that some criteria are
simply not met at this time: the ENHS-HP reactivity control system utilizes control slabs that are
not redundant, so criteria 26 and 27 cannot be met [Ref. 1]. For this report five criterion were
selected to discuss the challenges facing the ENHS-HP design, analogous of Generation IV
design concerns, and how mPower has potentially addressed them. Given the proprietary nature
of mPower’s submissions to the NRC, some aspects of the comparison will be simplistic or
assumed.
Criterion 32-Inspection of reactor coolant pressure boundary. Components which
are part of the reactor coolant pressure boundary shall be designed to permit (1)
periodic inspection and testing of important areas and features to assess their
structural and leak-tight integrity, and (2) an appropriate material surveillance
program for the reactor pressure vessel.
Challenges: The ENHS-HP reactor components are packed tightly and meant to stay in place
once initially assembled. Coupled with the intended long core life, this allows limited
opportunity for inspection from the inside. This is in contrast to LWR designs that inherently
have more opportunities due to shorter refueling cycles. Even the mPower design benefits from
a four year refueling interval; much shorter than the ENHS-HP. However, the integral nature of
the ENHS-HP means fewer components and connections to the primary vessel, so exterior
inspection methods will likely be simpler and more thorough. This scheme may not detect
internal defects like that experienced in the Davis-Besse reactor vessel head (an internal
corrosion attack which degraded the integrity of the reactor pressure boundary). A robust pre-
construction testing regime to support the assertions of material compatibility over the extended
life cycle will be required before a long-term license is issued. Also, instrumented systems to
detect material degradation will be of greater importance. Stress and strain gauges, sonic

9. transducers, and vibration monitoring may help support the case for alternative inspection
techniques and methods. Lastly, when the system is opened up the transparency of flibe may be
of benefit compared to liquid metals like sodium that would be opaque at operating temperature;
visual inspection is at least possible. The mPower, by comparison, assumes a typical LWR
inspection and testing methodology with the exception of being on an extended schedule to
match the longer refueling periodicity.
One method to alleviate the NRC’s concerns might be to intentionally reduce the first
units’ refueling interval in order to conduct more thorough in-vessel inspections. The period
would not even need to be specifically for refueling. While this would inhibit some of the
proposed benefits of proliferation resistance and long operating life, opening up the plant on-site
to conduct periodic inspections may pay back any lost operating revenue and increase
maintenance costs by providing objective evidence to support the proposed longer operating life
and lead to a stronger technical basis for allowing reduced inspection intervals for subsequent
units.
Criterion 34--Residual heat removal. A system to remove residual heat shall be
provided. The system safety function shall be to transfer fission product decay heat
and other residual heat from the reactor core at a rate such that specified acceptable
fuel design limits and the design conditions of the reactor coolant pressure
boundary are not exceeded.
Suitable redundancy in components and features, and suitable interconnections,
leak detection, and isolation capabilities shall be provided to assure that for onsite
electric power system operation (assuming offsite power is not available) and for
offsite electric power system operation (assuming onsite power is not available) the
system safety function can be accomplished, assuming a single failure.
Challenge: The ENHS-HP has two primary methods of removing decay heat. In a normal state
the design uses the regular heat flow path to bleed heat into the tertiary system: residual heat
from operation plus decay heat are transferred into the flibe, then through natural circulation to
the S-CO2 heat exchanger, then from the S-CO2 system to the ultimate heat sink. In the event of
an accident that defeats the S-CO2 system, the reactor vessel air cooling systems (RVACS) is

10. initiated. The RVACS works by drawing cold air through a stack into the annulus between the
reactor vessel outer wall and the underground silo that contains it. Convection (major factor) and
conduction (minor factor) transfer heat from the vessel wall to the air, which is buoyed up into
the exhaust stack. From conceptual design calculations the vessel height is a dominant factor in
achieving sufficient decay heat removal. With estimated decay heat of 1-5% of total power, the
ENHS-HP design can accommodate adequate heat removal with terrestrial ambient conditions
and achieve passive cooling [Ref. 10].
The mPower decay heat removal method is analogous to that found in most Generation
III+ designs. By incorporating an auxiliary steam condenser and automatic reactor coolant
system depressurization with passive injection, high and low pressure decay heat removal
systems, coupled with large volumes of water available for cooling, a design which maximizes
the amount of inventory in the vessel, and utilizing pool height differences to drive natural
circulation, the mPower achieves its design goal of passive decay heat removal. The company
claims the design is capable of coping for up to 14 days under station black out conditions [Ref.
2].
Criterion 38-Containment heat removal. A system to remove heat from the reactor
containment shall be provided. The system safety function shall be to reduce
rapidly, consistent with the functioning of other associated systems, the containment
pressure and temperature following any loss-of-coolant accident and maintain them
at acceptably low levels.
Suitable redundancy in components and features, and suitable interconnections,
leak detection, isolation, and containment capabilities shall be provided to assure
that for onsite electric power system operation (assuming offsite power is not
available) and for offsite electric power system operation (assuming onsite power is
not available) the system safety function can be accomplished, assuming a single
failure.
Challenge: One of the first issues the ENHS-HP will need to overcome is how to communicate
the lack of a containment. In their conceptual design the designers claim this criterion is not
applicable since they chose a confinement, vice containment, strategy. While truly meeting the

11. Figure 3: RVACS system diagram [10]
letter of the law, they will likely face criticism
about meeting the intent. Assuming the
confinement strategy includes appropriate
filtration, and is sized to remove the maximum
residual and decay heat through the RVACS, there
appears no reason it couldn’t meet the intent of
this criterion. The NRC has proposed a
methodology for incorporating a confinement
strategy into a design through six design questions
to justify the decision [Ref. 11]. DOE operates
numerous facilities with confinement, and research
reactors have used confinement successfully for
decades.
Criterion 41-Containment atmosphere cleanup. Systems to control fission products,
hydrogen, oxygen, and other substances which may be released into the reactor
containment shall be provided as necessary to
reduce, consistent with the functioning of other associated systems, the
concentration and quality of fission products released to the environment following
postulated accidents, and to control the concentration of hydrogen or oxygen and
other substances in the containment atmosphere following postulated accidents to
assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and features, and
suitable interconnections, leak detection, isolation, and containment capabilities to
assure that for onsite electric power system operation (assuming offsite power is not
available) and for offsite electric power system operation (assuming onsite power is
not available) its safety function can be accomplished, assuming a single failure.
Challenge: The mPower design uses passive hydrogen recombinators to ensure flammable limits
are not exceeded in containment in the event of an accident. Like all of large LWR designs the
containment itself acts to bottle up any release gases. These features are reasonable well

12. understood and tested through decades of experience and lessons learned from events like
Fukushima-Daiichi. For Generation IV designs, the designer must communicate how the
systems meet the intent of this criterion.
Without zircaloy and water in the primary system, the potential for hydrogen generation
is practically eliminated. S-CO2 and flibe are inert and therefore pose no flammability threat.
The fuel is designed to capture a large gas fission product inventory over the course of 20 years,
so the design is inherently more resilient at gas capture. This fuel region corresponds to the
adiabatic section of the heat pipes and is contained in the monolith, hence protecting it from
missile impact. While the ENHS-HP design seems to have adequately met this criteria, other
liquid metal reactor designs use sodium as a coolant and will require more substantial
justification to demonstrate how sodium-water interactions will be mitigated.
Criterion 62-Prevention of criticality in fuel storage and handling. Criticality in the
fuel storage and handling system shall be prevented by physical systems or
processes, preferably by use of geometrically safe configurations.
Challenge: The ENHS-HP is designed to be very robust and account for safety during assembly
and transport. In transit the absorbers and/or control slabs are latched in place. Interlocks
prevent withdrawing until a start-up temperature of 350°C is reached. The fuel rods are also
immobilized in the monolith. Crush scenarios will likely be part of the design analysis, but the
probability is reduced since there are fewer opportunities for overhead maintenance in the vessel.
The flibe is theoretically undisplaced by water (vessel is sealed), and the fuel is designed to
operate in the fast spectrum and unmoderated, reducing the potential for criticality by those
means. More significant concerns would be associated with the fuel development and
fabrication, and factory placement, which leads to a new set of challenges for the applicant: how
to parse out information between the reactor design and associated licensing with that of the
manufacturing license. It is unclear whether the NRC will license a design without a greater
understanding of the supporting processes. This may be especially true of the long-term waste
philosophy; such uncertainty stopped licensing activities at the NRC in the wake of Fukushima-
Daiichi and the lack of a comprehensive strategy for high level waste in the United States [Ref.
12].

13. Brayton Cycle Analysis: One innovation being pursued by Generation IV designers is
use of the Brayton thermodynamic cycle to improve efficiency of the secondary system.
Generally two working fluids are proposed: helium, when used as the primary coolant in a high
temperature gas-cooled reactor, or S-CO2 in an indirect cycle. The proposed benefits of S-CO2
include increased thermodynamic efficiency (since S-CO2 is denser it requires less energy to
compress than helium), being relatively inert and non-toxic, and having a greater estimated
reliability in the cycle due to fewer and smaller turbomachines. The potential heat range of S-
CO2 compatible systems could lead to hydrogen production through high temperature
electrolysis, although further advances in material compatibility and longevity are necessary to
realize this benefit [Ref. 10].
The basic principles of S-CO2 cycle are based on ideal gas laws no longer applying since
isentropic compression and expansion are not accurate assumptions at the high temperature and
pressure conditions for CO2 as a supercritical fluid; real gas effects must be accounted for.
Instead, the system is characterized by polytropic compression and expansion through a series of
machines. The inlet and outlet conditions of the series of turbines and compressors are balanced
to tune the efficiency of the cycle. At the basic level it is desirable to keep the fluid temperature
low in compression since the specific volume is proportional to temperature; this contributes to
reduced compressor work and may be achieved through use of intercoolers. When combined
with high operating pressures, near 20 MPa, this also allows for smaller turbomachinery for a
given system and reduction in compression work by up to 15% compared to helium. [Ref. 10]

14. The temperature-entropy diagram above [Ref. 13] provides an overview of the ideal
Brayton cycle. The diagram also shows one of the design choices for those selecting the S-CO2
cycle: use of a single shaft machine versus a split-shaft, or multiple-machine, design. Since both
the turbine and compressors rotate at the same speed, inlet/outlet conditions for each individual
machine are not optimized. However, the overall cost of the system is reduced since only one
rotating machine is being fabricated with multiple parts. There are a reduced number of seals,
bearings, and casings to be purchased, which saves capital costs. And once tuned, the system is
at peak efficiency for baseload operations and maximizing inventory control. Some designs use
reheating and intercooling to increase the overall efficiency at the cost of extra components.
Intercoolers would provide discrete changes to the compression section of the diagram above
(between points 6, 1, 2, 2a) while reheating provides an additional jump before the reactor side
(between points 2a and 3).

15. In order to advance the likelihood of realizing S-CO2 Brayton cycles, development of
appropriate balance-of-plant codes must continue, with additional benchmarking to real-world
systems. Equations of state for S-CO2 may require additional analysis, but most sources appear
to agree that sufficient data exists to continue code development with a high degree of
confidence. The turbomachinery presented in conceptual designs is being developed for other
industries, so the industrial capability to fabricate such components exists now. However,
assumed efficiencies in the range of 90% could be improved; it is likely the further research into
higher temperature applications (above 700°C) will improve upon current designs. Additionally,
the compatibility of S-CO2 with proposed materials like Inconel and other high-temperature
alloys will be necessary to understand creep rates and corrosion effects; the current research
indicates that such concerns are within a realistic safety envelop for the proposed material uses
[Ref. 10].
Conclusion: Both the mPower and ENHS-HP designs advertise substantial benefits and
improvements upon existing LWR technology. It is clear that designs like mPower will have an
easier road ahead for licensing, given their familiarity to existing reactors. Their main advantage
comes in the form of reduced initial cost, and therefore potentially opening up new markets for
nuclear energy as compared to large LWR designs. From a technology standpoint there appears
to be little significant improvement in plant operating efficiencies, although the improvements in
safety are certainly in-line with NRC expectations and public desires.
The ENHS-HP represents an almost revolutionary take on nuclear energy production.
While such an idea could yield increasing benefits as fabrication capacity and utilization
increase, the initial hurdle of regulatory acceptance and starting up one-of-a-kind manufacturing
are significant barriers to market entry. Market forces (like emission regulation) and government
involvement (like government cost share and research grants) may help ease some financial
burden from an applicant.
The NRC has made efforts to come alongside industry stakeholders in finding solutions
for many of the challenges stated earlier. Public meetings to collaborate on proposed
methodologies and new ideas have strengthened both the regulatory consistency (good for
industry) and effectiveness of the requirements (good for the regulator). Efforts like the
combined NRC-Westinghouse-Southern Nuclear Company Inspections, Test, Analyses, and
Acceptance Criteria (ITAAC) Closure Pilot Program in 2010 provided an opportunity for all

16. parties to benefit from demonstrating new concepts together [Ref. 14]. However, applicants
must also understand that any activity involving rulemaking, such as certifying a new reactor
design, is no small matter. The agency’s many parts provide a great deal of diligence before
making such long-lasting decisions. There may be a subtle warning in that it took almost two
decades from the time ITAAC were officially adopted in 10CFR Part 52 till a pilot was
conceived to determine how ITAAC closure would actually occur.
While the NRC is open to test cases for SMR application challenges like common
infrastructure and manufacturing licenses [Ref. 15], there is something to be said for how DOE
has managed unique and novel nuclear applications. Over the course of its history they have
gained more experience with novel reactor designs than almost anyone; almost all of their
reactors are effectively one-off experiments or demonstration facilities. The framework of
nuclear safety under 10 CFR Part 830 provides a tailored approach to DOE facilities which may
provide more favorable circumstances for unique applications. With space at the national
laboratories generally available, a supportive Office of Nuclear Energy with experience in test
reactor facilities, and significant research and development resources in close proximity, it is no
wonder some designs are pursuing initial siting on DOE land. Before throttling their funding,
B&W and TVA were quickly moving towards an early site permit on the Clinch River at Oak
Ridge, TN. While the proposal was for an initial license through the NRC for commercial
electricity production, there may have been benefit to building a true demonstration plant that
could be later converted for commercial operations. After developing the required analyses and
documentation to support a DOE authority to operate, the barrier to NRC approval may have
been lower and not that far off.
Applicants such as the ENHS-HP may benefit was further research into unique regulatory
applications to reduce their initial licensing burden. Activities like NRC pilots or working
through DOE may yield benefits compared to purely traditional licensing processes. The
technical concerns discussed here may appear overwhelming, but likely have achievable
solutions given sufficient effort. The regulatory challenges appear much more significant, but
also have reasonable solutions if the right conditions exist.

17. Sited References:
[1] Greenspan, Ehud, et al, “Solid-Core Heat-Pipe Nuclear Battery Type Reactor”, University of
California Berkley, September 30, 2008.
[2] Temple, Robert, “B&W mPower Program IAEA SMR Technical Meeting”, September 3,
2013, accessed at: http://publicaa.ansi.org/sites/apdl/NESCCDocs/NSECC%2012-036%20-
%20BM%20mPower%20Reactor%20Design%20Overview.pdf, accessed on July 10, 2015.
[3] U.S. Nuclear Regulatory Commission, Code of Federal Regulations, Title 10, Part 50,
Appendix A “General Design Criteria for Nuclear Power Plants”, accessed at:
http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-appa.html, accessed on
August 18, 2015.
[4] Brown, N.; Carelli, M.; Conway, L.; Dzodzo, E.; Greenspan, E.; Hossain, Q.; Saphier, D.;
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General References:
[1] Oh, Chang; Lillo, Thomas; Windes, William; Totemeir, Terry; Moore, Richard,
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Testing Material Compatibility”, Idaho National Engineering and Environment Laboratory,
INEEL/EXT-04-02437, October 2004.
[2] V. Dostal, M.J. Driscoll, P. Hejzlar, “A Supercritical Carbon Dioxide Cycle for Next
Generation Nuclear Reactors” March 2004,