them must be processed.
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
133.1.2. Fresh Fuel shipment
Fuel rods for B-VHTR and fuel pebbles for P-VHTR are put into containers and shipped from
fuel fabrication facilities to reactor sites. Adequate C/S system such as sealing and PP need
to be applied to containers to ensure continuity of knowledge according the sensitivity grade
of the nuclear material being shipped. Note that there are no current domestic or internationally
licensed shipping container for transporting large quantities of HALEU fuels.
3.1.3. Fresh Fuel receiving
Broken fresh fuel elements should be segregated and must be stored separately by the user
for shipment back to the supplier for recycling as un-irradiated scrap. The C/S system for fresh
fuel shipment must be confirmed upon fresh fuel receiving. The nuclear material in the broken
fresh fuel elements is not attractive because the amounts are small and the material is still in
the form of coated fuel particles.
3.2. System elements related to reactor site of type B-VHTR
PR of B-VHTR is based on item accountancy. It is possible to imprint an ID on each fuel block,
so the safeguards approach has many similarities with the safeguards of LWRs. All system
elements related to a reactor site are confined within the reactor building as shown in Figure
11 [26]. All movements of fuel can be monitored by the surveillance cameras. Fuel storage
racks of the fresh fuel storage and spent fuel storage areas are sealed after handling fuel
therein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors
in the door valve. Movement of the fuel handling machine is slow due to its mass of more than


therein. Fuel inventory in the reactor core is verified by measuring the fuel flow with detectors
in the door valve. Movement of the fuel handling machine is slow due to its mass of more than
100 tons. This movement can be followed by the surveillance cameras whose data should be
continuously transferred to mitigate potential Cyber-attacks.
3.2.1. Fresh fuel storage on site
Fuel blocks are assembled by inserting fuel rods into pre-formed holes in the graphite blocks
in the reactor building. The on-site movement of fuel blocks of the B-VHTR is shown in Figure
11. The fuel blocks are stored in the fresh fuel storage rack until such time as the blocks
scheduled for reloading are returned to the reactor core. An adequate C/S system such as
surveillance cameras and PP should be applied to the fresh fuel storage area, the refueling
machine, and the spent fuel storage area for continuity of knowledge. Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
14Figure 11: Movement of fuel blocks in reactor site of B-VHTR [26]
3.2.2. Refueling Machine for fresh fuel loading and spent fuel discharging
This paragraph refers to HTTR as this is considered fully representative of B-VHTR [27].
StandpipeVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper
15The fresh fuel blocks are taken into the refueling
machine from the fresh fuel storage, and then the
refueling machine is lifted and moved onto the door
valve over the reactor with the crane. The fresh fuel
blocks are loaded into the vertical empty space from
where the spent fuels have been taken out. The IDs
of fuel blocks are confirmed at time of loading of fresh


blocks are loaded into the vertical empty space from
where the spent fuels have been taken out. The IDs
of fuel blocks are confirmed at time of loading of fresh
fuel. The spent fuel blocks in the reactor are taken into
the revolver-rack of the refueling machine and moved
to a spent fuel storage facility by the crane before the
fresh fuels are loaded. The control rod driving device
and the pair of control rods must be removed before
refueling. Replaceable side reflectors and fuel blocks
are handled using the refueling machine. They are
passed through the door valve and the stand pipe at
the upper part of the reactor core for any refueling.
Fuel reloading in light water reactors (LWRs) is
performed in water that provides a radiation shielding
effect. However, the coolant of B-VHTR is helium and
has no radiation shielding effect. For this reason, the
fuel exchange for B-VHTR is performed by remote
control of the gripper of the refueling machine, since
the fuel cannot be directly viewed. It is also necessary
to incorporate a radiation shielding function in the
refueling machine because it will contain the spent
fuel block in the revolver-rack. For this reason, its
mass exceeds 100 tons. When the refueling machine
is moved from the upper part of the reactor, the coolant (helium) in the reactor should not be
allowed to leak. A door valve is provided between the refueling machine and the standpipe to
prevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is
shown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door


prevent leakage of the coolant (helium) in the reactor to outside. The position of door valve is
shown in Figure 12 [27]. Neutron detectors and gamma ray detectors are attached to the door
valve, since the door valve is necessary to move out core components (anything such as spent
fuel blocks, replaceable side reflectors and irradiated experimental material from the reactor).
3.2.3. Reactor Core
The core consists of hexagonal columns of fuel blocks, control rod guide blocks and
surrounding replaceable side reflector, constituted of blocks. The permanent reflectors
surround the replaceable side reflectors. Fuel blocks are stacked vertically in several stages,
and replaceable reflectors are placed above and below them. In order to accommodate the
decrease in reactivity associated with fuel depletion as the reactor is operated, by design the
reactor core is loaded with adequate excess reactivity at the beginning of operation. Each fuel
block is engraved with a unique ID and loaded to a predetermined position in the reactor core.
After a certain period of operation, the spent fuel block is taken out through the stand pipe
using the refueling machine. The coolant flows through the flow paths in the graphite blocks
and is heated. The heated coolant is brought into a hot plenum and guided to outside of the
reactor pressure vessel at a temperature of 700 to 950 °C.
The control rods are suspended from the control rod drive mechanism in standpipes above the
core and inserted into the core or reflector, as needed. Control rod guide columns for inserting
control rods are provided in the core.
Any undeclared movement of the refueling machine would be detected by surveillance
cameras. Furthermore, irradiation of undeclared material is detectable with the neutron and


control rods are provided in the core.
Any undeclared movement of the refueling machine would be detected by surveillance
cameras. Furthermore, irradiation of undeclared material is detectable with the neutron and
gamma ray detectors attached in the door valve used for introducing and removing materials
into and from the core. The combination of neutron and gamma ray detectors, shown in Figure
Figure 12: Door valve and refueling
machine [27]Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
1613 [26] makes it possible to distinguish the nature of materials introduced into the core or
removed from it as nuclear materials and non-nuclear materials. Data obtained by both
detectors should be continuously transferred to safeguards inspectorates to avoid Cyber-
attacks or other tampering.
3.2.4. Spent fuel storage on site
The spent fuel blocks are stored for a certain period in racks of the spent fuel storage facility
that includes a water-cooling system in order to remove decay heat. The movement of spent
fuel blocks can be detected by an adequate C/S system such as sealing the lid on the top of
the storage racks and monitoring them with further surveillance cameras.

Figure 13: Neutron detectors and gamma ray detectors installed in the door valve for B-VHTR [26]
3.2.5. On-site radioactive waste storage
Substances that do not contain nuclear fuel materials, such as activation products, are stored
in the on-site radioactive waste storage facility, so their attractiveness from the PR viewpoint
is low. However, such materials should be protected from a PP viewpoint.
3.2.6. On-site radioactive waste storage
The spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal


3.2.6. On-site radioactive waste storage
The spent fuel blocks in storage are put in fuel transfer casks for shipping to the final disposal
or to the reprocessing plant after cooling for a certain period in the spent fuel storage on site.
Continuity of Knowledge (CoK) is maintained by use of adequate C/S systems, such as sealing
transfer casks, and adequate PP is also applied, such as protection by guards. The spent fuel
blocks are not attractive as sources of explosive nuclear materials used for NED due to the
poor quality of the materials and the great difficulty of reprocessing. But they may be attractive
from the view point of “radiological sabotage" due to their high radioactivity content.
3.3. System elements related to reactor site of P-VHTR
For safeguards purposes, P-VHTR is regarded as a quasi-bulk type facility. In the past,
however, there have been cases where safeguards were implemented by assigning IDs to
pebbles at the research reactor level, but not for online monitoring during the re-loading
procedure. However, it is usually sufficient for safeguards to just count/keep track of the
number of fresh fuel and spent fuel pebbles as they are moved from and to their respective
storage systems. The operating temperatures and high pressure of the system would make it
difficult to divert fuel out of the core.
3.3.1. Fresh fuel storage on site
 IAEAVery-High-Temperature Reactor (VHTR)                 PR&PP White Paper
17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S
system and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded


17The containers with fuel pebbles are stored in the fresh fuel storage under an adequate C/S
system and PP for P-VHTR. These fuel pebbles are moved to the charging room to be loaded
into the reactor core. The number of fuel pebbles should be counted if it is possible, and the
movement of the fuel pebbles from the fresh fuel storage to the charging room should be
observed via surveillance cameras. Diversion or otherwise acquisition of fuel pebbles is not
attractive due to the difficulty of recovering the nuclear material from fuel elements and
because the amount of nuclear material in them is small.
3.3.2. Recirculation of irradiated fuel pebbles
The fuel pebbles have no identification
numbers and are loaded randomly into
the reactor core. The amount of nuclear
material in every fresh fuel pebble is the
same (heavy metal loading and
uranium enrichment level). If initially
fueled entirely with fresh fuel pebbles,
P-VHTR cores would become critical
with a small total volume of fuel.
Therefore, graphite balls and boron
balls containing no fuel are loaded into
the core along with the fresh fuel in
order to maintain the desired height of
fuel in the core. With fuel depletion,
graphite balls and boron balls are
removed, and fresh fuel pebbles are
loaded in, as the core evolves from the
initial loading core to the equilibrium
core. Figure 14 shows the movement of
the fuel pebbles in the reactor [28]. Fuel
pebbles are taken out from the core
through the fuel pebble discharging
tube. Failed fuel pebbles are separated


the fuel pebbles in the reactor [28]. Fuel
pebbles are taken out from the core
through the fuel pebble discharging
tube. Failed fuel pebbles are separated
and are stored in the scrap containers.
Sound fuel pebbles are led to the
dosing wheel where their fuel burnup
levels are measured. The fuel burnup is
evaluated by measuring the Cesium-
137 gamma ray peak with a gamma
spectrometer. However, it has recently
been suggested that Cs-137 would not necessarily be a good burnup indicator, and Zr-95, Nb-
95, and La-140 may provide more appropriate burnup instead [29]. Further research is needed.
The fuel pebbles that have achieved a predetermined burnup level are discharged through the
discharge tube and are led to containers in the discharge compartment as spent fuel pebbles.
On the other hand, fuel pebbles that have not reached the predetermined burnup level are
transported pneumatically to the upper part of the core and reloaded at the top of the core.
This reloading is repeated until the fuel pebble reaches the predetermined burnup level. The
number of reloading cycles is typically between 5 to 15. The precise figure depends on the
specific design, reloading pattern and target burnup levels. High fuel burnup is achievable due
to the highly stable characteristics of coated fuel particles and due to nearly continuous fuel
loading. It is higher than the burnup of LWRs as well as B-VHTR. A burnup level of 100 GWd/T
is achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance
features due to large isotopic fraction of high content in plutonium that produces a high level


is achievable for the spent fuel of P-VHTR, and it results in superior proliferation resistance
features due to large isotopic fraction of high content in plutonium that produces a high level
of decay heat. The physical inventory verification in the reactor core is performed by controlling
the number of fresh fuel pebbles loaded and accounting for the spent fuel pebbles discharged
Figure 14: Movement of the fuel pebbles
for P-VHTR [28]Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
18and the number of failed fuel pebbles discharged to the scrap container. Access to the reactor
cell will be controlled by an adequate C/S system and PP.
3.3.3. Spent fuel storage on site
The spent fuel pebbles in containers are stored for a certain period in the on-site spent fuel
storage. The containers are cooled in order to remove decay heat. The movement of a
container can be observed using an adequate C/S system, such as sealing the containers and
monitoring the storage area with surveillance cameras. The amount of fissile nuclear material
(U-235 and Pu-239) in the spent fuel pebbles is small due to high burnup and high content of
decay heat-generating Pu isotopes. One of interesting discussions is the treatment of
damaged pebbles. In general, the damaged pebbles are added to the spent fuel storage, i.e.
there is no separate waste storage of broken pebbles planned for the PBMR design. Damaged
pebbles are always to be expected to occur during irradiation in the reactor and cannot be
returned for further cycles through the core, so they had to be classified as spent fuel. However,


pebbles are always to be expected to occur during irradiation in the reactor and cannot be
returned for further cycles through the core, so they had to be classified as spent fuel. However,
since those pebbles are less burnt, they are potentially more attractive in terms of Pu quality.
3.3.4. Radioactive waste storage on site
Substances that do not contain nuclear fuel materials, such as activation products, are stored
here, so their attractiveness from the PR viewpoint is low. However, these waste materials still
need to be protected from a PP viewpoint.
3.3.5. Spent fuel shipping
The spent fuel pebbles in containers will be transferred to the final disposal or to the
reprocessing plant after cooling for a certain period in the spent fuel storage area on site. COK
is ensured using an adequate C/S system such as sealing the containers and monitoring the
movement of the containers with surveillance cameras. The spent fuel pebbles are not
attractive from the point of view of nuclear materials for use for NEDs, but they may be
attractive from the view point of “radiological sabotage" due to their high radioactivity content.
See the section 5.2 for more discussion.
3.4. System elements related to reprocessing site or final disposal site of spent
fuel for B-VHTR and P-VHTR
The treatment of spent fuel of both B-VHTR and P-VHTR can be divided into (1) direct final
disposal and (2) reprocessing. The direct disposal option is attractive because the coatings of
coated fuel particles themselves are “containers” for the fission products and the fuel itself
possesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR


coated fuel particles themselves are “containers” for the fission products and the fuel itself
possesses high mechanical and chemical stability. Thus, the direct final disposal of the VHTR
fuel has reduced environmental and public impact.
Furthermore, the reprocessing of VHTR fuel is not considered attractive. The reason is that
metric tons and tens of cubic meter quantities of carbon encasing coated fuel particles would
have to be removed using either grind-leach, burn-leach of electrolysis in nitric acid if
reprocessing were to be performed. However, these technologies have still not been
demonstrated at industrial level. For this reason, spent fuel of VHTR has low attractiveness for
diversion / acquisition and / or processing as nuclear material. Spent fuels from the VHTR may
potentially still be attractive for radiological sabotage due to their high content in radioactive
materials that results from their high fuel burnup levels. The physical robustness of VHTR fuel
is favorable in this respect, making it more difficult for a potential adversary to achieve
widespread dispersal. The proliferation resistance features corresponding to the reprocessing
of the spent fuel of VHTR mentioned-above are valid not only for spent fuels of LEU-fuel, but
also for that of LEU / Th, LEU/Pu (MOX), Pu / Th MOX with high burnup. Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
193.5. Diversion targets
The key proliferation resistance feature of the VHTR is the fuel itself. The extraction of a
significant quantity (SQ) of either indirect-use U-235 from LEU (75 kg) or direct-use U-233 and
plutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of


plutonium (both 8kg) from VHTR fuel will require the processing of metric tons and tens of
cubic meter quantities of carbon encasing coated particles using either grind-leach, burn-leach,
or electrolysis in nitric acid. A background report [14] that supported the compilation of the
original VHTR white paper (published in 2011) discussed diversion targets for the two fuel
forms, prismatic block and pebble. The following discussion is quoted from the background
report [14] with some modifications using the PBMR [16] and the GT-MHR [6-8] as example
plants for the P-VHTR and the B-VHTR respectively.
“Using the PBMR as an example, the diversion of an indirect-use significant quantity (75
kilograms) of U-235 in LEU in fresh pebbles would require, for the equilibrium core with a
pebble loading of 9 grams of LEU at 9.6% enrichment, 75,000/(9 * 0.096) = 86,806 pebbles or
~17.4 MT of fuel pebbles, which should be quite readily detectable even over time since that
is ~20 percent of a core loading.” “By comparison, for the prismatic core GT-MHR or MHTGR
using fuel elements with inscribed serial numbers for visual tracking, the diversion of an
indirect-use significant quantity (75 kilograms) of U-235 in LEU in fuel elements containing
~3.43 kilograms of LEU on average at 19.8% enriched would require 75/(3.43 * 0.198) = ~111
fuel elements or 13.5 MT of fuel elements, which would be ~15–16% of a GT-MHR core loading


fuel elements or 13.5 MT of fuel elements, which would be ~15–16% of a GT-MHR core loading
or ~17% of the MHTGR core loading.” “Thus, the mass ratio for the diversion of indirect-use
U-235 in LEU between fresh pebbles and fresh GT-MHR fuel elements is 17.4/13.5 = ~1.29
so that 29% more pebbles by mass would have to be diverted to obtain 75 kilograms of indirect-
use U-235 in LEU.”
“Because the fuel elements of PBMRs are quite difficult to track, the use of LEU-fueled PBMRs
has been examined by several researchers from the aspect of the attractiveness for diversion
of fully burned spent fuel, one-cycle-irradiated pebbles, and the use of special production
pebble.
“The calculation results for the plutonium isotopic fractions in the PBMR fully burned spent fuel
would likely be very close to those for the prismatic VHTR spent fuel where the prismatic fuel
is to be discharged at a burn-up exceeding 100 GWD/MT (or MWD/kg). The PBMR and
prismatic VHTR spent fuel will have slightly different plutonium isotopic compositions resulting
from differences in the thermal-neutron and epithermal-neutron energy spectra due to a
different moderator-to-fissile atom ratio and additional thermal and epithermal neutron self-
shielding due to the higher-density fuel compacting used in the prismatic fuel.” …..” It is
expected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular
HTR will have plutonium isotopic fractions very close to the values calculated for the PBM in


expected, however, that the spent LEU fuel from both the GA GT-MHR and Areva Modular
HTR will have plutonium isotopic fractions very close to the values calculated for the PBM in
Table 3.6.1.” It appears that the Pu will be of reactor grade in all cases by applying the fissile
material type metric of PRPP WG.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
20Table 1: Calculated plutonium isotopic fractions for PBMR spent fuel as a function of initial
enrichment and discharge burn-up [Table 4.1 from [14]]
“Because the PBMR recirculates a pebble up to six times through the core before it is
discharged to spent fuel storage at full burn-up (~92 GWD/MT), the question arises about the
diversion of an irradiated pebble after one cycle or the use of special pebbles designed as
target elements to produce plutonium.” “The analysis of the PBMR by PBMR (Pty) Ltd. [16]
shows in Figure 15 [14] the plutonium build-up per pebble and the relative isotopic content as
a function of recirculation.”
Figure 15: Plutonium build-up in a PBMR fuel element in an equilibrium core [14]
“Figure 15 indicates that at full burn-up each pebble will contain about 0.11 grams of plutonium
with the isotopics indicated, and, from this, it can be inferred that, at full burn-up (120 GWD/MT
in the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60–
70 grams of plutonium of similarly degraded isotopics.” The diversion of 1 SQ of direct-use Pu


in the GT-MHR), the prismatic fuel elements can be estimated to contain on the order of 60–
70 grams of plutonium of similarly degraded isotopics.” The diversion of 1 SQ of direct-use Pu
from pebbles at full burn-up requires 8,000/0.11 = 72,727 pebbles or ~14.4 MT of fuel pebbles.
It takes 8,000/65 = 123 prismatic fuel elements to secure 1 SQ of direct-use Pu.
“However, the LEU pebble in a PBMR is recirculated up to six times while the fuel element in
a GT-MHR or MHTGR is typically reloaded only once. From Figure 15, the plutonium content
of a pebble after its initial irradiation is given as ~0.047 grams (~74% Pu-239), whereas for the
GT-MHR there are no data quoted for the one-cycle-burned prism, but it is inferred that the
plutonium loading would be ~50 grams with less favorable isotopics than in the pebble after
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
21one cycle of irradiation. From this, a rough comparison can be made that it would take at least
~1050 pebbles diverted after the first cycle to equal the amount of less favorable plutonium in
a prismatic fuel element removed from a GT-MHR after the first irradiation.” Table 3.6-2 shows
a summary table indicating the amount of material needed to collect an SQ.
Table 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel
SQ 75 kg 8kg


a summary table indicating the amount of material needed to collect an SQ.
Table 2: SummaryDiversion Target U-235 from Fresh LEU Pu from Spent fuel
SQ 75 kg 8kg
Equivalent pebbles 86806 (17.4 MT) 72727 (14.4 MT)
Equivalent blocks 111 (13.5 MT) 123 (15.0 MT)Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
224. Proliferation Resistance Considerations Incorporated into Design
The fuel in reactor cores of B-VHTR and P-VHTR is not as accessible and visible as the fuel
in an LWR. Therefore, physical inventory verification of nuclear materials in the reactor cores
is carried out by measurement of fuel flows into and from the core. Major Material Balance
Areas and Key Measurement Points are shown in Figure 16 as an example. Also, adequate
C/S is necessary. Adequate counter-measures against cyberattacks are required to maintain
CoK by C/S.
Figure 16: Material Balance Areas and Key Measurement Points of B-VHTR and P-VHTR
Design Information Verification (DIV) and C/S are implemented to avoid concealment of fuel.
Direct transfer of the C/S signal to IAEA is recommended to enhance proliferation resistance.
As noted previously, the key proliferation resistance feature of the VHTR is the fuel itself. To
obtain a significant quantity of either indirect-use U-235 from LEU or direct-use plutonium, one
must process metric tons and tens of cubic meter quantities of carbon encasing fuel using
either grind-leach or burn-leach of electrolysis in nitric acid.
The high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature


either grind-leach or burn-leach of electrolysis in nitric acid.
The high burnup of the spent fuel of the VHTRs is also a key proliferation resistance feature
due to the high isotopic fraction of even plutonium isotopes generating large amounts of decay
heat and high dose rate. However, it is controversial.
Historically, it has been argued that the technical difficulty of fabricating nuclear weapons
depends on the isotopic composition of plutonium, in particular the amount of Pu-240. Although
there are several references, the one that summarizes the key points is by Pellaud [30].
For nuclear safeguards verification activities there is no distinction for Pu with less than 80%
Pu-238. However, the heat generated by Pu isotopic containing more than a few percent of
Pu-238 would substantially increase the technical difficulty related to the fabrication phase
(weaponization). Using a set of figures of merit (FOM) for attractiveness, Bathke, et al. [31]
estimated that about 8% Pu-238 is required to render the plutonium isotopic unattractive for
an unadvanced proliferant state that requires reliably high-yield nuclear devices, however it
remains attractive for both technologically advanced states, which can handle it, and
subnational groups for which high reliability might not be a requirement. However, these
arguments are founded on the assumption that the proliferants demand reliable yield. In the
case of unadvanced proliferant or non-state actors who do not pay attention to the yield, high
reliability might not be their requirement.
With those reasons considered, the current GIF PRPP WG methodology adopts weapon
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper


reliability might not be their requirement.
With those reasons considered, the current GIF PRPP WG methodology adopts weapon
Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
23grade, reactor grade, and deep-burn grade for Pu categorization [1]. The fact that Pu in
HTGRs’ spent fuel can achieve deep-burn is one of the notable features.
4.1. Concealed diversion or production of material
Diversion of large quantities of nuclear materials (U-235, plutonium or U-233) is detectable by
spent fuel accountancy based on radiation monitoring or fuel element counting, by C/S on fuel
storage, or by recorded reactivity deviations in reactor operations. The VHTR does not produce
readily accessible, attractive fissile material. The technologies for reprocessing coated fuel
particles are complicated and still require development.
4.1.1. Diversion of unirradiated nuclear material items
Once the fuel has been encased within fuel kernel of coated fuel particle and furthermore into
fuel elements (such as fuel compacts for B-VHTR or fuel pebbles for P-VHTR), diversion
becomes difficult. The latter (fuel elements) consist of coated fuel particles encased within
graphitized carbon. Note that fresh fuel fabrication should be performed under surveillance.
Once in fuel assembly (compact ball) form, the nuclear material is more difficult to retrieve due
to difficulty of separation of nuclear material from large amounts of graphite and of the strength
of the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B-
VHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S.


of the coatings of particles. Fabricated fresh fuel can be stored under C/S measures for B-
VHTR and P-VHTR. The theft during transportation of fresh fuel can be detected by the C/S.
The raw constituents are observed under the same C/S applied for fuel fabrication of LWR.
4.1.2. Diversion of irradiated nuclear material items
4.1.2.1. B-VHTR
The major irradiated nuclear material items are spent fuel blocks. Diversion of Pu is possible
by discharging fuel blocks after a short time of reactor operation and then reprocessing them.
The fuel blocks are unloaded through standpipes over the reactor pressure vessel. For such
an operation, both the refueling machine and the door valve are required because it is
necessary to maintain isolation between the reactor coolant and the outside atmosphere for
the B-VHTR. Unexplained or illicit movements of the refueling machine and the door valve by
the crane can be detected by surveillance cameras. Also, undeclared movements of
prematurely discharged fuel blocks are detected by the neutron detector and the gamma ray
detector in the door valve. Any discharged material from the reactor pressure vessel can be
identified as nuclear material or not. Nuclear material is indicated when signals of both neutron
and gamma ray are detected. If the material is non-nuclear, such as a surveillance sample,
then no neutron source is detected. Undeclared discharging of experimental nuclear materials
is detected in the same manner as fuel blocks.
4.1.2.2.  P-VHTR
Diversion of Pu may be possible using the continuous fuel loading feature through early
discharging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are


Diversion of Pu may be possible using the continuous fuel loading feature through early
discharging of fuel pebbles from the reactor core before even-mass-number Pu isotopes are
accumulated. However, this would be detected by the burnup measuring detectors.
Furthermore, it is technically difficult because the reprocessing process of VHTR fuel is still not
established and detection of this diversion route is possible if an appropriate C/S system is in
place.
4.1.3. Undeclared production of nuclear material
4.1.3.1. B-VHTR Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
24Undeclared production of nuclear material may be possible through the irradiation of fertile
nuclear material in irradiation holes in the core or replaceable side reflectors of B-VHTR. The
materials would be loaded and unloaded through standpipes over the reactor pressure vessel.
In the B-VHTR, it is not possible to directly access and to visually observe the fuel in the core
or the replaceable side reflectors as would be possible in LWRs where water above the core
is used as radiation shielding. For these reasons, a handling machine with radiation integrated
shielding function, such as the refueling machine, is necessary for any undeclared production
of nuclear material. Any unexplained or illicit movement of handling machines can be detected
by surveillance cameras in the reactor building. Moreover, ton quantities of fertile material
would need to be loaded illicitly to generate a SQ, and it is difficult to envisage this being
practical to achieve without detection.
It should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors


practical to achieve without detection.
It should be noted that B-VHTR could be used in a mode similar to that of the Magnox reactors
for producing weapon-grade plutonium. In this case, rod-type Magnox fuel containing metal
uranium would be inserted into some cooling holes of the graphite blocks instead of using
ordinary B-VHTR fuel rods based on coated particle fuel. In this way the difficulty of
reprocessing of VHTR fuel would be avoided, as the reprocessing methods for Magnox fuel
are well established. However, the reactors would have to be operated with low reactor coolant
outlet temperatures to protect the integrity of the Magnox fuel and ton quantities of Magnox-
type fuel would need to be irradiated. This would imply giving up efficient power production,
which would be detectable. It might to worth thinking that this mode of operation could be
dangerous because of accumulation of Wigner energy in the graphite blocks, but further study
is needed.
4.1.3.2. P-VHTR
The inlet pipes of fresh fuel pebbles, in the fuel charging room, can be used for loading target
pebbles and for the access to the core region of a P-VHTR. However, these pipes cannot be
easily used for loading illicit material for the undeclared production of nuclear materials due to
the length of, and many curves in, the fuel loading path. Pebbles with diameter of 6 cm could
be loaded into the pipe. It is very important to confirm in Design Information Verification that
there are no access holes into the pipes except at the fresh fuel pebble loading location,
precluding any other pipe access into the reactor core. Irradiation of fertile materials covered


there are no access holes into the pipes except at the fresh fuel pebble loading location,
precluding any other pipe access into the reactor core. Irradiation of fertile materials covered
with graphite or carbon that look like fuel pebbles is possible. But such pseudo-fuel spheres
may break during movement through the core and would be difficult to remove. Furthermore,
such pseudo-fuel without ceramic coatings would release unexpected high radioactivity into
the primary cooling system at high temperature operation, which would be detectable. In
addition, there would result many operational problems, which would also be detectable and
require explanation. Tton quantities of heavy metal would need to be irradiated in the core to
generate 1 SQ. Finally, it is important to recognize that the presence of target breeding pebbles
in the core will alter the balance between fresh fuel demand and energy production in a way
that is detectable long before a significant quantity of fissile material is accumulated [32-35].
In addition, the identification of such breeding pebbles by the gamma measurement is much
more difficult than for regular fuel.
4.2. Breakout
As mentioned in section 3.4, reprocessing has not yet been demonstrated for the coated fuel
particles at industrial scale. In the presence of multi-lateral contractual provisions, for example
adhering to the guidance of the international Nuclear Suppliers Group (NSG), for the supply of
fresh fuels and the take-back of spent fuels for an exported VHTR, the issue of breakout is
further mitigated since there will be either no such material, or limited quantities of material, to
be reprocessed in the host states.
4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper


be reprocessed in the host states.
4.2.1. Diversion of existing nuclear material Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
25As mentioned in Section 3.4, the key proliferation resistance feature is the use of coated fuel
particles embedded within a graphite matrix. Therefore, diverting existing nuclear material from
VHTR fuels is difficult, lengthy and costly, regardless of the implementation of safeguards and
PP for the fuels. Since the reprocessing technology is not developed to industrial level,
extraction of nuclear material is significantly difficult. Moreover, the high burnup of spent fuel
from VHTRs is also a key proliferation resistance feature due to the presence of plutonium
isotopes that produce large amounts of decay heat. Pu in high burnup spent fuel contains
considerable even-numbered Pu isotopes, i.e. Pu-238, 240 and 242, whose decay heat
negatively affects use as a NED. Note that the diversion of raw material before being coated
with carbon and silicon carbide would be the easiest pathway for the processing of nuclear
materials to be used in the fabrication of nuclear explosive devices. However, this is not a
VHTR-specific problem, but a concern for all types of nuclear reactor systems.
4.2.2. Production of the necessary weapons usable nuclear material
As mentioned in section 3.4, the key proliferation resistance feature of VHTRs is the use of
coated fuel particles embedded within graphite matrix. It is necessary to process metric tons
and tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of
nuclear material necessary for production of weapons.
4.3. Pu Production in clandestine facilities


and tens of cubic meter quantities of carbon encasing the fuel kernels to obtain the amount of
nuclear material necessary for production of weapons.
4.3. Pu Production in clandestine facilities
High quality graphite with very low impurity levels is used in the technology of the B-VHTR and
P-VHTR.  This high quality graphite can be used for gas-cooled reactors in which weapons-
grade plutonium can be produced from natural uranium. Therefore, the consumption of large
amounts of nuclear-grade graphite should be controlled. For that reason, nuclear grade
graphite is controlled according to NSG lists part 1.
Operation of the clandestine facilities (reactor and fuel reprocessing) could be detected by
environmental sampling under the international safeguards regimes.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
265. Physical Protection Considerations Incorporated into Design
This section provides a qualitative overview discussion of the aspects of VHTR systems and
their design that create potential benefits or problems from the point of view of potential threat
by sub-national actors.
5.1. Theft of material for nuclear explosives
Plutonium in the spent fuel of LEU cycles and U-233 in that of future LEU/Th cycles are
attractive for the NED production. However, these nuclear materials in spent fuels are
accompanied with fission products, which are highly radioactive and make it difficult for
terrorists to steal the materials. Moreover, the nuclear materials are encased inside the coated
fuel particles. In these coated particles, the material of interest would be quite dilute so that the
theft of a significant quantity would require the theft of metric tons of contaminated graphite
and/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant


theft of a significant quantity would require the theft of metric tons of contaminated graphite
and/or graphitized carbon containing the coated fuel particles. Obtaining access to a significant
quantity of plutonium or U-233 in the stolen spent fuels would require substantial effort for
reprocessing. Furthermore, plutonium with a high inventory of the plutonium isotopes other
than Pu-239 is not attractive for the manufacturing of NEDs (e.g. high decay heat). U-233 with
hundreds of parts per millions of U-232 is not attractive due to high radioactivity and to the
necessity of further chemical cleaning to remove radioactive decay products. For those
reasons, the intrinsic qualities of VHTR spent fuel make it undesirable as a target for theft by
a sub-national group for use as nuclear explosive.
5.2. Radiological sabotage
VHTRs are designed such that the fuel temperature is maintained below fuel-damaging
temperatures under all conditions of normal operations and accident situations, including
beyond-design-basis events.  The design vision is that, even if the safety-related reactor cavity
cooling system were to malfunction, decay heat in the core would still be removed to the
external wall of the reactor vessel. As a result, the fuel temperatures in the core do not exceed
the levels that would cause the loss of the primary containment provided by the SiC coatings
over the fuel kernels.
The ultimate radiological sabotage act for reactors is that of an insider or an intruder trying to
cause radiological exposure by inducing a large power excursion. For both P-VHTR and B-
VHTR designs, appropriate physical protection and controls must be in place to prevent such
acts. These designs have several safety benefits from the very high temperature tolerance of


VHTR designs, appropriate physical protection and controls must be in place to prevent such
acts. These designs have several safety benefits from the very high temperature tolerance of
the fuel and the strong negative temperature power coefficient.
Another relevant discussion is that both VHTRs are extremely resilient to this kind of terrorist
attacks because passive heat removal, or reactor cavity cooling system (RCCS), by air cooling,
water or a combination of both is available when a loss of coolant happens.
The high burnup levels in the spent fuel of both VHTR types is one of the key proliferation
resistance features due to high radioactivity. However, spent fuels of VHTRs may be attractive
for Radiological Dispersion Device (RDD) due to high radioactivity resulting from the high
burnup. Below is the discussion of RDD for both P-VHTR and B-VHTR:
In the case of P-VHTR, the quasi-bulk fuel form may be attractive for terrorists when
considering the possibility of dispersal of the spent fuel. Protection of spent fuel on the
reactor site will be important. This should also be considered in PP when transporting
spent fuel by land.
In the case of B-VHTR, the item-type fuel allows its PP to be similar to that of LWRs.
Moreover, TRISO is considered to be very resistant to scattering and therefore more
robust against RDD-type terrorism than LWRs.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
27Finally, some points to be considered for the PP of VHTR are listed referring to the previous
VHTR white paper:
Quality controls at the fuel fabrication plant in the supplier nation.


27Finally, some points to be considered for the PP of VHTR are listed referring to the previous
VHTR white paper:
Quality controls at the fuel fabrication plant in the supplier nation.
Proper maintenance, inspection, and protection of (1) the helium supply and the helium
supply station to prevent the introduction of corrosive chemicals, (2) the primary coolant
contaminant monitoring equipment to detect the introduction of such chemicals, and
(3) the helium purification system to remove contaminants.
Careful maintenance, inspections, testing and protection of reactivity control systems
to assure the capability to achieve safe hot and cold shutdown and, if required,
accomplish the same function from a secure remote location.
Physical protection is required of and controlled access to fresh and spent fuel storage
locations, the inbound and outbound transportation loading systems, the transportation
of the fresh fuel from the fuel fabrication facility, and the spent fuel to its processing or
disposal facilities.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
286. PR&PP Issues, Concerns and Benefits
The key areas of known strengths of the VHTR concept at this time are its robust fuel form,
with fissile material strongly diluted in carbonaceous material, high burnup and the use of the
once-through LEU fuel cycle, which all make VHTR fuel unattractive for proliferation purposes.
When considering PR, B-VHTR will have item-based safeguards applied, while P-VHTR
safeguards are quasi-bulk, so differing safeguards approaches will be required is relatively
difficult.
Regarding PP, typical reactor site protections on the reactor, control systems, and fresh and


safeguards are quasi-bulk, so differing safeguards approaches will be required is relatively
difficult.
Regarding PP, typical reactor site protections on the reactor, control systems, and fresh and
spent fuel storage will be required. It can be concluded that VHTRs are extremely resilient to
terrorist attacks because RCCS is available when a loss of coolant happens.
For system designers, program policy makers, and external stakeholders who read this white
paper are encouraged to evaluate PR&PP features using the GIF PRPP WG methodology
from an early stage of design, and keep updating them as designs progress.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
297. References
[1] Evaluation Methodology for Proliferation Resistance and Physical Protection of Generation IV
Nuclear Energy Systems, GIF/PRPPWG/2006/005, Revision 6, prepared by the Proliferation
Resistance and Physical Protection Evaluation Methodology Expert Group of the Generation IV
International Forum (GIF), September 15, 2011.
[2] PRPP Working Group and System Steering Committees of the Generation IV International Forum.
Proliferation Resistance and Physical Protection of the Six Generation IV Energy Systems.
Technical Report GIF/PRPPWG/2011/002, Generation IV International Forum (GIF), 2011.
[3] IAEA, Proliferation Resistance Fundamentals for Future Nuclear Energy Systems, IAEA STR-
332, IAEA Department of Safeguards, IAEA, Vienna (2002).
[4] Gen IV International Forum, GIF R&D Outlook for Generation IV Nuclear Energy Systems: 2018
Update (2019).
[5] M. A. Fütterer, et al., "The High Temperature Gas-Cooled Reactor," Encyclopedia of Nuclear


Update (2019).
[5] M. A. Fütterer, et al., "The High Temperature Gas-Cooled Reactor," Encyclopedia of Nuclear
Energy, Elsevier, pp. 512-522, 2021, https://doi.org/10.1016/B978-0-12-409548-9.12205-5
[6] M.B. Richards et al., Part 1 -- H2-MHR Pre-Conceptual Design Report: SI-Based Plant,
GA-A25401, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006.
[7] M.B. Richards et al., Part 2 -- H2-MHR Pre-Conceptual Design Report: HTE-Based Plant,
GA-A25402, General Atomics, Idaho National Laboratory, and Texas A&M University, April 2006.
[8] General Atomics, Gas-Turbine Modular Helium Reactor (GT-MHR) Conceptual Design
Description Report, GA Document No. 910720, Revision 1, July 1996, transmitted by letter from
Laurence L. Parme (GA) to Raji Tripathi (USNRC), "GT-MHR Conceptual Design Description
Report," GA/NRC-337-02, General Atomics, San Diego, CA, August 6, 2002.
[9] L. Lommers et al., “AREVA HTR Concept for Near-Term Deployment,” Nuclear Engineering and
Design, 251, pp. 292-296, October 2012. https://doi.org/10.1016/j.nucengdes.2011.10.030.
[10] Brochure: ANTARES - The AREVA HTR-VHTR Design,


[10] Brochure: ANTARES - The AREVA HTR-VHTR Design,
https://www.yumpu.com/en/document/read/32557580/antares-the-areva-htr-vhtr-design-smr
[11] V. Petrunin et al., "Analysis of questions concerning the nonproliferation of fissile materials for
low-and medium-capacity nuclear power systems," Atomnaya Energiya 105, Issue 3, pp. 123-
127, September 2008 (in English. pp. 159-164, Atomic Energy 105, Springer, New York,
ISSN 1063-4258 (Print), 1573-8205 (Online)).
[12] K. Kunitomi, et al., "JAEA's VHTR for Hydrogen and Electricity Cogeneration: GTHTR300C,"
Nuclear Engineering and Technology 39, pp. 9-20, February 2007.
[13] Chang Keun Jo, Hong Sik Lim, and Jae Man Noh, "Preconceptual Designs of the 200MWth Prism
and Pebble-bed Type VHTR Cores," PHYSOR-2008, International Conference on the Physics of
Reactors “Nuclear Power: A Sustainable Resource,” Casino-Kursaal Conference Center,
Interlaken, Switzerland, September 14-19, 2008.
[14] D. Moses, “Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical
 Protection (PR&PP),” ORNL/TM-2010/163, Oak Ridge National Laboratory, August 2010.
[15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting,


[15] Presentations by PBMR (Pty) Ltd. to the U.S. Nuclear Regulatory Commission Public Meeting,
PBMR Safety and Design Familiarization, February 28-March 3, 2006.
[16] Johan Slabber, PBMR (Pty) Ltd., "PBMR Nuclear Material Safeguards," Paper No. B14,
Proceedings of the Conference on High Temperature Reactors, Beijing, China, September, 22-
24, 2004, International Atomic Energy Agency, Vienna (Austria).
[17]   E. Mulder, MM. van tStaden et al, “The Coupled Neutronics and Thermo-Fluid Dynamics Design
Characteristics of the Xe-100 200 MWth Reactor,” Proceedings of the Conference on High
Temperature Reactors, Las Vegas, Nevada, USA, November, 7-10, 2018.Very-High-Temperature Reactor (VHTR)                 PR&PP White Paper
30[18] Dong Y (2012) China’s activities in HTGRs HTR-10 and HTR-PM. In: IAEA Course on High
Temperature Gas Cooled Reactor Technology. Beijing, China, 22-26 October 2012.
[19] Z. Zhang et al., “The Shandong Shidao Bay 200 MWe High-Temperature Gas-Cooled Reactor
Pebble-Bed Module (HTR-PM) Demonstration Power Plant: An Engineering and Technological
Innovation,” Engineering 2, pp. 112-118, March 2016.
http://dx.doi.org/10.1016/J.ENG.2016.01.020


Innovation,” Engineering 2, pp. 112-118, March 2016.
http://dx.doi.org/10.1016/J.ENG.2016.01.020
[20] Y. Xu and K. Zuo, "Overview of the 10 MW high temperature gas cooled reactor—test module
project," Nuclear Engineering and Design 218, pp. 13–23, October 2002.
[21] The image of XE-100 Reactor downloaded from:
https://x-energy.com/media/xe-100
[22] M. A. Fütterer, F. von der Weid, P. Kilchmann, A High Voltage Head-End Process for Waste
Minimization and Reprocessing of Coated Particle Fuel for High Temperature Reactors, Proc.
ICAPP’10, paper 10219, San Diego, CA, USA, 13-17 June 2010.
[23] Correct reference for SCWR WP (TBC)
[24] Nuclear Fuel Industries, Ltd, “HTGR fuel manufacturing process 1. Fuel particle process,”
https://www.nfi.co.jp/product/prod03.html (accessed 2020-7-16).
[25] K, Juergen et al., “Upgrading (V)HTR fuel elements for generation IV goals by SiC encapsulation,”
Kerntechnik, 77(5), 351-355, 2012
[26] Kiyonobu Yamashita, Fujio Miyamoto, Sigeaki Nakagawa, and Toshiyuki Tanaka, "Safeguards
Concept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor
System," Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997.


Concept for the High Temperature Engineering Test Reactor Using Unattended Fuel Flow Monitor
System," Journal of Nuclear Materials Management, Volume XXV, Number 4, August 1997.
[27] S. Saito, et al., “Design of High Temperature Engineering Test Reactor (HTTR),” JAERI 1332,
1994, https://doi.org/10.11484/jaeri-1332.
[28] U. Cleve et al., “The Technology of High-Temperature-Reactors,” Proceedings of ICAPP 2011,
Paper 11076, Nice, France, May 2-5, 2011.
[29] T. Shiba, et al., “Proliferation Resistance and Safeguardability of Very High Temperature
Reactor,” IAEA Symposium on International Safeguards, 5-8 November 2018, Vienna
International Centre, Vienna (Austria).
[30] B. Pellaud, “Proliferation Aspects of Plutonium Recycling,” Journal of Nuclear Material
Management, XXXI, 1, pp. 30-38, 2002
[31] C. Bathke, et al., “The Attractiveness of Materials in Advanced Nuclear Fuel Cycles for Various
Proliferation and Theft Scenarios,” Nuclear Technology, Vol 179, Issue 1, 2012
[32] A.M. Ougouag and H. D. Gougar, “Preliminary Assessment of the Ease of Detection of Attempts
at Dual Use of a Pebble-Bed Reactor,” Transactions of the Winter 2001 Annual Meeting of ANS,
Reno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001.


Reno, NV, Trans. ANS 85, pp. 115-117, Nov. 2001.
[33] Ougouag, A. M., S. M. Modro, W. K. Terry, and H. D. Gougar “Rational Basis for a Systematic
Identification of Critical Components and Safeguard Measures for a Pebble-Bed Reactor”
Transactions of the Winter 2002 Annual Meeting of ANS, Washington, DC, Trans. ANS 87, pp.
367-368, Nov. 2002.
[34] A.M. Ougouag, W. K. Terry and H. D. Gougar, “Examination of the Potential for Diversion
or5Clandestine Dual Use of a Pebble-Bed Reactor to Produce Plutonium,” Proceedings of HTR
2002, 1st International Topical Meeting on High temperature Reactor Technology (HTR), Petten,
Netherlands, April 22-24, 2002.
[35] A. M. Ougouag, H. D. Gougar, and T.A. Todd, “Evaluation of the Strategic Value of Fully Burnt
PBMR Spent Fuel,” May 2006, Idaho National Laboratory, INL/EXT-06-11272Very-High-Temperature Reactor (VHTR)          PR&PP White Paper
31APPENDIX 1: VHTR Major Design Parameters
Appendix VHTR.A – VHTR Major Reactor Design Parameters
Major Reactor
ParametersFramatome
SC-HTGRGeneral
Atomics
GT-MHRX-Energy
Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA


SC-HTGRGeneral
Atomics
GT-MHRX-Energy
Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA
GTHTR300COKBM   GT-
MHRKAERI
NHDD
Thermal Power  (MW-th) 625 600 200 250 600 600 200
Thermal Efficiency (%) in
Electricity Generation~40 ~48 40 (inferred) 40 ~50 ~48 None, H 2
production
Primary Coolant Helium Helium Helium Helium Helium Helium Helium
Moderator High-
Temperature
GraphiteHigh-
Temperature
GraphiteHigh-
Temperature
Graphitized
Carbon with
Graphite
ReflectorHigh-
Temperature
Graphitized
Carbon with
Graphite
ReflectorHigh-
Temperature
GraphiteHigh-
Temperature
GraphiteHigh-
Temperature
Graphite or
Graphitized
Carbon with
Reflector
Power Density (MW/m3) ~6.3 (inferred) 6.3 4.95 (max) ~3.22 5.4 6.3 2.27-3.0
pebble, 5.68
prismatic
Fuel Materials LEUO 2 TRISO-
coated
particlesUC 0.5O1.5
TRISO-
coated
particles;
LEUC 0.5O1.5
(19.8%) fissile
and
UNatC0.5O1.5
fertileLEUO 2 TRISO-
coated particlesLEUO 2
TRISO-


(19.8%) fissile
and
UNatC0.5O1.5
fertileLEUO 2 TRISO-
coated particlesLEUO 2
TRISO-
coated
particlesLEUO 2
TRISO-coated
particlesPuO 1.8 ,
LEUCO  or
mixed
uranium-
plutonium
oxide (MOX)LEUO 2
TRISO-
coated
particlesVery-High-Temperature Reactor (VHTR)          PR&PP White Paper
32Appendix VHTR.A – VHTR Major Reactor Design Parameters (Continued)
Major Reactor
ParametersFramatome
SC-HTGRGeneral
Atomics
GT-MHRX-Energy
Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA
GTHTR300COKBM   GT-
MHRKAERI
NHDD
Core Inlet
Temperature/Pressure
(ºC/MPa)325/6.0 490/7.07 260/~6.1 250/~7.0 587-666/6.9
(electrical
production) &
594/5.1 (H 2
production)490/7.07 490/~7.0
Core Outlet
Temperature/Pressure
(ºC/MPa)750 for
electricity
generation)850/7.0 750/~6.0 750/~7.0 850-950/6.9
(electrical
production)
&950/5.1 (H 2


generation)850/7.0 750/~6.0 750/~7.0 850-950/6.9
(electrical
production)
&950/5.1 (H 2
production)850/7.0 950/~7.0
Neutron Energy Spectrum Thermal
peaking just
below 0.3 eVThermal
peaking just
below 0.3 eVThermal peaking
just below
0.3  eVThermal
peaking just
below 0.3 eVThermal
peaking just
below 0.3 eVThermal
peaking just
below 0.3 eVThermal
peaking just
below 0.3 eV
Appendix VHTR.B – A Comparison of VHTR Fuel Cycle Parameters
Fuel Cycle
ParametersAreva
Modular HTRGeneral
Atomics GT-
MHRX-Energy
Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA
GTHTR300COKBM   GT-
MHRKAERI
NHDD
Reactor Thermal
Power (MW-th)625 600 200 250 600 600 200
Reactor Electrical
Power (MWe)
Generation~250, 186 for
cogeneration
with process
heat use262 to 286
(varied
assumptions
documented)80 (inferred) 100 per reactor
in two reactors
per module274-302
depending on
outlet T, 87-
202 depending 262 to 286
(varied
assumptions


in two reactors
per module274-302
depending on
outlet T, 87-
202 depending 262 to 286
(varied
assumptions
documented)Only H 2
productionVery-High-Temperature Reactor (VHTR)          PR&PP White Paper
33on H 2
production
Fuel type
 -Form
 -Fertile material
 -Fissile materialLEU
Ceramic
coated particle
U-238
U-235LEU
Ceramic coated
particle
U-238
U-235LEU
Ceramic coated
particle
U-238
U-235LEU
Ceramic
coated particle
U-238
U-235LEU
Ceramic
coated particle
U-238
U-235Pu initially
Ceramic
coated particle
None
PuLEU
Ceramic
coated particle
U-238
U-235
Enrichment (%) ~15 19.8 in fissile
particles, 0.7
(UNat) in fertile
particles10 8.5 in the
equilibrium
core~14 Pure Pu 9.6 pebble,
15.5 prismaticVery-High-Temperature Reactor (VHTR)          PR&PP White Paper
34Appendix VHTR.B – A Comparison of VHTR Fuel Cycle Parameters (Continued)
Fuel Cycle
ParametersAreva
Modular HTRGeneral
Atomics GT-
MHRX-Energy
Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA
GTHTR300COKBM   GT-
MHRKAERI


Xe-100Huaneng
Group &
CNEC/INET
HTR-PMJAEA
GTHTR300COKBM   GT-
MHRKAERI
NHDD
Source of Fissile
Material (inputs
are assumed
since not given in
available
documentation)U.S. or
European
enrichment
plants
(inferred)U.S. or
European
enrichment
plants
(inferred)U.S. or European
enrichment
plants (inferredUndefined Undefined Russian
excess
weapons Pu;
other U and
Pu in later
versionsUndefined
Fuel Inventory
(MT) Not given 4.68 initial core,
2.26 each
reload~2.0 in
equilibrium core~2.9 in
equilibrium
coreNot given ~1.8 in
equilibrium
cycleNot given
Discharge Burn-
up (GWD/MT)150 121 for LEU
cycle175 90 120 ~120-150 153
Refueling
frequency
(months)18 18 Continuous on
lineContinuous on
line24/18
(electrical)/18
(H2)18 Pebble
continuous;
Recycle Approach Baseline is
once-throughBaseline is
once-throughBaseline is once-
throughBaseline is
once-throughBaseline is
recyclingNo recycle,
deep-burnBaseline is
once-through
Recycle
TechnologyTo be
developedTo be
developedTo be developed To be
developedTo be
developedNo recycle, -
deep-burnTo be
developed


once-through
Recycle
TechnologyTo be
developedTo be
developedTo be developed To be
developedTo be
developedNo recycle, -
deep-burnTo be
developed
Recycle efficiency To be
determinedTo be
determinedTo be
determinedTo be
determinedTo be
determinedNo recycle,
deep-burnTo be
determinedVery-High-Temperature Reactor (VHTR)         PR&PP White Paper
35APPENDIX 2: Summary of PR relevant intrinsic design features. Reference IAEA-
STR-332. Please refer to IAEA-STR-332, for full explanations and complete
definitions of terms and concepts.
Summary of PR relevant Intrinsic
design featuresB-VHTR
(GT-MHR, HTTR)P-VHTR
(PBMR)
Features reducing the attractiveness of the technology for nuclear weapons programmes
1. The Reactor Technology needs an
enrichment Fuel Cycle phaseYes Yes
2. The Reactor Technology produces
SF with low % of fissile plutoniumHigher burnup than LWR SF resulting in
low % fissile plutonium.Fully burn pebbles have higher burnup
than LWR SF resulting in low % fissile
plutonium.
3. Fissile material recycling performed
without full separation from fission
productsNo recycling No recycling
Features preventing or inhibiting diversion of nuclear material
4. Fuel assemblies are large & difficult
to dismantleYes. Fuel pellets are inserted in holes in
fuel blocks. There is no fuel assembly in P-VHTR.
Fuel pebbles are small but to acquire 1
SQ of U-235 or Pu would require a large


fuel blocks. There is no fuel assembly in P-VHTR.
Fuel pebbles are small but to acquire 1
SQ of U-235 or Pu would require a large
number of pebbles (tens of thousands).
5. Fissile material in fuel is difficult to
extractTRISO fuel is difficult to reprocess. TRISO fuel is difficult to reprocess.
6. Fuel cycle facilities have few points
of access to nuclear material,
especially in separated formFuel blocks are replaced after one cycle
of irradiation, no reprocessing.Fuel cycle facilities mainly involve
pebble handling but no reprocessing,
and remote operations are required.
7. Fuel cycle facilities can only be
operated to process declared feed
materials in declared quantitiesN/A N/A
Features preventing or inhibiting undeclared production of direct-use material
8. No locations in or near the core of
a reactor where undeclared target
materials could be irradiatedIrradiate target material in moderator
blocks or control rod is a possibility.
Ton quantities of fertile material
needed to generate 1SQ would be
difficult to conceal and would affect
reactor operation. The core is an open cavity filled with
fuel pebble. There is no space for
control rod to hide the target
materials. There is no space to hide
target pebbles and no means to
harvest the target pebbles after
irradiation. Proliferator has difficulty to
distinguish the target materials from
fuel pebble. Ton quantities of fertile
material needed to generate 1SQ
would be difficult to conceal and would
affect reactor operation.
9. The core prevents operation of the
reactor with undeclared target


material needed to generate 1SQ
would be difficult to conceal and would
affect reactor operation.
9. The core prevents operation of the
reactor with undeclared target
materials (e.g. small reactivity
margins)The large number of fuel blocks
required to accumulate 1 SQ of Pu
makes operation of the reactor with
undeclared target easy to detect. It
might be possible to replace the
control rods with the target materials.It is easy to detect diversion because
the core is designed with little excess
reactivity. It is possible to introduce U-
238 pebbles for breeding, but would be
difficult to carry-out, owing to the large
number of pebbles involved. Very-High-Temperature Reactor (VHTR)         PR&PP White Paper
36Summary of PR relevant Intrinsic
design featuresB-VHTR
(GT-MHR, HTTR)P-VHTR
(PBMR)
10. Facilities are difficult to modify for
undeclared production of nuclear
materialThe particle-fuelled reactor is difficult
to modify to use other fuel for
undeclared production of nuclear
material..The large number of fuel pebbles
involved in any undeclared production
makes the activity diifcult to carry out..
11. The core is not accessible during
reactor operationNot accessible and very high radiation
environment.Not accessible and very high radiation
environment.
12. Uranium enrichment plants (if
needed) cannot be used to produce
HEUExpect international safeguards in
place to deter HEU production.Expect international safeguards in
place to deter HEU production.
Features facilitating verification, including continuity of knowledge
13. The system allows for
unambiguous Design Information


place to deter HEU production.Expect international safeguards in
place to deter HEU production.
Features facilitating verification, including continuity of knowledge
13. The system allows for
unambiguous Design Information
Verification (DIV) throughout life
cycleDIV should be straight-forward. DIV should be straight-forward.
14. The inventory and flow of nuclear
material can be specified and
accounted for in the clearest possible
mannerFuel blocks are amenable to item-
counting. Fuel pebbles are treated in bulk for
accounting. Although it is in a closed
system, nuclear material  in the
pebbles always move due to online
refueling through a pipe.
15. Nuclear materials remain
accessible for verification the greatest
practical extentFuel blocks are identifiable by serial
numbers. However, since there is no
water shielding like LWRs, inspectors
cannot directly see the fuel block
loaded in the core.Verification of pebbles may pose
challenges.
16. The system makes the use of
operation and safety/related sensors
and measurement systems for
verification possible, taking in to
account the need for data
authenticationRadiation monitors and interlocks for
fuel transfer machinery can also be
used for safeguards. Measurement systems needed to
count and authenticate fuel pebbles
for operation can also be used for
safeguards. Devices that measure the
reactivity and the burnup will also be
important for safeguards.
17. The system provides for the
installation of measurement
instruments, surveillance equipment
and supporting infrastructure likely to
be needed for verificationSystem elements are similar to LWRs
and should be amendable to
installation of safeguards equipment..Though system elements are similar to


instruments, surveillance equipment
and supporting infrastructure likely to
be needed for verificationSystem elements are similar to LWRs
and should be amendable to
installation of safeguards equipment..Though system elements are similar to
LWRs fuel accounting is different and
item counting is not practical. The
system is a candidate for the
application of safeguards-by-design.37THE GENERATION IV INTERNATIONAL FORUM
Established in 2001, the Generation IV International Forum (GIF) was created as a co-operative
international endeavor seeking to develop the research necessary to test the feasibility and performance
of fourth generation nuclear systems, and to make them available for industrial deployment by 2030. The
GIF brings together 13 countries (Argentina, Australia, Brazil, Canada, China, France, Japan, Korea, Russia,
South Africa, Switzerland, the United Kingdom and the United States), as well as Euratom – representing
the 27 European Union members and the United Kingdom – to co-ordinate research and develop these
systems. The GIF has selected six reactor technologies for further research and development: the gas-
cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten salt reactor (MSR), the sodium-
cooled fast reactor (SFR), the supercritical-water-cooled reactor (SCWR) and the very-high-temperature
reactor (VHTR).