The Multi-Mission Molten Salt Reactor (MSR):

Versatility in Achieving Excess Fissile Destruction, Resource Extension, & Proliferation Resistance Mission Requirements

by, Bruce N. Hoglund, 1995


Background Abstract:

Molten Salt Reactors (MSRs) were abandoned by the USA in the 1970's due to political competition with Liquid Metal Fast Breeder Reactors (LMFBRs) despite MSRs' inability to melt down, their inherent proliferation resistance features (fluid fuel reactor, thorium and 232U/233U), and potential superior breeding and overall fuel cycle utilization advantages. The world is very different since Oak Ridge National Laboratory's Director (Alvin Weinberg) was fired in 1972 for championing MSRs and nuclear safety. Issues that were annoyances to the nuclear establishment then have now stopped continued deployment of reactors in the Western countries. The Developing World needs environmentally acceptable energy and is actively building nuclear plants the West has stopped building. The West now worries about the fissile material glut and unemployed nuclear weapon designers proliferation potentials, and Global Climate Changes from fossil fuels. A solution to these problems is the Molten Salt Reactor.


A brief history of the Molten Salt Reactor (MSR) is given, but the main focus is proliferation of nuclear materials and how the MSR can reduce, if not eliminate, the proliferation possibility via its use of the Thorium Fuel Cycle, as opposed to the current Uranium-Plutonium Fuel Cycle. The destruction of the excess Weapons Fissile Material (Highly Enriched Uranium, HEU, & Bomb-Grade Plutonium) by emplacement in a MSR and converting the excess material's neutrons into proliferation resistant Uranium-233/232 (233U & 232U) is examined. Various postulated proliferation means are discussed and where possible, quantitatively shown not to be possible with MSRs. A good source of references for the interested student of MSRs, thorium fuel cycle, and proliferation prevention.



Belying their ugly name, Molten Salt Reactors (MSRs) provide an optimal solution to today's nuclear problems of; increased safety requirements, nuclear resource extension, proliferation resistance, and reduced waste streams. Furthermore, MSRs can provide immediate solutions to legacy nuclear problems such as excess fissile destruction and long-lived actinide (also called Transuranic [TRU]) wastes burning, including spent fuel. Application of Molten Salt Technology (MST) to these problems is limited only by institutional and political problems caused mainly by a lack of understanding of MSRs and MST.

Russia has acute Nuclear problems, so it seems reasonable that the first MSR should be built there. This project would have the beneficial result of providing capital inflows, similar to those provided to nascent democracies of Post-war Germany and Japan, to a project that would not only be non-military, but would consume significant fractions of weapon materials, weapon producing infrastructure and personnel. Trans-boarder environmental and security problems posed by unsafe power reactors (such as Chernobyl style reactors), long-lived actinides, and possible fissile smuggling could be mitigated by the development and construction of MSRs in Russia.

Development and deployment of MSRs for the purpose of excess fissile destruction and power production is entirely within existing levels of previously demonstrated MST at Oak Ridge National Laboratory (ORNL - near Knoxville, Tennessee, USA) during the MSR's development period from 1946-1980+. Past abandonment of MSR was due to non-technical reasons, as the MSR program at ORNL was extremely successful in demonstrating a reactor with much higher levels of safety and resource utilization than other reactors. In fact, a MSR still holds the world record for the longest continuous chain reaction1. MSRs are not limited to doing just one thing well, they can do all things well!

Operational MSRs, for the initial purposes of excess weapon fissile destruction and legacy actinide waste burning, would provide necessary prototype design and operational experience of MSRs which could, with no modifications to the physical plant (only minor ad hoc chemistry changes in the liquid fuel), serve as competitive power reactors with enhanced safety, resource utilization, and proliferation resistance.


Molten Salt Reactor History

The first MSR was proposed as a power source for a nuclear powered deep penetration bomber during the early Cold War period before the invention of ICBMs (InterContinental Ballistic Missiles)2. Experimental efforts investigating molten salts began in 1947, and after 3 years determined molten fluorides were best suited due to their low vapor pressure at jet-engine temperatures, good heat transfer properties, immunity to radiation damage, and lack of violent reactions with air or water.3 Although the Aircraft Nuclear Project (ANP) was more of a political and military goal than a realistic technological goal, the spin-off MSR technology (ORNL's Molten Salt Reactor Experiment - MSRE) is extremely useful for non-military goals of long-lived waste destruction and safer power production. Unfortunately, due to MSR knowledge being largely confined to ORNL and the competition for funding during the Liquid Metal Fast Breeder Reactor (LMFBR) development of the early 1970's caused the MSR to be abandoned4. Today, it is realized that a breeder reactor will probably not be necessary until after 2050 AD, and recent press about "excess fissile" from the end of the Cold War only reinforces the point. Past abandonment of the MSR was based nominally on assessments of how well the MSR could breed fissile materials (its weakest ability), not on today's criteria of greater safety, fuel cycle simplicity, proliferation resistance, and overall economy. On these criteria, a MSR is a hands down winner.

"The broadest and perhaps most important conclusion from the MSRE is that this was a quite practical reactor. It ran for long periods of time, yielding valuable information, and when maintenance was required it was accomplished safely and without excessive delay." 5


Mission: Excess Fissile Destruction

Excess fissile materials from burgeoning spent fuel and end of the Cold War should not be reason to continue the excessive consumption of uranium resources by current, inefficient Light Water Reactor (LWR) technology which are ad hoc adaptations of submarine reactors. Excess fissile from the end of the Cold War should be viewed as an inheritance to invest wisely 6. To take virtually pure fissile, either Highly Enriched Uranium (HEU) or weapon grade plutonium, and 'blend it down' with depleted Uranium-238 (238U) is to throw away many billions of dollars of enrichment, separation, and purification. Even the wasteful blending-down of plutonium with uranium to produce Mixed Oxide (MOX) fuel elements is simple in concept, but can not be done by any American commercial MOX plants, as none exist due to the expense and lack of market. Creation of an American MOX facility would be expensive, and politically risky as environmental and non-proliferation groups view the creation of a MOX facility as the necessary first step towards creation of the hated "Plutonium Economy". It would also go against 2 decades of American policy towards the use of plutonium and non-proliferation.

The issues and benefits of using a MSR for the destruction of excess fissile materials were explored by others, who concluded that there were no technical reasons which prevent the construction of a MSR to convert HEU and weapons Pu into proliferation resistant Uranium-233 (233U) fuel 7. 233U is uranium, as are current LWR fuels, so there is no change needed in fabricating the fuel (except it must be done remotely; this is the non-proliferation feature of 233U; see below for more information). Furthermore, 233U an excellent thermal reactor fuel due to its lower parasitic capture cross-section and higher neutron yield than current Uranium-235 (235U) or Plutonium-239 (239Pu) fuels. 233U's superiority should allow for higher burn-ups, and therefore lower fuel costs in existing reactors. Furthermore, 233U can be isotopically denatured to any level with 238U and provides significant deterrence in the form of the high energy gamma from a daughter of 232U decay, and significant gammas produced directly from 233U decay. These gammas may prove lethal enough to absolutely prevent direct handling, even by the most suicidal of clandestine bomb makers. Even if this turns out not to be the case, and more research into this area is definitely needed, the unavoidable contamination of all 233U fuels with 232U and 232U's Thallium-208 (208Tl) daughter's 2.6 MeV gamma provides an effective means of locating 233U anywhere on the planet.


An Example Pu Burning Regime

The weapon material burning MSR could initially be fueled with ~2/3 HEU and ~1/3 Pu. This mixture is well within the operational range of MSRs and is a minor extension of already demonstrated MSR operation 8, 9. This mixture will initially provide approximately half the fissions in 235U and the other half in the 239Pu. Since there will be only minor amounts of 238U in the HEU (~8% 238U in HEU enriched to 92%), there will only be very small amounts of new plutonium produced, which will stop once the HEU make-up feed is replaced with a plutonium feed. This MSR will not have the extensive chemical processing necessary to be a full breeder, and will operate as a high converter; ~0.8 - 0.9 conversion ratio, which is well beyond LWRs and would greatly extend fissile supplies 10. For example, a conversion rate of 0.8 in a 1 GWe MSR would only require a daily input of ~0.5 kg fissile / day F1. This fissile material make up could be plutonium from excess weapons stocks or possibly spent fuel directly 11. Just the recently declared 200 tons of excess US weapons fissile material alone could provide for over 1,000 reactor years of 1 GWe MSR operation. Feeding the reactor pure 239Pu as the small amount of makeup material in a high converter MSR, with no chemical processing, would not compromise reactor safety due to reduced delayed neutrons or higher rates of reactivity of plutonium because Pu would only account for ~20% of the fissile 12, with the remainder fissile (the 0.8 of the conversion ratio) coming from the bred 233U from the 232Th salt, and 235U from 233U fissions, within the core salt. Furthermore, the greater prompt thermal (negative) reactivity coefficient of the MSR allows for better control (safer), and the fluid nature of the fuel with the accompanying dispersion of the Pu prevents hot spots or other areas of uneven burn up as is the case with MOX fuels 13. Should the Pu feed rate of the MSR prove too small to achieve satisfactory rates of plutonium destruction, then the MSR could operate with a 'once through quickly' cycle (see Fig. 1), and temporary storage of removed salt (uranium removed via F2 sparge, but with entrained Pu isotopes) for later burning. (see Fig. 1 below)

Fig. 1. A schematic of a possible processing scheme so as to increase the destruction and denaturing rate of weapon's plutonium.


This cycle has the advantage of allowing more weapon's plutonium to be processed through a MSR than if complete Pu burning were done, because ~1/3 - 1/4 of the neutron reactions in 239Pu are captures and not fissions. Neutron capture by 239Pu degrades its usefulness as a bomb material so the mission of weapon plutonium destruction can be said to be partially met, much as the burning of Pu in LWRs does. There will also be 238Pu, depending upon the salt's irradiation time and composition, due to buildup from 235U captures and 239Pu (n, 2n) reactions. 238Pu complicates bomb design due to its large heat release. Representative compositions of various plutonium compositions are shown in Table 1. According to knowledgeable bomb designers, reactor grade Pu does not appreciably increase the difficulty in designing a bomb, but it increases the amount of plutonium required 14. Note, the MSR's values are "steady state" amounts entrained in the salt and are not removed from the reactor.


Table 1: Plutonium compositions of various reactors compared to bomb grade

Nuclear Weapon

MOX/Reactor Grade

Molten Salt Reactor

Plutonium Isotope

Pu % composition

Pu % composition

Pu % composition





















Source for Nuclear Weapon and MOX/Reactor Grade Pu: Page 45, Table 2-2 of National Academy of Sciences (NAS) "Management & Disposition of Excess Weapons Plutonium: Reactor-Related Options" (1995).
Source for Molten Salt Reactor Pu (after 15 years of operation): Page 24, Table 10 of "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207.


The stored salt which contains the degraded Pu is difficult to separate from the salt, and contains significant amounts of fission products such as zirconium and Rare Earths.15, 16 These fission products provide high radiation levels (similar to spent fuel) so as to deter clandestine attempts to separate the Pu (which is difficult). This temporarily stored salt is very similar to the "Spiking Option" of the NAS Plutonium Disposition study in that it contains similar radiation and chemical separation barriers. Unlike the spent fuel of LWRs however, at a later date the temporarily stored salt can be put directly back in a either the original MSR, or into a new MSR with the addition of only enough fissile material (233U, 235U, &/or 239Pu) for criticality, to complete the burning of the Pu isotopes. If the earlier removed 233U/235U fissile is added for resumption of criticality, then it is possible to achieve a complete burn-up of the remaining plutonium (and other transuranics) in the salt, so that there need not be any removal of plutonium, and hence a potential proliferation concern, from the salt at any time.

As currently proposed, the Pu-burning MSR will not need any processing beside the already demonstrated and easy to accomplish sparging with helium and fluorine gas, and collecting the generated volatile and aerosol fission products as the MSR operates.

"The MSRE processing plant, if run in a semicontinuous fashion, is large enough to process a 1000 MW(e) reactor on a three-year cycle." & "With the addition of extractive processing, such a reactor, or one very similar, becomes a high-performance breeder."17

A possible slight modification of the reference design's large amount of graphite moderator in the core, heterogeneous style MSRE type design, might be a more homogenous arrangement (where the graphite is largely, or completely, removed from the core and the moderation comes from the salt itself). This is not unreasonable as the salt itself provides 1/2 - 2/3 the moderation of graphite, due to the large amounts of LiF and BeF2 (berylium and fluorine are the main moderators; lithium, despite its low mass is a poor moderator, and for the MSR must be enriched to at least 99.9% Li-7 so as to reduce parasitic neutron capture in Li-6 and H-3 [tritium] production). However, the homogeneous molten salt systems were abandoned due to "inferior breeding performance" {relative graphite moderated systems} and higher fissile inventories 18. Of possible additional benefit could be the fact that the harder spectrum (due to the decreased moderation as provided by the graphite in the core), might aid in increasing plutonium through-put (due to increased captures versus fissions), and may increase the amount of 232U produced as its production is increased by fast, unmoderated neutrons.


Mission: Resource Extension

The construction of a domestic MOX facility, while providing a 'sink' for excess weapons plutonium, would not meaningfully extend the uranium resource. In fact, the MOX facility would have to be shut down if additional plutonium production from spent fuel reprocessing facilities were not also constructed, as the weapons' plutonium would be quickly exhausted. A MOX fabrication and reprocessing facility will require governmental funding, as no commercial enterprise will undertake it 19. (It would help continuation of the current Nuclear Energy Industry as the "razor blade" business, where the money is made selling the services and fuel elements, not in the original reactor.) 20 Complete recycle of plutonium from commercial reactors via spent fuel reprocessing and MOX fuel element fabrication facilities only extends the uranium resource by 38% 21. Table 2 shows that the resource requirements of the MOX and Reprocessing duo are over 6 times that of an MSR with no chemical processing (CMSR). MSR's resource extension is also supported by Russian investigators of MSR converter reactors (CMSR; no chemical processing) at the "Kurchatov Institute" (Moscow, Russia) where a LWR required 5 times the resources that a CMSR did, and a thorium fueled LWR (a Light Water Breeder Reactor type) 2.5 times the resources 22.


Table 2: Uranium Resource Requirements of LWRs and MSRs (with no extensive chemical processing)






No Recycle
(Once through)


Recycle U


Recycle U & Pu

(Full Reprocessing)


Denatured MSR


Converter MSR



Uranium (U3O8) requirement, tons












Note: The resource requirements are in "Yellow Cake" from the operation of a 1 GWe plant at 60% capacity for 30 years.
Source U cycle: Page 9, "The Liquid Metal FAST BREEDER REACTOR: An Environmental and Economic Critique", T.B. Cochran (1974).
Source for DMSR & CMSR: Pages 30 - 33, & Table 17 of ORNL/TM-7207, "Conceptual Design Characteristics of a Denatured Molten Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes.
Adjustments to data in Table 17 for the DMSR & CMSR made for 60% LMFBR plant utilization factor, conversion to tons, & 50% resource reduction due to fissile credit at end of MSR life (fissile inventory within remaining, useful salt). CMSR requirements based on DMSR, but reduced by 29% greater efficiency due to removed dilution of salt by 238U dilutant.

A commercial MOX fabrication facility would probably cost as much as a MSR and take as long to deploy, yet without concomitant construction of spent fuel reprocessing facilities, the MOX facilities would become an useless "White Elephant" once the excess weapons material was consumed. Furthermore, reprocessing would not provide a significant extension of the uranium resource due to the low conversion rates and poor thermal efficiency of the base LWRs (see Table 2). Even if the LWRs were re-cored with thorium (increases Conversion Rate, CR) instead of 238U, it is unlikely an aqueous reprocessing facility ('Water based'; the common type today, versus the "Dry" fluoride processing) could be created to reprocess the solid thorium fuel elements or that a MOX facility could be adapted to thorium based fuels. Past experience with aqueous reprocessing of solid thorium based fuels (Thorex), as demonstrated by the commercial and process failure at the West Valley Reprocessing Plant 23 illustrates the technical difficulty of reprocessing solid fuel elements based on thorium fuels. Much of the difficulty is due to the more difficult aqueous process (THOREX) thorium's chemistry introduces versus the easier uranium-plutonium (PUREX) processing24. Molten salt based processing methods, such as the "Dry" (because there is no water involved) Fluoride volatility processing or "Pyroprocessing", have the ability to avoid such problems but are not yet commercial.

The only way to avoid both the reprocessing and MOX fabrication facilities from becoming quickly obsolete due to resource depletion is to also create a massive breeder reactor program again. This has well documented technical, not to mention the more serious political liabilities, and is unlikely to begin soon enough to address current nuclear problems. Furthermore, the Integral Fuel Reactor (IFR - a combined LMFBR & on-site pyroprocessing fuel cycle), which has a very proliferation resistant fuel cycle due to the incomplete processing and subsequent retainment of substantial fission and actinide products via Molten Salt processing, is not viewed by many in the current LWR industry as being a useful, future source of fissile material. This is because the excess plutonium produced by the IFR is very "dirty" (highly contaminated with fission products and actinides, which is the anti-proliferation aspect of the fuel cycle). The "dirt" would provide additional difficulties in both the reprocessing and fabrication of MOX fuel elements due to the contamination with fission products (mainly gamma emitting zirconium and Rare Earths) and built-up actinides. Direct handling of the fuel elements would probably also be impossible, so all fabrication, inspection, transportation, and loading would have to be done remotely, which adds to the costs of an already marginally cost effective fuel cycle.

To solve the current problems of excess fissile and destruction of long-lived actinide wastes, politically difficult and technically demanding funding, design, and deployment of the above triad solutions; MOX Fabrication Facility, Reprocessing Facility, and Larger Scale IFR, must be achieved simultaneously. The MSR however, represents a complete solution. It can be initially fueled with HEU and Pu in roughly the proportions obtained from dismantled weapons (perhaps 2/3 HEU & 1/3 239Pu), and the small, annual fissile additions, to maintain operation of a MSR, could be met by additions of any fissile material from any source; excess weapons fissile, spent fuel, or even IFR bred "dirty" fuels. Should fissile fuels become scarce in the future, the MSR can flexibly adapt to any fuel as dictated by economics and availability. If necessary, a well designed MSR can, with the addition of only chemical processing to the existing physical plant, become a net breeder of 233U fuel 25. However, should economics and future developments prefer the use of fissile fuels produced outside a MSR (such as IFR, Hybrid fusion breeders, Accelerator-based Breeders, etc.), the MSR has unparalleled flexibility to efficiently utilize those fuels without modification to the basic physical plant or design, nor interruption in operation.

Most of the attention on the production (breeding) of new nuclear fuel is usually focused on the rates and amounts of nuclear fuel produced by various arrangements of reactor devices and fuel cycles. Rarely considered are the possible synergies between seemingly different energy sources, or of the quality of the proliferation resistance of the ultimately stockpiled, fabricated, and transported fuel. An excellent example of these synergies are between fusion and fission. Fusion is a neutron rich, but tritium poor reaction. Furthermore, producing sufficient tritium depends upon surrounding the fusion reaction with large amounts of Lithium-6 (less than 10% of the naturally occurring lithium) so as to "breed" tritium (or possibly helium-3 [He3]) via neutron reactions. Fission is a neutron poor reaction, and in the MSR's case, a tritium rich reaction. Additionally, the MSR needs lithium enriched in Lithium-7 (the opposite of the fusion reaction, and 7Li is >90% of the naturally occurring lithium), but the tritium is produced in ~50:50 ratios in both of the lithiums 26. Therefore, the MSR is able to utilize the lithium the fusion reaction rejects and produce the required tritium, and the fusion reaction is able to produce the 233U fuel the MSR needs from its excess, high energy neutrons!

Although there has been some interest in the coupling of these natural synergies 27, 28, 29 little mention seems to be made of the extremely proliferation resistant fuel that would result from the high-energy (fast) neutron regime common to all advanced 233U producers. The fast neutrons will cause greater 233U production, than thermal MSRs, so as to not only produce sufficient quantities of makeup fuel for MSRs (thereby relieving MSRs of having to achieve their weakest ability; Breeding), but to allow for a fuel that should have excellent non-proliferation aspects due to extremely high levels of 2.6 MeV gamma radiation from the high levels of 232U. Thorium based nuclear fuel production (233U) in a fast neutron regime enhances production of 232U. 232U is produced via a fast, 6.37 MeV threshold, neutron displacement reaction with 232Th; a (n, 2n) reaction 30. Not only do currently proposed fuel producing schemes such as IFR, Fusion-Fission Hybrids, and Accelerator-Based Breeders offer potentials of abundant future supplies of fuel for consumption in simple MSR converter reactors, but also the inherent proliferation resistance offered by high levels of 232U produced by the fast neutron spectra common to all. High 232U levels should allow for storage, transportation, and common utilization of this future nuclear fuel, with minimal need to dilute with 238U, the world over.


Mission: Proliferation Resistance (Fluid Fuels)

Producing weapons grade plutonium requires that the 238U fertile material be irradiated long enough so that there are sufficient neutron captures to create enough 239Pu to make chemical extraction practical, but not so long that the newly produced 239Pu captures neutrons and either fissions or transmutates into non-fissile 240Pu and 242Pu 31. Since there are no fuel elements in a Fluid Fuel Reactor {FFR} there is no mechanical damage done by neutron irradiation, and requirements {and therefore opportunities for clandestine diversions} for fuel fabrication, transportation or reprocessing are all but eliminated 32. No mechanical damage combined with the ability to constantly process the fuel means complete burn-ups are achievable. Furthermore, since the fluid is a homogeneous fluid, there are no subunits that can be specially treated or irradiated so as to clandestinely produce undetected fissile 33. Uranium (the 233U fuel) is easily removed from salts via the fluoride volatility process, so there need not be any fissile in any of the wastes, which could later be clandestinely "mined". Plutonium is not easily removed from the salt and due to its much higher neutron reaction rates suffers fission and isotopic conversion to non-fissile plutonium (non-bomb grade Pu) at a much more rapid rate than the uranium fuels; 233U and 235U. This is one of the reasons fluid fuel reactors have never been known to produce bomb material.


Mission: Proliferation Resistance (Uranium Fissile Dilution)

After many years of operation, the fuel salt will contain all the long-lived isotopes of uranium. It seems that this natural denaturing of uranium has not been fully considered, as when the MSR was studied as a non-proliferation system 238U was added to denature (isotopically dillute) the salt. This has the disadvantage of slightly reducing neutron economy and producing plutonium (although at levels far below LWRs). If the original MSR is operating in a secure area on bomb grade material, there would be no need to denature the fissile within the salt. Instead, the Pu burning MSR would operate as a 'Pu to 233U converter via thorium neutron absorption. A rough approximation of how the uranium isotope composition would change over time is shown in Table 4 which takes the data from Engel's "Conceptual Design..." in "Table 9. Actinide inventories in DMSR [Denatured MSR] fuel salt" and converts to percentage uranium compositions, but with the amount of 238U not considered as uranium, but as plutonium feed (which it ultimately becomes). This gives an approximation of the natural denaturing that occurs in a fuel salt since there is no need for fuel element removals and reprocessing. Unfortunately, 232U buildups {see below for significance} were not calculated as it was considered only of minor importance due to the minimal effect 232U has on the nuclear performance of a reactor.


Table 4: Uranium Isotopic Composition Changes in "Thermal" & "Epithermal Spectrum MSRs

Thermal MSR

Epithermal MSR


Composition %

15 year
Composition %

30 year
Composition %

Composition %

5 year
Composition %

20 year
Composition %






























See note below

See note below

See note below




Source Thermal MSR: Page 23, Table 9 of ORNL/TM-7207, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, W.A. Rhoades, July 1980.
Source Epithermal MSR: Pages 653-655, Table 14-6 of "Fluid Fuel Reactors" (1958), Ed. J.A. Lane, H.G. MacPherson, & F. Maslan.
Note: Uranium-238 amounts were not shown for the Thermal MSR so as to obtain an approximation of the uranium isotopic changes, without the denaturing additions of 238U that were made to the Denatured-MSR from which the Thermal MSR data was obtained.

It is beyond the scope of this paper as to the effect of this uranium isotopic changes upon bomb design, and if the "dirty salt's" uranium could even become critical so as to be a bomb. It should be pointed out that both 234U and 236U have neutron absorption cross sections far higher than 238U so as to make them much more effective dilutants than 238U. Should 234U and 236U 's denaturing prove insufficient to dilute the fissile to below bomb capable, then 238U (depleted uranium) is easily added so as to complete the dilution - which does not have a large negative impact upon reactor economy or Transuranic Waste production 34.

It would be very desirable to have a "metric" of "allowable" uranium isotope concentrations based on the "threat" (likelihood) of bomb production diversions. Such a metric would greatly aid in the tailoring of the fuel mix (which is eminently possible with Fluid Fuel Reactors) so as to provide the desired proliferation-resistance, yet without undue loss of neutron economy or production of Transuranic Wastes. It may ultimately be determined that 238U isotopic dilution is unnecessary as the fuel salt will also contain amounts of 232U which, as it will be seen below, render clandestine bomb manufacture and employment all but impossible.


Mission: Proliferation Resistance (Spontaneous Fission)

It is well known among nuclear bomb designers that there are two basic types of bomb designs; "The Gun Type" and the "Implosion Type". Gun Type of bombs are generally considered simpler than Implosion Types (which are required if the rate of spontaneous fission is "too" high). Plutonium isotopes exhibit high enough levels of spontaneous fission so as to require implosion type bombs. Uranium isotopes, with the possible exception of 232U, do not exhibit high spontaneous fission rates, which allows the construction of either gun or implosion type weapons.

Table 5: Spontaneous Fission Rates in Uranium and Plutonium Isotopes

Uranium Isotope

Spontaneous Fission Rate

Plutonium Isotope

Spontaneous Fission Rate


(8 ± 5.5) x 1013 yr


(5 ± 0.6) x 1010 yr


??? yr


5.5 x 1015 yr


1.6 x 1016 yr


(1.34 ± 0.015) x 1011 yr


1.8 x 1017 yr


??? yr


2 x 1016 yr


(6.5 ± 0.7) x 1010 yr
Source: "The Nuclear Properties of the Heavy Elements: Fission Phenomena", Vol. III, E.K Hyde (1971).


It is doubtful if spontaneous fission will require implosion type bombs for 233U fissile, even with large amounts of 232U, as the Spontaneous Fission Rate is still orders of magnitude smaller than the plutonium isotopic contaminants.


Mission: Proliferation Resistance (High Energy Gamma Radiation)

The MSRE (The Molten Salt Reactor Experiment} was the first reactor to operate on 233U 35. The MSRE was fueled with 39 kilograms of 233U 36, which "... contained ~220 ppm 232U, which made it so radioactive as to practically prohibit any other use." 37 Although the 233U used in the MSRE came from relatively low burn-up fuel elements from various LWRs that had operated on 235U and thorium fuel elements (such as the Indian Point PWR) and were reprocessed by the Thorium-Uranium Recycle Facility (TURF) and only contained 222 ppm of 232U, it was sufficient to generate a gamma dose of 300 rem/h from a 450 gram (~1 pound) amount of the uranium oxide at a distance of 2 inches (5 cm) 38. This is an example of the unique proliferation resistance the thorium based fissile fuel, 233U, provides.

Unfortunately, the high energy gamma radiation does not come from the 233U directly, although 233U is the most gamma active of the fissile fuels. The gamma comes from a daughter decay product of a contaminate of the 233U fuel, 232U. The 2.614 MeV gamma ray, that is considered the biggest deterrent to weapon's use of 233U fuels, comes from the beta particle (ß-) decay of Thallium-208 (208Tl), which occurs 36% of the 232U decays 39. Although most of the precursor daughters of 208Tl are short lived, the immediate decay product of 232U, 228Th, is relatively long lived at 1.913 years 40. This has the effect of delaying the rate of gamma radiation until such time as the 228Th can build up from the decay of 232U. However, once there is even tiny amounts of 228Th, the gamma radiation becomes quite intense, but until there is at least some 228Th, the 232U contaminated 233U is relatively radiation free of 2.6 MeV gammas.


Graphs of Thallium-208 Precursors (232U & 228Th) Buildup and The Resulting Gamma Radiation

The graphs above shows the rate of 228Th, and the 232U and 228Th amount changes and resulting gamma radiation that would occur in a 6.7 kg mass of 233U, the approximate minimum to construct a weapon 2, 41. Figs. 3 and 4 illustrates the rapid gamma radiation increase, even though it comes from the secondary decay from the accumulated build up of 232U daughter, 228Th. One of the postulated means that fissile material could be clandestinely obtained is via the "Break-out Scenario", whereby a MSR is overtaken (due to direct terrorist attack, overthrow of a previously benign government, etc.). The uranium within the MSR could then be easily extracted via the fluoride volatility method (which is one of the beautiful features of control and processing of a MSR). The uranium fluoride (UF6) could be converted from the UF6 chemical form to UF4, via vapor-phase reduction with hydrogen 42, and then to uranium metal via reaction with calcium or magnesium 43. This uranium is considered bomb grade unless diluted by 238U, as in the Denatured Molten Salt Reactor (DMSR; where 233U & 235U is kept below 13% & 20% enrichments, respectively) 44. However, the undiluted MSR uranium would not be pure fissiles 233U and 235U, even if no 238U denaturant were added as there would be significant quantities of 232U, 234U, 236U, 237U, 238U, and 239U due to neutron captures and (n, 2n) reactions {See Table 4}; dependent upon MSR operating parameters. The 237U and 239U would present the greatest early danger to the would be clandestine bomb makers due to their short half-lives and gamma emissions. 233U is also the most gamma active of the fissiles and would present some degree of danger to bomb fabricators, possibly with a surface radiation as high as 430 rems/hr 3. This value should be considered highly speculative, but demands resolution as it is an immediate radiation from the 232U and not dependent upon the build up of a secondary, daughter material, such as 228 Th is for 232U. The better known and most significant deterrent would come from the 232U daughter 208Tl, but since that radiation is "delayed" due to the buildup time for the 228Th precursor, the gammas from 233U, 237U, and 239U could provide enough of a early deterrent until the 228Th could build up to the lethal level necessary to deter use of uranium derived from a MSR under breakout conditions (see Fig. 5). Further research is necessary to determine expected radiation levels from uranium derived from a MSR under a variety of operational conditions.


Fig. 5 Conceptual Plot of Gamma Radiation from Recently Separated (Sparged) MSR Uranium


Unfortunately none of these details were considered when the MSR was last considered during the late 1970's and early 1980's because of its inherent proliferation resistant features. The great differences between that earlier nuclear age and today's, 'Question and quantify everything' age, is illustrated by the following anecdote:


When "Old Timer" MSR specialists are asked, "Why didn't you better quantify the radiation hazard, and therefore the proliferation deterent of the 233U fuel?" Their answer is invaribly, "We knew it was lethal. Who cared how lethal!"


Mission: Proliferation Resistance (Gamma Radiation Detection & Shielding)

Gamma radiation can be reduced by placing sufficient mass between the source and the object so as to shield the object from the full radiation dose. Furthermore, the source can be made more difficult to detect by shielding due to the obscuring effect shielding has on radiation {please see Appendix page 5 for further discussion}. To achieve a level of 1% of the original unscattered (more significant for detection prevention than biological damage prevention) radiation amount, the bomb sized mass of 233U (6.7 kg 233U) would have to be surrounded by a 0.49 meter (~19 ins.) thickness of concrete {Appendix page 5}. While achievable, it does greatly complicate bomb design and perhaps prevent clandestine construction or movement a weapon constructed using 233U. Perhaps this is why no current nuclear weapon nation has any known 233U weapons, as secrecy of emplacement and number of weapons would be largely impossible, not to mention the hazard to weapon handling personnel.

It should be noted that a reduction of 1% of the radiation of a 233U bomb was just given as an example of the difficulty and the amount of concrete necessary. There are no known open literature sources as to the sensitivity of gamma detectors that could be mounted on planes and satellites for the detection and monitoring of 233U clandestine weapon creation and deployment. As any weapon designer, clandestine or not, will always take the simplest route towards creating a weapon that meets the needs and abilities of the creating organization, it is unlikely that anyone would choose the more difficult route posed by using 233U when there are many tons of weapon and reactor 239Pu, and the possibility to enrich naturally occurring uranium via laser, centrifuge, or even old-but-tested Calutrons.

Mission: Proliferation Resistance (232U build-up methods)


Contamination of 233U fuels is unavoidable due to the multiple pathways 232U is formed.

232Th -(n, 2n)--> 231Th --ß (25.5 hr)--> 231Pa --(n, g)--> 232Pa --ß (1.31 days)--> 232U

233U -(n, 2n)--> 232U

233Pa (n, 2n) -> 232Pa --ß (25.5 hr)--> 232U

Ý230Th --(n, g )--> 231Th --ß (25.5 hr)--> 231Pa --(n, g )--> 232Pa --ß (1.31 days)--> 232U

237Np -(n, 2n)--> 236Np --ß (22 hr 50%)--> 236Pu --» (2.85 yr )--> 232U


Ý NOTE: 230Th is a variable naturally occurring thorium isotope that depends upon the amount of co-located uranium as it is a daughter of the 238U decay chain. It can vary from 0 - ~100 ppm of the thorium isotope 4, 45.


Utilization of Thorium and/or Uranium-233 (233U) makes it impossible to design a reactor or fuel system that does not have some degree of contamination with 232U, as there is always a neutron flux that will produce 232U via one or more of the pathways above. The degree of contamination is entirely dependent upon actual reactor operation and overall system design. It is this lack of ability to quantitatively specify exact production amounts of 232U contamination, prior to detailed design studies, that prevents Thorium based 233U's wholesale adoption for proliferation prevention. More research and better data is needed to better quantify 232U production under a variety of reactor operations and conditions.

It should again be noted that there is no known way to easily extract Protactinium from molten salt fuels. So the proliferation danger posed by clandestine removals of Protactinium isotopes is moot as there are no easy ways to remove Pa from molten salt; detection would be very easy. 46


Mission: Proliferation Resistance (Salt Amounts to Steal)

It is sometimes suggested that the salt containing the Protactinium (Pa) could be quickly and secretly removed from the reactor so as to produce a source of 233U without the 232U contamination problem. This argument is flawed due to many details that render such a scheme impossible. First, even in a reactor with very high concentrations of Pa in the salt, such as a Molten Salt Breeder Reactor (MSBR) design, there would only be 1.54 x 10-5 mole Pa/mole salt!47 To remove a kilogram of 233U at least a kilogram of Pa would have to be removed and allowed to decay (assuming 100% recovery factors, and neglecting undesirable 231Pa, 232Pa, and 234Pa isotopes that will also be present to varying extents and whose decay will produce 232U and 234U contaminates). This would require the removal of ~279,000 moles of salt.5

Based on the salt composition of the Molten Salt Breeder Reactor (MSBR) {see footnote 6 & reference 48} the salt had a mass of 62.2 grams salt / mole of salt. Since the 233U thief requires 279,000 moles of salt, we can see they will have to steal 17,353 kilograms (17 tons) of red hot and highly radioactive salt to obtain their 1 kilogram of 233U.


Mission: Proliferation Resistance (Protactinium separation difficulties)

To overcome this rather difficult chore, it is often suggested that the 233U thieves will 'skim off' (on site) the Pa from the salt while the MSR operates. Although there may be vast future improvements in Pa removal from molten salts, it must be remembered that one of the reasons the MSBR was not considered was due to the difficulty of Pa removal at quick enough and high enough rates. A quick glance at the "Conceptual reductive extraction flow sheet for processing a MSBR"49 will prove otherwise to those who suggest Pa removal is a trivial task. Also, even with all of the complexity and optimization of the Pa removal system the efficiency with which Pa was removed was not 100% efficient 50, and in fact "represented a steady state at very nearly optimal conditions" where "a small error in this amount [reductant flow into the system]" could cause the "return of all of the protactinium to the reactor", where it would produce 232U due to 231Pa (n, g) and 233Pa (n, 2n) reactions. The exact amounts and rates of which all depend upon the many reactor design and operation variables.

Even if easier and more efficient Pa removal systems are developed, there would be an increase in proliferation risk only to the extent that the system could immediately remove Pa as it was formed with complete efficiency. Given this unlikely scenario, the thieves would then have to get access to the entire flow of molten salt, which in the planned 1,000 MWe MSBR was circulated through 4 circuits pumped by 84,000 liters/min. (22,000 gal/min.) pumps 51 for a total of 43,000,000 kg (94,800,000 lbs) per hour. Since ~1 kg of 233U is burned for 2.2 x 107 kWh (thermal) 52, and the MSBR was expected to be of 2,250,000 kW (thermal) size and have a breeding rate of 1.06, the thieves could, at most, obtain 0.006 kg/hr of 233U 7, 53 without impacting criticality conditions (stopping the nuclear reaction).

Even if the 233U thieves had complete access to the reactor chemical processing facility, they would still need to chemically process out parts per million quantities of highly radioactive Pa in a flow of 704 °C (1,300 °F) radioactive salt at a rate of 43,000,000 kg (94,800,000 lbs) per hour for 167 hours ( 1 kg 233U / 0.006 kg/hr) at 100% efficiency to obtain their 1 kilogram of 233U. This will produce the purest possible 233U, but it will still contain some 232U, which will largely depend upon neutron fluxes, energy spectrum distribution, and ages and concentrations of thorium and protactinium. No known technology is able to process that much salt at such high efficiencies, but the development of this hypothetical process removes the only real barrier a MSBR has to becoming an immediate economic reality. Any thieves with access to such a wondrous technology would do better to sell their services freely to legitimate reactor operators, and use the cash they receive to buy the world versus blowing it up!

Mission: Proliferation Resistance (Salt Heat & Radiation)

Another hurdle for the pilferer of salt is the heat which is released by the salt. Even with a 1 hour cooling period to allow the decay of the high energy releasing, short lived isotopes, the salt still releases ~350 W/liter (~10 kW/ft3), not to mention the associated radioactivity 54, 55. Obviously not a material that someone would put in their pocket. Even if the thieves attempted to steal the salt in the 200 ft3 Pa Decay Tank where the concentrations of Pa are highest 8 the heat release is 1 kW/liter (28 kW/ft3), which makes it even hotter. The 233Pa, which has a half-life of 27 days, emits a fairly strong gamma in 34% of the decays of 0.312 MeV56. While not as strong a gamma as 208Tl's 2.6 MeV gamma, it still requires substantial shielding to avoid detection. There will also be a small amount of 234Pa in the salt, and it has a ~0.9 MeV gamma it emits in 70% of the time.57

Gamma emission of the salt was never quantified during the MSR's development period, as non-proliferation aspects of Molten Salt were assumed to be self-evident back then, but research conducted about radiation stability of the fluoroborate secondary coolant salt illustrates the radiation magnitude. The fuel salt (with the fissile) was estimated to give the fluoroborate salt a ~0.25 W/g dose (through the heat exchanger tube walls).58 The fluoroborate secondary salt was found to be very stable at that irradiation level, but human tissue would not fare so well, as a 0.25 W/g absorbed gamma dose is equal to 90,000,000 rads/hr (900,000 grays/hr)9 (1,000 rads whole body is 100% lethal; >5,000 rads, death occurs in hours)59.


Mission: Excess Military Fissile (Plutonium) Burial

To avoid perceived high cost and long lead time obstacles to new reactor or even MOX fabrication facilities, it is suggested that the Pu be mixed with already existing highly radioactive wastes and buried. To believe that this option will be simple, cheap, and quick displays total ignorance of the history of burying the less controversial civilian nuclear wastes; burial of the much more emotional weapon materials is unlikely be quicker, simpler and cheaper. Furthermore, it is not clear in the scientific community if geologic isolation of nuclear materials can be assured for the 'beyond human time scales' involved, nor is it assured that buried fissile materials will be safe from either human thefts or natural accumulations and subsequent nuclear excursions. It should be pointed out that plutonium-239, with a half-life of 24,400 years, would still be 87% present in the grave-robbed pyramids of Egypt had it be buried there, and its decay product would be the easy to construct a nuclear bomb material, uranium-235!

Any study of the ongoing Yucca Mountain debacle, with accusations by government scientists that it could explode60,61 is decades late and billions of dollars over budget62, should disabuse those who believe weapons plutonium burial would be a quick, easy, cheap task. Meanwhile the spent fuel is piling up at the nation's utilities63 so badly that 20 states are suing the Federal Government for failing to honor a commitment to developing a permanent waste storage area despite having collected a penny ($0.01) per 10 kilowatt fee to pay for the repository.64 Press reports of the thousands of shipments of nuclear materials that will need to be transported65, along with serious questions about the ability to retain hazards for the 10,000 years the EPA requires, let alone the possible need to retain the nuclear materials for up to 250,000 years, as suggested by a recent report by the National Academy of Sciences66 have eroded public confidence and led Sen. Richard Bryan (D-NV) to declare,

"A repository will never be built at Yucca Mountain."67


"It's nothing more than a high level swindle ... a $10 billion scam perpetuated by our own federal government,"

said Michigan Attorney General Frank Kelley68 of the burial of spent nuclear fuel at Yucca Mountain, and the same might be said about any plan to bury weapons plutonium.



Mission: Excess Fissile Burial (Quasi-Technical Issues)

"Plutonium can only be destroyed by neutrons; burial only hides it. Plutonium needs to be viewed differently than it is today, perhaps best expressed by the statement; 'Plutonium should not be considered a waste, nor a resource, but instead, an endowment'."

Professor Wolfe Häfele during a talk at The First Annual "Alvin Weinberg Lecture", Oak Ridge National Laboratory, April 25, 1995.


Many arguments are made against the storage of nuclear materials. These arguments fall into 2 categories:

Question about the form and composition of the materials
Questions about the possible hazards of greater than human-time-scale storage.


The first questions the current nuclear infrastructure and political/bureaucratic rules and the risks they may allow. Such questions then lead to rules such as, 'No reprocessing of spent fuel due to proliferation hazards'. However, if there are no fissile materials, nor long term heat producers such as Transuranic wastes, then there are no questions about unintentional criticality {nuclear "burnings" or explosions}, proliferation thefts, or transuranic heat powered migrations (also answers 2nd question). Plans to mix excess fissile (weapons' plutonium) with HLWs (High Level Wastes) so that the radiation can provide a barrier to possible future proliferation underestimate the many possible present chemical separations methods (let alone future ones) or that HLWs decay faster than fissiles so that the weapons quality of the buried material improves, like a fine wine, over time.69

The second question is the most difficult as there are a myriad of technical and social issues of the viability of waste packages, and the geological, hydrological, and social conditions for thousands of years are beyond our current ability to forcast. Because this questioning of long term safety operates at the boundary, or beyond, of what is known about long term effects, the questioner can freely cause sufficient doubt as there is little likelihood that we will soon find answers to questions that are beyond human experience. One of the best examples of this is ORNL's Gordon Michaels' paper, "Potential Benefits of Waste Transmutation to the U.S. High-Level Waste Repository" [AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, NV July 1994, pages 8 - 21]. This paper examines possible problems with long term storage viability of high level wastes (HLWs), such as spent fuel [which the paper examines specifically] or the similar weapons plutonium mixed with HLWs, due to the heat released by the actinides (~80% of total in the first 1,000 years). The best way to eliminate worries and endless debates at the fringe of the known, such as the possibilities of fission excursions {nuclear reactions/explosions due to water and fissile self assembling}, or potential later thefts of bomb material, is to eliminate the root sources of such problems; the "un-natural" actinides, such as Plutonium, via neutron bombardment.


To summarize the weapons' plutonium burial options; they are naive, more costly and will take longer to conduct than advertised, will not permanently remove the weapon material from later use, have long term consequences, and do nothing to solve other vexing nuclear problems such as spent fuel.


Mission: MSR Deployment

The MSR is usually viewed by those unfamiliar with it as a single reactor type with the unusual and suspicious, circulation of fuel through-out its primary loop. This view only serves to illustrate the general ignorance of MSRs, as the MSR can take many forms, and can be later modified by the addition of chemical processing. Although MSR designs in the USA and Japan have generally been conservative modifications of the reference ORNL MSBR design, the Russians have considered much more imaginative designs. These include; a MSR without graphite in the core (homogeneous core) and a graphite reflector, a High Temperature MSR using only graphite as the construction material and radiant heat exchanger, and natural and gas-lift augmented circulation MSRs.70

Even once a physical design is built, the MSR can be functionally changed by changing the chemistry of the circulating fuel, and by the addition of online fuel salt processing. This option, also confuses many, as it is generally unknown that there is a sort of "triage" of MSR processing. Some of the processing that can be done (added later) are:71

� Sparging with helium {removes noble gasses and some tritium and noble metals}
� Plating out on metal surfaces in reactor and heat exchanger {Semi- and noble metals}
� Sparging with F2 gas {removes bromine and iodine}
� Reductive extraction with Bi-Li alloy {other fission products}

The minimum processing that should be done is the sparging with He, as that is simple and removes large fractions of fission products. The plating-out of the noble and semi-noble metals on MSR walls (or mesh strainers) will happen naturally, although its rate is sensitive to the UF3/UF4 ratio, which is also controlled to prevent corrosion and uranium oxide formation should there be an ingress of air or water.72 The other two processing schemes are only required if single salt breeding should become necessary due to large increases in fissile prices. Otherwise, the expense, especially the "Reductive extraction with Bi-Li alloy", is unwarranted with today's low cost fissile supplies.

Basic MSR processing is simply allowing the natural (or usually accelerated by bubbling, or sparging, the liquid salt with helium gas) bubbling out of the noble gasses such as krypton and xenon. This feature has the benefit of not only removing important neutron poisons such as xenon-135, so that the need for excess reactivity for xenon and other fission product over-ride is all but removed, but also removes large fractions of fission products that could create a radiological hazard should an unexpected type of nuclear accident occur. Other potentially dangerous fission products in an accident, such as iodine and strontium, "... showed no tendency to escape from the salt."73

Unlike other reactors, most of a MSR's fission products would not be present in the reactor, but in isolated, separate storage.74 A large fraction of the decay heat caused by fission product decay is also continuously removed when the fission products are removed, which minimizes, if not eliminates the core cooling problems associated with emergency shutdowns. An additional level of protection is also provided by the 10 times larger ratio of heat capacity of the salt versus a dry Light Water Core, thereby making the "China syndrome" a moot point.75 Even greater safety redundency is provided by fluid fuel reactors' option of draining a malfunctioning core into the drain tanks. This is in sharp contrast to solid fuel reactors which retain all of their fission products and whose solid, fixed cores, with their large radioactive inventories have been dispersed during accidents. The worst postulated accident for a MSR is a leak. Given the experience of the MSRE where the salt was found to contain the fission products76, a leak would not be a major disaster.

Of course some may argue that the MSR may never succeed as it does not adhere to the current model of Nuclear Energy as the "razor blade" business, whereby profits are made by selling the "razor" (the reactor) at cost, but charging dearly for the "razor blades" (the fuel elements and services).77 However, it seems likely that safe, clean, economical energy will ultimately win over narrower profit horizons.

There are no technical obstacles for the deployment of a 'sealed' MSR that operates with no chemical processing. Such a reactor could operate for ~30 years with no processing other than allowing the noble gases to naturally bubble out of the salt and periodic additions of fissile material. Since it would be advantageous due to environmental, economic, and resource conservation concerns to achieve as much reclamation of the expensive salt after (or during) the 30 year operation period, a parallel development of various salt processing methods should occur78 (much of which has already been done as part of the Integral Fuel Reactor [IFR] project's fuel cycle studies). If this strategy of attempting to remove very long term fission product buildups in the salt is pursued, the salt can continue service in another reactor after the first MSR has reached the end of its life. This would not only greatly reduce decommisioning costs and difficulties (as the vast majority of the radioactivities go with the salt and removed fission products), but allow the expense of the salt to be amortized over >100 year periods!


MSR Suggested Deployment

In the past, the emphasis was always on the big power projects (such as 1 GWe power plants) so as to attain the lowest unit installed costs. It is generally recognized today that large plants, while they may promise low unit costs, have higher hidden costs in the form of expensive delays, supply-demand mismatches, disruptive and expensive down-times, large financing requirements and risks, and unexpected scaling problems. The MSR may be unique among reactors in that it does not seem to be as sensitive to lower unit costs via larger sizes. Perhaps this is due to the material and component fabrication savings due to the lack of high-pressure vessels and pipes for reactor core, circulation, and safety systems. It is interesting to note that even a large MSR as was planned for the breeding MSR (the 1 GWe MSBR) had 4 separate pumps and heat exchangers for the salt. Each of these would have been ~250 MWe in size. It is for this reason, and the desire to avoid unnecessary risks, delays, and unknowns inevitably associated with large power systems, that I propose a series of modular MSR converter reactors of about 250 MWe in size. Such a small size would lend itself to modular, mass production where the first successful MSR deployed for the purpose of burning excess weapon material in Russia could also serve as a prototype for commercial production and deployment. The only obstacles are institutional, not technical.



Footnote 1.

1,000 MWe * 1 MWth/44% MWe ÷ 925 MW-day / kilogram fissile = 2.457 g/day. Since the conversion ratio is 0.8, the net fissile feed would be 2.457 kg/day * (1-0.8) = 0.491 kg/day or 179 kg/yr; NOTE: 925 MW*day/kilogram fissile from pg. 17, NUCLEAR REACTOR ENGINEERING (1981) (Back)

Footnote 2.

~222 ppm U232 (232U) is most often the concentration mentioned in papers describing experiments involving 233U fuels. It therefore seems to be a minimum, nominal amount; a sort of worst case scenario unless heroic measures are taken to avoid 232U buildups, e.g., irradiate Thorium in 'pure' thermal neutron flux, choose thorium suppies that have no associated 230Th, irradiate for short periods to avoid 233U buildup & associated hard flux. (Back)

Footnote 3.

Page 45 of "Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options", National Academy of Sciences, 1995, National Academy Press, gives the formula to approximately calculate the surface gamma dose for a sphere as:

D (grays/hr) = 0.5 x Da x (ut / um)

"Da is the rate of gamma energy release in the solid material (J/kg-hr), ut is the mass-energy absorption coefficient for tissue" and um is the mass-energy absorption coefficient for the metal (uranium) at the 0.3 MeV energy of the emitted gamma for U233. The 0.3 MeV gamma is a sum of the gammas emitted by U233 in the range of 0.275 MeV to 0.375 MeV as given by LANL's Internet site which has online the values from "ENSDF DATED 790323" & "NOTE: THE PRECISE E-GAMMA VALUES FOR THIRTY OF THE GAMMA RAYS ARE THOSE REPORTED BY C. W. REICH ET AL., INT. J. APPL. RADIAT. ISOT. 35, 185 (1984)."

The values for mass-energy absorption coefficient for the metal (um) & tissue ( ut) are from the CD-ROM American National Standard ANSI/ANS 6.4.3-1991, pg. 14 & 15.

The calculations for the surface gamma radiation of a bomb-sized 233U sphere are:

Da = bomb mass x 1 mole U233/233 g x #atoms/mole x % emit x energy gamma x decay constant

Da = 6700 g x 1 mole U233/233 g x 6.023 x 1023 atoms/mole x 21.6%% x 0.3 MeV x ln 2/(159000 x 24 x 365) hr

Da = 5.61 x 1014 MeV/kg-hr x 1.602 x 10-13 J/MeV = 89.8 J/kg-hr

D = 0.5 x 89.8 J/kg-hr x (0.0317 cm2/g) /(0.325 cm2/g)

D = 4.38 grays/hr = 438 rems/hr (Back)


Footnote 4.

The ratio of Th230:Th232 can be calculated by:

Th230/Th232 = 1.76 x 10-5 * rU,Th.

Pg 284, "NUCLEAR CHEMICAL ENGINEERING", Benedict, Pigford, & Levi (1981). (Back)


Footnote 5.

1 kg Pa = 1,000 grams Pa = (1000 grams Pa / 233 grams Pa / mole) = 4.29 mole Pa

4.29 moles Pa = 4.29 moles Pa / 1.54 x 10-5 mole Pa / 1 mole salt = 279,000 moles salt (Back)


Footnote 6:

Calculations of masses of various Molten Salt Constituents in the Reference ORNL Molten Salt Breeder Reactor (MSBR)

71.7% LiF

0.717 mole LiF/salt mole x 26 g/mole LiF =

18.64 g/salt mole

16.0% BeF2

0.16 mole BeF2/salt mole x 47 g/mole BeF2 =

7.52 g/salt mole

12.0% ThF4

0.12 mole ThF4/salt mole x 308 g/mole ThF4 =

36.96 g/salt mole

0.30% UF4

0.003 mole UF4/salt mole x 309 g/mole UF4 =

0.927 g/salt mole



Footnote 7:

2.25 x 106 kW(th) / 2.2 x 107 kWh(th) / kg U233 = 0.1 kg U233 / hour. Since the breeding rate for new U233 to consumed U233 is (1.06 - 1) = 0.06, the rate of new U233 produced is: 0.1 kg U233 consumed / hour * 0.06 new U233 / consumed U233 = 0.006 kg/hr. (Back)


Footnote 8:

1.41 x 10-3 mole % versus the 1.54 x 10-5 mole %, or 92 times greater (so as to reduce the amount they need to steal by a factor of 92). (Back)


Footnote 9:

0.25 W/g � (1,000 g/kg) � (3,600 J/W�hr) =

900,000 J/(kg � hr) � (1 gray/(J/kg) =

900,000 grays/hr � (100 rads/gray) =

90,000,000 rads/hr

(Source: Page 566, "Nuclear Reactor Engineering", 3rd Ed., Samuel Glasstone & Alexander Sesonske (1981), Van Nostrand Reinhold Company). (Back)






1. Page 35, "Basis and Objectives of the Los Alamos Accelerator-Driven Transmutation Technology Project", C.D. Bowman, [AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, NV July 1994, pages 22 - 43]. (Back)

2. Page 97, "THE FIRST NUCLEAR ERA: The Life and Times of a Technological Fixer", A.M.Weinberg (1994), 291 pages. (Back)

3. Page 33, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

4. Pages 377 - 378, "The Molten Salt Adventure", by H.G. MacPherson, NUCLEAR SCIENCE AND ENGINEERING, Vol. 90, pgs 374-380 (1985). (Back)

5. Page 41, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

6. During a talk, "On Plutonium", for The ORNL Distinguished Lecture Series: The First Annual "Alvin Weinberg Lecture, 20 Apr 95, Professor Wolf Häfele said, "Plutonium should not be considered a waste, nor a resource, but instead, an endowment." With this view, the plutonium can act as "a catalyst" to convert into 233U using Thorium. (Back)

7. "MOLTEN SALT REACTORS FOR BURNING DISMANTLED WEAPONS FUEL", U. Gat, J.R. Engel, & H.L. Dodds, Dec. 92, Nuclear Technology, pages 390 - 394}. (Back)

8. Page 41, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72.} (Back)

9. Page 31, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

10. Page 33, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

11. Page 35, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979; Where denatured U was assumed for the calculations, & pg 379, "The Molten Salt Adventure", by H.G. MacPherson, NUCLEAR SCIENCE AND ENGINEERING, Vol. 90, pgs 374-380 (1985) (Back)

12. Page 36, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979; because the 239Pu would substitute for the depleted, or natural, uranium feed which becomes 239Pu due to neutron absorption. (Back)

13. Pages 118 - 121, "Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options", National Academy of Sciences, 1995, National Academy Press. (Back)

14. Page 44, "Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options", National Academy of Sciences, 1995, National Academy Press. (Back)

15. Page 41, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

16. Page 337, "The Development Status... ORNL-4812} table of fission product mole fractions in the fuel salt assuming 25 day removal times. (Back)

17. Page 112, NUCLEAR APPLICATIONS & TECHNOLOGY, Vol. 8 February 1970. (Back)

18. Page 35, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

19. Page 118, "Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options", National Academy of Sciences, 1995, National Academy Press. (Back)

20. Page 29, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

21. Page 9, "The Liquid Metal FAST BREEDER REACTOR: An Environmental and Economic Critique", by Thomas B. Cochran (1974). (Back)

22. Page 141, "The Results of the Investigations of Russian Research Center - "Kurchatov Institute" on Molten Salt Applications to Problems of Nuclear Energy Systems", Vladimir M. Novikov [AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, NV July 1994, pages 138 - 147]. (Back)

23. Page 122-123, "Radioactive Waste: Politics, Technology and Risk", by Ronnie D. Lipschutz, A Report of the Union of Concerned Scientists, Ballinger Publishing Company, Cambridge, Massachusetts, 1980. (Back)

24. Page 514, "Nuclear Chemical Engineering", 2nd Ed., by Manson Benedict, Thomas H. Pigford, & Hans Wolfgang Levi, published by, McGraw Hill Book Company, (1981). (Back)

25. Page 112, "MOLTEN-SALT REACTORS - HISTORY, STATUS, AND POTENTIAL", by M.W. Rosenthal, P.R. Kasten, and R.B. Briggs, in NUCLEAR APPLICATIONS & TECHNOLOGY, Vol. 8, Feb 70. (Back)

26. Page 65, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207 , 156 pages. (Back)

27. "Symbiotic system of a fusion and a fission reactor with very simple fuel reprocessing", BLINKIN, V. L.; NOVIKOV, V. M., AB (Akademiia Nauk SSSR, Institut Atomnoi Energii, Moscow, USSR), Nuclear Fusion, vol. 18, July 1978, p. 893-900. (Back)

28. "Optimization of the fission-fusion hybrid concept", SALTMARSH, M. J.; GRIMES, W. R.; SANTORO, R. T., Oak Ridge National Lab., TN., Plasma Physics, April 1979. (Back)

29. "Design of a helium-cooled molten salt fusion breeder", MOIR, R. W.; LEE, J. D.; FULTON, F. J.; HUEGEL, F.; NEEF, W. S., JR.; SHERWOOD, A. E.; BERWALD, D. H.; WHITLEY, R. H.; WONG, C. P. C.; DEVAN, J. H., TRW Energy Development Group, Redondo Beach, Calif., GA Technologies, Inc., ORNL, Presented at the 6th Topical Meeting on the Technology of Fusion Energy, San Francisco, 3-7 Mar. 1985, Nuclear and High-Energy Physics, 02/1985. (Back)

30. Page 378, "Nuclear Chemical Engineering", 2nd Ed., by Manson Benedict, Thomas H. Pigford, & Hans Wolfgang Levi, published by, McGraw Hill Book Company, (1981). (Back)

31. Page 28, "Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options", National Academy of Sciences (NAS), 1995, National Academy Press. (Back)

32. Page 31, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

33. Page 94, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207, 156 pages. (Back)

34. Page 34, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization Without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

35. Page 31, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)


37. Page 126, "EXPERIENCE WITH MSRE", by Haubenreich & Engel, in NUCLEAR APPLICATIONS & TECHNOLOGY, Vol. 8, Feb 70. (Back)


39. Pages B-478 & B-487 , CRC Handbook of Chemistry and Physics, 53rd Ed. (Back)

40. Page B-506, ibid. (Back)

41. Page 550, " Nuclear Chemical Engineering", 2nd Ed., by Manson Benedict, Thomas H. Pigford, & Hans Wolfgang Levi, published by, McGraw Hill Book Company, (1981). (Back)

42. Page 274, " Nuclear Chemical Engineering", 2nd Ed., by Manson Benedict, Thomas H. Pigford, & Hans Wolfgang Levi, published by, McGraw Hill Book Company, (1981). (Back)

43. Page 275 - 279, ibid. (Back)

44. Page 34, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

45. Page 378, "NUCLEAR CHEMICAL ENGINEERING" (1981) & pg 515, "NUCLEAR REACTOR ENGINEERING" (1981). (Back)

46. Page 94, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207, 156 pages. (Back)

47. Page 173, fig. 3, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

48. Pages 143, & 192 Table II, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72; MS mole wt. is ~64. (Back)

49. Pages 173, fig. 3, ibid. (Back)

50. Page 175, ibid. (Back)

51. Page 192, ibid. (Back)

52. Page 17, "Nuclear Reactor Engineering", ibid. (Back)

53. Page 196, TABLE IV, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

54. Page 170, ibid. (Back)

55. Page 350, ibid. (Back)

56. Page B-514, CRC Handbook. (Back)

57. Page B-515, CRC Handbook. (Back)

58. Page 145, "The Development Status..." ORNL-4812. (Back)

59. NOTE: The measurement "Rad" was used, where Rem (Radiation equivalent in man - an adjustment made by multiplying by the radiation's Quality Factor, which for gamma radiation is 1; therefore my "sloppy" interchangeable usage of rad & rem, and gray & sievert.) is actually the appropriate unit as it measures absorbed dose's effect on human tissues. Page 580 - 581, "The Effects of Nuclear Weapons", Dept. of the Army Pamphlet #50-3, (Mar 77). (Back)

60. Associated Press (AP) Online Report: "Atomic Waste May Erupt", AP 4 Mar 95 22:22 EST V0431. (Back)

61. Associated Press (AP) Online Report: "Nuclear Waste Could Blast", AP 23 Mar 95 20:27 EST V0715. (Back)

62. Associated Press (AP) Online Report: "Gov't Nuclear Plan Threatened", AP 25 Jun 95 12:20 EDT V0750. (Back)

63. Associated Press (AP) Online Report: "Nuclear Waste Piling Up", AP 24 Jun 95 11:34 EDT V0405. (Back)

64. Associated Press (AP) Online Report: "States Sue Over Nuclear Waste", AP 06/20 18:05 EDT V0167. (Back)

65. Associated Press (AP) Online Report: "Nuclear Fuel Shipments Rapped", AP 17 Jan 95 20:09 EST V0557. (Back)

66. Associated Press (AP) Online Report: "Scientists Wary Of Nuke Waste", From Message-ID: <>, Date: Wed, 2 Aug 95 0:40:17 PDT. (Back)

67. Associated Press (AP) Online Report: "Gov't Nuclear Plan Threatened", AP 25 Jun 95 12:20 EDT V0750. (Back)

68. Associated Press (AP) Online Report: "Nuclear Waste Piling Up", AP 24 Jun 95 11:34 EDT V0405. (Back)

69. Page 238, "Potential Role of ABC-Assisted Repositories In U.S. Plutonium And High-Level Waste Disposition", David Berwald, Anthony Favale, and Timothy Myers of Grummand Aerospace Corporation, and Jerry McDaniel of Bechtel. AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, NV July 1994, pages 236 - 247. (Back)

70. Page 142, "The Results of the Investigations of Russian Research Center - "Kurchatov Institute" on Molten Salt Applications to Problems of Nuclear Energy Systems", Vladimir M. Novikov [AIP Conference Proceedings 346, International Conference on Accelerator-Driven Transmutation Technologies and Applications, Las Vegas, NV July 1994, pages 138 - 147]. (Back)

71. Page 33, "Molten-Salt Reactors for Efficient Nuclear Fuel Utilization without Plutonium Separation", J.R. Engel, W.A. Rhoades, W.R. Grimes, and J.F. Dearing, NUCLEAR TECHNOLOGY, Vol. 46, Nov 1979. (Back)

72. Page 94, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207, 156 pages. (Back)

73. Page 411, "The Development Status..." ORNL-4812. (Back)

74. Page 406, ibid. (Back)

75. Page 403 , ibid. (Back)

76. Page 370, ibid. (Back)

77. Page 29, "The Development Status of MOLTEN-SALT BREEDER REACTORS", ORNL-4812, Aug. 72. (Back)

78. Page 94, "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor With Once-Through Fueling", J.R. Engel, W.R. Grimes, H.F. Bauman, H.E. McCoy, J.F. Dearing, & W.A. Rhoades, (1980), ORNL/TM-7207, 156 pages. (Back)

© Copyright, Bruce N. Hoglund, 1995

History of above paper:

This paper was to be my contribution for part of a joint paper & international conference. Due to personal, institutional, political reasons, it never was. Therefore, it has never had the sort of editing or peer review it should have. However, I believe its contents are accurate and factual. I am placing it into the public domain of the Internet so that others may learn about otherwise poorly known subjects: Molten Salt Reactors, the Thorium-Uranium-233/232 Fuel Cycle, and Proliferation features of both. I give free, responsible use of this material so long as it is properly attributed.


Comments or Questions? Please email me, Bruce Hoglund <>


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