THE SAFETY RISKS OF USING MIXED-OXIDE FUEL

IN VVER-1000 REACTORS:

AN OVERVIEW

 

Edwin S. Lyman, PhD

Scientific Director

May 20, 2000

 

 

Introduction

 

The United States and Russia are in the final stages of negotiating an agreement that would establish a framework for the disposal of approximately 34 tonnes each of plutonium withdrawn from nuclear weapons programs. According to the U.S. Government, the agreement stipulates that 25 tonnes of weapons-grade plutonium (WG-Pu) in the U.S. and 33 tonnes of WG-Pu in Russia would be incorporated into mixed-oxide (MOX) fuel and irradiated in existing nuclear reactors. The MOX fuel would be substituted for a fraction of the low-enriched uranium (LEU) that normally fuels the reactors. The remainder of the WG-Pu (9 tonnes in the U.S. and 1 tonne in Russia) would be combined with high-level nuclear wastes after being stabilized in a ceramic matrix, through a process known as immobilization.

 

The agreement calls for development of an infrastructure in both the U.S. and Russia that would be initially capable of processing 2 tonnes of WG-Pu per year. As a result of its reluctance to declassify the isotopic composition of WG-Pu, Russia will add about 12% reactor-grade plutonium (RG-Pu) to the WG-Pu before making it into MOX, for a total of 37 tonnes under the agreement. Therefore, a capacity of about 2.3 tonnes of total plutonium per year in Russia will be required.

 

The plutonium disposal infrastructure will consist of three main facilities: a plant to convert plutonium warhead components into oxide powder, a MOX fuel fabrication plant and an immobilization plant. In addition, modifications to the nuclear reactors designated to receive MOX fuel may be required.

 

Recognizing that more WG-Pu may in the future be declared surplus to the weapons program, the agreement will also require that a plan for increasing the annual disposal capacity in both countries to 5 tonnes of WG-Pu be developed within a year. According to the director of the Office of Fissile Materials Disposition (MD) of the U.S. Department of Energy (DOE), Laura Holgate, both governments hope that the capacity expansion plan will be completed in sufficient time so that it can be implemented immediately, and the 2 tonne per year stage can be bypassed. Funding for implementation of the agreement in Russia, which according to current estimates will cost nearly two billion dollars, has not yet been secured, but ultimately will have to be provided by the U.S. and other Western nations.

 

To meet the initial MOX disposition target, the U.S. is planning to use four pressurized-water reactors (PWRs) in the U.S. --- two at the McGuire nuclear station in the state of North Carolina and two at the Catawba nuclear station in the state of South Carolina. Both nuclear stations are owned by Duke Power, a private utility. In order to use MOX, the operating licenses of these plants will have to be amended by the U.S. Nuclear Regulatory Commission (NRC). A second utility, Virginia Power, was originally part of the private consortium that received the MOX contract, but withdrew in March of this year, stating that its participation no longer made economic sense for the company.

 

Russia plans to use all seven of its operating VVER-1000 reactors (four at Balakovo, two at Kalinin and one at NovoVoronezh), which are similar in concept to Western PWRs, as well as the BN-600 fast neutron reactor at Beloyarsk. Russia will require more reactors to process a similar amount of MOX fuel than the U.S. will because it is planning to use a lower MOX core fraction. In order to increase the rate of MOX disposition of Russian plutonium per the follow-on agreement, either the core fraction in the Russian reactors would have to be increased or Russian MOX fuel would have to be exported to other countries such as Ukraine, Western Europe and Canada. The preferred alternative of the Ministry of Atomic Energy (Minatom) is to build a generation of new fast breeder reactors, known as BN-800s, but the U.S. continues to oppose this alternative, which poses additional proliferation risks.

 

Increasing the amount of WG-Pu to be immobilized is another alternative for increasing the disposal rate that would be safer than the MOX options. However, Minatom vehemently opposes this approach because of a misguided belief that WG-Pu should not be "thrown away." As this report will show, Minatom's insistence in using MOX fuel could well have disastrous consequences for Russia and Eastern Europe.

 

 

The Risks of MOX in Western PWRs

 

The substitution of MOX for LEU in LWRs raises serious safety risks that have not been adequately assessed in either the U.S. or Russia. These risks, which apply broadly to both WG-MOX and RG-MOX, can be grouped into two categories.

 

First, the probabilities of certain severe accidents may increase when MOX is used. Fundamentally, the introduction of MOX fuel into LWRs reduces the effectiveness of the materials used to absorb neutrons in the core, such as the control rods and the boron dissolved in the coolant. This makes it more difficult to control the nuclear reactions in the core and reduces the margin available to safely shut down the reactor if problems arise. At the same time, the "delayed neutron fraction," a parameter which determines the speed at which the power level of the reactor responds to changes in conditions, is smaller when MOX fuel is used. This means that the operator not only has less control over transients but has less time to respond to them as well.

 

Measures can be taken to restore some of the lost control worth, such as increasing the number of control rods or the concentration of boron in the coolant. However, these are only partial fixes and may introduce other problems as well.

 

Second, the consequences of a severe accident (as measured in latent cancer fatalities and early fatalities from acute radiation exposure) involving containment failure or containment bypass (i.e. a steam generator tube rupture) will be greater if MOX fuel is in the core. This is because MOX cores have higher concentrations of actinides, including isotopes of plutonium, americium and curium. Most of these are alpha-particle emitters with large radiotoxicities if inhaled or ingested.

 

A recent report by the Nuclear Control Institute (NCI) analyzes a number of these issues with regard to WG-MOX use in U.S. PWRs. [1] This report calculates that in the event of a severe, Chernobyl-type accident with containment failure or bypass, at a PWR with a 40% WG-MOX core, the number of latent cancer fatalities would be about 25% greater than for a PWR with an all-LEU core. Depending on the population density in the vicinity of the accident, this could correspond to hundreds to thousands of additional cancer deaths.

 

The report also finds that the use of MOX could have serious negative effects on other aspects of PWR operation. Three examples are:

 

Overcooling transients and pressurized thermal shock. Certain types of accident initiators, such as a break in a main steam line, can result in a rapid temperature drop in the primary coolant system. In PWRs, when the temperature of the coolant drops, the reactivity increases (the "moderator temperature coefficient" is negative). Therefore, even if the reactor is immediately "scrammed" (that is, the nuclear reaction is stopped by inserting control rods), the reactivity and core power will increase and automatic safety systems must be activated.

 

One potential outcome of an overcooling transient is a severe accident known as pressurized thermal shock (PTS), which is of particular concern with respect to VVER-1000s (see below). PTS can occur when a steel reactor pressure vessel has been embrittled by exposure to neutron radiation. If the reactor vessel is cooled down below a certain temperature (which depends on the vessel material, the radiation exposure time and many other factors) while it is still pressurized, small cracks can rapidly expand and cause the vessel to split open. Once this happens, cooling of the fuel cannot be assured and a meltdown is likely. In addition, as the pressure vessel breaks apart it can also breach the containment, resulting in a large radiological release. After a main steam line break, active intervention by operators is required to prevent the reactor from entering a state at which PTS can occur.

 

There are a number of ways in which MOX fuel can increase the risk that an overcooling transient will progress to a severe accident. First, because the moderator temperature coefficient is more negative in MOX cores and the delayed neutron fraction is smaller, the rate at which the power increases will be greater, reducing the time margin for activation of safety systems.

 

Second, the risk of PTS will be higher if MOX fuel is in the core, since the rate at which the temperature decreases will be greater. This is because the decay heat of MOX fuel is smaller than that of LEU fuel immediately after the reactor is scrammed. An analysis of the main steam line break accident by Westinghouse shows that the temperature drops into the range at which PTS can occur after six minutes for a partial MOX core, whereas for LEU fuel the dangerous temperature range is avoided entirely. [2]

MOX will also lead to an increase in PTS risk by accelerating the neutron embrittlement of the reactor pressure vessel, since plutonium-239 fission produces a greater number of "fast" neutrons that have a high probability of escaping from the core than uranium-235 fission. While reactor operators may try to minimize this problem by placing MOX fuel elements away from the reactor vessel (i.e. near the center of the core), this creates additional power-peaking problems, since the neutron flux is higher near the center of the core. This has caused fuel management problems in at least one nuclear plant in Germany. It is unclear whether this difficulty can be adequately resolved.

Reactivity insertion accidents. In a hypothetical accident in which a control rod is ejected from a PWR core, a sudden increase in reactivity will result within adjacent fuel assemblies. This can cause rapid increases in temperature and power in fuel rods that can cause the cladding to burst open or, at a higher level, the fuel itself to fragment. At a minimum, this could cause a release of fission products into the primary coolant. At worst, the expulsion of fuel particles could block the flow of coolant, leading to higher temperatures, more fuel damage, and a potential meltdown.

 

It has now been documented that MOX fuel of the type produced by the French company Cogema, after being irradiated for more than three annual cycles, is more vulnerable to failure during a reactivity insertion accident than LEU fuel at a similar burnup. (Cogema is a key participant in the U.S. MOX disposal consortium and will design and operate the U.S. MOX fabrication plant. It is anxious to play a similar role in Russia as well.) This conclusion results from the Cabri test reactor experiment series in France, in which a MOX fuel pin that had been irradiated for four annual cycles underwent a severe failure, whereas an LEU pin with similar characteristics did not fail.[3]

 

The failure is understood to be a consequence of the inhomogeneous nature of MOX fuel, which contains plutonium-rich clumps. These clumps experience extremely high local burnups and consequently accumulate large quantities of fission gas in small volumes. When the fuel pin is subject to a reactivity insertion, the gas can be rapidly released, resulting in fragmentation of the fuel and rupture of the cladding. Two other experiments with MOX at lower burnups were carried out as well. Although these rods did not fail, the fission gas released into the gap between the fuel and cladding was considerably higher than that seen from LEU fuel rods at similar burnup. Production of inhomogeneous fuel is a fundamental flaw of the French MOX fuel fabrication process.

 

Loss-of-coolant accidents and station blackouts. Probabilistic risk assessments carried out for Western PWRs find that loss-of-coolant accidents (in which a rupture in the primary coolant system reduces the amount of water available to cool the core), and station blackouts (in which off-site power is lost) are among the largest contributors to core melt and early containment failure. The likelihood that these events will lead to such severe consequences is related to the extent to which damage occurs to the fuel cladding before the emergency core cooling systems are activated. This in turn strongly depends on the centerline temperature of the fuel rods.

 

The centerline temperature of MOX fuel rods is about 50C higher than that of LEU fuel under similar operating conditions. As a result, the rate of cladding corrosion during the initial stages of a loss-of-coolant accident will be more rapid for MOX fuel, and faster-acting safety injection systems may be necessary to prevent fuel damage and core melt.

 

 

The Risks of Using MOX in VVER-1000s

 

According to the International Atomic Energy Agency, compared with Western-style PWRs, the VVER-1000 has a number of design features which make it more vulnerable to experiencing severe accidents that could lead to radiological releases. These include design of the plant layout, which does not ensure sufficient physical separation of diverse safety systems to prevent common-mode failures. These design flaws are compounded by a number of other issues that affect safety, including the type and quality of materials used in construction, the reliability of instrumentation and control systems, the inadequacy of the maintenance during the current difficult economic situation and the lack of thorough and independently reviewed safety documentation (accident analyses, emergency operating plans and severe accident management guidelines).

 

Although Western nations have joined in efforts with states in Eastern Europe and the former Soviet Union to correct some of the deficiencies in reactors of Soviet design, most of the emphasis has been on reactors of the RBMK or VVER-440 type, which are considered to be of higher risk than the VVER-1000 series. In no instance, however, has the West committed to financing upgrades of Soviet-designed reactors to conform to Western safety standards, an endeavor which has been estimated to cost hundreds of millions of dollars (billions of rubles) per reactor.

 

To the extent that safety margins are smaller in VVER-1000s than in Western PWRs, the further reduction in margin associated with use of MOX will be of even greater concern for VVER-1000s.

With regard to the additional public health consequences associated with MOX use, the results for VVER-1000s should be very similar to those cited above for U.S. PWRs --- about a 25% increase in latent cancer fatalities following a severe accident with containment failure or bypass. The relatively high concentration of actinides in MOX cores compared to LEU cores is a generic feature of light-water reactors. A recent study by the Kurchatov Institute calculated that at a burnup of 60,000 megawatt-days per tonne, a WG-MOX VVER-1000 fuel assembly would have around twice as much plutonium and around three times as much americium and curium as an LEU assembly.[4] These results are nearly identical to those we found for U.S. PWRs.

 

In fact, the consequences may be even greater than those calculated for U.S. PWRs. This is because Minatom intends to mix WG-Pu with about 12% RG-Pu to obscure the isotopic composition of Russian WG-Pu. This will increase the quantity of higher plutonium isotopes in the unirradiated MOX fuel to be used in Russia. As a result, the concentrations of americium and curium isotopes in the irradiated fuel will be higher than if WG-Pu was used, and the associated radiation exposures following an accident would be higher.

 

Because the most severe consequences of an accident at a MOX-fueled VVER-1000 would result from a failure or bypass of the containment, unresolved issues associated with VVER-1000 containment and steam generator integrity will become even more urgent if MOX fuel is used. The Nuclear Energy Agency (NEA) has raised questions about the long-term integrity of the pre-stressed concrete containments at VVER-1000s, and has cited the need for research "to develop a model which could predict the loss of performance as a function of loss of pre-stress...".[5]

 

The risk of containment bypass is a significant concern, especially with regard to the integrity of the steam generator collectors. A leak in a steam generator establishes a pathway from the primary to the secondary coolant systems and hence to the environment. VVER-1000 steam generator collectors experienced numerous failures in the mid-1990s, and despite temporary fixes the NEA states that "assurance of the VVER-1000 steam generator collector integrity remains one of the most important safety issues."

 

A number of the deficiencies of VVER-1000s should be of particular concern with regard to the use of MOX fuel:

 

Pressurized thermal shock. Foremost among them is the embrittlement of the reactor pressure vessels (RPV) and the risk of pressurized thermal shock (PTS).

 

According to the NEA, one of the most urgent open issues for VVER-1000 safety is an evaluation of the integrity of the reactor pressure vessel "which takes into account the actual condition of vessel material under operating conditions."[6] The radiation resistance of RPV materials is highly sensitive to slight variations in the composition of the steel used in the vessel and welds, as well as reactor-specific details of the operating history. For instance, some VVER-1000 vessel weld joints contain more than 1.5% nickel and have significantly reduced radiation resistance. The NEA cites in particular the lack of a detailed understanding of the conditions to which the weld areas are exposed during reactor operation, the lack of non-destructive assay methods that could locate micro-cracks in the RPV and assess the extent of its embrittlement, and the lack of verified PTS assessment methods.

To cope with this problem, Russian engineers have developed a process known as "annealing," in which heat treatment is applied to the RPV and welds to restore their crack resistance. However, the benefits of annealing have not been decisively verified and it is not known how long they last. In addition, annealing itself can have a detrimental effect on the RPV because it can cause "sensitization" of stainless steel, a process that significantly reduces resistance to localized corrosion.

 

Compounding the uncertainties in the integrity of VVER-1000 RPVs is the lack of adequate analyses of overcooling transients that could initiate PTS, such as the main steam line break.

 

The IAEA reaffirmed its concerns regarding VVER-1000 RPV integrity and PTS risk in the report of a 1999 conference.[7] The author has also been privately informed by an official from the IAEA's Department of Nuclear Safety that all VVER-1000 RPVs are embrittled. Moreover, the official has little confidence in the effectiveness of the annealing procedure to mitigate this problem.

 

Given these concerns, the potential increase in the risk of pressurized thermal shock as a result of using MOX fuel must be fully evaluated. Since this risk is associated with plant aging, it will continue to become more acute if VVER-1000 lifetimes must be extended to accommodate the plutonium disposal program.

 

Reactivity insertion accidents. There are differences between VVER-1000 fuel rods and PWR rods which could exacerbate the problems seen in high-burnup PWR MOX rods during the Cabri reactivity insertion accident tests. For example, unlike Western PWR fuel pellets, VVER-1000 pellets have a hole in the center. The hole was intended to reduce temperatures in the center of the fuel pellet and to provide extra volume for fission gas to accumulate, thus reducing the internal rod pressure. However, in practice this does not seem to work. Because the center of the fuel pellet is the hottest region, the gas accumulating there reaches a higher temperature than is typical in PWRs, so that "the internal fuel rod pressure is considerably higher than in a PWR for a given filling pressure and plenum gas volume."[8] According to the IAEA, "the merits of the hollow UO₂ pellet design used in WWER [VVER] fuel rods are debatable." Fission gas release fractions in high-burnup VVER-1000 fuel rods of up to 70% have been observed, compared to typical values of 5% or less for PWR fuel.[9]

 

The higher fission gas release in LEU VVER-1000 fuel rods compared to PWR fuel rods raises questions about whether MOX can be used in VVER-1000s without significant changes to the fuel rod design, since one would expect fission gas release in VVER-1000 MOX to be even higher. This would result in an even greater vulnerability to clad rupture during reactivity insertion accidents and LOCAs than was observed in PWR fuel in the Cabri test series.

 

Russian engineers claim that the cladding used in VVER-1000 fuel rods, which is a zirconium-niobium alloy, is more resistant to failure during reactivity insertion accidents, even for high burnup fuel, because it does not corrode and become embrittled during irradiation as easily as conventional PWR cladding, which is pure zirconium. However, if true, this property would not prevent the type of MOX fuel failure seen in the Cabri tests, since the MOX rod that failed was not heavily oxidized.

 

Conclusions

 

The risk to public health and safety associated with using MOX fuel in both U.S. PWRs and Russian (and Ukrainian) VVER-1000s is considerable and must be fully assessed before plans to go forward are undertaken.

 

In our view, these additional health and safety risks are unnecessary to achieve the goal of warhead plutonium disposition. The immobilization alternative is capable of safely increasing the inaccessibility of plutonium without incurring the risks associated with reactor irradiation. However, short-sighted technocrats and bureaucrats in both the U.S. and Russia, encouraged and lobbied by nuclear industry representatives from Western Europe and Japan, appear determined to load MOX fuel into reactors no matter how large the risk. The only hope of stopping this dangerous program lies in the principled resistance of the residents of communities which would be most directly affected by a Chernobyl-type accident in their backyards.

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Edwin Lyman, Scientific Director of the Nuclear Control Institute (NCI) since 1995, received a PhD in physics from Cornell University in 1992. Prior to joining NCI, he was a researcher at Princeton University's Center for Energy and Environmental Studies.