Monday, February 23, 2015

Metal Fires in Fast Reactors: Part I

The origin of this post was from an extensive quotation the first part of of a Sandia NL report on liquid sodium cooled fast reactors.  The question I am addressing is a simple one, how safe is the IFR. Although the backers of the IFR tout its safety, they also advocate for more government investment in IFR r&D.  Although they do not admit it, among the issues requiring futher R&D, as this report clearly shows is safety.

"Metal Fire Implications for Advanced Reactors, Part 1: Literature Review," is an important report for anyone who is interested in the IFR or other fast sodium cooled reactors. The issue of fire in Sodium cooled reactors is an important one, which desirves serious attention. IFR advocates argue that the IFR is highly safe. My own review of the ABTR answered many of my questions about IFR safety, but I am not a nuclear safety expert, and my findings should not be the last word on IFR safety. The good thing about the current Sandia report is that it comes from Sandia rather than Argonne, and therefore the writers cannot be accused of IFR cheerleading. The report is well written, and it is quite approachable by none scientists, who are looking for more information about the problem of IFR/LMFBR safety. The report demonstrates that progress has been made on LMFBR safety, but does not support claims that further LMFBR safety research is unneeded. I do not intend to post all of the post, but rather to call the readers attention to some passages, in the hope that the readers interests will be ignited.

Metal Fire Implications for Advanced Reactors, Part 1: Literature Review

By Tara J. Olivier, Ross F. Radel, Steven P. Nowlen, Thomas K. Blanchat, & John C. Hewson

Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

The anticipated nuclear power renaissance hinges on public acceptance and a demonstrated treatment of potential safety issues, particularly for advanced reactor designs. The Advanced Burner Reactor (ABR) uses a liquid sodium primary coolant as do certain other advanced reactor concepts. In contrast to today’s Light Water Reactors (LWRs), liquid-metal- cooled reactors present a unique risk; namely, potential metal fires involving the sodium coolant.
Fire is a significant contributor to total nuclear plant risk for current generation LWRs. Given “passively safe” advanced designs, some elements of plant risk will diminish substantially. Fires could represent the dominant risk contributor, especially given the unique characteristics of metal fires such as very high temperatures and fire suppression challenges. Fast breeder reactors all over the world use liquid sodium as a coolant and there has been experimental and analytical research done related to sodium fires as early as the 1950’s. The research has included fundamental studies, work on droplet combustion, pool burning, suppression, and large-scale sodium fire experiments. However, there are gaps in our understanding of the basic combustion behavior and combustion mechanics due to the complexities involved. These gaps have led to little progress in understanding the basic combustion behaviors for sodium. (Makino 2006). Many of these same concerns were noted as far back as 1972 (Newman 1972).

New technologies have substantially improved fire computer modeling capabilities, but to apply these tools to a sodium fire will require some additional model development and validation work. Unfortunately, most of the experiments performed in the past cannot be used to support model development today. Clear definition of the experimental boundary and initial conditions are necessary to create the modeled conditions, and most of the experimental results lack this information. “Reports of precise conditions in experiments are rare in the literature,” so the heat transfer evaluations have almost been impossible (Makino 2006).

This report includes four elements. First, a comprehensive review will define the current state of knowledge for metal fires. This will include actual metals fire experience in various applications. Second, an assessment of advanced reactor concept designs and identification of the unique metal fire safety and hazards was completed. A number of potential safety scenarios exist and will be grouped as to potential importance and representative physics to prioritize the specific research directions that will maximize breadth of applicability to emerging reactor designs. Third, a detailed review of sodium combustion research and potential approaches to the design and conduct of future experiments will be presented. Fourth, Appendix A presents an annotated bibliography of relevant literature identified during extensive literature review.

This chapter describes past sodium fires at nuclear reactors and other sodium facilities. The incidents discussed in this chapter were chosen to highlight the most significant issues surrounding sodium fires. These issues include design defects at startup (Monju), pipe bursts (BN-600), sodium spray fires (Almeria), and sodium-concrete interactions (ILONA)1.

2.1 Monju Prototype Fast Breeder Reactor

The Monju Prototype Fast Breeder Reactor (FBR) first reached criticality in 1994. Powered operation began in 1995, and a series of power raising tests were performed, with a planned full-power test planned for June 1996. Monju is a loop-type 280 MWe sodium-cooled reactor with mixed oxide fuel (Mikami 1996). During normal operation, the inlet and outlet sodium temperatures in the primary coolant loop are 397 °C and 529 °C, respectively. Sodium temperatures in the secondary coolant loop range between 325-505 °C.

During a scheduled power rating test (40% electrical power) on December 8, 1995, a high sodium temperature alarm sounded at the outlet of the secondary side of the intermediate heat exchanger (IHX) (Mikami 1996). At the same time, smoke detectors sounded in the same area, closely followed by a sodium leak detection alarm. Operators began normal plant shut-down procedures, but after increased smoke was observed 50 minutes later, it was decided to manually trip the reactor. This shutdown occurred approximately 1.5 hours after the initial alarms sounded.

Investigations later confirmed that a sodium leak and fire had occurred, ultimately, the source of the leak was traced to a damaged temperature sensor (pictured in Figure 1). The sensor consists of thermocouple wires housed in a protective well tube. It was found that the tip of the well tube had broken off and the thermocouple was bent at an angle of 45 degrees toward the downstream flow direction.

A microscopic inspection of the flow tube was performed to determine the root cause of the leak. It was concluded that the breakage of the well tube was caused by high cycle fatigue due to flow induced vibration in the direction of sodium flow. It was found that the problems were rooted in the design of the well tube. Although designers applied ASME standards to prevent resonant vibrations, they failed to take into account the sharp taper of the Monju tube design. As a result, the vortex-induced vibration could not be prevented. The design has subsequently been re-evaluated.

In addition to replacing all similarly designed temperature sensors, aspects of sodium fire response and emergency operation procedures were also modified at the Monju site. For example, the reactor will be shut down immediately if a sodium leak is confirmed in the future. A summary of the Monju Improvement Plan is shown in Table 1 (Mikami 1996). As this event was confined to the secondary coolant loop, there was no radiological release that affected either the general public or the plant personnel. However, it has resulted in over a decade of safety reviews in order to re-establish both technical surety and public confidence in the plant. The Monju plant is scheduled to resume operation in mid-2008.
The brief discussion of the ABTR is an excellent indicator of the actual developmental status of the IFR. Once again we see clearly that the IFR is not ready for commercial implementation. 
3.1.4 Advanced Breeder Test Reactor

The ABTR is a sodium-cooled, pool-type reactor based on experience gained from the Experimental Breeder Reactor-II (Chang 2006). It is a 95 MWe design with an estimated 38 percent plant efficiency. The ABTR was developed as a test bed for a similar commercial design-the Advanced Breeder Reactor. The ABTR design uses a 20 percent TRU, 80 percent uranium metal fuel clad with HT-9 stainless steel. A summary of plant specifications is provided in Table 8.

There are numerous objectives of the ABTR design, including demonstration of reactor- based transmutation of trans-uranics as part of an advanced fuel cycle, qualification of the trans-uranic-containing fuels and advanced structural materials needed for a full-scale ABR, and supporting the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. ABTR designers also have the following objectives:

• To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in advanced breeder reactors;
• To demonstrate improved technologies for safeguards and security;
• To support development of the U.S. infrastructure for design, fabrication and construction,
testing and deployment of systems, structures and components for the ABR.

3.3 Sodium Fire Consequences

3.3.2 Core Voiding

A fundamental difference between water and sodium-cooled reactors is the void reactivity coefficient. If the water around the core is voided (boiled, drained) in a water-cooled (thermal) reactor during operation, the power level will automatically drop. The reactor is therefore said to have a negative void reactivity coefficient. In contrast, if sodium is voided in certain sodium-cooled fast reactors (particularly large reactors), it will cause the power level of the reactor to rapidly increase. This reactor is said to have a positive void reactivity coefficient. When the reactor power increases, it can lead to additional boiling and voiding until fuel melts. This positive feedback can lead to extremely rapid surges in reactor power, potentially damaging or melting fuel and cladding.

Multiple events can lead to core voiding during operation, and great care is taken in the proposed new reactors to ensure that these events are prevented. They include sodium boiling, loss of coolant accidents (LOCA), and gas bubble entrainment within the sodium. Sodium fires could lead to sodium boiling if an undercooling event is initiated without scram (reactor shutdown). A severe leak in the secondary system, perhaps coupled with cable fires could lead to this situation. A large leak in the primary system could also disrupt flow enough to induce sodium boiling in the core. A sodium leak in the primary system could also lead to either a LOCA or gas bubble entrainment event. A large primary leak could potentially uncover a portion of the core. If gas is pulled back into a leak in the primary system, the resulting bubbles could also reach the core.

3.3.3 Loss of Heat Sink

A loss of heat sink event can be triggered by sodium leaks in the steam generators. As stated above, the standard procedure in response to these leaks is to drain one or both sides of the steam generator. In the event that multiple steam generators are compromised, reactor cooling must be accomplished with backup safety systems. In the case of the new generation of reactors, these safety systems are generally passive in nature (i.e. they require no operator intervention). These systems ultimately rely on natural circulation driven by core decay heat, and so are also independent of cable fires or loss of site power. In addition to these engineered safety features, the inherent high heat capacity of the sodium and structural elements of the reactor will provide valuable time for operators to restore the system to normal.

3.3.4 Loss of Engineered Safety Systems

The inherent mobility of a fire can cause a fire to become a threat to an entire reactor system. Numerous examples exist of cable fires causing serious problems in a nuclear power plant. Perhaps the most famous of these is the 1975 Browns Ferry fire, where all of the normal core-cooling functions were lost due to a cable fire (Nowlen 2001). However, operators were able to maintain core cooling with a control rod drive pump not included in plant procedures. The fire at Greifswald burned for about 92 minutes causing a station blackout and the loss of all active means of cooling the core (Nowlen 2001). As a result, a pressurizer relief valve opened and failed to close. This situation persisted for at least five hours and led to depletion of the secondary and primary side coolant inventories. The plant was ultimately recovered through initiation of low pressure pumps, the recovery of off-site power, and the recovery of one auxiliary feedwater pump.

These and other incidents demonstrate the need for next-generation sodium-cooled reactors to consider the potential impact of fire on safety systems to maintain core cooling, including the passive safety systems. Every adverse situation cannot be anticipated or avoided. However, if the reactor safety systems operate independent of the plant operators and electrical systems, then these systems can likely maintain cooling until plant personnel put out fires and regain control of the situation.

There is one additional factor that is unique to metal fires that may need to be addressed. Conventional (i.e., non-metal) fires are not generally considered a threat to primary plant piping components used in a light-water reactor (Nowlen, Najafi, et al. 2005). This would include the primary piping itself and other piping equipment such as large valves, check valves, and water- filled vessels (e.g., storage tanks). However, sodium fires burn at much higher temperatures than do other types of fires. Hence, metal fires could represent a threat to components and equipment not normally considered fire-vulnerable. For metal-cooled reactors, the performance of plant safety systems and equipment under fire conditions, including the passive safety systems, should be evaluated in this context.

3.4 Summary

Of all the new reactor designs proposed in the Generation IV program, sodium fast reactors have the largest experience base. Thanks in large part to this experience, numerous engineered safety features and safety procedures have been built into the next generation of sodium-cooled reactors. These features are designed to reduce the likelihood and consequence of sodium release and fire.

The risk of sodium release and fire exist during the three stages of a reactor lifetime; startup, day-to-day operation, and refueling and maintenance. Based on past experience, design and manufacturing defects have generated the greatest risk of sodium leakage and fire at reactor startup. Pipes, welds, and steam generator tubes are the most likely components to fail during routine operation. Thermal and mechanical fatigue must be avoided to minimize the chance of these failures. Refueling and maintenance accidents are generally caused by a combination of improper procedures and human error. The experience gained in existing reactors should help to minimize the chance of these leaks.

Sodium fires at any facility can cause serious problems beyond the immediate burn area. However, a sodium fire at a nuclear reactor can have consequences beyond those possible at non- nuclear facilities. The most notable consequences of sodium fire at a nuclear power plant include smoke in the control room, core voiding, reactor under-cooling, loss of heat sink, and loss of engineered safety systems. Sodium fires burn much hotter than other types of fires and might therefore threaten plant equipment, such as piping elements that are not normally considered vulnerable to fire damage. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation of sodium-cooled reactors. However, some unique considerations do come into play with sodium fires.
The report conclusion will serve to demonstrate that IFR safety research still has aways to go.


This report documents the results of the initial stage of the “Metal Fire Implications for Advanced Reactors” Laboratory Directed Research and Development project. Efforts to date have included an extensive literature search to cover the sodium fire recorded accidents, the proposed LMFBR designs and safety concerns and sodium fire combustion experiments and research.

Past experiences/accidents with sodium fires at nuclear and non-nuclear sodium facilities were investigated to identify the types of hazards that must be accounted for when designing the next generation of sodium-cooled nuclear reactors. The risk of sodium release and fire exists primarily during the three stages of a reactor lifetime; startup, day-to-day operation, and refueling and maintenance. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation of sodium-cooled reactors.

A need also exists to improve the state-of-the-art fire modeling codes to include the sodium fire combustion phenomenon. The past experiments did not record the details of the boundary conditions for both pool and spray fire scenarios. A lot of the experiments were small scale compared to the amount of sodium that could be involved in a HCDA. There exists a need to understand the phenomenon of inter-droplet interactions in a spray fire scenario. There has not been any experimental work to address this. Fire is one of the key parameters in a NPP risk analysis. With the GNEP program making progress forward, expertise in metal fires is essential for Sandia National Laboratories.


Anonymous said...

With many module 50 MW to 100 MW class LFTR spatially distributed in an electrical power network, there will be accidents. To think otherwise is simply unrealistic. LFTR have many potentially dangerous failure modes, just like any other type of nuclear reactor. For LFTR, its fuel salt is particularly hazardous, with hard Gamma emissions, which means that cleaning up after a LFTR would be extremely difficult. What use is a melt plug and drainage pipe from a LFTR, if debris from a disintegrated wall separating primary and secondary core circuits becomes blocked, so that the dump tank is ineffective in an even of gross overheating (for example)? The high neutron flux of a LFTR will cause severe embrittlement of the reactor container, with a risk of fracture and formation of cracks. The ORNL reactor did not operator sufficiently long and at sufficient power for such problems to be identified.

The danger with LFTR is that its advocates seem to be looking at the World through rose-tinted glasses, i.e. unrealistic.

With potential advances in much safer LENR for power generation, the role of LFTR is better as a transmutation apparatus for disposing of high-level nuclear waste generated from convention BWR. The USA has 77000 tonnes of high-level nuclear waste, and Japan has 17000 tonnes of such waste, mostly stored at nuclear reactor sites (such as Fukushima Dai'ichi).

The problem with LFTR is that there is presently too much uncritical hype; slick salesmen. The reality is that Thorium LFTR will be useful, but has its severe disadvantages also, and a realistic view on matters is urgently required.

Charles Barton said...

First, LFTR advocacy is not the same as molten salt advocacy. By7 focusing solely on a long term prospect, the LFTR, you ignore Uranium Cycle thermal reactors without thorium. Your argument linking the core melt valve drainage systen to the heat exchange seems trivial. The drain could feed directly from the core. Thirdly oRNL knew more than you seem to about core matirials, and had developed a nickel alloy core material that was resistant gt to embitterment. Terrestrial Energy has offered an alternattive approach. Build simple low cost cores that are replaced after 7 years of continuous use.

Once again you offer speculative assurtions, rather than ideas linked to well researched documents. You ignore many of the MSR safety ideas discussed in Nuclear Green. Finally you accuse MSR advocates of being slick salesmen, while advocating cold fusions, an idea regarded by much of the science community as suspect at best.


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