Fire in Sodium cooled reactors: A Sandia Literature Review

"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

Abstract
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.

1. INTRODUCTION
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.

2. PREVIOUSLY RECORDED SODIUM FIRE ACCIDENTS
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.

5. CONCLUDING REMARKS

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.

The History of Reactor Safety: Chernobyl

Nuclear safety has been an important concern of reactor designers since the first reactors were designed. The earliest reactor designers realized that some reactors designs were safer than others. Safety flaws were known to exist in the reactors built in the Hanford, Washington complex. For that reason the complex was situated in a relatively remote lightly inhabited area.

The safety problem with the Hanford reactor stemmed from the use of water cooling in a graphite reactor. If the reactor heated up to much, the water inside the reactor could could boil within its pipes. Water inside the graphite moderated core, tends to act as a break on the chain reaction. When water is removed from the water cooled graphite reactor the break is released, and the chain reaction speeds up. That adds more heat to the reactor core, and the added heat boils off more water. This process can build up very rapidly until there is a steam explosion that could potentially destroy the reactor core.

Why would anyone ever build such an unsafe reactor? The answer in the case of the United States, was that the Hanford Reactors were built because the American Government believed that it had a military necessity to do so. In 1942, the Germans were known to be developing nuclear technology, and that technology was assumed to have a military purpose. Few of the people who were involved at the time believed that nuclear safety was the most important issue. The danger to national survival that would emerge if Germany was the first to develop a nuclear weapon was the primary concern. Thus the decision was made to build unsafe reactors at Hanford because the reactors would be an important source of weapons grade fissionable Plutonium. The Hanford reactor design was never copied elsewhere in the United States, and when the United States Government around 1950 made the cold war decision to expand its plutonium output, a different, safer reactor design was chosen.

The political leadership of the Soviet Union dictated that nuclear safety was not an issue, When Yuri Andreyev took an examination to become a Soviet nuclear plant operator, he was asked to describe how a reactor could explode. He answered the question by describing three different scenario. The examiner criticized Andreyev 's answer,

"Keep it in your mind, man -- Soviet reactors cannot explode,"

the examiner told him.

With this attitude the Soviet Union chose to copy the Hanford design for plutonium production reactors. Later these military heated coolant water from these reactors was used to produce steam that drove electrical generators. This experiment was considered successful, and the Soviet leadership chose to develop a new class of power generating civilian reactors that was based on the old, unsafe Hanford design, During the 1950's the Soviets began producing a slightly modified Plutonium production reactor, that could also generate steam for electrical generation. These reactors were designed with little regard to the nuclear safety concerns, but they were cheap and easy to manufacture, so the Soviets adopted the design for a single purpose power reactor, the RBMK.

All of the safety problems of the original Hanford design were still present in the RBMK. While thinking about nuclear accident containment had advanced in the West, the Soviets failed to see the point. Western researchers had investigated how radioisotopes could escape into the environment in a nuclear accident, and how that escape could be prevented. Western power reactors reactor designs included a system of multiple barriers to the release of radioactive materials into the environment in case of an accident. The Western nuclear containment system was tested at Three Mile Island when a an error by a reactor operator turned a minor nuclear incident into a core meltdown. The TMI accident was contained without and known illnesses or deaths as a consequence. Some radiation was released in the form of radio active gases, but radio active gasses are also released by burning natural gas in the home. Despite the evidence that the Western containment system prevented radiation related deaths and illness, the Soviets slightly modified RBMK reactor building design to improve containment, but this modification still left RBMK containment far short of Western standards.

In addition to its known safety problems, the RBMK design included a serious hidden safety flaw. Sonja Schmid, in a study of Soviet nuclear safety practices notes,

Soviet design choices display a similar circular pattern: while the “Sovietness” of the graphite-water reactor (the “Chernobyl’ type” reactor) was invoked to legitimize its development, mass production and implementation, the reactor design itself then served as proof of Soviet technological prowess.

In addition to the already noted RBMK safety flaws, there was a defect in the design of its control rods. There was a graphite extension of the control rod, that entered the reactor prior to the neutron absorbing metal that formed the main body of the control rod. The control rod channels were normally filled with water, which as I have already noted acted as a break on the chain reaction. The presence of more graphite actually accelerated the chain reaction, as did the displacement of water in the control channels. As a result, as the control rods entered the reactor, there was a massive spike in core reactivity. Power output rose quickly to an estimated 30 Billion thermal watts, ten times what the RBMK was designed to handle. As this process unfolded, the heat of the reactor core increased dramatically, fuel elements and full channels began to rupture and pressure built up in the coolant water tubes As steam began to escape the rupturing tubes, steam pressure inside the reactor case increased, until the case explosively ruptured. A two thousand ton plate that covered the top of the reactor was blown off by the force of the explosion, opening the upper surface of the reactor to the sky. Three seconds later, an even more violent explosion occurred ripped the core apart, flinging chunks of burning radioactive matter high into the air.

Under ordinary circumstances graphite does not burn, but the graphite in the Chernobyl was heated white hot, and the mass of superheated graphite was exposed to an airflow, that ignited it. Then all of the Soviet nuclear mistakes, all of the Soviet arrogance that vainly assumed that nothing bad could happen to Soviet reactors came home with horrific force.

The Chernobyl reactor disaster was not simply a result of poor reactor design, but it was the result of an ideological system that believed that Soviet reactors could not fail. Mistakes made by the Chernobyl operators played a very large role in the accident. Those mistakes were the result of a test requirement imposed on the reactor staff, by the authoritarian Soviet system. Reactor safety issues were simply ignored in the performance of the test, and the reactor staff drove the reactor to and past its breaking point. It is unlikely that there would have been a Chernobyl, had there not been a Soviet Union.

Reactor safety and dam safety

Reactor safety and dam safety
(Cross posted with updates from bartoncii)

Critics of nuclear power often resort to irrational arguments. For example the words "Three Mile Island," are assumed to have great value in forming their argument, yet defenders of nuclear power point to the Three Mile Island incident, as an example of the safety of nuclear reactor. Infect there were no direct or indirect deaths associated with the "Three Mile Island" incident. A number of court cases, alleging that radiation exposure from the Three Mile Incident caused cancer, but none of the cancer victims could show that they had been exposed to radiation. Numerous research studies conducted in the area of the Three Mile Island reactor incident, showed that no increase in cancer cases developed in the Three Mile Island area after the incident. Thus the Three Mile Island incident demonstrated that even in the event of a major reactor accident, American reactors are designed to protect the public against dangerous exposures to radiation and radioactive materials.

During the 28 years since the Three Mile Island accident there have been important advances in reactor safety. New reactor designs include passive safety features, that would have prevented the TMI accident if they had been included in the Three Mile Island reactors. Critics of Nuclear power steadfastly refuse to recognize advances in nuclear safety. Indeed, most creatics of power, do not appear to understand the difference between reactor designs.

Buy are nuclear critics really concerned about human safety? If they were concerned about safety, then why haven't they looked at dam safety issues. Dams are far more dangerous than reactors.

Am October 2007 news report stated that the US Army Corps of Engineers believes that the Mosul Dam, the largest dam in Iraq is in imminent danger of a collapse that would trigger the biggest flood Iraq has seen since the time of Noah. The report suggest that the flood could kill as many as 500,000 people. In September 2006, the US Army Corps of Engineers discobered that the dam was in danger.

"In terms of internal erosion potential of the foundation, Mosul Dam is the most dangerous dam in the world," the a corps report warned. "If a small problem [at] Mosul Dam occurs, failure is likely." The was built on water-soluble gypsum, and it is leaking like a seive. The corp describe the dam site as "fundamentally flawed".

A collapse of the Mosul dam, would unleash its trillion-gallon lake which would put the nearly 2 million people of the Iraqi city of Mosul under 20 metros (70 feet) of water and parts of Baghdad under 4.5 metros (15 feet) of water, according to Abdulkhalik Thanoon Ayoub, the Musol dam manager. As many as 500,000 people could be killed by the dam failure.

The Wolf Creek Dam:
Most dangerous dam in the United States?

Closer to home, two dams on the Cumberland River Wolf Creek Dam, and Center Hill Dam, are believed to be at high risk for collapse, due to water seepage underneath the dams' foundations. Last year, following the discovery of new leaks, dam engineers announced they were lowering the lake behind Wolf Creek Dam by another ten feet to prevent its collapse. The lake was already 33 feet below its maximum capacity. The dam has been classified as a “high risk” for failure by the US Corp of Engineers. Repair work was underway for 2 years prior to the January 2007 panic. Down the Cumberland river from the Wolf Creek dam are the cities of Nashville, the home of 600.000 people, and Clarksville, home to another 125,000. Low lying areas of those cities would be inundated by any flood unleashed by a Wolf Çreek Dam failure.

In 1983, following unusually heavy rains that filled Powell Reservoir, engineers at the Glen Canyon Dam on the Colorado River, discovered to their horror that water rushing through the dam spillwar was rapidly eroding it, imperiling the huge dam. In order to save the dam, the engineers had to shut the spillway down, and they decided to do that by blocking it with 1"thick, marine grade plywood. That's right folks, all that prevented the collapse of a huge dam, that impounded a 185 mile long lake, was a 1" thick sheet of plywood. Nothing the engineers can do will prevent the situation from happening again. Next time the plywood might not hold. The Glen Canyon spillways were not designed for the discharge of water for prolonged periods of time. High volume use of these spillways for more than a week or two would most likely lead to their catastrophic failure. Once the spillways failed, complete breaching of the dam could occur in a matter of hours. The Bureau of Reclamation did a study of what would happen if the Glen Canyon Dam failed. The study found in the event that the dam was overtopped or breached, a five hundred feet high flood would reached the Grand Canyon. The flood would be 230 feet high when it reached Lake Mead. The study concludes: “ The failure of Glen Canyon Dam due to overtopping would produce catastrophic flooding with unprecedented flood depths and discharges all the way to Lake Mead and Hoover Dam. Even if Hoover Dam did not fail, there would be unprecedented flooding downstream of Lake Mead as well”.

In 1928, the collapse of the St. Francis dam outside Los Angeles unleashed a wall of water 135 feet high that killed 450 people. A 38,000 acre-feet of water was impounded behind the dam. The water weighed almost 52 million tons. The flood destroyed 1,200 homes and demolished 10 bridges. Other American dams have broken with loss of life. For example, the Baldwin Hills Reservoir dam broke in 1963. Five people died. In 1972, coal slurry impoundment dam #3, at Buffalo Creek Hollow in West Virginia broke, killing 125 people, and injuring another 1121. The 1976 collapse of the Teton Dam, on the Teton Riber in Idaho killed 11 people. The Kelly Barnes Dam, located in Stephens County, Georgia, collapsed in 1977 killing 39 people. The most catastrophic flood that occurred as the result of am American dam failure was the Johnstown, Pennsylvania flood, of 1889, which killed over 2,200 people.

Even the imfamous Johnstown flood pales in comparison to the 1975 failure of 62 Huai River dams in China. The Banqiao Dam was designed to withstand a 1000 year flood, which was estimated to be one from a storm that would drop 0.53 meters of rain over a three day period.
In August of 1975 catastrophic 3'+ a day rainfall produced a 1-in-2,000 year flood in the Huai River basin. The Shimantan Dam reservoir filled to twice its capacity and broke on August 8, 1975. A half hour later, the Banqiao Reservoir overtopped the Banqiao Dam, and quickly destroyed that dam's structure. A wall of water 6 miles wide, and 23 feet high rushed down river at a speed of 30 mph. Numerous down stream dams were breached, and an area of 4,600 square miles temporarily turned into a lake. As many as 235,000 people died as a result of the flood and its aftermath.

The Banqiao Dam after its failure

Finally the huge Three Gorges Dam, poses the mother of all flood dangers. The dam is located near a simismic fault. Dam critics have argued that dam designers have underestimated the stresses an earthquake would impose on the dam. The Three Gorges Dam was designed to withstand a "thousand year flood." But then so was the failed Banqiao Dam. Critics of the Three Gorges Dam have pointed to numerous optimistic assumptions, which its designers made. What is so frightening about the Three Gorges Dam is that 75 millions people live down stream from it, and would be threatened by its failure.

'This is a geologically risky area and the dam definitely increases those risks,' said Chen Guojie, a geologist at the Institute of Mountain Hazards in Chengdu. Chinese officials recently admitted that the huge weight of the water behind the dam had started to erode the Yangtze river's banks in many places, which, together with frequent fluctuations in water levels, had triggered a series of landslides. Landslides have caused waves 50 metres high on lake waters.

Officials also said that another four million people in the area would have to be relocated from around the dam because of safety issues. A failure of the dam would unleash a flood 186 meters (600 feet) high. The lake behind the Three Gorges Dam is 600 km (400 miles) long. A flood of that magnitude would be unimaginable.

Despite repeated dam failures and the large loss of human life, and the dangers of further catistrophic failures, nuclear critic like Helen Caldicutt worries only about reactors. Critics of nuclear power catastrophize about its dangers, even though no one has ever died as the result of the operation of a civilian power reactor in the United States. The same critics remain strangely silent about the dangers posed by supposable renewable green energy sources like hydroelectric power dams, despite the deaths of thousands of people, and the danger dams pose to millions of lives.

George Parker - A world-class experimenter

I am sure that my father will be pleased to learn that I have posted his tribute to George Parker on the Internet. I had not realized prior to reading this tribute that George Parker, like my father was a native of the northern border of East Tennessee. Parker was from Johnson City, and my father grew up in Jellico. Parker was a scientist who died with his boots on. He was still making discoveries in 1997, and is listed as a co-author of a number of ORNL publications as late as 1994. George Parker died on September 6, 1997 after a career at ORNL that lasted more than 50 years. He retired in 1996, but continued to work as a consultant, until he was too sick with to work any longer. .It is quite clear that George Parker was an icon at ORNL, and was perhaps the world's greatest authority of fission products release during his lifetime. Thus the articles that George Parker wrote with my father on nuclear safety, carried his great authority along with my father's very considerable authority. (Cross posted on bartoncii.)

By C.J. Barton, Sr.

Many people in this area are undoubtedly grieved, as I am, by the passing of George W. Parker.

One of my regrets in regard to George is that I never got around to writing the George Parker story that I discussed with Barbara Lyon when she was editor of the ORNL Review. The following is a substitute for the story that I envisioned at that time.

George got his start in working with fission products at the University of Chicago. I have heard it said that when criticality was first achieved in a pile there, George was in a nearby laboratory stirring a pot containing some fission products. He soon transferred to Oak Ridge, close to his former home in or near Johnson City, and started the career in what became the Oak Ridge National Laboratory that would span a period of more than half a century.

Lots of fission products soon were produced in the Graphite Reactor and George became involved in work with some highly radioactive materials. George never was one to worry about exposure to radioactivity, and his former co-workers George Creek, Paul Lantz, and Bill Martin have told some pretty hairy tales of their experiences. One health physicist told me recently that when they started getting after George about the high [radiation] exposure levels on by his badge, he began leaving it in his office.

I first met George in the early spring of 1949, about six months after I started work in the Chemical Development Division at Y - 12. He and his associate, Paul Lantz helped me to develop the radioactive tracer techniques that made possible the rapid progress of research on the separation of hafnium from zirconium and the eventual availability of caddying for the nuclear reactor submarine.

This help was typical of George. He never was too busy to assist others who came to him for help. He not only had a store of experience with radioactive materials to draw on, he had over the years stashed away a fabulous collection of equipment that he would share with others.

George was one of the first experimenters to recognize the importance of measuring the release of fission products from overheated reactor fuel and he started such studies in 1955. A fire in a British Magnox reactor that occurred about 1957 demonstrated the importance of this work, and when I joined George's group in 1960, his attention was focused on uranium metal fuel such as that used in Magnox reactors and the Graphite Reactor.

I helped some in that research and I recall carrying hot samples to Hugh Parker's analytical laboratory on the end of a 10-foot-long rod. However, we learned that I could make my most important contribution to George's program by writing,

We co-authored a chapter on fission product release from overheated reactor fuels that was published in Volume 2 of "The Technology of Nuclear Reactor Safety" that was edited and published by MIT. That was considered to be a rather prestigious document at the time it was published in 1973.

George's reputation and wide acquaintance among foreign experimenters enabled us to provide an international flavor to the three symposiums on reactor chemistry that we organized and helped to hold in Gatlinburg in the early '60s. We enjoyed the contacts that these meetings provided as well as the fellowship with a wide variety of experimenters.

After several years in which my principal mission seemed to be to present George Parker's work in a readable form, I moved on to other work. George was a big help to me when I established a glove box labo­ratory in one comer of Building 4501, where his work was located, and which he helped to design.

We had lunch together most days, a custom that started when we worked together. The fellowship that we enjoyed in these lunches was enhanced at times by the presence or experimenters from abroad who came to work with George for varying lengths of time. Their presence demonstrated the extent of George's international reputation.

In summary, George was a uniquely talented person whom I considered to be a world-class experimenter. He had an unusual ability to visualize studies that needed to be done, design equipment needed to do the job and get it constructed, and then to perform the necessary research.

His reputation as an experimenter was well enough known among the people who controlled the purse strings that he was able to continue as an ORNL employee long beyond the normal retirement age.

Finally, I have lost a valued friend and the scientific community has lost an exceptionally gifted experimenter.