BACKGROUND

Of the 104 operational nuclear power reactors in the United States, thirty-five are boiling water reactors (BWR). General Electric is the sole designer and manufacturer of BWRs in the United States. The BWR's distinguishing feature is that the reactor vessel serves as the boiler for the nuclear steam supply system. The steam is generated in the reactor vessel by the controlled fissioning of enriched uranium fuel which passes directly to the turbogenerator to generate electricity.

LACK OF CONTAINMENT INTEGRITY DURING A NUCLEAR ACCIDENT

The purpose of a reactor containment system is to create a barrier against the release of radioactivity generated during nuclear power operations from certain "design basis" accidents, such as increased pressure from a single pipe break. It is important to understand that nuclear power plants are not required by the Nuclear Regulatory Commission (NRC) to remain intact as a barrier to all possible accidents or "non-design basis" accidents, such as the melting of reactor fuel. All nuclear reactors can have accidents which can exceed the design basis of their containment.

But even basic questions about the the GE containment design remain unanswered and its integrity in serious doubt. For example, 23 of these BWRs use a smaller GE Mark I pressure suppression containment conceived as a cost-saving alternative to the larger reinforced concrete containments marketed by competitors. A large inverted light-bulb-shaped steel structure called "the drywell" is constructed of a steel liner and a concrete drywell shield wall enclosing the reactor vessel--this is considered the "primary" containment.. The atmosphere of the drywell is connected through large diameter pipes to a large hollow doughnut-shaped pressure suppression pool called "the torus", or wetwell, which is half-filled with water. In the event of a loss-of-coolant-accident (LOCA), steam would be released into the drywell and directed underwater in the torus where it is supposed to condense, thus suppressing a pressure buildup in the containment.

The outer concrete building is the "secondary" containment and is smaller and less robust (and thus cheaper to build) than the containment buildings used at most reactors.

As early as 1972, Dr. Stephen Hanauer, an Atomic Energy Commission safety official, recommended that the pressure suppression system be discontinued and any further designs not be accepted for construction permits. Hanauer's boss, Joseph Hendrie (later an NRC Commissioner) essentially agreed with Hanauer, but denied the recommendation on the grounds that it could end the nuclear power industry in the U.S.

Here are copies of the three original AEC memos, including Hendrie's:

November 11, 1971: outlines problems with the design and pressure suppression system containment.

September 20, 1972: memo from Steven Hanauer recommends that U.S. stop licensing reactors using pressure suppression system

September 25, 1972: memo from Joseph Hendrie (top safety official at AEC) agrees with recommendation but rejects it saying it "could well mean the end of nuclear power..."

In 1976, three General Electric nuclear engineers publicly resigned their prestigious positions citing dangerous shortcomings in the GE design.

An NRC analysis of the potential failure of the Mark I under accident conditions concluded in a 1985 report that Mark I failure within the first few hours following core melt would appear rather likely."

In 1986, Harold Denton, then the NRC's top safety official, told an industry trade group that the "Mark I containment, especially being smaller with lower design pressure, in spite of the suppression pool, if you look at the WASH 1400 safety study, you'll find something like a 90% probability of that containment failing." In order to protect the Mark I containment from a total rupture it was determined necessary to vent any high pressure buildup. As a result, an industry workgroup designed and installed the "direct torus vent system" at all Mark I reactors. Operated from the control room, the vent is a reinforced pipe installed in the torus and designed to release radioactive high pressure steam generated in a severe accident by allowing the unfiltered release directly to the atmosphere through the 300 foot vent stack. Reactor operators now have the option by direct action to expose the public and the environment to unknown amounts of harmful radiation in order to "save containment." As a result of GE's design deficiency, the original idea for a passive containment system has been dangerously compromised and given over to human control with all its associated risks of error and technical failure.

As we have now seen at Fukushima, Japan, in March 2011, this containment design failed catastrophically when hydrogen built up in the outer containment buildings until three of them exploded. The outer containment building was neither large enough nor strong enough to withstand these explosions.

VULNERABILITY OF IRRADIATED FUEL POOLS

The irradiated (sometimes called "spent") fuel pools in GE Mark I reactors are above the reactor core and outside the primary containment system. This design was chosen for efficiency, not safety--the fuel rods in the reactor are lifted by crane and simply moved over to the fuel pool. The explosions at Fukushima that caused severe damage to the containment buildings (as can be seen in the above satellite photo taken March 18, 2011) also exposed and compromised the fuel pools providing a direct pathway for release of radioactivity into the air. While there was substantial amounts of fuel in the Fukushima pools, in the U.S. pools are typically packed even more densely, meaning even higher potential radiation risks if they are compromised.

DETERIORATION OF BWR SYSTEMS AND COMPONENTS

It is becoming increasingly clear that the aging of reactor components poses serious economic and safety risks at BWRs. A report by NRC published in 1993 confirmed that age-related degradation in BWRs will damage or destroy many vital safety-related components inside the reactor vessel before the forty year license expires. The NRC report states "Failure of internals could create conditions that may challenge the integrity the reactor primary containment systems." The study looked at major components in the reactor vessel and found that safety-related parts were vulnerable to failure as the result of the deterioration of susceptible materials (Type 304 stainless steel ) due to chronic radiation exposure, heat, fatigue, and corrosive chemistry. One such safety-related component is the core shroud and it is also an indicator of cracking in other vital components through the reactor made of the same material.

Core Shroud Cracking

The core shroud is a large stainless steel cylinder of circumferentially welded plates surrounding the reactor fuel core. The shroud provides for the core geometry of the fuel bundles. It is integral to providing a refloodable compartment in the event of a loss-of-coolant-accident. Extensive cracking of circumferential welds on the core shroud has been discovered in a growing number of U.S. and foreign BWRs. A lateral shift along circumferential cracks at the welds by as little as 1/8 inch can result in the misalignment of the fuel and the inability to insert the control rods coupled with loss of fuel core cooling capability. This scenario can result in a core melt accident. A German utility operating a GE BWR where extensive core shroud cracking was identified estimated the cost of replacement at $65 million dollars. The Wuergassen reactor, Germany's oldest boiling water reactor, was closed in 1995 after wary German nuclear regulators rejected a plan to repair rather than replace the reactor's cracked core shroud.

Rather than address the central issue of age related deterioration, U.S. BWR operators now opt for a dangerous piecemeal approach of patching cracking parts at least cost but increased risk.

Paul Gunter, NIRS, March, 1996, updated by Michael Mariotte, NIRS, March 2011