Editors Note:This Study by Pitt details the opinions of many experts before the disaster at Fukushima, and shows how narrow-minded the methods of evaluating damage and impact to public health really are.
…In order to understand the meltdown accident, we must go back to its origins. A nuclear power reactor is basically just a water heater, evolving heat from fission processes in the fuel. This heats the water surrounding the fuel, and the hot water is used to produce steam. The steam is then employed as in coal- or oil-fired power plants to drive a turbine which turns a generator (sometimes called a “dynamo”) which produces electric power. There are two different types of reactors in widespread use in the United States, pressurized water reactors PWRs and boiling water reactors BWRs….
…In the PWR, the heated water is pumped out of the reactor to separate units called “steam generators,” where the heat in this water is used to produce steam. In the BWR, the steam is produced directly in the reactor so there is no need for a steam generator…
…The containment provides a broad range of protection for the reactor against external forces, such as a tornado hurling an automobile, a tree, or a house against it, an airplane flying into it, or a large charge of chemical explosive detonated against it. In a meltdown accident, however, the function of the containment is to hold the radioactive material inside…
…Two other mechanisms for breaking open the containment have been discussed. One of these is a steam explosion, which has received considerable research attention7 and was publicized in the fictional movie “The China Syndrome.” The worst situation is in a meltdown where the molten fuel falls into a pool of water at the bottom of the reactor vessel, producing so much steam so suddenly that the top of the reactor vessel would be blown off and hurled upward with so much force that it would break open the top of the containment building. This is a highly unlikely scenario…
…The movie also makes an issue of groundwater contamination following a meltdown accident. In actuality, were molten fuel to suddenly come into contact with groundwater, the latter would flash into steam, which would build up a pressure to keep the rest of the groundwater away. There would thus be little contact until the molten fuel cooled and solidified many days later. It would then be in the form of a glassy mass that would be highly insoluble in water, so there would be relatively little groundwater contamination. If that were judged to be a problem, there would be plenty of time to construct barriers to permanently isolate the radioactivity from groundwater thereafter. It is difficult to imagine a situation in which there would be any adverse health effects from groundwater contamination…
…If we are interested in detectable deaths that can be attributed to an accident, we must limit our consideration to acute radiation sickness, which can be induced by very high radiation doses, about a half million millirems in one day resulting in death within a month. This is a rather rare disease: there were three deaths due to it in the early years among workers in U.S. government nuclear programs, but there have been none for over 25 years now…
Data about the Effects of Melt-down and Cu on RPV integrity
Reactor pressure vessels, which contain the nuclear fuel in nuclear power plants, are made of thick steel plates that are welded together. Neutrons from the fuel in the reactor irradiate the vessel as the reactor is operated. This can embrittle the steel, or make it less tough, and less capable of withstanding flaws which may be present. Embrittlement usually occurs at a vessel’s “beltline,” that section of the vessel wall closest to the reactor fuel.
Pressurized water reactors (PWRs) are more susceptible to embrittlement than are boiling water reactors (BWRs). BWR vessels generally experience less neutron irradiation and therefore less embrittlement.
If cold water is pumped into a BWR vessel, the steam in the vessel will condense and reduce the internal pressure. BWRs may however be susceptible to overpressurization of the reactor pressure vessel at low temperatures under certain conditions.
NUREG‑1511, “Reactor Pressure Vessel Status Report” discusses the issue of vessel structural integrity and provides a plant‑by‑plant status. Updates were published in October 1996 (Supplement 1) and in October 2000 (Supplement 2).
Control rod drive mechanism nozzles and other vessel head penetration nozzles welded to the upper reactor vessel head are subject to another phenomenon – primary water stress corrosion cracking. The issue is a potential safety concern because a nozzle with sufficient cracking could break off during operation. This would compromise the integrity of the reactor coolant system pressure boundary – one of three primary barriers that protect the public from exposure to radiation.
Additional information on this experience may be accessed on NRC’s web site at: http://www.nrc.gov/reactors/operating/ops-experience/bottom-head-penetration-leakage.html.
Concept of Service Life Evaluation in Japan
• In Japan, there is no license period determined for nuclear plant. Based on the periodical inspection conducted every year, it can be operated another one year.
• As for plant life management for aged nuclear reactors, the Ministry of International Trade and Industry (MITI) showed the idea by the document ” Basic Idea on Plant Life Management (PLM)” in 1996, which indicated that 60 years operation could be possible by taking possible countermeasures. Various kinds of aging researches have been conducted since then.The review from the point of PLM was performed by utilities as self-imposed tests on nuclear plants before operation period of a plant became 30 years.
• In 1999, PLM was incorporated to Periodical Safety Review (PSR). PLM reviews have been done for 9 plants as a part of PSR up to now.
• In 2002, PSR has been placed in the safety regulations. Government research organizations, universities as well as utilities and industries are performing related aging researches. Nuclear and Industrial Safety Agency (NISA) in the Ministry of Economy, Trade and Industry (METI) is taking a major role for conducting PLM related research projects.
• So, service life can be evaluated through PSR by the most up-to-dated knowledge on material aging, integrity evaluation methodologies, reliability of non-destructive tests.
• JAERI has been contributing to the material aging study. In the future, probabilistic approach is thought to be very effective PLM evaluation method, and hence PFM methodology is important as a key technology for providing an indication of reliability.
TEPCO’s Report on the findings of Xe-135
On November 1, as a sampling result by the gas control system that we have newly installed for Unit 2, some Xe135, a nuclear fission product, was detected. Considering its short half-life (around 9 hours), it was produced by a recent nuclear fission, not produced before the March 11. Even though a nuclear reactor is not in a critical state, a very small amount of nuclear fission always happens. The Xe-135 was detected as a result of the measuring in a high temperature.
Xe135 density comparison of spontaneous fission and actual measurement
1. Presumed radioactive density of the Xe135 by spontaneous fission
Even in a non-critical (shut down) state, a nuclear reactor usually has spontaneous fissile nuclides in it. Well-known are Cm-242 and Cm-244, and currently in Nuclear Reactor of Unit 2, the fission is as follows:
Cm242: 8.3E8 times/sec
Cm244: 7.4E8 times/sec (Appendix 1-1)
We don’t take into consideration the nuclear fission caused by the U-235 absorption of neutron. This is an evaluation conservative enough to show that the Xe135 detected this time was produced in a non-critical state.
Although Xe135 is also produced by Xe134 (n, γ) reaction, we don’t evaluate it due to the presumably small contribution. (Appendix 1-2)
Yield of Xe135 produced by Cm-242: 2.66% Generation speed: 2.2E7 p/sec
Yield of Xe135 produced by Cm-244: 1.22% Generation speed 9.0E6 p/sec
That follows in total: 3.1E7 p/sec
(Source：Fission product yields, http://www-nds.iaea.org/wimsd/fpyield.htm#T5）
On August 3, 2001, the NRC issued Bulletin 2001-01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,” to licensed holders of U.S. pressured water reactors (PWR) following the discovery of cracked and leaking nozzles in 2000 and 2001. In the bulletin, the staff requested information from PWR licensees about the structural integrity of these nozzles at their facilities. In response to the bulletin, licensees provided their plans for inspecting their nozzles and the outside surfaces of their upper reactor vessel heads to determine whether any nozzles were leaking. Inspections by licensees during the fall of 2001 revealed vessel head penetration nozzle cracks at Three Mile Island Unit 1, Crystal River Unit 3, North Anna Unit 1, and Oconee Unit 3.
On August 9, 2002, the NRC issued Bulletin 2002-02, “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs.” The bulletin suggested that visual inspections of upper reactor vessel heads and their nozzles may need to be supplemented with non-visual non-destructive examinations to assure that the structural integrity and leakage integrity of the nozzles is maintained. Bulletin 2002-02 requested that PWR licensees provide information about their inspection programs and plans to supplement existing visual inspections with volumetric and surface examinations. Licensees responded with descriptions of inspection plans for at least the first refueling outage following the issuance of the bulletin. Many did not offer any long-term inspection plans, but instead opted to follow guidance being developed by the Materials Reliability Program, which is an industry-sponsored research organization.
Inspections performed at several PWRs in 2002 including those performed at the Davis-Besse Nuclear Plant, found leakage and cracks in vessel head penetration nozzles or J-groove welds that have required repairs or prompted the replacement of the vessel head. As a result of continuing concerns regarding licensee inspection programs in this area, the NRC issued an Order on February 11, 2003, to all PWR licensees in the U.S. The Order requires specific inspections of the vessel head and associated penetration nozzles based on their susceptibility to primary water stress corrosion cracking. The Order may be accessed at the following address on NRC’s website: http://www.nrc.gov/reactors/operating/ops-experience/vessel-head-degradation/vessel-head-degradation-files/order-rpv-inspections.pdf .
Twenty-six units were identified by the Electric Power Research Institute’s Materials Reliability Program as having a high susceptibility to nozzle cracking. Inspections by licensees performed after issuance of the latest bulletin and order, revealed nozzle or J-groove weld cracks and/or leaks at Oconee Unit 2, North Anna 2, Arkansas Nuclear One Unit 1, St. Lucie Unit 2, Milestone Unit 2, and Beaver Valley Unit 1. The utilities owning the Oconee, Surry, Davis-Besse and North Anna nuclear stations have replaced or are in the process of replacing their upper reactor vessel heads. Approximately twenty other units have announced plans to have their upper reactor vessel heads replaced within the next few years.
- Masashi Goto – Nuclear Reactor Designer teams up with ex-Hitachi Engineer Tanaka – PCV/RPV pipes broke after EQ not IC shutdown (enformable.com)
- Japanese Researcher Says Reactor 3′s “Re-Melting” Released Large Amount Of Radiation From MOX Fuel (enformable.com)
- #Fukushima I Nuke Plant Reactor 2 Achieves “Cold Shutdown” (ex-skf.blogspot.com)
- #Fukushima I Nuke Plant: Water Entombment Is Back on the Table (ex-skf.blogspot.com)
- TEPCO Dumps 565-Page Report on Early Days of Crisis, Says No Re-Melting of Reactor 3 Fuel (ex-skf.blogspot.com)