A. United States Nuclear Power Plants

1. Maine Yankee
2. Connecticut Yankee
3. Three Mile Island
4. Other Citations about U.S. Nuclear Power Plants
5. Safety Issues at U.S. Nuclear Power Plants

a. Reactor Embrittlement
b. Spent Fuel Cladding Failure
c. Steam Generator Degradation Mechanisms
d. LORCAs and Spent Fuel Cooling
e. Hot Particles
f. Spent Fuel Storage and Disposal (Dry Casks/Multi Purpose Casks, etc.)
g. MOX
6.  Nuclear Regulatory Commission Publications (if not listed in any of the above categories)
• There are thousands of documents and research papers characterizing the liquid, gaseous and solid waste effluents and emissions from nuclear power plants both in the United States and Europe. Only a few of these research papers are cited in this Website.
• Every atomic power generating station is the potential source point of huge releases of anthropogenic radionuclides to the environment of the same order of magnitude as the Chernobyl accident on April 26, 1986.

Maps of locations of Nuclear Power Reactors: NORTH AMERICA
 

1. Maine Yankee

Churchill, J.H., Hess, C.T. and Smith, C.W. (1980). Measurement and computer modeling of radionuclide uptake by marine sediments near a nuclear power reactor. Health Physics. 38. pg. 327-340.

• Isocuric mapping showed ¹³⁷Cs concentration in sediment near the Maine Yankee Atomic Power Company plant outflow (prior to its removal to the bottom of Montsweag Bay) to 25,000 pCi/kg in sediment in June, 1975.
• After removal of liquid effluent diffuser from Bailey's Cove, peak concentrations of ¹³⁷Cs in sediment dropped to 5000 pCi/kg.

• In a split survey of 131 samples of sediment and other media (seaweed, shellfish, etc.) where both the state and Maine Yankee report contamination in split samples, the Maine Yankee Atomic Power Company radiological surveillance reported 16 examples of anthropogenic radioactivity between 1989 and 1995, all of ¹³⁷Cs in sediments at one location, Foxbird Island. The peak concentration of ¹³⁷Cs in sediment was reported as 495 pCi/kg.
• The state of Maine, Division of Health Engineering reported slight levels of contamination in 41 samples, noting the presence of ^110mAg, ⁶⁰Co, one sample with ¹³¹I, and ¹³⁷Cs, which was the predominant anthropogenic nuclide. The peak concentrations of ¹³⁷Cs in sediment was noted as 540 pCi/kg.
• This survey, along with the NRC sponsored annual radiological reports issued by Maine Yankee and other nuclear power generators, continue to document the pristine and nearly uncontaminated environments surrounding Maine Yankee Atomic Power Company and other United States reactors, which seemed to have escaped most, if not all, of weapons testing derived stratospheric fallout, as well as all Chernobyl derived contamination.

• Total ¹³⁷Cs release at 8 New England nuclear power plants: 1974: 86.73 Ci; 1979: 0.35 Ci; 1984: 2.70 Ci.
• Total ³H release in 1977: 7,426.3 Ci.
• Total X-135 release in 1975: 830,093 Ci.

• Pre-operational surveys of field soil and sediment samples in the Maine Yankee Atomic Power Company vicinity revealed significantly higher levels of ¹³⁷Cs in many samples than were found in many post-operational field soil and sediment samples.
• Post-operational surveys of Bailey's Cove did record a significant impact from Maine Yankee Atomic Power Company derived activation products (⁵⁸Co, ⁶⁰Co), with peak concentrations of ⁵⁸Co up to 5,620 pCi/kg near the plant outfall.
• One hot particle was noted containing 7,700 pCi of ⁶⁰Co, and had a total activity of 9,000 pCi in a mass less than 20 µg. (pg. 18)
• Most of the extensive pre-operational nuclear weapons testing derived radiocesium as well as post-operational reactor derived radiocesium documented in this report have miraculously disappeared in later Maine Yankee Atomic Power Company environmental radiological summaries.

• "Among the problems associated with cultivation of bivalves in heated discharge water is the accumulation within the soft tissues of these filter feeding mollusks of vast quantities of pollutants (viruses, bacteria, heavy metals, pesticides, radionuclides, etc.). Concentrations of such pollutants can reach levels several orders of magnitude above those encountered in the surrounding water." (pg. 167).
• "...elevated temperatures encountered at varying distances from the discharge waters of the studied nuclear generating facility had an adverse effect on the growth, survival and recruitment of the experimentally-cultured mussels." (pg. 183).
• "...trace amounts of ⁵⁸Co, ⁶⁰Co, ¹³⁴Cs, ¹³⁷Cs, ⁵⁴Mn, ⁹⁵Zr, ⁹⁵Nb and ⁴⁰K were detected in both the shells and soft tissues of the mussels cultured in these waters." (pg. 187).

• Another anomaly in the Maine Yankee Atomic Power Company soap opera; for detailed comments on this citation, see RAD 12: Maine Yankee Atomic Power Company: radioactive waste inventories.

• Typical gross beta air concentrations in January through December of 1991: +/- 1,000 µBq/m³.
• Non-routine measurement in sediment and biotic media noted as ten times the background.
• A positive measurement is defined as three times greater than the standard deviation.
• Algae show trace amounts of ⁶⁰Co and ^110mAg on one occasion in 1990 out of a total of three samples for the year.
• Six sediment samples were taken at each of two locations during 1990; all showed slightly elevated levels of ¹³⁷Cs (to 220 pCi/kg dry weight near the old plant outfall).
• Sampling of bottom sediments near the plant diffuser discharge about 100 ft below the surface of Montsweag Bay is habitually avoided in all reports.
• No gamma-emitting radionuclides were detected in four fish samples from two locations in 1990.
• No environmental samples are reported to have been tested by the licensee for any alpha-emitting anthropogenic radionuclides in this or any other annual environmental monitoring report.
• The annual radiological environmental monitoring reports issued since the plant diffuser was moved to the bottom of Montsweag Bay indicate that the Maine Yankee area is nearly a pristine environment without deposition from nuclear weapons testing or Chernobyl fallout, and with almost no impact from plant discharges.
• The laboratory testing for the environmental samples provided by Maine Yankee for these annual reports was done by the Yankee Atomic Environmental Laboratory, a subsidiary of the Yankee Atomic Electric Company, the same company now embroiled in a controversy pertaining to falsified computer programs and emergency core cooling system capabilities at the Maine Yankee Atomic Power Company.
• For more information on the allegations pertaining to Maine Yankee Atomic Power Company and the Yankee Atomic Electric Company click on any of the following:
• Whistleblower's letter
• Brief to Jay McCloskey
• Letter to the FBI
• Summary Notice for a Petition for Indictment
• For another discrepancy in Maine Yankee Atomic Power Company radiological surveillance reporting, see McCarthy, W.J., et al. (1978) in RAD 12: Maine Yankee Atomic Power Company.
• This report illustrates why the NRC and its licensees should be the subject of an investigation by an independent council appointed by the attorney general for the purposes of examining the many irregularities in the environmental radiological survey programs sponsored by the NRC. It is extremely unlikely that nuclear power facilities under the supervision of the Nuclear Regulatory Commission are the sole locations in the northern hemisphere that have never been impacted by stratospheric fallout from the nuclear weapons tests of the 1950's and 1960's.

United States Nuclear Regulatory Commission. (March 1998). Haddam Neck Historical Review Team Report. US NRC, Washington, DC. http://www.nrc.gov/OPA/reports/hnhistm.htm.

• A comprehensive review of one of the largest series of nuclear accidents at any operational US reactor.

A chilling testimony to the capability of the NRC to substitute phony "scoping surveys" for accurate documentation of radiological effluents using media-specific, nuclide-specific spectroanalyses.

The Three Mile Island accident is a model of the misinformation pertaining to NRC operated nuclear facilities, and provides a preview of the deceptions that can be expected in the documentation of future nuclear accidents in the U.S.A.

Immediately after the Three Mile Island accident, supposedly knowledgeable officials released a statement, prior to any understanding of the release dynamics of the accident within the Three Mile Island reactor core, that the only radioisotope released, other than inert gases was 15 Ci of ¹³¹I. After a year or more of intense study, it was discovered that most of the fuel had melted into the lower reactor vessel core support area. Conditions which allow such melting would necessarily lead to a substantial vaporization and release of volatile radioisotopes such as cesium-137. In view of the liquefaction of the reactor fuel during the TMI accident, it is extremely unlikely that the source term release for TMI was limited to only 15 Ci of ¹³¹I.

• RADNET's inventory of citations in the Three Mile Island accident files will be posted here later.
• RADNET browsers are solicited for comments, information or additional citations on the Three Mile Island accident; any response would be greatly appreciated.

Report of the President's Commission on the Accident at Three Mile Island.  An on-line version of this report posted by a concerned citizen.
 

4. Other Citations about U.S. Nuclear Power Plants

Bedford, Henry F. (1990). Seabrook Station: Citizen politics and nuclear power. The University of Massachusetts Press, Amherst, MA. IS.

Biewald, Bruce and White, David. (January 15, 1999). Stranded nuclear waste: Implications of electric industry deregulation for nuclear plant retirements and funding decommissioning and spent fuel. Synapse Energy Economics, Inc., Cambridge, MA. http://www.citact.org/nucrep.html.

Linsalata, P., Wrenn, M.E., Cohen, N. and Singh, N.P. (1980). ^239,240Pu and ²³⁸Pu in sediments of the Hudson River estuary. Environmental Science and Technology. 14(2). pg. 1519-1523.
 

1976

Indian Point, NY

River sediments

239,240Pu

236 pCi/kg dry sediment

• Lower deposition levels away from plant, typically 14 - 64 pCi/kg.

• An excellent general history of nuclear power development, this text discusses the evasions and deceptions which characterize the early years of the industry and which continue to form the basis of contemporary rationalizations of the nuclear energy fiasco.
• The second section of this publication critiques the second generation of "inherently safe" reactors, discusses plutonium disposal options, energy policies and provides a general summary of past nuclear accidents, including Chernobyl.
• Useful appendices follow the text.
• The summary and recommendations about nuclear energy phase-out and carbon dioxide emissions reductions preface the text. The authors suggest vitrification of plutonium and suggest radical changes in the DOE current high-level waste management program.

Nuclear Waste News. (November 5, 1998). Decommissioning:  NRC oks Trojan reactor shipment; state, DOT approval pending. IAC-ACC-NO:  53200310 ND. Nuclear Waste News. 45(18).

• Docket No. 71-9271.

• A U.S. government information source on reactor waste inventories -- indispensable!
• This is the latest in the series of reports from the Oak Ridge National Laboratory, containing inventories of long-lived radionuclides which have accumulated due to the operation of commercial nuclear power plants in the United States (only).
• This report provides several basic definitions (pg. 3):

Spent Nuclear Fuel (SNF): Irradiated fuel discharged from a nuclear reactor.

High-Level Waste (HLW): Highly radioactive material resulting from the reprocessing of spent nuclear fuel.

Transuranic Wastes (TRUW): Radioactive wastes from fuel reprocessing and the fabrication of nuclear weapons, containing more than 100 nCi/g of alpha emitting isotopes with atomic numbers greater than 92 and having half-lives greater than 20 years.

• RADNET uses HLW (high-level waste) to mean, in terms of missing military high-level wastes, the total of accumulated SNF (spent nuclear fuel), HLW, and TRU (transuranic wastes) resulting from the weapons production process.
• Commercial HLW includes only SNF, as most commercial SNF has not been reprocessed, except at West Valley, New York (This exception illustrates the failed attempt to reprocess SNF from commercial reactors in this country.)
• Major commercial radioactive waste disposal sites are listed as located at: Barnwell, SC; Betty, NV; Salt Lake City, UT; Frankfurt, KY; Richland, WA; Sheffield, IL; and West Valley, NY.

Domestic Commercial Light Water Reactor Spent Nuclear Fuel Inventories (pg. 32-34):
 
• Boiling water reactors: Jan. 1, 1996: cumulative inventory of long-lived radionuclides: 8,100 x 10⁶ Ci (8,100,000,000 Ci).
• Pressurized water reactors: Jan. 1, 1996: cumulative inventory of long-lived radionuclides: 22,100 x 10⁶ Ci (22,100,000,000 Ci).
• Total commercial light water reactor spent nuclear fuel inventories: Jan. 1, 1996: cumulative inventory of long-lived radionuclides: 30,200 x 10⁶ Ci (30,200,000,000 Ci).
• Total commercial light water reactor spent nuclear fuel inventories: 2008: 37,800 x 10⁶ Ci (37,800,000,000 Ci), estimated.
• Permanently discharged spent nuclear fuel assemblies: Jan. 1, 1996: 103,944.
• Permanently discharged spent nuclear fuel assemblies: 2008: 200,000, estimated.
• Total commercial low-level waste generated as of Jan. 1, 1995: 5,995 x 10³ Ci (5,995,000 Ci).
• To obtain an approximate inventory of spent nuclear fuel, etc. at your local atomic power generating station divide the above data by 109 (There are 109 operating nuclear power plants in the U.S. as of Jan. 1, 1996).
• A model U.S. light water reactor has now accumulated 276,200,000 curies of spent fuel high-level waste as of Jan. 1, 1996.
• This report fails to include an inventory of greater than class C wastes (GTCC wastes), known as orphan wastes in the context of commercial spent fuel and low-level waste disposal, due to a lack of a designated facility for this type of waste. These highly radioactive reactor vessel components produced during the weapons production process have apparently been classified as low-level wastes and discarded in uncontained situations along with the missing high-level waste discussed elsewhere in RADNET. The missing liquid high-level wastes were also probably reclassified as low-level wastes prior to uncontained release.

Riccio, Jim and Brooks, Lisa. (1996). Nuclear lemons an assessment of America's worst commercial nuclear power plants. Fifth edition. Public Citizen: Critical Mass Energy Project.

• "Nuclear Lemons identifies the nation's most dangerous commercial nuclear reactors by rating them in each of a dozen safety, performance and economic areas .... by incorporating data on nuclear reactor safety-related systems collected by NRC." (pg. 11).
• The following criteria are used in Nuclear Lemons to rank each nuclear power reactor in the United States: capacity factor; enforcement discretion; forced outage rate; licensee event reports; operations and maintenance costs; safety system actuations; safety system failures; scrams; significant events; systematic assessment of licensee performance; violations; and worker exposure to radiation. (pg. 11).
• This is the fifth in a series of reports by the Critical Mass Energy Project on US nuclear power plants and is available in hard copy in two volumes; the second volume contains supplemental reactor-specific data upon which the reactor rankings are based. To contact Public Citizen: Critical Mass Energy Project, go to RADNET Section 13: RADLINKS: Part 2 and click on this link.

Stellfox, David. (May 20, 1999). First-cycle fuel at River Bend affected by mystery corrosion. Nucleonics Week. 40(20). pg. 2.

• "The thickest deposition of crud and all the fuel cladding failures at Entergy's River Bend occurred in first-cycle fuel, NRC said, adding the reason is 'not fully understood.'"
• "River Bend licensee Entergy Operations Inc., NRC, and fuel manufacturer General Electric are analyzing what caused the heavy crud buildup on the fuel during the unit's last operating cycle."
• "The crud depositions 'were different from those previously seen at River Bend or other GE facilities, in that the coating was less adherent and of much lower density, hence the greater thickness, and more porous,' NRC said in a May 14 report.  'At the most affected locations, the crud thickness on adjacent fuel rods was such that the open flow channel between the rods was significantly reduced.'"
• "Entergy spokeswoman Diane Park said 'actual leaks' were limited to seven fuel bundles Entergy had tentatively identified before entering the April 3 outage."
• "NRC said those failures were caused by 'deposition of an unusually thick layer of crud on the fuel in areas of particularly high heat flux.'"
• "Park said it was the amount of corrosion found on the other bundles, the non-leakers, that prompted the conservative decision to acquire new fuel -- some 112 new assemblies, according the NRC -- before restarting the reactor."

Weil, Jenny and Stellfox, David. (January 4, 1999). AEOD abolished, research office expanded under reorganization plan. Inside N.R.C. 21(1). pg. 1.

• "As expected, the biggest shakeup was to the Office for Analysis and Evaluation of Operational Data (AEOD) (INRC, 14 Sept. '98, 14). Created in 1979 after the partial meltdown at Three Mile Island-2 to independently assess operational events and provide feedback to NRC staff and licensees, AEOD was abolished last month and its functions redistributed to other existing offices.  Because AEOD was not statutorily established, no legislative action is required to break up the office."
• "Most of AEOD's functions will be moved to RES, but some responsibilities will be transferred to the Office of Human Resources, NRR, NMSS and the Executive Director for Operations (EDO)."
• "...today the kinds of problems being worked have more to do with aging of components and operational issues and do not require the same degree of large scale experiments."
• "Some former AEOD functions that will be moved to RES include the independent analysis and evaluation of plant performance data; event assessment activities; performance indicators program; the accident precursor program; and reliability, initiating events and common cause failure studies.  NMSS will pick up only one AEOD function -- responsibility of the nuclear materials event data base."

Yankee Nuclear Power Station. (1993). Supplement to Applicant's environmental report post operating license stage: Decommissioning environmental report. Yankee Atomic Power Company, Rowe, MA.

• These two reports provide a model for a licensee/NRC approach to decommissioning a nuclear power plant which was closed several years ago due to embrittlement of the reactor vessel.
• On site environmental surveillance notes soil core segments from seven locations surveyed in 1987 with concentrations to 1,150 pCi/kg (wet). Low-level waste inventory is listed as 5,172 curies (Table 6.2-1, in Decommissioning Environmental Report).
• While neither the Decommissioning Plan nor the Environmental Report notes GTCCW or spent fuel inventories, the Oak Ridge National Laboratory Integrated Data Base lists the following additional inventories of radioactive waste:

Reactor vessel internal components: 87 m³: 132,600 Ci

Reactor core baffle: 2.1 m³: 1,020,000 Ci

Reactor vessel: 187.9 m³: 4,700 Ci

Total decommissioning wastes 1993-1999: 1,159,536 Ci

• These decommissioning waste inventories do not include spent fuel, which will remain on site indefinitely as the Yankee Rowe facility follows the NRC "SAFESTOR" contingency plan for decommissioning: i.e. remove the low-level wastes and leave the spent fuel for the grandchildren. The spent fuel waste inventory for Yankee Rowe is not presently available, but since this facility is smaller than Maine Yankee (spent fuel inventory as of 1996: +200,000,000 Ci), the accumulated on site spent fuel inventory would be considerably less than this figure.
• Subtracting the 5,172 Ci of LLW from the Oak Ridge National Laboratory data leaves an inventory of 1,154,364 Ci greater than class C (GTCC) wastes as the decommissioning inventory of "orphan wastes" (not spent fuel, nor low-level wastes) . After the Yankee Nuclear Power Station was closed, the majority of GTCC reactor vessel components were removed from the reactor vessel itself and are now stored in the spent fuel pool.
• The 1982 nuclear waste policy act prohibits disposal of GTCC wastes as low-level wastes, however it should be noted for the information of any persons concerned with the decommissioning of Yankee Rowe, that as of the late winter of 1996, the current arrangement for disposal of the remaining sections of the reactor vessel, core baffle and reactor vessel components involve shipping the reactor vessel in its entirety by train for disposal in Barnwell, S.C. as low-level waste. These remaining reactor vessel components now contain between 5,000 and 6,000 curies of radioactivity; the rational for the disposal of these highly radioactive components in a landfill is that the greater than class C wastes will become class C low-level wastes upon combining the greater than class C wastes with a sufficient volume of low-level waste to reduce the overall curic average to just below the GTCC cut off point. In the case of Yankee Atomic, this cutoff point with the reactor vessel has been reached by filling the reactor vessel with cement. Litigation and negotiations are still continuing as of Jan. 1, 1997 as to when the Yankee Atomic reactor vessel will be shipped to S. Carolina for land burial. The Yankee Atomic reactor vessel is a typical example of what is called HOT C; the question for future decommissioning scenarios is how long the HOT C scam will be available to be used for siting GTCC wastes as low-level wastes at other low-level waste repositories.
• This will be the first commercial reactor vessel ever disposed of in its entirety in a landfill situation as well as the first massive highly radioactive unshielded unit ever transported by train through highly populated areas.
• The precedent for this policy of disposing of a highly radioactive reactor vessel as low-level waste was first inadvertently indicated when the 1987 TLG decommissioning report for the Maine Yankee Atomic Power Company reactor vessel inventory was reprinted as an appendix in a publicly available state of Maine report (See Vanags, 1992, A study of radioactive wastes). GTCC reactor vessel components containing in excess of 4,000,000 Ci were listed as destined for a landfill disposal, also in Barnwell SC, in slightly over 101 shipments as part of the proposed decommissioning (DECON) of the Maine Yankee facility. These GTCC components, containing millions of curies of radioactivity, are only 239 ft³ in volume, but were to be transported in one hundred and one shipments mixed with low-level waste to meet DOT, NRC, and other regulatory guidelines for the transshipment and disposal of "low-level waste" (See Maine Yankee Atomic Power Company Radioactive Waste Inventories, GTCC reactor vessel inventory, and/or "Question of the Day" listed in the notes.)
• RADNET readers interested in the topic of the decommissioning of the Yankee Atomic Power Company in Rowe, Massachusetts should note that the Maine Yankee Atomic Power Company is located in Wiscasset, Maine, and is still operational as of March 31, 1996. The old paradigm (1987) for disposing of Maine Yankee's reactor vessel is currently the only way for owners of Yankee Rowe to cheaply dispose of the Yankee Rowe reactor vessel in 1996 or 1997.
• RADNET readers who have additional information about public safety precautions which are being planned or will be implemented during the rail journey of the Yankee Rowe reactor vessel (+5,000 curies) on its way to a low-level waste disposal pit in South Carolina are requested to contact the Center for Biological Monitoring .
• The Yankee Rowe site, as is the case with most other mothballed U.S. commercial reactors, will remain a de facto high-level waste storage site indefinitely. The transfer of the highly radioactive GTCC reactor vessel components to the Yankee Rowe spent fuel pool for temporary storage complicate spent fuel pool decommissioning scenarios; Maine Yankee Atomic Power Company and other reactors with fully loaded spent fuel pools won't have the luxury of discarding their highly radioactive reactor vessel components in this location at the time of decommissioning or mothballing.

Government Accounting Office. (March 19, 1999). Nuclear regulation: Strategy needed to regulate safety using information on risk. GAO/RCED-99-95. GAO, Washington, DC.

• This report "requested by Sens. Joe Biden, D-Del., and Joe Lieberman, D-Conn., asserts that the NRC 'has not developed a comprehensive strategy that could move its regulation of the safety of nuclear plants to an approach that considers risk information.'" (The Energy Report, 27(17), April 26, 1999).
• "The GAO report finds that:  some utilities do not have current and accurate design information for their nuclear power plants needed for the risk-informed approach, and, neither the NRC nor the industry has standards that define the quality or adequacy of the risk assessments that utilities use to identify and measure risks to public health and the environment." (The Energy Report, 27(17), April 26, 1999).
• "In sum, the report concludes that there are a myriad of problems associated with the NRC's move to the new risk-based approach and that without the proper steps taken, the program will fail, jeopardizing oversight of the safety of the nation's nuclear plants." (The Energy Report, 27(17), April 26, 1999).

• "UCS [Union of Concerned Scientists] undertook a study to assess how the nuclear power industry is handling the pressures of aging equipment and shrinking budgets. For our focus group, we selected 10 plants that represent a cross section of the nuclear industry." (Executive Summary, p. v).
• "...plants' [NRC] internal auditors, a key element in the quality assurance programs that federal law requires, found none of the more than 200 problems reported last year." (Executive Summary, p. v).
• "A second significant finding was that far too many of the problems reported at the monitored plants resulted from workers' mistakes (35 percent of reported problems) or poor procedures (44 percent)." (Executive Summary, p. v).
• "At the LaSalle, Millstone, and Sequoyah plants, problems often remained undetected or uncorrected over a long period of time." (Executive Summary, p. v).
• This report is available on-line at http://www.ucsusa.org/publications/index.html.

• "As nuclear plants age, more fires associated with electrical circuits are showing up in a fire incident database ... 'We're starting to see more electrical issues as plants get older,' said Wayne Sohlman of Nuclear Electric Insurance Ltd. (NEIL)."
• "A NEIL-sponsored fire incident database, while limited at present, indicates that electrical wiring was the single largest category of material that ignited at nuclear plants.  Electrical cabling was also the first category for 'fire origin' in the database, and the first or largest category for fire ignition source was electrical malfunction."
• "One potential surprise, however, is that a significant number of fires -- 41, or the second largest category in the database -- are first reported by continuous fire watches, including hot-work watches, rather than automatic detection systems."

Curran, D. (August 1, 1991). Testimony of Diane Curran: Subject: Embrittlement of the reactor vessel at the Yankee Rowe nuclear power plant. Before the Subcommittee on Energy and the Environment, House Committee on Interior and Insular Affairs, U.S. House of Representatives, Washington, DC.

• This testimony refers to a petition cited below under Pollard, R. and Curran, D.
• "Yesterday, the Nuclear Regulatory Commission voted to deny a petition for emergency enforcement action and request for an adjudicatory hearing, filed by UCS [Union of Concerned Scientists] and NECNP [New England Coalition on Nuclear Pollution] on June 4, 1991, which charges that the Yankee Rowe nuclear power plant violates NRC regulatory standards for pressure vessel integrity. Memorandum and Order, CLI-91-11. The petition was based on the NRC Staff's own documents, which demonstrate that the Yankee Rowe vessel has become seriously embrittled by exposure to neutron irradiation over the course of its 31-year operating history. According to the Staff documents, the vessel significantly exceeds NRC's 'reference temperature' criteria and fails to meet the Commission's fracture toughness standards. Moreover, for virtually its entire operating life, Yankee Rowe has failed to comply with NRC requirements for routine vessel testing and inspection to determine its condition." (pg. 1-2).
• "The Commission has conceded that its calculations of the risk of operating Yankee Rowe are not based on any current data about the composition or condition of the Yankee Rowe vessel; that the Staff's analysis is fraught with uncertainty; and that these uncertain risk estimates are higher than well established NRC standards for safe operation. The Commission has asked the public to accept this high level of risk at Yankee Rowe until the Staff and Yankee Atomic can come up with information about the vessel that would have been submitted years ago if the regulations for testing and surveillance of the vessel had been enforced. This approach to regulation of nuclear power plants stands the NRC's regulatory scheme on its head and undermines whatever small confidence the public has left in this agency." (pg. 10).

• "Over the thirty years that the Yankee Rowe plant has operated, irradiation by the reactor core has embrittled the pressure vessel steel, rendering it vulnerable to cracking and rupture. If such cracking and rupture occur, they will almost certainly lead to a meltdown and uncontrolled release of radioactivity to the environment. As discussed in detail below, the Yankee Rowe vessel violates the Commission's standards for pressure vessel toughness and ductility." (pg. 1).
• "Moreover, the vessel has never been examined to determine the exact severity of those violations; or to determine the existence and size of cracks or flaws in the vessel wall." (pg. 1).
• "...the Yankee Rowe nuclear power plant fails to comply with an array of fundamental requirements for maintaining pressure vessel integrity in pressurized water reactors. ...Yankee Rowe's noncompliance with NRC requirements for pressure vessel integrity poses a safety risk of commensurate, if not graver, dimension than the suspicion of ECCS pipe cracking that caused the Commission to order 23 plant shutdowns in 1975." (pg. 26).
• "The issues raised by YAEC's [Yankee Atomic Electric Company] noncompliance with NRC regulations raise grave safety questions of tremendous public importance." (pg. 27).
• An important early warning of NRC failure to enforce its own regulations and of the complicity and willingness of YAEC to operate unsafe nuclear power plants in violation of federal regulation.
• This plant was later permanently shut due to these flaws in the reactor vessel.
• This important report by Pollard and Curran is essential reading for anyone concerned about reactor vessel embrittlement.
• See also the testimony of D. Curran cited above.

• "...the Nuclear Regulatory Commission (NRC) confirmed that age-related degradation in boiling water reactors (BWRs) will damage or destroy vital internal components well before the standard 40-year BWR license expires, ... This paper focuses on ... degradation of the internal components in BWR pressure vessels. This study found that the nuclear industry -- the regulated and the regulators alike -- is not prepared to deal with the grave age-related problems that lie ahead." (abstract).
• "Research has shown that a multitude of both large and small nuclear plant components are susceptible to a staggering variety of aging mechanisms. Reactor vessels, steam generators, piping, valves, heat exchangers, pumps, motors, instrumentation, electrical cables, seals, and supports are all degraded by erosion, fatigue, corrosion, radiation and thermal embrittlement, and vibration. Studies have also demonstrated that some types of degradation cannot be detected using the established methods of periodic testing and inspection." (pg. 1).
• "To date, the single most significant finding resulting from the NRC's research program is that the essential conditions that produce stress corrosion cracking -- including corrosion-susceptible materials, a corrosive environment, and tensile stresses -- are all present in BWRs. So far, most of the documented cracking has been found in one component, the core shroud. But 18 other BWR internal components are also known to be susceptible to corrosion, fatigue, creep, embrittlement, and erosion, or to a combination of these degradative mechanisms." (pg. 1).
• "Most BWRs experience core shroud cracking after only 20 years of operation -- not 40 or 60." (pg. 1).
• "The synergistic effects of multiple degraded components is still a largely unexplored but critical aspect of the BWR aging cycle." (pg. 1).
• This report includes the following definitions of degradation mechanisms as described in NUREG/CR-5754, 1993, (pg. 8):
• Stress Corrosion Cracking: SCC refers to the weakening of a BWR internal structural component because of deterioration caused by electrochemical reactions with the surrounding material.
• Creep: The progressive deformation of a structure under constant stress is known as creep.
• Fatigue: As a structure vibrates in response to dynamic loads, cracks develop in certain BWR internal components.
• Embrittlement: Exposure of internal components to high temperatures (thermal embrittlement) and prolonged exposure to fast neutron fluxes (radiation embrittlement) make a material more brittle and vulnerable to cracking.
• Erosion: The abrasive effects of bubbles and droplets in a liquid flow can weaken BWR internal components.
• This report also includes a "Summary of NRC data on core shroud cracking" at 19 U.S. reactors. (pg. 4).
• See RADNET Section 12: Maine Yankee Atomic Power Company: Collapse of a Pyramid Scheme for an extensive list of NRC publications and reports on safety issues at pressurized water reactors (PWR) such as the MYAPC.

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

United States Nuclear Regulatory Commission. (August 5, 1992). IE information notice no. 82-27:  Fuel rod degradation resulting from baffle water-jet impingement.  IN 82-27. Office of Inspection and Enforcement, U.S. NRC, Washington, D.C. http://www.nrc.gov/NRC/GENACT/GC/IN/1982/in82027.txt.

• "On May 6, 1982, Portland General Electric submitted a Licensee Event Report (LER) 344/82-06, describing abnormal fuel clad degradation identified during a pre-planned fuel inspection to locate suspected leaking fuel assemblies.  Fuel rod damage involved 17 fuel assemblies examined at the end of Cycle 4 operation.  Portions of fuel rods were found missing and loose fuel pellets were discovered and retrieved from reactor vessel internals and the refueling cavity.  Visual inspections revealed severe perimeter fuel rod failures in 8 fuel assemblies.  Failures in the remaining 9 assemblies were detected by sipping operations, but did not exhibit visual damage." (pg. 1).
• "In general, the water-jetting-induced rod motion causes fuel rod fretting because of abnormal clad wear against the Inconel grid assemblies, which consist of slotted straps interlocked in an "egg-crate" arrangement." (pg. 2).
• "Coolant cross-flow through the enlarged baffle gaps results in high velocity jetting because of this pressure differential.  The baffle water-jet then impinges on fuel rods and induces excessive rod motion, producing severe clad degradation. " (pg. 3).

• "During shutdown inspection activities after Cycle 7 at Salem Unit 2 and Cycle 9 at Beaver Valley Unit 1, the licensees at both plants discovered numerous fuel rods that had developed fretting wear and perforation.  The fuel vendor attributed the degradation to grid-to-rod fretting resulting from flow-induced vibration of the fuel bundles.  All but one of the affected fuel assemblies were Westinghouse twice-burned VANTAGE 5H fuel located next to the core baffle or with a history of previous operation at a peripheral location.  The fretting wear occurred at the zircaloy mid-grid spacers rather than at lower grid locations where debris-induced fretting wear typically occurs.  In some of the affected assemblies, secondary hydriding also was evident." (pg. 1).
• "Recent operating experience of pressurized-water reactors has identified debris-induced fretting as a leading cause of fuel failure.  However, current experience also indicates that a new type of vibrational fretting is emerging." (pg. 3).

"This vibrational fretting involves the natural frequency and flow condition for fuel assemblies adjacent to the core baffle." (pg. 3).

• The following email correspondence was sent by CBM on April 27, 1999, referencing the material in these proceedings.
• To whom it may concern:  In writing to Ray Shadis of Friends of the Coast about the current lack of documentation of the environmental impact of reactor operations and decommissioning at MYAPC, I enclosed the following list of questions which I am sending out as a separate email.  Any comments or answers to these questions would be greatly appreciated.

Observation:

In reading the minutes of an Advisory Committee on Reactor Safeguards meeting on reactor fuels, onsite fuel storage and decommissioning on Friday April 24, 1998, (http://www.nrc.gov/ACRS/rrs1/Trans_Let/index_top/ACRS_sub_tran/Reactor_Fuels/rf980423), I was struck by the fact that NRC committee members really don't have a complete understanding of fuel cladding failure processes.  Some general questions that I have, which pertain to fuel cladding failure at MYAPC, include the following:

1. Was most of MYAPC fuel cladding failure due to grid to rod fretting?
2. Did MYAPC mix any vendor fuels or were all the fuels used from one vendor?
3. If they did use mixed fuels, did excessive cross flow play any role in fuel failure?
4. How much of the fuel cladding failure was due to manufacturing defects?
5. Were there any instances of total fuel cladding failure with accompanying pellet spillage?
6. The confidential inventory makes reference to a pipe with a fuel assembly enclosed in cement:  is this an example of a fragmented fuel assembly and if so, how and when did this occur?
7. Since the confidential MYAPC spent fuel pool inventory indicates pellets in numerous filters, is there any other source of these pellets other than fuel cladding failure?
8. Did any fuel overheat to the extent that it would show clad ballooning?
9. Was any fuel cladding failure due to flow induced vibrations and would these vibrations be the main source of grid to rod fretting at MYAPC?
10. Was any of the fretting induced by debris, e.g. from welding of the thermal shield?
11. Was most debris-induced fretting at the lower grid locations; was there any damaged at the mid-grid spacers?
12. Did bad end cap welds play any role in MYAPC fuel cladding failure incidents?
13. What is the total inventory of missing fuel pellets from MYAPC fuel assemblies and how many assemblies are missing pellets?
14. What percentage of these fuel assembly pellet losses were successfully captured by reactor coolant filters?
15. Once loose in the water systems are these pellets susceptible to fragmentation or mass wasting?
16. To what extent do fission products released from fuel cladding failure incidents become entrained in CRUD layers within the reactor containment?
17. To what extent does fuel cladding failure increase fission gas and how much of this gas escaped MYAPC reactor containment?  For example was cesium iodide released in an aerosol rather than a particulate form, and how far does it travel from the stacks after emission?
18. I have many other questions, but ultimately the most important question of all is what quantities of reactor-derived fission and activation products have spread beyond the reactor containment and where are they now located?
19. What NRC LERs and other reports provide documentation of these fuel cladding failures at MYAPC?

Seabrook fuel cladding failure accident, summer of 1998:

1. The minutes cited above also make reference (see page 24 and 67) to a fuel cladding failure accident in Seabrook, NH, which is particularly significant in that it occurred in fresh fuel.  If anybody receiving this email message has any further information about the Seabrook accident, please contact the Center for Biological Monitoring.  We would be particularly interested in locating any licensee event reports (LERs) or any other NRC documentation of this accident.
2. Again, what is the source term of this fuel cladding failure accident?
3. How much radioactivity passed beyond Seabrook fission product barriers?

Barbito, Karin and Rogosky, Donna. (January 31, 1999). Steam generators; remote visual inspection; an eye for steam generator maintenance. Nuclear Engineering International. pg. 21.

• "Westinghouse has developed Steam Generator Secondary Side Maintenance Guidelines to encourage utilities to develop a plan to proactively monitor and maintain its steam generators.  The emphasis is on visual inspection."
• "In response to steam generator wrapper support structure failures in foreign plants and degradation of tube support plates at US facilities, the US Nuclear Regulatory Commission (USNRC) issued Information Notice 96-09 in February 1996 followed by a supplement in July 1996.  In December 1997, Generic Letter 97-06 was issued concerning the same issue.  These documents emphasise the importance of developing a  maintenance plan, part of which includes thorough visual inspections of steam generator secondary side internals to evaluate structural integrity."
• "Westinghouse Electric Company, a leader in outage services, developed the Steam Generator Secondary Side Maintenance Guidelines in response to degradation concerns, as discussed above, and including:
• Structural integrity issues (outlined in NRC letters) - wrapper support structure degradation.
• Tube degradation (described in an NEI document).
• Loose parts.
• Corrosive deposits.
• Sludge accumulation.
• Fouling.
• Wear.
• These conditions can result in a reduction in main steam pressure, hydrodynamic instabilities, stress corrosion cracking, lower steam pressure, tube rupture, replacement of steam generators prior to expected life, plant shutdown etc."
• "The top of the tubesheet, tubelane and annulus region can be a collection point for foreign material.  If not removed, these foreign objects can cause tube wear and potentially a primary to secondary leak."

"Visual inspections are required in the upper bundle region to determine their general condition of the tubes and tube support plates.  Plant operating data has confirmed approximately 80% of corrosion product transport deposits in the tube support plate/upper tube bundle region of the steam generator.  These deposits can lead to the concentration of potentially aggressive chemical species or form a ledge type structure blocking flow holes which increases pressure drop."

• "NRC staff has been meeting with industry officials for months trying to iron out a means to regulate steam generators that satisfies both the agency's statutory need for safety assurances and industry's economic need for flexibility to continue to run their generators with flawed tubes."
• "For industry to obtain that flexibility, it needs to eliminate or change the current requirement in PWR technical specifications requiring that all tubes be plugged or repaired if tube flaws (initially wastage, now stress corrosion cracking) exceed 40% to 50% throughwall."

"But trying to find a generic regulatory vehicle or mechanism which accommodates both parties' needs, gets the parties out of cycle-by-cycle reviews, and passes legal muster has proven elusive."

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

• Licensee/Facility: Entergy Operations, Inc., Arkansas Nuclear 2, Russelville, Arkansas. Dockets: 50-368 PWR/CE. Notification: MR Number: 4-97-0013. Date: 01/31/97 SRI.
• SUBJECT: PRESSURE TEST OF ANO, UNIT 2, STEAM GENERATOR TUBES.
• "Arkansas Nuclear One (ANO), Unit 2, recently received the results of pressure tests that were performed on two steam generator tubes (R70C98 and R16C56), which were removed from Steam Generator A during a recent forced outage to repair a steam generator tube leak (PNO-IV-96-061, MR 4-96-0128). Both tubes burst at approximately 3200 psig, which was significantly below the test pressure of 4750 psig needed to satisfy the Regulatory Guide 1.121 structural integrity criteria of three times the primary-to-secondary normal operating differential pressure."
• "Both tubes were found during the forced outage to contain single axial cracks at the first eggcrate support on the hot-leg side of the steam generator. For Tube R70C98, analysts found the bobbin coil data showed the defect as a distorted support indication. The motorized rotating pancake coil (MRPC) examination data indicated a 1.15 inch long flaw, with a throughwall depth of 81 percent. The length of the flaw in Tube R16C56 was found by MRPC to be 1.13 inches, and the depth was found to be 89 percent by bobbin coil and 78 percent by MRPC examination. Examination of these tubes during the previous refueling outage, 2R11, which was completed in November 1995, did not reveal any degradation."

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

• "The steam generator (SG) tubes in pressurized water reactors have a number of important safety functions. These tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the RCPB, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system; the SG tubes are also relied upon to isolate the radioactive fission products in the primary coolant from the secondary system. In addition, the SG tubes are relied upon to maintain their integrity, as necessary, to be consistent with the containment objectives of preventing uncontrolled fission product release under conditions resulting from core damage severe accidents."
• "In this regulatory guide, tube integrity means that the tubes are capable of performing their intended safety functions consistent with the licensing basis, including applicable regulatory requirements."
• "Concerns relating to the integrity of the tubing stem from the fact that the SG tubing is subject to a variety of corrosion and mechanically induced degradation mechanisms that are widespread throughout the industry. These degradation mechanisms can impair tube integrity if they are not managed effectively."
• "Title 10 of the Code of Federal Regulations establishes the fundamental regulatory requirements with respect to the integrity of the SG tubing. Specifically, several General Design Criteria (GDC) in Appendix A, 'General Design Criteria for Nuclear Power Plants,' to 10 CFR Part 50, 'Domestic Licensing of Production and Utilization Facilities,' are applicable to the integrity of the steam generator tubes."
• "These guidelines are intended to provide licensees with the flexibility to adjust the specifics of the program elements within the constraints of these guidelines to reflect new information, new NDE technology, new degradation mechanisms or defect types, changes in flaw growth rates, and other changing circumstances. Licensees must develop and implement steam generator defect specific management (SGDSM) strategies to fully achieve this flexibility. SGDSM strategies involve an integrated set of program elements, paralleling those in this regulatory guide, that address specific defect types."
• "The tube inspections are followed by assessments of tube integrity performance relative to performance criteria. Performance criteria acceptable to the NRC staff are given in Regulatory Position 2 of this regulatory guide. These performance criteria address three areas of tube integrity performance:   structural integrity, operational leakage integrity, and accident-induced leakage integrity."
• "Performance criteria acceptable to the NRC for accident leakage integrity are identified in Regulatory Position 2.3. These involve accident leakage rates consistent with those assumed in the licensing basis accident analyses for purposes of demonstrating that the accident consequences are in accordance with 10 CFR Part 100, or some fraction thereof, and GDC-19."
• "The objective of SG tube inspection is to provide sufficient information concerning the defect types present in the SGs, the tubes that contain defects, and the size of these defects such that when implemented in conjunction with the other programmatic elements of this regulatory guide, there is reasonable assurance that the tube integrity performance criteria in Regulatory Position 2 are being maintained throughout the time period between SG tube inspections."

Ford et. al. (1974). An assessment of the emergency core cooling systems rule making hearings.

• See complete citation in RAD12: Maine Yankee: Public Safety Bibliography.

• "This article presents the methodology, findings, and conclusions of a study conducted by the U.S. Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operation Data (AEOD) on loss of spent fuel pool (SFP) cooling." (abstract, pg. 237).

Airozo, Dave. (January 18, 1999). Agency staff says 'hot particle' rules do more harm than good. Inside N.R.C. 21(2). pg. 4.

• "Hot particles are tiny, usually microscopic, particles that most commonly contain cobalt-60 or fission products.  They apparently become electrically charged as a result of radioactive decay and tend to 'hop' from one surface to another.  Particles from leaking reactor fuel, commonly known as 'fuel fleas,' are among the hot particles found at reactor sites."
• "Because they are highly radioactive beta or beta-gamma emitters with relatively high specific activity, the hot particles deposit very large, highly nonuniform doses to very small amounts of tissue if they land on a worker's skin."
• "Because the [10 CFR] Part 20 worker whole-body skin exposure limit of 50 rem per year is not particularly relevant to the type of exposure and consequences arising from hot particle skin contact, NRC has used enforcement discretion in deciding what to do about worker exposures that exceed the 50-rem limit when the exposure is caused by a hot particle."
• "The overall cost of a hot particle control program runs from $200,000 to $2-million annually per reactor site, according to an Electric Power Research Institute (EPRI) report cited by the NRC staff."
• "Under the staff's plan, the monitoring practices, which NRC and the industry say add 3-5 person-rem per reactor outage per site, would be loosened and, instead of trying to meet the 50-rem limit set in Part 20, utilities would shoot for limiting hot particle skin exposures to 300 rads averaged over an area of 1 cm²."
• "If an exposure exceeded the 300-rad target, the utility would have to report that to the NRC and tell the agency what steps would be taken to fix any problems that caused the excess exposure.  The excess exposure would not, however, be considered an overexposure in regulatory terms."

"The staff also proposed an overall 1,000-rad dose cap, aimed at providing greater assurance that extremely high hot particle doses don't occur."

Many controversial issues are part of the current debate on how and where to dispose of reactor-derived spent fuel assemblies.  Among the most important controversies, aside from the final location of spent fuel, involves the design of dry casks to hold spent fuel when reactor spent fuel pools become filled to capacity as is now the case at a number of U.S. reactors.  As MYAPC, Connecticut Yankee and other facilities undergo decommissioning, spent fuel now stored underwater in fuel pools will be transferred to independent spent fuel storage installations (ISFSIs) while awaiting the unlikely construction of a final repository at Yucca Mt.  One of the most important stages in this process of storing and/or disposing of spent fuel is the development of appropriate "dry casks" to replace wet storage, reactor spent fuel pools not being designed to hold spent fuel for long periods of time.  A number of new dry cask designs are now being considered by the NRC for licensing.  This new model of dry cask is called a multi-purpose container (MPC) and is meant to be used not only for onsite storage of spent fuel but for its transport and final geological emplacement.  An important annoying detail for the nuclear industry is the fact that most dry casks now in use are obsolete and cannot be used for transport for final geological disposal.  Rather, those utilities such as Northern States Power, which have already purchased and are using the older dry casks, will have to take the spent fuel out of these casks in an underwater environment and transfer the spent fuel into new multi-purpose casks prior to any transport of spent fuel to a monitored retrievable storage facility such as that now being proposed in congress as a temporary alternative to final geological disposal in Yucca Mt.

Numerous controversial issues attend spent fuel storage including spent fuel pool safety, obsolete dry cask safety issues, NRC design and licensing criteria for new MPCs, transportation safety issues and final geological repository safety issues.  Presently, there are no licensed multi-purpose canisters available to transport spent fuel to a temporary monitored retrievable storage (MRS) facility, if such a facility is authorized by congress.  MPCs as well as ISFSIs are very expensive components of the back end of the nuclear fuel cycle, and even if the safety issues attending appropriate MPC design are resolved, the political issues of transport and disposal of spent fuel are not, and funding of new MPCs would greatly exceed all the funds collected to date by the Department of Energy for a final geological repository.  Most controversial of all is the fact that the contents of the spent fuel pools at US reactors include a variety of highly radioactive wastes which cannot be sited as "standard spent fuel" in newly designed MPCs.  Typical spent fuel pool contents that are not destined for MPCs include failed fuel assemblies, fuel assemblies which have been altered, fuel assemblies which have had significant damage but are not considered "failed," neutron sources initially used to start the chain reaction, highly radioactive filters which contain spent fuel pellets and activation products from the reactor containment and a wide variety of debris and other equipment which is too radioactive to site as low-level waste.  One of the upcoming problems with any "low-level" waste facility is that NRC concentration averaging policies allow much of this GTCC waste and spent fuel debris to be diluted with class A low-level wastes and then sited as class C low-level waste.  To review the contents of a relatively "clean" reactor's spent fuel pool, see Maine Yankee Atomic Power Company's recently released spent fuel pool inventory, much of which is not destined for a geological repository.  At least MYAPC has the advantage of not being burdened with obsolete dry casks that will have to be replaced with much more expensive multi-purpose canisters as is the case at a number of other US reactors.  Citations pertaining to MPC safety and design issues and spent fuel disposal in general will be posted in this section of RADNET as they become available.

United States Nuclear Regulatory Commission. (January 1997). Standard review plan for dry cask storage systems. NUREG-1536. Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, U.S. NRC, Washington, D.C. http://www.nrc.gov/NRC/NUREGS/SR1536/index.html.
 

g. MOX

Leventhal, P. and Dolley, S. (March 1, 1999). The reprocessing fallacy:  An update. Presented to the special panel session on spent fuel reprocessing, Waste Management 99 Conference, Tucson, Arizona. Nuclear Control Institute, Washington, DC. http://www.nci.org/pl-wm99.htm.

Lyman, E.S. (January 21, 1999). Public health consequences of substituting mixed-oxide for uranium fuel in light-water reactors. Nuclear Control Institute, Washington, DC. http://www.nci.org/moxsum.htm.

• The following quotations are all from the Executive Summary of this report.
• "Under one approach, known as "can-in-canister" immobilization (CIC), plutonium will be incorporated into chemically stable ceramic discs.  These discs will in turn be embedded in canisters of 'vitrified' (glassified) high-level radioactive waste (VHLW) at the Defense Waste Processing Facility (DWPF) at the Savannah River Site in South Carolina."
• "Cost and public health impact were major considerations in the process that DOE used to select MOX and immobilization from the large number of disposition options ... DOE argued that there are no decisive differences between the MOX [mixed plutonium-uranium oxide] and immobilization options with regard to any of its evaluation criteria ... However, this report concludes that DOE's evaluation is inaccurate.  We find that the public health risks associated with the MOX approach are significantly greater than those associated with CIC.  This is due primarily to our findings that the consequences of severe accidents involving LWRs [light-water reactors] with MOX cores are likely to be greater than those involving LEU [low-enriched uranium oxide] cores."
• "The total inventory of highly radiotoxic actinides, including plutonium-239 (Pu-239), americium-241 (Am-241), and curium-242 (Cm-242), is significantly greater in MOX cores than in LEU cores throughout the operating cycle.  Our analysis shows that the public health consequences of some severe accidents will be greater for reactors fueled with MOX."
• "For the case considered in this study we find that, compared to an LEU core, a full [weapons grade] WG-MOX core will contain about three times the amount of Pu-239, seven times as much Am-241 and seven times as much Cm-242 at the end of an operating cycle (i.e. just before the reactor is shut down for reloading). For MOX fabricated with reactor-grade plutonium (RG-Pu), Am-241 and Cm-242 inventories are greater by additional factors of 4 and 3, respectively."

"The use of WG-MOX in U.S. PWRs is not likely to lower the probability that a severe loss-of-containment accident may occur and may in fact increase it significantly."
• "The ability of high-burnup MOX fuels in current use to withstand severe accident conditions is inferior to that of LEU fuel."
• "A MOX-fueled PWR may have a greater risk of experiencing pressurized thermal shock of the pressure vessel."
• "Ice-condenser containments may be more vulnerable to early failure in a severe accident than large dry containments."
• "A severe accident at a PWR with a reactor-grade MOX (RG-MOX) core would cause up to twice as many latent cancer fatalities (LCFs) as would an accident at a PWR with a WG-MOX core."
• "Licensing of U.S. reactors to use MOX will have to take place primarily on a site-specific level. In addition, an NRC finding that MOX use poses "no significant hazards" under 10 CFR 50.92 clearly would not be justified."
• "Limitations on MOX fuel burnup to below 36 GWD/MT should be imposed unless high burnup safety issues are resolved."
• "The U.S. plan to encourage Russia to use WG-MOX in Russian and Ukrainian VVER-1000 LWRs poses even greater risks than the plan for U.S. domestic use of WG-MOX."
• "Risks associated with irradiation of WG-MOX in both U.S. LWRs and Russian VVER-1000s could be averted if both nations implemented an all-immobilization policy for the entire stockpile of excess WG-Pu. The use of MOX is unnecessary and should be avoided.

• This report contains a summary of the Los Alamos National Laboratory report LA-UR-96-4764 Gallium in weapons-grade plutonium and MOX fuel fabrication.
• An excellent concise report on the MOX fuel controversy. This report is available in hard copy or on the Internet from IEER, see RAD 13: RADLINKS, Part II-A.

Minns, J.L. and Masnik, M.T. (April 1998). Staff responses to frequently asked questions concerning decommissioning of nuclear power reactors: Draft report for comment. NUREG-1628. Division of Reactor Program Management, Office of Nuclear Reactor Regulation, US NRC, Washington, DC.

• See RAD 12-F: Maine Yankee: Decommissioning Debacle Bibliography for comments on this publication.

• Another milestone in the incompetent and deficient NRC oversight of its licensees.

• This citation is reviewed and annotated in the bibliography of RAD12: Maine Yankee Atomic Power Company. Please refer to United States Nuclear Regulatory Commission; all NRC reports in this section are listed in the order of their date of publication. A number of other NRC publications pertaining to nuclear power stations are reviewed in the bibliography in RAD12: Section 2-D: Public Safety Bibliography.

United States Nuclear Regulatory Commission. (1986). Residual radionuclide contamination within and around commercial nuclear power plants. NUREG/CR-4289. U. S. NRC, Washington D.C.

• This report addresses "radionuclides (both activation and fission products) transported from the reactor pressure vessel and deposited in all other contaminated systems of each nuclear plant" (pg. iii).
• This report contains extensive documentation of residual radionuclide concentrations, distributions and inventories at seven nuclear power plants (four shut down and three operating facilities). For example, transuranic radionuclides at the Humboldt Bay generating station include ^239,240Pu to 230 pCi/kg of surface soil.
• Nuclide concentrations are often given in concentrations of pCi/g due to the contamination levels being documented.

United States Nuclear Regulatory Commission. (1998). Evaluation of loss of offsite power events at nuclear power plants: 1980-1996. NUREG/CR-5496. Office for Analysis and Evaluation of Operational Data, U.S. NRC, Washington, D.C.

• Inside N.R.C. on December 7, 1998, reports that this report "updates a similar station blackout report for the period of 1969-1985 (NUREG-1032).  The original blackout report noted that 'plant-centered' events accounted for most off-site power losses, and the most recent study supports that finding." (pg. 18).
• "...loss of off-site power from grid-related events was very small -- only five during the 16-year period and none during the 1990s.  Only one event occurred during the report period that caused total and sustained voltage loss to all safety buses due to a grid disturbance.  That was a 1985 fire near Turkey Point." (pg. 18).
• "Most plant-centered initiating events while on line were caused by equipment faults, 58%, with human error accounting for 23%.  But during shutdown, the opposite holds, with human error being the major factor 58% of the time." (pg. 18).

• See comments on this publication in RAD 12: Maine Yankee, Part 5-F: Decommissioning Debacle Bibliography.