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locations of Nuclear Power Reactors: WORLD MAP
A. United States Nuclear Power Plants
Three Mile Island
Other Citations about U.S. Nuclear Power Plants
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.)
Nuclear Regulatory Commission Publications
(if not listed in any of the above categories)
B. Canadian Nuclear Power Plants
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.
C. Russian Nuclear
D. Japanese Nuclear Power Plants
E. European Nuclear Power Plants
G. Biological Monitoring
(see extensive citations in RAD7: Plume Pulse Pathways)
|A. United States Nuclear
of locations of Nuclear Power Reactors: NORTH AMERICA
MYAPC: Case Study of an Aging Pressurized Water
(see RADNET Section 12)
Click on Maine Yankee Atomic Power
Company for information on a case study devoted to the many safety,
legal, economic, and decommissioning issues surrounding the operation of
a specific U.S. nuclear power generating station at Wiscasset, Maine. Recent
developments at MYAPC include the following:
The discovery of circumferential cracking in 60% of MYAPC
steam generator tubes in 1995 as well as the confirmation of allegations
of fraudulent computer codes for emergency core cooling systems and containment
analyses have opened a floodgate of NRC inspection reports, safety assessments
and investigative reports which dramatically illustrate a multiplicity
of problems associated with aging pressurized water reactors.
RADNET Section 12 organizes Maine Yankee Atomic Power
Company citations into four categories: safety, economic, legal and decommissioning
issues. Of particular interest is the startling NRC documentation of:
Ongoing generic safety issues (steam tube degradation,
fire barrier deficiencies, crossed cables, etc.)
Inherent design flaws in U.S. pressurized water reactors
not previously noted.
The inability of NRC inspectors to observe existing long-standing
safety defects which were only discovered via extensive repairs or stepped
The complicity of the NRC and licensee employees in a
series of illegal for-profit power up-rates based on fraudulent computer
The difficulties of collecting sufficient funds for decommissioning
in an era of energy deregulation and the current predicament of a closed
nuclear power station with 169 million dollars on hand for decommissioning
(plus an additional 60 million dollars in the mill rate fund for spent
fuel disposal). MYAPC waste storage, disposal and decommissioning costs
are expected to be in excess of 1 billion dollars excluding 152 million
dollars spent on the attempt to revive this aging facility.
The recent decision to terminate repair attempts completes
the collapse of the MYAPC pyramid scheme into economic, legal and technological
The primary documentation of the MYAPC debacle derives
from NRC inspections, reports and investigations despite NRC complicity
in the problems at this plant. The positive role played by the NRC in following
up allegations made by the whistleblower's letter of Dec. 4, 1995 raises
hope that the NRC will follow through on its most important obligation
of the 21st century: closing unsafe, uneconomical and aging U.S. nuclear
energy generating stations.
The citations pertaining to MYAPC as an aging nuclear
facility should be of interest to anyone concerned with the public safety
implications and economic impact of pressurized water reactors (as well
as boiling water reactors) as long as even a single one of these technological
dinosaurs remains in service. The MYAPC debacle marks the twilight of the
nuclear energy era in America.
The MYAPC facility closed in May of 1997. Post-closure
developments are listed in a fifth section of RAD 12: "Twilight of
a Nuclear Era: The Maine Yankee Nuclear Power Company". The decommissioning
of the MYAPC facility along with the Connecticut Yankee facility provide
an excellent opportunity to examine NRC and licensee attempts to deal with
the dilemmas of dismantling typical pressurized water reactors. The MYAPC
decommissioning plan may be accessed at the NRC-OPA:
Yankee Documents link in RADLINKS: Part II: D-5.
The MYAPC Reactor Vessel
Inventory including reactor vessel GTCC internal components (+/- 4,170,000
Ci at 2 years cooling) are now posted in RADNET Section 12: Part 5: Decommissioning
Debacle and are of particular interest to anyone concerned with decommissioning
costs, scenarios or the issue of what actually will be sited in a "low-level"
radioactive waste landfill.
Note that NRC generated reports published after mid-1998 that do not
relate directly to MYAPC will be cited and annotated below in this section.
Those prior to mid-1998 may be found in RAD12 under the sections: public
safety bibliography, economic
issues bibliography, legal issues
bibliography and decommissioning
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.
Division of Health Engineering. (1996). Maine Yankee
environmental monitoring: Summary of other media. Unpublished, publicly
available research, Augusta, ME.
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
England, R.W. and Mitchell, E. (1987). Estimates of
environmental accumulations of radioactivity resulting from routine operation
of New England nuclear power plants (1973-84). (Report No. 1). A report
of the Nuclear Emissions Research Project, Whittemore School of Business
and Economics, University of New Hampshire, Durham, NH.
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,
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.
Hess, C.T. and Smith, C.W. (1976). Radioactive isotopic
characterization of the environment near Wiscasset, Maine using pre and
post-operational surveys in the vicinity of the Maine Yankee nuclear reactor.
Technical Note ORP/EAD-76-3. U.S. Environmental Protection Agency, Washington,
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.
Lutz, R.A., Incze, L.S., and Hess, C.T. (1980). Mussel
culture in heated effluents: Biological and radiological implications.
In: Mussel culture and harvest: A North American perspective (ed.
R.A. Lutz). Elsevier, Amsterdam.
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
up to 5,620 pCi/kg near the plant outfall.
One hot particle was noted containing 7,700 pCi of
and had a total activity of 9,000 pCi in a mass less than 20 µg.
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.
|June 1977-Jan. 1978
||Bailey's Cove, Wiscasset Maine
||Mytilus edulis, soft tissue
||320 pCi/kg mean
|June 1977-Jan. 1978
||Bailey's Cove, Wiscasset Maine
||Mytilus edulis, shells
||150 pCi/kg mean
|June 1977-Jan. 1978
||Bailey's Cove, Wiscasset Maine
||Mytilus edulis, shells
||211 pCi/kg mean
McCarthy, W.J., Ryder, D.L. and Antonitis, J.D. (1978).
concentrations in New England seaweeds following the Chinese nuclear bomb
test of March, 1978. Report No. 342. U.S. Department of Energy,Washington
D.C. pg. 57-77.
"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,
40K were detected in both the shells and soft tissues of
the mussels cultured in these waters." (pg. 187).
Yankee Atomic Electric Company.
(1991). Maine Yankee Atomic Power Station: Maine Yankee Atomic Power
Company: Annual radiological environmental monitoring report: January -
December 1990. Yankee Atomic Electric Company, Bolton, MA.
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
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
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:
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.
Atomic Power Company Decommissioning Plan
In early September 1997, Connecticut Yankee filed a
decommissioning plan with the NRC, which the NRC has posted in its entirety
Decommissioning this facility, which recently closed, is estimated to cost
426.7 million 1996 dollars and be completed by 2004. This decommissioning
plan allegedly includes decontamination and removal of all plant structures
and systems except for the spent fuel storage building. The site
is supposed to be available for unrestricted use in 2004. Recent revelations
of extensive on-site contamination (see New
York Times article on contamination at Connecticut Yankee) due
to leaking spent fuel in the early years of operation may complicate this
decommissioning plan. This 13 page proposal is extremely brief and provides
only a sketchy description of decommissioning activities. One particularly
interesting component of this brief proposal is that the reactor vessel
may be removed with the highly active GTCC internals intact and then disposed
of as low-level waste because the radioactivity in the entire vessel package
averages out to a class C category. "This allows the vessel including the
internals to be qualified for normal conditions of transport" i.e. as low-level
waste. This is a good example of possible shortcuts to be used in getting
rid of orphan GTCC wastes which are too radioactive to put in a low-level
waste site by themselves: mix the GTCC wastes with enough low-level waste
and presto, you have low-level waste.
United States Nuclear Regulatory Commission. (March
1998). Haddam Neck Historical Review Team Report. US NRC, Washington,
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.
Harold Denton, director of the Office of Nuclear Reactor
Regulation for the NRC during the 1979 Three Mile Island accident, has
donated his personal papers regarding his involvement with the accident
to the Pennsylvania
State Archives. They will be made available to the public sometime
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.
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).
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.
||Indian Point, NY
||236 pCi/kg dry sediment
Makhijani, A. and Saleska, S. (1996). The nuclear power
deception: U.S. nuclear mythology from electricity "too cheap to meter"
to "inherently safe" reactors. Institute for Energy and Environmental
Research, Tacoma Park, MD.
Lower deposition levels away from plant, typically 14
- 64 pCi/kg.
Makhijani, A. and Makhijani, A. (1995). Fissile materials
in a glass, darkly: Technical and policy aspects of the disposition of
plutonium and highly enriched uranium. Institute for Energy and Environmental
Research, Tacoma Park, MD.
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).
Oak Ridge National Laboratory. (1995). Integrated Data
Base Report 1994 U.S. Spent Nuclear Fuel and Radioactive Waste Inventories,
projection and characteristics. Report No. DOE/RW-0006, Rev. 11. Oak
Ridge National Laboratory, Oak Ridge, TN http://cid.em.doe.gov/.
Oak Ridge National Laboratory. (December 1997). Integrated
Data Base Report, 1996: U.S. Spent Nuclear Fuel and Radioactive Waste Inventories,
projections, and characteristics (revision 13). Report No. DOE/RW-0006,
Rev. 13. Oak Ridge National Laboratory, Oak Ridge, TN.
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
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
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.
Smeloff, Ed and Asmus, Peter. (1997). Reinventing electric
utilities: Competition, citizen action, and clean power. Island Press,
Washington, DC. IS.
"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:
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).
United States Nuclear Regulatory Commission. (December
1998). Report on waste burial charges. NUREG-1307. U.S.
NRC, Washington, D.C. http://www.nrc.gov/NRC/NUREGS/SR1307/r8/index.html
"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
Weil, Jenny and Stellfox, David. (January 4, 1999).
AEOD abolished, research office expanded under reorganization plan. Inside
N.R.C. 21(1). pg. 1.
Yankee Nuclear Power Station. (1993). Decommissioning
Plan. Yankee Atomic Power Company, Rowe, MA.
"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
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
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:
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
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
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.
|5. Safety Issues at U.S. Nuclear
Government Accounting Office. (March 19, 1999). Nuclear
regulation: Strategy needed to regulate safety using information on
risk. GAO/RCED-99-95. GAO, Washington, DC.
Lochbaum, David. (June 1998).
report on safety in America's nuclear power industry. Union of Concerned
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
"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).
Stellfox, David. (February 15, 1999). Database suggests
electrical fires more common as plants age. Inside N.R.C. 21(4).
"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,
"...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,
"At the LaSalle, Millstone, and Sequoyah plants, problems often remained
undetected or uncorrected over a long period of time." (Executive Summary,
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
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.
Pollard, R.D. and Curran,
D. (June 4, 1991). Petition for emergency enforcement action and request
for public hearing. Before the U.S. Nuclear Regulatory Commission.
Union of Concerned Scientists.
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).
Pollard, R. (September 1995).
nuclear power plants -- showing their age: Case study: core shroud cracking.
Union of Concerned Scientists.
"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
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.
Pollard, Robert. (December 1995).
U.S. nuclear power
plants -- showing their age: Case study: Reactor pressure vessel embrittlement.
U.S. Congress Office of Technology Assessment. (1993).
nuclear power plants: Managing plant life and decommissioning.
"...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.
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.
|b. Spent Fuel Cladding
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.
United States Nuclear Regulatory Commission. (October
12, 1993). NRC information notice 93-82: Recent fuel and core performance
problems in operating reactors. IN 93-82. Office of Nuclear Reactor
Regulation, U.S. NRC, Washington, DC. http://www.nrc.gov/NRC/GENACT/GC/IN/1993/in93082.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.
"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).
United States Nuclear Regulatory
Commission. (April 1998). Proceedings of U.S. NRC Advisory Committee
on Reactor Safeguards Meeting on Reactor Fuels, Onsite Fuel Storage, and
Decommissioning, Friday, April 24, 1998. U.S. NRC, Washington, D.C.
"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).
See the latest on CBM's work on this subject in Section
12: Maine Yankee: Part 5-E: Decommissioning
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.
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
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
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
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
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
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
3. How much radioactivity passed beyond Seabrook fission product
|c. Steam Generator Degradation
Barbito, Karin and Rogosky, Donna. (January 31, 1999).
Steam generators; remote visual inspection; an eye for steam generator
maintenance. Nuclear Engineering International. pg. 21.
Stellfox, David. (March 1, 1999). Staff search in vain
for regulatory vehicle for steam generator plans. Inside N.R.C.
21(5). pg. 2.
"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
Tube degradation (described in an NEI document).
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
"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."
United States Nuclear Regulatory Commission. (September
1993). Boiling-water reactor internals aging degradation study.
United States Nuclear Regulatory Commission. (April 28,
1995). Generic Letter 95-03: Circumferential Cracking of Steam Generator
United States Nuclear Regulatory Commission. (February
3, 1997). Region IV morning report, page 9, Subject: pressure test of
ANO, unit 2, steam generator tubes. U.S. NRC, Washington, D.C.
"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."
United States Nuclear Regulatory Commission. (February
6, 1997). Proposed generic communication: Degradation of steam generator
United States Nuclear Regulatory Commission. (December
Draft regulatory guide DG-1074: Steam generator tube integrity.
Office of Nuclear Regulatory Research, U.S. NRC, Washington, D.C. http://www.nrc.gov/NRC/RG/DG/1074/DG-1074.html.
Licensee/Facility: Entergy Operations, Inc., Arkansas Nuclear 2, Russelville,
Arkansas. Dockets: 50-368 PWR/CE. Notification: MR Number: 4-97-0013. Date:
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."
"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
"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
"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
|d. LORCAs and Spent Fuel Cooling
Ford et. al. (1974). An assessment of the emergency
core cooling systems rule making hearings.
Ibarra, J.G., Jones, W.R., Lanik, G.F., Ornstein, H.L.
and Pullani, S.V. (July-September, 1996). Assessment of spent fuel cooling.
Safety: Technical Progress Journal. 37(3). pg. 237-255.
"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).
"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
|f. Spent Fuel Storage
and Disposal (Dry Casks/Multi Purpose Casks, etc.)
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
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.
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.
Toevs, J.W. and Beard, C.A. (February, 1997). Gallium
in weapons-grade plutonium and MOX fuel fabrication. Science for Democratic
Action. An IEER (Institute for Energy and Environmental Research) publication.
5(4). pg. 11.
The following quotations are all from the Executive Summary of this
"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
"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
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.
|6. Nuclear Regulatory Commission
Publications (if not listed in any of the above categories)
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,
Santana, H., Stryker, W.J. and Childs, J.E. (December
31, 1998). Office of the Inspector General event inquiry: NRC
staff's handling of harassment and intimidation (H&I) complaints at
Millstone: Case No. 99-01S. OIG Report 99-01s. Office of the
Inspector General, U.S. NRC, Washington, DC. http://www.nrc.gov/OPA/reports/oig9901s.htm.
See RAD 12-F: Maine Yankee: Decommissioning
Debacle Bibliography for comments on this publication.
United States Nuclear Regulatory Commission. Reactor
safety study. WASH-1400. U.S. NRC,Washington, D.C.
Another milestone in the incompetent and deficient NRC oversight of
United States Nuclear Regulatory Commission. Evaluation
of station blackout accidents at nuclear power plants. NUREG-1032.
U.S. NRC, Washington, D.C.
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).
radionuclide contamination within and around commercial nuclear power plants.
NUREG/CR-4289. U. S. NRC, Washington D.C.
United States Nuclear Regulatory Commission. (July 1996).
waste: Production, storage, disposal. NUREG/BR-0216.
Office of Public Affairs, 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.
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
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).
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,
United States Nuclear Regulatory Commission. (August 1998).
regulatory guide DG-4006: Demonstrating compliance with the radiological
criteria for license termination. Office of Nuclear Regulatory Research,
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.
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