LICENSE RENEWAL FOR NUCLEAR POWER PLANTS
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Since the 1973 oil embargo, demand for electricity in the United
States has increased at a slower rate than historic-demand
projections would have predicted for the economic growth of the
past 18 years. Nonetheless, the demand has increased by 60% over
this time. Commercial nuclear power plants, some of which have been
operating since the 1960s, have met a significant portion of this
increased demand. Today nuclear generating capacity in the United
States totals over 100 gigawatts, representing over 20% of the
nation's total capacity for generating electricity. In addition to
their energy contribution, U.S. nuclear plants have offered several
environmental advantages over plants fired by fossil fuels,
including the absence of emissions of carbon dioxide (one of the
greenhouse gases), sulfur dioxide (a possible cause of acid rain),
and nitrogen oxides (contributors to urban smog and acid rain).
Although nuclear energy has played an important role over the past
three decades, it is approaching a crossroads. One decision must be
made concerning the feasibility of building future nuclear plants,
and another must be made concerning the need to extend the
operating life of existing nuclear plants.
Although the most recent order for a commercial nuclear power plant
in the United States was placed in 1978, the plants ordered before
that year have continued to be completed and connected to the
nation's electrical grid. Today five plants remain to be completed.
It is recognized that electric utilities probably will not order
additional nuclear plants unless regulatory requirements are
restructured to reduce their complexity and the time required for
license approval and construction. Progress is being made; however,
DOE, the electric utility industry, the Nuclear Regulatory
Commission (NRC), and the Congress are taking major steps toward
establishing designs, regulatory processes, and a legal framework
for future nuclear plants. For example, in late May 1992 the U.S.
House of Representatives passed legislation similar to a previously
passed Senate bill and similar to new NRC procedures that will
permit one-step licensing of nuclear power plants. Concurrently,
the industry and DOE are working on designs for passive-water
reactors and advanced reactor concepts to be certified by the NRC.
Because licenses of the existing plants are for 40 years, they will
begin to expire at the end of the century, before the advanced
nuclear plants can be available, even under the license-reform
rules. Nonetheless, the United States is on the verge of a
significant increase in nuclear power production capability, an
increase that will be vital until advanced plants take their place
in power generation. This increase is to come from the renewal of
the operating licenses for the existing nuclear plants. Instead of
permanently shutting down dozens of nuclear plants and removing an
important source of electricity, the NRC and utilities may extend
their operating life 20 years. License renewal is the second most
important decision that must be made at the nuclear energy
crossroads.
The figure here shows the rapid loss of nuclear generating capacity
that will occur after the year 2010 if no additional plants are
built or licenses extended. However, it is anticipated that up to
75% of the 112 existing plants, or some 80 plants, will apply for
license renewal. In addition to helping meet demand for
electricity, renewal of licenses can lead to major economic
savings. For example, it has been estimated that if the licenses
were renewed for 20 years for all of the existing nuclear plants,
a savings of up to $200 billion could be realized. The principal
reason is that the utilities have amortized the capital cost of the
plants over the initial 40-year license period. Increased operating
and maintenance costs for an aging plant during a renewal period
will be much less than the capital investment required to build new
plants, nuclear or fossil. Additionally, the economic
infrastructure of the utility industry would experience significant
relief if replacement of these plants were deferred by up to 20
years.
Current federal regulations permit license renewal, but they do not
give guidance on the requirements. Anticipating that nuclear
utilities would request renewal of plant licenses, the NRC began
several years ago to establish regulatory policies, technical
bases, and procedures appropriate to license renewal. The NRC is
now completing that process, and the regulatory bases to approve
these renewals and to oversee the plants during the renewal period
are essentially all in place. For this effort, the NRC has drawn
heavily on ORNL to conduct supporting research.
The regulatory requirements for license renewal fall into two
categories--technical and environmental. The NRC places such
requirements in the U.S. Code of Federal Regulations. Plants
seeking license renewal must satisfy review requirements both at
the time renewal is granted and during the renewal period. As in
earlier phases of developing nuclear power, ORNL has contributed
significantly to both areas by supplying data, analysis, and
technical interpretations to the NRC. The NRC requirements and some
of ORNL's contributions are presented below.
TECHNICAL REQUIREMENTS
The technical requirements for license renewal are based on two key
principles: (1) the existing current licensing basis (CLB) for each
operating reactor provides an acceptable level of safety for
operation during the renewal term, and (2) each plant's licensing
basis must be maintained during the renewal period, using existing
or new programs that focus on the management of age-related
degradation of plant systems, structures, and components (SSCs).
Evaluators of a renewal application will determine if the applicant
has taken the required steps toĽ document the CLB for that
plant--that is, state all the codes, standards, and regulatory
guides that apply to that specific plant;
- identify the SSCs--safety equipment plus those components
that may affect the performance of safety equipment;
- complete an assessment of the plant to verify that it
complies with the CLB at the beginning of the renewal
period;
- establish a program that can identify and monitor
age-related degradation of SSCs throughout the renewal
period; and
- establish a program to demonstrate that the plant is in
compliance with the CLB throughout the renewal period.
The CLB is composed of the original licensing requirements (e.g.,
codes, standards, regulatory guides) plus requirements that have
been added for the plant during its current license period.
Compliance with the CLB ensures that at least the current margins
of safety are maintained throughout the renewal period. One major
emphasis for the NRC has been to identify exactly which parts of
the plant should be required in the SSCs to avoid or mitigate the
consequences of hypothetical accidents.
ORNL has been a major participant in conducting research on the
age-related degradation of many of these structures and components.
This work has included identifying degradation mechanisms,
developing methods for monitoring degradation, evaluating
approaches (rules, criteria, and limits) for mitigating the effects
of degradation, and establishing technical bases for rules limiting
the effects of aging. In particular, ORNL researchers have
addressed concrete structures, pressure vessels, and engineered
safety systems components.
CONCRETE STRUCTURES STUDIED AT ORNL
The containment building and the basemat, on which the reactor
sits, are the two most important concrete structures in a nuclear
power plant in terms of safety. Material and structural degradation
resulting from aging and environmental influences must be
understood and managed to ensure the integrity of these components
and the associated defense against release of radiation to the
environment.
ORNL addresses concrete structures in the NRC-sponsored Structural
Aging Program, led by Dan Naus and Barry Oland, both of the
Engineering Technology Division. The program's chief goal is to
establish technical bases for regulatory criteria, which will
identify for license reviewers and licensees potential structural
safety issues and provide acceptance criteria.
ORNL researchers are studying the aging and environmental
influences on the properties of both concrete and steel-reinforcing
materials, potential degradation mechanisms, structural inspection
and monitoring techniques, repair methods, and procedures for
structural evaluations, especially of the containment building and
basemat.
The ORNL group has compiled extensive information from domestic and
international sources for nuclear and non-nuclear civil structures
and research activities. The information is being compiled into
comprehensive data bases on long-term material properties, concrete
aging mechanisms, inspection, repair, and structural integrity
experience. One important part of this effort is ORNL's development
of a Structural Materials Information Center, which makes these
data bases available in handbook and electronic formats. It will be
an important resource for assessing concrete structures during
license renewal evaluation.
Other developments include (1) an aging assessment methodology,
which can be used to rank concrete structures in terms of safety
significance and resistance to environmental damage, and (2) a
methodology to assess the current condition and predict the
lifetime reliability of concrete structures. Other guidelines on
inspection and repair will be issued by ORNL.
REACTOR PRESSURE VESSEL RESEARCH AT ORNL
Reactor pressure vessel (RPV) research continues to receive high
priority from the NRC because of its importance to safe plant
operation and other factors. ORNL, through the Heavy-Section Steel
Technology (HSST) Program, has been the lead laboratory for this
research for more than 25 years. The principal factors of concern
are that vessels are subjected to active aging mechanisms, such as
increasing embrittlement caused by neutron radiation; can be
potentially exposed to stresses, such as sudden decreases in
temperatures and pressures, under accident conditions; contain
fabrication flaws; and cannot be easily replaced to extend the
plant's life.
Before the creation of the NRC, the Atomic Energy Commission (AEC)
initiated research to obtain data and develop methods to ensure
that adequate margins of safety existed for the thick-wall pressure
vessels in commercial nuclear power plants. In 1966 ORNL was chosen
to be the lead laboratory for that research and has provided
continuous support to the AEC and NRC. The work has produced many
of the methods that are incorporated in codes and standards used by
the nuclear industry and the NRC, such as the reference fracture
toughness procedures in the American Society of Mechanical
Engineers (ASME) code and the NRC Regulatory Guide 1.154, which
gives guidelines for evaluating RPV integrity under pressurized
thermal-shock (PTS) scenarios in which a hot, pressurized,
irradiated vessel is suddenly chilled by the introduction of
cooling water.
Current work on RPVs at ORNL consists of three programs: the HSST
Program, managed by Bill Pennell of the Engineering Technology
Division; the Heavy-Section Steel Irradiation (HSSI) Program,
managed by Bill Corwin of the Metals and Ceramics Division; and the
Surveillance Data Bases, Analysis, and Standardization Program,
managed by Frank Kam of the Computing and Telecommunications
Division. ORNL's leadership in these areas is recognized worldwide,
and the ORNL staff maintain strong relationships with their
counterparts in many countries and at the International Atomic
Energy Agency.
The first sidebar to this article addresses the HSST Program and
some of the issues it examines. The HSST goal is to provide the NRC
with the best available technology to ensure margins of safety
against fracture of the RPV under hypothetical scenarios. The PTS
scenario has gained the most attention because it combines the
elements of fracture-prevention analysis, material properties,
fabrication factors, and aging effects resulting from irradiation.
Mechanisms responsible for age-related degradation of pressure
vessels include prolonged radiation exposure, sustained pressure
and thermal loadings, and transient pressure and thermal loadings.
Understanding these mechanisms and the materials making up vessels
is the key to ensuring that a vessel will not rupture under
prescribed conditions. In this area, ORNL has developed a series of
advanced data bases and computer codes for use in making fracture
predictions.
ORNL researchers are continuing to conduct advanced studies of the
fracture characteristics of the materials, fracture behavior of
structures, and analytical methods applicable to complex
structures. Careful performance of many large-scale fracture
experiments and their supporting analyses have defined the validity
of and margins for the current RPV assessment methods. These
efforts now include a focus on conditions that would exist beyond
the initial license period.
RPV life is based not exclusively on time, but rather mostly on the
embrittlement of the RPV steels, which depends on the amount, rate,
and temperature of neutron radiation. In other words, the
life-limiting conditions for a vessel depend on the neutron
exposure and the conditions under which it is accumulated. Of
course, during the approved period of extended operation, the
accumulated radiation exposure will be greater than for the plant's
initial 40 years. Current limits on allowable embrittlement are
expected to apply in principle during renewal periods, and the
current research is aimed at verifying that these limits will not
be exceeded during a 20-year extension. With respect to monitoring
RPV aging, one key need is in-reactor surveillance programs that
give accurate data on the neutron exposure and vessel embrittlement
during operation. Frank Kam and his project team are working
closely with the NRC to improve surveillance programs by combining
the best available dosimetry techniques with the periodic testing
of tensile and fracture specimens made of the RPV steel and weld
materials that are exposed to actual reactor operation. They also
maintain for the NRC a carefully documented national surveillance
data base on which regulatory guidance for embrittlement
assessments is based.
If the calculated embrittlement for the vessel in a specific plant
reaches the allowable limit, regulations permit the owner to
thermally anneal the vessel to remove a portion of the accumulated
embrittlement. In work for the HSSI program, Bill Corwin and Randy
Nanstad are verifying the correct times and temperatures for
annealing to achieve acceptable levels of recovery for vessel steel
and welds that may be more than 20 cm (8 in.) thick. Their
experiments on welds removed from the vessel of the cancelled
Midland Nuclear Plant Unit No. 2 will determine the degree of
recovery from annealing and reembrittlement rates during subsequent
radiations for prototypically thick welds. The HSSI Program staff
are carefully combining these results with those from earlier and
smaller specimen studies to provide the NRC with the best available
basis for establishing annealing standards. The NRC is completing
development of a regulatory guide to permit utilities to anneal
their RPVs and to continue operation after annealing even under
license renewal.
Radiation-induced embrittlement is not the only age-related
degradation mechanism affecting RPVs. Material damage caused by
thermal and mechanical stress cycles over time results in
degradation that must be factored into the allowable life
assessments. By knowing the plant's operating history, the owner
can compute accumulated damage resulting from thermal aging,
fatigue, and other factors to determine the remaining allowable
life for the vessel, as defined by the code criteria in the CLB.
Thus, accurate records of a plant's operating and maintenance
histories are recognized by utilities as important for license
renewal. Nanstad's group also continues to study the aging of these
steels under long-term thermal exposure conditions.
ENGINEERED SAFETY SYSTEMS COMPONENTS STUDIED AT ORNL
ORNL researchers are studying aging of nuclear reactor plant
components, such as valves, pumps, steam generators, and vessel
internals. Because the condition of these aging SSCs must be
considered throughout license renewal periods, the research is
focused on (1) developing guidelines for assessing the condition of
components at the end of the current license period, (2) devising
methods for monitoring age-related degradation during the license
renewal period, and (3) where possible, identifying approaches to
mitigate age-related degradation effects.
For the past eight years, ORNL has been a lead laboratory for the
NRC in this area and has made some major breakthroughs, especially
in developing techniques for nonintrusive monitoring of valves.
This work has led to patented techniques and substantial technology
transfer to industry. Don Casada, the manager of this research,
describes these developments in the second sidebar to this article
on p. 95. The research at ORNL is part of the NRC's overall Nuclear
Plant Aging Research Program.
ENVIRONMENTAL EFFECTS
A proposed NRC rule will cover environmental effects that should be
addressed during the process of license renewal. Over the past few
years, a multidisciplinary team led by Rich McLean, Lance McCold,
and Johnnie Cannon, all of the Energy Division, has intensively
studied the potential environmental impacts of extended nuclear
plant operation (see the interview on p. 114). The study identified
more than 100 issues, made detailed assessments of potential
environmental impacts, and examined the relationships of the issues
to the National Environmental Policy Act (NEPA). The final goal was
to identify the many issues that can be treated on a generic basis
and the remaining few that must be handled on a plant-by-plant
basis.
The results of the study were published for public comment in late
1991. The public-comment period recently ended, and ORNL is helping
the NRC formulate responses. After the NRC makes its final
responses, this part of the license extension rule will be
complete. The NRC will document the final rule as a modification to
Part 51 of Element 10 of the U.S. Code of Federal Regulations (10
CFR 51).
IMPLEMENTING THE LICENSE RENEWAL PROCESS
Upon completion of the generic environmental impact statement for
extending nuclear plant operation, all the regulatory instruments
will be in place to allow applications for license renewal. The
first license renewal application was expected to be for the
pressurized-water reactor of the Yankee Atomic Electric Company's
(YAEC) Yankee Nuclear Power Station in Rowe, Massachusetts (see
photo on p. 88). Commonly called Yankee Rowe, it was the oldest
operating nuclear power plant in the United States, and its license
was scheduled to expire in 2000. However, YAEC recently decided to
retire this plant because of economic considerations, including the
cost of the extensive efforts that would have been required to
qualify its embrittled reactor vessel for continued service.
The first application to renew a license for a boiling-water
reactor is expected to come from the Northern States Power Company
(NSPC) for its Monticello Nuclear Power Plant, whose license is
scheduled to expire in 2010. Both of these plants have in recent
years been the focus of advanced studies by DOE and the Electric
Power Research Institute to assess the issues of aging and
potential procedures for license renewal. The NSPC is expected to
submit its application for license renewal in early 1993.
The lead time between submittal of a renewal application and
expiration of the initial license period is expected to be about 15
years. About 5 years will be required to obtain approval for all
aspects of the application, and the other 10 years are allowed for
the plant owner to provide an alternative source of electricity
should the application be denied or if the owner decides not to
extend the plant's operation. The license renewal is to become
active when the approval is granted and to extend through the
renewal period. For example, if the renewal is approved 10 years
before expiration of the initial 40-year license, the renewal will
cover the remaining 10 years plus the approved renewal period.
In conclusion, meeting the growth in demand for electricity in the
next several decades in the United States can be easier and less
costly if the operating licenses of existing nuclear power plants
are renewed for another 20 years. ORNL has played an important role
by providing the NRC with technical expertise, data, and analyses
on which to base regulatory criteria and guidelines for license
renewal. The development of the regulatory procedures is now
essentially complete. It is expected that ORNL will continue to
contribute to the development of nuclear power technologies,
including the advanced reactors that will be certified by the NRC
for the next generation of nuclear power plants.
Claude E. Pugh
(keywords: nuclear power, nuclear power plants)
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Date Posted: 2/7/94 (ktb)