[Federal Register Volume 64, Number 190 (Friday, October 1, 1999)]
[Proposed Rules]
[Pages 53270-53275]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-25582]
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Proposed Rules
Federal Register
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This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 64, No. 190 / Friday, October 1, 1999 /
Proposed Rules
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150--AG26
Emergency Core Cooling System Evaluation Models
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend
its regulations to allow holders of operating licenses for nuclear
power plants to reduce the assumed reactor power level used in
evaluations of emergency core cooling system (ECCS) performance. Under
the proposed rule, licensees would be given the option to apply a
reduced margin for ECCS evaluation or to maintain the value of reactor
power currently mandated in the regulation. This action would allow
interested licensees to pursue small, but cost-beneficial, power
uprates and would reduce unnecessary regulatory burden without
compromising the margin of safety of the facility.
DATES: The comment period expires on December 15, 1999. Comments
received after this date will be considered if it is practical to do so
but the NRC is able to assure consideration only for comments received
on or before this date.
ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, D.C. 20555-0001, Attention: Rulemakings and
Adjudications Staff, Mail Stop O-16C1.
Deliver written comments to: One White Flint North, 11555 Rockville
Pike, Rockville, Maryland between 7:30 a.m. and 4:15 p.m. on Federal
workdays.
Documents related to this rulemaking may be examined at the NRC
Public Document Room, 2120 L Street, NW. (Lower Level), Washington,
D.C. Documents also may be viewed and downloaded electronically via the
interactive rulemaking Web site established by NRC for this rulemaking
(see the discussion under Electronic Access in the Supplementary
Information section). Obtain single copies of the environmental
assessment and the regulatory analysis from the NRC contact given
below.
FOR FURTHER INFORMATION CONTACT: Mr. Joseph E. Donoghue, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, D.C. 20555-0001; telephone: 301-415-1131; or by Internet
electronic mail to jed1@nrc.gov.
SUPPLEMENTARY INFORMATION:
Background
A holder of an operating license (i.e., the licensee) for a light-
water power reactor is required by regulations issued by the NRC to
submit a safety analysis report that contains an evaluation of
emergency core cooling system (ECCS) performance under loss-of-coolant
accident (LOCA) conditions. 10 CFR 50.46, ``Acceptance criteria for
emergency core cooling systems for light-water nuclear power
reactors,'' requires that ECCS performance under LOCA conditions be
evaluated and that the estimated performance satisfy certain criteria.
Licensees may conduct an analysis that ``realistically describes the
behavior of the reactor system during a LOCA'' (often termed a ``best-
estimate analysis''), or they may develop a model that conforms with
the requirements of Appendix K to 10 CFR Part 50. Most ECCS evaluations
are based on Appendix K requirements. The opening sentence of Appendix
K establishes the requirement to conduct ECCS analyses at a specified
power level: ``It shall be assumed that the reactor has been operating
continuously at a power level at least 1.02 times the licensed power
level (to allow for such uncertainties as instrumentation error).''
Licensees have proposed using instrumentation that would reduce the
uncertainties associated with measurement of reactor power when
compared with existing methods of power measurement. This would justify
a reduced margin between the licensed power level and the power level
assumed for ECCS evaluations. The proposed rule would revise this
provision in Appendix K, thereby allowing licensees the option of using
a value lower than 102 percent of licensed power in their ECCS analyses
where justified.
Several licensees have expressed interest in using updated
feedwater flow measurement technology discussed later in ``Calorimetric
Uncertainty and Feedwater Flow Measurement'' as a basis for seeking
exemptions from the Appendix K power level requirement and to implement
power uprates. One licensee, Texas Utilities Electric Company (TUE),
has obtained an exemption from the Appendix K requirement for Comanche
Peak Units 1 and 2 and is pursuing an increase in licensed power based,
in part, on more accurate feedwater flow measurement capability. The
prospect of additional exemption requests from other licensees provides
the impetus for the proposed rule.
The objective of this rulemaking is to reduce an unnecessarily
burdensome regulatory requirement. Appendix K was issued to ensure an
adequate performance margin of the ECCS in the event a design-basis
LOCA were to occur. The margin is provided by conservative features and
requirements of the evaluation models and by the ECCS performance
criteria. The existing regulation does not require that the power
measurement uncertainty be demonstrated, but rather mandates a 2-
percent margin to account for uncertainties, including those expected
to be involved with measuring reactor power. By allowing licensees to
justify a smaller margin for power measurement uncertainty, the
proposed rule does not violate the underlying purpose of Appendix K.
The intent of Appendix K, to ensure sufficient margin to ECCS
performance in the event of a LOCA, would still be met because of the
substantial conservatism of other Appendix K requirements. The proposed
rule would not significantly affect plant risk, as discussed in the
section entitled, ``ECCS Evaluation Conservatism.''
Another objective is to avoid unnecessary exemption requests. As
discussed above, a licensee has obtained an exemption from the 2-
percent margin requirement in 10 CFR Part 50, Appendix K. It is likely
that additional exemption requests will be submitted. Revising the rule
to remove the need for licensees to obtain exemptions is considered by
the NRC to be a prudent regulatory action.
[[Page 53271]]
If adopted, the proposed rule would give licensees the option of
applying a reduced margin between the licensed power level and the
assumed power level for ECCS evaluation, or maintaining the current
margin of 2-percent power. As discussed in the section entitled ``ECCS
Evaluation Conservatism,'' the NRC has concluded that the 2 percent
power margin requirement in the existing rule appears to be based
solely on considerations associated with power measurement extant at
the time of the original ECCS rulemaking. If licensees can show that
the uncertainties associated with power measurement instrumentation
errors are less than 2 percent, thereby justifying a smaller margin,
then the current rule unnecessarily restricts operation.
Making this change to the rule would give licensees the opportunity
to use a reduced margin if they determine that there is a sufficient
benefit. Licensees could apply the margin to gain benefits from
operation at higher power, or the margin could be used to relax ECCS-
related technical specifications (e.g., pump flows). Another potential
benefit would be in modifying fuel management strategies (e.g.,
possibly by altering core power peaking factors). However, the proposed
rule by itself does not allow increases in licensed power levels.
Because licensed power level for a plant is a technical specification
limit, proposals to raise the licensed power level must be reviewed and
approved under the license amendment process. The license amendment
request should include a justification of the reduced power measurement
uncertainty and the basis for the modified ECCS analysis, including the
justification for reduced power measurement uncertainty, should then be
included in documentation supporting the ECCS analysis (see Section-by-
Section Analysis).
In the short term, the NRC intends to grant exemptions to the
assumed power level provision of Appendix K for properly supported
exemption requests. In addition to satisfying the provisions of 10 CFR
50.12, properly supported exemption requests are expected to quantify
the uncertainties associated with measuring reactor thermal power that
are associated with the current 2-percent power margin.
In the longer term, the NRC intends to review the affected safety
analysis guidance and will evaluate the impact of the proposed rule on
those safety analyses. Further, the NRC is considering the need for
specific guidance to help licensees appropriately account for power
measurement uncertainty in safety analyses. However, the NRC expects
that power uprate amendment requests based on the proposed rule will
address the suitability of non-LOCA analyses for operation at proposed
higher power levels.
In addition to comments on the proposed rule, the NRC is seeking
comments on the specific issues set forth below under ``Issues for
Public Comment.''
Conservatisms in Appendix K ECCS Evaluation Model
Appendix K defines conservative analysis assumptions for ECCS
performance evaluations during design-basis LOCAs. Large safety margins
are provided by conservatively selecting the ECCS performance criteria
as well as conservatively establishing ECCS calculational requirements.
The major analytical parameters and assumptions that contribute to the
conservatisms in Appendix K are set forth in Sections A through D of
the rule: (A) ``Sources of Heat During the LOCA'' (the 102-percent
power provision is a key factor), (B) ``Swelling and Rupture of the
Cladding and Fuel Rod Thermal Parameters,'' (C) ``Blowdown Phenomena,''
and (D) ``Post-blowdown Phenomena: Heat Removal by ECCS.'' In each of
these areas, several assumptions are typically used to ensure
substantial conservatism in the analysis results. For instance: under
``Sources of Heat During the LOCA,'' decay heat is modeled on the basis
of an American Nuclear Society standard with an added 20-percent
penalty, and the power distribution shape and peaking factors expected
during the operating cycle are chosen to yield the most conservative
results. In ``Blowdown Phenomena,'' the rule requires use of the Moody
model and the discharge coefficient that yields the highest peak
cladding temperature. ``Post'Blowdown Phenomena; Heat Removal by the
ECCS,'' requires that the analysis assume the most damaging single
failure of ECCS equipment.
One of several conservative requirements in Section A is to assume
that the reactor is operating at 102 percent power when the LOCA occurs
``to allow for such uncertainties as instrumentation error. . . .''
(Appendix K, Section I.A., first sentence, emphasis added). The phrase,
``such as,'' suggests that the two percent power margin was intended to
address uncertainties related to heat source considerations beyond
instrument measurement uncertainties. However, the basis for the
required assumption of 102 percent power (2 percent power margin) does
not appear to be contained in the rulemaking record for the ECCS rules,
10 CFR 50.46 and Appendix K. These rules were adopted in 1974 (39 FR
1001, January 4, 1974), and were preceded by a formal rulemaking
hearing which ultimately resulted in a Commission decision on the
proposed rulemaking, CLI-73-39, 6 AEC 1085 (December 28, 1973). Neither
the statement of considerations (SOC) for the final rule nor the
Commission decision appear to provide specific basis for the required
assumption of 102 percent power.
The SOC for the final 1974 rule discusses the 102 percent power
assumption in general terms, and does not mention instrumentation
uncertainty:
The Commission believes that the implementation of the new
regulations will ensure an adequate margin of performance of the
ECCS should a design basis LOCA ever occur. This margin is provided
by conservative features of the evaluation models and by the
criteria themselves. Some of the major points that contribute to the
conservative nature of the evaluations and the criteria are as
follows:
(1) Stored heat. The assumption of 102 percent of maximum power,
highest allowed peaking factor, and highest estimated thermal
resistance between the UO2 and the cladding provides a
calculated stored heat that is possible but unlikely to occur at the
time of a hypothetical accident. While not necessarily a margin over
the extreme condition, it represents at least an assumption that an
accident happens at a time which is not typical.
39 FR at 1002 (first column).1 Thus, while the pre-accident
power level assumption is connected with the modeling of the rate of
heat generation after the LOCA occurs, a clear basis for the 102
percent assumed power level requirement is not provided, nor does the
SOC explain whether there are other uncertainties besides
instrumentation uncertainties for which the 102 percent assumed power
level is intended to compensate.
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\1\ This statement in the SOC was taken unchanged from Section I
of the Commission's ECCS decision. See CLI-73-39, 6 AEC 1085, 1093-
94 (December 28, 1973).
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The Commission's decision in the ECCS rulemaking hearing also does
not explain whether the 102 percent assumed power level was intended to
address uncertainties other than instrumentation uncertainties. Section
I of the Commission decision was the basis for the SOC discussion on
the 102 percent assumed power level (See 6 AEC at 1093-94). Section
III. A. of the Commission's decision, ``Required and Acceptable
Features of the Evaluation Model,'' does not offer a detailed technical
the basis for the power level chosen, but instead uses the language
ultimately adopted in the final Appendix K rule:
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For the heat sources listed in paragraphs 1 to 4 below it shall
be assumed that the reactor has been operating continuously at a
power level at least 1.02 times the licensed power level (to allow
for such uncertainties as instrumentation error), with the maximum
peaking factor allowed by the technical specifications.
6 AEC at 1100. Thus, the Commission's decision does not shed further
light on the basis for the 102 percent assumed power level, nor whether
the Commission had in mind uncertainties other than those associated
with the instrumentation for measurement of power level.
NRC review of the ECCS rulemaking hearing record did not disclose
presentations relating to quantification of power measurement
uncertainties, or the magnitude of other uncertainties that the 102
percent assumed power level may have been intended to address. The
Commission decision (CLI-73-39, 6 AEC 1085, December 28, 1973) cited
three documents in the rulemaking hearing record. The first, cited in
the Commission decision as Exhibit 1113, was ``Supplemental Testimony
of the AEC Regulatory Staff on the Interim Acceptance Criteria for
Emergency Core Cooling Systems for Light-Water Cooled Power Reactors,''
(filed October 26, 1972). In Section 10 of the document, stored energy
in the fuel was considered, specifically the expected power
distributions in fuel rods. The 102-percent power analysis requirement
is not discussed. The second item, cited in the Commission decision as
Exhibit 1137 was ``Redirect and Rebuttal Testimony of Dr. Donald H. Roy
on Behalf of Babcock & Wilcox,'' (October 26, 1972) in which the
characteristic of the decay heat release following reactor shutdown was
discussed. In this document, the 102-percent assumption is associated
with the predicted decay heat generation rate. The over-power condition
is associated with a ``design-basis maneuvering operation,'' but the
basis for the value of power chosen for the analysis (i.e., 102
percent) is not disclosed. Finally, in the ``Concluding Statement of
Position of the Regulatory Staff--Public Rulemaking Hearing on:
Acceptance Criteria for Emergency Core Cooling Systems for Light-Water
Cooled Nuclear Power Reactors,'' April 16, 1973 (the Concluding
Statement), the power level assumption is included as part of the
proposed rule itself. The proposed rule language clearly states that
the power level assumption is to ``allow for instrumentation error.''
The term ``such as'' does not appear here. It is unclear when or why
the proposed language in this regard was changed to its current form.
The power level assumption is mentioned again in the Concluding
Statement indirectly in association with power level changes before the
LOCA and the effect on decay heat generation. But it is discussed most
directly with regard to initial stored energy in the fuel. In the
discussion on stored energy, the 102-percent assumption is attributed
to ``uncertainties inherent in the measurement of the operating power
level of the core,'' (page 144 of the Concluding Statement). Reasons
for choosing 102-percent as the value are not discussed.
When Appendix K was first issued, as is the case today, the thermal
power generated by a nuclear power plant was determined by steam plant
calorimetry, which is the process of performing a heat balance around
the nuclear steam supply system (called a calorimetric). The heat
balance depends upon measurement of several plant parameters, including
flow rates and fluid temperatures. The differential pressure across a
venturi installed in the feedwater flow path is a key element in the
calorimetric measurement. Licensees have proposed using instrumentation
other than a venturi-based system to obtain feedwater flow rate for
calorimetrics. The lower uncertainty associated with the new
instrumentation is information that was apparently not available during
the original Appendix K rulemaking.
In view of the regulatory history for Appendix K, the Commission
now believes that the 2-percent margin embodied in the requirement for
a 102-percent assumed power level in Appendix K was based solely on
uncertainties associated with the measurement of reactor power level.
Proposed Reduction in 102 Percent Assumed Power Level
The Commission believes that other requirements of Appendix K
modeling include substantial conservatisms of much greater magnitude
than the 2 percent margin embodied in the requirement for a 102 percent
assumed power level. This point was discussed in ``Conservatisms in
Appendix K ECCS Evaluation Model,'' above.
The Commission is also aware of new information gained since the
1974 rulemaking which shows that the Appendix K model contains
substantial conservatisms. Evidence from experiments designed to
simulate LOCA phenomena suggest that these conservatisms added hundreds
of degrees Fahrenheit to the prediction of peak fuel cladding
temperature than would actually occur during a LOCA. The significant
conservatism was necessary when the rule was written because of a lack
of experimental evidence at that time with respect to the relative
effects of analysis input parameters, including pre-accident power
level. Since that time, there has been substantial additional research
on LOCA. NUREG-1230, ``Compendium of ECCS Research for Realistic LOCA
Analysis,'' December 1988, contains the technical basis for improved
understanding of LOCA progression and ECCS evaluation gained after the
ECCS rule was issued. The NUREG includes a discussion of the basis for
uncertainties in detailed fuel bundle power calculations as part of the
consideration of overall calculational uncertainty inherent in best-
estimate evaluations. Chapters 7 and 8 of the NUREG include
consideration of the changes in licensed power level that could result
from application of best-estimate evaluation methods. The discussion
includes an estimated sensitivity of predicted peak clad temperature
associated with changes in pre-accident power level. From that
estimate, the NRC expects peak cladding temperature changes of
approximately 15 deg.F to result from 1-percent changes in plant power
level that could result from the proposed rule.
In view of: (i) Substantial conservatisms embodied in the Appendix
K requirements for ECCS evaluations, (ii) new information developed
since the 1974 rulemaking which shows additional conservatism in the
Appendix K modeling requirements beyond that understood by the
Commission when it adopted the 1974 rule, and (iii) the relative
insensitivity of the calculated clad temperatures to assumed power
level, the Commission concludes that it is acceptable to allow a
reduction in the currently-required 102 percent power level assumption
if justified by the actual power level measurement instrumentation
uncertainty. Accordingly, the Commission proposes to amend the Appendix
K requirement for an assumed 102 percent power level. The proposed rule
would allow a licensee to use an assumed power level of less than 102
percent (but not less than 100 percent), provided that the licensee has
determined that the uncertainties in the measurement of core power
level justifies the reduced margin.
Calorimetric Uncertainty and Feedwater Flow Measurement
The NRC staff has approved an exemption to the 102-percent power
level requirement for Comanche Peak Units 1 and 2. The basis for the
action is application of upgraded feedwater flow measurement technology
at the
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plant. As indicated, the prospect of additional licensees requesting
similar action has prompted the proposed rule. Other methods, systems,
or analyses could be used as the basis for demonstrating reduced power
measurement uncertainty.
In most nuclear power plants, operators obtain a continuous
indication of core thermal power from nuclear instruments, that provide
a measurement of neutron flux. The nuclear instruments must be
periodically calibrated to counteract the effects of changes in flux
pattern, fuel burnup, and instrument drift. Steam plant calorimetry,
which is the process of performing a heat balance around the nuclear
steam supply system (called a calorimetric), is used to determine core
thermal power and is the basis for the calibration. The differential
pressure across a venturi installed in the feedwater flow path is a key
element in the calorimetric measurement. Some plants use this
calorimetric value directly to indicate thermal power; the nuclear
instruments are used as anticipatory indicators for transients and for
reactivity adjustments made with the control rods.
The system in use at Comanche Peak Units 1 and 2 is the Leading
Edge Flowmeter (LEFM), manufactured by Caldon, Inc. The LEFM system is
an ultrasonic flow meter that measures the transit times of pulses
traveling along parallel acoustic paths through the flowing fluid. LEFM
technology has been employed in non-nuclear applications, such as
petroleum, chemical, and hydroelectric plants for several years. This
operating experience will provide reliability data, supplementing data
from nuclear applications. Additional information on the Comanche Peak
Appendix K exemption and on the Caldon, Inc. LEFM system appears in
safety evaluations issued by the NRC staff on March 8, 1999, and May 6,
1999.
ABB Combustion Engineering has expressed interest in the proposed
rule because its flow-measuring system, known as Crossflow (which is
also an ultrasonic flow-measuring device), is expected to be part of a
licensee exemption request in the near future.
Issues for Public Comment
The NRC is seeking comments from the public on the following issues
related to this proposed rule:
1. The current rule states that the required 2-percent analysis
margin is to account for ``such uncertainties as instrumentation error.
. . .'' (emphasis added). This suggests that the 2-percent margin was
intended to account for other sources of uncertainty in addition to
instrumentation error. However, explicit documentation of the basis for
the value of the margin does not appear to be contained in the
rulemaking record for the original 1974 ECCS rulemaking. The Commission
is interested in whether there are other sources of uncertainty,
relevant to sources of heat following a LOCA, that should be considered
when licensees seek to reduce the margin in the Appendix K requirement
for assumed power. If other contributors are suggested, a clear
technical justification should accompany the suggestion.
2. Are there rulemaking alternatives to this proposed rule that
were not considered in the regulatory analysis for this proposed rule?
3. What criteria should be used for determining whether a proposed
reduction in the 2 percent power margin has been justified, based upon
a determination of instrumentation error? For example, should a
demonstrated instrumentation error of 1 percent in power level be
presumptive of an acceptable reduction in assumed power margin of 1
percent?
4. How should the proposed rule address cases in which licensees
determine that power measurement instrument error is greater than 2
percent?
Section-by-Section Analysis
Appendix K to Part 50--ECCS Evaluation Models (I)(A)--Sources of heat
during the LOCA
This section would be amended by removing words from the first
sentence in the section to specifically associate the power level
requirement with instrumentation error, and by adding a sentence
immediately following the first sentence in the section. The new
sentence indicates that licensees may assume a power level lower than
102 percent, but not less than 100 percent, provided that the proposed
lower alternative value can be shown to account for core thermal power
measurement instrumentation uncertainty.
Appendix K, Part II (1)(a) requires that the values of analysis
parameters or their basis be sufficiently documented to allow NRC
review. The requirement applies to all analysis input parameters,
including those related to other plant instrumentation, such as
temperature and pressure. Changes to other inputs are documented in the
same manner as the power measurement uncertainty would be documented
under the proposed rule. NRC review and approval is not necessarily
needed to change a parameter in an approved ECCS evaluation model.
Estimated changes in ECCS performance due to revised analysis inputs
are reported under Sec. 50.46 (a)(3), at least annually. As discussed
in the Statement of Considerations for Appendix K (53 FR 36001,
September 16, 1988), the annual reports keep NRC apprised of changes.
This should ensure that the NRC staff can judge a licensee's assessment
of the significance of changes and maintain cognizance of modifications
made to NRC-approved evaluation models. The licensee must include
revised parameters and other changes in the ECCS evaluation as required
by Sec. 50.46 (a)(3) when a single change or an accumulation of changes
is expected to affect peak cladding temperature by 50 deg.F or more.
The basis for the revised analysis parameter (i.e., the assumed power
level) should be included in documentation of the evaluation model, as
required by Appendix K, Part II (1)(a).
In most cases, the NRC expects that the analysis supporting the
power measurement uncertainty, as well as the description of the
relevant instrumentation and associated plant-specific parameters
involved in the uncertainty analysis, would be submitted for NRC review
and approval before being used. These requests are expected because
most licensees have adopted Generic Letter 88-16, ``Removal of Cycle-
Specific Parameter Limits from Technical Specifications.'' The generic
letter provided guidance for licensees to transfer cycle-specific
parameters from their technical specifications to a Core Operating
Limits Report (COLR). Licensees following the generic letter guidance
added an administrative requirement to their technical specifications
that specifically identifies NRC-reviewed and approved methods used to
determine core operating limits (e.g., topical reports). Because a
number of core operating limits are based on LOCA analysis results,
ECCS evaluation methods are included in the technical specification
list. Therefore, most licensees opting to use the relaxation in the
proposed rule would need to revise technical specifications to include
a reference to an NRC-approved topical report that includes the
uncertainty analysis justifying reduced power measurement uncertainty.
An additional technical specification consideration for licensees
pursuing changes based on the proposed rule could involve nuclear
instruments (NI) requirements. Existing plant technical specifications
include surveillance requirements to calibrate the power range NIs
based on the calorimetric measuring reactor thermal power. The
[[Page 53274]]
NIs provide the indication of reactor power used as an input for safety
systems. Licensees obtaining the relaxation offered in the proposed
rule are expected to change some operating parameter of the plant,
whether it be power level, required ECCS flow, etc. By incorporating
the justification of reduced uncertainty in power measurement in the
basis for their ECCS analysis, licensees would be placing a condition
on an input to the calorimetric. The NI calibration required by the
plant licensee would then be based on a calorimetric assuming the
reduced power measurement uncertainty. If, for some reason, during the
course of plant operation the reduced uncertainty did not apply (e.g.,
the new feedwater flow meter became inoperable), the calorimetric would
no longer be a valid source of calibration for the NIs. Licensees would
need to take action to maintain compliance with their technical
specification, for example, by using an alternate input to the
calorimetric. The power measurement uncertainties associated with the
alternate input would then apply and the plant would need to adjust its
operating condition (possibly lower its operating power level) to
satisfy the proposed rule and to maintain the validity of applicable
safety analyses.
Referenced Documents
Copies of GL-88-16 and CLI-73-39 are available for inspection and
copying for a fee at the NRC Public Document Room, 2120 L Street, NW.
(Lower Level), Washington, D.C.
Electronic Access
You may also submit comments via the NRC's interactive rulemaking
Web site, ``Rulemaking Forum,'' through the NRC home page (http://
ruleforum.llnl.gov). This site enables people to transmit comments as
files (in any format, but WordPerfect version 6.1 is preferred), if
your Web browser supports that function. Information on the use of the
Rulemaking Forum is available on the Web site. For additional
assistance on the use of the interactive rulemaking site, contact Ms.
Carol Gallagher, telephone: 301-415-5905; or by Internet electronic
mail to cag@nrc.gov.
Plain Language
The Presidential memorandum dated June 1, 1998, entitled, ``Plain
Language in Government Writing,'' directed that the government's
writing be in plain language. This memorandum was published June 10,
1998 (63 FR 31883). In complying with this directive, editorial changes
have been made in this proposed amendment to improve readability of the
existing language of the provisions being revised. These types of
changes are not discussed further in this document. The NRC requests
comment on the proposed rule specifically with respect to the clarity
and effectiveness of the language used. Comments should be sent to the
address listed under the ADDRESSES caption of the preamble.
Voluntary Consensus Standards
The National Technology Transfer Act of 1995, Pub. L. 104-113,
requires that Federal agencies use technical standards that are
developed or adopted by voluntary consensus standards bodies unless the
use of such a standard is inconsistent with applicable law or otherwise
impractical. In this proposed rule, the NRC is proposing to provide
holders of operating licenses for nuclear power plants with the option
of reducing the assumed reactor power level used in ECCS evaluations.
This proposed action constitutes a modification to an existing
government-unique standard, 10 CFR part 50, appendix K issued by the
NRC on January 4, 1974. The NRC is not aware of any voluntary consensus
standard that could be adopted instead of the proposed government-
unique standard. The NRC will consider using a voluntary consensus
standard if an appropriate standard is identified. If a voluntary
consensus standard is identified for consideration, the submittal must
explain how the voluntary consensus standard is comparable and why it
should be used instead of the proposed government-unique standard.
Finding of No Significant Environmental Impact: Availability
The NRC has determined under the National Environmental Policy Act
of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR
Part 51, that this regulation, if adopted, would not be a major Federal
action significantly affecting the quality of the human environment
and, therefore, an environmental impact statement is not required.
The proposed action is likely to result in relatively small changes
to ECCS analyses or to the licensed power of nuclear reactor
facilities. The NRC staff expects that no significant environmental
impact would result from the proposed rule, because licensee actions
based on the proposed rule would not significantly increase the
probability or consequences of accidents; no changes would be made in
the types of any effluents that may be released off site; and there
would be no significant increase in occupational or public radiation
exposure. Therefore, there are no significant radiological
environmental impacts associated with the proposed action. The proposed
action does not involve non-radiological plant effluents and has no
other environmental impact. Therefore, there are no significant non-
radiological environmental impacts associated with the proposed action.
The determination of the environmental assessment is that there
would be no significant offsite impact on the public from this action.
However, the general public should note that the NRC welcomes public
participation. Also, the NRC has committed itself to complying in all
its actions with Executive Order (E.O.) 12898, ``Federal Actions To
Address Environmental Justice in Minority Populations and Low-Income
Populations,'' dated February 11, 1994. The NRC has determined that
there are no disproportionately high and adverse impacts on minority
and low-income populations. In the letter and spirit of E.O. 12898, the
NRC is requesting public comments on any environmental justice
considerations or questions that the public thinks may be related to
this proposed rule, but that somehow were not addressed. The NRC uses
the following working definition of environmental justice:
Environmental justice means the fair treatment and meaningful
involvement of all people, regardless of race, ethnicity, culture,
income, or educational level with respect to the development,
implementation and enforcement of environmental laws, regulations, and
policies. Comments on any aspect of the environmental assessment,
including environmental justice, may be submitted to the NRC as
indicated under the ADDRESSES heading.
The draft environmental assessment is available for inspection at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, D.C. Single copies of the environmental assessment are
available from Mr. Joseph Donoghue, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-
0001, telephone: 301-415-1131, or by Internet electronic mail to
JED1@nrc.gov.
Paperwork Reduction Act Statement
This proposed rule increases the burden on licensees opting to use
a reduced power level assumption for ECCS analysis (i.e., below 102%)
to include the change in their annual
[[Page 53275]]
report required under 10 CFR 50.46 (a)(3)(ii). The public burden for
this information collection is estimated to average one-half hour per
response. Because the burden for this information collection is
insignificant, Office of Management and Budget (OMB) clearance is not
required. Existing requirements were approved by the Office of
Management and Budget, approval number 3150-0011.
Public Protection Notification
If a means used to impose an information collection does not
display a currently valid OMB control number, the NRC may not conduct
or sponsor, and a person is not required to respond to, the information
collection.
Regulatory Analysis
The Commission has prepared a regulatory analysis on this
regulation. Interested persons may examine a copy of the regulatory
analysis at the NRC Public Document Room, 2120 L Street NW. (Lower
Level), Washington, D.C. Single copies of the analysis are available
from Mr. Joseph Donoghue, Office of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555-0001, telephone:
301-415-1131, or by Internet electronic mail to [email protected]
Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this proposed rule, if adopted,
would not have a significant economic impact on a substantial number of
small entities. This proposed rule would affect only the licensing and
operation of nuclear power plants. The companies that own these plants
do not fall within the definition of ``small entities'' found in the
Regulatory Flexibility Act or within the size standards established by
the NRC in 10 CFR 2.810.
Backfit Analysis
The NRC has determined that the backfit rule in 10 CFR 50.109 does
not apply to this proposed rule and that a backfit analysis is not
required for this proposed rule because the change does not involve any
provisions that would impose backfits as defined in 10 CFR
50.109(a)(1). The proposed rule would establish an alternative approach
for ECCS performance evaluations that may be voluntarily adopted by
licensees. Licensees may continue to comply with existing requirements
in Appendix K. The proposed rule does not impose a new requirement on
current licensees and therefore, does not constitute a backfit as
defined in 10 CFR 50.109(a)(1).
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
peactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
Accordingly, we propose to amend 10 CFR part 50 as follows:
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for Part 50 continues to read as follows:
Authority: Sections 102, 103, 104, 105, 161, 182, 183, 186, 189,
68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec.
234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135,
2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202,
206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842,
5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235), sec. 102, Pub.
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a, and Appendix Q also issued under sec. 102, Pub. L. 91-190,
83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
2. Appendix K to Part 50 is amended by revising the introductory
paragraph of I. A., ``Sources of heat during the LOCA,'' to read as
follows.
Appendix K to Part 50--ECCS Evaluation Models
I. Required and Acceptable Features of the Evaluation Models
A. Sources of heat during the LOCA. For the heat sources listed
in paragraphs I. A. 1 to 4 of this appendix it must be assumed that
the reactor has been operating continuously at a power level at
least 1.02 times the licensed power level (to allow for
instrumentation error), with the maximum peaking factor allowed by
the technical specifications. An assumed power level lower than the
level specified in this paragraph (but not less than the licensed
power level) may be used provided the proposed alternative value has
been demonstrated to account for uncertainties due to power level
instrumentation error. A range of power distribution shapes and
peaking factors representing power distributions that may occur over
the core lifetime must be studied. The selected combination of power
distribution shape and peaking factor should be the one that results
in the most severe calculated consequences for the spectrum of
postulated breaks and single failures that are analyzed.
* * * * *
Dated at Rockville, Maryland, this 27th day of September, 1999.
For the Nuclear Regulatory Commission.
Kenneth R. Hart,
Acting, Secretary of the Commission.
[FR Doc. 99-25582 Filed 9-30-99; 8:45 am]
BILLING CODE 7590-01-P
