[Federal Register Volume 59, Number 201 (Wednesday, October 19, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-25916]
[[Page Unknown]]
[Federal Register: October 19, 1994]
VOL. 59, NO. 201
Wednesday, October 19, 1994
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AE97
Shutdown and Low-Power Operations for Nuclear Power Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend
its regulations to require power reactor licensees to: Assure that
uncontrolled changes in reactivity, reactor coolant inventory, and loss
of subcooled state in the reactor coolant system when subcooled
conditions are normally being maintained, will not occur when the plant
is in either a shutdown or low power condition; assure that containment
integrity is maintained or can be reestablished in a timely manner as
needed to prevent releases in excess of the current limits in the
regulations when the plant is in either a shutdown or low power
condition; establish controls in technical specifications limiting
conditions for operation and surveillance requirements or plant
procedures required by technical specifications administrative controls
for equipment which the licensee identifies as necessary to perform
their safety function when the plant is in a shutdown or low power
condition; evaluate realistically the effect of fires stemming from
activities conducted during cold shutdown or refueling conditions,
determine whether such fires could realistically prevent accomplishment
of the normal decay heat removal capability, and if so, either provide
measures to prevent loss of normal decay heat removal or establish a
contingency plan that would ensure that an alternate decay heat removal
capability exists; and for licensees of PWRs only, provide
instrumentation for monitoring water level in the RCS during midloop
operation. The proposed amendments would provide substantial additional
protection to public health and safety from the risk of a core-melt
accident.
DATES: The comment period expires January 3, 1995. Comments received
after this date will be considered if it is practical to do so, but the
Commission 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, DC 20555-0001, ATTN: Docketing and Service
Branch.
Deliver comments to: 11555 Rockville Pike, Rockville, Maryland,
between 7:45 am and 4:15 pm Federal workdays.
Copies of comments received may be examined and copied for a fee at
the NRC Public Document Room, 2120 L Street, NW (Lower Level),
Washington, DC.
FOR FURTHER INFORMATION CONTACT: Gary M. Holahan, Director, Division of
Systems Safety and Analysis, Office of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:
(301) 504-2884.
SUPPLEMENTARY INFORMATION:
Background
Over the past several years, the Nuclear Regulatory Commission
(NRC) staff has become increasingly concerned about the safety of
operations during the shutdown of nuclear power reactors. The loss of
decay heat removal (DHR) during shutdown and refueling has been a
continuing problem. In 1980, DHR was lost at the Davis-Besse plant when
one residual heat removal (RHR) pump failed and the second pump was out
of service. After reviewing the Davis-Besse event and studying the
operating requirements that existed at the time of the event, the NRC
issued Bulletin 80-42 and Generic Letter (GL) 80-43 calling for new
technical specifications to ensure that one RHR system is operating and
a second is available (i.e., operable) for most shutdown conditions.
The Diablo Canyon event of April 10, 1987, highlighted the fact that
midloop operation was a particularly sensitive condition with respect
to operability of the residual heat removal pumps. In this event, the
reactor coolant system was overdrained during midloop operation. The
resulting low water level in the reactor vessel caused vortexing and
air entrainment and loss of both residual heat removal pumps. After
reviewing the event, the staff issued GL 88-17, recommending that
licensees address numerous generic deficiencies to improve the
reliability of the DHR capability. More recently, the incident
investigation team's report on the loss of AC power at the Vogtle plant
(NUREG-1410) emphasized the need for risk management of shutdown
operations. Furthermore, discussions with foreign regulatory
organizations (i.e., French and Swedish authorities) about their
evaluations regarding shutdown risk have reinforced previous NRC staff
findings that the core-damage probability (CDP) for shutdown operation
can be a fairly substantial fraction of the total CDP. Because of these
concerns regarding operational safety during shutdown, the NRC
conducted a careful, detailed evaluation of safety during shutdown and
low-power operations which is documented in NUREG-1449.
Objective
The NRC staff's comprehensive evaluation of shutdown and low-power
operations, documented in NUREG-1449, included observations and
inspections at a number of plants, analysis of operating experience,
deterministic safety analysis, and insights from probabilistic risk
assessments. It was observed that shutdown risks have been reduced at
many plants through improvements to outage programs. However, the
improvements have been unevenly and inconsistently applied across the
industry. From this evaluation, the NRC has concluded that public
health and safety have been adequately protected during the period that
plants have been in shutdown and low power conditions; but that
substantial safety improvements are possible and NRC requirements are
warranted for the following reasons:
(1) A regulatory requirement would set minimum standards for all
plants and would ensure that safety improvements already made by
industry will be applied consistently throughout the industry and will
not be eroded in the future.
(2) A regulatory requirement would further reduce risk by improving
safety in the areas of fire protection for all plants and midloop
operation for PWRs.
(3) Significant precursor events involving loss of DHR capability
continue to occur despite efforts to resolve the problem.
(4) Some controls, including regulatory controls, have been
significantly lacking and have in the past allowed plants to enter
circumstances that would likely challenge safety functions with minimal
mitigation equipment available and containment integrity not
established.
The NRC has identified possible regulatory actions to address these
problems and subjected them to a regulatory analysis which also
addresses the requirements for a backfit analysis under 10 CFR
50.109.\1\ These actions have been evaluated within the framework of
the Commission's Safety Goal Policy, (51 FR 30028; August 21, 1986) to
determine whether or not they would result in a substantial increase in
the overall protection of the public health and safety.
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\1\The current regulatory analysis only addresses the LCO and SR
option for controls for specific equipment relied upon during
shutdown and low-power operations, whereas the proposed rule allows
for incorporation of controls included in technical specifications
limiting conditions for operation and surveillance requirements in
accordance with 10 CFR 50.36(c)(2) and (3), or plant procedures
required by technical specifications administrative controls
pursuant to 10 CFR 50.36(c)(5). The staff plans to revise the
regulatory analysis to incorporate consideration of other
alternatives as appropriate for equipment controls during shutdown
and low-power operations. In addition, the staff will consider the
following in the revised regulatory analysis: (1) insights gained
from the recent NRC PRAs for shutdown and low-power operations at
Surry and Grand Gulf; (2) industry improvements made in outages; (3)
comments received from ACRS, CRGR and the Commission; (4) specific
industry comments on the draft regulatory analysis documented in a
letter from NUMARC dated January 11, 1994, in a letter from NEI
dated March 28, 1994 and in a letter from GEOG dated April 8, 1994.
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The NRC has observed that many shutdown operations may take place
with the containment partially open. Therefore, cost-effective
regulatory actions are appropriate to ensure substantial reduction in
core-damage probability, and an improvement in the likelihood of
containment isolation, when necessary. These actions would
substantially increase the overall protection of public health and
safety.
Operating Experience
The NRC staff reviewed operating experience at nuclear power plants
to ensure that its evaluation encompassed the range of events
encountered during shutdown and low-power operations including:
licensee event reports (LERs), studies performed by the Office for
Analysis and Evaluation of Operational Data (AEOD), and various
inspection reports to determine the types of events that take place
during refueling, cold and hot shutdown, and low-power operations.
The NRC staff also reviewed events that occurred at foreign nuclear
power plants using information found in the foreign events file
maintained for AEOD at the Oak Ridge National Laboratory (ORNL). The
AEOD compilation included the types of events that applied to U.S.
nuclear plants and those not found in a review of U.S. experience.
In performing this review, the NRC staff found that the more
significant events for pressurized-water reactors (PWRs) were the loss
of residual heat removal, potential pressurization, and boron dilution
events. The more important events for boiling-water reactors (BWRs)
were the loss of coolant, the loss of cooling, and potential
pressurization. Generally, the majority of important events involved
human error and procedural errors. The NRC staff documented this review
in NUREG-1449. In addition, the NRC staff selected 10 events from the
AEOD review for further assessment as precursors to potential severe
core-damage accidents. This assessment is fully documented in NUREG-
1449.
Further, undesirable events continue to occur during shutdown
operations. Recent operating experiences during shutdown include (1)
entry into midloop operation with a degraded RHR pump at a PWR on
December 11, 1993, (2) the discovery of a large, undetected nitrogen
gas bubble in the RCS during extended cold shutdown at a PWR on
December 17, 1993, (3) a hydrogen burn in an empty pressurizer caused
by welding activities during cold shutdown at a PWR on February 3,
1994, and (4) the loss of one train of RHR 2 days after shutdown due to
outage activities at a BWR on March 17, 1994. These recent events
reinforce the previous assessment of shutdown operations documented in
NUREG-1449.
Industry Work
The industry has addressed outage planning and control with
programs that include workshops, Institute of Nuclear Power Operations
(INPO) inspections, Electric Power Research Institute (EPRI) support,
as well as enhanced training and procedures. One activity (a formal
initiative proposed by the Nuclear Management and Resources Council
(NUMARC)) has produced for the utilities a set of guidelines to use for
self-assessment of shutdown operations (NUMARC 91-06).2 This high-
level guidance addresses many, but not all, of the areas in outage
planning that need improvement. Detailed guidance on developing an
outage planning program is outside the scope of the NUMARC effort. The
NRC staff believes that NUMARC 91-06 represents a significant and
constructive step, effects of which have already been realized by many
utilities using the draft guidance in recent outages.3 For
example, on the basis of its review of operating experience and pilot
team inspections, the staff observed that industry efforts and
improvements have been made which should reduce risk in the shutdown
and low-power operations area. Some licensees were observed to have in-
depth contingency planning for backup cooling; other licensees were
found to have well-planned and tightly conducted outages run by outage-
experienced, operationally oriented personnel; and other licensees had
developed well-defined strategies and procedures for plant and hardware
configurations, including fuel offload, midloop operation in PWRs, use
of nozzle dams in PWRs, venting in PWRs, electrical equipment, onsite
sources of ac power, containment status and control, and such key
instrumentation as RCS temperature, reactor water level, and RCS
pressure. Further, industrys defense-in-depth concept for safety
functions and outage strategy contained in NUMARC 91-06 have been
recognized as excellent self-improvements in the shutdown and low-power
operations area. However, implementation of these efforts and
improvements has been unevenly and inconsistently applied, as observed
at several site inspections conducted by the staff.
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\2\These guidelines serve as the basis for an industry-wide
program that has been implemented at all plants.
\3\NUMARC 91-06 is available from Nuclear Energy Institute, 1776
Eye Street NW., Suite 400, Washington, DC 20006-3708.
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Safety Importance
The NRC's staff's rationale for proposing the requirements
described previously is that they will provide substantial safety
improvements, and the costs of implementation are justified in view of
the benefits to be provided. This judgment is based on a qualitative
assessment supplemented by a quantitative analysis. The considerations
that principally support the proposed action are as follows:
(1) The improvements reflect the NRC safety philosophy of ``defense
in depth'' in that they address: (a) Prevention of credible challenges
to safety functions through improvements in operations and fire
protection; and (b) mitigation of challenges to redundant protection
systems, through improved procedures, training, improved controls on
plant equipment and contingency plans.
(2) Accident sequences during shutdown which are as rapid and
severe as those that might occur during power operation should be
addressed with commensurate requirements. This is supported by the
staff's engineering analysis of accidents during shutdown conditions
documented in NUREG-1449.
(3) The improvements being proposed are aimed directly at problems
that have been repeatedly observed in operating experience, e.g., loss
of decay heat removal, loss of ac power, loss of RCS inventory, fires,
personnel errors, poor procedures and poor planning, and lack of
training.
Only a very limited number of probabilistic risk assessment (PRA)
studies covering shutdown conditions have been performed and those
studies contain considerable uncertainty. The uncertainty is due
largely to the predominant role played by operators and other licensee
staff in shutdown events and recovery from them. Human reliability is
difficult to quantify, especially under unfamiliar conditions which are
often not covered in training or procedures. The collection of PRA
studies discussed in NUREG-1449 gives some insight into the likely
range of shutdown risks for the spectrum of current plants. The mean
CDP for shutdown events appears to be in the range of 6E- 05 to 7E-06
per reactor-year. Although detailed uncertainty analysis is not
available for most of the PRAs covering shutdown conditions, some
insight can be gained by examining the uncertainty analysis in NUREG-
1150 where the CDP uncertainty ranges (5th and 95th percentiles) are
approximately one order of magnitude. From this limited information,
the staff concludes that a reasonable estimate of the range of CDP is
1E-04 to 1E-06 per reactor-year.
On the basis of the analysis of operating experience in NUREG-1449,
including the accident sequence precursor analysis, the NRC staff
identified the following as dominant event sequences during shutdown:
loss of all ac power, loss of RCS inventory, and loss of reactor vessel
level control in PWRs. These sequences have been modeled as part of the
regulatory analysis of proposed improvements in shutdown and low-power
operations. Core-damage probabilities for these sequences are point
estimates built from best estimates of each step in the sequence. No
uncertainty analysis was performed because of the lack of reliable
statistical data for shutdown conditions. However, a sensitivity study
has been performed to assess the effect of uncertain assumptions on the
overall results of the analysis. The results of the sensitivity study
show that despite sensitivity to changes in PRA assumptions, the
estimated changes in risk associated with the proposed improvements
remain significant even when inputs are changed significantly.
The results of the analysis of the dominant event sequences
indicate potential reductions in core-damage probability of greater
than 5E-05 per reactor-year for each PWR's improvement, and
approximately 1E-05 per reactor-year for improvement to BWRs. As
previously stated, the staff recognizes that significant improvement in
core-damage probability has already been achieved through recent
industry actions, however, the proposed rule would place a regulatory
``footprint'' on outage safety and codify improvements made by industry
to ensure that (1) reductions in risk already achieved are not eroded
in the future and (2) consistency and uniform achievement of the safety
improvements is realized throughout the industry. The proposed rule
would also set minimum standards for all plants and further reduce risk
by improving safety in the areas of fire protection for shutdown decay
heat removal and effective reactor vessel water level instrumentation
for PWRs in midloop operation.
Containment capability and releases of radioactivity for accident
sequences during shutdown are also evaluated as part of the regulatory
analysis. From that work, the NRC has concluded that an intact
containment will effectively prevent early releases from shutdown
accidents. Large, dry PWR containments should remain intact if closed
before being challenged. Severe core-damage accidents in open
containments or in containments that fail are expected to have offsite
consequences similar to severe core-damage accidents initiating from
power operations. Onsite consequences within a few hundred meters of
open or failed containments may be more severe at shutdown than at
power. The potential dose to the public for a severe core-damage
accident without an effective containment was estimated to be 2E+06
person-rem (2E+04 person-Sv).
Basis for Commission Position
The NRC proposes to resolve concerns regarding shutdown and low-
power operations by rulemaking that would require power reactor
licensees to:
(1) Assure that uncontrolled changes in reactivity, reactor coolant
inventory, and loss of subcooled state in the reactor coolant system
when subcooled conditions are normally being maintained, will not occur
when the plant is in either a shutdown or low-power condition;
(2) Assure that containment integrity is maintained or can be
reestablished in a timely manner as needed to prevent releases in
excess of the guidelines of 10 CFR Part 100 when the plant is in either
a shutdown or low-power condition;
(3) Identify that equipment necessary to make the reactor
subcritical or critical in a controlled manner and maintain it
subcritical in a shutdown condition, and establish controls in either
technical specifications limiting conditions for operation and
surveillance requirements in accordance with the requirements of 10 CFR
50.36(c)(2) and (3) or plant procedures required by technical
specifications administrative controls pursuant to 10 CFR 50.36(c)(5)
for that equipment such that they will ensure each safety function when
the plant is in a shutdown or low power condition;
(4) Prior to (and throughout the shutdown refueling outage as
necessary to accommodate unforeseen contingencies) entering cold
shutdown or a refueling condition, evaluate realistically available
fire-protection features and the outage plan for possible fires
stemming from activities conducted during cold shutdown or refueling
conditions, determine whether such fires could realistically prevent
accomplishment of the normal decay heat removal capability during cold
shutdown or refueling conditions, and if so, either take measures to
prevent loss of normal decay heat removal by such fires during cold
shutdown or a refueling condition, or have a contingency plan in place
that will ensure an alternate decay heat removal capability exists and
that will describe the general steps to connect the alternate decay
heat removal system to the reactor coolant system (RCS); and
(5) For licensees of PWRs only, provide instrumentation for
monitoring water level in the RCS during midloop operation.
The technical basis for the NRC's staff's position is derived from
the NRC staff's comprehensive evaluation of shutdown and low-power
issues in NUREG-1449, ``Shutdown and Low-Power Operations at Nuclear
Power Plants in the United States.'' NUREG-1449 was published as a
draft report for comment in February 1992. The comment period on the
draft NUREG-1449 ended on April 30, 1992, and a large number of
comments were received from utilities and industry organizations. The
NRC staff addressed the comments in the final report (NUREG-1449) which
was issued in September 1993. The principal findings from NUREG-1449
that support the NRC regulatory position in this proposed rule are the
following:
(1) Accident sequences during shutdown can be as rapid and severe
as those during power operations.
(2) All PWR containments and BWR (boiling-water reactor) Mark III
primary containments are capable of offering significant protection if
the containment is closed or can be closed quickly. However, analyses
show that the steam and radiation environment in the containment, which
can result from an extended loss of DHR or LOCA, would make it
difficult to close the containment in many cases. BWR Mark I and II
secondary containments offer less protection against an accident, but
this is offset by a significantly lower likelihood of core damage in
BWRs than in PWRs.
(3) Outage planning is crucial to safety during shutdown conditions
since it establishes (a) if and when a licensee will enter
circumstances likely to challenge safety functions and (b) the level of
mitigation equipment available.
(4) Using technical specifications to control the availability of
safety-related equipment is appropriate because (i) operators are
trained and accustomed to operating the facility in accordance with
approved procedures within the clear limits set by technical
specifications and (ii) technical specifications establish clear and
enforceable regulatory requirements.\4\
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\4\The NUREG-1449 analysis only addressed the use of technical
specifications for control of specific equipment relied upon during
shutdown and low-power operations. The proposed rule allows for
incorporation of controls using either technical specifications
limiting conditions for operation and surveillance requirements in
accordance with the requirements of 10 CFR 50.36(c) (2) and (3), or
plant procedures required by technical specifications administrative
controls pursuant to 10 CFR 50.36(c)(5).
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(5) Although maintenance activities that can increase the potential
for fire are performed during shutdown, current NRC requirements in the
area of fire protection do not apply to shutdown conditions.
(6) Operating experience continues to show that the ability to
maintain control of RCS level in PWRs during draindown and steady-state
operation has been a problem. The principal contributor to events
during some shutdown configurations has been identified as poor quality
and reliability of reactor vessel level instrumentation. This problem
is most significant during midloop operation, where a small variation
in level can lead to a loss of DHR. PRAs have consistently found a
higher risk associated with midloop operation than with other
operational states.
The requirements being proposed by the NRC are aimed directly at
problems that have been repeatedly observed in operating experience,
such as loss of decay heat removal, loss of ac power, loss of RCS
inventory, fires, personnel errors, poor procedures, poor planning, and
poor training. The proposed requirements reflect the NRC safety
philosophy of defense in depth, in that they address: (1) prevention of
credible challenges to safety functions through improvements in
operations, fire protection and water level instrumentation in PWRs and
(2) mitigation of challenges to redundant protection systems, through
improved equipment controls.
Equipment controls must be included in either technical
specifications limiting conditions for operation and surveillance
requirements in accordance with the requirements of 10 CFR 50.36(c)(2)
and (3), or plant procedures required by technical specifications
administrative controls pursuant to 10 CFR 50.36(c)(5). Requirements
for specific equipment availability using plant procedures would be
established by the licensee in a way that provides maximum flexibility
by: (1) permitting the use of non-safety as well as safety equipment to
provide safety functions; (2) permitting reduced decay heat levels to
be a factor in developing such mitigating strategies as the selection
of protective features and determination of when to put such protective
features into service; and (3) allowing changes regarding the
availability of equipment during the outage to be made without prior
NRC review and approval. This particular resolution path has not been
evaluated explicitly in the regulatory analysis; but the NRC believes
that this approach to controlling mitigative equipment can produce a
safety benefit comparable to that for the LCO approach.
Relationship to Existing Requirements
Technical Specifications
Section 50.67(c)(3)(iii) of the proposed rule may result in changes
to plant-specific technical specifications as well as to the standard
technical specifications documented in NUREG-1430, NUREG-1431, NUREG-
1432, NUREG-1433, and NUREG-1434 (STS for Babcock & Wilcox plants,
Westinghouse plants, Combustion Engineering plants, General Electric
BWR/4 plants, and General Electric BWR/6 plants, respectively). Section
50.67(c)(3)(iii) of the proposed rule requires identified equipment
controls during shutdown or low-power conditions to be established in
technical specifications or plant procedures required by technical
specifications administrative controls in support of specific safety
functions, including such support functions as electric power. Section
50.67(c)(3)(ii) states that the controls must reflect sufficient
redundancy in systems, subsystems, components, and features to ensure
that, for the onsite electric power system in operation (assuming
offsite power is not available), safety functions can be accomplished,
assuming a single failure. LCOs currently used at some plants do not
cover all of the safety functions recommended in the proposed rule. For
some systems, under some conditions, standard technical specifications,
as well as current plant-specific technical specifications, lack the
redundancy called for in the proposed rule.
Fire Protection
The principal regulation covering fire protection is 10 CFR 50.48.
It requires all plants to have a fire protection plan that satisfies
General Design Criterion (GDC) 3 of Appendix A to 10 CFR Part 50.
Appendix R to 10 CFR Part 50 gives specific requirements to be
satisfied in complying with the regulation for plants licensed before
1979. Additionally, guidance for satisfying the regulation is found in
the branch technical positions referenced in the regulation. However,
this guidance was developed to ensure that the plant could be brought
to a hot shutdown condition from power operation during a fire and does
not address the condition of being in a shutdown or refueling mode at
the time of a fire. Further, fire-protection criteria established by
the regulations only require that at least one train of those systems
important for ensuring an adequate level of DHR during cold shutdown
and refueling be capable of being restored to service within 72 hours
of a fire. In addition, NRC guidelines for performing a fire hazards
analysis do not address shutdown and refueling conditions, or the
potential impact a fire may have on the capability to maintain shutdown
cooling.
With the proposed requirements in the area of fire protection
during cold shutdown or refueling conditions, it is the Commission's
intent to supplement current requirements for fire protection with
additional requirements to ensure that decay heat removal capability is
not lost because of a fire during cold shutdown or refueling
conditions. If the evaluation required by the proposed rule shows that
fires would prevent accomplishment of normal decay heat removal
capability, the licensee must either take measures to prevent the loss
of normal decay heat removal by such fires or have a contingency plan
in place that will ensure that an alternate decay heat removal
capability exists during cold shutdown or a refueling condition. The
contingency plan should describe the general steps to connect the
alternate decay heat removal system to the RCS. The NRC staff
recognizes that this could be done by revising existing regulations to
include detailed supplemental requirements. However, the proposed
requirements state that realistic fires during cold shutdown and
refueling conditions should be evaluated rather than the more
conservative fires that are analyzed under Appendix R. This realistic
evaluation of available fire-protection features and the outage plan
for possible fires should serve as the basis for further appropriate
action. Permanent hardware fixes need not be employed as an option to
reduce the risk of fire during cold shutdown and refueling conditions.
On the contrary, if the evaluation results in the conclusion that some
changes must be made, the licensee should consider less onerous options
to reduce the risk of fire such as: (a) modifying or relocating the
activities that might cause the fire; (b) constructing temporary fire
barriers; or (c) revising plant procedures.
Instrumentation
The NRC believes the proposed action regarding installation in a
PWR of new reactor vessel water level instrumentation, including an
alarm, is a cost-justified substantial safety enhancement and the costs
of implementation are justified in the view of the substantial benefit
that is provided.\5\ This action stems from a desire to eliminate
losses of the RHR system due to air ingestion caused by operator error
when lowering water level to achieve a midloop condition. The
additional level instrumentation would supplement the improved level
instrumentation adopted voluntarily by all affected licensees in
response to GL 88-17, ``Loss of Decay Heat Removal.''
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\5\The staff's regulatory analysis includes the assumption that
BWR water level instrumentation will be operable during cold
shutdown and refueling operations in accordance with current
standard technical specifications. The results of the analysis
support the conclusion that improvements in BWR water level
instrumentation used during shutdown operations are not warranted.
Recent concerns with the accuracy of BWR water level instrumentation
are being addressed by utilities with actions in response to NRC
Bulletin 93-03, dated May 28, 1993. Those actions will ensure that
BWR water level instrumentation will function as assumed in the
regulatory analysis.
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Expected Achievement
The NRC notes that, based on the available evidence, no undue
public risk exists without the promulgation of the rule for shutdown
and low-power operations. The proposed rule would strengthen safety by
preventing accidents and mitigating accidents, and thereby reduce the
likelihood of a core-damage accident and the offsite releases due to
loss of a key safety function during shutdown or low-power operations.
Significant improvements have already been achieved in this regard
through the implementation of the NUMARC guidelines; however, the
proposed rule would place a regulatory ``footprint'' on outage safety
and codify improvements made by industry to ensure that (1) reductions
in risk already achieved are not eroded in the future and (2)
consistency and uniform achievement of the safety improvements is
realized throughout the industry. The proposed rule would also set
minimum standards for all plants and further reduce risk by improving
safety in the areas of fire protection for shutdown decay heat removal
and effective reactor vessel water level instrumentation for PWRs in
midloop operation. Moreover, the overall risk may also be reduced by
additional improvements in severe accident management, given the
assumption that core damage occurs, whether from an event during an
outage or during power operations. Therefore, the proposed rule should
be viewed as being in the same accident prevention context as the ATWS
rule (10 CFR 50.62) and the station blackout rule (10 CFR 50.63) in
that it recognizes, as the other two rules recognize, multiple failure
possibilities resulting from common cause effects that should be
addressed.
Comments Requested
Section 50.67(c)(3)(i) of the proposed rule calls for the
identification of equipment necessary to (a) make the reactor
subcritical or critical in a controlled manner and maintain the reactor
subcritical in a shutdown condition, (b) maintain RCS inventory and
capability to add makeup water to the reactor vessel, (c) remove decay
heat from the reactor, (d) monitor water level in the reactor vessel,
and (e) maintain or reestablish containment integrity when the plant is
in a shutdown or low-power condition. Further, Section 50.67(c)(3)(ii)
of the proposed rule requires licensees to establish controls for the
equipment identified such that they will perform their safety function
when the plant is in a shutdown or low power condition. The controls
must reflect sufficient redundancy in systems, subsystems, components,
and features to ensure that, for the onsite electric power system in
operation (assuming offsite power is not available), safety functions
can be accomplished, assuming a single failure, for all conditions
except refueling operations (with water level above the reactor in
excess of a lower limit established in applicable technical
specifications or plant procedures). Section 50.67(c)(3)(iii) of the
proposed rule specifies that the controls required by paragraph
(c)(3)(ii) be included in technical specifications limiting conditions
for operation and surveillance requirements in accordance with the
requirements of 10 CFR 50.36(c)(2) and (3), or plant procedures
required by technical specifications administrative controls pursuant
to 10 CFR 50.36(c)(5). The NRC would like to receive comments
describing the possible alternate methods for equipment controls.
Additionally, the current regulatory analysis only addresses LCO and SR
changes within the technical specifications, and does not reflect the
risk reduction already achieved by industry through voluntary actions.
The Commission requests information as to steps that licensees have
already taken to reduce risk during shutdown and low-power operations.
Finally, the NRC would like to receive comments on the use of
probabilistic risk assessment (PRA) information and the calculation of
the value of offsite dose (accident consequence) in the cost/benefit
analysis.
Availability of Documents
Copies of all NRC documents, including generic issue (GI) notices
are available for public inspection and copying for a fee at the NRC
Public Document Room (PDR) at 2120 L Street, N.W. (Lower Level)
Washington, DC 20555-0001.
Copies of NUREGs-1150, 1410, 1430, 1431, 1432, 1433, 1434, and 1449
may be purchased from the Superintendent of Documents, U.S. Government
Printing Office, by calling (202) 275-2060 or by writing to the
Superintendent of Documents, U.S. Government Printing Office, Mail Stop
SSOP, Washington, DC 20402-9328. Copies are also available from the
National Technical Information Service, 5825 Port Royal Road,
Springfield, VA 22161.
Criminal Penalties
For purposes of section 223 of the Atomic Energy Act of 1954, as
amended (AEA), the Commission proposes to issue the proposed rule under
one or more of sections 161b, 161i, or 161o of the AEA. Willful
violations of the rule are subject to criminal enforcement.
Finding of No Significant Environmental Impact: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR Part 51, that this rule, if adopted, does not
degrade the environment in any way. The actions resulting from this
rule, if adopted, would reduce the core damage frequency and risks
during shutdown and low-power operations. Therefore, the Commission
concludes that there will be no significant impact on the environment
from this proposed rule. This discussion constitutes the environmental
assessment and finding of no significant impact for this proposed rule;
a separate assessment has not been prepared.
Paperwork Reduction Act Statement
This proposed rule amends information collection requirements that
are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et
seq.). The rule has been submitted to the Office of Management and
Budget for review and approval of the information collection
requirements.
The public reporting burden for this collection of information is
estimated to average 3160 hours per respondent, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. Send comments regarding this burden estimate
or any other aspect of this collection of information, including
suggestions for reducing the burden, to the Information and Records
Management Branch (T-6 F 33), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of
Management and Budget, Washington, DC 20503.
Regulatory Analysis
The Commission has prepared a draft regulatory analysis\6\ for this
proposed rule that examines the costs and benefits of the alternatives
considered. This analysis is documented in a report entitled,
``Regulatory Analysis in Accordance with 10 CFR 50.109: Requirements
for Shutdown and Low-Power Operations at Nuclear Power Plants,'' and is
available for inspection in the NRC Public Document Room, 2120 L
Street, N.W. (Lower Level), Washington, DC. Single copies of the
analysis may be obtained from Kulin Desai, Division of Systems Safety
and Analysis, Office of Nuclear Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, Telephone: (301) 504-
2835.
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\6\The current regulatory analysis only addresses the LCO and SR
Option for controls for specific equipment relied upon during
shutdown and low-power operations, whereas the proposed rule allows
for incorporation of controls including technical specifications
limiting conditions for operation and surveillance requirements in
accordance with 10 CFR 50.36(c)(2) and (3), or plant procedures
required by technical specifications administrative controls
pursuant to 10 CFR 50.36(c)(5). The staff plans to revise the
regulatory analysis to incorporate consideration of other
alternatives as appropriate for equipment controls during shutdown
and low-power operations. In addition, the staff will consider the
following in the revised regulatory analysis: (1) insights gained
from the recent NRC PRAs for shutdown and low-power operations at
Surry and Grand Gulf; (2) industry improvements made in outages; (3)
comments received from ACRS, CRGR and the Commission; and (4)
specific industry comments on the draft regulatory analysis
documented in a letter from NUMARC dated January 11, 1994, in a
letter from NEI dated March 28, 1994 and in a letter from CEOG dated
April 8, 1994.
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The Commission requests public comments on the proposed rule, draft
Regulatory Guide, ``Shutdown and Low-Power Operations at Nuclear Power
Plants,'' and the draft report documenting the regulatory analysis,
entitled, ``Regulatory Analysis in Accordance with 10 CFR 50.109:
Requirements for Shutdown and Low-Power Operations at Nuclear Power
Plants.''
Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980, (5
U.S.C. 605(b)), the Commission certifies that, if promulgated, this
proposed rule 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 scope of the definition of
``small entities'' as given in the Regulatory Flexibility Act or the
Small Business Size Standards in regulations issued by the Small
Business Administration at 13 CFR Part 121.
Backfit Analysis
As required by 10 CFR 50.109, a backfit analysis has been performed
for the proposed rule. The backfit analysis on which this determination
is based is included in the report entitled, ``Regulatory Analysis in
Accordance with 10 CFR 50.109: Requirements for Shutdown and Low-Power
Operations at Nuclear Power Plants,'' dated December 1993. The backfit
analysis approach emphasized a qualitative estimation supplemented by a
quantitative analysis for bounding conditions as reflected in the
regulatory analysis. The backfit analysis and the regulatory analysis
will be revised based on comments received from the public. The
Commission has determined, based on this analysis, that backfitting to
comply with the requirements of this proposed rule will provide a
substantial increase in protection to public health and safety because
it would: (1) reduce the frequency of events caused by poor planning
and control of activities during outages; (2) ensure availability of
key safety functions during shutdown and low-power operations at all
plants; (3) ensure that a method of decay heat removal remains viable
in the event of a fire in any plant area during cold shutdown or
refueling conditions; and (4) provide accurate instrumentation for PWRs
to use when draining the reactor coolant system to a midloop
configuration to avoid air binding and eventual loss of residual heat
removal pumps. The Commission has further determined the cost of
implementing the new requirements is justified for PWRs in view of the
increase in protection attributable to the proposed backfits but plans
to specifically reassess BWRs following consideration of comments on
this proposed rulemaking.
List of Subjects
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
For the reasons given in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended and 5 U.S.C. 553, the NRC is proposing to adopt
the following amendments to 10 CFR Part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for Part 50 continues to read as follows:
Authority: Secs. 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. 1244, 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 as amended by Pub. L. 102-486, Sec. 2902, 106 Stat 3123 (42
U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68
Stat. 936, 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. In Sec. 50.8 paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35,
50.36, 50.36a, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61,
50.63, 50.64, 50.65, 50.67, 50.71, 50.72, 50.75, 50.80, 50.82, 50.90,
50.91, and appendices A, B, E, G, H, I, J, K, M, N, O, Q, and R to this
part.
3. A new Sec. 50.67 is added to read as follows:
Sec. 50.67 Shutdown and low-power operations.
(a) Applicability. This section applies to all holders of operating
licenses for commercial nuclear power plants.
(b) Definitions. For the purposes of this section:
Cold Shutdown means that plant state in which the reactor is
subcritical, KEffective is less than .99, the reactor coolant
system temperature is less than or equal to 200 deg.F, and all reactor
vessel head closure bolts are fully tensioned.
Low Power Condition means that the plant is operating with the
reactor critical and the main generator isolated from the grid because
the output breaker connecting the unit to the utility power grid is
open.
Midloop Operation means that plant operational state in which the
plant is in a shutdown condition, fissionable fuel assemblies are
present within the reactor vessel, and the reactor coolant system (RCS)
water level is below the top of the flow area of the hot legs at the
junction with the reactor vessel.
Outage Plan means that written plan of activities to be conducted
during a shutdown or low power condition.
Refueling Condition means that plant state in which the reactor is
subcritical with fissionable fuel assemblies present within the reactor
vessel, and one or more reactor vessel head closure bolts are less than
fully tensioned.
Shutdown Condition means that plant state in which the reactor is
subcritical with fissionable fuel assemblies present within the reactor
vessel.
Technical Specifications, Administrative Controls, Limiting
Conditions for Operation, and Surveillance Requirements are as defined
in 10 CFR 50.36.
(c) General Requirements. All licensees must:
(1) Provide reasonable assurance that uncontrolled changes in
reactivity, uncontrolled changes in reactor coolant inventory, and loss
of subcooled state in the reactor coolant system when subcooled
conditions are normally being maintained will not occur when the plant
is in either a shutdown or low power condition.
(2) Assure that containment integrity is maintained or can be
reestablished in a timely manner as needed to prevent releases in
excess of the guidelines of 10 CFR part 100 when the plant is in a
shutdown or low power condition.
(3)(i) Identify that equipment (including electric power and
compressed air) necessary to:
(A) Make the reactor subcritical or critical in a controlled manner
and maintain it subcritical in a shutdown condition;
(B) Maintain reactor coolant system inventory and capability to add
makeup water to the reactor vessel;
(C) Remove decay heat from the reactor;
(D) Monitor water level in the reactor vessel; and
(E) Maintain or reestablish containment integrity when the plant is
in a shutdown or low power condition;
(ii) Establish controls for the equipment identified in paragraph
(c)(3)(i) of this section such that they will perform their safety
function when the plant is in a shutdown or low power condition. The
controls must reflect sufficient redundancy in systems, subsystems,
components, and features to ensure that, for the onsite electric power
system in operation (assuming offsite power is not available), safety
functions can be accomplished, assuming a single failure, for all
conditions except refueling operations (with water level above the
reactor in excess of a lower limit established in applicable technical
specifications or plant procedures); and
(iii) The controls required by paragraph (c)(3)(ii) of this section
must be included in either:
(A) Technical specifications limiting conditions for operation and
surveillance requirements in accordance with the requirements of 10 CFR
50.36(c) (2) and (3), or
(B) Plant procedures required by technical specifications
administrative controls pursuant to 10 CFR 50.36(c)(5).
(4)(i) Prior to (and throughout the shutdown refueling outage as
necessary to accommodate unforeseen contingencies) entering cold
shutdown or a refueling condition, evaluate realistically available
fire protection features and the outage plan for possible fires
stemming from activities conducted during cold shutdown or refueling
conditions, and determine realistically whether such fires could
prevent accomplishment of normal decay heat removal capability during
cold shutdown or refueling conditions. If the evaluation shows that
such fires would prevent accomplishment of normal decay heat removal
capability, the licensee must either:
(A) Take measures to prevent the loss of normal decay heat removal
by such fires during cold shutdown or a refueling condition; or
(B) Have a contingency plan in place that will ensure an alternate
decay heat removal capability exists and that will describe the general
steps to connect the alternate decay heat removal system to the RCS.
Plant staff must be trained in the implementation of the contingency
plan.
(ii) Any departures from the outage plan during the shutdown or
refueling outage shall be evaluated in the manner also described above
and appropriate measures implemented.
(d) Requirements for licensees of PWRs. All licensees of
pressurized-water reactors must provide instrumentation for monitoring
water level in the RCS during midloop operation. The accuracy of the
instrumentation shall not be affected by changes in pressure in the RCS
or connected systems. The installed instrumentation shall include
visible and audible indications in the control room to alert operators
before water level falls below a prescribed limit.
(e) Implementation. (1) All licensees must comply with paragraph
(c) of this section by no less than 6 months before the first refueling
outage that starts either 12 months or more after the effective date of
this section or 12 months or more after issuance of the Commission's
regulatory guide giving details and examples of approaches to satisfy
these requirements (whichever is later).
(2) If the licensee chooses to install or modify systems,
structures, or components to comply with the requirements of paragraph
(c) of this section, such hardware installation and/or modification
must be completed by the end of the first refueling outage that starts
either 12 months or more after the effective date of this section or 12
months or more after issuance of the Commission's regulatory guide
giving details and examples of approaches to satisfy these requirements
(whichever is later).
(3) All licensees must submit technical specifications required by
paragraph (c)(3)(iii) within 6 months after issuance of the final
regulatory guide providing guidance on compliance with the requirements
of this section.
(4) All licensees of PWRs, except as noted in paragraph (e)(5) of
this section, must comply with paragraph (d) of this section by the end
of the first refueling outage that starts either 12 months or more
after the effective date of this section or 12 months or more after
issuance of the Commission regulatory guide giving details and examples
of approaches to satisfy this requirement (whichever is later).
(5) The requirement in paragraph (e)(4) of this section does not
apply to those plants that have completely defueled for final shutdown
but still retain an operating license (i.e., those plants that are
preparing for decommissioning).
Dated at Rockville, Maryland, this 14th day of October, 1994.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Acting Secretary of the Commission.
[FR Doc. 94-25916 Filed 10-18-94; 8:45 am]
BILLING CODE 7590-01-P