[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