Shutdown and Low-Power Operations for Nuclear Power Reactors

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