Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63769

Unit Number of

members

Mississippi ................................ 2

Louisiana .................................. 2

Tennessee ................................ 2

North Carolina .......................... 2

Kentucky ................................... 2

Michigan ................................... 2

North Dakota ............................ 2

Wisconsin ................................. 2

Maryland ................................... 1

Virginia ...................................... 1

Georgia ..................................... 1

South Carolina .......................... 1

Alabama .................................... 1

Delaware ................................... 1

Texas ........................................ 1

Pennsylvania ............................ 1

Oklahoma ................................. 1

New Jersey ............................... 1

Eastern Region (New York,

Massachusetts, Connecticut,

Florida, Rhode Island,

Vermont, New Hampshire,

Maine, West Virginia, District

of Columbia, and Puerto Rico 1

Western Region (Montana, Wyoming, Colorado, New Mexico, Idaho, Utah, Arizona,

Wash-ington, Oregon, Nevada, California, Hawaii, and

Alaska) .................................. 1

* * * * *

Dated: October 19, 2000.

Barry L. Carpenter,

Deputy Administrator, Livestock and Seed

Program.

[FR Doc. 00–27411 Filed 10–24–00; 8:45 am]

BILLING CODE 3410–02–P

NUCLEAR REGULATORY

COMMISSION

10 CFR Parts 50 and 72

RIN 3150–AF98

Reporting Requirements for Nuclear

Power Reactors and Independent

Spent Fuel Storage Installations at

Power Reactor Sites

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final rule.

SUMMARY: The Nuclear Regulatory

Commission (NRC) is amending the

event reporting requirements for nuclear

power reactors to reduce or eliminate

the unnecessary reporting burden

associated with events of little or no

safety significance. This final rule

continues to provide the Commission

with reporting of significant events

where Commission action may be

needed to maintain or improve reactor

safety or to respond to heightened

public concern. This final rule also

better aligns event reporting

requirements with the type of

information NRC needs to carry out its

safety mission, including revising

reporting requirements based on

importance to risk and extending the

required reporting times consistent with

the time that information is needed for

prompt NRC action. Also, NUREG–

1022, Revision 2, ‘‘Event Reporting

Guidelines, 10 CFR 50.72 and 50.73,’’ is

being made available concurrently with

the amendments.

DATES: The final rule is effective January

23, 2001.

ADDRESSES: Documents related to this

action may be examined, and/or copied

for a fee, at the NRC’s Public Document

Room, located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland. Documents created

or received at the NRC after November

1, 1999 are also available electronically

at the NRC’s Public Electronic Reading

Room on the Internet at http://

www.nrc.gov/NRC/ADAMS/index.html.

From this site, the public can gain entry

into the NRC’s Agencywide Document

Access and Management System

(ADAMS), which provides text and

image files of NRC’s public documents.

For further information contact the PDR

Reference staff at 1–800–397–4209, 301–

415–4737 or by email to pdr@nrc.gov.

FOR FURTHER INFORMATION CONTACT:

Dennis P. Allison, Office of Nuclear

Reactor Regulation, Washington, DC

20555–0001, telephone (301) 415–1178,

e-mail dpa@nrc.gov.

SUPPLEMENTARY INFORMATION:

Contents

I. Background

II. Analysis of Comments

III. Discussion

1. Objectives

2. Section by Section Discussion of Final

Amendments

3. Revisions to Event Reporting Guidelines

in NUREG–1022

4. Reactor Oversight

5. Enforcement

6. Electronic Reporting

7. State Input

8. Plain Language

IV. Environmental Impact: Categorical

Exclusion

V. Backfit Analysis

VI. Regulatory Analysis

VII. Paperwork Reduction Act Statement

VIII. Regulatory Flexibility Act Certification

IX. Small Business Regulatory Enforcement

Fairness Act

X. National Technology Transfer and

Advancement Act

XI. Final Amendments

I. Background

The reporting requirements in

Sections 50.72 and 50.73 have been in

effect, with minor modifications, since

1983. Experience has shown a need for

change in several areas. On July 23,

1998 (63 FR 39522), the NRC published

in the Federal Register an advance

notice of proposed rulemaking (ANPR)

to announce a contemplated rulemaking

that would modify reporting

requirements for nuclear power reactors.

Among other things, the ANPR

requested public comments on several

concrete proposals for modification of

the event reporting rules. Public

meetings were held to discuss the ANPR

at NRC Headquarters on August 21,

1998, in Rosemont, Illinois on

September 1, 1998, and at NRC

Headquarters on November 13, 1998.

A proposed rule was published in the

Federal Register on July 6, 1999 (64 FR

36291), including a conforming change

to Section 72.216. Concurrently, a draft

revision to the associated event

reporting guidelines was made available

for public comment (NUREG–1022,

Draft Revision 2). A public meeting was

held at NRC Headquarters on August 3,

1999, to discuss the proposed rule and

draft guidelines. Public comments were

due on September 20, 1999. Additional

public meetings were held on February

25, and March 22, 2000, to discuss

public comments.

II. Analysis of Comments

The comment period for the proposed

rule expired September 20, 1999.

Twenty-seven comment letters were

received, representing comments from

24 nuclear power plant licensees

(utilities), two organizations of utilities,

and one State agency.

In addition to the written comments

received, the proposed rule was the

subject of a public meeting on August 3,

1999, as discussed above under the

heading ‘‘Background,’’ and comments

made at that meeting have also been

considered.

Most commenters expressed support

for amending the rules in accordance

with the objectives discussed in the

proposed rule. However, they objected

to some of the specific provisions. Many

comments also provided specific

recommendations for changes to the

proposed rules. The resolution of

comments is summarized below. This

summary addresses the principal

comments (i.e., comments other than

those that are: minor or editorial in

nature; supportive of the approach

described in the proposed rules; or

applicable to another area or activity

outside the scope of sections 50.72 and

50.73).

Comment A (Do not require reporting

of degraded components): The proposed

rule included a new component

63770 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

reporting criterion. It would have

required reporting ‘‘Any event or

condition that resulted in a component

being in a degraded or non-conforming

condition such that the ability of the

component to perform its specified

safety function is significantly degraded

and the condition could reasonably be

expected to affect other similar

components in the plant.’’ The term

‘‘significantly degraded’’ was defined by

providing several examples of

reportable and non-reportable events.

The stated purpose was to ensure that

design basis or other discrepancies

would continue to be reported if the

capability to perform a specified safety

function is significantly degraded and

the condition has generic implications.

Most commenters strongly objected to

the proposed component reporting

criterion. Among other things, they

indicated:

(1) The proposed component

reporting criterion is not needed

because, after deleting the requirement

to report a condition that is outside the

design basis of the plant, any significant

events would still be captured by the

other existing criteria.

(2) The proposed component

reporting criterion would be unclear

and subject to widely varying

interpretation with regard to the

meaning of the term ‘‘significantly

degraded’’ and the term ‘‘could

reasonably be expected to apply to other

similar components.’’

(3) The proposed component

reporting criterion would be overly

burdensome. For example, it would

become necessary to screen all single

component failures for reportability.

(4) The proposed component

reporting criterion would be contrary to

the stated objectives of the rulemaking.

For example, it would result in many

additional reports for events with little

or no safety- or risk-significance.

Response: In the final rule, the

proposed reporting criterion has been

retained, but modified to address the

concerns about unnecessary burden and

clarity expressed in the comments. It

requires reporting any event or

condition that as a result of a single

cause could have prevented the

fulfillment of a safety function for two

or more trains or channels in different

systems that are needed to:

(1) Shut down the reactor and

maintain it in a safe shutdown

condition;

(2) Remove residual heat;

(3) Control the release of radioactive

material; or (4) Mitigate the

consequences of an accident.

Events covered by this criterion may

include cases of procedural error,

equipment failure, and/or discovery of a

design, analysis, fabrication,

construction, and/or procedural

inadequacy. However, licensees are not

required to report an event pursuant to

this criterion if the event results from:

(1) A shared dependency among

trains or channels that is a natural or

expected consequence of the approved

plant design; or

(2) Normal and expected wear or

degradation.

Subject to the two exclusions stated

above, this criterion, as modified, is

needed to capture those events with

enough generic significance that a single

cause could have prevented the

fulfillment of the safety function of

multiple trains or channels, but the

event:

(1) Would not be captured by

§§ 50.73(a)(2)(v) and 50.72(b)(3)(v)

[event or condition that could have

prevented fulfillment of the safety

function of structures and systems

needed to * * *] because the affected

trains or channels are in different

systems; and

(2) Would not be captured by

§ 50.73(a)(2)(vii) [common cause

inoperability of independent trains or

channels] because the affected trains or

channels are either:

(i) Not assumed to be independent in

the plant’s safety analysis; or

(ii) Not both considered to be

inoperable.

The criterion, as modified, would not

be unclear because it uses the term

‘‘could have prevented fulfillment of the

safety function,’’ which is already used

in a previously existing criterion.

The criterion, as modified, is not

considered overly burdensome because

it is estimated to result in fewer reports

than the previous requirement to report

a condition outside the design basis of

the plant. It is not necessary to screen

all single component failures for

reportability.

The criterion, as modified, is

considered consistent with the

objectives of the rulemaking for the

same reasons.

Comment B (Do not change the term

‘‘any engineered safety feature

[ESF] * * *’’): In the proposed rule, the

term ‘‘any engineered safety feature

(ESF), including the reactor protection

system (RPS),’’ which defines the

systems for which actuation must be

reported, would be replaced by a

specific list of systems. It was

recognized that this proposal to list the

systems in the rule was controversial

and public comment was specifically

invited in this area. In particular, three

principal alternatives to the proposed

rule were identified for comment. They

are:

Alternative 1, status quo. The rule

would continue to require reporting for

actuation of ‘‘any ESF.’’ The guidance

would continue to infer that reporting

should include the systems on a list

which is similar to the list in the

proposed rule.

Alternative 2, plant-specific list. The

rule would require that licensees

develop a plant-specific, risk-informed

list.

Alternative 3, pre-1998 practice. The

rule would continue to require reporting

actuation of ‘‘any ESF.’’ The guidance

would indicate that this includes those

systems identified as ESF’s in each

plant’s final safety analysis report

(FSAR).

The comments may be summarized as

follows:

(1) Most commenters objected to the

proposed rule, which would replace the

term ‘‘any engineered safety feature

(ESF), including the reactor protection

system (RPS)’’ with a list of specific

systems. The reasons cited by the

commenters include the following:

(a) Providing an all-inclusive list of

systems in the rules is inappropriate.

(b) Each facility’s FSAR specifies

equipment that is designated as ESF

equipment.

(c) Plant-specific differences exist in

the safety-related status of their systems.

(d) The risk-significance of a

particular system can vary greatly

between plants, due to a wide variety of

design differences. An all-inclusive list

would increase the burden for some

plants whose equipment on the list was

not ESF equipment or equipment with

a suitably high risk-significance.

(e) There are a number of specific

problems with the proposed list.

Specific examples were provided.

(2) Most commenters recommended

in favor of Alternative 3, returning to

the pre-1998 practice of reporting

actuation for only those systems that are

designated as ESFs in each facility’s

FSAR. They stated that this option best

meets the goal of clarity and simplicity.

(3) One commenter recommended in

favor of Alternative 1 (status quo),

where the reporting guidelines contain

a list of systems similar to the list

proposed for the rule. It stated that the

facility’s internal reporting procedures

already reflect the current practice. Any

benefit that might be obtained by

returning to the pre-1998 practice would

be so slight that it would not justify the

cost of changing the procedures.

(4) Some commenters indicated that

there are problems with the status quo

that need to be solved. For example, the

reporting guidelines should exclude

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63771

reporting of reactor water cleanup

system (RWCU) isolations that routinely

occur during system restoration

following maintenance outages, due to

rapid pressurization following valve

opening.

(5) Most commenters objected to

Alternative 2 (developing a plantspecific, risk-informed list of systems).

They stated that this would require a

significant expenditure of resources and

it is unclear as to whether or how it

would meet the NRC’s needs better than

Alternative 3 (returning to pre-1998

practice). They also noted that there is

a separate initiative to ‘‘risk-inform’’ 10

CFR Part 50. This may result in

development of plant-specific lists of

systems based on risk significance.

However, the commenters do not

believe the necessary criteria have been

adequately established to make that

shift as part of this rulemaking to

modify 10 CFR 50.72 and 50.73. They

recommended that later, as part of the

rule change to ‘‘risk-inform’’ Part 50, the

NRC should evaluate whether or not it

is appropriate to ‘‘risk-inform’’ ESF

systems subject to the event reporting

requirements of 10 CFR 50.72 and 50.73.

Response: (1) The NRC believes

providing a list of systems is the best

approach because it will obtain

consistent reporting of events that result

in actuation of highly risk-significant

systems. Consistent reporting for such

events is needed to support estimating

equipment reliability parameters and is

important to several aspects of the

NRC’s general move towards more riskinformed regulation.

Commenters stated that the risksignificance of the systems varies

depending on plant design. As

discussed below under the headings

‘‘(e)(i)’’ through ‘‘(e)(vii),’’ a number of

items have been removed from the list

based on specific comments. The NRC

believes that these systems remaining

on the list are of sufficient risk

significance to warrant reporting of a

system actuation. The principal reason

for reporting an actuation of one of these

systems is that it is indicative of an

unplanned plant transient that the NRC

needs to evaluate to determine if action

is necessary to address a safety problem.

In this context, the NRC’s need to

evaluate the event is independent of

classification of the system. For

example, a valid actuation of the

auxiliary feedwater (AFW) system at a

pressurized water reactor (PWR) means

there was a transient that involved an

abnormal plant parameter, such as low

steam generator level, which initiated

the actuation. This is the reason the

NRC needs to evaluate the event, and it

is independent of how the AFW system

happens to be classified at the particular

plant.

The classification of systems in the

FSARs has evolved over the years. For

example, in earlier PWR designs the

auxiliary feedwater system was not

considered to be an ESF, and this is

reflected in early FSARs. Later, although

the system’s function and importance

did not change, it came to be considered

an ESF, and this is reflected in later

FSARs. Since the function and

importance is the same regardless of

classification, it does not make sense to

exclude reporting for actuation of the

auxiliary feedwater system based on its

classification in the FSAR.

Furthermore, this approach is

estimated to result in a net reduction in

the number of events reported under

this criterion. Some licensees will make

additional reports involving highly risksignificant systems. However, these

additional reports will be outweighed

by the elimination of reports involving

systems with lesser risk-significance.

(a) Commenters indicated that

providing an all-inclusive list of systems

in the rules is inappropriate. However,

the NRC does not believe the list is all

inclusive. It contains only systems that

are highly risk-significant and omits

systems of lesser risk-significance, even

if the systems of lesser risk-significance

are designated as ESFs. The NRC also

believes the list is appropriate because

it provides consistent reporting of

events that result in actuation of these

highly risk-significant systems and, at

the same time, a net reduction in

reporting burden.

(b) Commenters stated that each

facility’s FSAR specifies equipment that

is designated as ESF equipment.

However, the NRC believes that those

lists are not consistent or risk-informed.

For example, at several plants,

emergency diesel generators (EDGs),

which are highly risk-significant, are not

identified as ESFs. At several

pressurized water reactors (PWRs), the

AFW system which is highly risksignificant, is not identified as an ESF.

At most boiling water reactors (BWRs),

the reactor core isolation cooling (RCIC)

system, which is highly risk-significant,

is not identified as an ESF. On the other

hand, most plants identify systems with

lesser risk-significance, such as fuel

building ventilation and filtration

systems, as ESFs.

(c) Commenters stated that plantspecific differences exist in the safetyrelated status of systems. However, the

NRC does not believe that this fact bears

directly on the question of which

system actuations should be reported.

There is no need to report the actuation

of all safety related systems, and there

is no reason to exclude reporting for the

actuation of a non-safety-related system

if it is highly risk-significant simply on

the basis that it has not been classified

by the licensee as an ESF.

(d) Commenters stated that the risksignificance of a particular system can

vary greatly among plants. They further

stated that an all-inclusive list would

therefore increase the burden for some

plants whose equipment on the list was

not ESF equipment or equipment with

a suitably high risk-significance. The

NRC agrees with the general statement

that the risk-significance of a particular

system can vary greatly among plants.

However, the systems on the list are

virtually always of high risksignificance. While it is true that, as a

result of the list, some licensees will

make additional reports, any additional

reports will involve systems that are

highly risk-significant. Also, these

additional reports will be outweighed

by the elimination of reports involving

systems with lesser risk-significance.

Thus, the net effect is a reduction in

reporting.

(e) Commenters provided several

specific examples of items they

considered to be problems with the list.

These examples are:

(i) In the proposed rule, the feedwater

coolant injection (FWCI) system was

characterized as an example of an

emergency core cooling system (ECCS).

Commenters stated that FWCI systems

are not considered to be ECCS. The NRC

believes that clarification is warranted.

In the final rule, FWCI is not

characterized as an ECCS. However, it is

included as a separate item in the list.

(ii) The proposed rule would have

required reporting actuations of the

RCIC system. Commenters stated that

RCIC is included in the Improved

Standard Technical Specifications

(ISTS) because it meets criterion 4 of 10

CFR 50.36, based on its contribution to

the reduction of overall plant risk. They

further stated that RCIC is not credited

in the plant’s safety analysis. The NRC

believes that RCIC is highly risksignificant and, therefore, it remains on

the list in the final rule.

(iii) Commenters stated that nonreportable exceptions should be allowed

for systems that are considered to be

ESFs, yet have lower levels of risk

significance (control room ventilation

systems, reactor building ventilation

systems, fuel building ventilation

systems, auxiliary building ventilation

systems, RWCU isolations during

restoration from maintenance, etc.). The

NRC agrees. The final rule eliminates

unnecessary reporting for systems that

are considered to be ESFs, yet have

lower levels of risk significance. It also

63772 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

eliminates reporting for RWCU

isolations during restoration from

maintenance because they are routine

and are of low risk and safety

significance.

(iv) Commenters stated that the list

inappropriately includes ‘‘associated

support systems’’ for BWR Division 3

EDGs. The NRC agrees. In the final rule

the term ‘‘associated support systems’’

has been eliminated for BWR Division 3

EDGs, and other EDGs as well.

(v) Commenters stated that the list

inappropriately includes station

blackout diesel generators (and black

start gas turbines that serve a similar

purpose) that are not safety related. The

NRC agrees. The final rule does not

require reporting for station blackout

diesel generators (and black start gas

turbines that serve a similar purpose).

(vi) Commenters stated that although

the term ‘‘anticipated transient without

scram (ATWS) mitigating systems’’ is

clear to those licensees that have

dedicated systems (i.e. AMSAC), a great

deal of confusion exists for those that

have no dedicated system. Due to the

lack of clarity, it could be interpreted

that any system that might be used

during an ATWS would fall into this

category (i.e. feedwater systems,

borating systems, control rods, etc.).

Extensive clarification would be needed

to eliminate this ambiguity. The NRC

agrees that clarification is warranted. In

the final rule this item has been

eliminated. Reporting is not needed for

actuations for a system such as AMSAC.

The reports needed for other systems are

captured by other items on the list.

(vii) Commenters stated that it is

unclear as to whether the service water

entry applies only to emergency service

water systems (i.e., those that don’t

operate unless there is an accident) or

also to the standby service water

systems that only run to remove heat

from the residual heat removal (RHR)

heat exchangers. The NRC agrees. In the

final rule this item has been clarified to

indicate that reporting is required for

emergency service water (ESW) systems

that do not normally run and that serve

as ultimate heat sinks. In addition, this

item has been deleted from the list of

systems for which telephone

notification is required under section

50.72 because an ESW actuation by

itself does not indicate the type of

transient that the NRC needs to

evaluate. However, ESW system

actuations are reportable only under

section 50.73 because the information is

needed to support the NRC staff’s

equipment reliability estimates.

(2) As stated by commenters,

Alternative 3 would provide clarity and

simplicity. However, the NRC believes

that adoption of the list of systems in

the final rule also provides clarity and

simplicity.

(3) Although one commenter

recommended in favor of Alternative 1,

the NRC believes that this alternative

would invite variable interpretation.

The event reporting guidelines would

contain a list of systems, whereas the

rule would require reporting the

actuation of ‘‘any ESF.’’

(4) Some commenters stated that,

under previous requirements it was

necessary to report reactor water

cleanup system isolations that routinely

occur during system restoration

following maintenance outages, due to

rapid pressurization following valve

opening. The list of systems eliminates

these unneeded reports because it limits

the reporting of containment isolation

signals to those that affect multiple

systems or multiple main steam

isolation valves (MSIVs).

(5) As indicated in the comments,

with respect to Alternative (2), the

project to ‘‘risk-inform’’ 10 CFR Part 50

may, in the future, lead to development

of plant-specific lists of systems based

on importance to risk and, as part of that

project, it may be appropriate to

consider whether or not the

applicability of this reporting criterion,

as well as other reporting criteria,

should be based on such lists. It is

expected that at that time the criteria

necessary for development of the list

will have been adequately established.

Comment C (Eliminate reporting for

historical events): The proposed rule

would have eliminated the requirement

for a telephone notification, under 10

CFR 50.72, for:

(1) ‘‘Any event, found while the

reactor is shutdown, that, had it been

found while the reactor was in

operation, would have resulted in the

nuclear power plant, including its

principal safety barriers, being seriously

degraded or being in an unanalyzed

condition that significantly

compromises plant safety,’’ and

(2) ‘‘Any event or condition that alone

could have prevented the fulfillment of

the safety function of structures or

systems that are needed to: (A) Shut

down the reactor and maintain it in a

safe shutdown conditions; (B) remove

residual heat; (C) Control the release of

radioactive material; or (D) Mitigate the

consequences of an accident’’ if the

condition no longer exists at the time of

discovery.

The proposed rule would also have

eliminated the requirement for a written

licensee event report (LER), under 10

CFR 50.73, for:

(1) ‘‘Any operation or condition

prohibited by the plant’s Technical

Specifications,’’ if the condition has not

existed within three years of the date of

discovery, and

(2) ‘‘Any event or condition that alone

could have prevented the fulfillment of

the safety function of structures or

systems that are needed to: (A) Shut

down the reactor * * *; (B) remove

residual heat; (C) Control the release of

radioactive material; or (D) Mitigate the

consequences of an accident,’’ if the

condition has not existed within three

years of the date of discovery.

With regard to 10 CFR 50.73, public

comment was specifically invited on

whether such historical events and

conditions should be reported (rather

than being excluded from reporting, as

proposed). Public comment was also

invited on whether the three year

exclusion of such historical events and

conditions should be extended to all

written reports required by section

50.73(a) (rather than being limited to

these two specific reporting criteria, as

proposed).

Most commenters supported the

revisions to 10 CFR 50.72 that eliminate

reporting of historical events. They

stated that no safety significance exists

for 10 CFR 50.72 reporting of historical

events.

Most commenters also supported: (1)

the elimination of written LERs for

historical events for the two cases

proposed; (2) extending the exclusion to

all written reports required under

section 50.73(a); and (3) using two years

as a cutoff point, rather than three years.

They stated that two years encompasses

one refueling cycle of operation.

Significant effort can be expended

searching back in history for historical

events. Reporting historical events more

than two years old provides a low safety

benefit and unnecessarily increases the

reporting burden. It was recognized that

three years is consistent with the time

period that performance indicators are

tracked under the new oversight

process. However, most commenters

stated that no safety significance exists

for 10 CFR 50.73 reporting of historical

events which occurred more than two

years ago.

Response: The final rule eliminates

the requirement to provide a telephone

notification or a written LER for a

historical event for the reasons

discussed above.

The cutoff date for reporting of

historical events remains at 3 years, as

was indicated in the proposed rule. The

3-year cutoff is necessary because the

NRC staff tracks performance indicators

for a period of 3 years, in order to

include a refueling outage as well as an

extended period of operations, which

provides more stable performance

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63773

indicators. The additional burden of

searching back for 3 years to determine

if a condition existed within three years

of the date of discovery, instead of only

2 years, is minimal because this type of

event is rarely identified. Thus, it is

considered justified in order to provide

better performance indicators.

Comment D (Time limits for

reporting): The proposed rule would

have continued to require reporting

within one hour after occurrence for

declaration of an Emergency Class, or

for deviation from the plant’s Technical

Specifications authorized pursuant to 10

CFR 50.54(x). Reporting of other events

that are reportable by telephone under

10 CFR 50.72 would be reportable

within 8 hours after occurrence, rather

than within 1 hour or 4 hours as was

previously required. Submittal of

written LERs would be required within

60 days after discovery, rather than

within 30 days as previously required.

Public comment was specifically

invited on the question of whether

additional levels should be used to

better correspond to particular types of

events. For example, 10 CFR 50.72

previously required reporting within 4

hours for events that involve low levels

of radioactive releases, and events

related to safety or environmental

protection that involve a press release or

notification of another government

agency. These types of events could be

maintained at 4 hours so that

information is available on a more

timely basis to respond to heightened

public concern about such events. In

another example, events related to

environmental protection are sometimes

reportable to another agency, which is

the lead agency for the matter, with a

different time limit, such as 12 hours.

These types of events could be reported

to the NRC at approximately the same

time as they are reported to the other

agency.

Most comments on the proposed rule

supported the proposal to use just three

basic levels of required reporting times

in 10 CFR 50.72 and 10 CFR 50.73 (1

hour, 8 hours, and 60 days), as indicated

in the proposed rule, in the interest of

simplicity. They indicated that

additional levels of reporting are not

needed. They also agreed with the

revised reporting times based on

importance to risk and extending the

required reporting times consistent with

the need for prompt NRC action.

Additionally, they noted that the

increased time for submittal of LERs

will allow for completion of required

engineering evaluations after event

discovery, provide for more complete

and accurate LERs, and result in fewer

LER revisions and supplemental reports.

One comment letter, from the State of

North Carolina, recommended

maintaining the required reporting time

at 4 hours for:

(1) Any airborne radioactive release

that, when averaged over a time period

of 1 hour, results in concentrations in an

unrestricted area that exceed 20 times

the applicable concentration specified

in Appendix B to Part 20, Table 2,

Column 1;

(2) Any liquid effluent release that,

when averaged over a time of 1 hour,

exceeds 20 times the applicable

concentration specified in Appendix B

to Part 20, Table 2, Column 2, at the

point of entry into the receiving waters

(i.e., unrestricted area) for all

radionuclides except tritium and

dissolved noble gases;

(3) Any event requiring the transport

of a radioactively contaminated person

to an offsite medical facility for

treatment; and

(4) Any event or situation, related to

the health and safety of the public or

onsite personnel, or protection of the

environment, for which a news release

is planned or notification to other

government agencies has been or will be

made. Such an event may include an

onsite fatality or inadvertent release of

radioactively contaminated materials.

The letter indicated that the

information from such events are of

interest to the public and public

officials. Furthermore, the State’s

Division of Radiation Protection (DRP)

provides independent advice to State

decision-makers as part of its emergency

preparedness function. Any delay in

providing the information to the DRP

may prevent or delay decisions on

public health or public announcements.

State agencies may be able to get the

information from licensees, even under

the proposed rule. However, this can be

difficult to do when an incident is

actually occurring unless the NRC’s

rules mandate the reporting within a

prompt and well-defined period of time.

Similar comments were received from

the State of Illinois regarding the ANPR.

Response: After consideration of the

comments, and the potential need for

NRC action, the final rule employs four

basic levels of required reporting times

in 10 CFR 50.72 and 10 CFR 50.73 (1

hour, 4 hours, 8 hours, and 60 days).

Although this is not as simple as using

just three levels, as was indicated in the

proposed rule, it allows more flexibility

in matching the required reporting time

to the potential need for NRC action.

The final rule requires 4-hour

reporting, if the event was not reported

in 1 hour, for an event or situation,

related to the health and safety of the

public or onsite personnel, or protection

of the environment, for which a news

release is planned or notification to

other government agencies has been or

will be made. Such an event may

include an onsite fatality or inadvertent

release of radioactively contaminated

materials. This is the same as previously

required. These reports are needed

promptly because they involve events

where there may be a need for the NRC

to respond to heightened public

concern.

The final rule also requires 4-hour

reporting, if the event was not reported

in 1 hour, for unplanned transients.

These are events where there may be a

need for the NRC to take a reasonably

prompt action, such as partially

activating its response plan to monitor

the course of the event. In summary,

they are:

(a) An event that resulted or should

have resulted in ECCS discharge into

the reactor coolant system (RCS) as a

result of a valid signal, except when it

results from and is part of a pre-planned

sequence during testing or operation.

Previously this was a 1-hour report.

(b) Initiation of a shutdown required

by the plant’s Technical Specifications.

Previously this was a 1-hour report.

(c) A reactor scram or reactor trip

when the reactor is critical, except when

it results from and is part of a preplanned sequence during testing or

operation. Previously, actuation of any

engineered safety feature (ESF),

including the reactor protection system

(RPS), was a 4-hour report.

Three criteria are deleted from § 50.72

because they are not needed in order to

obtain prompt notification of events.

They are retained in § 50.73, however,

because they are needed in order to

obtain written LERs. In summary, they

are:

(a) A natural phenomenon or other

external event that poses an actual

threat to plant safety, or significantly

hampers site personnel in the

performance of duties necessary for safe

operation. Events of this type are

captured by declaration of an

Emergency Class, which is reportable

within 1 hour.

(b) An internal event that poses an

actual threat to plant safety, or

significantly hampers site personnel in

the performance of duties necessary for

safe operation, including fires, toxic gas

releases, or radioactive releases. Events

of this type are captured by declaration

of an Emergency Class, which is

reportable within 1 hour.

(c) An airborne radioactive release, or

liquid effluent release, that exceeds

specific limits. Releases that are large

enough to warrant prompt notification

are captured by declaration of an

63774 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

Emergency Class, which is reportable

within 1 hour after the declaration.

Releases that involve a public

announcement or notification to another

agency are reportable within 1 hour

after the announcement or notification.

For the remaining events reportable

under § 50.72, the final rule requires 8-

hour reporting, if not reported in 1 hour

or 4 hours. These are events where there

may be a need for the NRC to take an

action within about a day, such as

initiating a special inspection or

investigation. In summary, they are:

(a) The plant including its principal

safety barriers being in a seriously

degraded condition, or the plant being

in an unanalyzed condition that

significantly degrades plant safety.

(b) A valid actuation of any system

listed in paragraph (b)(3)(iv)(B), except

when the actuation results from and is

part of a pre-planned sequence during

testing or reactor operation.

(c) An event or condition that at the

time of discovery could have prevented

fulfillment of the safety function of

structures or systems needed to shut

down the reactor, remove residual heat,

control the release of radioactive

material, or mitigate an accident.

(d) Transport of a radioactively

contaminated person to an offsite

medical facility.

(e) A major loss of emergency

assessment capability, offsite response

capability, or offsite communications

capability.

Comment E (Eliminate all reporting of

invalid ESF actuations): The proposed

rule would have eliminated the

requirement for a telephone notification,

under 10 CFR 50.72, for an ESF

actuation if it is an invalid automatic

actuation or an unintentional manual

actuation. It was stated that invalid

actuations are generally less significant

than valid actuations because they do

not involve plant conditions (e.g., low

reactor coolant system pressure) that

would warrant system actuation.

Instead, they result from other causes

(such as a dropped electrical lead

during testing).

The proposed rule would not have

eliminated the requirement for a written

LER for such events. It was stated that

there is still a need for reporting,

because the reports are used in making

estimates of equipment reliability

parameters, which in turn are needed to

support the Commission’s move

towards risk-informed regulation.

Most commenters indicated that

invalid ESF actuations should not be

reported under 10 CFR 50.73 unless the

actuation impacts the plant such that

other reporting criteria are

independently met. They stated that

contrary to the NRC’s expectations,

reporting of invalid actuations will not

provide the information needed to

estimate equipment reliability

parameters. This information should be

collected by other less burdensome

mechanisms, such as the Equipment

Performance and Information Exchange

(EPIX) system and Maintenance Rule

reports.

Response: The NRC disagrees with

many of the comments. Invalid

actuations do provide information

needed in estimating equipment

reliability because they constitute

unplanned demands. The response to

unplanned demands may or may not

differ significantly from those of

planned test demands. Thus, in making

reliability estimates, the results from

unplanned demands are compared

against those from planned test

demands to determine whether or not it

is appropriate to combine them. As

indicated in the Commission Paper

SECY–97–101, May 7, 1997, ‘‘Proposed

Rule, 10 CFR 50.76, Reporting

Reliability and Availability Information

for Risk-significant Systems and

Equipment,’’ Attachment 3, this is one

of the categories of information that the

NRC relies upon in order to make

equipment reliability estimates.

As also discussed in SECY–97–101,

EPIX is a voluntary program which does

not provide a break out of invalid

actuations and their results. The fact

that ESF actuations are reported in

written LERs was one of the key factors

in making the determination that the

NRC could work around weaknesses in

the EPIX data in order to develop

reliability estimates.

Reports developed under the

maintenance rule, 10 CFR 50.65, are not

submitted to the NRC.

Regardless, the Commission agrees

that a reduction in unnecessary burden

is warranted. Accordingly, the final rule

takes the following approach:

(a) The requirement to provide a

telephone notification under § 50.72 for

an invalid ESF actuation is eliminated,

as was indicated in the proposed rule.

(b) The requirement to report these

events under § 50.73 is retained.

However, the licensee is given the

option of providing a telephone report

rather than a written LER. This is far

less burdensome. In this case, the

telephone notification has the same due

date as the LER would have (60 days)

because the information is not needed

immediately.

Comment F (Eliminate reporting of

high pressure coolant injection (HPCI)

inoperability): As indicated in the 1983

Statements of Considerations for 10 CFR

50.72 and 50.73, failure or inoperability

of a single train system, such as the

HPCI system in BWRs, is considered to

constitute an ‘‘event or condition that

alone could have prevented the

fulfillment of the safety function of

structures or systems that are needed to:

(A) Shut down the reactor * * *; (B)

Remove residual heat; (C) Control the

release of radioactive material; or (D)

Mitigate the consequences of an

accident.’’

Most commenters indicated that

inoperability of HPCI does not of itself

constitute a condition that would

prevent the fulfillment of a safety

function. Therefore, there is no benefit

in reporting of HPCI inoperability if it

has no affect on the ability to fulfill a

safety function. BWR design considers

HPCI inoperability and provides

supporting systems such as reactor core

isolation cooling (RCIC), Core Spray,

and automatic depressurization system

(ADS). This is supported by the

relatively long Allowed Outage Time for

HPCI in the Standard Technical

Specifications (i.e., 14 days). If, in the

event of HPCI inoperability, it can be

shown that these systems are available

and capable of fulfilling the safety

function without HPCI, the event should

not be reportable. Reporting HPCI

inoperability in these cases has no

meaning for event reporting and appears

to be solely a data gathering exercise.

Additionally, the reactor oversight

process uses a performance indicator for

Safety System Functional Failures based

on 10 CFR 50.73 reports. These

indicators count failures of single train

systems (such as HPCI), assuming that

the event report documents a safety

system failure. Reporting HPCI

inoperability when there is no impact

on the overall capability to fulfill the

safety function (e.g., remove residual

heat) will result in an overly

conservative and detrimental

assessment of this indicator.

Response: As indicated in the 1983

Statements of Considerations for 10 CFR

50.72 and 50.73, the purpose of this

reporting criterion is to capture failure,

inoperability, etc. on the basis of a

structure or system. Thus, if an event or

condition could have prevented

fulfillment of the safety function of a

system (i.e., by that system), it is

reportable even if other system(s) could

have performed the same safety

function(s).

Also, in its assessment of plant

performance, the NRC uses a

performance indicator that includes

failure or inoperability of single train

systems such as HPCI. Thus,

elimination of the requirement to report

such events would be contrary to one of

the objectives of the rulemaking—to

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63775

maintain consistency with the NRC’s

actions to improve integrated plant

performance.

Comment G (Allow 8 hours after

discovery for telephone reporting):

Section 50.72(b)(3) states ‘‘* * * the

licensee shall notify the NRC as soon as

practical and in all cases, within eight

hours of the occurrence of any of the

following: * * *’’ The comment letter

states that this should be revised to say

‘‘* * * the licensee shall notify the NRC

as soon as practical and in all cases,

within eight hours of the occurrence or

discovery of any of the following:

* * *.’’ The addition of the term ‘‘or

discovery’’ provides for those events

that are discovered to have occurred in

the past, remained undetected for

sometime, and presently exist.

Response: The NRC disagrees.

Addition of the term ‘‘or discovery,’’ as

suggested by the comment, is not

necessary. As they have in the past, the

reporting guidelines address those

limited cases, such as discovery of an

existing but previously unrecognized

condition, where it may be necessary to

undertake an evaluation in order to

determine if an event or condition is

reportable. In other cases, where

telephone reporting is required, the

event should be reported as soon as

practical and not later than the specified

time limit.

Comment H (Eliminate telephone

reporting for non-critical scrams): Most

commenters recommended that

telephone reporting of RPS actuation

(reactor scrams) be limited to those

occurring from a critical condition.

Response: The NRC partially

disagrees. A valid scram, even from a

subcritical condition, is indicative of an

event with enough significance that the

NRC should screen and/or review it on

the day it occurs, rather than waiting for

submittal of a written LER. However,

telephone reporting under section 50.72

has been eliminated for invalid scrams

from a subcritical condition.

Comment I (Limit reporting to

conditions that do prevent fulfillment of

a required function): Regarding section

50.72(b)(2)(v), which indicates that

licensees shall report: ‘‘Any event or

condition that at the time of discovery

could have prevented the fulfillment of

the safety function of structures or

systems that are needed to: * * *,’’ this

should be revised to read as follows:

‘‘Any event or condition that at the time

of discovery is preventing the ability to

fulfill the safety function of structures or

systems that are needed to: * * *’’

This change is required to reflect the

correct tense of the existence of an event

or condition, rather than past

speculation. Because of past confusion

pertaining to the interpretation of this

area, it is suggested that further

discussion be included in the

statements of consideration explaining

that ‘‘is preventing’’ represents actual

conditions and does not imply that

further failures should be speculated.

Response: The NRC does not agree.

The term ‘‘could have prevented’’

reflects the meaning of the rule. It

means that, at the time of discovery, the

condition could have prevented

fulfillment of the function (for example,

had there been a demand for the

function). This includes but is not

limited to the case where, at the time of

discovery, the condition is actually

preventing fulfillment of the function.

This Statement of Considerations and

the reporting guidelines indicate that, in

evaluating reportability under this

criterion, it is not necessary to postulate

an additional random single failure.

Comment J (Human performance data

in LERs): Section 50.73(b)(2)(ix)(J)

previously required that the narrative

section of an LER include the following

specific information as appropriate for

the particular event:

‘‘(1) Operator actions that affected the

course of the event, including operator

errors, procedural deficiencies, or both,

that contributed to the event.

(2) For each personnel error, the

licensee shall discuss:

(i) Whether the error was a cognitive

error (e.g., failure to recognize the actual

plant condition, failure to realize which

systems should be functioning, failure

to recognize the true nature of the event)

or a procedural error;

(ii) Whether the error was contrary to

an approved procedure, was a direct

result of an error in an approved

procedure, or was associated with an

activity or task that was not covered by

an approved procedure;

(iii) Any unusual characteristics of the

work location (e.g., heat, noise) that

directly contributed to the error; and

(iv) The type of personnel involved

(i.e., contractor personnel, utilitylicensed operator, utility non-licensed

operator, other utility personnel).’’

The proposed amendment would

have changed section 50.73(b)(2)(ii)(J) to

simply state: ‘‘For each human

performance related problem that

contributed to the event, the licensee

shall discuss the cause(s) and

circumstances.’’ It was stated that the

current rule is more detailed than

necessary. Details would continue to be

provided in the reporting guidelines, as

indicated in section 5.2.1 of the draft of

Revision 2 to NUREG–1022.

Most commenters recommended that,

instead of adopting the wording in the

proposed rule, section 50.73(b)(2)(ii)(J)

be revised to state: ‘‘For each root cause

personnel error, the licensee shall

discuss the cause(s) and

circumstances.’’ They stated that the

shift from ‘‘personnel error’’ and the

implied ‘‘root cause’’ to ‘‘human

performance related problem’’ and

‘‘contributing factors’’ would greatly

increase the scope of investigation and

burden to the licensee. They also stated

that it is only appropriate to require

discussion of personnel error root

causes.

Response: The intent of the proposed

change was to clarify the requirements,

not to expand them. Accordingly, the

final rule states ‘‘For each human

performance related root cause, the

licensee shall discuss the cause(s) and

circumstances.’’ This limits the

requirement to discussion of root causes

of the event. It would not be

appropriate, or consistent with the

previous requirement discussed above,

to limit the requirement to discussion of

personnel error root causes, as opposed

to procedural deficiency root causes, for

example.

Comment K (Do not require additional

availability data in LERs): Section

50.73(b)(3) requires that the assessment

of safety consequences in an LER

include the availability of systems or

components that could have performed

the same functions as systems or

components that failed during the event.

Proposed section 50.73(b)(3)(ii) would

add a requirement that the assessment

also include the availability of systems

or components that: ‘‘Are included in

emergency or operating procedures and

could have been used to recover from

the event in case of an additional failure

in the systems actually used for

recovery.’’

Most commenters objected to this new

provision, on the grounds that it adds

significant burden without adding

value. They stated that reporting should

be based on existing plant conditions.

Emergency operating procedures

provide direction for use of many plant

systems. If additional failures must be

postulated, multiple systems would be

required to be included in the LER for

each safety function. There exists an

infinite combination of failures that

could be postulated. This unbounded

requirement would result in a large

amount of additional information that

would be of minimal use. The

assessment of the safety consequences

and implications of the event would

become cluttered with hypothetical

additional failures and possible plant

responses. Some commenters stated that

the proposed requirement would require

licensees to speculate on actions that

could have been taken, and it would

63776 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

add significant burden with no added

value.

Response: The purpose of the

proposed change was to ensure that

LERs contain sufficient information to

support a risk assessment of the event.

Usually there is enough information, or

there is nearly enough information and

the NRC staff can telephone the licensee

to obtain any additional information

needed. Section 50.73(b)(2)(6) requires

that LERs include ‘‘The name and

telephone number of a person within

the licensee’s organization who is

knowledgeable about the event and can

provide additional information

concerning the event and the plant’s

characteristics.’’ Further, Section

50.73(c) provides that the NRC may

require submittal of additional

information if necessary for complete

understanding of an unusually complex

or significant event.

However, for those events that occur

when the plant is shutdown, it has been

difficult to obtain enough information

because it cannot be assumed that

equipment that is normally operable

and available during operation is

available during plant shutdown.

Accordingly, in the final rule there is a

requirement for additional availability

information. To eliminate unnecessary

burden, the requirement for additional

availability data is limited to shutdown

events. Also, it is revised to simply

require providing the availability of

systems needed to shut down the

reactor and maintain safe shutdown

conditions, remove residual heat,

control the release of radioactive

material, or mitigate an accident. This

will eliminate potential difficulties in

deciding what combinations of failures

should be postulated for the purpose of

deciding which systems to address.

Comment L (The rule should stand

alone): Licensees must use both the rule

and NUREG–1022, Rev. 2, to determine

reportability of conditions. The rule

should be a stand-alone document

written simply enough to be understood

without the need for a 100+ page

guidance document.

Response: The NRC does not agree

that it is necessary to eliminate the

detailed event reporting guidelines and/

or include a similar level of detail in the

rule. Generally speaking, the rule

language cannot be precise enough to

cover all the situations that might be

governed by the rule and require

clarification. Furthermore, in response

to the ANPR, most commenters

expressed the need for timely guidance

on the final rule. Finally, the NRC has

reviewed the guidelines and modified

them where necessary to ensure they are

consistent with the final rule.

Comment M (The terms ‘‘significant’’

and ‘‘serious’’ are not defined in the

rule): One commenter stated that the

terms ‘‘significantly affects’’ and

‘‘seriously degraded’’ are not defined

anywhere in the proposed rule.

Response: The NRC does not agree

that it is necessary to define these terms

in the rule. The term ‘‘unanalyzed

condition that significantly affects plant

safety,’’ which was used in the proposed

rule, is changed to ‘‘unanalyzed

condition that significantly degrades

plant safety’’ in the final rule, to make

it clear that only matters with a negative

effect on safety are reportable. Its

meaning is defined by the same

examples that have served since 1983 to

define the term ‘‘unanalyzed condition

that significantly compromises plant

safety.’’ These are: (1) Multiple

functionally related safety grade

components out of service; (2)

accumulation of voids that could inhibit

the ability to adequately remove heat

from the reactor core, particularly under

natural circulation conditions; and (3)

voiding in instrument lines that results

in erroneous indication causing the

operator to misunderstand the true

condition of the plant. Also, two new

examples have been added. They are: (1)

Discovery that a system required to meet

the single failure criterion does not do

so; and (2) discovery that the fire

protection system does not protect at

least one safe shutdown train in the

event of fire in a given area. All of these

examples are discussed in the Statement

of Considerations for the final rule as

well as the reporting guidelines.

The term ‘‘condition of the nuclear

power plant, including its principal

safety barriers, being seriously

degraded’’ is defined by guidance that is

very similar to the guidance which has

defined it since 1983. Specifically, the

guidance states that this criterion

applies to material (e.g., metallurgical or

chemical) problems that cause abnormal

degradation of or stress upon the

principal safety barriers (i.e., the fuel

cladding, reactor coolant system

pressure boundary, or the containment)

such as:

(1) Fuel cladding failures in the

reactor, or in the storage pool, that

exceed expected values, or that are

unique or widespread, or that are

caused by unexpected factors.

(2) Welding or material defects in the

primary coolant system which cannot be

found acceptable under ASME Section

XI, IWB–3600, ‘‘Analytical Evaluation of

Flaws’’ or ASME Section XI, Table

IWB–3410–1, ‘‘Acceptance Standards.’’

(3) Serious steam generator tube

degradation.

(4) Low temperature over pressure

transients where the pressuretemperature relationship violates

pressure-temperature limits derived

from Appendix G to 10 CFR Part 50

(e.g., TS pressure-temperature curves).

(5) Loss of containment function or

integrity, including containment leak

rate tests where the total containment

as-found, minimum-pathway leak rate

exceeds the limiting condition for

operation (LCO) in the facility’s TS.

This guidance is discussed in further

detail below under the heading

‘‘Principal safety barrier seriously

degraded.’’

Comment N (False elevated sense of

problems): In addition to the points

discussed above under the heading

‘‘Comment E,’’ some commenters stated

that reporting of invalid actuations will

convey a false elevated sense of

problems to the general public, causing

undue alarm for situations that actually

represent little or no safety or risk

significance. Therefore, the new rule

should not require invalid actuations to

be reported.

Response: The NRC does not agree

that it is necessary to eliminate

reporting for invalid actuations in order

to avoid conveying a false elevated

sense of problems to the general public.

As discussed in the response to

Comment E, there is a need for reporting

of these events because they are used in

making estimates of equipment

reliability parameters, which in turn are

needed to support the NRC’s move

towards risk-informed regulation.

Invalid actuations have been reportable

since 1983 under the previous rules,

pursuant to both sections 50.72 and

50.73. No undue public alarm about

such invalid actuations has been

apparent to the NRC. The commenters

did not identify any specific situation or

provide any anecdotal evidence that

reporting such invalid actuations has

caused undue public alarm.

Comment O (Eliminate reporting of

missing fire barriers): One commenter

stated that the proposed rule notice at

Page 36299, first column, the example

pertaining to missing or degraded fire

barriers basically equates such

conditions with degraded principal

safety barriers (i.e., fuel cladding,

reactor coolant pressure boundary, and

containment). This is inappropriate and

should be deleted.

Response: The NRC does not agree.

The example indicates that a condition

is reportable, as an unanalyzed

condition that significantly affects plant

safety, ‘‘if fire barriers are found to be

missing such that the required degree of

separation for redundant safe shutdown

trains is lacking.’’ This would mean

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63777

that, if a fire occurs in the given area,

no safe shutdown trains would be

protected to an acceptable degree.

Because Probabilistic Risk Assessment

(PRA) studies continue to indicate that

fire is a dominant contributor to risk,

the inability to guarantee one train of

safe shutdown capability, as required, is

considered to be a condition that

significantly degrades safety.

Comment P (Applicability of the

backfit rule—no basis was stated): One

commenter stated that in the proposed

rule at Page 36303, Section VI., Backfit

Analysis, the NRC stated that 10 CFR

50.109 does not apply without giving

any basis for the claim.

Response: The discussion below,

entitled Backfit Analysis, has been

modified to provide the basis for the

conclusion that 10 CFR 50.109 does not

apply.

Comment Q (Modify ‘‘unanalyzed

condition that significantly affects

safety’’): Most commenters stated that in

section 50.72(b)(2)(ii)(B), the phrase

‘‘significantly affects plant safety’’ has

no positive or negative connotation.

Reword the section to read, ‘‘The

nuclear power plant being in an

unanalyzed condition that significantly

degrades plant safety.’’

Response: The NRC agrees. The

phrase is revised as recommended for

the reason stated.

Comment R (Recognize risksignificance factors): One commenter

stated that Section 50.73(a)(1) fails to

recognize any risk significance factors.

Response: The NRC does not agree.

Section 50.73(a)(1) is general in nature

and indicates that, unless otherwise

specified in section 50.73, the licensee

shall report an event if it occurred

within the last three years regardless of

the plant mode or power level, and

regardless of the significance of the

structure, system, or component that

initiated the event. Risk factors are

recognized elsewhere in section 50.73.

For example, the requirement to report

an event or condition that could have

prevented fulfillment of the safety

function of structures or systems is

limited to those structures or systems

that are needed to perform specific

safety functions. The list of systems for

which actuation must be reported is

based on risk-significance. Lack of

significance is the reason for the

elimination of reporting for late

surveillance tests where the equipment,

when tested, is functional. It is also the

basis for eliminating several other

requirements, such as immediate

notification under section 50.72 for

many invalid actuations.

Comment S (Modify ‘‘operation or

condition prohibited by TS’’): Section

50.73(a)(2)(i)(B) should be revised to

read, ‘‘Any operation or condition

occurring within three years of the date

of discovery which was prohibited by

the plant’s CURRENT Technical

Specifications.’’ This rewrite would

direct plants that recently converted to

Improved Standard Technical

Specifications to apply the current

requirements to the identified

condition, rather than having to

consider the previous requirements

under old Technical Specifications

which are no longer applicable.

Response: The NRC agrees. The issue

involves the following scenario. A

licensee discovers a historical operation

or condition that was prohibited by the

TS in effect at the time the operation or

condition occurred. However, the

prohibition has subsequently been

removed from the TS. The event is not

considered significant because

subsequently the operation or condition

was found to be acceptable and the

Technical Specifications have been

revised to permit it. Accordingly, the

final rule eliminates the requirement to

report such events.

Comment T (Reporting burden would

not be decreased): In addition to the

points discussed above under the

heading ‘‘Comment A,’’ one commenter

disagreed with the NRC’s assessment

that the proposed rule would represent

an overall decrease in burden. This

disagreement was based on the

following points:

(a) (Telephone notifications are less

burdensome than written LERs):

Although the proposed rule would have

decreased the number of phone-in

reports pursuant to 10 CFR 50.72, the

commenter believes this burden is very

small when compared with the burden

of processing and submitting Licensee

Event Reports (LERs) pursuant to 10

CFR 50.73.

(b) (Actuation of systems that are

currently excluded systems would

become reportable): In the proposed

rule, systems that were excluded from

reporting requirements via previous

rulemaking because they represented

little or no safety significance have been

reinstated (e.g., Reactor Water Cleanup

System). Such action will now lead to

reporting all isolations, even those with

no safety significance.

(c) (Systems not classified as ESF

would be treated as ESF): Systems that

are not classified as Engineered Safety

Features (ESF) will now be treated as

ESF (e.g., Reactor Core Isolation Cooling

System).

(d) (Invalid actuations would be

added to the reporting requirements):

Invalid actuations are now included in

the reporting requirements. The impact

of this change is that the clarifications

for what used to be reportable have been

deleted. Therefore, the proposed rule

would treat all isolations or movements

of a component as reportable regardless

of safety significance.

(e) (The requirements for human

performance data would be increased):

The scope of information requested for

human performance events has

substantially increased, going well

beyond previous direct root cause to

now include associated contributing

factors.

Response: The NRC believes that

reporting burden will be decreased for

the reasons described in the regulatory

analysis. With regard to the specific

bases cited for this comment:

(a) The NRC agrees that a telephone

notification is less burdensome than a

written LER. However, this does not

mean that the reporting burden would

be increased, or maintained, unless

there is some increase in the number of

LERs required under the final rule. This

is not the case.

(b) The NRC does not agree that the

proposed rule would have made

actuation of previously excluded

systems reportable. The previously

excluded systems are: (i) Reactor water

clean-up system; (ii) control room

emergency ventilation system; (iii)

reactor building ventilation system; (iv)

fuel building ventilation system; or (v)

auxiliary building ventilation system.

None of these appeared on the proposed

list of systems for which actuation

would be reportable.

(c) The NRC believes that system

actuations added by adoption of the

proposed list of systems are outweighed

by system actuations eliminated.

(d) The NRC does not agree that

invalid actuations are being added to

the reporting requirements, because

they were already in the reporting

requirements.

(e) See the response to Comment J.

Comment U (Incentive to disable

safety systems): In addition to the points

discussed above under the heading

‘‘Comment E,’’ one commenter

indicated that reporting of invalid

system actuations provided an incentive

to disable safety systems.

Response: The NRC does not agree

that it is necessary to eliminate

reporting for invalid actuations to avoid

creating an incentive to disable safety

systems during maintenance activities

to avoid the possibility of reporting an

inadvertent actuation.

As discussed in the response to

Comment E, there is a need for reporting

of these events because they are used in

making estimates of equipment

reliability parameters, which in turn are

63778 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

needed to support the NRC’s move

towards risk-informed regulation. Also,

in the final rule, licensees are not

required to provide an immediate

notification under Section 50.72 for an

invalid system actuation. Furthermore,

in the final rule licensees have the

option of providing a telephone

notification within 60 days, rather than

submitting a written LER, for an invalid

system actuation. These changes

provide a drastic reduction in the

burden of reporting for invalid system

actuations. This burden reduction

mitigates against any incentive to

disable safety systems during

maintenance in order to avoid the

possibility of reporting an invalid

actuation.

Comment V (Amend 10 CFR

76.120(d)(2) to allow 60 days): One

commenter noted that the NRC plans to

consider the idea of expanding the 60-

day deadline for written reports to other

regulations. The commenter

recommended amending 10 CFR

76.120(d)(2) to allow 60 days for written

reports required under that regulation.

Response: The NRC continues to plan

to evaluate the need for rulemaking to

modify 10 CFR Parts 72 and 73,

including the suggestion that 60 days be

allowed for written reports required

under 10 CFR 72.75 and 73.71. As part

of that effort, the NRC will also consider

the suggestion that 60 days be allowed

for written reports required under 10

CFR 76.120(d)(2).

Comment W (Enforcement levels):

Some commenters indicated that the

proposed characterization of

Enforcement Level III for failure to

provide a required 1-hour or 8-hour

non-emergency telephone notification is

too harsh in most cases. They indicated

that in most cases the information

provided in these non-emergency

notifications has low safety significance.

Response: As discussed further below

under the heading ‘‘Enforcement,’’ the

philosophy of the Enforcement Policy

changes is to base the significance of the

reporting violation on its impact on the

NRC’s ability to provide proper

oversight of licensee activities.

Accordingly, in some cases, Severity

Level III is appropriate for failure to

make a required telephone notification

and in other cases it is not.

Comment X (LER format and content):

One commenter recommended that the

NRC reconsider a ‘‘check the box’’

approach. The commenter indicated

that such an approach could be crafted

to make LER data entry easier, more

consistent, and less ambiguous, without

making LERs more difficult for the

general public to understand.

Response: The NRC does not believe

it is feasible to adopt a ‘‘check the box’’

in the final rule because the proposed

rule did not include a proposal along

those lines and development of a sound

system would take considerable time,

delaying issuance of the final rule.

Comment Y (Coordinate with

performance indicator efforts): One

commenter suggested careful

coordinated consideration among the

NRC staff responsible for this

rulemaking and those responsible for

performance indicator efforts to ensure

that reports submitted under 10 CFR

50.73(a)(2)(v) are not being misapplied.

Response: The NRC agrees and the

suggested coordination has taken place,

and will continue in the future as well.

As a result, it is not expected that the

NRC will misapply reports submitted

under 10 CFR 50.73(a)(2)(v).

Comment Z: One commenter

recommended that telephone

notifications due within 8 hours should

only be required when activation of the

NRC emergency response organization

is actually required.

Response: The NRC does not agree

that this is a feasible approach because

activation of the NRC’s emergency

response organization is not a simple

function of the reporting criterion under

which an event is considered to be

reportable. For example, the emergency

response organization is sometimes

activated for events which, at the time

of reporting, are considered to

correspond to lower levels of Emergency

Classes or non-emergency reporting

criteria.

Comment AA (Do not include criteria

for reporting degraded steam generator

tubes): The Statement of Considerations

for the proposed rule and the Draft

Revision 2 to NUREG–1022 would

indicate that steam generator tube

degradation is considered serious, and

thus reportable as a seriously degraded

reactor coolant system boundary, if the

tubing fails to meet specific

performance criteria involving margin

against burst and accident induced

leakage rate. Most commenters proposed

that this guidance be deleted. They

stated that the position was based on a

Draft Regulatory Guide (DG–1074,

Steam Generator Tube Integrity) that has

not been approved. Discussions between

the industry and the NRC are being held

to define the steam generator program

and Technical Specification

requirements. Some of the examples

provided in the proposed section are

contrary to agreements that have been

made between the industry and the NRC

staff. Recognizing that these agreements

are still evolving, the proposed revisions

to the rule(s) and NUREG–1022 must

agree with the final positions on steam

generator issues.

Response: The details have been

removed from the Statement of

Considerations. The details in the final

Revision 2 to NUREG–1022 have been

modified to reflect the NRC staff’s

current thinking. The guidance is

consistent with the steam generator tube

integrity performance criteria and

reporting guidelines currently under

discussion. This reporting is needed to

permit the staff to determine if further

inquiry or action might be needed

before the plant is restarted.

The NRC does not agree that it is

necessary to delay issuance of this

reporting guidance pending staff

endorsement of the NEI 97–06 initiative.

The NUREG–1022 guidance merely

provides reasonable examples of

degraded steam generator tube

conditions which the NRC needs to

evaluate. If it is determined in the future

that different detailed guidance is

needed, it can be issued at that time.

III. Discussion

1. Objectives

The purposes of sections 50.72 and

50.73 remain the same because the basic

needs remain the same. The essential

purpose of section 50.72 is ‘‘ * * * to

provide the Commission with

immediate reporting of * * *

significant events where immediate

Commission action to protect the public

health and safety may be required or

where the Commission needs timely

and accurate information to respond to

heightened public concern.’’ (48 FR

39039; August 29, 1983). Section 50.73

‘‘* * * identifies the types of reactor

events and problems that are believed to

be significant and useful to the NRC in

its effort to identify and resolve threats

to public safety. It is designed to

provide the information necessary for

engineering studies of operational

anomalies and trends and patterns

analysis of operational occurrences. The

same information can be used for other

analytic procedures that will aid in

identifying accident precursors.’’ (48 FR

33851; July 26, 1983).

The objectives of these final

amendments are as follows:

(1) To better align the reporting

requirements with the NRC’s needs for

information to carry out its safety

mission. An example is extending the

required initial reporting times for some

events, consistent with the time at

which the reports are needed for NRC

action.

(2) To reduce unnecessary reporting

burden, consistent with the NRC’s

needs. An example is eliminating the

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63779

reporting of design and analysis defects

and deviations with little or no risk-or

safety-significance.

(3) To clarify the reporting

requirements where needed. An

example is clarifying the criteria for

reporting design or analysis defects or

deviations.

(4) Any changes should be consistent

with NRC actions to improve integrated

plant assessments. For example, reports

that are needed in the assessment

process should not be eliminated.

2. Section by Section Discussion of Final

Amendments

General requirements and reportable

events [section 50.72(a)(1) and section

50.73(a)(1)]. The term ‘‘if it occurred

within 3 years of the date of discovery’’

is added to eliminate reporting for

conditions that have not existed during

the three years before discovery. Such a

historical event has less significance,

and assessing reportability for earlier

times can consume considerable

resources. For example, assume that a

procedure is found to be unclear and, as

a result, a question is raised as to

whether the plant was ever operated in

a prohibited condition. If operation in

the prohibited condition is likely, the

answer would be reasonably apparent

based on the knowledge and experience

of the plant’s operators and/or a review

of operating records for the past three

years. The effort required to review all

records older than three years in order

to rule out the possibility is not

warranted.

A sentence is added to indicate that

for an invalid actuation reported under

section 50.73(a)(2)(iv) the licensee may,

at its option, provide a telephone

notification to the NRC Operations

Center within 60 days after discovery of

the event in lieu of submitting a written

LER. For this type of event, a telephone

notification will provide the

information needed and impose less

burden than an LER.

General requirements [section

50.72(a)(5)]. The requirement to inform

the NRC of the type of report being

made (i.e., Emergency Class declared,

non-emergency 1-hour report, or nonemergency 8-hour report) is revised to

refer to paragraph (a)(1) instead of

referring to paragraph (a)(3) to correct a

typographical error.

Required initial reporting times

[sections 50.72(a)(5), (b)(1), (b)(2), and

new section 50.72(b)(3); and sections

50.73(a)(1) and (d)]. In the final

amendments, declaration of an

Emergency Class continues to be

reported immediately after notification

of appropriate State or local agencies

and not later than 1-hour after

declaration. This includes declaration of

an Unusual Event, the lowest

Emergency Class.

Deviations from Technical

Specifications authorized pursuant to 10

CFR 50.54(x) continue to be reported as

soon as practical and in all cases within

1 hour of occurrence. These two criteria

capture those events where there may be

a need for immediate action by the NRC

to protect public health and safety.

The requirement to report an event or

situation, related to the health and

safety of the public or onsite personnel,

or protection of the environment, for

which a news release is planned or

notification to other government

agencies has been renumbered from

section 50.72(b)(2)(vi) to section

50.72(b)(2)(xi). In other respects this

reporting criterion is unchanged, and

the event is reportable within 4 hours,

if not reported within 1 hour. This

provides the information at the time it

may be needed to respond to heightened

public concern.

The requirement to report a natural

phenomenon or other external event

that poses an actual threat to plant

safety or significantly hampers site

personnel in the performance of duties

necessary for safe operation in section

50.72(b)(1)(iii) is deleted. Events of this

type are captured by declaration of an

Emergency Class, which is reportable

within 1 hour.

The requirement to report an internal

event that poses an actual threat to plant

safety, or significantly hampers site

personnel in the performance of duties

necessary for safe operation, including

fires, toxic gas releases, or radioactive

releases in section 50.72(b)(1)(vi) is

deleted. Events of this type are captured

by declaration of an Emergency Class,

which is reportable within 1 hour.

The requirement to report an airborne

radioactive release or liquid effluent

release that exceeds specific limits in

section 50.72 (b)(2)(iv) is deleted.

Releases that are large enough to

warrant prompt notification are

captured by declaration of an

Emergency Class, which is reportable

within 1 hour after the declaration.

Releases that involve a news release or

notification to other government

agencies are reportable within 4 hours

of the occurrence.

The remaining non-emergency events

that are reportable by telephone under

10 CFR 50.72 are reportable as soon as

practical and in all cases within 4 hours

or 8 hours (instead of within 1 hour or

4 hours as was previously required).

This reduces the unnecessary burden of

rapid reporting, while:

(1) Capturing, within 4 hours, those

events where there may be a need for

the NRC to take a reasonably prompt

action, such as partially activating its

response plan to monitor the course of

the event.

(2) Capturing, within 8 hours, those

events where there may be a need for

the NRC to take an action within about

a day, such as initiating a special

inspection or investigation.

See the response to Comment D,

above, for further discussion.

Written LERs are due within 60 days

after discovery of a reportable event or

condition (instead of within 30 days as

was previously required). Changing the

time limit from 30 days to 60 days does

not imply that licensees should take

longer than they previously did to

develop and implement corrective

actions. They should continue to do so

on a time scale commensurate with the

safety significance of the issue.

However, for those cases where it does

take longer than thirty days to complete

a root cause analysis, this change will

result in fewer LERs that require

amendment (by submittal of an

amended report).

The term ‘‘within 30 days of the

discovery of a reportable event or

situation’’ is deleted from section

50.73(d). This provision is redundant to

the provisions of section 50.73(a)(1),

which requires that a licensee submit an

LER within 60 days after discovery of an

event described in section 50.73(a).

Retaining the time limit, which is now

60 days, in section 50.73(d) would

create a conflict with sections 20.2201

and 20.2203 which require licensees to

submit LERs for the events described in

those sections within 30 days and in

accordance with section 50.73(d).

Operation or condition prohibited by

technical specifications [section

50.73(a)(2)(i)(B)]. This criterion is

modified to eliminate reporting if the

Technical Specification is

administrative in nature. Violations of

administrative Technical Specifications

have generally not been considered to

warrant submittal of an LER, and since

1983 when the LER rule was issued the

NRC’s event reporting guidelines have

excluded almost all cases of such

reporting. This change makes the plain

wording of the rule consistent with that

guidance.

Also, this criterion is modified to

eliminate reporting if the event

consisted solely of a case of a late

surveillance test where the oversight is

corrected, the test is performed, and the

equipment is found to be functional.

This type of event has not proven to be

significant because the equipment

remained functional.

Finally, this criterion is modified to

eliminate reporting of an operation or

63780 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

condition that occurred in the past and

was prohibited at that time if, prior to

discovery of the event, the Technical

Specifications were revised such that

the operation or condition is no longer

prohibited. Such an event would have

little or no significance because, by the

time of discovery, the operation or

condition would have been determined

to be acceptable and thus permissible

under current Technical Specifications.

The NRC expects licensees to include

violations of the Technical

Specifications in their corrective action

programs, which are subject to NRC

audit.

Condition of the nuclear power plant,

including its principal safety barriers,

being seriously degraded [former

sections 50.72(b)(1)(ii) and (b)(2)(i),

replaced by new section

50.72(b)(3)(ii)(A); and section

50.73(a)(2)(ii), renumbered to

50.73(a)(2)(ii)(A)]. Previously, 10 CFR

50.72(b)(1)(ii) and (b)(2)(i) provided the

following distinction. During operation,

a seriously degraded plant, including its

principal safety barriers, was reportable

within one hour. An event discovered

while shutdown that had it been

discovered during operation would have

resulted in a seriously degraded plant,

including its principal safety barriers,

was reportable within 4 hours. The new

10 CFR 50.72(b)(3)(ii)(A) eliminates the

distinction because there are no longer

separate 1-hour and 4-hour categories of

non-emergency reports for this criterion.

There are only 8-hour non-emergency

reports for this criterion.

Unanalyzed condition that

significantly degrades plant safety

[sections 50.72(b)(1)(ii)(A) and (b)(2)(i),

replaced by new section

50.72(b)(3)(ii)(B); and section

50.73(a)(2)(ii)(A), renumbered to

50.73(a)(2)(ii)(B)]. Previously, 10 CFR

50.72(b)(1)(ii)(A) and (b)(2)(i) provided

the following distinction. During

operation, an unanalyzed condition that

significantly compromised plant safety

was reportable within 1 hour. An event

discovered while shut down that had it

been discovered during operation would

have resulted in an unanalyzed

condition that significantly

compromised plant safety was

reportable within 4 hours. The new 10

CFR 50.72(b)(2)(ii)(B) eliminates this

distinction because there are no longer

separate 1-hour and 4-hour categories of

non-emergency reports for this reporting

criterion. There are only 8-hour nonemergency reports for this criterion.

In addition, the new 10 CFR

50.72(b)(2)(ii)(B) and 50.73(a)(2)(ii)(B)

refer to a condition that significantly

degrades plant safety rather than a

condition that significantly

compromises plant safety. This is an

editorial change intended to better

reflect the nature of the criterion.

Condition that is outside the design

basis of the plant [old section

50.72(b)(2)(ii)(B); and old section

50.73(a)(2)(ii)(B)]. This criterion is

deleted. A condition outside the design

basis of the plant is still required to be

reported if it is significant enough to

qualify under one or more of the

following criteria.

Plant safety significantly degraded. If

a condition outside the design basis of

the plant (or any other unanalyzed

condition) is significant enough that, as

a result, plant safety is significantly

degraded, the condition is reportable

under sections 50.72(b)(2)(ii)(B) and

50.73(a)(2)(ii)(B) [i.e., an unanalyzed

condition that significantly degrades

plant safety].

As was previously indicated in the

1983 Statements of Considerations for

10 CFR 50.72 and 50.73, with regard to

an unanalyzed condition that

significantly compromises plant safety,

‘‘The Commission recognizes that the

licensee may use engineering judgment

and experience to determine whether an

unanalyzed condition existed. It is not

intended that this paragraph apply to

minor variations in individual

parameters, or to problems concerning

single pieces of equipment. For

example, at any time, one or more

safety-related components may be out of

service due to testing, maintenance, or

a fault that has not yet been repaired.

Any trivial single failure or minor error

in performing surveillance tests could

produce a situation in which two or

more often unrelated, safety-grade

components are out-of-service.

Technically, this is an unanalyzed

condition. However, these events

should be reported only if they involve

functionally related components or if

they significantly compromise plant

safety,’’ (48 FR 39042; August 29, 1983

and 48 FR 33856, July 26, 1983).

‘‘When applying engineering

judgment, and there is a doubt regarding

whether to report or not, the

Commission’s policy is that licensees

should make the report,’’ (48 FR 39042;

August 29, 1983).

‘‘For example, small voids in systems

designed to remove heat from the

reactor core which have been previously

shown through analysis not to be safety

significant need not be reported.

However, the accumulation of voids that

could inhibit the ability to adequately

remove heat from the reactor core,

particularly under natural circulation

conditions, would constitute an

unanalyzed condition and would be

reportable,’’ (48 FR 39042; August 29,

1983 and 48 FR 33856, July 26, 1983).

‘‘In addition, voiding in instrument

lines that results in an erroneous

indication causing the operator to

misunderstand the true condition of the

plant is also an unanalyzed condition

and should be reported,’’ (48 FR 39042;

August 29, 1983 and 48 FR 33856, July

26, 1983).

Furthermore, beyond the examples

given in 1983, examples of reportable

events include discovery that a system

required to meet the single failure

criterion does not do so.

In another example, if fire barriers are

found to be missing, such that the

required degree of separation for

redundant safe shutdown trains is

lacking, the event is reportable. On the

other hand, if a fire wrap, to which the

licensee has committed, is missing from

a safe shutdown train but another safe

shutdown train is available in a

different fire area, protected such that

the required separation for safe

shutdown trains is still provided, the

event is not reportable.

Structure or system not capable of

performing its specified safety function.

If a design or analysis defect or

deviation (or any other event or

condition) is significant enough that, as

a result, a structure or system is not

capable of performing its specified

safety functions, the condition is

reportable under sections 50.72(b)(3)(v)

and 50.73(a)(2)(v) [i.e., an event or

condition that could have prevented the

fulfillment of the safety function of

structures or systems that are needed to:

shut down the reactor * * *; remove

residual heat; control the release of

radioactive material; or mitigate the

consequences of an accident].

For example, in one case an annual

inspection indicated that some bearings

were wiped or cracked on both

emergency diesel generators (EDGs).

Although the EDGs were running prior

to the inspection, the event was

reportable because there was reasonable

doubt about the ability of the EDGs to

operate for an extended period of time,

as required.

Train inoperable longer than allowed.

If a design or analysis defect or

deviation (or any other event or

condition) is significant enough that, as

a result, one train of a multiple train

system controlled by the plant’s TS is

not capable of performing its specified

safety functions for a period of time

longer than allowed by the TS, the

condition is reportable under section

50.73(a)(2)(i)(B) [i.e., an operation or

condition prohibited by TS].

For example, if it is found that an

exciter panel for one EDG lacks

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63781

1The LCO typically employs La, which is defined

in Appendix J to 10 CFR Part 50 as the maximum

allowable containment leak rate at pressure Pa, the

calculated peak containment internal pressure

related to the design basis accident. Minimumpathway leak rate means the minimum leak rate

that can be attributed to a penetration leakage path;

for example, the smaller of either the inboard or

outboard valve’s individual leak rates.

appropriate seismic restraints because of

a design, analysis, or construction

inadequacy and, as a result, there is

reasonable doubt about the EDG’s ability

to perform its specified safety functions

during and after a Safe Shutdown

Earthquake (SSE), the event would be

reportable.

Or, for example, if it is found that a

loss of offsite power could cause a loss

of instrument air and, as a result, there

is reasonable doubt about the ability of

one train of the auxiliary feedwater

system to perform its specified safety

functions for certain postulated steam

line breaks, the event would be

reportable.

Principal safety barrier seriously

degraded. If a condition outside the

design basis of the plant (or any other

event or condition) is significant enough

that, as a result, a principal safety

barrier is seriously degraded, it is

reportable under sections

50.72(b)(3)(ii)(A) and 50.73(a)(2)(ii)(A)

[i.e., any event or condition that results

in the condition of the nuclear power

plant, including its principal safety

barriers, being seriously degraded]. This

reporting criterion applies to material

(e.g., metallurgical or chemical)

problems that cause abnormal

degradation of or stress upon the

principal safety barriers (i.e., the fuel

cladding, reactor coolant system

pressure boundary, or the containment)

such as:

(i) Fuel cladding failures in the

reactor, or in the storage pool, that

exceed expected values, or that are

unique or widespread, or that are

caused by unexpected factors.

(ii) Welding or material defects in the

primary coolant system which cannot be

found acceptable under ASME Section

XI, IWB–3600, ‘‘Analytical Evaluation of

Flaws’’ or ASME Section XI, Table

IWB–3410–1, ‘‘Acceptance Standards.’’

(iii) Serious steam generator tube

degradation.

(iv) Low temperature over pressure

transients where the pressuretemperature relationship violates

pressure-temperature limits derived

from Appendix G to 10 CFR Part 50

(e.g., TS pressure-temperature curves).

(v) Loss of containment function or

integrity, including containment leak

rate tests where the total containment

as-found, minimum-pathway leak rate

exceeds the limiting condition for

operation (LCO) in the facility’s TS.1

Common cause inoperability of

independent trains or channels. If a

condition outside the design basis of the

plant (or any other event or condition)

is significant enough that, as a result,

independent trains or channels become

inoperable, it would be reportable under

section 50.73(a)(2)(vii) [i.e., an event

where a single cause or condition

caused independent trains or channels

to become inoperable]. For example, in

one case it was found that independent

circuit breakers, required to operate

after a Loss-of-Coolant Accident

(LOCA), were not qualified for the

expected radiation levels (and were thus

considered inoperable). In another

example, a wiring error caused

independent containment isolation

valves to be incapable of properly

closing (i.e., they would not close tightly

because they would stop closing based

on limit switch operation rather than

torque).

Single Cause that Could Have

Prevented Fulfillment of the Safety

Functions of Trains or Channels in

Different Systems. Finally, a condition

outside the design basis of the plant (or

any other event or condition) would be

reportable if it is significant enough

that, as a result of a single cause, it

could have prevented the fulfillment of

a safety function for two or more trains

or channels in different systems that are

needed to:

(1) Shut down the reactor and

maintain it in a safe shutdown

condition;

(2) Remove residual heat;

(3) Control the release of radioactive

material; or

(4) Mitigate the consequences of an

accident.

This new criterion is contained in

sections 50.73(a)(2)(ix)(A) and (B), as

discussed below.

Single Cause that Could Have

Prevented Fulfillment of the Safety

Functions of Trains or Channels in

Different Systems. [new sections

50.73(a)(2)(ix)(a) and (B)]. This new

criterion requires reporting any event or

condition that as a result of a single

cause could have prevented the

fulfillment of a safety function for two

or more trains or channels in different

systems that are needed to:

(1) Shut down the reactor and

maintain it in a safe shutdown

condition;

(2) Remove residual heat;

(3) Control the release of radioactive

material; or

(4) Mitigate the consequences of an

accident.

Events covered by this new criterion

may include cases of procedural error,

equipment failure, and/or discovery of a

design, analysis, fabrication,

construction, and/or procedural

inadequacy. However, licensees are not

required to report an event pursuant to

this criterion if the event results from:

(1) A shared dependency among

trains or channels that is a natural or

expected consequence of the approved

plant design; or

(2) Normal and expected wear or

degradation.

Subject to the two exclusions stated

above, this criterion captures those

events where a single cause could have

prevented the fulfillment of the safety

function of multiple trains or channels,

but the event:

(1) Would not be captured by

§§ 50.73(a)(2)(v) and 50.72(b)(3)(v)

[event or condition that could have

prevented fulfillment of the safety

function of structures and systems

needed to . . .] because the affected

trains or channels are in different

systems; and

(2) Would not be captured by

§ 50.73(a)(2)(vii) [common cause

inoperability of independent trains or

channels] because the affected trains or

channels are either:

(i) Not assumed to be independent in

the plant’s safety analysis; or

(ii) Not both considered to be

inoperable.

This new criterion is closely related to

§§ 50.73(a)(2)(v) and 50.72(b)(3)(v)

[event or condition that could have

prevented fulfillment of the safety

function of structures and systems

needed to: shut down the reactor and

maintain it in a safe shutdown

condition; remove residual heat; control

the release of radioactive material; or

mitigate the consequences of an

accident]. Specifically:

The meaning of the term ‘‘could have

prevented the fulfillment of the safety

function’’ is essentially the same for this

new criterion as it is for §§ 50.73(a)(2)(v)

and 50.72(b)(3)(v) [i.e., there was a

reasonable expectation of preventing the

fulfillment of the safety function(s)

involved]. However, in contrast to

§§ 50.73(a)(2)(v) and 50.72(b)(3)(v),

reporting under this new criterion

applies to trains or channels in different

systems. Thus, for this new criterion,

the safety function that is affected may

be different in different trains or

channels.

In contrast to §§ 50.73(a)(2)(v) and

50.72(b)(3)(v), reporting under this new

criterion applies only to a single cause.

Also, in contrast to §§ 50.73(a)(2)(v) and

50.72(b)(3)(v), this new criterion does

not apply to an event that results from

63782 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

2Or, alterantively, there was reasonable doubt

that the safety function would have been fulfilled

if the affected trains has been called upon to

perform them.

a shared dependency among trains or

channels that is a natural or expected

consequence of the approved plant

design. For example, this new criterion

does not capture failure of a common

electrical power supply that disables

Train A of AFW and Train A of High

Pressure Safety Injection (HPSI),

because their shared dependency on the

single power supply is a natural or

expected consequence of the approved

plant design.

Similar to §§ 50.73(a)(2)(v) and

50.72(b)(3)(v), this new criterion does

not capture events or conditions that

result from normal and expected wear

or degradation. For example, consider

pump bearing wear that is within the

normal and expected range. In the case

of two pumps in different systems, this

new criterion categorically excludes

normal and expected wear. In the case

of two pumps in the same system,

normal and expected wear should be

adequately addressed by normal plant

operating and maintenance practices

and thus should not indicate a

reasonable expectation of preventing

fulfillment of the safety function of the

system.

This criterion pertains only to written

LERs required by 10 CFR 50.73.

Telephone notifications are not required

under this criterion.

It is estimated that the combination of

removing the previous requirement to

report a condition outside the design

basis of the plant and adding this

criterion will, on balance, result in

fewer reports. In addition, the events

reportable under this criterion are

events that would likely have been

considered reportable under the

previous requirement to report a

condition outside the design basis of the

plant.

An example of an event that would be

reportable under this criterion is as

follows. During testing, two

containment isolation valves failed to

function as a result of improper air gaps

in the solenoid operated valves that

controlled the supply of instrument air

to the containment isolation valves. The

valves were powered from the same

electrical division. Thus,

§ 50.73(a)(2)(vii) [common cause

inoperability of independent trains or

channels] would not apply. The two

valves isolated fluid process lines in

two different systems. Thus

§ 50.73(a)(2)(v) [condition that could

have prevented fulfillment of the safety

function of a structure or system] would

apply only if engineering judgment

indicates there was a reasonable

expectation of preventing fulfillment of

the safety function for redundant valves

within the same system.2 However, this

new criterion would certainly apply if a

single cause (such as a design

inadequacy) induced the improper air

gaps, thus preventing fulfillment of the

safety function of two trains or channels

in different systems.

Another example of an event

reportable under this criterion is as

follows. A motor operated valve in one

train of a system was found with a crack

75 percent through the stem. Although

the valve stem did not fail, engineering

evaluation indicated that further

cracking would occur which could have

prevented fulfillment of its safety

function. As a result, the train was not

considered capable of performing its

specified safety function and the valve

stem was replaced with a new one.

The root cause was determined to be

environmentally assisted stress

corrosion cracking which resulted from

installation of an inadequate material

some years earlier. The same inadequate

material had been installed in a similar

valve in a different system at the same

time. The similar valve was exposed to

similar environmental conditions as the

first valve.

The condition is reportable under this

new criterion if engineering judgment

indicates that there was a reasonable

expectation of preventing fulfillment of

the safety function of both affected

trains. This depends on details such as

whether the second valve stem was also

significantly degraded and, if not,

whether any future degradation of the

second valve stem would have been

discovered and corrected, as a result of

routine maintenance programs, before it

could become problematic.

Additional examples may be found in

event reporting guidelines in NUREG–

1022, Revision 2.

Condition not covered by the plant’s

operating and emergency procedures

[former section 50.72(b)(1)(ii)(C); and

former section 50.73(a)(2)(ii)(C)]. This

criterion is deleted because it does not

result in worthwhile reports aside from

those that are captured by other

reporting criteria such as:

(1) An unanalyzed condition that

significantly degrades plant safety;

(2) An event or condition that could

have prevented the fulfillment of the

safety function of structures or systems

that are needed to: shutdown the reactor

and maintain it in a safe shutdown

condition; remove residual heat; control

the release of radioactive material; or

mitigate the consequences of an

accident;

(3) An event or condition that results

in the condition of the nuclear power

plant, including its principal safety

barriers, being seriously degraded;

(4) An operation or condition

prohibited by the plant’s TS;

(5) An event or condition that results

in actuation of an ESF;

(6) An event that poses an actual

threat to the safety of the nuclear power

plant or significantly hampers site

personnel in the performance of duties

necessary for the safe operation of the

nuclear power plant.

External event that poses an actual

threat or significantly hampers

personnel [former section

50.72(b)(1)(iii), deleted; and section

50.73(a)(2)(iii)]. This criterion requires

reporting a natural phenomenon or

other external event that poses an actual

threat to plant safety, or significantly

hampers site personnel in the

performance of duties necessary for safe

operation. Section 50.72(b)(1)(iii) is

deleted because it is redundant to

section 50.72(a)(1)(i). That is, events of

this type are captured by declaration of

an Emergency Class, which is reportable

within 1 hour. Section 50.73(a)(2)(iii) is

retained in order to ensure submittal of

an LER. This provision is not redundant

because there is no criterion in section

50.73 that generally requires an LER for

declaration of an Emergency Class.

System actuation [old sections

50.72(b)(1)(iv) and (b)(2)(ii), replaced by

new sections 50.72(b)(2)(iv)(A),

(b)(2)(iv)(B), and (b)(3)(iv); and section

50.73(a)(2)(iv)]. Previously, sections

50.72(b)(1)(iv) and (b)(2)(ii) provided

the following distinction: an event that

results or should have resulted in ECCS

discharge into the reactor coolant

system as a result of a valid signal was

reportable within 1 hour; any other

engineered safety feature (ESF)

actuation, including reactor protection

system (RPS) actuation, was reportable

within 4 hours. The new 10 CFR

50.72(b)(2)(iv)(A) requires reporting an

event that results or should have

resulted in ECCS discharge into the

reactor coolant system as a result of a

valid signal within 4 hours. The new

section 50.72(b)(2)(iv)(B) requires

reporting a reactor scram during critical

operation within 4 hours. The new

section 50.72(b)(3)(iv) requires reporting

other ESF actuations within 8 hours.

See the response to Comment D, above,

for further discussion.

The new section 50.72(b)(2)(iv)

eliminates telephone reporting for

invalid actuations, except for actuation

of the RPS when the reactor is critical.

These events are not significant and

thus telephone reporting is not needed.

The final amendments do not eliminate

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63783

the requirement for reporting of an

invalid actuation under 10 CFR 50.73.

There is still a need for reporting of

these events because they are used in

making estimates of equipment

reliability parameters, which in turn are

needed to support the Commission’s

move towards risk-informed regulation.

However, for an invalid actuation

reported under section 50.73(a)(2)(iv),

other than actuation of the RPS when

the reactor is critical, section 50.73(a)(1)

provides the option of making a

telephone report to the NRC Operations

Center within 60 days instead of

submitting a written LER. The telephone

report is far less burdensome. Sixty days

is an appropriate time because the

information is not needed immediately.

(See the response to Comment E above

for further discussion of this need.)

Previously, the rules generally

required reporting the actuation of any

ESF including the RPS. The final rule,

instead, generally requires reporting for

actuation of specific listed systems.

These systems are:

(1) Reactor protection system (RPS)

including: reactor scram or reactor trip.

(2) General containment isolation

signals affecting containment isolation

valves in more than one system or

multiple main steam isolation valves

(MSIVs).

(3) Emergency core cooling systems

(ECCS) for pressurized water reactors

(PWRs) including: high-head,

intermediate-head, and low-head

injection systems and the low pressure

injection function of residual (decay)

heat removal systems.

(4) ECCS for boiling water reactors

(BWRs) including: high-pressure and

low-pressure core spray systems; highpressure coolant injection system; low

pressure injection function of the

residual heat removal system.

(5) BWR reactor core isolation cooling

system; isolation condenser system; and

feedwater coolant injection system.

(6) PWR auxiliary or emergency

feedwater system.

(7) Containment heat removal and

depressurization systems, including

containment spray and fan cooler

systems.

(8) Emergency ac electrical power

systems, including: emergency diesel

generators (EDGs); hydroelectric

facilities used in lieu of EDGs at the

Oconee Station; and BWR dedicated

Division 3 EDGs.

(9) Emergency service water (ESW)

systems that do not normally run and

that serve as ultimate heat sinks. ESW

system actuations are reportable under

section 50.73 only.

This approach provides for consistent

reporting of actuations for these highly

risk-significant systems. At the same

time, it eliminates reporting of

actuations for systems of lesser risksignificance, such as actuation of

ventilation systems that are considered

to be ESFs.

Section 50.72 excludes reporting for

an actuation that resulted from and was

part of a pre-planned sequence during

testing or reactor operation. It further

excludes reporting of an invalid

actuation, except for a reactor scram or

reactor trip when the reactor is critical.

A valid actuation is one that results

from either a ‘‘valid signal’’ or an

intentional manual initiation. A ‘‘valid

signal’’ is one that results from actual

plant conditions or parameters

satisfying the requirements for system

actuation. An invalid actuation is one

that does not meet the criteria for being

valid.

Section 50.73 also excludes reporting

for an actuation that resulted from and

was part of a pre-planned sequence

during testing or reactor operation. It

further excludes reporting of an invalid

actuation that occurred when the system

was properly removed from service or

an invalid actuation that occurred after

the safety function had been already

completed.

For those invalid actuations which

are reportable under section 50.73, a

licensee may provide a telephone

notification within 60 days, rather than

submitting an LER. This option to

provide a telephone notification rather

than an LER does not apply, however,

to a reactor scram or reactor trip that

occurs while the reactor is critical.

Event or condition that could have

prevented the fulfillment of the safety

function of structures or systems that

are needed to: shut down the reactor

and maintain it in a safe shutdown

condition; remove residual heat; control

the release of radioactive material; or

mitigate the consequences of an

accident [former section 50.72(b)(2)(iii),

replaced by new sections 50.72(b)(3)(v)

and (b)(3)(vi); and sections 50.73(a)(2)(v)

and (a)(2)(vi)]. The phrase ‘‘event or

condition that alone could have

prevented the fulfillment of the safety

function of structures or systems * * *’’

is clarified by deleting the word

‘‘alone’’. This clarifies the requirements

by more clearly reflecting the principle

that it is necessary to consider other

existing plant conditions in determining

the reportability of an event or

condition under this criterion. For

example, if one train of a two train

system is incapable of performing its

safety function for one reason, and the

other train is incapable of performing its

safety function for a different reason, the

event is reportable.

The term ‘‘at the time of discovery’’ is

added to section 50.72(b)(3)(v) to

eliminate telephone notification for a

condition that no longer exists or no

longer has an effect on required safety

functions. For example, it might be

discovered that at some time in the past

both trains of a two train system were

incapable of performing their safety

function, but the condition was

subsequently corrected and no longer

exists. In another example, while the

plant is shutdown, it might be

discovered that during a previous

period of operation a system was

incapable of performing its safety

function, but the system is not currently

required to be operable. These events

are considered significant, and an LER

is required, but there is no need for

telephone notification.

A new paragraph, section

50.72(b)(3)(vi) is added to clarify section

50.72. The new paragraph explicitly

states that telephone reporting is not

required under section 50.72(b)(2)(v) for

single failures if redundant equipment

in the same system was operable and

available to perform the required safety

function. That is, although one train of

a system may be incapable of

performing its safety function, reporting

is not required under this criterion if

that system is still capable of performing

the safety function. This is the same

principle that was and continues to be

stated explicitly in section

50.73(a)(2)(vi) with regard to written

LERs.

Airborne radioactive release or liquid

effluent release [former section

50.72(b)(2)(iv), deleted; and section

50.73(a)(2)(viii), retained; and former

section 50.73(a)(2)(ix), deleted]. These

criteria require reporting releases of

radioactive material at a very low level

because, for a power reactor, such a

release would indicate a breakdown in

the licensee’s programs to control

releases—not because of the impact of

such a release.

Section 50.72(b)(2)(iv) is deleted

because immediate notification is not

needed for releases at such a low level.

Declaration of an Emergency Class,

which occurs at a somewhat higher (but

still low) level, captures releases that are

large enough to warrant immediate

notification. Declaration of an

Emergency Class is reportable within 1

hour under section 50.72(a)(1)(i).

Section 50.73(a)(2)(viii) is retained in

order to ensure submittal of an LER.

Even if the release is very small, the

NRC needs to review the event and

consider whether action is needed to

ensure the cause is addressed at other

plants as appropriate. There is no

criterion in section 50.73 that generally

63784 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

3The examples refer to those published in

NUREG–1600, ‘‘General Statement of Policy and

Procedure for NRC Enforcement Actions,’’ dated

May 1, 2000.

requires an LER for declaration of an

Emergency Class.

Section 50.73(a)(2)(vix) is deleted

because it is not correct. It indicated

that reporting under section

50.73(a)(2)(viii) satisfied the

requirements of section 20.2203(a)(3).

However, some events captured by

section 20.2203(a)(3) are not captured

by section 50.73(a)(2)(viii).

Internal event that poses an actual

threat or significantly hampers

personnel [former section

50.72(b)(1)(vi), deleted; and section

50.73(a)(2)(x)]. This criterion requires

reporting an internal event that poses an

actual threat to plant safety, or

significantly hampers site personnel in

the performance of duties necessary for

safe operation, including fires, toxic gas

releases, or radioactive releases. Section

50.72(b)(1)(vi) is deleted because it is

redundant to section 50.72(a)(1)(i). That

is, events of this type are captured by

declaration of an Emergency Class,

which is reportable within 1 hour.

Section 50.73(a)(2)(x) is retained in

order to ensure submittal of an LER.

This provision is not redundant because

there is no criterion in section 50.73 that

generally requires an LER for

declaration of an Emergency Class.

Major loss of emergency assessment

capability, offsite response capability, or

communication capability [former

section 50.72(b)(2)(v), replaced by new

section 50.72(b)(3)(xii)]. The new

section is modified by adding the word

‘‘offsite’’ in front of the term

‘‘communications capability’’ to make it

clear that the requirement does not

apply to internal plant communication

systems.

Contents of LERs [section

50.73(b)(2)(ii)(F)]. Paragraph (F) is

revised to correct the address of the

NRC Library.

Spent fuel storage cask problems

[former sections 50.72(b)(2)(vii) and

72.216(a)(1), (a)(2), and (b)]. The

provisions of section 50.72(b)(2)(vii) are

deleted because these reporting criteria

are redundant to the reporting criteria

contained in sections 72.216(a)(1) and

(a)(2). Repetition of the same reporting

criteria in different sections of the rules

added unnecessary complexity and was

inconsistent with the current practice in

other areas, such as reporting of

safeguards events as required by section

73.71.

Sections 72.216(a)(1) and (a)(2) place

upon general licensees the same

reporting criteria as are placed on

specific licensees under sections

72.75(b)(2) and (b)(3). To avoid

duplication in Part 72, sections

72.216(a)(1) and (a)(2) are deleted and

section 72.216(c) is abridged to simply

require that the general licensee shall

make initial and written reports in

accordance with sections 72.74 and

72.75. These changes eliminate a

reference in section 72.216(a) to section

50.72(b)(2)(vii), now deleted, which had

established the time limit for initial

notification by general licensees. The

same time limit is placed on general

licensees by including them within the

scope of section 72.75(b). Section

72.216(b) is also deleted because its

requirements for a written report are

encompassed by section 72.75(d)(2).

Exemptions [section 50.73(f)]. The

provisions of this section are deleted

because the exemption provisions in

section 50.12 provide for granting of

exemptions when they are warranted.

Including another, section-specific

exemption provision in section 50.73

adds unnecessary complexity to the

rules.

3. Revisions to Event Reporting

Guidelines in NUREG–1022

A report, NUREG–1022, Revision 2,

‘‘Event Reporting Guidelines, 10 CFR

50.72 and 50.73,’’ is being made

available concurrently with the final

amendments to 10 CFR 50.72 and 50.73.

The report is available for inspection in

the NRC Public Document Room or it

may be viewed and downloaded

electronically via the interactive

rulemaking web site established by the

NRC for this rulemaking, as discussed

above under the heading ADDRESSES.

Single copies may be obtained from the

contact listed above under the heading

FOR FURTHER INFORMATION CONTACT. In

the report, guidance that is considered

to be new or different in a meaningful

way, relative to that provided in

NUREG–1022, Revision 1, is indicated

by underlining the appropriate text.

4. Reactor Oversight

The NRC is implementing revisions to

the process for oversight of operating

reactors, including inspection,

assessment, and enforcement processes.

In connection with this effort, the NRC

has considered the kinds of event

reports that would be eliminated by the

final rules and concluded that the

changes are consistent with the

oversight process.

In connection with the proposed rule,

public comment was invited on whether

or not this is the case. In particular, it

was requested that if any examples to

the contrary are known they be

identified. None were identified.

5. Enforcement

The NRC intends to modify its

existing enforcement policy in

connection with the final amendments

to sections 50.72 and 50.73. The

philosophy of the changes is to base the

significance of the reporting violation

on the impact on the NRC’s ability to

provide proper oversight of licensee

activities. For example, a late report

may impact the ability of the NRC to

fulfill its obligations of fully

understanding issues that are required

to be reported in order to accomplish its

public health and safety mission, which

in many cases involves reacting to

reportable issues or events. As such, the

NRC intends to revise the Enforcement

Policy, NUREG–1600 3 as follows:

(1) Supplement I.C—Examples of

Severity Level III violations.

(a) Example 11 will be revised to read

as follows—A failure to provide the

required 1-hour telephone notification

of an emergency action taken pursuant

to 10 CFR 50.54(x).

(b) An additional example will be

added that will read as follows—A

failure to provide a required 1-hour, 4-

hour or 8-hour non-emergency

telephone notification pursuant to 10

CFR 50.72, that substantially impacts

agency response.

(c) An additional example will be

added that will read as follows—A late

4-hour or 8-hour notification that

substantially impacts agency response.

(2) Supplement I.D—Examples of

Severity Level IV violations.

(a) Example 4, will be revised to read

as follows—A failure to provide a

required 60-day written LER pursuant to

10 CFR 50.73.

These changes in the Enforcement

Policy will be consistent with the

overall objective of the rule change of

better aligning the reporting

requirements with the NRC’s reporting

needs. The Enforcement Policy changes

will correlate the Severity Level of the

infractions with the relative importance

of the information needed by the NRC.

Section IV.A.3 of the Enforcement

Policy provides that the Severity Level

of an untimely report may be reduced

depending on the individual

circumstances. In deciding whether the

Severity Level should be reduced for an

untimely 1-hour, 4-hour, or 8-hour nonemergency report, the impact that the

failure to report had on any agency

response will be considered. For

example, if a delayed 8-hour reportable

event impacted the timing of a followup

inspection that was deemed necessary,

then the Severity Level will not

normally be reduced. Similarly, a late

notification that delayed the NRC’s

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63785

ability to perform an engineering

analysis of a condition to determine if

additional regulatory action was

necessary will generally not be

considered for disposition at a reduced

Severity Level.

6. Electronic Reporting

The NRC is currently in the process

of implementing an electronic

document management and reporting

program, known as the Agency Wide

Document Access and Management

System (ADAMS) that will provide for

electronic submittal of many types of

reports, including LERs. Accordingly,

no separate rulemaking effort to provide

for electronic submittal of LERs is

necessary.

7. State Input

Many States (Agreement States and

Non-Agreement States) have agreements

with power reactors to inform the States

of plant issues. State reporting

requirements are frequently triggered by

NRC reporting requirements.

Accordingly, the NRC sought State

comment on issues related to the

proposed amendments via letters to

State Liaison Officers as well as by a

specific request in the Federal Register

notice on the proposed rule. Comments

on the proposed rule were received from

one State agency, as discussed above

under the heading ‘‘Comment D.’’

8. Plain Language

The President’s Memorandum dated

June 1, 1998, entitled, ‘‘Plain Language

in Government Writing,’’ directed that

the Federal Government’s writing be in

plain language. The NRC requested

comments on the proposed rule

specifically with respect to the clarity

and effectiveness of the language used.

A number of suggestions aimed at

improving the clarity and effectiveness

of the language used were received and

incorporated into the final rule.

IV. Environmental Impact: Categorical

Exclusion

The NRC has determined that this

proposed regulation is the type of action

described in categorical exclusion 10

CFR 51.22(c)(3)(iii). Therefore, neither

an environmental impact statement nor

an environmental assessment has been

prepared for this proposed regulation.

V. Backfit Analysis

The NRC has determined that the

backfit rule, 10 CFR 50.109, does not

apply to information collection and

reporting requirements such as those

contained in the final rule because they

do not impose backfits as defined in 10

CFR 50.109(a)(1). Therefore, a backfit

analysis has not been prepared.

However, as discussed below, the NRC

has prepared a regulatory analysis for

the proposed rule, which examines the

costs and benefits of the proposed

requirements in this rule. The

Commission regards the regulatory

analysis as a disciplined process for

assessing information collection and

reporting requirements to determine

that the burden imposed is justified in

light of the potential safety significance

of the information to be collected.

VI. Regulatory Analysis

The NRC prepared a draft regulatory

analysis for the proposed rule to

examine the costs and benefits of the

alternatives considered by the NRC, and

public comments on this analysis were

requested in connection with the

proposed rule. As discussed above

under the heading ‘‘Comment T,’’ some

commenters disagreed with the

proposition that the rule would reduce

reporting burden. These comments were

addressed by incorporating changes into

the final rule, such that the assumptions

in the draft regulatory analysis are

sustained, and no changes have been

made to the regulatory analysis. The

regulatory analysis is available for

inspection in the NRC Public Document

Room or it may be viewed and

downloaded electronically via the

interactive rulemaking web site

established by NRC for this rulemaking,

as discussed above under the heading

ADDRESSES. Single copies may be

obtained from the contact listed above

under the heading FOR FURTHER

INFORMATION CONTACT.

VII. Paperwork Reduction Act

Statement

This final rule amends information

collection requirements that are subject

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.). This rule has

been reviewed and approved by the

Office of Management and Budget,

approval numbers 3150–0011 and 3150–

0104.

The annual public reporting burden

for the currently existing reporting

requirements in 10 CFR 50.72 and 50.73

is estimated to average about 700 hours

per nuclear power reactor, including the

time for reviewing instructions,

searching existing data sources,

gathering and maintaining the data

needed, and completing and reviewing

the information collection. It is

estimated that the proposed

amendments would impose a one-time

implementation burden of about 200

hours per reactor. The recurring annual

information collection burden is

estimated to be reduced by 132 hours

per reactor.

Send comments on any aspect of this

information collection, including

suggestions for reducing this burden, to

the Records Management Branch (T–

6E6), U.S. Nuclear Regulatory

Commission, Washington, DC 20555–

0001 or by Internet electronic mail to

BJS1@NRC.GOV; and to the Desk

Officer, Office of Information and

Regulatory Affairs, NEOB–10202,

(3150–0011 AND 3150–0104); Office of

Management and Budget, Washington,

DC 20503.

Public Protection Notification

If a means used to impose an

information collection does not display

a currently valid OMB control number,

the NRC may not conduct or sponsor,

and a person is not required to respond

to, an information collection.

VIII. Regulatory Flexibility Act

Certification

In accordance with the Regulatory

Flexibility Act (5 U.S.C. 605(b)), the

Commission certifies that this rule does

not have a significant economic impact

on a substantial number of small

entities. This proposed rule affects 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’’ set

forth in the Regulatory Flexibility Act or

the size standards established by the

NRC (10 CFR 2.810).

IX. Small Business Regulatory

Enforcement Fairness Act

In accordance with the Small

Business Regulatory Enforcement

Fairness Act of 1996, the NRC has

determined that this action is not a

major rule and has verified this

determination with the Office of

Information and Regulatory Affairs of

OMB.

X. National Technology Transfer and

Advancement Act

The National Technology Transfer

and Advancement Act of 1995, Pub. L.

104–113, requires that Federal agencies

use technical standards developed or

adopted by voluntary consensus

standards bodies unless the use of such

a standard is inconsistent with

applicable law or otherwise impractical.

There are no consensus standards

regarding the reporting of safety

information by nuclear power plant

licensees to regulatory authorities that

would apply to the requirements

imposed by this rule. Thus, the

provisions of the Act do not apply to

this rule.

63786 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

1Other requirements for immediate notification

of the NRC by licensed operating nuclear power

rectors are contained elsewhere in this chapter, in

particular §§ 20.1906, 20.2202, 50.36, 72.216, and

73.71.

2These Emergency Classes are addressed in

Appendix E of this part.

3Commercial telephone number of the NRC

Operations Center is (301) 816–5100.

4Requirements for ERDS are addressed in

Appendix E, Section VI.

XI. Final Amendments

List of Subjects

10 CFR Part 50

Antitrust, Classified information,

Criminal penalties, Fire prevention,

Intergovernmental relations, Nuclear

power plants and reactors, Radiation

protection, Reactor siting criteria,

Reporting and recordkeeping

requirements.

10 CFR Part 72

Criminal penalties, Manpower

training programs, Nuclear materials,

Occupational safety and health,

Reporting and recordkeeping

requirements, Security measures, Spent

fuel.

For the reasons set out 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 adopting the following amendments

to 10 CFR Part 50 and 10 CFR Part 72.

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. 444, as amended (42 U.S.C.

2132, 2133, 2134, 2135, 2201, 2232, 2233,

2236, 2239, 2282); secs. 201, as amended,

202, 206, 88 Stat. 1242, as amended, 1244,

1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.

95–601, sec. 10, 92 Stat. 2951 (42 U.S.C.

5851). Section 50.10 also issued under

secs. 101, 185, 68 Stat. 955 as amended

(42 U.S.C. 2131, 2235), sec. 102, Pub. L.

91–190, 83 Stat. 853 (42 U.S.C. 4332).

Sections 50.13, 50.54(D.D.), 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.

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. Section 50.72 is amended by

revising paragraphs (a) and (b) to read

as follows:

§ 50.72 Immediate notification

requirements for operating nuclear power

reactors.

(a) General requirements.1 (1) Each

nuclear power reactor licensee licensed

under § 50.21(b) or § 50.22 of this part

shall notify the NRC Operations Center

via the Emergency Notification System

of:

(i) The declaration of any of the

Emergency Classes specified in the

licensee’s approved Emergency Plan; 2

or

(ii) Those non-emergency events

specified in paragraph (b) of this section

that occurred within three years of the

date of discovery.

(2) If the Emergency Notification

System is inoperative, the licensee shall

make the required notifications via

commercial telephone service, other

dedicated telephone system, or any

other method which will ensure that a

report is made as soon as practical to the

NRC Operations Center.3

(3) The licensee shall notify the NRC

immediately after notification of the

appropriate State or local agencies and

not later than one hour after the time the

licensee declares one of the Emergency

Classes.

(4) The licensee shall activate the

Emergency Response Data System

(ERDS) 4 as soon as possible but not

later than one hour after declaring an

Emergency Class of alert, site area

emergency, or general emergency. The

ERDS may also be activated by the

licensee during emergency drills or

exercises if the licensee’s computer

system has the capability to transmit the

exercise data.

(5) When making a report under

paragraph (a)(1) of this section, the

licensee shall identify:

(i) The Emergency Class declared; or

(ii) Paragraph (b)(1), ‘‘One-hour

reports,’’ paragraph (b)(2), ‘‘Four-hour

reports,’’ or paragraph (b)(3), ‘‘Eighthour reports,’’ as the paragraph of this

section requiring notification of the nonemergency event.

(b) Non-emergency events—(1) Onehour reports. If not reported as a

declaration of an Emergency Class

under paragraph (a) of this section, the

licensee shall notify the NRC as soon as

practical and in all cases within one

hour of the occurrence of any deviation

from the plant’s Technical

Specifications authorized pursuant to

§ 50.54(x) of this part.

(2) Four-hour reports. If not reported

under paragraphs (a) or (b)(1) of this

section, the licensee shall notify the

NRC as soon as practical and in all

cases, within four hours of the

occurrence of any of the following:

(i) The initiation of any nuclear plant

shutdown required by the plant’s

Technical Specifications.

(ii)–(iii) [Reserved]

(iv)(A) Any event that results or

should have resulted in emergency core

cooling system (ECCS) discharge into

the reactor coolant system as a result of

a valid signal except when the actuation

results from and is part of a pre-planned

sequence during testing or reactor

operation.

(B) Any event or condition that results

in actuation of the reactor protection

system (RPS) when the reactor is critical

except when the actuation results from

and is part of a pre-planned sequence

during testing or reactor operation.

(v)–(x) [Reserved]

(xi) Any event or situation, related to

the health and safety of the public or

onsite personnel, or protection of the

environment, for which a news release

is planned or notification to other

government agencies has been or will be

made. Such an event may include an

onsite fatality or inadvertent release of

radioactively contaminated materials.

(3) Eight-hour reports. If not reported

under paragraphs (a), (b)(1) or (b)(2) of

this section, the licensee shall notify the

NRC as soon as practical and in all cases

within eight hours of the occurrence of

any of the following:

(i) [Reserved]

(ii) Any event or condition that results

in:

(A) The condition of the nuclear

power plant, including its principal

safety barriers, being seriously

degraded; or

(B) The nuclear power plant being in

an unanalyzed condition that

significantly degrades plant safety.

(iii) [Reserved]

(iv)(A) Any event or condition that

results in valid actuation of any of the

systems listed in paragraph (b)(3)(iv)(B)

of this section, except when the

actuation results from and is part of a

pre-planned sequence during testing or

reactor operation.

(B) The systems to which the

requirements of paragraph (b)(3)(iv)(A)

of this section apply are:

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63787

5Actuation of the RPS when the reactor is critical

is reportable under paragraph (b)(2)(iv)(B) of this

section.

(1) Reactor protection system (RPS)

including: Reactor scram and reactor

trip. 5

(2) General containment isolation

signals affecting containment isolation

valves in more than one system or

multiple main steam isolation valves

(MSIVs).

(3) Emergency core cooling systems

(ECCS) for pressurized water reactors

(PWRs) including: High-head,

intermediate-head, and low-head

injection systems and the low pressure

injection function of residual (decay)

heat removal systems.

(4) ECCS for boiling water reactors

(BWRs) including: High-pressure and

low-pressure core spray systems; highpressure coolant injection system; low

pressure injection function of the

residual heat removal system.

(5) BWR reactor core isolation cooling

system; isolation condenser system; and

feedwater coolant injection system.

(6) PWR auxiliary or emergency

feedwater system.

(7) Containment heat removal and

depressurization systems, including

containment spray and fan cooler

systems.

(8) Emergency ac electrical power

systems, including: Emergency diesel

generators (EDGs); hydroelectric

facilities used in lieu of EDGs at the

Oconee Station; and BWR dedicated

Division 3 EDGs.

(v) Any event or condition that at the

time of discovery could have prevented

the fulfillment of the safety function of

structures or systems that are needed to:

(A) Shut down the reactor and

maintain it in a safe shutdown

condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident.

(vi) Events covered in paragraph

(b)(3)(v) of this section may include one

or more procedural errors, equipment

failures, and/or discovery of design,

analysis, fabrication, construction, and/

or procedural inadequacies. However,

individual component failures need not

be reported pursuant to paragraph

(b)(3)(v) of this section if redundant

equipment in the same system was

operable and available to perform the

required safety function.

(vii)–(xi) [Reserved]

(xii) Any event requiring the transport

of a radioactively contaminated person

to an offsite medical facility for

treatment.

(xiii) Any event that results in a major

loss of emergency assessment capability,

offsite response capability, or offsite

communications capability (e.g.,

significant portion of control room

indication, Emergency Notification

System, or offsite notification system).

* * * * *

3. Section 50.73 is amended by

revising sections (a), (b)(2)(ii)(F),

(b)(2)(ii)(J), (b)(3), (d), and (e) and by

removing and reserving paragraph (f) to

read as follows:

§ 50.73 Licensee event report system.

(a) Reportable events. (1) The holder

of an operating license for a nuclear

power plant (licensee) shall submit a

Licensee Event Report (LER) for any

event of the type described in this

paragraph within 60 days after the

discovery of the event. In the case of an

invalid actuation reported under

§ 50.73(a)(2)(iv), other than actuation of

the reactor protection system (RPS)

when the reactor is critical, the licensee

may, at its option, provide a telephone

notification to the NRC Operations

Center within 60 days after discovery of

the event instead of submitting a written

LER. Unless otherwise specified in this

section, the licensee shall report an

event if it occurred within three years of

the date of discovery regardless of the

plant mode or power level, and

regardless of the significance of the

structure, system, or component that

initiated the event.

(2) The licensee shall report:

(i)(A) The completion of any nuclear

plant shutdown required by the plant’s

Technical Specifications.

(B) Any operation or condition which

was prohibited by the plant’s Technical

Specifications except when:

(1) The Technical Specification is

administrative in nature;

(2) The event consisted solely of a

case of a late surveillance test where the

oversight was corrected, the test was

performed, and the equipment was

found to be capable of performing its

specified safety functions; or

(3) The Technical Specification was

revised prior to discovery of the event

such that the operation or condition was

no longer prohibited at the time of

discovery of the event.

(C) Any deviation from the plant’s

Technical Specifications authorized

pursuant to § 50.54(x) of this part.

(ii) Any event or condition that

resulted in:

(A) The condition of the nuclear

power plant, including its principal

safety barriers, being seriously

degraded; or

(B) The nuclear power plant being in

an unanalyzed condition that

significantly degraded plant safety.

(iii) Any natural phenomenon or other

external condition that posed an actual

threat to the safety of the nuclear power

plant or significantly hampered site

personnel in the performance of duties

necessary for the safe operation of the

nuclear power plant.

(iv)(A) Any event or condition that

resulted in manual or automatic

actuation of any of the systems listed in

paragraph (a)(2)(iv)(B) of this section,

except when:

(1) The actuation resulted from and

was part of a pre-planned sequence

during testing or reactor operation; or

(2) The actuation was invalid and;

(i) Occurred while the system was

properly removed from service; or

(ii) Occurred after the safety function

had been already completed.

(B) The systems to which the

requirements of paragraph (a)(2)(iv)(A)

of this section apply are:

(1) Reactor protection system (RPS)

including: reactor scram or reactor trip.

(2) General containment isolation

signals affecting containment isolation

valves in more than one system or

multiple main steam isolation valves

(MSIVs).

(3) Emergency core cooling systems

(ECCS) for pressurized water reactors

(PWRs) including: high-head,

intermediate-head, and low-head

injection systems and the low pressure

injection function of residual (decay)

heat removal systems.

(4) ECCS for boiling water reactors

(BWRs) including: high-pressure and

low-pressure core spray systems; highpressure coolant injection system; low

pressure injection function of the

residual heat removal system.

(5) BWR reactor core isolation cooling

system; isolation condenser system; and

feedwater coolant injection system.

(6) PWR auxiliary or emergency

feedwater system.

(7) Containment heat removal and

depressurization systems, including

containment spray and fan cooler

systems.

(8) Emergency ac electrical power

systems, including: emergency diesel

generators (EDGs); hydroelectric

facilities used in lieu of EDGs at the

Oconee Station; and BWR dedicated

Division 3 EDGs.

(9) Emergency service water systems

that do not normally run and that serve

as ultimate heat sinks.

(v) Any event or condition that could

have prevented the fulfillment of the

safety function of structures or systems

that are needed to:

63788 Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations

(A) Shut down the reactor and

maintain it in a safe shutdown

condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident.

(vi) Events covered in paragraph

(a)(2)(v) of this section may include one

or more procedural errors, equipment

failures, and/or discovery of design,

analysis, fabrication, construction, and/

or procedural inadequacies. However,

individual component failures need not

be reported pursuant to paragraph

(a)(2)(v) of this section if redundant

equipment in the same system was

operable and available to perform the

required safety function.

(vii) Any event where a single cause

or condition caused at least one

independent train or channel to become

inoperable in multiple systems or two

independent trains or channels to

become inoperable in a single system

designed to:

(A) Shut down the reactor and

maintain it in a safe shutdown

condition;

(B) Remove residual heat;

(C) Control the release of radioactive

material; or

(D) Mitigate the consequences of an

accident.

(viii)(A) Any airborne radioactive

release that, when averaged over a time

period of 1 hour, resulted in airborne

radionuclide concentrations in an

unrestricted area that exceeded 20 times

the applicable concentration limits

specified in appendix B to part 20, table

2, column 1.

(B) Any liquid effluent release that,

when averaged over a time period of 1

hour, exceeds 20 times the applicable

concentrations specified in appendix B

to part 20, table 2, column 2, at the

point of entry into the receiving waters

(i.e., unrestricted area) for all

radionuclides except tritium and

dissolved noble gases.

(ix)(A) Any event or condition that as

a result of a single cause could have

prevented the fulfillment of a safety

function for two or more trains or

channels in different systems that are

needed to:

(1) Shut down the reactor and

maintain it in a safe shutdown

condition;

(2) Remove residual heat;

(3) Control the release of radioactive

material; or

(4) Mitigate the consequences of an

accident.

(B) Events covered in paragraph

(a)(2)(ix)(A) of this section may include

cases of procedural error, equipment

failure, and/or discovery of a design,

analysis, fabrication, construction, and/

or procedural inadequacy. However,

licensees are not required to report an

event pursuant to paragraph (a)(2)(ix)(A)

of this section if the event results from:

(1) A shared dependency among

trains or channels that is a natural or

expected consequence of the approved

plant design; or

(2) Normal and expected wear or

degradation.

(x) Any event that posed an actual

threat to the safety of the nuclear power

plant or significantly hampered site

personnel in the performance of duties

necessary for the safe operation of the

nuclear power plant including fires,

toxic gas releases, or radioactive

releases.

(b) * * *

(2) * * *

(ii) * * *

(F) The Energy Industry Identification

System component function identifier

and system name of each component or

system referred to in the LER.

(1) The Energy Industry Identification

System is defined in: IEEE Std 803–1983

(May 16, 1983) Recommended Practice

for Unique Identification in Power

Plants and Related Facilities—

Principles and Definitions.

(2) IEEE Std 803–1983 has been

approved for incorporation by reference

by the Director of the Federal Register

in accordance with 5 U.S.C. 552(a) and

1 CFR part 51.

(3) A notice of any changes made to

the material incorporated by reference

will be published in the Federal

Register. Copies may be obtained from

the Institute of Electrical and

Electronics Engineers, 445 Hoes Lane,

P.O. Box 1331, Piscataway, NJ 08855–

1331. IEEE Std 803–1983 is available for

inspection at the NRC’s Technical

Library, which is located in the Two

White Flint North Building, 11545

Rockville Pike, Rockville, Maryland

20852–2738; and at the Office of the

Federal Register, 800 North Capitol

Street, Suite 700, NW, Washington, DC

2001.

* * * * *

(J) For each human performance

related root cause, the licensee shall

discuss the cause(s) and circumstances.

* * * * *

(3) An assessment of the safety

consequences and implications of the

event. This assessment must include:

(i) The availability of systems or

components that could have performed

the same function as the components

and systems that failed during the event,

and

(ii) For events that occurred when the

reactor was shutdown, the availability

of systems or components that are

needed to shutdown the reactor and

maintain safe shutdown conditions,

remove residual heat, control the release

of radioactive material, or mitigate the

consequences of an accident.

* * * * *

(d) Submission of reports. Licensee

Event Reports must be prepared on

Form NRC 366 and submitted to the

U.S. Nuclear Regulatory Commission, as

specified in § 50.4.

(e) Report legibility. The reports and

copies that licensees are required to

submit to the Commission under the

provisions of this section must be of

sufficient quality to permit legible

reproduction and micrographic

processing.

(f) [Reserved]

* * * * *

PART 72—LICENSING

REQUIREMENTS FOR THE

INDEPENDENT STORAGE OF SPENT

NUCLEAR FUEL AND HIGH-LEVEL

RADIOACTIVE WASTE

4. The authority citation for part 72

continues to read as follows:

Authority: Secs. 51, 53, 57, 62, 63, 65, 69,

81, 161, 182, 183, 184, 186, 189, 68 Stat. 929,

930, 932, 933, 934, 935, 954, 955, as

amended, sec. 234, 83 Stat. 444, as amended

(42 U.S.C. 2071, 2073, 2077, 2092, 2093,

2095, 2099, 2111, 2201, 2232, 2233, 2234,

2236, 2237, 2238, 2282); sec. 274, Pub. L. 86–

373, 73 Stat. 688, as amended (42 U.S.C.

5841, 5842, 5846); Pub. L. 95–601, sec. 10, 92

Stat. 2951 as amended by Pub. L. 102–486,

sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851);

sec. 102, Pub. L. 91–190, 83 Stat. 853 (42

U.S.C. 4332); secs. 131, 132, 133, 135, 137,

141, Pub. L. 97–425, 96 Stat. 2229, 2230,

2232, 2241, sec. 148, Pub. L. 100–203, 101

Stat. 1330–235 (42 U.S.C. 10151, 10152,

10153, 10155, 10157, 10161, 10168).

Section 72.44(g) also issued under

secs. 142(b) and 148(c), (d), Pub. L. 100–

203, 101 Stat. 1330–232, 1330–236 (42

U.S.C. 10162(b), 10168(c), (d)). Section

72.46 also issued under sec. 189, 68

Stat. 955 (42 U.S.C. 2239); sec. 134, Pub.

L. 97–425, 96 Stat. 2230 (42 U.S.C.

10154). Section 72.96(d) also issued

under sec. 145(g), Pub. L. 100–203, 101

Stat. 1330–235 (42 U.S.C. 10165(g)).

Subpart J also issued under secs. 2(2),

2(15), 2(19), 117(a), 141(h), Pub. L. 97–

425, 96 Stat. 2202, 2203, 2204, 2222,

2224, (42 U.S.C. 10101, 10137(a),

10161(h)). Subparts K and L are also

issued under sec. 133, 98 Stat. 2230 (42

U.S.C. 10153) and sec. 218(a), 96 Stat.

2252 (42 U.S.C. 10198).

5. Section 72.216 is revised to read as

follows:

§ 72.216 Reports.

(a) [Reserved]

Federal Register / Vol. 65, No. 207 / Wednesday, October 25, 2000 / Rules and Regulations 63789

(b) [Reserved]

(c) The general licensee shall make

initial and written reports in accordance

with §§ 72.74 and 72.75.

Dated at Rockville, Maryland, this 18th day

of October, 2000.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook

Secretary of the Commission.

[FR Doc. 00–27283 Filed 10–24–00; 8:45 am]

BILLING CODE 7590–01–P

NATIONAL CREDIT UNION

ADMINISTRATION

12 CFR Part 792

The Production of Nonpublic Records

and Testimony of NCUA Employees in

Legal Proceedings and the Privacy Act

AGENCY: National Credit Union

Administration (NCUA).

ACTION: Final rule.

SUMMARY: NCUA is making minor and

technical revisions to its regulation

implementing the Privacy Act of 1974

(PA). The revised rule is updated to

conform to current law governing the

method an individual may use to

establish his or her identity to obtain

access to protected records and the

requirements for the release of medical

records. The revised rule changes time

limits so that they conform more closely

to those under the Freedom of

Information Act (FOIA) and clarifies

that the agency maintains four, rather

than three, systems of records subject to

exemptions under the PA. The revision

also updates the rule to reflect

organizational changes within NCUA

and corrects cross-references in Subpart

C, which governs the production of

nonpublic records and the testimony of

NCUA employees in legal proceedings.

DATES: This rule is effective November

24, 2000.

FOR FURTHER INFORMATION CONTACT:

Dianne M. Salva, Staff Attorney,

Division of Operations, Office of

General Counsel, at the above address or

telephone: (703) 518–6540.

SUPPLEMENTARY INFORMATION: On June

12, 2000, NCUA published a proposal to

revise its PA regulation, announcing

that it would accept comments from the

public for a period of 60 days. 65 FR

36797, June 12, 2000. All comments

submitted to NCUA were favorable.

As part of a government-wide

initiative, NCUA reviewed its practices

related to privacy and personal

information in federal records. In its

review of the agency systems of records,

it identified several changes in record

keeping practices and agency

organization. As a result, NCUA revised

its systems notices to make them clearer

and simpler and to eliminate

redundancies. 65 FR 3486, January 21,

2000. Now, as a result of its review of

the PA regulation, NCUA is updating

the regulation to reflect current law,

terminology and organizational

functions and clarify which of its

systems of records are subject to PA

exemptions. The changes are minor and

technical and streamline the regulation

to make it clearer and simpler to use.

Regulatory Procedures

Paperwork Reduction Act

This regulation imposes no additional

information collection, reporting or

record keeping requirements.

Regulatory Flexibility Act

Pursuant to section 605(b) of the

Regulatory Flexibility Act (5 U.S.C.

605(b)), NCUA certifies that this rule

will not have a significant economic

impact on a substantial number of small

entities. NCUA expects that this

amended rule will not: (1) Have

significant secondary or incidental

effects on a substantial number of small

entities; or (2) create any additional

burden on small entities. These

conclusions are based on the fact that

the regulation’s changes are minor and

are intended to simplify and clarify

agency record keeping and disclosure

procedures. Accordingly, a regulatory

flexibility analysis is not required.

Executive Order 13132

Executive Order 13132 encourages

independent regulatory agencies to

consider the impact of their regulatory

actions on state and local interests. In

adherence to fundamental federalism

principles, NCUA, an independent

regulatory agency as defined in 44

U.S.C. 3502(5), voluntarily complies

with the executive order. This rule is

procedural in nature and will not have

substantial direct effects on the states,

on the relationship between the national

government and the states, or on the

distribution of power and

responsibilities among the various

levels of government. NCUA has

determined that the rule does not

constitute a policy that has federalism

implications for purposes of the

executive order.

Small Business Regulatory Enforcement

Fairness Act

The Small Business Regulatory

Enforcement Fairness Act of 1996 (Pub.

L. 104–121) provides generally for

congressional review of agency rules. A

reporting requirement is triggered in

instances where NCUA issues a final

rule as defined by section 551 of the

Administrative Procedures Act. 5 U.S.C.

551. NCUA has recommended to the

Office of Management and Budget that

it determine that this is not a major rule,

and is awaiting its determination.

The Treasury and General Government

Appropriations Act, 1999—Assessment

of Federal Regulations and Policies on

Families

The NCUA has determined that this

rule will not affect family well-being

within the meaning of section 654 of the

Treasury and General Government

Appropriations Act, 1999, Public Law

105–277, 112 Stat. 2681 (1998).

Agency Regulatory Goal

NCUA’s goal is to promulgate clear

and understandable regulations that

impose minimal regulatory burden. We

find that these amendments are

understandable and minimally

intrusive.

List of Subjects in 12 CFR Part 792

Administrative practice and

procedure, Archives and records, Credit

unions, Information, Records.

By the National Credit Union

Administration Board on October 19, 2000.

Becky Baker,

Secretary of the Board.

For the reasons set out in the

preamble, the NCUA amends 12 CFR

part 792 as follows:

PART 792—REQUESTS FOR

INFORMATION UNDER THE FREEDOM

OF INFORMATION ACT AND PRIVACY

ACT, AND BY SUBPOENA; SECURITY

PROCEDURES FOR CLASSIFIED

INFORMATION

1. The authority citation for part 792

continues to read as follows:

Authority: 5 U.S.C. 301, 552, 552a, 552b;

12 U.S.C. 1752a(d), 1766, 1789, 1795f; E.O.

12600, 52 FR 23781, 3 CFR, 1987 Comp., p.

235; E.O. 12958, 60 FR 19825, 3 CFR, 1995

Comp., p. 333.

§ 792.41 [Amended]

2. In § 792.41, remove ‘‘§ 792.4(b)(2)’’

and add, in its place, ‘‘§ 792.32.’’

§ 792.47 [Amended]

3. In § 792.47(b), remove ‘‘§ 792.5’’

and add, in its place, ‘‘§ 792.19.’’

§ 792.49 [Amended]

4. In § 792.49, in the definition of

Nonpublic records, remove ‘‘§ 792.3’’

and add, in its place, ‘‘§ 792.11’’.

5. Amend § 792.55 by revising

paragraph (a)(3) to read as follows: