Reporting Requirements for Nuclear Power Reactors

NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 72

RIN 3150-AF98

Reporting Requirements for Nuclear Power Reactors

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission is proposing to amend the

event reporting requirements for nuclear power reactors: to update the

current rules, including reducing or eliminating the reporting burden

associated with events of little or no safety significance; and to

better align the rules with the NRC's needs for information 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 it is needed for prompt NRC action. Also, a

draft report, NUREG-1022, Revision 2, is being made available for

public comment concurrently with the proposed amendments.

DATES: Submit comments on or before September 20, 1999. Comments

received after this date will be considered if it is practical to do

so, but the Commission is able to ensure consideration only for

comments received on or before this date.

ADDRESSES: Mail comments to: Secretary, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001. ATTN: Rulemakings and

Adjudications Staff.

Deliver comments to: 11555 Rockville Pike, Rockville, Maryland,

between 7:30 a.m. and 4:15 p.m. Federal workdays.

Electronic comments may be provided via the NRC's interactive

rulemaking website through the NRC home page (http://www.nrc.gov). From

the home page, select ``Rulemaking'' from the tool bar at the bottom of

the page. The interactive rulemaking website can then be accessed by

selecting ``Rulemaking Forum.'' This site provides the ability to

upload comments as files (any format), if your web browser supports

that function. For information about the interactive rulemaking

website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail

[email protected]

Certain documents related to this rulemaking, including comments

received, the transcripts of public meetings held, the draft regulatory

analysis and the draft report NUREG-1022, Revision 2 may be examined at

the NRC Public Document Room, 2120 L Street, NW, (Lower Level),

Washington, DC. These same documents also may be viewed and downloaded

electronically via the interactive rulemaking web site established by

NRC for this rulemaking.

FOR FURTHER INFORMATION CONTACT: Dennis P. Allison, Office of Nuclear

Reactor Regulation, Washington, DC 20555-0001, telephone (301) 415-

1178, e-mail [email protected].

SUPPLEMENTARY INFORMATION:

Contents

I. Background

II. Rulemaking Initiation

III. Analysis of Comments

IV. Discussion

1. Objectives of Proposed Amendments

2. Discussion of Proposed Amendments

3. Revisions to Reporting Guidelines in NUREG-1022

4. Reactor Oversight

5. Reporting of Historical Problems

6. Reporting of Component Problems

7. Enforcement

8. Electronic Reporting

9. Schedule

10. State Input

V. Environmental Impact: Categorical Exclusion

VI. Backfit Analysis

VII. Regulatory Analysis

VIII. Paperwork Reduction Act Statement

IX. Regulatory Flexibility Certification

X. Proposed Amendments

I. Background

Section 50.72 has been in effect, with minor modifications, since

1983. Its essential purpose 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 has also been in effect, with minor modification,

since 1983. Its essential purpose is to identify ``* * * 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).

II. Rulemaking Initiation

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

whether the NRC should proceed with rulemaking to modify the event

reporting requirements in 10 CFR 50.72, ``Immediate notification

requirements for operating nuclear power reactors,'' and 50.73,

``Licensee event report system,'' and several concrete proposals were

provided for comment.

A public meeting was held to discuss the ANPR at NRC Headquarters

on August 21, 1998. The ANPR was also discussed, along with other

topics, at a public meeting on the role of industry in nuclear

regulation in Rosemont, Illinois on September 1, 1998. The public

comment period on the ANPR closed on September 21, 1998. A comment from

the Nuclear Energy Institute (NEI) proposed conducting ``table top

exercises'' early in the development and review process to test key

parts of the requirements and guidance for clarity and consistency.

That comment was accepted and a third public meeting was held on

November 13, 1998 to discuss issues of clarity and consistency in the

contemplated approach. Transcripts of these meetings are available for

inspection in the NRC Public Document Room or they may be viewed and

downloaded electronically via the interactive rulemaking web site

established by NRC for this rulemaking, as discussed above under the

heading

[[Page 36292]]

ADDRESSES. Single copies may be obtained from the contact listed above

under the heading For Further Information Contact.

III. Analysis of Comments

The comment period for the ANPR expired September 21, 1998. Twenty-

one comment letters were received, representing comments from sixteen

nuclear power plant licensees (utilities), two organizations of

utilities, two States and one public interest group. A list of comment

letters is provided below. The comment letters expressed support for

amending the rules along the general lines of the objectives discussed

in the ANPR. Most of the letters also provided specific recommendations

for changes to the contemplated amendments discussed in the ANPR. In

addition to the written comments received, the ANPR has been the

subject of three public meetings as discussed above under the heading

BACKGROUND, and comments made at those meetings have also been

considered.

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 ANPR; or applicable to another area or activity outside the

scope of sections 50.72 and 50.73).

Comment 1: Several comments recommended amending 10 CFR 50.73 to

allow 60 days (instead of the current 30 days) for submittal of

Licensee Event Reports (LERs). They indicated that this would allow a

more reasonable time to determine the root causes of events and lead to

fewer amended reports.

Response: The comments are accepted for the reason stated above.

The proposed rule would change the time limit to 60 days.

Comment 2: Two comments suggested a need to establish starting

points for reporting time clocks that are clear and not subject to

varied interpretations.

Response: The reporting guidelines in this area have been reviewed

for clarity. Some editorial clarifications are proposed in section 2.5

of the draft of Revision 2 to NUREG-1022, which is being made available

for public comment concurrently with the proposed rule, as discussed

below under the heading ``Revisions to Reporting Guidelines in NUREG-

1022.''

Comment 3: Many comments opposed adopting a check the box approach

for human performance and other information in LERs (as was proposed in

the ANPR, with the objective of reducing reporting burden). They

indicated that adopting a check the box approach would result in

substantial implementation problems, and recommended continuing to rely

on the narrative description which provides adequate information. One

comment opposed the idea of a check the box approach on the grounds

that it would make LERs more difficult for the general public to

understand. A few comments supported the check the box approach.

Response: The intent of the check the box approach was to reduce

the effort required in reporting; however, the majority of comments

indicate this would not be the case. Accordingly, the proposed rule

does not reflect adoption of a check the box approach.

Comment 4: Several comments opposed codifying the current

guidelines for reporting human performance information in LERs (i.e.,

adding the detailed guidelines to the rule, as was proposed in the

ANPR). They recommended leaving the rule unchanged in this regard,

indicating that sufficient information is being provided under the

current rule and guidelines.

Response: The comments are partially accepted. The proposed rule

would not codify the reporting guidelines (as proposed in the ANPR) for

the reasons stated above.

However, the proposed rule would simplify the requirement. It is

not necessary to specify the level of detail provided in the current

rule. Accordingly, the amended paragraph would simply require a

discussion of the causes and circumstances for any human performance

related problems that contributed to the event. 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. This draft report is

being made available for public comment concurrently with the proposed

rule, as discussed below under the heading ``Revisions to Reporting

Guidelines in NUREG-1022.''

Comment 5: Several comments opposed codifying a list of specific

systems for which actuation must be reported (by naming the systems in

10 CFR 50.72 and 50.73, as was proposed in the ANPR). They indicated

that a system's contribution to risk can vary widely from plant to

plant, which precludes construction of a valid universal list. They

recommended that, instead, actuation be reported only for those systems

that are specified to be engineered safety features (ESFs) in the final

safety analysis report (FSAR).

Response: The proposed rule would include a list of systems for

which actuation would be reported. However, the concern is recognized

and public comment will be specifically invited on several alternatives

to the proposed rule.

Comment 6: Several comments opposed changing the criteria in 10 CFR

50.72 and 50.73 which require reporting any event or condition that

alone could have prevented the fulfillment of the safety function of

structures or systems * * *. The change proposed in the ANPR would have

substituted the phrase ``alone or in combination with other existing

conditions'' for the word ``alone'' in this criterion. The comments

indicated that this would add confusion, the rule as currently worded

is sufficiently clear, and the need to consider other existing plant

conditions in evaluating reportability is understood and uniformly

implemented. They recommended leaving the rule unchanged in this

regard.

Response: The comments are partially accepted. The requirement

would not be changed by substituting the phrase ``alone or in

combination with other existing conditions'' for the word ``alone'' in

this criterion (as proposed in the ANPR).

However, the proposed amendments would change the rules by deleting

the word ``alone,'' so that they would require reporting ``any event or

condition that could have prevented fulfillment of the safety function

of structures or systems * * *.'' This would simplify the wording,

rather than making it more complicated. It is not intended to change

the meaning of the requirement, but to make the meaning more apparent

in the wording of the rule. The following points, which are relevant to

this question, would continue to be made clear in the reporting

guidelines. See section 3.2.7 of the draft of Revision 2 to NUREG-1022,

which is being made available for public comment concurrently with the

proposed rule, as discussed below under the heading ``Revisions to

Reporting Guidelines in NUREG-1022.''

(1) It is not necessary to assume an additional random single

failure in evaluating reportability. (If such an assumption were

necessary, inoperability of a single train would generally be

reportable under this criterion.)

(2) It is necessary to consider other existing conditions in

determining reportability. (For example, if Train A fails at a time

when Train B is out of service for maintenance, the event is

reportable.)

(3) The event is reportable regardless of whether or not a system

was called upon to perform its safety function. (For example, if an

emergency core cooling system [ECCS] was incapable of performing its

specified safety

[[Page 36293]]

functions, the event is reportable even if there was no call for the

ECCS function.)

(4) The event is reportable regardless of whether or not a

different system was capable of performing the safety function. (For

example, if the onsite power system failed, the event is reportable

even if the offsite power system was available and capable of

performing its safety functions.)

Comment 7: Several comments recommended changing 10 CFR 50.72 and

50.73 to exclude reporting an invalid actuation of an ESF. (An invalid

actuation is one that does not result from a plant condition that

warrants ESF initiation.)

Response: The comments are partially accepted. The proposed

amendments would eliminate the requirement for telephone notification

of an invalid actuation under 10 CFR 50.72. Invalid actuations are

generally less significant than valid actuations because they do not

involve plant conditions (e.g., low reactor coolant system pressure)

conditions that would warrant system actuation. Instead, they result

from other causes such as a dropped electrical lead during testing).

However, the proposed amendments would not eliminate the

requirement for a written report of an invalid actuation under 10 CFR

50.73. There is still a need for reporting of invalid actuations

because they are needed to make estimates of equipment reliability

parameters, which in turn are needed to support the Commission's move

towards risk-informed regulation. This is discussed further in a May 7,

1997 Commission paper, SECY-97-101, ``Proposed Rule, 10 CFR 50.76,

Reporting Reliability and Availability Information for Risk-significant

Systems and Equipment,'' Attachment 3.

Comment 8: Several comments recommended changing 10 CFR 50.72 and

50.73 to limit certain reports to current events and conditions. That

is, they recommended that an event or condition that could have

prevented the fulfillment of the safety function of structures or

systems * * * be reported:

(1) By telephone under 10 CFR 50.72(b)(2)(iii) only if it currently

exists, and

(2) By written LER under 10 CFR 50.73(a)(2)(v) only if it existed

within the previous two years.

For a ``historical'' event or condition of this type (i.e., one

which might have been significant at one time but has since been

corrected) there is less significance than there is for a current event

and, thus, immediate notification under 50.72(b)(2)(iii) is not

warranted. With regard to 50.73(a)(2)(v), two years encompasses at

least one operating cycle. Considerable resources are expended when it

is necessary to search historical records older than this to make past

operability determinations, and this is not warranted by the lesser

significance of historical events older than two years.

Response: The comments are partially accepted, for the reasons

stated above. That is, under the proposed rules, an event or condition

that could have prevented the fulfillment of the safety function of

structures or systems * * * would be reported by telephone under 10 CFR

50.72(b)(2)(iii) only if it exists at the time of discovery. An event

or condition that could have prevented the fulfillment of the safety

function of structures or systems * * * would be reported by written

LER under 10 CFR 50.73(a)(2)(v) only if it existed within the previous

three years.

In addition, although not recommended in the comments, under the

proposed rule an operation or condition prohibited by the plant's

Technical Specifications would be reported under 50.73(a)(2)(i)(B) only

if it existed within the previous three years. For this criterion as

well, considerable resources are expended when it is necessary to

search historical records older than three years to make past

operability determinations, and this is not warranted by the lesser

significance of historical events older than three years.

Three years is proposed, rather than two years as suggested in the

comments, because the NRC staff trends plant performance indicators

over a period of three years to ensure inclusion of periods of both

shut down and operation.

Comment 9: Several comments opposed using the term risk-significant

(or significant) in the absence of a clear definition.

Response: The term ``significant'' would be used in two criteria in

the proposed rules. In the first criterion, sections 50.72 and 50.73

would require reporting an unanalyzed condition that significantly

affects plant safety. In this context the term ``significant'' would be

defined by examples, five of which are discussed below under the

heading ``Condition that is outside the design basis of the plant.'' In

the second criterion, section 50.73 would require reporting when a

component's ability to perform its safety function is significantly

degraded and the condition could reasonably be expected to affect other

similar components in the plant. Again, the term ``significant'' would

be defined by examples, six of which are discussed below under the

heading ``Significantly degraded components.''

Comment 10: Several comments recommended changing 10 CFR 50.72 and

50.73 to exclude reporting of an unanalyzed condition that

significantly compromised plant safety on the basis that it is

redundant to other reporting criteria.

Response: The comment is not accepted. Several types of worthwhile

reports have been identified that could not readily be captured by

other criteria as discussed further below under the heading ``Condition

that is outside the design basis of the plant.''

Comment 11: Several comments recommended amending 10 CFR 50.72 and

50.73 to exclude reporting of a seriously degraded principal safety

barrier on the basis that it is redundant to other reporting criteria.

Response: The comments are not accepted. This criterion captures

some worthwhile reports that would not be captured by other criteria,

such as significant welding or material defects in the primary coolant

system. However, some clarifications are proposed in Section 3.2.4 of

the draft reporting guidelines, to better indicate which events are

serious enough to qualify for reporting under this criterion.

Comment 12: One comment recommended that, with regard to a

condition or operation prohibited by the plant's Technical

Specifications, reporting should be eliminated for violation of all

administrative Technical Specifications.

Response: The comment is partially accepted. The proposed rule

would eliminate reporting for Technical Specifications that are

administrative in nature. The reporting guidelines would not change. As

stated in the current reporting guidelines in NUREG-1022, Revision 1,

failure to meet administrative Technical Specifications requirements is

reportable only if it results in violations of equipment operability

requirements, or had a similar detrimental effect on a licensee's

ability to safely operate the plant. For example, operation with less

than the required number of people on shift would constitute operation

prohibited by the Technical Specifications. However, a change in the

plant's organizational structure that has not yet been approved as a

Technical Specification change would not. An administrative procedure

violation or failure to implement a procedure, such as failure to lock

a high radiation area door, that does not have a direct impact on the

safe operation of the plant, is generally not reportable under this

criterion.

[[Page 36294]]

Comment 13: One comment recommended changing 10 CFR 50.73 to

require that LERs identify: (1) How many opportunities to detect the

problem were missed and (2) corrective actions to prevent future

misses.

Response: No changes are proposed. If missed opportunities are

identified and are significant to the event, they should be captured by

the current requirements to provide a comprehensive description of the

event and to describe corrective actions if they are significant to the

event.

Comment 14: With regard to design issues, one comment recommended

including language in the rules or their statements of considerations

encouraging a voluntary report under 10 CFR 50.9 for a newly discovered

design issue which is not otherwise reportable at the plant where first

discovered (because the affected systems can still perform their

specified safety functions) but which might have a significant impact

on generic design issues at other plants.

Response: A statement encouraging submittal of voluntary LERs is

included in the reporting guidelines. In addition, the guidelines would

indicate that any significant degradation that could reasonably be

expected to affect multiple similar components in the plant should be

reported.

Comment 15: Several comments opposed placing a condition, related

to systematic non-compliance, on the elimination of reporting of late

surveillance tests (as proposed in the ANPR) under 10 CFR 50.73. The

condition would be burdensome because licensees would need to track

instances of missed surveillance tests in given time periods.

Response: The proposed rule does not contain this condition.

Reporting for the purpose of identifying systematic non-compliance is

not needed because NRC resident inspectors routinely review plant

problem lists, and thus would be aware of any systematic non-compliance

in this area if it occurs.

Comment 16: One comment recommended changing the rules to allow

licensees to rely on notifications made to resident inspectors, which

could eliminate the need to make a telephone notification via the

emergency notification system (ENS) and/or submit a written LER, at

least for some events or conditions. They indicated, for example, this

should be adequate where the event is a decision to issue a news

release.

Response: No changes are proposed. Telephone notifications to the

NRC Operations Center, when required, are needed to ensure that the

event can be promptly reviewed. This includes notification of the NRC

Headquarters Emergency Officers and the Regional Duty Officer and

consideration of whether to activate NRC incident response procedures.

Written LERs, when required, are needed to ensure that events can be

systematically reviewed for safety significance.

Comment 17: Some comments opposed amending 10 CFR 50.73 to require

additional information regarding equipment availability for shutdown

events (as proposed in the ANPR) to support staff probabilistic risk

assessments (PRAs). They indicated that it is rare that sufficient

information is not available in an LER.

Response: The proposed rule would require such information.

Frequently, when shutdown events are subjected to a probabilistic risk

analysis, it is necessary to call the plant to determine the status of

systems and equipment. The proposed rule would eliminate much of that

need.

Comment 18: Several comments recommended deleting 10 CFR

50.72(b)(2)(i), ``Any event found while the reactor is shut down, 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.'' The comments indicated

that because the plant would be shutdown, there is no need for

immediate NRC action.

Response: The requirement for telephone reporting would not be

entirely eliminated because, if a principal safety barrier is

significantly degraded or a condition that significantly affects plant

safety exists; the event may be significant enough that the NRC would

need to initiate actions [such as contacting the plant to better

understand the event and/or initiating a special inspection or

investigation] within about a day even if the plant is shutdown.

However, in the proposed rule this specific criterion would be

combined with 10 CFR 50.72(b)(1)(ii), ``Any event or condition during

plant operation that results in the condition of the nuclear power

plant, including its principal safety barriers, being seriously

degraded or * * * '' Also, the term ``unanalyzed condition that

significantly compromises plant safety'' would be deleted. In

combination with other changes, this would result in the following

criterion for telephone notification ``Any event or condition that

results in the condition of the nuclear power plant, including its

principal safety barriers, being seriously degraded.''

Comment 19: Some comments recommended that the NRC use enforcement

discretion during the rulemaking process to provide early relief with

regard to reporting a condition outside the design basis of the plant

and/or a late surveillance test (condition or operation prohibited by

Technical Specifications).

Response: The current rules will continue to apply until final

revised rules are issued and become effective. However in

dispositioning any violation, the risk-and safety-significance of the

violation will be an important consideration. Establishing an interim

enforcement discretion policy would involve the same critical elements

as developing the revised rule and guidance including a provision for

public comment. This would complicate the rulemaking process, and

essentially constitute a prediction of its final outcome, which may or

may not turn out to be correct.

Comment 20: Several comment letters opposed the idea of tying

enforcement criteria (i.e., violation severity levels) to reporting

criteria. They indicated this could have an unintended adverse effect

on reporting and the resources consumed because in matching an event

with a reporting criterion, a licensee would essentially be forced to

make a preliminary determination of severity level.

Response: The comments are not accepted. The proposed changes to

the enforcement criteria, are discussed below under the heading

``Enforcement.''

Comment 21: As requested by the ANPR, a number of comments

identified reactor reporting requirements other than sections 50.72 and

50.73 where changes are warranted.

Response: Comments regarding changes to reactor reporting

requirements other than sections 50.72 and 50.73 will be addressed in a

separate action. A Commission paper on that subject was submitted on

January 20, 1999, SECY-99-022, ``Rulemaking to Modify Reporting

Requirements for Power Reactors'' and the Commission issued a Staff

Requirements Memorandum on March 19, 1999 directing the staff to

proceed with planning and scheduling.

Comment 22: One comment recommended changing the required initial

reporting time for some events to `` * * * within 8 hours or by the

beginning of the next business day,'' instead of simply specifying `` *

* * within 8 hours.'' The comment indicated it does not appear that

the

[[Page 36295]]

NRC takes action on these events during non-business hours.

Response: The comment is not accepted. The NRC needs these reports

in time to call the plant to find out more about the event and/or

initiate a special inspection or an investigation, if warranted, within

a day. Sometimes these actions are taken during non-business hours.

Comment 23: One comment recommended that an event or condition that

could have prevented fulfillment of the safety function of structures

or systems. * * * should be reportable only when the time limits of the

TS are exceeded. It indicated that if the time limits are not exceeded

the event is not significant enough to warrant reporting.

Response: The comment is not accepted. Generally, standard TS

require commencement of shutdown within one hour if an important

system, such as emergency ac power, is inoperable. However, the stated

reason for allowing one hour before commencing the shutdown is to

provide time to prepare for an orderly shutdown. Also, the condition

might have lasted much longer than one hour before it was discovered.

Finally, an event that results in a safety system failure (or inability

to perform its function) is generally significant enough to warrant NRC

review.

Comment 24: One comment from the State of Ohio recommended that,

although rule changes are not necessary, emphasis should be placed on

positive notification of State and local agencies of emergency

conditions before calling the NRC.

Response: The comment is accepted. It arose from a weakness in the

NRC's response to an event at the Davis-Besse plant. Because there were

considerable difficulties in establishing telephone communications with

the plant at the time of the event, NRC Operations Center personnel

requested that the licensee remain on the line and said that the NRC

would notify the State. However, the NRC did not do so in a timely

manner. Training and procedure changes have been implemented to ensure

this type of problem will not reoccur.

Comment 25: One comment letter, from the State of Illinois, stated

the following: ``In section 50.72 of the advance notice of proposed

rulemaking, seven non-emergency events listed as (f), are proposed to

be reported in eight hours instead of one hour. Of those seven events,

six (specifically, (ii), (iii), (iv), (v), (vi), and (vii)) would

probably be classified as emergency events under existing emergency

plans at an Illinois site * * *. This will cause reporting confusion

during an event at a time when clarity is necessary. These six events

should all be reported as emergency events, not non-emergency events.

EAL thresholds in licensee emergency plans should be required to

reflect them clearly. All of these events would affect the State of

Illinois' response and our emergency plans. NRC must reconsider the

categories of non-emergency events in the context of the current

guidance to licensees for classifying EALs to ensure there is a clear

distinction between emergency and non-emergency reportable events.''

Response: Section 50.72 has been reviewed, and appears to be clear

in this regard. It indicates the following:

(1) Any declaration of an Emergency Class is reportable pursuant to

10 CFR 50.72(a)(1)(i) and (a)(3),

(2) The conditions listed in paragraph (b)(1), ``One-hour

reports,'' are reportable pursuant to paragraph (b)(1) if not reported

as a declaration of an Emergency Class under paragraph (a), and

(3) The conditions listed in paragraph (b)(2), ``Eight-hour

reports, are reportable pursuant to paragraph (b)(2), if not reported

under paragraphs (a) or (b)(1).

Comment 26: One comment letter, from the State of Illinois, opposed

relaxing the required initial reporting time from 4 hours to 8 hours

for the following types of events:

(i) Airborne radioactive release that results in concentrations

over 20 times allowable levels in an unrestricted area;

(ii) Liquid effluent in excess of 20 times allowable concentrations

released to an unrestricted area;

(iii) Radioactively contaminated person transported to an offsite

medical facility for treatment;

(iv) News release or other government agency notification related

to the health and safety of the public or onsite personnel, or

protection of the environment.

The comment further indicated: ``It is of paramount importance that

those charged with regulating and monitoring the public impact of

radiological releases are being kept informed of unplanned releases in

a timely manner. Illinois law requires that we perform independent

assessments, decide what actions may be necessary to protect the

public, and assist in informing the public regarding any radiological

risk. Should follow-up action to a release be necessary, then the less

time that has elapsed, the better the state is able to respond in a

timely and appropriate manner. We oppose any reduction in notification

requirements for unplanned radiation releases from a site regardless of

the source or quantity.

Timeliness is also important for items of obvious public interest.

News of seemingly small events spreads quickly, particularly in local

communities around the power plants. Delayed reporting of such events

means that we will be unprepared to respond to queries from local

officials, or the media, with a resultant loss of public confidence.

Therefore, we also oppose any reduction in notification requirements

for newsworthy events.''

Response: In the interest of simplicity, the proposed amendments

would maintain just three basic levels of required reporting times in

10 CFR 50.72 and 50.73 (1 hour, 8 hours, and 60 days). However, the

concern is recognized and public comment is specifically invited on the

question of whether additional levels should be introduced to better

correspond to particular types of events, as discussed below under the

heading ``Required Initial Reporting Times.'' Also, if in a final rule

the NRC should relax the time limit to 8 hours, a State would not be

precluded from obtaining reports earlier than 8 hours.

Comment 27: Two comment letters addressed coordination with States.

The comment letter from Florida Power & Light Company stated ``The

NRC's Public workshop on August 21, 1998, touched on a number of

examples where opportunities exist to reduce reporting burdens. An

industry representative commented that licensees sometimes have to

report the same event to state agencies and the NRC provided one such

example. FPL concurs with the recommendation that the time requirement

for reporting an event to the NRC and to the state should be consistent

wherever practical and possibly in some cases eliminated.''

The comment letter from Northeast Nuclear Energy Company stated

``Northeast Nuclear Energy Company agrees with extending the non-

emergency prompt notifications to eight hours. This would help to

eliminate unnecessary reports and retractions. However, it is necessary

to have the individual states closely involved with the rule change

since they may have requirements that are more restrictive or conflict

with the proposed rulemaking. For example, in Connecticut all 10 CFR

50.72 reports require notification of the state within one hour.''

Response: The ANPR specifically requested State input. In addition,

a letter requesting input was sent to each State. Written comments were

received from the State of Ohio and the State of Illinois. In addition,

representatives

[[Page 36296]]

from several States attended one of the public meetings on the ANPR.

The NRC will continue to solicit State input as the rulemaking process

proceeds.

Comment 28: One comment recommended eliminating two of the

requirements for immediate followup notification during the course of

an event, section 50.72(c)(2)(i), the results of ensuing evaluations or

assessments of plant conditions, and section 50.72(c)(2)(ii), the

effectiveness of response or protective measures taken. The comment

indicated that the requirements continue to apply after the event and

that they require reporting even if, for example, the result of a

further analysis does not change the initial report.

Response: The comment is not accepted. The requirements for

followup reporting apply only during the course of the event. Followup

reports are needed while the event is ongoing. For example, if an

analysis is completed during an ongoing event, and it confirms an

earlier estimate of how long it will take to uncover the reactor core

if electric power is not restored, that information may very well be

useful for the purpose of evaluating the need for protective measures

(evacuation).

Comment 29: One comment recommended clarifying the reporting

requirements for problems identified by NRC inspectors.

Response: No changes are proposed. The current reporting guidelines

include a paragraph making it clear that an event must be reported via

telephone notification and/or written LER, as required, regardless of

whether it had been discussed with NRC staff personnel or was

identified by NRC personnel.

Comment 30: Several comments recommended changing the requirements

in 50.46(a)(iii)(2) for reporting errors in or corrections to ECCS

analyses.

Response: These comments will be addressed in a separate action

(along with other comments on reporting requirements other than

sections 50.72 and 50.73).

Comment 31: Some comments raised issues regarding plant-specific

reporting requirements contained in Technical Specifications (or other

parts of the operating license). One suggestion was that 10 CFR 50.72

and 50.73 should be changed to address these issues. Another suggestion

was that a Generic Letter be issued indicating that the NRC would be

receptive to requests for license amendments to eliminate specific

reporting requirements.

Response: No changes are proposed for sections 50.72 and 50.73,

which identify generic reporting requirements. It is not feasible or

appropriate to address the specific reporting requirements contained in

individual operating licenses in this format.

The idea of issuing a generic communication to specific requests

for license amendments will be addressed (along with other comments on

reporting requirements beyond the scope of sections 50.72 and 50.73) in

a separate action.

Comment 32: One comment recommended that in section 50.72(b)(1)(v),

the word ``offsite'' be added before ``communications capability'' to

make it clear that what must be reported is a loss of communications

with outside agencies, not internal plant communications systems.

Response: The comment is accepted. In the proposed rule the word

``offsite'' would be added.

Comment 33: Several comments suggested that the NRC should define

its needs relative to the information provided in LERs.

Response: The essential purpose of the LER rule is to identify 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. The rule is designed to provide the

information necessary for engineering studies of operational anomalies

and trends, and patterns analysis of operational occurrences. To this

end, the information required in LERs is generally needed to understand

the event, its significance, and its causes in order to determine

whether generic or plant specific action is needed to preclude

recurrence. Some further specific functions are discussed below.

It is necessary to identify and analyze events and conditions that

are precursors to potential severe core damage, to discover emerging

trends or patterns of potential safety significance, to identify events

that are important to safety and their associated safety concerns and

root causes, to determine the adequacy of corrective actions taken to

address the safety concerns, and to assess the generic applicability of

events.

The NRC staff reviews each LER to identify those individual events

or generic situations that warrant additional analysis and evaluation.

The staff identifies repetitive events and failures and situations

where the frequency or the combined significance of reported events may

be cause for concern. The NRC staff reviews past operating history for

similar events and initiates a generic study, as appropriate, to focus

upon the nature, cause, consequences and possible corrective actions

for the particular situation or concern.

The NRC staff uses the information reported in LERs in confirming

licensing bases, studying potentially generic safety problems,

assessing trends and patterns of operational experience, monitoring

performance, identifying precursors of more significant events, and

providing operational experience to the industry.

The NRC determines whether events meet the criteria for reporting

as an Abnormal Occurrence Report to Congress or for reporting to the

European Nuclear Energy Agency (NEA).

The information from LERs is widely used within the nuclear

industry, both nationally and internationally. The industry's Institute

of Nuclear Power Operation (INPO) uses LERs as a basis for providing

operational safety experience feedback data to individual utilities

through such documents as significant operating experience reports,

significant event reports, significant events notifications, and

operations and maintenance reminders. U.S. vendors and nuclear steam

system suppliers, as well as other countries and international

organizations, use LER data as a source of operational experience data.

Comment 34: Some comments indicated that the licensing basis should

be defined.

Response: No changes are proposed. The term ``licensing basis'' is

not explicitly used in the event reporting rules or the draft reporting

guidelines. It can come into play, via Generic Letter (GL) 91-18,

``Information to Licensees Regarding two NRC Inspection Manual Sections

on Resolution of Degraded and Nonconforming Conditions and on

Operability,'' in determining what the ``specified safety function'' of

a system is. This relates to whether an event is reportable as an event

or condition that could have prevented the fulfillment of the safety

function of structures or systems * * * and/or an operation or

condition prohibited by the plant's technical specification (TS).

However, any unsettled details regarding exactly which commitments are

included in the licensing basis (for example because of differences

between the definitions in GL 91-18 and 10 CFR 54.3) are not of a

nature that would change the determination of whether or not a system

is capable of performing its specified safety functions (i.e.,

operable).

[[Page 36297]]

Comment 35: Several comments recommended conducting tabletop

exercises (public meetings) early in the drafting process, involving

licensees, inspectors, and headquarters personnel to discuss the draft

amendments and associated and guidance.

Response: The Commission agrees. The recommended public meeting was

held on November 13, 1998.

Comment 36: Several comments recommended conducting a workshop

(public meeting) early during the public comment period to discuss the

proposed rule and draft guidance.

Response: The Commission agrees. The recommended workshop has been

added to the schedule.

Comment 37: Several comments recommended that the reporting

guidelines be revised concurrently with the rules.

Response: The Commission agrees. Draft guidelines are being made

available for comment concurrent with the proposed rules.

Comment 38: Several comment letters recommended reviewing

enforcement criteria at the same time the rule is being developed to

ensure consistent application of enforcement to reporting.

Response: The comment is accepted. The Enforcement Policy is being

reviewed concurrently with development of the rule.

IV. Discussion

1. Objectives of Proposed Amendments

The purpose of sections 50.72 and 50.73 would remain the same

because the basic needs remain the same. The objectives of the proposed

amendments would be as follows:

(1) To better align the reporting requirements with the NRC's

current reporting needs. An example is extending the required initial

reporting times for some events, consistent with the need for timely

NRC action. Another example is changing the criteria for reporting

system actuations, to obtain reporting that is more consistent with the

risk-significance of the systems involved.

(2) To reduce the reporting burden, consistent with the NRC's

reporting needs. An example is eliminating the reporting of design and

analysis defects and deviations of 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) To maintain consistency 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 Proposed Amendments

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 non-emergency 8-hour report)

would be 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 sections 50.73(a)(1) and (d)]. In the proposed amendments,

declaration of an emergency class would continue to be reported

immediately after notification of appropriate State or local agencies

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) would 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.

Non-emergency events that are reportable by telephone under 10 CFR

50.72 would be reportable as soon as practical and in all cases within

8 hours (instead of within 1 hour or 4 hours as is currently required).

This would reduce the burden of rapid reporting, while still capturing

those events where there may be a need for the NRC to contact the plant

to find out more about the event and/or initiate a special inspection

or investigation within about a day.

Written LERs would be due within 60 days after discovery of a

reportable event or condition (instead of within 30 days as is

currently 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 would result

in fewer LERs that require amendment (by submittal of an additional

report).

The Performance Indicator (PI) program and the future risk-based

performance indicator program provide valued input to regulatory

decisions (e.g. Senior Management Meetings). Adding 30 days to the

delivery of data supplying these programs would result in the reduction

in the currency and value of these indicators to senior managers. With

respect to the Accident Sequence Precursor program, the additional 30

days will add a commensurate amount of time to each individual event

assessment since Licensee Event Reports (LERs) are the main source of

data for these analyses. The delivery date for the annual Accident

Sequence Precursor report would also slip accordingly. The NRC staff

would have to make more extensive use of Immediate Notifications (10

CFR 50.72) and event followup to compensate in part for the Licensee

Event Report (LER) reporting extension.

In the interest of simplicity, the proposed amendments would

maintain just three basic levels of required reporting times in 10 CFR

50.72 and 50.73 (1 hour, 8 hours, and 60 days). However public comment

is specifically invited on the question of whether additional levels

should be introduced to better correspond or particular types of

events. For example, 10 CFR 50.72 currently requires 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.

Operation or condition prohibited by TS [section

50.73(a)(2)(i)(B)]. The term ``during the previous three years'' would

be added to eliminate written LERs for conditions that have not existed

during the previous three years. Such a historical event would now have

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 should 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 very considerable

[[Page 36298]]

effort required to review all records older than three years, in order

to rule out the possibility, would not be warranted.

In addition, this criterion would be modified to eliminate

reporting if the technical specification is administrative in nature.

Violation of administrative technical specifications have generally not

been considered to warrant submittal of an LER, and since 1983 when the

rule was issued the staff's reporting guidance has excluded almost all

cases of such reporting. This change would make the plain wording of

the rule consistent with that guidance.

Finally, this criterion would be 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.

Condition of the nuclear power plant, including its principal

safety barriers, being seriously degraded [current sections

50.72(b)(1)(ii) and (b)(2)(i), replaced by new section 50.72(b)(2)(ii),

and section 50.73(a)(2)(ii)]. Currently, 10 CFR 50.72(b)(1)(ii) and

(b)(2)(i) provide the following distinction: a qualifying event or

condition during operation is initially reportable in one hour; a

condition discovered while shutdown that would have qualified if it had

it been discovered during operation is initially reportable in four

hours. The new 10 CFR 50.72(b)(2)(ii) would eliminate the distinction

because there would no longer be separate 1-hour and 4-hour categories

of non-emergency reports for this criterion. There would only be 8-hour

non-emergency reports for this criterion.

Unanalyzed condition that significantly compromises plant safety

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

50.73(a)(2)(ii)(A); replaced by new section 50.72(b)(2)(ii)(B), and

section 50.73(a)(2)(ii)(B)]. Currently, 10 CFR 50.72(b)(1)(ii)(A) and

(b)(2)(i) provide the following distinction: a qualifying event or

condition during operation is initially reportable in one hour; a

condition discovered while shutdown that would have qualified if it had

it been discovered during operation is initially reportable in four

hours. The new 10 CFR 50.72(b)(2)(ii)(B) would eliminate the

distinction because there would no longer be separate 1-hour and 4-hour

categories of non-emergency reports for this reporting criterion. There

would only be 8-hour non-emergency reports for this criterion.

In addition, the new 10 CFR 50.72(b)(2)(ii)(B) and

50.73(a)(2)(ii)(B) would refer to a condition that significantly

affects 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 [current

Section 50.72(b)(2)(ii)(B) and section 50.73(a)(2)(ii)(B)]. This

criterion would be deleted. However, a condition outside the design

basis of the plant would still be reported if it is significant enough

to qualify under one or more of the following criteria.

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 would not be capable of performing its specified safety

functions, the condition would be reportable under sections

50.72(b)(2)(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: (A) Shut down * * *].

For example, during testing of 480 volt safety-related breakers,

one breaker would not trip electrically. The cause was a loose

connection, due to a lug that was too large for a connecting wire.

Other safety related breakers did not malfunction, but they had the

same mismatch. The event would be reportable because the incompatible

lugs and wires could have caused one or more safety systems to fail to

perform their specified safety function(s).

Another example is as follows. 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 would be reportable because there was reasonable doubt about the

ability of the EDGs to operate for an extended period of time, as

required.

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, and thus the train is

inoperable longer than allowed by the TS, the condition would be

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

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 a certain postulated

steam line breaks, the event would be reportable.

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 affected, the condition would be reportable

under sections 50.72(b)(2)(ii)(B) and 50.73(a)(2)(ii)(B) [i.e., an

unanalyzed condition that significantly affects 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.''

\1\

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\1\ 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.'' \2\

---------------------------------------------------------------------------

\2\ 48 FR 39042, August 29, 1983.

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``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

[[Page 36299]]

unanalyzed condition and would be reportable.'' \3\

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\3\ 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.'' \4\

---------------------------------------------------------------------------

\4\ 48 FR 39042, August 29, 1983 and 48 FR 33856, July 26, 1983.

---------------------------------------------------------------------------

Furthermore, beyond the examples given in 1983, examples of

reportable events would 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 would be 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 would not be

reportable.

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 would be reportable

under sections 50.72(b)(2)(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) Steam generator tube degradation in the following

circumstances:

(1) The severity of degradation corresponds to failure to maintain

structural safety factors. The structural safety factors implicit in

the licensing basis are those described in Regulatory Guide 1.121.

These safety factors include a margin of 3.0 against gross failure or

burst under normal plant operating conditions, including startup,

operation in the power range, hot standby, and cooldown, and all

anticipated transients that are included in the plant design

specification.

(2) The calculated potential primary-to-secondary leak rate is not

consistent with the plant licensing basis. The licensing basis accident

analyses typically assume [for accidents other than a steam generator

tube rupture (SGTR)] a 1 gpm primary-to-secondary leak rate concurrent

with the accident to demonstrate that the radiological consequences

satisfy 10 CFR Part 100 and GDC-19. In these instances, degradation

which may lead to leakage above 1 gpm under accident conditions, other

than a SGTR, would exceed the threshold. For some units, the staff has

approved accident leakages above 1 gpm subject to updating the

licensing basis accident analyses to reflect this amount of leakage and

subject to risk implications being acceptable.\5\

---------------------------------------------------------------------------

\5\ In addition, if the extent of degradation is great (i.e., if

many tubes are degraded or defective), a telephone notification and

a written LER should be provided. The plant's TS typically provide

specific requirements indicating when reporting is required (based

on the number of tubes degraded or defective in terms of ``percent

inspected'') and those requirements should be used to determine

reportability.

---------------------------------------------------------------------------

(iv) Low temperature over pressure transients where the pressure-

temperature 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.\6\

---------------------------------------------------------------------------

\6\ The 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. Minimum-pathway 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.

---------------------------------------------------------------------------

Finally, a condition outside the design basis of the plant (or any

other event or condition) would be reportable if a component is in a

degraded or non-conforming condition such that the ability of a

component to perform its specified safety function is significantly

degraded and the condition could reasonably be expected to apply to

other similar components in the plant. This new criterion is contained

in section 50.73(a)(2)(ii)(C) as discussed below.

As a result, these proposed amendments would focus the reporting of

conditions outside the design basis of the plant to the safety

significant issues while reducing the number of reports under the

current rules in order to minimize the reporting of less significant

issues. In particular, the proposed amendments will help ensure that

significant safety problems that could reasonably be expected to be

applicable to similar components at the specific plant or at other

plants will be identified and addressed although the specific licensee

might determine that the system or structure remained operable, or that

technical specification requirements were met. The proposed rules will

provide that, consistent with the NRC's effort to obtain information

for engineering studies of operational anomalies and trends and

patterns analysis of operational occurrences, the NRC would be able to

monitor the capability of safety-related components to perform their

design-basis functions.

Significantly degraded component(s) [section 50.73(a)(2)(ii)(C)].

This new reporting criterion would require reporting if a component is

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 apply to

other similar components in the plant. It would be added 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. On the other hand,

if the degradations are not significant or the condition does not have

generic implications, reporting would not be required under this

criterion.

For example, at one plant several normally open valves in the low

pressure safety injection system were routinely closed to support

quarterly surveillance testing of the system. In reviewing the design

basis and associated calculations, it was determined that the

capability of the valves to open in the event of a large break loss-of-

coolant accident (LOCA) combined with degraded grid voltage during a

surveillance test was degraded. The licensee concluded that the valves

would still be able to reopen under the postulated conditions and

considered them operable. However, that conclusion could not be

supported using the conservative standards established by Generic

Letter 89-10. Pending determination of final corrective action,

administrative procedures were implemented to preclude closing the

valves. The event would be reportable because the

[[Page 36300]]

capability of a component to perform its specified safety functions was

significantly degraded and the same condition could reasonably be

expected to apply to other similar components.

In another example, during a routine periodic inspection, jumper

wires in the valve operators for three valves were found contaminated

with grease which was leaking from the limit switch gear box. The cause

was overfilling of the grease box, as a result of following a generic

maintenance procedure. The leakage resulted in contamination and

degradation of the electrical components which were not qualified for

exposure to grease. This could result in valve malfunction(s). The

conditions were corrected and the maintenance procedures were changed.

The event would be reportable because the capability of several similar

components to perform their specified safety functions could be

significantly degraded.

In a further example, while processing calculations it was

determined that four motor operated valves within the reactor building

were located below the accident flood level and were not qualified for

that condition. Pending replacement with qualified equipment, the

licensee determined that three of the valves had sufficiently short

opening time that their safety function would be completed before they

were submerged. The fourth valve was normally open and could remain

open. After flooding, valve position indication could be lost, but

valve position could be established indirectly using process parameter

indications. The event would be reportable because the capability of

several similar components to perform their specified safety functions

could be significantly degraded.

An example of an event that would not be reportable is as follows.

The motor on a motor-operated valve (MOV) burned out after repeated

cycling for testing. This event would not be reportable because it is a

single component failure, and while there might be similar MOVs in the

plant, there is not a reasonable basis to think that other MOVs would

be affected by this same condition. On the other hand, if several MOVs

had been repeatedly cycled and then after some extended period of time

one of the MOVs was found inoperable or significantly degraded because

of that cycling, then the condition would be reportable.

Minor switch adjustments on MOVs would not be reported where they

do not significantly affect the ability of the MOV to carry out its

design-basis function and the cause of the adjustments is not a generic

concern.

At one plant the switch on the radio transmitter for the auxiliary

building crane was used to handle a spent fuel cask while two

protective features had been defeated by wiring errors. A new radio

control transmitter had been procured and placed in service. Because

the new controller was wired differently than the old one, the drum

overspeed protection and spent fuel pool roof slot limit switch were

inadvertently defeated. While the crane was found to be outside its

design basis, this condition would not be reportable because the switch

wiring deficiency could not reasonably be expected to affect any other

components at the plant.

Condition not covered by the plant's operating and emergency

procedures [section 50.72(b)(2)(ii)(C), and section

50.73(a)(2)(ii)(C)]. This criterion would be deleted because it does

not result in worthwhile reports aside from those that would be

captured by other reporting criteria such as:

(1) An unanalyzed condition that significantly affects 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:

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;

(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 any of the

systems listed in the rules, as amended;

(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.

Manual or automatic actuation of any engineered safety feature ESF

[current sections 50.72(b)(1)(iv) and (b)(2)(ii), replaced by new

sections 50.72(b)(2)(iv), and section 50.73(a)(2)(iv)]. Currently,

sections 50.72(b)(1)(iv) and (b)(2)(ii) provide the following

distinction: an event that results or should have resulted in ECCS

discharge into the reactor coolant system is initially reportable

within 1 hour; other ESF actuations are initially reportable within 4

hours. The new 10 CFR 50.72(b)(2)(iv) would eliminate this distinction

because there would no longer be separate 1-hour and 4-hour categories

of non-emergency reports for this criterion. There would only be 8-hour

non-emergency reports for this criterion.

The new section 50.72(b)(2)(iv) would eliminate telephone reporting

for invalid automatic actuation or unintentional manual actuation.

These events are not significant and thus telephone reporting is not

needed. However, the proposed amendments would not eliminate the

requirement for a written report 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. (See 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).

The term ``any engineered safety feature (ESF), including the

reactor protection system (RPS),'' which currently defines the systems

for which actuation must be reported in section 50.72(b)(2)(iv) and

section 50.73(a)(2)(iv), would be replaced by a specific list of

systems. The current definition has led to confusion and variability in

reporting because there are varying definitions of what constitutes an

ESF. For example, at some plants systems that are known to have high

risk significance, such as emergency ac power, auxiliary feedwater, and

reactor core isolation cooling are not considered ESFs. Furthermore, in

many cases systems with much lower levels of risk significance, such as

control room ventilation systems, are considered to be ESFs.

In the proposed amendments actuation would be reportable for the

specific systems named in sections 50.72(b)(2)(iv) and 50.73(a)(2)(iv).

This would result in consistent reporting of events that result in

actuation of these highly risk-significant systems. Reasonable

consistency in reporting actuation of highly risk-significant systems

is needed to support estimating equipment reliability parameters, which

is important to several aspects of the move towards more risk-informed

regulation, including more risk-informed monitoring of plant

performance.

The specific list of systems in the proposed rule would also

eliminate reporting for events of lesser significance, such as

actuation of control room ventilation systems.

The specific list of systems in the proposed rule is similar to the

list of systems currently provided in the reporting guidelines in

NUREG-1022,

[[Page 36301]]

Revision 1, with some minor revisions. It is based on systems for which

actuation is frequently reported, and systems with relatively high

risk-significance based on a sampling of plant-specific PRAs (see Draft

Regulatory Guide DG-1046, ``Guidelines for Reporting Reliability and

Availability Information for Risk-Significant Systems and Equipment in

Nuclear Power Plants,'' particularly Tables C-1 through C-5).

This proposal to list the systems in the rule is controversial and

public comment is specifically invited in this area. In particular,

three principal alternatives to the proposed rule have been identified

for comment:

(1) Maintain the status quo. Under this alternative, the rule would

continue to require reporting for actuation of ``any ESF.'' The

guidance would continue to indicate that reporting should include as a

minimum the system on the list.

(2) Require use of a plant-specific, risk-informed list. Under this

alternative, the list of systems would be risk-informed, and plant-

specific. Licensees would develop the list based on existing PRA

analyses, judgment, and specific plant design. No list would be

provided in the rule.

(3) Return to the pre-1998 situation (i.e., before publication of

the reporting guidance in NUREG-1022, Revision 1). Under this

alternative, the rule would continue to require reporting for actuation

of ``any ESF.'' The guidance would indicate that reporting should

include those systems identified as ESF's for each particular plant

(e.g., in the FSAR).

With regard to this third alternative, it may be noted that this

approach has the advantage of clarity and simplicity. There would be no

need to develop a new list, and this is the practice that was followed

from 1984-1997 without creating major problems. However, the lists of

ESFs are not based on risk-significance. For example, emergency diesel

generators (EDGs) are known to be highly risk-significant; however, at

six plants, the EDGs are not considered to be ESFs. Similarly,

auxiliary feedwater (AFW), systems at pressurized water reactors (PWRs)

are known to be highly risk-significant; however, at a number of plants

these systems are not considered to be ESFs. Also, reactor core

isolation cooling (RCIC) systems at boiling water reactors (BWRs) are

known to be highly risk significant; however, at a number of plants

these systems are not considered to be ESFs. In contrast, at many

plants, systems with much lower levels of risk significance, such as

control room ventilation systems, are considered to be ESFs.

Event or condition that could have prevented fulfillment of the

safety function of structures or systems that * * * [current sections

50.72(b)(1)(ii) and (b)(2)(i), replaced by new sections 50.72(b)(2)(v)

and (vi), and sections 50.73(a)(2)(v) and (vi)] The phrase ``event or

condition that alone could have prevented the fulfillment of the safety

function of structures or systems.* * * '' would be 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'' would be added to section

50.72(b)(2)(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 some time ago 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 would be required, but there would be no need

for telephone notification.

The phrase ``occurring within three years of the date of

discovery'' would be added to section 50.73(a)(2)(v) to eliminate

written LERs for conditions that have not existed during the previous

three years. Such a historical event would now have less significance,

and assessing reportability for earlier times can consume considerable

resources. For example, assume that during a design review a

discrepancy is found that affects the ability of a system to perform

its safety function in a given specific configuration. If it is likely

that the safety function could have been prevented, the answer should

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 very considerable effort required to review all

records older than three years, in order to rule out the possibility,

would not be warranted.

A new paragraph, section 50.72(b)(2)(vi) would be added to clarify

section 50.72. The new paragraph would explicitly state 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 is currently stated explicitly in section

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

Major loss of emergency assessment capability, offsite response

capability, or communication capability [current section

50.72(b)(2)(v), new section 50.72(b)(2)(xiii)]. The new section would

be 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.

Airborne radioactive release * * * and liquid effluent release * *

* [section 50.72(b)(2)(viii) and sections 50.73(a)(2)(viii) and

50.73(a)(2)(ix)]. The statement indicating reporting under section

50.72(b)(2)(viii) satisfies the requirements of section 20.2202 would

be removed because it would not be correct. For example, some events

captured by section 20.2202 would not be captured by section

50.72(b)(2)(viii). Also, the statement indicating that reporting under

section 50.73(a)(2)(viii) satisfies the requirements of section

20.2203(a)(3) would be deleted because it would not be correct. Some

events captured by section 20.2203(a)(3) would not be captured by

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

The proposed extension of reporting deadlines to 8 hours in section

50.72 and 60 days in section 50.73 raises questions about whether

similar changes should be made to Parts 20, 30, 40, 70, 72 and 76. The

merits of such changes, which may vary for different types of

licensees, will be addressed in separate actions.

Contents of LERs [sections 50.73(b)(2)(ii)(F) and

50.73(b)(2)(ii)(J)]. Paragraph (F) would be revised to correct the

address of the NRC Library.

Paragraph (J) currently requires that the narrative section include

the following specific information as appropriate for the particular

event:

``(1) Operator actions that affected the course of the event,

including operator

[[Page 36302]]

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,

utility-licensed operator, utility non-licensed operator, other utility

personnel).''

The proposed amendment would change section 50.73(b)(2)(ii)(J) to

simply require that the licensee discuss the causes and circumstances

for each human performance related problem that contributed to the

event. It is not necessary to specify the level of detail provided in

the current rule, which is more appropriate for guidance. 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. This draft

report is being made available for public comment concurrently with the

proposed rule, as discussed below under the heading ``Revisions to

Reporting Guidelines in NUREG-1022.''

Spent fuel storage cask problems [current sections 50.72(b)(2)(vii)

and 72.16(a)(1), (a)(2), (b) and (c)]. Section 50.72(b)(2)(vii) would

be deleted because these reporting criteria are redundant to the

reporting criteria contained in sections 72.216(a)(1), (a)(2), and (b).

Repetition of the same reporting criteria in different sections of the

rules adds unnecessary complexity and is inconsistent with the current

practice in other areas, such as reporting of safeguards events as

required by section 73.71.

Also, a conforming amendment would be made to section 72.216. This

is necessary because section 72.216(a) currently relies on section

50.72(b)(2)(vii), which would be deleted, to establish the time limit

for initial notification. The amended section 72.216 would refer to

sections 72.74 and 72.75 for initial notification and followup

reporting requirements.

Assessment of Safety Consequences [section 50.73(b)(3)]. This

section currently requires that an LER include an assessment of the

safety consequences and implications of the event. This assessment must

include the availability of other systems or components that could have

performed the same function as the components and systems that failed

during the event. It would be modified by adding a requirement to also

include the status of components and systems 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.'' This information is needed to better support the

NRC's assessment of the risk-significance of reported events.

Exemptions [section 50.73(f)]. This provision would be deleted

because the exemption provisions in section 50.12 provide for granting

of exemptions as warranted. Thus, including another, section-specific

exemption provision in section 50.73 adds unnecessary complexity to the

rules.

3. Revisions to Reporting Guidelines in NUREG-1022

A draft report, NUREG-1022, Revision 2, ``Event Reporting

Guidelines, 10 CFR 50.72 and 50.73,'' is being made available for

public comment concurrently with the proposed amendments to 10 CFR

50.72 and 50.73. The draft 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

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 draft

report, guidance that is considered to be new or different is a

meaningful way, relative to that provided in NUREG-1022, Revision 1, is

indicated by redlining the appropriate text.

4. Reactor Oversight

The NRC is developing revisions to 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 proposed rules

and believes that the changes would not have a deleterious effect on

the oversight process. Public comment is invited on whether or not this

is the case. In particular, it is requested that if any examples to the

contrary are known they be identified.

5. Reporting of Historical Problems

As discussed above, provisions would be added to sections

50.73(a)(2)(i)(B) and 50.73(a)(2)(v) to eliminate reporting of a

condition or event that did not occur within three years of the date of

discovery. (See the response to Comment 8, the discussion under the

heading ``Operation or condition prohibited by TS,'' and the discussion

under the heading ``Event or condition that could have prevented

fulfillment of the safety function of structures or systems that * * *

'') Public comment is invited on whether such historical events and

conditions should be reported (rather than being excluded from

reporting, as proposed). Public comment is 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).

6. Reporting of Component Problems

As discussed above, a new reporting criterion would be added to

require reporting if a component is 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 apply to other similar components in

the plant. (See the response to Comment 14 and the discussion under the

heading ``Significantly degraded component(s) [section

50.73(a)(2)(ii)(C)].'') Public comment is invited on whether this

proposed new criterion would accomplish its stated purpose--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.

Public comment is also invited on whether the proposed new criterion

would be subject to varying interpretations by licensees and

inspectors.

7. Enforcement

The NRC intends to modify its existing enforcement policy in

connection with the proposed amendments to sections 50.72 and 50.73.

The philosophy of the proposed changes is to base the significance of

the reporting violation on: (1) The reporting requirement, which will

require reporting within time frames more commensurate with the

significance of the underlying issues than the current rule; and (2)

the impact that a late report may have on the ability of the NRC to

[[Page 36303]]

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, Rev. 1 as follows:

(1) Appendix B, Supplement I.C--Examples of Severity Level III

violations.

(a) Example 14 would be revised to read as follows--A failure to

provide the required one hour telephone notification of an emergency

action taken pursuant to 10 CFR 50.54(x).

(b) An additional example would be added that would read as

follows--A failure to provide a required 1-hour or 8-hour non-emergency

telephone notification pursuant to 10 CFR 50.72.

(c) An additional example would be added that would read as

follows--A late 8-hour notification that substantially impacts agency

response.

(2) Appendix B, Supplement I.D--Examples of Severity Level IV

violations.

(a) Example 4, would 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 would 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 would correlate the Severity Level of the infractions

with the relative importance of the information needed by the NRC.

Section IV.D 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 or 8-hour non-emergency report the impact that

the failure to report had on any agency response would 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 would not normally be reduced. Similarly, a late notification

that delayed the NRC's ability to perform an engineering analysis of a

condition to determine if additional regulatory action was necessary

would generally not be considered for disposition at a reduced Severity

Level. Additionally, late reports filed in cases where the NRC had to

prompt the licensee to report would generally not be subject to

disposition at reduced Severity Level and the Severity Level for

failure to submit a timely Licensee Event Report (LER) would not be

reduced to a minor violation.

In accordance with Appendix C of the Enforcement Policy, `` Interim

Enforcement Policy for Severity Level IV Violations Involving

Activities of Power Reactor Licensees,'' the failure to file a 60-day

LER would normally be dispositioned as a Non-Cited Violation (NCV).

Repetitive failures to make LER reports indicative of a licensee's

inability to recognize reportable conditions, such that it is not

likely that the NRC will be made aware of operational, design and

configuration issues deemed reportable pursuant to 10 CFR 50.73, will

be considered for categorization at Severity Level III. This

disposition may be warranted since such licensee performance impacts

the ability of the NRC to fulfill its regulatory obligations.

8. Electronic Reporting

The NRC is currently planning to implement an electronic document

management and reporting program, known as the Agency-wide Document

Access and Management System (ADAMS), that will in general provide for

electronic submittal of many types of reports, including LERs.

Accordingly, no separate rulemaking effort to provide for electronic

submittal of LERs is contemplated.

9. Schedule

The current schedule is as follows:

08/99--Conduct public workshop to discuss proposed rule and draft

reporting guidelines (separate notice with workshop details will be

published later this month).

August 5, 1999--Public comments due to OMB

September 7, 1999--Receive OMB approval

September 20, 1999--Public comments due to NRC

10/01/99--Provide final rule and guidelines to NRC staff rulemaking

group

11/05/99--Provide final rule and guidelines to the formal concurrence

chain

01/14/00--Provide final rule and guidelines to CRGR and ACRS

02/11/00--Complete briefings of CRGR and ACRS

03/10/00--Provide final rule and guidelines to Commission

04/07/00--Publish final rule and guidelines

10. 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 seeks State comment on issues

related to the proposed amendments to power reactor reporting

requirements.

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 requests comments on

this proposed rule specifically with respect to the clarity and

effectiveness of the language used. Comments should be sent to the

address listed above.

V. 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.

VI. 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 proposed rule. 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.

VII. Regulatory Analysis

The Commission has prepared a draft regulatory analysis on this

proposed rule. The analysis examines the costs and benefits of the

alternatives considered by the Commission. The draft 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.''

The Commission requests public comment on this draft analysis.

Comments on the draft analysis may be

[[Page 36304]]

submitted to the NRC as discussed above under the heading ADDRESSES.

VIII. Paperwork Reduction Act Statement

This proposed rule would amend information collection requirements

that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501

et seq.). This rule has been submitted to the Office of Management and

Budget for review and approval of the information collection

requirements.

The public reporting burden for the currently existing reporting

requirements in 10 CFR 50.72 and 50.73 is estimated to average about

790 hours per response (i.e., per commercial nuclear power reactor per

year) 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, after which there would be a recurring

annual burden reduction of about 200 hours per reactor per year. The

U.S. Nuclear Regulatory Commission is seeking public comment on the

potential impact of the information collection contained in the

proposed rule and on the following issues:

Is the proposed information collection necessary for the proper

performance of the NRC, including whether the information will have

practical utility?

Is the estimate of burden accurate?

Is there a way to enhance the quality, utility, and clarity of the

information to be collected?

How can the burden of the information collection be minimized,

including the use of automated collection techniques?

Send comments on any aspect of this proposed information

collection, including suggestions for reducing this burden, to the

Information and Records Management Branch (T-5 F33), U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001 or by Internet

electronic mail to [email protected]; and to the Desk Officer, Office of

Information and Regulatory Affairs, NEOB-10202, (3150AF98), Office of

Management and Budget, Washington, DC 20503.

Comments to OMB on the information collections or on the above

issues should be submitted by August 5, 1999. Comments received after

this date will be considered if it is practical to do so, but

consideration cannot be ensured for comments received after this date.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to

respond to, an information collection unless it displays a currently

valid OMB control number.

IX. Regulatory Flexibility Certification

In accordance with the Regulatory Flexibility Act (5 U.S.C.

605(b)), the Commission certifies that this rule will not, if

promulgated, 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).

X. Proposed 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, and 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 proposing to

adopt 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:

Sec. 50.72 Immediate notification requirements for operating nuclear

power reactors.

(a) General requirements.7 (1) Each nuclear power

reactor licensee licensed under Sec. 50.21(b) or Sec. 50.22 of this

part shall notify the NRC Operations Center via the Emergency

Notification System of:

---------------------------------------------------------------------------

\7\ Other requirements for immediate notification of the NRC by

licensed operating nuclear power reactors are contained elsewhere in

this chapter, in particular Secs. 20.1906, 20.2202, 50.36, 72.74,

72.75, and 73.71.

---------------------------------------------------------------------------

(i) The declaration of any of the Emergency Classes specified in

the licensee's approved Emergency Plan; 8 or

---------------------------------------------------------------------------

\8\ These Emergency Classes are addressed in Appendix E of this

part.

---------------------------------------------------------------------------

(ii) Of those non-Emergency events specified in paragraph (b) of

this section.

(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.9, 10

---------------------------------------------------------------------------

\9\ Commercial telephone number of the NRC Operations Center is

(301) 816-5100.

\10\ [Reserved]

---------------------------------------------------------------------------

(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) 11 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

[[Page 36305]]

system has the capability to transmit the exercise data.

---------------------------------------------------------------------------

\11\ Requirements for ERDS are addressed in Appendix E, Section

VI.

---------------------------------------------------------------------------

(5) When making a report under paragraph (a)(1) of this section,

the licensee shall identify:

(i) The Emergency Class declared; or

(ii) Either paragraph (b)(1), ``One-Hour Report,'' or paragraph

(b)(2) ``Eight-Hour Report,'' as the paragraph of this section

requiring notification of the Non-Emergency Event.

(b) Non-emergency events--(1) One-Hour reports. If not reported as

a declaration of the 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 Sec. 50.54(x)

of this part.

(2) Eight-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 eight 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) 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 affects plant safety.

(iii) Any natural phenomenon or other external condition 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 plant.

(iv)(A) Any event or condition that results in intentional manual

actuation or valid automatic actuation of any of the systems listed in

paragraph (b)(2)(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)(2)(iv)(A) of this section apply are:

(1) Reactor protection system (reactor scram, reactor trip).

(2) 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.

(3) ECCS for boiling water reactors (BWRs) including: high-pressure

and low-pressure core spray systems; high-pressure coolant injection

system; feedwater coolant injection system; low pressure injection

function of the residual heat removal system; and automatic

depressurization system.

(4) BWR isolation condenser system and reactor core isolation

cooling system.

(5) PWR auxiliary feedwater system.

(6) Containment systems including: containment and reactor vessel

isolation systems (general containment isolation signals affecting

numerous valves and main steam isolation valve [MSIV] closure signals

in BWRs) and containment heat removal and depressurization systems,

including containment spray and fan cooler systems.

(7) Emergency ac electrical power systems, including: emergency

diesel generators (EDGs) and their associated support systems;

hydroelectric facilities used in lieu of EDGs at the Oconee Station;

safety related gas turbine generators; BWR dedicated Division 3 EDGs

and their associated support systems; and station blackout diesel

generators (and black-start gas turbines that serve a similar purpose)

which are started from the control room and included in the plant's

operating and emergency procedures.

(8) Anticipated transient without scram (ATWS) mitigating systems.

(9) Service water (standby emergency service water systems that do

not normally run).

(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)(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 this paragraph if redundant equipment in

the same system was operable and available to perform the required

safety function.

(vii) [Reserved]

(viii)(A) 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.

(B) 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.

(ix) Any 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 including fires, toxic gas releases, or radioactive

releases.

(x) Any event requiring the transport of a radioactively

contaminated person to an offsite medical facility for treatment.

(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.

(xii) 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:

Sec. 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. Unless otherwise specified in

this section, the licensee shall report an event 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 occurring within three years of the

date of discovery which was prohibited by the plant's Technical

Specifications, except when:

(1) The technical specification is administrative in nature; or

[[Page 36306]]

(2) The event consists 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 capable of performing its specified safety

functions.

(C) Any deviation from the plant's Technical Specifications

authorized pursuant to Sec. 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;

(B) The nuclear power plant being in an unanalyzed condition that

significantly affects plant safety; or

(C) 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.

(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 (reactor scram, reactor trip).

(2) 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.

(3) ECCS for boiling water reactors (BWRs) including: high-pressure

and low-pressure core spray systems; high-pressure coolant injection

system; feedwater coolant injection system; low pressure injection

function of the residual heat removal system; and automatic

depressurization system.

(4) BWR isolation condenser system and reactor core isolation

cooling system.

(5) PWR auxiliary feedwater system.

(6) Containment systems including: containment and reactor vessel

isolation systems (general containment isolation signals affecting

numerous valves and main steam isolation valve [MSIV] closure signals

in BWRs) and containment heat removal and depressurization systems,

including containment spray and fan cooler systems.

(7) Emergency ac electrical power systems, including: emergency

diesel generators (EDGs) and their associated support systems;

hydroelectric facilities used in lieu of EDGs at the Oconee Station;

safety related gas turbine generators; BWR dedicated Division 3 EDGs

and their associated support systems; and station blackout diesel

generators (and black-start gas turbines that serve a similar purpose)

which are started from the control room and included in the plant's

operating and emergency procedures.

(8) Anticipated transient without scram (ATWS) mitigating systems.

(9) Service water (standby emergency service water systems that do

not normally run).

(v) Any event or condition occurring within three years of the date

of discovery that 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 (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 this paragraph 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) 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)(1) The Energy Industry Identification System component function

identifier and system name of each component or system referred to in

the LER.

(i) 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.

(ii) IEEE Std 803-1983 has been approved for incorporation by

reference by the Director of the Federal Register.

(2) 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,

345 East 47th Street, New York, NY 10017. 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; and at the Office of the Federal Register, 1100 L

Street, NW, Washington, DC.

* * * * *

(J) For each human performance related problem that contributed to

the event, 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 the availability of systems or

components that:

[[Page 36307]]

(i) Could have performed the same function as the components and

systems that failed during the event, or

(ii) 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.

* * * * *

(d) Submission of reports. Licensee Event Reports must be prepared

on Form NRC 366 and submitted within 60 days of discovery of a

reportable event or situation to the U.S. Nuclear Regulatory

Commission, as specified in Sec. 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:

Sec. 72.216 Reports.

(a) [Reserved]

(b) [Reserved]

(c) The general licensee shall make initial and written reports in

accordance with Secs. 72.74 and 72.75.

Dated at Rockville, Maryland, this 25th day of June, 1999.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook,

Secretary of the Commission.

[FR Doc. 99-16934 Filed 7-2-99; 8:45 am]

BILLING CODE 7590-01-P