Reduction of Reporting Requirements Imposed on NRC Licensees

SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend

its regulations to reduce reporting requirements currently imposed on

water-cooled nuclear power reactor, research and test reactor, and

nuclear material licensees. This action would reduce the regulatory

burden on NRC licensees. The proposed rule would implement an NRC

initiative to review its current regulations with the intent to revise

or eliminate duplicative or unnecessary reporting requirements. The

proposed amendments would: (1) Eliminate the current requirement for

licensees to submit summary reports of containment leakage rate tests

to the NRC (10 CFR Part 50--Appendix J), but preserve the requirements

in Secs. 50.72 and 50.73 under which licensees currently report any

instances of leakage exceeding authorized limits in the technical

specifications of the license; (2) revise 10 CFR 55.25 to refer

licensees to a similar reporting requirement in 10 CFR 50.74(c) and

require notification of operator incapacity only in case of permanent

disability or illness; and (3) eliminate the requirement for quarterly

submittal of safeguards event logs presently contained in 10 CFR

73.71(c)(2) and Appendix G to Part 73.

DATES: The comment period expires December 19, 1994. Comments received

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

Commission is able to assure consideration only for comments received

on or before this date.

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

Commission, Washington, DC 20555-0001, Attention: Docketing and Service

Branch. Comments may be delivered to One White Flint North, 11555

Rockville Pike, Rockville, MD, between 7:45 a.m. and 4:15 p.m. on

Federal workdays.

Copies of the draft regulatory analysis, the finding of no

significant impact, the supporting statement submitted to OMB, and

comments received may be examined at the NRC Public Document Room, 2120

L Street NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Naiem S. Tanious, Office of Nuclear

Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC

20555. Telephone (301) 415-6103.

Background

On January 7, 1994, the Executive Director for Operations (EDO)

sent to the Commission SECY-94-003, ``Plan for Implementing Regulatory

Review Group Recommendations.'' The Commission approved these

recommendations for reducing regulatory burden on its licensees. This

proposed rule is one of several rulemaking and other regulatory actions

that the NRC staff is developing to implement those recommendations.

During the NRC staff review of the regulations, Federal Register

notices were published on February 24, 1992 (57 FR 6299) and June 19,

1992 (57 FR 27394) that solicited the views of the public, the nuclear

power industry, and other interested parties regarding reduction of the

regulatory burden and reporting requirements. Comments were received in

response to those notices. A summary of the comments received that are

pertinent to this action is included in this document.

Discussion

These proposed amendments would: (1) Eliminate the current

requirement for licensees to submit summary reports of containment

leakage rate tests to the NRC (10 CFR Part 50-Appendix J), but preserve

the requirements in Secs. 50.72 and 50.73 under which licensees

currently report any instances of leakage exceeding authorized limits

in the technical specifications of the license; (2) revise 10 CFR 55.25

to refer licensees to a similar reporting requirement in 10 CFR

50.74(c) and require notification of operator incapacity only in case

of permanent disability or illness; and (3) eliminate the requirement

for quarterly submittal of safeguards event logs presently contained in

10 CFR 73.71(c)(2) and Appendix G to Part 73.

Although these proposed reduction in reporting requirements were

discussed in Federal Register notices published on February 24, 1992

(57 FR 6299) and June 19, 1992 (57 FR 27394), the public is again

invited to submit comments. Specifically, the NRC requests comments and

supporting rationale on the appropriateness of eliminating or

consolidating these reporting requirements and whether the public

health and safety will be adversely affected by these changes.

Elimination of Reporting Requirements from 10 CFR Part 50, Appendix J

10 CFR Part 50, Appendix J, currently requires all water-cooled

nuclear power reactor licensees to conduct containment leakage testing.

The containment leakage tests demonstrate that the containment system

meets all the leakage criteria specified in the technical

specifications of the licenses. Currently, Section V.B. of Appendix J

requires licensees to submit a summary report of the results of all

leak rate tests and any associated corrective actions. Under this

proposed rulemaking, licensees of water-cooled nuclear power reactors

will continue to conduct containment leakage testing and to prepare the

summary report. However, they would not be required to submit the

summary report to the NRC. They would still be required to report to

the NRC instances of leakage in excess of authorized limits, via a

written licensee event report,1 as now required by

Sec. 50.73(a)(2)(ii). If such a leakage condition is found during

operation, an immediate notification by telephone is required by

Sec. 50.72(b)(1)(ii). If the leakage condition is found during shutdown

the telephone notification is required by Sec. 50.72(b)(2)(i).

\1\These reports would be required when total containment as-

found, minimum pathway leak rate exceeds the limiting condition for

operation (LCO) in the facility's technical specification.

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The NRC believes that the elimination of the requirement to submit

the summary report to the NRC of leakage tests when these results are

within acceptance limits would have no impact on the overall health and

safety of the public. Because these tests have been performed and

evaluated frequently by the nuclear power industry, any

misinterpretation of testing requirements is highly unlikely. Moreover,

licensees would still be required to prepare the summary reports and

make those reports available for review and inspection at the

respective plant sites. Having these reports available at the plant

sites should be sufficient for normal record reviews, and for any

necessary in-depth reviews. Therefore, the NRC proposes to eliminate

the requirement to report results of tests within specified limits.

Consolidation of 10 CFR 50.74 and 10 CFR 55.25 Reporting Requirements

If an operator licensed pursuant to 10 CFR 55, becomes ill or

disabled to the point that he or she no longer can safely perform their

duties, the reactor licensee is required to report the occurrence of

disability under both 10 CFR 50.74(c) and 10 CFR 55.25. The NRC is

proposing to require only a single report by eliminating the reporting

requirements in 10 CFR 55.25 and modifying 10 CFR 55.25 to refer

facility licensees to 10 CFR 50.74(c).

In addition, when 10 CFR Part 55 was promulgated, the intent of

Sec. 55.25 was to receive reports only of permanent or potentially

permanent illness or disability of licensed operators that would

prevent them from safely carrying out their responsibilities. However,

this intent, is not explicitly stated in either Sec. 55.25 or

Sec. 50.74(c). To remove this ambiguity, the word ``permanent'' is

added in both Secs. 50.74(c) and 55.25. (A more detailed discussion on

``permanent'' versus ``temporary,'' illness, or disability can be found

in the NRC publication NUREG-1262,2 ``Answers to Questions at

Public Meetings Regarding Implementation of Title 10, Code of Federal

Regulations, Part 55 on Operators' Licenses,'' November 1987, page 21,

question 91).

\2\Copies of NUREG-1262 may be purchased from the Superintendent

of Documents, U.S. Government Printing Office, Mail Stop SSOP,

Washington, DC 20402-9328. Copies are also available from the

National Technical Information Service, 5285 Port Royal Road,

Springfield, VA 22161. A copy is also available for inspection and

copying for a fee in the NRC Public Document Room, 2120 L Street,

NW. (Lower level), Washington, DC 20555-0001.

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Public Comments

Only two comments were received concerning the reporting

requirements for power reactor licensees. Neither suggested elimination

of any power reactor reporting requirement. However, both suggested

that the redundant requirements of 10 CFR Parts 50 and 55 addressing

illness or disability of licensed operator be consolidated in 10 CFR

50.74.

Elimination of Reporting Requirements in 10 CFR Part 73.71(c)(2)

10 CFR Part 73.71(c)(1) requires that licensees maintain a current

log for recording safeguards events. An event that must be recorded in

the log is defined in Appendix G, Part 73 as ``Any failure,

degradation, or discovered vulnerability in a safeguard system. * *

*.''3 10 CFR 73.71(c)(2) requires that a copy of the log be

submitted quarterly to the NRC.

\3\The full definition in 10 CFR Part 73, Appendix G, Section II

is: (a) Any failure, degradation, or discovered vulnerability in a

safeguard system that could have allowed unauthorized or undetected

access to a protected area, material access area, controlled access

area, vital area, or transport had compensatory measures not been

established. (b) Any other threatened, attempted, or committed act

not previously defined in Appendix G with the potential for reducing

the effectiveness of the safeguard system below that committed to in

a licensed physical security or contingency plan or the actual

condition of such reduction in effectiveness.

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The NRC proposes to eliminate the requirement that licensees submit

copies of the safeguard event logs. Until recently, the NRC staff

published an annual report which contained trending analysis of log

events. However, the NRC now believes that the greatest benefits of

dissemination of these statistics on safeguards equipment performance

and lessons learned about the causes and prevention of safeguards

equipment malfunctions have been realized, and that continuing to

publish that report is no longer cost effective. However, licensees

will still be required to enter events in the logs, and make those logs

available for review and inspection at the respective plant sites.

Having the logs available at the plant site should be sufficient for

normal record reviews, and any necessary in-depth reviews. Therefore,

the NRC believes that public health and safety will not be adversely

affected if the logs are no longer submitted to the NRC.

Public Comments

The former Nuclear Management and Resources Council, now known as

the Nuclear Energy Institute (NEI), commented that power reactor

licensees should be deleted from the list of licensees subject to the

provisions of 10 CFR 73.71(c). According to NEI, comparisons among

plants using the data provided in the logs are not meaningful because

the number of events reported by each site is dramatically influenced

by a number of site-specific variables such as the number and design of

system components and unique physical arrangements. NEI stated that

comments received from industry were almost unanimous in advising that

licensees receive insignificant information from the NRC's quarterly

``Safeguards Events Analysis Report.'' NEI further commented that the

real benefit in recording safeguards events lies in its usefulness as a

management tool to measure a plant's specific performance, independent

of other facilities.

One licensee commented that if the requirement to submit a log to

the NRC were not deleted, the frequency of submittal should be reduced

from 4 times each year to 2 times each year as required for submittal

of fitness-for-duty performance data in 10 CFR 26.71(d). The licensee

noted that timeliness would not be adversely impacted in a significant

way by annual or semiannual rather than quarterly reporting. The

licensee also suggested that evaluation of trends is more meaningful

when based on events over 6 months or a year rather than only 3 months.

The NRC believes that, in the early years of this program, there

was considerable benefit from comparisons of the performance of a

site's security equipment with the performance of the rest of the

industry, notwithstanding differences in site-specific variables.

However, the NRC now believes that the greatest benefits have been

realized and that continuing the program as a regulatory tool has a

diminishing cost benefit. As such, the NRC agrees with the comments

that the primary benefit in logging events is the usefulness of the log

as a means for the licensees to track and trend the performance of the

safeguards systems at their own plants. In fact, the NRC has already

discontinued publication of the ``Safeguards Events Analysis Report.''

Although the NRC is proposing to eliminate the requirement that

licensees submit their safeguards event logs, licensees would still be

required to enter events into their logs and maintain those logs on

site for review by the NRC inspectors.

Written Reports

This proposed rule would not require additional written reports. On

the contrary, under this proposed rule, reporting will be reduced for

all licensees under 10 CFR Parts 50, 55, and 73.

Environmental Impact: Categorical Exclusion

The NRC has determined that this proposed rule is the type of

action described in the categorical exclusion, 10 CFR 51.22(c)(3)(iii).

Therefore, neither an environmental impact statement nor an

environmental assessment has been prepared for this regulation.

Paperwork Reduction Act Statement

This proposed rule amends information collection requirements that

are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et

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

Budget for review and approval of the paperwork requirements.

Because the rule will relax existing information collection

requirements, the public burden for this collection of information is

expected to be reduced by approximately 10 hours per licensee. This

reduction includes the time required for reviewing instructions,

searching existing data sources, gathering and maintaining the data

needed and completing and reviewing the collection of information. Send

comments regarding the estimated burden reduction or any other aspect

of this collection of information, including suggestions for reducing

this burden, to the Information and Records Management Branch (T-6

F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001;

and to the Desk Officer, Office of Information and Regulatory Affairs,

NEOB-10202 (3150-0011, 3150-0018, 3150-0002), Office of Management and

Budget, Washington, DC 20503.

Regulatory Analysis

The Commission has prepared a draft regulatory analysis on this

proposed regulation. 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, 2120 L Street

NW. (Lower Level), Washington, DC. Single copies of the draft analysis

may be obtained from Naiem S. Tanious, telephone (301) 415-6103. The

Commission requests public comment on the draft regulatory analysis.

Comments on the draft analysis may be submitted to the NRC as indicated

under the ADDRESSES heading.

Regulatory Flexibility Certification

In accordance with the Regulatory Flexibility Act of 1980 (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 the nuclear power

reactors, research and test reactors, and some material licensees. The

companies and organizations that own these plants do not fall within

the scope of the definition of ``small entities'' set forth in the

Regulatory Flexibility Act of the size standards established by the NRC

(56 FR 56671; November 6, 1991).

Backfit Analysis

The NRC has determined that the backfit rule 10 CFR 50.109, does

not apply to this proposed rule because these amendments do not involve

any provisions which would impose backfits on licensees as defined in

Sec. 50.109(a)(1). Information collection and reporting requirements

are not subject to the backfit rule; moreover, the changes proposed in

this rulemaking relax existing requirements.

List of Subjects

10 CFR Part 50

Antitrust, Classified information, Criminal Penalties, Fire

protection, Intergovernmental relations, Nuclear power plants and

reactors, Radiation protection, Reactor siting criteria, Reporting and

recordkeeping requirements.

10 CFR Part 55

Criminal Penalties, Manpower training programs, Nuclear power

plants and reactors, Reporting and recordkeeping requirements.

10 CFR Part 73

Criminal Penalties, Hazardous materials transportation, Export,

Import, Nuclear materials, Nuclear power plants and reactors, Reporting

and recordkeeping requirements, Security measures.

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; the Nuclear Waste Policy Act of 1982, as

amended; and 5 U.S.C. 553; the Commission is proposing to adopt the

following amendments to 10 CFR Parts 50, 55, and 73.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION

FACILITIES

1. The authority citation for 10 CFR Part 50 continues to read as

follows:

Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68

Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,

83 Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,

2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88

Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.

2951 as amended by Pub. L. 102-486, sec. 2902, 106 Stat 3123, (42

U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68

Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.

91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd),

and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42

U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued

under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,

50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83

Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued

under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58,

50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42

U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939

(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,

68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued

under sec. 187, 68 Stat. 955 (42 U.S.C 2237).

2. In Sec. 50.74, paragraph (c) is revised to read as follows:

Sec. 50.74 Notification of change in operator or senior operator

status.

* * * * *

(c) Permanent disability or illness as described in Sec. 55.25 of

this chapter.

3. In 10 CFR Part 50 Appendix J, Section III, paragraphs A.1. (a),

(b), and (d); Section IV. paragraph A., and Section V. paragraphs A.

and B., are revised to read as follows:

Appendix J to Part 50--Primary Reactor Containment Leakage Testing for

Water-Cooled Power Reactors.

* * * * *

III. Leakage Testing Requirements.

* * * * *

A. Type A test-1. Pretest requirements. (a) Containment

inspection in accordance with V. A. shall be performed as a

prerequisite to the performance of Type A tests. During the period

between the initiation of the containment inspection and the

performance of the Type A test, no repairs or adjustments shall be

made so that the containment can be tested in as close to the ``as

is'' condition as practical. During the period between the

completion of one Type A test and the initiation of the containment

inspection for the subsequent Type A test, repairs or adjustments

shall be made to components whose leakage exceeds that specified in

the technical specification as soon as practical after

identification. If during a Type A test, including the supplemental

test specified in III.A.3.(b), potentially excessive leakage paths

are identified which will interfere with satisfactory completion of

the test, or which result in the Type A test not meeting the

acceptance criteria III.A.4.(b) or III.A.5.(b), the Type A test

shall be terminated and the leakage through such paths shall be

measured using local leakage testing methods. Repairs and/or

adjustments to equipment shall be made and Type A test performed.

The corrective action taken and the change in leakage rate

determined from the tests and overall integrated leakage determined

from local leak and Type A tests shall be included in the summary

report required by V.B.

(b) Closure of containment isolation valves for the Type A test

shall be accomplished by normal operation and without any

preliminary exercising or adjustments (e.g., no tightening of valve

after closure by valve motor). Repairs of maloperating or leaking

valves shall be made as necessary. Information on any valve closure

malfunction or valve leakage that require corrective action before

the test, shall be included in the summary report required by V.B.

* * * * *

(d) Those portions of the fluid systems that are part of the

reactor coolant pressure boundary and are open directly to the

containment atmosphere under post-accident conditions and become an

extension an extension of the boundary of the containment shall be

opened or vented to the containment atmosphere prior to and during

the test. Portions of closed systems inside containment that

penetrate containment and rupture as a result of a loss of coolant

accident shall be vented to the containment atmosphere. All vented

systems shall be drained of water or other fluids to the extent

necessary to assure exposure of the system containment isolation

valves to containment air test pressure and to assure they will be

subjected to the post accident differential pressure. Systems that

are required to maintain the plant in a safe condition during the

test shall be operable in their normal mode, and need not be vented.

Systems that are normally filled with water and operating under

post-accident conditions, such as the containment heat removal

system, need not be vented. However, the containment isolation

valves in the systems defined in III.A.1.(d) shall be tested in

accordance with III.C. The measured leakage rate from these tests

shall be included in the summary required by V.B.

* * * * *

IV. Special Testing Requirements.

A. Containment modification. Any major modification, replacement

of a component which is part of the primary reactor containment

boundary, or resealing a seal-welded door, performed after the

preoperational leakage rate test shall be followed by either a Type

A, Type B, or Type C test, as applicable for the area affected by

the modification. The measured leakage from this test shall be

included in the summary report required by V.B. The acceptance

criteria of III.A.5.(b), III.B.3., or III.C.3., as appropriate,

shall be met. Minor modifications, replacements, or resealing of

seal-welded doors, performed directly prior to the conduct of a

scheduled Type A test do not require a separate test.

* * * * *

V. Inspection and Reporting of Tests.

A. Containment inspection. A general inspection of the

accessible interior and exterior surfaces of the containment

structures and components shall be performed prior to any Type A

test to uncover any evidence of structural deterioration which may

affect either the containment structural integrity or leak-

tightness. If there is evidence of structural deterioration, Type A

tests shall not be performed until corrective action is taken in

accordance with repair procedures, non destructive examinations, and

tests as specified in the applicable code specified in Sec. 50.55a

at the commencement of repair work. Such structural deterioration

and corrective actions taken shall be included in the summary test

report required by V.B.

B. Report of Test Results. 1. The preoperational and periodic

tests must be documented in a readily available summary report that

will be made available for inspection, upon request, at the nuclear

power plant. The summary report shall include a schematic

arrangement of the leakage rate measurement system, the

instrumentation used, the supplemental test method, and the test

program selected as applicable to the preoperational test, and all

the subsequent periodic tests. The report shall contain an analysis

and interpretation of the leakage rate test data for the Type A test

results to the extent necessary to demonstrate the acceptability of

the containment's leakage rate in meeting acceptance criteria.

2. For each periodic test, leakage test results from Type A, B,

and C tests shall be included in the summary report. The summary

report shall contain an analysis and interpretation of the Type A

test results and a summary analysis of periodic Type B and Type C

tests that were performed since the last type A test. Leakage test

results from type A, B, and C tests that failed to meet the

acceptance criteria of III.A.5(b), III.B.3, and III.C.3,

respectively, shall be included in a separate accompanying summary

report that includes an analysis and interpretation of the test

data, the least squares fit analysis of the test data, the

instrumentation error analysis, and the structural conditions of the

containment or components, if any, which contributed to the failure

in meeting the acceptance criteria. Results and analyses of the

supplemental verification test employed to demonstrate the validity

of the leakage rate test measurements shall also be included.

PART 55--OPERATORS' LICENSES

4. The authority citation for 10 CFR Part 55 continues to read as

follows:

Authority: Secs. 107, 161, 182, 68 Stat. 939, 948, 953, as

amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2137, 2201,

2232, 2282); secs. 201, as amended, 202, 88 Stat. 1242, as amended,

1244 (42 U.S.C. 5841, 5842).

Sections 55.41, 55.43, 55.45, and 55.59 also issued under sec.

306, Pub. L. 97-425, 96 Stat. 2262 (42 U.S.C. 10226). Section 55.61

also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236,

2237).

5. Section 55.25 is revised to read as follows:

Sec. 55.25 Incapacitation because of disability or illness.

If, during the term of the license, the licensee develops a

permanent physical or mental condition that causes the licensee to fail

to meet the requirements of Sec. 55.21 of this part, the facility

licensee shall notify the Commission, within 30 days of learning of the

diagnosis, in accordance with Sec. 50.74(c). For conditions for which a

conditional license (as describing in Sec. 55.33(b) of this part) is

requested, the facility licensee shall provide medical certification on

Form NRC 396 to the Commission (as described in Sec. 55.23 of this

part).

PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS

6. The authority citation for 10 CFR Part 73 continues to read as

follows:

Authority: Secs. 53, 161, 68 Stat. 930, 948, as amended, sec.

147, 94 Stat. 780 (42 U.S.C. 2073, 2167, 2201); sec. 201, as

amended, 204, 88 Stat. 1242, as amended, 1245 (42 U.S.C. 5841,

5844).

Section 73.1 also issued under secs. 135, 141, Pub. L. 97-425,

96 Stat. 2232, 2241 (42 U.S.C, 10155, 10161). Section 73.37(f) also

issued under sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841

note). Section 73.57 is issued under sec. 606, Pub. L. 99-399, 100

Stat. 876 (42 U.S.C. 2169).

7. In Sec. 73.71, paragraph (c)(2) is deleted, paragraph (c)(1) is

redesignated as paragraph (c), and paragraph (d) is revised to read as

follows:

Sec. 73.71 Reporting of safeguards events.

* * * * * *

(d) Each licensee shall submit to the Commission the 30-day written

reports required under the provisions of this section that are of a

quality which will permit legible reproduction and processing. If the

facility is subject to Sec. 50.73 of this chapter, the licensee shall

prepare the written report of NRC Form 366. If the facility is not

subject to Sec. 50.73 of this chapter, the licensee shall not use this

form but shall prepare the written report in letter format. The report

must include sufficient information for NRC analysis and evaluation.

8. In 10 CFR Part 73, Appendix G, the title of Section II is

revised to read as follows:

Appendix G to Part 73--Reportable Safeguards Events

* * * * *

II. Events to be recorded within 24 hours of discovery in the

safeguards event log.

* * * * *

Dated at Rockville, Maryland, this 20th day of October, 1994.

For the Nuclear Regulatory Commission.

James M. Taylor,

Executive Director for Operations.

[FR Doc. 94-27126 Filed 11-1-94; 8:45 am]

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