[Federal Register Volume 59, Number 211 (Wednesday, November 2, 1994)]
[Proposed Rules]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-27126]
[[Page Unknown]]
[Federal Register: November 2, 1994]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 59, No. 211 / Wednesday, November 2, 1994 /
Proposed Rules
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50, 55, AND 73
RIN 3150-AF18
Reduction of Reporting Requirements Imposed on NRC Licensees
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule.
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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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