[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
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 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]
BILLING CODE 7590-01-P