Reduction of Reporting Requirements Imposed on NRC Licensees

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: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its

regulations to reduce reporting requirements currently imposed on

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

nuclear material licensees. This rule reduces the regulatory burden on

NRC licensees; and partially implements a recent NRC initiative to

revise or eliminate duplicative or unnecessary reporting requirements.

The amendments will: 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; 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 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.

EFFECTIVE DATE: April 13, 1995.

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

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

20555. Telephone (301) 415-6103.

SUPPLEMENTARY INFORMATION:

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

final rule is one of several rulemakings and other regulatory actions

currently being developed by the NRC staff to implement the Regulatory

Review Group recommendations to eliminate duplicative or unnecessary

reporting requirements. The NRC believes that this action will reduce

the regulatory burden on NRC licensees without causing adverse effects

on the protection of public health and safety.

On November 2, 1994 (59 FR 54843), the NRC published the notice of

proposed rulemaking that reduces reporting requirements on licensees

under Parts 50, 55, and 73. Specifically, the proposed amendments were

intended to: (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. The public comment period expired December 19, 1994.

Analysis of Public Comments on the Proposed Rule

The NRC received seven comments: one from Nuclear Energy Institute

(NEI), an organization that represents the nuclear power industry, five

from the nuclear power industry, and one from Ohio Citizens for

Responsible Energy, Inc. (OCRE). The comments from NEI and the nuclear

power industry are supportive of the proposed rule to reduce the

reporting requirements. OCRE opposes the proposed rule. However, all

commenters believe that elimination of these reports will not adversely

impact public health and safety. The following section addresses the

public comments received and provides NRC's response to them.

Of the six comments received which favor the proposed rule, several

of those endorsing the rule pointed out that the proposed changes

eliminate unnecessary or redundant requirements and conserve both NRC

and licensee resources. Two of the commenters felt that the NRC should

assess additional reporting requirements to determine whether they can

be eliminated or reduced in frequency. As discussed in the background

section of this rulemaking, the NRC has underway several regulatory

activities to implement the Regulatory Review Group's recommendations

to eliminate duplicative or unnecessary reporting requirements. This

rulemaking is limited to the requirements set out in the proposed

rulemaking.

Licensees do not Need to Assemble the Summary Report

One commenter from the nuclear power industry states that the

requirement to generate but not submit a summary report for the

containment leakage tests provides no additional benefit and is an

unnecessary burden since the summary report contains data readily

available from other sources. The commenter suggests that the

requirement to generate the summary report be eliminated.

The NRC disagrees. The NRC believes that the results of containment

leakage tests, the licensee analysis verifying the acceptability of the

results, as well as any necessary interpretations of the results, is

necessary information which might not be documented absent this

documentation requirement. Furthermore, the assembly of a summary

report will provide access by NRC inspectors and auditors to this

information in a more timely fashion.

Public Participation in the NRC Regulatory Process Will Diminish

OCRE opposes the proposed rule because it believes that adoption of

the rule will diminish the public's access to information. OCRE states

that the public's health and safety is not the only factor to consider

when NRC proposes to eliminate some licensee reports. Access to these

reports, OCRE states, is vital for effective public participation in

the regulatory process.

To that end, OCRE has filed a petition for rulemaking with the NRC

(59 FR 30308, June 13, 1994). The purpose of the petition is to

establish public right-to-know provisions which would ensure public

access to licensee-held information.

In each case where the NRC considers eliminating a reporting

requirement, the NRC first considers the public health and safety

impact of the proposed elimination. If there is no direct impact on

public health and safety, the NRC also considers the reduced

administrative burden on the licensee and the extent to which the

proposed elimination will deprive the public of important health and

safety information. OCRE's comments have raised the generic issue of

the incremental and cumulative effect of this and similar

[[Page 13616]] rulemakings in depriving the public of access to

licensee information that was previously available from the NRC. In

that regard, OCRE has directly presented this issue to the Commission

through its petition for rulemaking referenced above and the NRC finds

that this generic issue is better addressed in the context of that

petition, rather than in individual rulemakings such as this one. The

NRC also finds that the effect of this rulemaking will be to reduce the

administrative burden on licensees and that the loss of the information

in this particular case will not adversely affect the public interest

in access to information regarding adequate protection of the public

health and safety.

Having considered all comments received and other input, the NRC

has determined that the following final rule should be promulgated.

Written Reports

This final rule would not require additional written reports. On

the contrary, under this final rule, reporting will be reduced for all

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

Criminal Penalties

For purposes of Section 223 of the Atomic Energy Act of 1954, as

amended, relating to willful violations of requirements notice is

hereby given that these amendments are being adopted and promulgated

pursuant to Sections 161b, 161i, or 161o of the Act.

Environmental Impact: Categorical Exclusion

The NRC has determined that this final 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 final rule amends information collection requirements that are

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

seq.). These requirements were approved by the Office of Management and

Budget, approval numbers 3150-0011, -0018, and -0002.

Because the rule will relax existing information collection

requirements, the annual public burden for this collection of

information is expected to be reduced by approximately 20 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, -0018, -

0002), Office of Management and Budget, Washington, DC 20503.

Regulatory Analysis

The Commission has prepared a regulatory analysis on this final

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

considered by the Commission. The Commission requested public comment

on the draft regulatory analysis, but no comments were received.

Therefore, no changes to the draft regulatory analysis have been made.

The draft regulatory analysis is adopted as the final regulatory

analysis without change. The analysis is available for inspection in

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

Washington, DC.

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 have a

significant economic impact on a substantial number of small entities.

This final 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 final rule because these amendments do not involve

any provisions which would impose backfits on licensees as defined in

Sec. 50.109(a)(1). In addition, information collection and reporting

requirements are not subject to the backfit rule.

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.

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 10 CFR 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

* * * * * [[Page 13617]]

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 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 report 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 report

required by V.B.

B. Recordkeeping 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. 10 CFR 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 described 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 Sec. 1701, 106 Stat.

2951, 2952, 2953 (42 U.S.C. 5841, 5844, 2297f).

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. Section 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 [[Page 13618]] 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 on 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 2nd day of March, 1995.

For the Nuclear Regulatory Commission.

James M. Taylor,

Executive Director for Operations.

[FR Doc. 95-6210 Filed 3-13-95; 8:45 am]

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