ML20211K998

Recommends That Commission Approve Issuance of Proposed Rev to 10CFR50,App J, Leakage Tests for Containments of Light Water Cooled Nuclear Power Plants & Draft Reg Guide,Task Ms 021-5

ML20211K998
Person / Time
Issue date:	05/29/1986
From:	Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
Shared Package
ML20210N314	List: ML20210N309 ML20211K998
References
FOIA-87-14, RTR-REGGD-01.XXX, RTR-REGGD-1.XXX, TASK-MS-021-5, TASK-MS-21-5, TASK-RE, TASK-RINV, TASK-SE SECY-86-167, SECY-86-67, NUDOCS 8607010132
Download: ML20211K998 (128)

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May 29,1986 '%..... SECY-86-167 RULEMAKING ISSUE

. (Notation Vote)

For: The Commissioners From: Victor Stello,.i..

Executive Director for Operations

Subject:

ISSUANCE OF PROPOSED REVISION TO 10 CFR 50, APPENDIX J PLUS RELATED DRAFT REGULATORY GUIDE MS 021-5

Purpose:

To obtain Commission approval 1) for an exemption from S 50.109(a)(3) and 2) to publish for public comment a proposed rule to update 10 CFR 50, Appendix J, Leakage Tests for Containments of Light-Water-Cooled Nuclear Power Plants, and a related draft regulatory guide.

Issue: Issuance of these two documents on containment leakage testing is for the purpose of updating the existing 1973 regulation and endors-ing a related national standard. Other related, longer term, and broader issues are currently under review by the NRC staff, such as containment function, degree of integrity required, and valida-tion of that integrity under conditions other than the standard LOCA. These may or may not impact the test program covered by Appendix J. However, applying PRA insights and other results of the severe accident program is still a few years away, and the degree of impact, if any, on Appendix J cannot be estimated at~

this time. If affected, Appendix J is likely to be only one of many NRC positions that may have to be reviewed in the context of the conclusions of these studies. The proposed rule and draft regulatory guide are needed now by the NRC licensing and enforce-ment staffs in order to improve uniformity and efficiency in the regulation of this inservice inspection and test program and to conform to the current state of the art. However, an exemption is needed from S 50.109(a)(3) in order to continue with this effort.

Brckground: A. Appendix J of 10 CFR'50 was originally issued for public com-ment as a proposed rule on August 27, 1971 (36 FR 17053);

published in final form on February 14, 1973 (38 FR 4385); and became effective on March 16, 1973. The only amendment to this Appendix since 1973 was a limited one, on Type B (penetra-tion) test requirements that was published for comment on January 11, 1980 (45 FR 2330); published in final form ,

September 22, 1980 (45 FR 62789); and became effective on f_.

- October 22, 1980. .

Contact:

/^7 Gunter Arndt 443-7893

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The Commissioners 2 s- .s This revision of Appendix J will provide greater flexibility in applying alternative requirements due to variations in plant design and reflect changes due to: (1) experience in applying the existing requirements; (2) advances in containment leakage testing methods; (3) interpretive questions; (4) simplifying

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the text; (5) various external / internal comments since 1973; and (6) exemption requests received and approved.

B. The~ draft regulatory guide has'as its basis the 1981 standard ANSI /ANS 56.8, " Containment System Leakage Testing Requirements,"

that detailed the then-current state-of-the-art in containment leakage testing procedures and data reduction and analysis.

The standard is being endorsed in a guide rather than the rule.

This approach limits the rule to test criteria, and leaves endorsement of detailed test procedures and statistical data reduction techniques to a guide that can be revised as the testing technology changes. The draft guide is being presented along with the proposed rule revision. This is in compliance with a standing Commission request to provide at least an out-line of any regulatory guide (s) that will result from a proposed rule or rule revision. The rule and guide should both be issued simultaneously for public commen't and, eventually, as effective documents.

C. The nature of this rule revision and its related guide would permit their publication for public comment by the EDO. How-ever they are being submitted to the Commission for its infor-mation and for a notation vote due to prior interest expressed by the Commission in this rule. (Chilk to Dircks, 6/9/81 memo on Sequoyah-2. Commission requested staff to reconsider treat-ment of Appendix J waivers by proposing a rule change.)

D. A backfit analysis (Enclosure 9) is provided as a part of this package in accordance with 5 50.109. It will'also be announced in the Federal Register Notice for the proposed rule.

The analysis does not conclude that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit. It does conclude, however, that the direct and indirect costs of implementation are justified due to better, more uniform tests and test reports, greater confidence in the reliability of the test results, fewer exemption requests, and fewer interpretive debates.

The proposed rule is intended to be applied to the entire popu-lation of nuclear power reactors and clearly constitutes a backfit. .

This revision of the 1973 rule is not being proposed by the NRC staff on the basis of any substantial increase in safety N . . . . . - . . . . . . . . . . . - . . . . . . . . _ . . . . , . . -

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.s 3 or decrease in costs. Instead, it is being proposed as both safety and cost neutral. Justification for the revision is based on the need to conform present' testing capabilities to the current state of the art, and to use the best available procedures, thereby not freezing a stale (1972) technology.

The revision will keep _ rule requirements unambiguous, current, useful, consistent with practice, and flexible enough to accommodate differing plant designs.

For the benefit of the public, licensees, and the NRC staff, this proposed rule should be issued at this time for public comment. Therefore, in order to issue both the rule and regu-latory guide for public comment,_an exemption from the require-ments of the backfit rule, 6 50.109(a)(3), is requested for both the rule and the regulatory guide.

Discussion: A. The enclosed Federal Register notice for Appendix J requests public input beyond the immediate scope of this proposed rule.

The Invitation to Comment section lists a number of questions

+ on which comments are being solicited. The Federal Register notice also lists some assumptions inherent in Appendix J.

The NRC staff intends to use the public input both in develop-ing the final rule and as early input on the related issues addressed in the Federal Register Notice.

l The Federal Register Notice contains information on the scope of this rulemaking, and also addresses related, broader, longer term issues. The Appendix J revision scope is limited to cor-rections and clarifications, excluding new criteria. The pro- .

I posed rulemaking package addresses the current, urgent needs of those involved with conducting and regulating containment system leakage rate tests.

i B. The regulatory guide is needed to set forth the acceptability of the test procedures and data reduction techniques of ANSI / {

, ANS 56.8, " Containment System Leakage Testing Requirements."-

An NRC endorsement of this standard is needed together with the revision of Appendix J, since the standard presents means by which licensees can meet _the criteria of the proposed Appen-dix J. This standard also replaces ANSI N 45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear Reactors,"

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that is still referenced in the existing Appendix J.

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1. The Chief Counsel for Advocacy of the Small Business Administration will be inforned of the certification re-garding the economic impact on small entities and the reasons for it as required by the Regulatory Flexibility Act.

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ictor Stello, Jr.

Executive Director for Operations Enclcsures:

1. Federal Register Notice of Rulemaking

2. Environmental Assessment

3. OMB Reporting Review Package

4. Federal Register Notice of Regulatory Guide Availability

5. Proposed Draft Regulatory Guide

6. Draft Public Announcement

7. Draft Congressional Letters

8. ACRS 7/11/85 Letter

9. S 50.109 Backfit Analysis Commissioners' comments or consent should be provided directly to the Office of the Secretary by c.o.b. Monday, June 16, 1986.

Commission Staff Office comments, if any, should be submitted to the Commissioners NLT Monday, June 9, 1986, with an infor-mation copy to the Office of the Secretary. If the paper is of such a nature that it requires additional time for analytical review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected.

DISTRIBUTION: .

Commissioners OGC -

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ENCLOSURE 1 l

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Federal Register Notice

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, , ' NUCLEAR REGULATORY COMISSION ,

10 CFR Part 50 General Revision of Appendix J AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY

The Nuclear Regulatory Commission is proposing to amend its

. regulations to update the criteria and clarify questions of interpretation in regard to. leakage rate testing of containments of light-water-cooled nuclear power plants. The proposed rule would aid the licensing and en-forcement staff by eliminating conflicts, ambiguities, and a lack of uni-formity in the regulation of the inservice inspection program.

OATE: Comment period expires . Comments received after this date will be considered if it is practical to do so, but assurance i

' of consideration cannot be given except for comments received on or

! before this date.

l ADDRESSES: Written comments may be submitted to the Rules and Procedures h

Branch, Division of Rules and Records, Office of Administration, U.S.

Nuclear Regulatory Commission, Washington, DC 20555. Comments may also l

be delivered to Room 4000, Maryland National Bank Butiding, Bethesda, Maryland from 8:15 a.m. to 5:00 p.m. Monday through Friday. Copies of comments received may be examined at the NRC Public Document Room, 1717 H Street NW., Washington, DC.

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Copies of draft regulatory guide MS 021-5 may be obtained from the , ,

Nuclear Regulatory Commission, Document Management Branch, Washington, D: 20555.

FOR FURTHER INFORMATION CONTACT: Mr. E. Gunter Arndt, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 443-7893.

BACKGROUND SUPPLEMENTARY INFORMATION:

Appendix J of 10 CFR Part 50 was originally issued for public comment as a proposed rule on August 27, 1971 (36 FR 17053); published in final form on February 14, 1973 (38 FR 4385); and became effective on March 16, I 1973. The only amendment to this Appendix since 1973 was a limited one, on Type B (penetration) test requirements that was published for comment on January 11, 1980 (45 FR 2330); published in final form September 22, 1980 (45 FR 62789); and becaer. effective on October 22, 1980.

This revision of AppenJix J has been in preparation for some time.

i It will provide greater (1exibility in applying alternative requirements due to variations in plant design and reflects changes based on:

(1) experience in applying the existing requirements; (2) advances in f

containment leak testing methods; (3) interpretive questions;

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(4) simplifying the text (5) various external / internal comments since

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1973; and (6) exemption requests received and approved.

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. -, This proposed revision is for the purpose of updating the existing regulation. Other related, longer term, and broader issues are currently under review by the NRC staff, such as containment function, degree of int,egrity required, and validation of that integrity under conditions other than postulated in this rule. In order to better understand its function and scope, assumptions inherent in Appendix J are prasented as follow:

1. Certain levels of radiation exposure at the plant site boundary shall not be exceeded under (a) operating or (b) design basis accident conditions.

2. Certain levels of radiation exposure to plant operating personnel shall not be exceeded under (a) operating or (b) design basis accident conditions.

3. All four exposure levels (la, Ib, 2a, 2b) may be different, but can be calculated.

4. Defense-in-depth will be used for protection against these levels of exposures. As the final barrier, a containment system is re-quired in order to maintain any or all of these exposure limits.

5. The required degree of containment system leaktightness for design basis accidents can be (a) calculated, (b) specified, (c) built, (d) maintained, (e) inspected.

6. A generic inspection program can be defined that verifies the required leaktightness of the containment following construction and periodically throughout plant life.

7. NRC regulations should require such an inspection program, and define the test requirements and acceptance criteria.

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8. A standard loss-of-coolant accident is assumed as the i ,

design basis accident. Since the containment isolation system is an engineered safety feature, only safety grade systems and components

'are relied upon to define the containment boundary that must be exposed to the containment pneumatic test pressure for the integrated leak rate test. In addition, all safety grade systems are assumed to be subject to a potential single active failure, and must be locally leak rate tested accordingly.

9. Pneumatic testing to peak calculated accident pressure is O

adequate without testing for, or at, accident temperatures or radiation levels.

10. Shielding tests need not be performed.

11. Periodic testing provides adequate confidence in the level of containment system integrity. Continuous monitoring of all individual isolation barriers is not necessary.

The scope of this revision to Appendix J is limited to However, corrections and clarifications, and excludes new criteria.

this notice also addresses related, broader, longer term activities.

Following is information of some of these other related activities that are not reflected in this proposed rulemaking.

In order to better identify the availability of containment leakage integrity, concepts of " continuous containmentleakagemonitoring"(suchasnegativecontainment operatingpressure)and"relativelyfrequentgrosscontinment

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integrity check" (such as a low pressure pumpup just prior to 4

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operation to check for openings) are under consideration by the NRC staff.

These would identify large breaches of the containment system boundary, during, or just prior to, normal operating conditions. It should be noted they wou'Id only test the normal operating containment atmosphere boundary, ,

not the Appendix J, post-accident boundary including isolation valves.

Comments on these or alternative concepts, and what effect, if any, they would have on the proposed Appendix J requirements, are also being solicited in the following section of this preamble.

Past practice has been to implement the provisions of Appendix J by

• means of licensees' technical specifications. Currently, a Technical Specification Improvement Project (TSIP) is underway to reevaluate the NRC's philosophy and utilization of the technical specifications. While the proposed revision described herein assumes implementation of Appendix J by licensee's technical specifications, the work of the TSIP may lead to some changes in this form of implementation.

Another program is presently being conducted to identify current I NRC regulatory requirements that have marginal importance to safety and to recommend appropriate actions to modify or to eliminate these unneces-sary requirements. A Federal Register notice was published on October 3, 1984, to announce the initiation of the program (49 FR 39066). As a part of the program, regulatory requirements associated with containment leak-tightness are being evaluated. The risk and cost effectiveness of contain-

! ment leaktightness requirements will be examined to determine their value with respect to plant safety and possible alternative requirements.

Any resulting changes to existing regulations will be made through normal rulemaking procedures, including ACRS review and public comment.

Comments on the questions posed in this notice will also provide early, useful input to these associated activities.

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INVITATION TO COMMENT

• Coments from all interested persons on 'a11' aspects of this revision and on the risk and cost effectiveness of containment leaktightness in 1) general are requested by the coment expiration date in order that:

the final. revision will reflect consideration of all points of view, and

2) the staff's assessment of the risk importance of containment leaktightness can benefit from such comments. Especially requested are comments which address the following questions:

(1) the extent to which these positions in the proposed rule are already in use; (2) the extent to which those in use, and those not in use but proposed, are desirable;

"* ~ "(3) whether there continues to be a further need for this regulation; (4) estimates of the costs and benefits of this proposed revision, as a whole and of its separate provisions; (5) if the existing rule or its proposed revision were completely voluntary, how many licensees would adopt either version in its entirety and why; (6) since the NRC is planning a broader, more comprehensive review of containment functional and testing requirements in the next year or two, whether it is then still worthwhile to go forward with this proposed revision as an interim updating of the existing regulation; (7)whetherthetechnicalspecificationlimitsonallowablecontainment leakage should be relaxed and if so, to what extent and why, or if not, why not; (8) what risk-important factors influence containment performance under severe accident conditions, to what degree these factors.are considered in the current containment testing requirements.,and what approaches should be considered in addressing factors not presently covered; 6

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' * (9) what other approaches to validating containment integrity could be used that might provide detection of leakage paths as soon as they occur, whether they would result in any adjustments to the Appendix J

. test program and why; (10) what effect " leak-before-break" assumptions could have on the leak-age rate test program. Current accident assumptions use instant aneous complete breaks in piping systems, resulting in a test

" Leak-program based on pneumatic testing of vented, drained lines.

- before-break" assumptions presume that pipes will fail more gradually, leaking rather than instantly emptying. .

(11) how to effectively adjust Type A test results to reflect individual Type 8 and C test results obtained from inspections, repairs, adjust-ments, or replacements of penetrations and valves in the years in between Type A tests. Such an additional criterion, currently outside the scope of this proposed revision, would provide a more meaningful tracking of overall containment leaktightness on a more continuous basis than once every several years. The only existing or proposed criterion for Type 8 and C tests performed outside the outage in which a Type A test is performed is that the sum of Type 8 and C tests must not exceed 60% of the allowable containment leakage. Cur-rently being discussed by the NRC staff are:

a. All Type 8 and C tests performed during the same outage as a Type A test, or performed during a specified time period (nomi-nally 12 months) prior to a Type A test, be factored into the determination of a Type A test "as found" condition.

b. If a particular penetration or valve fails two consecutive Type B or C tests, the frequency of testing that penetration must be 1

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increased until two satisfactory B or C tests are obtained at the nominal test frequency. Concurrently, existing requirements' to increase the frequency of Type A tests due to consecutive ,

"as found" failures are already being relaxed in the proposed revision of Appendix J. Instead, attentic:* would be focused on correcting component degradation, no matter when tested, and the "as found" Type A test would reflect the actual condition of the overall containment boundary.

c. Increases or decreases in Type 8 or C "as found" test results (over the previous "as left" Type B or C test results) shall be added to or subtracted from the previous "as left" Type A test result.

measures If this sum exceeds 0.75 L, but is less than 1.0 L,,

shall be taken to reduce the sum to no more than 0.75 L,. This will not be considered a reportable condition.

If this sum exceeds 1.0 L,, measures shall be taken to reduce the sum to no more than 0.75 L,. This will be considered a reportable condition.

l The existing requirements that the sum of all Type B and C tests be no greater than 0.60 L,shall also remain in effect.

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Major Channes The following are the major changes proposed in this rulemaking.

1. Level of detail. The level of detail addressed in the proposed rev'ision'of Appendix J has been limited. This revision of the regulation defines general containment system leakage test criteria.

2. Editorial. For increased clarity, an expanded and revised Table of Contents and set of definitions has been provided, conforming to current usage. The text has also been revised to conform to " plain English" objectives.

• 3. Interpretations. Some changes have been made to resolve past questions of interpretation (e.g., definitions of " containment isolation valves").

4. Greater flexibility. A major problem with Appendix J has been the lack of a provision for dealing with plants already built where design features are incompatible with Appendix J requirements (e.g. , air lock testing). As a result, provision has been made in this revision for consideration by the NRC staff of alternative leakage test requirements when necessary.

5. Type A test pressure. The option of performing periodic reduced pressure testing in lieu of testing at full calculated accident pressure has been dropped. This change reflects the opinion that extrapolating low pressure leakage test results to full pressure leakage Reasonable argument can test results has turned out to be unsuccessful.

be made for low pressure testing. However, the NRC staff believes that the peak calculated accident pressure (a) has always been the intended reference test pressure, (b) is consistent with the typical'. practice for NRC staff evaluations of accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 9

[7590-013 accordance with Regulatory Guide 1.3 and 1.4, (c) provides at least a . .

nominal check for gross low pressure leak paths that a low pressure leak does not provide for high pressure leak paths, (d) directly represents technical specification leakage rate limits, and (e) provides' greater confidence in containment system leaktight integrity. For these reasons, the full, rather than reduced, pressure has been retained as the test pressure.

6. Type A test frequency. The test frequency has been uncoupled from the 10 year inservice inspection period used by the ASME Boiler &

• Pressure Vessel Code for mechanical systems. A different time base is used, but the frequency has remained essentially the same.

7. Type A test duration. The duration has been dropped from the test criteria in Appendix J. It is considered as part of the testing procedures, and is a function of the state of the testing technology and the level of confidence in it.

8. Type A test "as is" clarification. Appendix J originally noted in III. A.1(a) that the containment was to be "... tested in as close to the 'as is' condition as practical." This is re-emphasized and clarified by the explicit requirements that have been added to measure, record, and report "as found" and "as left" leakage rates.

Type A test allowable leakage rate prorating. Seventy-five 9.

percent of the allowable leakage rate represents the "as left" Type A test acceptance criterion, leaving 0.25 of the allowable leakage rate as a margin for deterioration until the time of the next regulatory scheduled Type A test, when the "as-found"-leakage rate criterion is 1.0 of the allowable leakage rate. .

10. Quantification of allowable leakage rates. It should be noted that ILo change has been made to the way in which the allowable test 10

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leakage rates are quantified. The_ regulation still refers to the individual plant technical specifications for these values. Debate continues, however, on what these values should be and whether they can be generically specified, rather than individually specified for each site and plant.

11. Refocusing of corrective actions. When a reportable problem is identified, a Corrective Action Plan is to be submitted. It identifies the problem to the NRC staff, and notes the cause, what was or will be done to correct it, and what will be done to prevent its recurrence.

Increased local leakage testing frequency may be necessary. Appendix J originally addressed increased test frequency only for Type A tests. This revision applies adjustment of test frequency directly to identified problem areas.

12. The final paragraph of the proposed amendment specifies a date by which an implementation schedule must be submitted, rather than by which it must be implemented. This is because the ease with which licensees will be able to implement all the provisions of the amendment will be highly plant specific depending on plant design, outage and testing schedules, and amount of technical specification changes needed.

FINDING 0F N0 SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY The Commission has detemined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of;the human environment and therefore an environmental impact statement is'not 11

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required. There will be no radiological environmental impact offsite, but there may be an occupational radiation exposure onsite of about 3.0 man-rem per year of plant operation for inspection personnel (about 0.4%

increase). Alternatives to issuing this revision were considered and found not acceptable. The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the environmental assessment and the finding of no significant impact are available from Mr. E. Gunter Arndt, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555, Telephone (301) 443-7893.

PAPERWORK REDUCTION ACT STATEMENT This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980(44USC3501etseq.).

This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

REGULATORY ANALYSIS The Commission has prepared a' draft regulatory analysis on the proposed revision. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft analysis is available for inspection and copying in the NRC Public Document Room, 1717 H Street, NW, Washington, DC. The Commission requests public comment on the draft analysis. Comments may be submitted to the NRC as indicated under the Addresses heading.

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. , BACKFIT ANALYSIS The Comission has prepared a backfit analysis on the proposed revi-

- sion. The analysis is required .under 10 CFR Part 50, Section 50.109, as of October 21, 1985, for:the management of backfitting for power reacters.

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The analysis is available for inspection and' copying in the NRC Public s ,

} Document Room,1.717 H Street NW, Washington, DC. The Commission requests public comment on the analysis. Coments may be submitted to the NRC as j, -

indicated under the Addresses heading.

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3 ld Lh The analysis does not conclude that there is a substantial increase in the overall protection of the public health and safety or the comon defense and security to be derived from the backfit. It does conclude, however, that the direct and indirect costs of implementation are justified due to better, more. uniform tests and test reports, greater confidence in the l reliability of the test results,' fewer exemption requests, and fewer I

• interpretive debates. For these easons, which are presented in greater detail in the backfit analysis, the Comission has granted an exemption l

i frodtherequirementforadeterminationpursuanttoS50.109(a)(3).

This section requires a determination that the rule will result.in a sub-stantial increase in.the overall protection of the public health and safety i  %

or the comon defense and security to be derived from the backfit.

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REGULATORY FLEXIBILITY CERTIFICATION 1* - In'accordance with the Regulatory Flexibility Act of 1980,

' (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if s r i promulgated, have a significant economic impact on a substantial number 13

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[7590-01] , , .g of small entities. This proposed rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the-scope of the definition of "small entities" set ~

for,th in. the Regulatory Flexibility Act or the Small Business. Size Stand-ards set out in regulations issued by the Small Business Administration at' 13 CFR Part 121.

LIST OF SUBJECTS IN 10 CFR PART 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection,- Reactor' siting criteria, Reporting and recordkeeping requirements.

RELATED REGULATORY GUIDE The notice of availability of a draft regulatory guide on the same subject " Containment System Leakage Testing" (MS 021-5) is also being published in the notice section of this Federal Register. The draft regulatory guide contains specific guidance on acceptable leakage test methods, procedures, and analyses that may be used to implement these requirements and criteria.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the_ Energy _ Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing;to adopt the  !

l following amendments to 10 CFR Part 50. .

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- - PART 50 -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for Part 50 continues to read as follows:

AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.'936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1246, as amended (42 U.S.C. 5841, 5842, 5846),

unless otherwise noted.

Section 50.7 also issued under Pub. L.95-601, sec. 10.92 Stat. 2951

. (42 U.S.C. 5851). Sections 50.57(d), 50.58, 50.91, and 50.92 also issued under Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2133, 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). Sections 50.100-50.102 also issued under sec. 186, 68 Stat. 955 (42 U.S.C. 2236).

For the purposes cI se . 223, 68 Stat. 958, as amended (42 U.S.C. 2273); $$ 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54, 1

and 50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); $$ 50.10(b) and (c) and 50.54 are issued under sec.

j 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(i)); and $$ 50.55(e),

l 50.59(b), 50.70, 50.71, 50.72, 50.73, and 50.78 are issued under sec.

161o, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).

2. Appendix J is revised to read as follows:

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Leakage Tests for Containments of , ,

' Light-Water-Cooled Nuclear Power Plants

• Table of Contents I. INTRODUCTION II. DEFINITIONS III. GENERAL LEAK TEST REQUIREMENTS A. Type A Test

• 1. Preoperational Test

2. Periodic Test . . . . . . - .

3. Test Frequency

4. Test Start and Finish

5. Test Pressure

6. Pretest Requirements

7. Verification Test

8. Acceptance Criteria f

9. Retesting

10. Permissible Periods- for Testing

' 8. Type B Test

- 1. Frequency j

i

2. Pressure l

l 3. Air Locks

4. Acceptance Criteria

[

C. Type C Test

1. Frequency i .

2. Pressure / Medium

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[7590-01]

3. Acceptance Criteria

4. Valves That Need Not Be Type C Tested IV. SPECIAL LEAK TEST REQUIREMENTS A. Containment Modification or Maintenance B. f-jultiple Leakage Barriers or Subatmospheric Containments V. TEST METHOD, PROCEDURES, AND ANALYSES A. Type A, B, and C Test Details

- B. Combination of Periodic Type A, B, and C Tests VI. REPORTS A. Submittal B. Content VII. APPLICATION A. Applicability B. Effective Date 17

' [7590-01) .

3

- . c, I. INTRODUCTION i .

~

One of the conditions of all operating licenses for light-water-cooled power reactors as specified in 5 50.54(o) of this part is that primary containments meet the leak test requirements set forth in this Appendix. ' The tests ensure that (a) leakage through the primary contain-ments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the Technical Specifications and (b) inservice inspection of penetrations and isolation valves is per-formed so that proper maintenance and repairs are made during their service

. life. This Appendix identifies the general re'quirements and acceptance criteria for preoperational and subsequent periodic leak testing.2 II. DEFINITIONS ACCEPTANCE CRITERIA Standards against which test results are to be compared for

.s establishing the functional acceptability of the containment system as a leakage limiting boundary.

"AS FOUND" LEAKAGE RATE The leakage rate prior to any needed repairs or adjustments to.the leakage barrier being tested.

"AS LEFT" LEAKAGE RATE The leakage rate following any needed repairs or adjustments to the 1eakage barrier being tested.

l CONTAINMENT INTEGRATED LEAK RATE TEST (CILRT)

The combination of a Type A test and its verification test.

T..

1 Specific guidance concerning acceptable leakage test methods, procedures, and analyses that may be used to implement these requirements and criteria will be provided in a regulatory guide that is being issued in draft form for public comment with the designation MS 021-5. Copies of the regulatory guide may be obtained from the Nuclear Regulatory Commission, Document Management Branch, Washington, DC 20555.

18

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.i '. [7590-01]

. . c, CONTAINMENT ISOLATION SYSTEM FUNCTIONAL TEST A test to verify the proper performance of the isolation system by normal operation of the valves. For automatic containment isolation sys,tems,.a test of the automatic isolation system performed by actuation of the containment isolation signals.

C0'NTAINMENT ISOLATION VALVE Any valve defined in General Design Criteria 55, 56, or 57 of Appen-dix A " General Design Criteria for Nuclear Power Plants," to this part.

CONTAINMENT LEAK TEST PROGRAM

, _The comprehensive testing of the containment system that includes Type A, B, and C tests.

CONTAINMENT SYSTEM The principal barrier, after the reactor coolant pressure boundary,,

to prevent the release of quantities of radioactive material that would have a significant radiological effect on the health of the public. It includes:

(1) the primary containment, including access openings and penetrations.

(2) containment isolation valves, pipes, closed systems, and other components used to effect isolation of the containment atmosphere from the outside environs, and

! (3) those systems or portions of systems that by their functions i extend the primary containment boundary to include their system boundary.

This definition does not include boiling water reactors' (BWR) reactor buildings or pressurized water reactors' (PWR) shield buildings. Also excluded from the provisions of this Appendix are the interior barriers

[ 19 l

[7590-01] .. .,

such as the BWR Mark II drywell floor and the drywell perimeters of the BWR Mark 111 and the PWR' ice condenser.

i

-L,(WEIGHT PERCENT /24 HR)

. The maximum allowable Type A test leakage rate in units of weight percent per 24-hour period at pressure P ac as specified in the Technical Specifications.

L,,(WEIGHT PERCENT /24 HR)

The measured Type A test leakage rate in units of weight percent per 24-hour period at pressure Pac, obtained from testing the containment system in the state as close as practical to that that would exist under design basis accident conditions (e.g., vented, drained, flooded, or pressurized).

LEAK 4

I An opening that allows the passage of a fluid.

LEAKAGE The quantity of fluid escaping from a leak.

LEAKAGE RATE The rate at which the contained fluid escapes from the test volume at a specified test pressure.

MAXIMUM PATHWAY LEAKAGE RATE The maximum leakage rate that can be attributed to a penetration l

leakage path (e.g.. the larger, not total, leakage of two valves in series). This generally assumes a single active failure of the better of two leakage barriers in series when performing Type B or C tests.

MINIMUM PATHWAY LEAKAGE RATE 4

The minimum leakage rate that can be attributed to a pqnetration leakage path (e.g. , the smallest leakage of two valves in series). This 20

' ' [7590-01]

is used when correcting the measured value of containment leakage rate from the Type A test (L,,) to obtain the overall integrated leakage rate and generally assumes no single active failure of redundant leakage barriers under these test conditions.

OVERALL INTEGRATED LEAKAGE RATE The total leakage rate through all leakage paths, including contain-ment welds, valves, fittings, and components that penetrate the contain-ment system, expressed in units of weight percent of contained air mass at test pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Pac (P5ig)

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident as specified in the technical specifications.

PERIODIC LEAK-TEST Test conducted during plant operating lifetime.

PREOPERATIONAL LEAK TEST Test conducted upon completion of construction of a primary or secondary containment, including installation of mechanical, fluid, electrical, and instrumentation systems penetrating these containment systems, and prior to the time containment integrity is required by the Technical Specifications.

PRIMARY CONTAINMENT The structure or vessel that encloses the major components of the reactor coolant pressure boundary as defined in S 50.2(v) of this part an'd is designed to contain accident pressure and serve as a leakage barrier against the uncontrolled release of radioactivity to the environ-ment. The term " containment" as used in this Appendix refers to the

' primary containment structure and associated leakage barriers. "

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STRUCTURAL INTEGRITY TEST . .

A pneumatic test that demonstrates the capability of a primary containment to withstand a specified internal design pressure load.

TYPE A TEST A test to measure the containment system overall integrated leakage rate under conditions representing design basis loss-of-coolant accident containment pressure and systems alignments (1) after the containment system has been completed and is ready for operation and (2) at periodic intervals thereafter. The verification test is not part of this definition - see CILRT.

TYPE B TEST

. A pneumatic test to detect and measure local leakage through the following containment penetrations:

(1) Those whose design incorporates resilient seals, gaskets, sealant compounds, expansion bellows, or fitted with flexible metal seal assemblies.

(2) Air locks, including door seals and door operating mechanism penetrations that are part of the containment pressure boundary.

TYPE C TEST A pneumatic test to measure containment isolation valve leakage rates.

VERIFICATION TEST Test to confirm the capability of the Type A test method and equip-1 l

1 l ment to measure L,.

~ ~

III? GENERAL-LEAK TEST ' REQUIREMENTS A. Type A Test (1) Preoperational Test. A preoperational Type A test must be ,

conducted on the containment system and must be preceded by:

22 I.

[7590-01]

(a) Type B and Type C tests, (b) A structural integrity test.

(2) Periodic Test. A periodic Type A test must be performed on the con.tainment system.

(3) Test Frequency. Unless a longer interval is specifically approved by the NRC staff, the interval between the preoperational and first periodic Type A tests must not exceed three years, and the interval between subsequent periodic Type A tests must not exceed four years. If the initial fuel loading is delayed so that the three year interval o between the first preoperational test and the first periodic test is exceeded, another preoperational test will be necessary. If such an addi-tional preoperational Type A test or an additional Type A test required by Sections III.A.8 or IV.A. of this Appendix is performed, the Type A test interval may be restarted.

(4) Test Pressure. The Type A test pressure must be equal to or greater than P ac at the start of the test but must not exceed the contain-ment design pressure and must not fall more than 1 psi below P ac f r the duration of the test, not including the verification test. The tett pressure must be established relative to the external pressure of the containment. This may be either atmospheric pressure or the subatmospheric pressure of a secondary containment.

(5) Pretest Requirements. Closure of containment isolation valves for the Type A test must be accomplished by normal operation and without any preliminary exercising or adjustments for the purpose of improving .

performance (e.g., no tightening of valve after closure by valve motor).

Repairs of malfunctioning or leaking valves must be made as.necessary.

Information on valve leakage that requires corrective action prior to, 23

[7590-01] . . ,

during, or after the test (see Section V.B.) must be included in the . ,

report submitted to the Commission as specified in Section VI of this Appendix.

- (6)' Verification Test. A leakage rate verification test must be performed after a Type A test in which the leakage rate meets the criteria in III. A.(7)(b)(ii). The verification test selected must be conducted for a duration sufficient to establish accurately the change in leakage rate between the Type A and verification tests. The results of the Type A test are acceptable if the sum of the verification test imposed

• leakage and the containment leakage rate calculated from the Type A test (L,,) does not differ from the leakage rate calculated from the verifica-

~

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tion test by more than 10.25 L,.

(7) Acceptance Criteria.

(a) For the preoperational Type A Test, the "as left" leakage rate must not exceed 0.75L,, as determined by a properly justified statistical analysis. The "as found" leakage rate does not apply to the preoperational test.

(b) For each periodic Type A test, the leakage rate, as deter-mined by a properly justified statistical analysis, must not exceed:

(i) L,, for the "as found" condition, (ii) 0.75L,, for the "as left" condition, (c) In meeting these Type A test acceptance criteria, isola-tion, repair, or adjustment to a leakage barrier that may affect the leakage rate through that barrier is permitted prior to or during the Type A test-provided:

~ -

(i) all potential leakage paths of the isolated, repaired, or adjusted leakage barrier are locally leak testable, and 24

[7590-01]

(ii) the local leakage rates are measured before and after the isolation, repair, or adjustment and are reported under Section VI of this Appendix.

(iii) All changes in leakage rates resulting from isola-tion, repair, or adjustment of leakage barriers subject to Type B or Type C testing are determined using the minimum pathway leakage method and added to the Type A test result to obtain the "as found" and "as left" containment leakage rates.

(d) The effects of isolation, repair, or adjustments to the

- containment boundary made after the start of the Type A test sequence on the Type A test results must be quantified and the appropriate analytical corrections made (this includes tightening valve stem packing, additional tightening of manual valves, or any action taken that will affect the leakage rates).

(8) Retesting.

(a) If, for any periodic Type A test, the as found leakage rate fails to meet the acceotance criterion of 1.0L,, a Corrective Action Plan that focuses attention on the cause of the problem must be developed and implemented by the licensee and then submitted together with the Containment Leak Test Report as required by Section VI of this Appendix.

The test schedule applicable to subsequent Type A tests (III. A.(3)) shall be submitted to the NRC staff for. review and approval. An as left Type A test that meets the acceptance criterion of 0.75L, is required prior to l plant startup.

i (b) If two consecutive periodic as found Type A tests exceed l

the as found acceptance criterion of 1.0L,: .

25 t

• [7590-01). -

(i) Regardless of the periodic retest schedule of.III.A.(3),

a Type A test must be performed at least every 24 months (based on the refueling cycle normally being about 18 months) unless an alternative lea,kage test program is acceptable to the NRC staff on some other defined basis. This testing must be performed until two consecutive periodic "as found" Type A tests. meet the acceptance criterion of 1.0L,after which the retest schedule specified in III.A.(3) may be resumed.

(ii) Investigation as to the cause and nature of the 1

Type A test failure might indicate that an alternative leakage test program such as more frequent Type B or Type C testing may be more appro-priate than the performance of two consecutive successful Type A leakage tests. The licensee may then submit a' Corrective Action Plan and an If this i

alternative leakage test program proposal for NRC staff review.

submittal is approved by the NRC staff, the licensee may implement the corrective action and alternative leakage test program in lieu of one or both of the Type A leakage tests required by Section III. A.(8)(b)(i).

(9) Permissible periods for testing. The performance of Type A tests must be limited to periods when the plant facility is secured in the shutdown condition under the administrative controls and safety procedures defined in the license.

B. Type B Test (1) Frequency.

(a) Type B tests, except tests for air locks, must be performed on containment penetrations during. shutdown _for_ref_ueli_ng or at other convenient intervals but in no case at intervals greater than

26

__ ~ - -- . - .. -_- __ .- . . . . _ _ _ - - - - _ _ - _ _ - _

[7590-01]

2 years. If opened following a Type A or B test, containment penetra-tions subject to Type B testing must be Type B tested prior to returning the reactor to an operating mode requiring containment integrity.

~

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(b) For containment penetrations empicying a continuous le,akage monitoring system that is at a pressure not less than P ac' leakage readings of sufficient sensitivity to permit comparison with Type B test leak rates must be taken at intervals specified in the Tech-nical Specifications. These leakage readings must be part of the Type B reporting of VI.A. When practical, continuous leakage monitoring systems must not be operating or pressurized during Type A tests. If the contin-uous leakage monitoring system cannot be isolated, such as inflatable air lock door seals, leskage into the containment must be accounted for and the Type A test results corrected accordingly.

(2) Pressure. Type B tests must be conducted, whether individually or in groups, at a pneumatic pressure not less than Pac eXCept as pro-vided in paragraph III.B.(3)(b) of this section or in the Technical Specifications.

(3) Air Locks.

l (a) Initial and periodic tests. Air locks must be tested prior to initial fuel loading and at least once each 6-month interval thereafter at an internal pressure not less than Pac. Alternatively, if I

there have been no air lock openings within 6 months of the last l

successful test at Pac, this interval may be extended to the next

! refueling outage or airlock opening (but in no case may the interval exceed 2 years). Reduced pressure tests must continue to be performed on

{ the air lock or its door seals at 6-month intervals. Opening of the air lock for the purpose of removing air lock testing equipment following an l

air lock test does not require further testing of the air lock.

27 i

[7590-01]

(b) ' Intermediate tests must be conducted as follows: , ,

(i) Air locks opened during periods when containment integrity is required by the plant's-Technical Specifications must be tested w'ithin 3 days after being opened. For air lock doors opened more frequently than once every 3 days, the air lock must be tested at least 4

once every 3 days during the period of frequent openings. Air locks opened during periods when containment integrity is not required by the I

/

plant's Technical Specifications need not be repeatedly tested during i such periods. However, they must be tested prior to the plant requiring

• containment integrity. For air lock doors having testable seals, testing the seals fulfills the intermediate test requirecents of this paragraph. __

In the event that this intermediate testing cannot be done at Pac, the test pressure must be as stated in the Technical Specifications.

(ii) Whenever maintenance other than on door seals has been performed'on an air lock, a complete air lock test at a test pressure of not less than P,c is required, if that maintenance involved the pressure retaining boundary.

(iii) Air lock door seal testing or reduced pressure testing may not be substituted for the initial or periodic full pressure test of the entire air lock required in paragraph III.B.(3)(a) of this Section. ,

(4) Acceptance Criteria.

(a) The sum of the as found or as left Type B and C test results must not exceed 0.60L,using maximum pathway leakage and including leakage rate readings from continuous leakage monitoring systems.

(b) Leakage measurements are ' acceptable ~if~0btained through ~

2 component leakage surveillance systems (e.g., continuous pressurization

+

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[7590-01]

of individual or clustered containment components) that maintain a pres-sure not less than P,c at individual test chambers of those same contain-ment penetrations during normal reactor operation. Similar penetrations not, included in the component leakage surveillance system are 'still subject to individual Type B tests.

(c) An air lock, penetration, or set of penetrations that fails to pass a Type B test must be ratested following determination of cause and completion of corrective action. Corrective action to corruct the leak and to prevent its future recurrence must be developed and implemented.

(d) Individual acceptance criteria for all air lock tests must be stated in the Technical Specifications.

C. Type C Test (1) Frequency. Type C tests must be performed on containment isola-tion valves during each reactor shutdown for refueling or at other convenient intervals but in no case at intervals greater than 2 years.

(2) Pressure / Medium.

(a) Containment isolation valves unless pressurized with a qualified water seal system must be pressurized with air or nitrogen at a pressure not less than Pac *

(b) Containment isolation valves, that are sealed with water from a qualified seal system, must be tested with water at a pressure not less than 1.10 Pac*

(3) Acceptance Criteria.

(a) The sum of the as found or as left Type B and C test results must not exceed 0.60L, using maximum pathway leakage and including leakage rate readings from continuous leakage monitoring systems.

29

[7590-01] ,

~ '

> c, (b) Leakage from containment isolation valves that are sealed . .

with water from a seal system may be excluded when determining the

- combined Type B and C leakage rate if:

(i) The valves have been demonstrated to have leakage rates that do not exceed those specified in the Technical Specifications, and (ii) The installed isolation valve seal system inventory is sufficient to ensure the sealing function for at least 30 days at a pressure of 1.10 Pac *

(4) Valves That Need Not Be Type C Tested.

(a) A containment isolation valve need not be Type C tested if 1

it can be shown that the valve does not constitute a potential contain-ment atmosphere leak path during or following an accident, considering a single active failure of a system component.

(b) Other valves may be excluded from Type C testing only when approved by the NRC staff under the provisions of paragraph VII.A. ~

(. IV. SPECIAL LEAK TEST REQUIREMENTS A. Containment Modification or Maintenance Any modification, repair, or replacement of a component that is part of the containment system boundary and that may affect containment integrity must be followed by either a Type A, Type B, or Type C test.

Any modification, repair, or replacement of a component subject to Type B The or Type C testing must also be preceded by a Type B or Type C test.

measured leakage from this test must be included 15 the report-to the -

Commission required by Section VI of this Appendix. Following structural changes or repairs that affect the pressure boundary, the licensee shall l

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[7590-01]

• C2

' - demonstrate whether or not a structural integrity test is needed prior to the next Type A test. The acceptance criteria of paragraphs III.A.(7),

III.B.(4), or III.C.(3) of this Appendix, as appropriate, must be met.

Typs A testing of certain minor modifications, repairs, or replacements may be deferred to the next regularly scheduled Type A test if local leakage testing is not possible and visual (leakage) examinations or non-destructive examinations have been conducted. These shall include:

Welds of attachments to the surface of the steel pressure retaining boundary; Repair cavities the depth of which does not penetrate the required design steel wall by more than 10%; Welds attaching to the steel pressure retaining boundary penetrations the nominal diameter of which does not exceed one inch.

B. Multiple Leakage Barrier or Subatmospheric Containments i The primary reactor containment barrier of a multiple barrier or subatmospheric containment shall be subjected to Type A tests to verify l

l that its leakage rate meets the requirements of this appendix. Other structures of multiple barrier or subatmospheric containments (e.g. ,

secondary containments for boiling water reactors and shield buildings for pressurized water reactors that enclose the entire primary reactor containment or portions thereof) shall be subject to individual tests in accordance with the procedures specified in the technical specifications.

V. TEST METHODS, PROCEDURES, AND ANALYSES A. Type A, 8 and C Test Details Leak test methods, procedures, and analyses for a steel, concrete, or combination steel and concrete containment and its penetrations and 31

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[7590-01] .

. isolation valves for light-water-cooled power reactors must be referenced or defined in the Technical Specifications.

B. , Combination of Periodic Type A. B and C Tests Type B and C tests are considered to be conducted in conjunction with the periodic Type A test when performed during the same outage as the Type A test. The licensee shall perform, record, interpret, and report

' the tests in such a manner that the containment system leak-tight status ,

is determined on both an as found basis and an as left basis, i.e., its

leak status prior to this periodic Type A test together with the related Type B and C tests and its status following the conclusion of these tests.- - - - - . . .

l- VI. REPORTS A. Submittal

1. The preoperational and periodic Type A tests, including sum-maries of the results of Type B and C tests conducted in conjunction with the Type A test, must be reported in a summary technical report sent not later than 3 months after the conduct of each test to the Commission in I

the manner specified in 5 50.4. The report is to be titled " Containment Leakage Test."

2. Reports of periodic Type B and C tests conducted at intervals intermediate to the Type A tests must also be submitted to the NRC as in Reports paragraph VI.A.1 at the time of the next Type A test submittal.

4 must be submitted to the NRC Regional Administrator within 30 days of completion of any Type B or C tests that fail to meet their as found _ _ _ _

acceptance criteria. '.

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[7590-01]

B. Content A Type A test Corrective Action Plan, when required under paragraph III.A.(8) of this Appendix, must be included in the report. Any correc-tive act' ion required for those Type B and C tests included as a part of the Type A test sequence must also be included in the report.

VII. APPLICATION A. Applicability The requirements of this Appendix apply to all operating nuclear

- power reactor licensees as specified in SS 50.54(o) of this part unless it can be demonstrated that alternative leak test requirements (e.g. , for certain containment designs, leakage mitigation systems, or different test pressures not specifically addressed in this Appendix) are' demon-strated to be adequate on some other defined basis. Alternative leak test requirements and the bases for them will be made a part of the plant Technical Specifications if approved by the NRC staff.

B. Effective Date f This Appendix is effective (30 days after publication). By (insert a date 180 days after the effective date of this revision), each licensee and each applicant for an operating license shall submit a plan

(

to the Director of the Office of Nuclear Reactor Regulation for imple-menting this Appendix. This submittal must include an implementation schedule, with a final implementation no later than (insert a date 48 months after the effective date of this revision). Until the licensee finally implements the provisions of this revision, the licensee shall continue to use in their entirety the existing Technical Specifications 33 l

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[7590-01] .

and the Appendix J on which they are based. Thereafter, the licensee . .

shall use in their entirety this revision and the Technical Specifica-tions conforming to this revision.

Dated at Washington, DC, this day of , 1985.

- For the Nuclear Regulatory Commission.

Samuel J. Chilk Secretary of the Commission C

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a Draft E9 03/27/86 Federal Register Notice Margin marked to show changes due to:

ACRS & CRGR comments 7/11/85 Regional & Office comments 8/30/85 i

NRR, DRA, ELD 10/15/85 0CM paper 12/24/85 OCM(backfitanalysis),

RES(editorial),and DRR (10 CFR 50.4 revision) 2/25/86 RES concurrence connents 3/13/86 ELDexemptionto50.109(a)(3) 3/25/86 RES concurrence editorial 3/27/86 l

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[7590-013_ .

NUCLEAR REGULATORY COMMISSION , ,

10 CFR Part 50 General Revisien of Appendix J ,

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule. l 4

SUM 4ARY: The Nuclear Regulatory Commission is proposing to amend its i

regulations to update the criteria and clarify questions of interpretation in regard to leakage rate testing of containments of light-water-cooled '

N nuclear power plants. The proposed rule would aid the licensing and en- g forcement staff by eliminating conflicts, ambiguities, and a lack of uni- f formity in the regulation of the inservice inspection program.

DATE: Comment period expires . Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except for comments received on or before this date.

ADDRESSES:

Written comments may be submitted to the Rules and Procedures Branch, Division of Rules and Records, Office of Administration, U.S. '

Nuclear Regulatory Commission, Washington, DC 20555. Comments may also D

1 a

l be delivered to Room 4000, Maryland National Bank Building, Bethesda, }

Copies of Maryland from 8:15 a.m. to 5:00 p.m. Monday through Friday.

comments received may be examined at the NRC Public Document Room, 1717 H Street NW., Washington, DC.

1

.o--

3 ,-,. m_m.,,,... ---.w--,s , - - - - . , - . .-_,,..e .,e - -- ,.,.--.----n -,---s

~ [7590-01)

Copies of draf t regulatory guide MS 021-5 may be obtained from the Nuclear Regulatory Commission, Document Management Branch, Washington, DC 20555.

FOR FURTHER INFORMATION CONTACT: Mr. E. Gunter Arndt, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 443-7893.

BACKGROUND SUPPLEMENTARY INFORMATION:

Appendix J of 10 CFR Part 50 was originally issued for public comment as a proposed rule on August 27,1971 (36 FR 17053); published in final form on February 14, 1973 (38 FR 4385); and became effective on March 16, 1973. The only amendment to this Appendix since 1973 was a limited one, on Type B (penetration) test requirements that was published for comment on January 11, 1980 (45 FR 2330); published in final form September 22, 1980 (45 FR 62789); and became effective on October 22, 1980.

This revision of Appendix J has been in preparation for some time, b Y

It will provide greater flexibility in applying alternative requirements S

due to variations in plant design and reflects changes based on:

(1) experience in applying the existing requirements; (2) advances in containment leak testing methods; (3) interpretive questions; (4) simplifying the text (5) various external / internal comments since 1973; and (6) exemption requests received and approved. ,

2

[7590-01] .

This proposed revision is for the purpose of updating the existing u

-Q regulation. Other related, longer term, and broader issues are currently h under review by the NRC staff, such as containment function, degree of integrity required, and validation of that integrity under conditions other than postulated in this rule. In order to better understand its function and scope, assumptions inherent in Appendix J are presented as b is follow:

1. Certain levels of radiation exposure at the plant site boundary shall not be exceeded under (a) operating or (b) design basis accident {

}

conditions.

2. Certain levels of radiation exposure to plant operating

-t personnel shall not be exceeded under (a) operating or (b) design basis accident conditions.

3. All four exposure levels (la, ib, 2a, 2b) may be different, but can be calculated.

4. Defense-in-depth will be used for protection against these levels of exposures. As the final barrier, a containment system is re- E quired in order to maintain any or all of these exposure limits. '

5. The required degree of containment system leaktightness for design basis accidents can be (a) calculated, (b) specified, (c) built, p

(d) maintained, (e) inspected.

6. A generic inspection program can be defined that verifies the required leaktightness of the containment following construction and periodically throughout plant life.

7. NRC regulations should require such an inspection program, and define the test requirements and acceptance criteria.

3

• [7590-01]

8. A standard loss-of-coolant accident is assumed as the Since the containment isolation system is an design basis accident.

engineered safety feature, only safety grade systems and components are relied upon to define the containment boundary that must be exposed to the containment pneumatic test pressure for the In addition, all safety grade systems integrated leak rate test.

are assumed to be subject to a potential single active failure, and must t,e locally leak rate tested accordingly.

9. Pneumatic testing to peak calculated accident pressure is adequate without testing for, or at, accident temperatures or radiation levels.

10. Shielding tests need not be performed.

Periodic testing provides adequate confidence in the level' y 11.

k of containment system integrity. Continuous monitoring of all D individual isolation barriers is not necessary.

The scope of this revision to Appendix J is limited to However, y corrections and clarifications, and excludes new criteria.

N I

this notice also addresses related, broader, longer term activities. kk j Following is information of some of these other related activities that are not reflected in this proposed rulemaking.

In order to better identify the availability of g containment leakage integrity, concepts of " continuous containment leakage monitoring" (such as negative containment l operating pressure) and "relatively frequent gross continment l

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integrity check" (such as a low pressure pumpup just prior to 4

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[7590-01) .. .,

operation to check for openings) are under consideration by the NRC staff. . .

These would identify large breaches of the containment system boundary, during, or just prior to, normal operating conditions. It should be noted ,

they would only test the normal operating containment atmosphere boundary, not the Appendix J, post-accident boundary including isolation valves.

. Comments on these or alternative concepts, and what effect, if any, they would have on the proposed Appendix J requirements, are also being solicited g N 1 in the following section of this preamble.  %

Past practice has been to implement the provisions of Appendix J by i

means of licensees' technical specifications. Currently, a Technical i

Specification Improvement Project (TSIP) is underway to reevaluate the NRC's philosophy and utilization of the technical specifications. While the proposed revision described herein assumes. implementation of i

-Appendix J by licensee's technical specifications, the work of the TSIP i may lead to some changes in this form of implementation.

a Another program is presently being conducted to identify current NRC regulatory requirements that have marginal importance to safety and to recommend appropriate actions to modify or to eliminate these unneces-sary requirements. A Federal Register notice was published on October 3, 2 N

1984, to announce the initiation of the program (49 FR 39066). As a part d of the program, regulatory requirt :nts associated with containment leak- ,

tightness are being evaluated. The risk and cost effectiveness of contain-1 ment leaktightness requirements will be examined to determine their value with respect to plant safety and possible alternative requirements.

Any resulting changes to existing regulations will be made through ,

normal rulemaking procedures, including ACRS review and public comment.  %

Comments on the questions posed in this notice will also provide early, {  ;

useful input to these associated activities. h 5 ,

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• P 8 4

,e >

' ' [7590-01)

~h -

s 4, . /

INVITATION TO COMMENT olh

? {l l .

'

• Comments' from all interested persons on all aspects of this revision A

'i 4 ' land on the risk and cost effectiveness of containment leaktightness in < "

1)

J%

general are reques'ted by the coment expiration date in order that: '

- 1 1 the final revision will reflect consideration of all points of view, and 2),the staff's assessment of the risk importance of containment if leaktightness' can benifit from such comments. Especially requested are i

comments which addritss the following questions:

(1) the extent to which these positions in the proposed rule are f already 5

>t in use; (2) the extent to which those in use, and those not in use but proposed, are desi.rable;

, y (3) whether there continues to be a further need for this regulation; (4) estimates of the costs and benefits of this proposed revision, as a i whole and of its'shparate provisions; (5) if the existing fule or its proposed revision were completely

• volontary, how many licensees would adopt either version in its sa entirety and why;J i~

J)

(6) Since the NRC is planning a broader, more comprehensive review of containment functional and testing requirements in the next year or two, whether it is then still worthwhile to go forward with this  %

proposed revision as an interim updating of the existing regulation; k

' (7) whether the technical specidication limits on allowable ' conta s leakage should be relaxed and if so,' to what extent and why, or if not, why not; s

(8) what risk-important factors influence containment performance under 2h V yh severe accident conditions, to what degree these factors,are k

considered in the current containment testing requirements, and what9.

! approaches should be considered in addressing factors not presentl Mg covered; D

$g)

,lb 6 e, 4

~ ~ ' " - . - - , - __

"i

.[7590-01]

' . n INVITATION TO Com ENT , ,

Comments from all interested persons on all aspects of this revision are requested by the comment expiration date in order that the final revision will reflect consideration of all points of view. Especially requested are comments which address the following questions: '

y (1) the extent to which these positions in the proposed rule are already D,.

in use; (2) the extent to which those in use, and those not in use but proposed, are desirable; (3) whether there continues to be a further need for this regulation; (4) estimates of the costs and benefits of this proposed revision, as a w whole and of its separate provisions; $

(5) whether the technical specification limits on allowable containment h leakage should be relaxed and if so, to what extent and why, or if not, why not;

-(6) what the effect would be on the leakage rate test program if the source ,%'

term used were to be based on NUREG 09562 methods rather than TID 148442; b'

i

' (7) if the existing rule or its proposed revision were completely voluntary,

l. how many licensees would adopt either version in its entirety and why; L '

(8) since the NRC is planning a broader, more comprehensive review of con-tainment functional and testing requirements in the next few years, whether it is then still worthwhile to go forward with this proposed revision as an interim updating of the existing regulation; I Copies of NUREG 0956, Reassessment of the Technical Bases for Estimating h 1

Source Terms, and Technical Information Document 14844, Calculation of gy Distance Factors for Power and Test Reactor Sites, may be'obtained from the Commission's Public Document Room,1717 H Street NW, Washington, DC p 20555.

6

[7590-01)

' . n (9) what other approaches to validating containment integrity could be used that might provide detection of leakage paths as soon as they occur, whether they would result in any adjustments to the Appendix J test program and why; (10) what effect " leak-before-break" assumptions could have on the leak-age rate test program. Current accident assumptions use instant aneous complete breaks in piping systems, resulting in a test "Le ak-program based on pneumatic testing of vented, drained lines.

before-break" assumptions presume that pipes will fail more gradually, leaking rather than instantly emptying.

(11) how to effectively adjust Type A test results to reflect individual Type B and C test results obtained from inspections, repairs, adjust-ments, or replacements of penetrations and valves in the years in between Type A tests. Such an additional criterion, currently outside the scope of this proposed revision, would provide a more meaningful tracking of overall containment leaktightness on a more continuous basis than once every several years. The only existing or proposed 4

criterion for Type B and C tests performed outside the outage in QC

,N which a Type A test,is performed is that the sum of Type B and C tests must not exceed 60% of the allowable containment leakage. Cur-rently being discussed by the NRC staff are:

a. All Type B and C tests performed during the same outage as a Type A test, or performed during a specified time period (nomi-nally 12 months) prior to a Type A test, be fa tored into the determination of a Type A test "as foend" condition.

b. If a particular penetration or valve fails two cons,ecutive Type B or C tests, t.he frequency of testing that penetration must be 7

[7590-01)

• increased until two satisfactory B or C tests are obtained at the nominal test frequency. Concurrently, existing requirements

• to increase the frequency of Type A tests due to consecutive "as found" failures are already being relaxed in the proposed revision of Appendix J. Instead, attention would be focused on correcting component degradation, no matter when tested, and the ,{

"as found" Type A test would reflect the actual condition of the overall containment boundary.

c. Increases or decreases in Type B or C "as found" test results (over the previous "as lef t" Type B or C test results) shall be k added to or subtracted from the previous "as left" Type A test result.

If this sum exceeds 0.75 L, but is less than 1.0 L,, measures This shall be taken to reduce the sum to no more than 0.75 L,.

will not be considered a reportable condition.

i If this sua exceeds 1.0 L,, measures shall be taken to reduce the sum to no more than 0.75 L ,. This will be considered a reportable condition.

The existing requirements that the sum of all Type B and C tests be no greater than 0.60 L, shall also remain in effect.

8

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Major Changes b

Y The following are the major changes proposed in this rulemaking. (

1. Level of detail. The level of detail addressed in the proposed revision of Appendix J has been limited. This revision of the regulation defines general containment system leakage test criteria.

2. Editorial. For increased clarity, an expanded and revised Table of Contents and set of definitions has been provided, conforming to current usage. The text has also been revised to conform to " plain English" objectives.

3. Interpretations. Some changes have been made to resolve past questions of interpretation (e.g., definitions of " containment isolation valves").

4. Greater flexibility. A major problem with Appendix J has been the lack of a provision for dealing with plants already built where design features are incompatible with Appendix J requirements (e.g. , air lock testing). As a result, provision has been made in this revision for consideration by the NRC staff of alternative leakage test f

requirements when necessary.

j Type A test pressure. The option of performing periodic 5.

l reduced pressure testing in lieu of testing at full calculated accident i

I pressure has been dropped. This change reflects the opinion that extrapolating low pressure leakage test results to full pressure leakage Reasonable argument can test results has turned out to be unsuccessful.

l be made for low pressure testing. However, the NRC staff believes that l

the peak calculated accident pressure (a) has always been the intended reference test pressure, (b) is consistent with the typical practice for NRC staff evaluations of accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in I

9

[7590-01) .

. n .

accordance with Regulatory Guide 1.3 and 1.4, (c) provides at least a . .

nominal check for gross low pressure leak paths that a low pressure leak t

does not provide for high pressure leak paths, (d) directly represents g,.sL technical specification leakage rate limits, and (e) provides~ greater g For these reasons, confidence in containment system leaktight integrity.

the full, rather than reduced, pressure has been retained as the test pressure.

6. Type A test frequency. The test frequency has been uncoupled from the 10 year inservice inspection period used by the ASME Boiler &

Pressure Vessel Code for mechanical systems. A different time base is used, but the frequency has remained essentially the same.

7. Type A test duration. The duration has been dropped from the test criteria in Appendix J. It is considered as part of the testing procedures, and is a function of the state of the testing technology and the level of confidence in it.

8. Type A test "as is" clarification. Appendix J originally noted in III. A.1(a) that the containment was to be "... tested in as close to the 'as is' condition as practical." This is re-emphasized and clarified by the explicit requirements that have been added to measure, record, and report "as found" and "as left" leakage rates.

Type A test allowable leakage rate prorating. Seventy-five 9.

percent of the allowable leakage rate represents the "as left" Type A test acceptance criterion, leaving 0.25 of the allowable leakage rate as a margin for deterioration until the time of the next regulatory scheduled Type A test, when the "as found" leakage rate criterion is 1.0 of the allowable leakage rate.

10. Quantification of allowable leakage rates. It should be noted that no change has been made to the way in which the allowable test 10

[7590-01]

. . leakage rates are quantified. The regulation still refers to the Debate individual plant technical specifications for these values.

continues, however, on what these values should be and whether they can be generically specified, rather than individually specified for each site and plant.

11. Refocusing of corrective actions. When a reportable problem is identified, a Corrective Action Plan is to be submitted. It identifies the problem to the NRC staff, and notes the cause, what was or will be done to correct it, and what will be done to prevent its recurrence.

Increased local leakage testing frequency may be necessary. Appendix J This originally addressed increased test frequency only for Type A tests.

revision applies adjustment of test frequency directly to identified problem areas.

12. The final paragraph of the proposed amendment specifies a date by which an implementation schedule must be submitted, rather than by which it must be implemented. This is because the ease with which licensees will be able to implement all the provisions of the amendment will be highly plant specific depending on plant design, outage and testing schedules, and amount of technical specification changes needed.

l FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY l The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in k Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a }

major Federal action significantly affecting the quality of the human h environment and therefore an environmental impact statement is not 11

[7590-01]

required. There will be no radiological environmental impact offsite, , ,

but there may be an occupational radiation exposure onsite of about 3.0 ,

man-rem per year of plant operation for inspection personnel (about 0.4%

increase). Alternatives to issuing this revision were considered and ,

I found not acceptable. The environmental assessment and finding of no j 3%

significantimpactonwhichthis'determinationisbasedareavailableforjQ '.O inspectionattheNRCPublicDocumentRoom,1717HStreetNW, Washington,'$

DC. Single copies of the environmental assessment and the finding of no significant impact are available from Mr. E. Gunter Arndt, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, Washington, DC 20555 Telephone (301) 443-7893.

PAPERWORK REDUCTION ACT STATEMENT This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 USC 3501 et seq.).

This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

REGULATORY ANALYSIS The Commission has prepared a draft regulatory analysis on the proposed revision. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft analysis is 3 x

available for inspection and copying in the NRC Public Document Room, C D

1717 H Street, NW, Washington, DC. The Commission requests public i

comment on the draft analysis. Comments may be submitted to the NRC as indicated under the Addresses heading.

12

[7590-01]

BACKFIT ANALYSIS '~' <

The Commission has prepared a backfit analysis on the proposed revision, u o.

The analysis is~ required under 10 CFR Part 50. Section 50.109., as of .N l October 21,1985, for the management of backfitting for power reactors.

f The analysis is available for inspection and copying in the NRC Public '

M Document Room, 1717 H Street NW, Washington, DC. The Commission requests , )

R public coment on the analysis. Comments on the analysis may be l submitted to the NRC as indicated under the Addresses heading.

The analysis does not conclude that there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit. It does conclude, however, that the direct and indirect costs of implementation are justified due to better, more uniform tests and test reports, greater com Jence in the reliability of the test results, fewer exemption requ .ts, and fewer interpretive debates. For these reasons, which are presented in greater detail in the backfit analysis, the Comission has.

granted an exemption from the requirement for a determination pursuant to o i50.109(a)(3). This section requires a determination that the rule will h result in substantial increase in the overall protection of the public h health and safety or the comon defense and security to be derived from the backfit.

REGULATORY FLEXIBILITY CERTIFICATION In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C.

605(b)), the Comission certifies that this rule will not, if promulgated, have a significant economic impact on substantial number 13 l

l

-(7b9u-01J ,

of small entities. This proposed rule affects only the licensing and *

• operation of nuclear power plants. The companies that own tt.ese plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business . Size Stand-ards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.

LIST OF SUBJECTS IN 10 CFR PART 50 Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

RELATED REGULATORY GUIDE The notice of availability of a draft regulatory guide on the same subject " Containment System Leakage Testing" (MS 021-5) is also being k N

The draft 4 published in the notice section of this Federal Register. b regulatory guide contains specific guidance on acceptable leakage test l

methods, procedures, and analyses that may be used to implement these

~

requirements and criteria.

1 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 l

of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR Part 50.

l 14

4 o,. . ..- (7590-01]

I PART 50 -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATI 4

1. The authority citation for Part 50 continues to read as follows: *N r l

AUTHORITY:

Secs.103,104,161,182,183,186,189, 68 Stati 936, 937, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C. 2133, 2134, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1246, as amended (42 U.S.C. 5841, 5842, 5846),

unless otherwise noted.

Section 50.7 also issued under Pub. L.95-601, sec. 10.92 Stat. 2951 Sections 50.57(d), 50.58, 50.91, and 50.92 also issued (42 U.S.C. 5851).

under Pub. L.97-415, 96 Stat. 2071, 2073 (42 U.S.C. 2133, 2239).

Section 50.78 also issued under sec.122, 68 Stat. 939 (42 U.S.C. 2152).

also issued under sec. 184, 68 Stat. 954, as amended Sections 50.80-50.81 Sections 50.100-50.102 also issued under sec. 186, (42 U.S.C. 2234).

68 Stat. 955 (42 U.S.C. 2236).

For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C. 2273); $$ 50.10(a), (b), and (c), 50.44, 50.46, 50.48, 50.54,.

and 50.80(a) are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); $$ 50.10(b) and (c) and 50.54 are issued under sec. 1611, 68 Stat. 949, as amended (42 U.S.C. 2201(i)); and $$ 50.55(e),

50.59(b), 50.70, 50.71, 50.72, 50.73, and 50.78 are issued under sec.

161o, 68 Stat. 950, as amended (42 U.S.C. 2201(o)).

2. Appendix J is revised to read as follows:

1 15

---

- _..._--,,--...,-...,..--,r-.-, . = . - - . . _ . . _ . . . ~ . . . . , . _ , _ - .~.,,,--.,---.---_,_,-.-,-_,,..y..

[7590-01)

.t

. , f 'I

~  !

)> i

's Leakage Tests for Containments of l Light-Water-Cooled Nuclear Power Plants Table of Contents I. INTRODUCTION II. DEFINITIONS III. GENERAL LEAK TEST REQUIREMENTS A. Type A Test

1. Preoperational Test

2. Periodic Test

3. Test Frequency

4. Test Start and Finish

5. Test Pressure

6. Pretest Requirements .

7. Verification Test

8. Acceptance Criteria

9. Retesting

10. Permissible Periods for Testing i

B. Type B Test

1. Frequency

2. Pressure .

3. Air Locks

4. Acceptance Criteria C. Type C Test

1. Frequency .

2. Pressure /Mediu:.

16

e y

• . .' [7590-01]

. 9 e

3. Acceptance Criteria

4. Valves That Need Not Be Type C Tested IV. SPECIAL LEAK TEST REQUIREMENTS A. Containment Modification or Maintenance

. 'ks, B. Multiple Leakage Barriers or Subatmospheric Containments V. TEST METHOD, PROCEDURES, AND ANALYSES A. Type A, B, and C Test Details B. Combination of Periodic Type A, B, and C Tests VI. REPORTS A. Submittal B. Content VII. APPLICATION A. Applicability I B. Effective Date l

l t

1 17 t

._~_- _. . _ _ , _ _ _ . _ __ _

[7590-01) , ,

I

1. . INTRODUCTION . .

One of the conditions of all operating licenses for light-water-cooled power reactors as specified in 5 50.54(o) of this part is that ,

primary containments meet the leak test requirements set forth in this Appendix. The tests ensure that (a) leakage through the primary contain-

'- ments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the Technical Specifications ,

I and (b) inservice inspection of penetrations and isolation valves is per-formed so that proper maintenance and repairs are made during their service 1

life. This Appendix identifies the general requirements and acceptance criteria for preoperational and subsequent periodic leak testing.2 II. DEFINITIONS ACCEPTANCE CRITERIA Standards against which test results are to be compared for a

I establishing the functional acceptability of the containment system as

! a leakage limiting boundary.

"AS FOUND" LEAKAGE RATE

! The leakage rate prior to any needed repairs or adjustments to the leakage barrier being tested.

"AS LEFT" LEAKAGE RATE The leakage rate following any needed repairs or adjustments to the leakage barrier being tested.

4 i CONTAINMENT INTEGRATED LEAK RATE TEST (CILRT)

The combination of a Type A test and its verification test.

f 3 Specific guidance concerning acceptable leakage test m  ;

i will be provided in a regulatory guide that is being Copiesissued of thein draft' form regulatory guidefor public comment with the designation MS 021-5.  :

may be obtained from the Nuclear Regulatory Commission Document Manageme

Branch, Washington, DC 20555. r 5

18

_ _ _ _ . . ~_- _ _ _ _ _ . . _ _ _ . _ _ _ _ _

[7590-01]

- . CONTAINMENT ISOLATION SYSTEM FUNCTIONAL TEST A test to verify the proper performance of the isolation system by normal operation of the valves. For automatic containment isolation systems, a test of the automatic isolation system performed by actuation of the containment isolation signals.

CONTAINMENT ISOLATION VALVE Any valve defined in General Design Criteria 55, 56, or 57 of Appen-i dix A " General Design Criteria for Nuclear Power Plants," to this part.

CONTAINMENT LEAK TEST PROGRAM The comprehensive testing of the containment system that includes t

Type A, B, and C tests.

I CONTAINMENT SYSTEM The principal barrier, after the reactor coolant pressure boundary, to prevent the release of quantities of radioactive material that would It have a significant radiological effect on the health of the public.

includes:

(1) the primary containment, including access openings and penetrations.

(2) containment isolation valves, pipes, closed systems, and other components used to effect isolation of the containment atmosphere from the outside environs, and (3) those systems or portions of systems that by their functions extend the primary containment boundary to include their system boundary. k This definition does not include boiling water reactors' (BWR) reactor fk Also

' - - - - - buildings or pressurized water reactors' (PWR) shield buildings.

excluded from the provisions of this Appendix are the interior barriers 19 4

i

,e--- . , . . - --.,, __r.- .,,,,,,-,,.,,,_--,_.__,,.,,-_.,--,,.-,_,y.., yy_-. ._~ -._,.--...,_.4--, ~ . .-- - , , ..%..,,,,,-m , - , - - . ,

(7590-01) ,

such as the BWR Mark 11 drywell floor and the drywell perimeters of the

BWR Mark III and the PWR ice condenser.

L,(WEIGHT PERCENT /24 HR)

The maximum allowable Type A test leakage rate in units of weight percent per 24-hour period at pressure P ac as specified in the Technical Specifications.

L,,(WEIGHT PERCENT /24 HR)

The measured Type A test leakage rate in units of weight percent per 24-hour period at pressure P,g, obtained from testing the containment system in the state as close as practical to that that would exist under design basis accident conditions (e.g., vented, drained, flooded, or pressurized).

LEAK An opening that allows the passage of a fluid.

LEAKAGE The quantity of fluid escaping from a leak.

LEAKAGE RATE The rate at which the contained fluid escapes from the test volume at a specified test pressure.

MAXIMUM PATHWAY LEAKAGE RATE The maximum leakage rate that can be attributed to a penetration leakage path (e.g., the larger, not total, leakage of two valves in series). This generally assumes a single active failure of the better of two leakage barriers in series when performing Type B or C tests.

MINIMUM PATHWAY LEAKAGE RATE The minimum leakage rate that can be attributed to a penetration This leakage path (e.g., the smallest leakage of two valves in seTies).

20

[7590-01) is used when correcting the measured value of containment leakage rate from the Type A test (L,,) to obtain the overall integrated leakage rate and generally assumes no single active failure of redundant leakage-barriers under these test conditions.

OVERALL INTEGRATED LEAKAGE RATE I

The total. leakage rate through all leakage paths, including contain-ment welds, valves, fittings, and components that penetrate the contain-ment system, expressed in units of weight percent of contained air mass at test pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

P,c(psig)

The calculated peak containment internal pressure related to the design

- basis loss-of-coolant accident as specified in the technical specifications.

PERIODIC LEAK TEST Test conducted during plant operating lifetime.

PREOPERATIONAL LEAK TEST Test conducted upon completion of construction of a primary or secondary containment, including installation of mechanical, fluid, electrical, and instrumentation systems penetrating these containment systems, and prior to the time containment integrity is required by the Technical Specifications.

PRIMARY CONTAINMENT The structure or vessel that encloses the major components of the reactor coolant pressure boundary as defined in l 50.2(v) of this part and is designed to contain accident pressure and serve as a leakage

- - - - barrier against the uncontrolled release of radioactivity to the environ-ment. The term " containment" as used in this Appendix refers to the ,

primary containment structure and associated leakage barriers.

21

[7590-01)

. c.

STRUCTURAL INTEGRITY TEST .

A pneumatic test that demonstrates the capability of a primary containment to withstand a specified internal design pressure load.

TYPE A TEST A test to measure the containment system overall integrated leakage rate under conditions representing design basis loss-of-coolant accident containment pressure and systems alignments (1) after the containment systen. has been completed and is ready for operation and (2) at periodic intervals thereafter. The verification test is not part of this definition - see CILRT.

TYPE 8 TEST A pneumatic test to detect and measure local leakage through the following containment penetrations:

(1) Those whose design incorporates resilient seals, gaskets, b sealant compounds, expansion bellows, or fitted with flexible metal seal p

N assemblies.

(2) Air locks, including door seals and door operating mechanism penetrations that are part of the containment pressure boundary.

TYPE C TEST

\

A pneumatic test to measure contairment isolation valve leakage rates.

VERIFICATION TEST  %

Test to confirm the capability of the Type A test method and equip- 4 ment to measure L,.

III. GENERAL LEAK TEST REQUIREMENTS A. Type A Test ~

A preoperational Type A test must be g (1) Preoperational Test. N conducted on the containment system and must be preceded by: Qh 22

[7590-01)

- . (a) Type B and Type C tests, (b) A structural integrity test.

A periodic Type A test must be performed on the (2) Periodic Test. k 5

containment system. Sb (3) Test Frequency. Unless a longer interval is specifically .

approved by the NRC staff, the interval between the preoperational and first periodic Type A tests must not exceed three years, and the interval If between subsequent periodic Type A tests must not exceed four years.

the initial fuel loading is delayed so that the three year interval between the first preoperational test and the first periodic test is If such an addi-exceeded, another preoperational test will be necessary.  %

~

A tio~nal preoperationa1 Type A test or an additional Type A test required by Sections III. A.8 or IV. A. of this Appendix is performed, the Type A test interval may be restarted.

The Type A test pressure must be equal to or (4) Test Pressure.

greater than P,e at the start of the test but must not exceed the contain-for the ment design pressure and must not fall more than 1 psi below P ac The test duration of the test, not including the verification test.

pressure must be established relative to the external pressure of the containment.

This may be either atmospheric pressure or the subatmospheric pressure of a secondary containment.

(5) Pretest Requirements. Closure of containment isolation valves for the Type A test must be accomplished by normal operation and without any preliminary exercising or adjustments for the purpose of improving l

- -performance-(e.g., no tightening of valve after closure by valve motor).

Repairs of malfunctioning or leaking valves must be made as necessary.

l Information on valve leakage that requires corrective action' prior to, I

23 i

[

) - --- .,-- - _ _, _ _ . _

[7590-01] . ,.

D

' during, or after the test (see Section V.8.) must be included in the . .

report submitted to the Commission as specified in Section VI of this Appendix.

A leakage rate verification test must be ,

(6) Verification Test.

performed after a Type A test in which the leakage rate meets the criteria in Ill. A.(7)(b)(ii). The verification test selected must be conducted for a duration sufficient to establish accurately the change in The results of leakage rate between the Type A and verification tests.

the Type A test are acceptable if the sum of the verification test imposed leakage and the containment leakage rate calculated from the Type A test i

(L,,) does not differ from the leakage rate calculated from the verifica-tion test by more than 20.25 L ,.

(7) Acceptance Criteria.

(a) For the preoperational Type A Test, the "as left" leakage rate must not exceed 0.75L,, as determined by a properly justified statistical analysis. The "as found" leakage rate does not apply to the

.preoperational test. b (b) For each periodic Type A test, the leakage rate, as deter-mined by a properly justified statistical analysis, must not exceed:

(1) L,, for the "as found" condition, (ii) 0.75L,, for the "as left" condition,

! (c) In meeting these Type A test acceptance criteria, isola-tion, repair, or adjustment to a leakage barrier that may affect the leakage rate through that barrier is permitted prior to or during the Type A test provided:

(i) all potential leakage paths of the isolated, repaired, or adjusted leakage barrier are locally leak testable, and 24 1

(7590-01]

. o

' ~

(ii) the local leakage rates are measured before and after the isolation, repair, or adjustment and are reported under Section VI of this Appendix.

(iii) All changes in leakage rates resulting from isola-tion, repair, or adjustment of leakage barriers subject to Type B or Type C testing are determined using the minimum pathway leakage method and added to the Type A test result to obtain the "as found" and "as left" containment leakage rates.

(d) The effects of isolation, repair, or adjustments to the containment boundary made after the start of the Type A test sequence on the Type A test results must be quantified and the appropriate analytical corrections made (this includes tightening valve stem packing, additional tightening of manual valves, or any action taken that will affect the leakage rates).

(8) Retesting.

(a) If, for any periodic Type A test, the as found leakage rate fails to meet the acceptance criterion of 1.0L,, a Corrective Action Plan that focuses attention on the cause of the problem must be developed and implemented by the licensee and then submitted together with the Containment Leak Test Report as required by Section VI of this Appendix.

The test schedule applicable to subsequent Type A tests (III.A.(3)) shall be submitted to the NRC staff for review and approval. An as left Type A k N

l test that meets the acceptance criterion of 0.75L,is required prior to plant startup.

- - - - - - (b)-If-two consecutive periodic as found Type A tests exceed the as found acceptance criterion of 1.0L,: , .

25 I

I

[7590-01]

(1) Regardless of the periodic retest schedule of III. A.(3),

a Type A test must be performed at least every 24 months (based on the refueling cycle normally being about 18 months) unless an alternative leakage test program is acceptable to the NRC staff on some other defined basis. This testing must be performed until two consecutive periodic "as b found" Type A tests meet the acceptance criterion of 1.0L, after which the retest schedule specified in III.A.(3) may be resumed.

(ii) Investigation as to the cause and nature of the

Type A test failure might indicate that an alternative leakage test program such as more frequent Type B or Type C testing may be more appro-priate than the performance of two consecutive successful Type A leakage tests. The licensee may then submit a Corrective Action Plan and an l

If this alternative leakage test program proposal for NRC staff review.

submittal is approved by the NRC staff, the licensee may implement the corrective action and alternative leakage test program in lieu of one or both of the Type A leakage tests required by Section III. A.(8)(b)(i).

i i

(9) Permissible periods for testing. The performance of Type A tests must be limited to periods when the plant facility is secured in the shutdown condition under the administrative controls and safety procedures defined in the license. ,

B. Type B Test (1) Frequency.

(a) Type B tests, except tests for air locks, must be performed on containment penetrations during shutdown for refueling or at other convenient intervals but in no case at intervals greater than 3 .

26 l

,-------c,,-w-rmm,-,,,--rr---------e=~+-w------r.--e-e-i

[7590 013 r

.- - 2 years. If opened following a Type A or 8 test, containment penetra-tions subject to Type B testing must be Type B tested prior to returning the reactor to an operating mode requiring containment integrity. '

4 (b) For containment penetrations employing a continuous

' leakage monitoring system that is at a pressure not less than P,c, i

leakage readings of sufficient sensitivity to permit comparison with Type B test leak rates must be taken at intervals specified in the Tech- '

nical Specifications. These leakage readings must be part of the Type B reporting of VI.A. When practical, continuous leakage monitoring systems must not be operating or pressurized during Type A tests. If the contin-b A

~

uous leakage monitoring system cannot be isolated, such as inflatable air lock doo'r seals, leakage i'nto the containment must be accounted for and

~

1

the Type A test results corrected accordingly.

(2) Pressure. Type B tests must be conducted, whether individually t

or in groups, at a pneumatic pressure not less than P,c except as pro-i

' vided in paragraph III.B.(3)(b) of this section or in the Technical Specifications.

(3) Air Locks.

(a) Initial and periodic tests. Air locks must be tested prior to initial fuel loading and at least once each 6-month interval l

Alternatively, if thereafter at an internal pressure not less than P,e. ,

there have been no air lock openings within 6 months of the last i

' successful test at P,c, this interval may be extended to the next i

refueling outage or airlock opening (but in no case may the interval N +

exceed.2. years). _ Reduc.ed_ pressure _ te_sts must_ continue to be performed on p

j the air lock or its door seals at 6-month intervals. Opening of the air lock for the purpose of removing air lock testing equipment following an

' air lock test does not require further testing of the air lock.

' 27

[7590-01] .

(b) Intermediate tests must be conducted as follows: ,

(i) Air locks opened during periods when containment integrity is required by the plant's Technical Specifications must be tested within 3 days after being opened. For air lock doors opened more frequently than once every 3 days, the air lock must be tested at least once every 3 days during the period of frequent openings. Air locks g

'2' opened during periods when containment integrity is not required by the b

plant's Technical Specifications need not be repeatedly tested during such periods. However, they must be tested prior to the plant requiring containment integrity. For air lock doors having testable seals, testing the seals fulfills the intermediate test requirements of this paragraph.

In the event that this intermediate testing cannot be done at Pac, the

! test pressure must be as stated in the Technical Specifications.

(ii) Whenever maintenance other than on door seals has been performed on an air lock, a complete air lock test at a test pressure of not less than P ac is required, if that maintenance involved the pressure retaining boundary.

(iii) Air lock door seal testing or reduced p* assure testing may not be substituted for the initial or periodic full-pressure test of the entire air lock required in paragraph III.B.(3)(a) of this Section.

(4) Acceptance Criteria.

(a) The sum of the as. found or as lef t Type B and C test results must ot exceed 0.60L,using maximum pathway leakage and including leakage rate readings from continuous leakage monitoring systems.

(b) Leakage measurements are acceptable if obtained through component leakage surveillance systems (e.g. , continuous pressurization 28

(7590-01]

of individual or clustered containment components) that maintain a pres-sure not less than P,g at individual test chambers of those same contain-ment penetrations during normal reactor operation. Similar penetrations not included in the component leakage surveillance system are still subject to individual Type B tests.

(c) An air lock, penetration, or set of penetrations thlit fails to pass a Type B test must be ratested following determinstion.of cause and completion of corrective action. Corrective action to correct the leak and to prevent its future recurrence must be developed and implemented.

(d) Individual acceptance criteria for all air lock tests must be stated in the Technical Specifications.

C. Type C Test (1) Frequency. Type C tests must be performed on containment isola-tion valves during each reactor shutdown for refueling or at other convenient intervals but in no case at intervals greater than 2 years.

(2) Pressure / Medium.

(a) Containment isolation valves unless pressurized with a qualified water seal system must be pressurized with air or nitrogen at a pressure not less than Pac *

(b) Containment isolation valves, that are sealed with water from a qualified seal system, must be tested with water at a pressure not less than 1.10 P ac' (3) Acceptance Criteria. --__. - -....__._ _ _ _____

(a) The sum of the as found or as left Type B and C test results must not exceed 0.60L, using maximum pathway leakage and including leakage rate readings from continuous leakage monitoring systems.

29

[7590-01) ,, ,

(b) Leakage from containment isolation valves that are sealed with water from a seal system may be excluded when determining the combined Type 8 and C leakage rate if:

(i) The valves have been demonstrated to have. leakage-rates that do not exceed those specified in the Technical Specifications, and (ii) The installed isolation valve seal system inventory is sufficient to ensure the sealing function for at least 30 days at a

~

pressure of 1.10 P,c.

(4) Valves That Need Not Be Type C Tested.  %-

(a) A containment isolation valve need not be Type C tested if k

=

it can be shown that the valve does not constitute a potential contain-ment atmosphere leak path during or following an accident, considering a single active failure of a system component.

[

(b) Other valves may be excluded from Type C testing only when approved by the NRC staff under the provisions of paragraph VII.A.

i IV. SPECIAL LEAK TEST REQUIREMENTS I- A. Containment Modification or Maintenance l

Any modification, repair, or replacement of a component that is part of the containment system boundary and that may affect containment integrity must be followed by either a Type A, Type B, or Type C test.

Any modification, repair, or replacement of a component subject to Type 8 or Type C testing must also be preceded by a Type 8 or Type C test. The measured leakage from this test must be included in the report to the Commission required by Section VI of this Appendix. Following structural changes or repairs that affect the pressure boundary, the licensee shall 30 l

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. - demonstrate whether or not a structural integrity test is needed pr'ior to the next Type A test. The acceptance criteria of paragraphs III.A.(7),

III.B.(4), or III.C.(3) of this Appendix, as appropriate, must be met.

Type A testing of certain minor modifications, repairs, or replacements may be deferred to the next regularly scheduled Type A test if local leakage testing is not possible and visual (leakage) examinations or non-destructive examinations have been conducted. These shall include:

Welds of attachments to the surface of the steel pressure retaining Q

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boundary; Repair cavities the depth of which does not penetrate the  %

required design steel wall by more than 10%; Welds attaching to the steel pressure retaining boundary penetrations the nominal diameter of

~ ~ ~ ~

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which does not exceed one inch.

B. Multiple Leakane Barrier or Subatmospheric Containments The primary reactor containment barrier of a cultiple barrier or subatmospheric containment shall be subjected to Type A tests to verify that its leakage rate meets the requirements of this appendix. Other structures of multiple barrier or subatmospheric containments (e.g., 4b secondary containments for boiling water reactors and shield buildings <

for pressurized water reactors that enclose the entire primary. reactor containment or portions thereof) shall be subject to individual tests in ,

s accordance with the procedures specified in the technical specifications. }a V. TEST METHODS, PROCEDURES, AND ANALYSES A. Type A B and C Test Detafis _ . _ . __ ___ . . _

Leak test methods, procedures, and analyses for a steel, concrete, g

• Tn or combination steel and concrete containment and its penetrations and 1 31

[7590-01) '

isolation valves for light-water-cooled power reactors must be referenced . .

or defined in the Technical Specifications. .

N B. Combination of Periodic Type A. 8. and C Tests Type 8 and C tests are considered to be conducted in conjunction with the periodic Type A test when performed during the same outage as the Type A test. The licensee shall perform, record, interpret', and report the tests in such a manner .that the containment system leak-tight status is determined on both an as found basis and an as left basis, i.e., its leak status prior to this periodic Type A test together with the related Type B and C tests and its status following the conclusion of these tests.

1 I

VI. REPORTS A. Submittal

1. The preoperational and periodic Type A tests, including sum-maries of the results of Type B and C tests conducted in conjunction with the Type A test, must be reported in a summary technical report sent not \s later than 3 months after the conduct of each test to the Commission in j{

l the manner specified in 5 50.4. The report is to be titled " Containment if Leakage Test."

i

2. Reports of periodic Type B and C tests conducted at intervals intermediate to the Type A tests must also be submitted to the NRC as in Reports paragraph VI.A.1 at the time of the next Type A test submittal.

must be submitted to the NRC Regional Administrator within 30 days of completion of any Type B or C tests that fail to meet their as found acceptance criteria.

l 32 I

[7590-01)

8. Content A Type A test Corrective Action Plan, when required under paragraph III.A.(8) of this Appendix, must be included in the report. Any correc-tive action required for those Type B and C tests included as a part of the Type A test sequence must also be included in the report.

VII. APPLICATION A. Aphlicability The requirements of this Appendix apply to all operating nuclear power reactor licensees as specified in $$ 50.54(o) of this part unless it can'be demonstrated that alternative leak test requirements (e.g. , for~ ~ ~

~~

certain containment designs, leakage, mitigation systems. or different test pressbres not specifically addressed'in this Appendix) are demon-i strat'ed t'o be adequate on some other defined basis. Alternative leak test requirements and the bases for them will be made a' part of the plant Technical Specifications if approved by the NRC staff.

B. Effective Date  %

This Appendix is effective (30 days after publication). By (insert $%

a date 180 days after the affective date of this revision), each lD licensee and each applicant for an operating license shall submit a plan to the Director of the Office of' Nuclear Reactor Regulation for imple-menting this Appendix. This submittal must include an implementation h A

P

' schedule, with a final implementation no later than (insert a date 48 g

Until_the_ licensee ____ M__

months after the effective date of this revision).-

-finally implements the provisions of this revision, the licensee shall lg M

continue to use in their entirety the existing Technical Specifications 33

(7590-013 , ,

p 0 s cs and the Appendix J on which they are based. Thereafter, the licensee h{, j2

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shall use in their entirety this revision and the Technical Specifica-  ? j{f t;

tions conforming to this revision. lin M.

Dated at Washington, DC, this day of , 1985.

For the Nuclear Regulatory Commission.

Samuel J. Chilk Secretary of the Commission 34

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FINDING OF NO SIGNIFICANT IMPACT O

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ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT; PROPOSED REVISION TO APPENDIX J OF 10 CFR PART 50 The Nuclear Regulatory Commission is proposing to amend its regulations to update the criteria and clarify questions of interpretation in regard to leakage testing of containments of light-water-cooled nuclear power

- plants.

Environmental Assessment Identific'ation of Proposed Action Appendix J of 10 CFR Part 50 was originally issued for public connent as a proposed rule on August 27, 1971 (36 FR 17053); published in final form on February 14,1973 (38 FR 4385); and became effective on March 16, 1973.

The only amendment to this Appendix since 1973 was a limited one, on Type on B (penetration) test requirements that was published for comment

- January 11,1980 (45 FR 2330); published in final form September 22, 1980 J

(45 FR 62789); and became effective on October 22, 1980.

This revision of Appendix J has been in preparation for some time. It will provide greater flexibility in applying alternative requirements due to variations in plant design and reflects changes based on: (1) experience in applying the existing requirements; (2) advances in containment leak testing methods; (3) interpretive questions; (4) simplifying the text, (5) various external / internal comments since 1973; and(6)exemptionrequestsreceivedandapproved.

I

Need for the Proposed Action Changes in the state-of-the-art of leakage testing, experience .with using the test criteria, and the evolution and variety of plant designs have made'it necessary to update the 1973 criteria.

Environmental Impacts of the Proposed Action The proposed revision of Appendix J will have no radiological environmental impact offsite. However, if the rule is promulgated in final form as now proposed, there will be an average increase in occupational radiation exposure onsite of about 3.0 man-rem per year of plant operation for inspection personnel (i.e., occupational radiation exposure is increased on average about 0.4%). This is due to the increase in the number of inspections in order to improve the confidence level in the data.

The amendment does not affect non-radiological plant effluents and has no other environmental impact. Therefore, the Comission concludes that there are no significant non-radiological environmental impacts associated with the proposed amendment.

Alternatives to the Proposed Action As required by Section 102(2)(E) of flEPA (42 U.S.C.A. 4332(2)(E)), the staff has considered possible alternatives to the proposed action. One 2

alternative was not to initiate a rulemaking proceeding. This is not acceptable as there would be increasing conflicts between the regulation and current testing procedures. This would only produce more' exemption requests; a further drain on applicant and staff resources. There would be no environmental impact change but problems . incurred in using the present rule would not be resolved.

t Issuing a regulatory guide and abolishing the rule was considered. This is not acceptable because a regulatory guide is non-mandatory. The staff feels that there could be an increase in exposure to the public if the testing were non-mandatory and containment integrity were not maintained.

The present approach of revising the existing rule was chosen as the best alternative. Revision of Appendix J will be beneficial to all. - The public will benefit from improved reliability of containment leakage integrity. The NRC staff will benefit from fewer exemption requests, clearer and more complete test criteria, increased regulatory flexibility, fewer interpretive debates, more useful test reports, and improved, more representative, and uniform testing programs. Utilities will derive the same benefits, as well as having test criteria that focus more accurately on problem areas and which could result in significant cost savings.

I Alternative Use of Resources No alternative use of resources was considered.

f

' 3

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Agencies and Persons Consulted , ,

The staff relied on an analysis perfomed by Science and Engineering Associates, and a study performed by Oak Ridge National Laboratory.

i Finding of no Significant Impact The Comission has determined not to prepare an environmental impact statement for the proposed amendment. .

Based on the foregoing environmental assessment, we conclude that the  ;

proposed action will not have a significant effect on the quality of the human environment.

For further details with respect to this action, see the Final Report by Science and Engineering Associates, dated April 1985, and NUREG/CR-3549,

" Evaluation of Containment Leak Rate Testing Criteria" which are available for public inspection at the Comission's Public Document Room,1717 H Street. N.W., Washington, D.C.

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02 REPORTING REVIEW O

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• ,j, , jy SUPPORTING STATEMENT FOR PROPOSED RULE 10 CFR 50, APPENDIX J, IVII.B.

A. JUSTIFICATION

1. Need for the Collection of Information.

The containments of all light-water-cooled reactor power plants must be periodically leakage tested. These leakage tests help assure that any radioactive materials released into the containment will be suitably contained and that releases to the outside environment will be small. The leakage test requirements are specified in 10 CFR 50, Appendix J, " Leakage Tests for Containments of Light-Water-Cooled Nuclear Power Plants."

Adoption of the amended rul'e would add the following reporting requirement in Paragraph VII.B. By 180 days after the effective date of the amendment, each licensee and each applicant for an operating license shall submit a plan to the Director of the Office of Nuclear Reactor Regulation for implementing the amended Appendix. This plan is in place of a universal, automatic compitance date, and provides more flexibility to licensees in implementing the revised rule.

2. Agency Use of Information.

Use of this implementation approach permits a more practical and orderly phasing-in of new Technical Specifications and Appendix J testing procedures than would otherwise be possible. NRC staff review and approval of implementation plan will be within 6 months of submittal.

3. Reduction of Burden Through Information Technology.

~

There are no information technology applications which would reduce the burden associated with this information collection.

2

4. Effort to Identify Duplication.

The information being requested is unique to NRC's activity and therefore there is no duplication. FILS was searched with negative results.

5. Effort to Use Similar Information.

No similar information exists, since this report is an implementation schedule requested in lieu of arbitrarily imposing the revised rule immediately.

6. Effort to Reduce Small Business Burden.

Nuclear power plant licensees, those affected by this reporting requirement, are not "small entities" as defined in the Regulatory Flexibility Act or The Small Business Administration at 13 CFR Part 121.

7. Consequences of Less Frequent Collection.

- As a one-time only report, there is no less frequent collection.

8. Circumstances Which Justify Variation From OBM Guidelines.

There are no special circumstances that require the collection to be conducted in a manner inconsistent with the guidelines in 5 CFR 1320.6.

9. Consultations Outside the NRC.

The proposed rule revision is being published in the Federal Register for public comment, so all interested parties outside I the NRC will have q ual opportunity to provide input to the final rule.

10. Confidentiality of Information.

l No assurance of confidentiality of information has been requested 1

or provided.

11. Justification for Sensitive Ouestions.

[

! No questions of a sensitive nature are being asked of the nuclear

____.J * *t al*"t l#*"****

3

. c,

12. Estimated Annualized Cost to the Federal Government.

Implementation schedules will be reviewed in lieu of exemption requests that would otherwise be submitted if the rule were made effective immediately and arbitrarily. However, taken over 3 years, the annualized cost of review and approval of 127 responses x 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> x $60.00/hr i 3 years would be $304,800.

13. Estimate of Burden.

a. As a one-time only report, the annualized burden to nuclear power plant licensees taken over 3 years would be 20,320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> - 3 years = 6773.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> annually,

b. As above, the annual cost, taken over 3 years would be 6773.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> x $60.00/ hour which equals $406,398.00 annually.

c. Estimated one-time reporting burden.

1. Number of respondents = 127 (93 operating + 34 underconstruction)

2. Number of responses per respondent = 1

3. Total annual responses = 127

4. Hours per response = 160 (2 weeks x 2 people to review rule, tech specs, plant schedules and layout, and write report) (Based on discussion with NRC Region).

5. Total hours = 20,320

14. Reasons for Change in Burden.

The rule change adds a one-time information collection 4

requirement.

15. Publication for Statistical Use.

This collection of information will not have any results that will be published for statistical use. ,

B. COLLECTIONS OF INFORMATION EMPLOYING STATISTICAL METHODS 4

Statistical methods are not used in the collection of information.

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ENCLOSURE 4 i

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FR NOTICE OF REG. GUIDE AVAILABILITY En /. y

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[7590-01]

NUCLEAR REGULATORY COM ISSION Draft Regulatory Guide; Issuance. Availability The Nuclear Regulatory Comission has issued for public coment a draft of a new guide planned for its Regulatory Guide Series. This series has been developed to describe and make available to the public methods acceptable to the NRC staff -of implementing specific parts of the Connission regulations and, in some cases, to delineate techniques used by the staff in evaluating specific problems or postulated accidents and to ,

provide guidance to applicants concerning certain of the infomation needed by the staff in its review of applications for pemits and licenses.

The draft guide, temporarily identified by its task number, MS 021-5 (which should be mentioned in all correspondence concerning this draft guide), is entitled " Containment System Leakage Testing" and is intended

- for Division 1, " Power Reactors." It is being developed to provide

f. guidance on procedures acceptable to the NRC staff for conducting

- - containment leakage tests. This draft guide endorses ANSI /ANS-56.8-1981,

• '" Containment System Leakage Testing Requirements."

t This draft guide is being issued to involve the public in the early l,

stages of the development of a regulatory position in this area. It has received complete staff review but'does not represent a final NRC staff position.

A separate regulatory analysis has not been prepared for this guide.

This is because an extensive analysis, including a contractor-generated cost / benefit analysis, has been prepared and made available in conjunction with the proposed revision to 10 CFR Part 50, Appendix J, that is

[7590-01] ,

~

l published for public coment elsewhere in this Federal Register. This . .

regulatory guide clarifies acceptable positions for implementing the criteria of the proposed revision to Appendix J. As such, it has been an inherent portion of the development package for the proposed Appendix J revision. Readers are therefore referred to the proposed Appendix J revision and to supporting documentation for a comprehensive perspective on the use of this guide.

Public coments are being solicited on the draft guide (including any

~

implementationschedule). Coments should be sent to the Secretary of the Comission. U.S. Nuclear Regulatory Comission. Washington, DC 20555, Attention: Docketing and Service Branch, by I

Although a time limit is given, coments and suggestions in ,-

connection with (1) items for inclusion in guides currently being developed or (2) improvements in all published guides are encouraged at any time.

Regulatory guides are available for inspection at the Comission's Public Document Room, 1717 H Street NW, Washington, DC. Requests for single copies of draft guides (which may be reproduced) or for placement on an automatic distribution list for single copies of future draft guides in specific divisions should be made in writing to the U.S. Nuclear Regulatory Comission, Washington, DC, 20555, Attention: Director.

Division of Technical Information and Document Control. Telephone requests cannot be accomodated. Regulatory guides are not copyrighted, and Comission approval is not required to reproduce them.

Dated at Rockville, Maryland, this day of 1985 Guy A. Arlotto. .

Director, Division of Engineering Technology, Office of Nuclear Regulatory Research 2

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• mm ENCLOSURE 5 O

Draft E3 12/27/85

~~ ~'

DRAFTREGULATORYGUIDE'M'S"021'-5 ~ ~ ~ ~ ~ ~

• G. Arndt x37893-

- o DRAFT E3 Decemberg,1985 REGULATORY GUIDE MS 021-5 CONTAINMENT SYSTEM LEAKAGE TESTING A. INTRODUCTION General Design Criteria 1, " Quality Standards and Records," 16, " Containment Design," 50, " Containment Design Basis," 52, " Capability for Containment Leakage Rate Testing," 53, " Provisions for Containment Testing and Inspection," 54,

" Piping Systems Penetrating Containment," 55, " Reactor Coolant Pressure Boundary Penetrating Containment," 56, " Primary Containment Isolation," and 57, " Closed Systems Isolation Valves," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utili-~ ~ ~~'

zation Facilities," require, in part, that the containment system bi~ des'igned and constructed.for periodic integrated and local leakage rate testing at con-tainment design pressure. On , 1985, the Commission published proposed ~ amendments (_FR ) to Appendix J, " Leakage Tests for Containments of Light-Water-Cooled Nuclear Power Plants," to 10 CFR Part 50, which defines the criteria for such testing. This guide describes a method acceptable to the NRC staff for complying with these amended regulations if they are promulgated as published.

B. DISCUSSION J

j Backaround

! American National Standard ANSI /ANS 56.8-19811 " Containment System

,! Leakage Testing Requirements," was prepared by the American Nuclear Society

Standards Committee, Working Group ANS 56.8, and published in 1981 as a replacement to ANSI N45.4-1972, " Leakage Rate Testing of Containment Structures

! for Nuclear Reactors" (ANSI-7.60).

3 Copies of the American National Standard, ANSI /ANS~ 56.8-19811 " Containment-System Leakage Testing Requirements," may be obtained from the American Nuclear Society, 555 North Kensington Avenue, LaGrange Park, IL 60525. It may be inspected at the Nuclear Regulatory Commission's Public Document Ro6m, 1717 H i Street NW., Washington, DC.

1 RG DIV TASK MS 021-5 Draft 12/27/85

The old ANSI N45.'4-1972 standard was endorsed and referenced without . .

exceptions in Appendix J to 10 CFR Part 50. The new ANSI /ANS 56.8-1981 standard has been considerably expanded and updated and it has become difficult to endorse the standard without some exceptions. As a result, the new standard is being endorsed in this regulatory guide instead of in the revised Appendix J to facilitate the listing of exceptions to the standard and their removal as the stan.dard is revised or errata sheets are issued.

t In revising Appendix J to 10 CFR Part 50, it was intended that the regula-tion limit itself to general test criteria and leave detailed testing techniques and analyses to the ANSI standard. This will permit the standard and its endorsing regulatory guide to be revised as the testing technology changes without affecting the basic test criteria in the regulation and without requiring the regulation to be frequently rewritten to keep it up to date with the i testing technology.

There will always be some debate over whether certain positions are properly regulatory criteria or details of the testing procedures. However, this division of requirements and procedures is believed to provide the most responsive arrangement that will ensure safe limits on containment system leakage while keeping current with technical advances in testing procedures and analysis methods. Also, by having the regulation address general test criteria and leaving the details of implementation to the standard, it is expected that fewer license exemptions will have to be filed than have been necessary under the previous regulation, thereby reducing an unproductive administrative burden on both licensees and NRC staff.

Discussion of Regulatory Positions In those areas in which the provisions of the referenced standard are insufficient for' licensing purposes or where special emphasis is desired, the staff has provided supplementary guidelines (recommendations) it considers needed. These are in the Regulatory Position. Brief reasons for including them are given below.

1. Conflict. This position eliminates the need to identify every difference between the standard and the regulation, and emphasizes how such differences should be handled.

2. Type A Test Requirement. Paragraph 3.3.7 requires that the leakage rate include an upper confidence limit (UCL). Since the Type A test results I Draft 12/27/85 2 RG DIV TASK MS 021-5 i

being corrected are at the UCL, the correction being applied to instrument For. clarity, minimum pathway leakage is error should also be at the UCL.

being explicitly mentioned.

Pressurizina Considerations _. Some plant designs have auxiliary steam 3.

lines penetrating primary containment for use during outages.~ Where possible, such lines should be isolated and vented to prevent the introduction of another energy source that may prove difficult to account for in the calculations.

4. Liquid Level Monitoring. The amount of error resulting from the neglect of changes in sump level is not related to the fact that the water condensed from the air. This is because the air mass equations, provided

'elsewhere in the standard, subtract out the effect of the water vapor changes in air mass but make no provision for the volume changes resulting from the conversion to (or from) water.

5. Type A Test Frequency. This position conforms to current practice and represents a practical and logical interpretation of the end of the test -- - - -

interval.

6. Verification Test. a. Based upon experience with the existing verification test critera, several clarifications needed to be documented.

For periodic Type A tests, consideration is being given to the future use of a zero-pressure test to verify the ability of the Type A test instrumentation to read the Type A test leakage. This is, however, still in the future, and to what extent it could supplement or replace the current verification test or current instrument selection guide criteria has still to be determined.

b. is subject to the same statistical errors as The measurement of W2 "

l 3

the measurement of the air mass values used in the calculation of the leak i rate. It is not likely that a believable determination of the step change Since 20 sets of data points are could be made with one air mass data set.

required to establish the leak rate (paragraph 5.4, page 11 of the standard),

it would be appropriate to require a minimum number of data sets for the ,

verification test also. The formula result is reformatted to more clearly represent the preceding text.

7. Type B and C Test Pressures. In order for the test results to be ~

a method -

valid, either the test differential pressure must be equal to P ac must exist to correctly extrapolate to Pac. At present, much controversy Until exists on how to extrapolate test results to a much higher pressure.

3 RG DIV TASK MS 021-5 Draft 12/27/85

such controversy can be satisfactorily resolved, it is prudent to perform all . ,

Type B and C tests at Pac *

8. Type-B and C Test Schedule. This position conforms to current practice, j and represents a practical and logical interpretation of the test interval.

9. Test Medium and Water Filled Systems. The NRC staff always' applies the single-failure criterion in the review of containment related systems.

10. Calibration. a., b. As presently written, the only requirement is for a one-point in situ check within six months of performing the preopera-tional or periodic test. This is insufficient to ensure system accuracy, linearity, sensitivity, and repeatability. The requirements of paragraph 4.2.1 of the standard for initial calibration data do not state when such calibration is to be performed. Therefore, as presently written, the instruments could be initially calibrated at the time of purchase (perhaps two years before performing the preoperational test) and then a one point check performed within six months of each test for the next forty years. In addition, installing equipment )

and performing calibrations six months prior to the preoperational Type A test will almost guarantee damage to the equipment prior to its use, as considerable construction work is still in progress at this time. Typically, licensees have the instruments calibrated within the six-month period, and then installed and checked within two weeks to a month prior to performing the test.

11. Temperature Measurement. a. The fraction of the containment volume assigned to each eensor should be able to be confirmed. Recommendations are provided for the basis for such confirmation.

b., c. The ANS 56.8 standard addresses temperature surveys, but clarifica-tion is needed regarding the use of fans and when dewpoint temperature surveys should be made. The temperature surveys should be made with the ventilation configuration to be used in the tests. Since operational heat sources are so different from preoperational conditions, the surveys should be rerun at least

~

for the first periodic Type A leakage rate test, as is stated for drybulb temperature surveys.

12. Absolute Test Method. The absolute test method should reflect spatial temperature variability over the containment volume when it exists. The current equation, although correct, inadequately defines the temperature term, permitting its calculation by allowing the assumption of a uniform density throughout the containment. This density may not, however, be uniform because Draft 12/27/85 4 RG DIV TASK MS 021-5

l the temperature may not be uniform. Although this assumption yields similar, j acceptable results with temperature distributions normally experienced, it is l still not a technically correct assumption. Therefore, the derivation of the temperature value to be used in the equation, Tj , is shown, using a -definition that allows accommodation of spatial temperature variation through the containment.

13. 'Reportina of Results. A uniform format for reporting Type A, B, and C test results is being encouraged in order to make better use of the data history being generated.

14. Flow Rate (Air, Water, Nitrogen). a., b. This recommendation is being made to avoid the use of an air discharge test method since there are many inherent inaccuracies in trying to capture and measure discharged air, l

e.g., leak paths from the tested volume other than that being metered.

, 15. Water Collection. The best determinant of leakage rate, whether the fluid used is air or water, is to measure the makeup required to maintain test pressure. In addition, certain piping configurations will not permit the collection of valve (water) leakage thus requiring that makeup be monitored.

16. Vacuum Retention. Pressure buildup or vacuum decay is not considered i to be an equivalent alternative to measuring a flow rate due to the added com-plexity of varying parameters (temperature, etc.) and loss of accuracy as the 4

pressure differential disappears. Also, the formula in the standard does not t provide the same results for positive or negative pressures.

i 17. Recordina of Leakage Rates. Clarification is provided for a concern I on packing leakage that has had to be clarified in the past.

18. Containment Atmosphere Stabilization, 19. Sensor Stability, 20.

Data Recording and Analysis. Use of these supplementary recommendations with

+

ANSI /ANS 5.8-1981 will allow discontinuing the use in licensing reviews of Bechtel Topical Report BN-TOP-1, " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants," Revision 1, November 7, 1972. It should be noted that these recommendations eliminate the requirement for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> periodic test. The preoperational test is still in-tended to be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be available as a baseline test. Position 20 -

presents a method of data analysis acceptable to the NRC staff for controlling the quality of the data obtained during the Type A and verification tests, and ,

determining test acceptability.

1 Draft 12/27/85 5 RG DIV TASK MS 021-5

_ _ _ .-.__- ~ _--. _ _ - _ . - _ . _ _ _ _ . . _ _ _ -_ _

f This method (Position 20.d) establishes additional conditions on the quality of the regression fit _obtained using the method of ANSI /ANS 56.8-1981.

Condition (a) represents a limit on the deviation from straight line behavior permitted in the data. Condition (b) is essentially a condition on the minimum test duration as a function of the observed scatter in the data about the regression line. This second condition is analogous to the requirement that ISG 5 0.25 L, but applied to data scatter rather than instrumentation errors.

Condition (a.1) is a standard statistical test used to measure if a higher order term in the regression analysis is warranted. When this condition is satisfied, the data is better fit by a parabola than by a straight line at the 95% confidence level. The right hand side of the inequality is a parametri- }

zation of the 95th percentile F distribution with one degree of freedom in the numerator and N-3 degrees of freedom in the denominator. Condition (a.2) is a limit on the magnitude of the ratio of the average time varying leak rate _

component to the constant component. This ratio is estimated by fitting the data with a parabola.

The left hand side of the inequality of Condition (b) is the coefficient of determination (the square of the correlation coefficient). The corresponding limit on the right hand side is derived using the following considerations:

the standard deviation of the data scatter about the regression line is compared to the estimate made of the instrumentation errors, the resultant chi-square is allowed to vary up to the parametrized 95 percentile of the chi-square distribution with N-2 degrees of freedom, and the condition is imposed that the ISG $ 0.25 L,.

This method does not change the way the leakage rate or upper confidence level are calculated. It imposes two additional conditions on the data behavior, and puts limits on how much data scatter and nonlinearity are acceptable.

6 RG DIV TASK MS 021-5 Draft 12/27/85

. _ _ . . . . . . _ . . . . . I

C. REGULATORY POSITION The procedures, requirements, measurements, and analytical techniques recommended by American National Standard ANSI /ANS 56.8-1981, " Containment-System Leakage Testing Requirements," together with its Appendices are generally acceptable to the NRC staff. They provide an adequate basis for complying with the Commission's regulations with regard to the leakage testing of contain-ment systems, subject to the following:

1. Conflict. If any provisions of the standard conflict with the require-ments of Appendix J to 10 CFR Part 50, the requirements of Appendix J govern.

2. Type A Test Requirement. Paragraphs 3.2.1.2, 3.2.1.3, and 3.2.6(a) should be supplemented by:

"The leakages shall be based on the minimum pathway and shall include instrumentation system error."

3. Pressurizina Considerations. In paragraph 3.2.1.7 (page 4) use the i

following instead of the second sentence:

"All possible sources of gas leakage (such as air, nitrogen, steam) into the containment from any system..."

4. . Liquid Level Monitoring. In paragraph 3.2.1.8 {page 4) the second paragraph should not be used.

5. Type A Test Freauency. Paragraph 3.2.3 (page 4) should be supplemented by:

"If the test interval ends while primary containment integrity is not required, the test interval may be extended provided a successful Type A test is completed prior to the plant requiring containment integrity."

6. Verification Test. a. Paragraph 3.2.6(b) (page 5) should contain the following additional material:

"(1). The purpose of the verification test is to verify the ability of the Type A test instrumentation to detect leakage rates approaching L,.

(2). The verification test should measure a change in the leakage rate or a change in the mass. However, a "one point check" is insufficient, and sufficient points should be used to establish a continuous definitive line slope extension following directly from the Type A test line plot.

7 RG DIV TASK MS 021-5 Draft 12/27/85 l

-(3). The start time for the verification test should be as soon as possible - .

following each Type A test.

(4). Data acquisition should not be interrupted without justification from the end of the successful Type A test to the start of the verification test. In some cases, this period of time could be several hours and should then be considered to be part of the Type A ,

test. Data acquisition, of course, should also not be interrupted without justification from the start to the finish of the verification i . test.

(5). The Type A test leakage rate to be used for comparison with the

~

verification test is that of the last valid Type A data point."

b. In Paragraph 3.2.6.(b)(2) (page 5), th'e method described is acceptable

! only if it is supplemented by a requirement for a statistically adequate number used in the equation-of air mass measurement data sets for the measurement of W2 t for the step-change verification test and if the formula result is < 0. p, _

where t is the time required to pump daily allowable leakage at the rate being p

pumped.

, 7. Type B and C Test Pressures. In paragraph 3.3.2 (page 6) the following l should be used in place of the last sentence of its first paragraph:

" Substituting a vacuum test for pressurization to P ac shall be permitted as long as the differential pressure across the item under test is at least pac."

8. Type B and C Test Schedule. Paragraph 3.3.4 (a) (page 6) should be -

supplemented by the following sentence:

"If the two- year interval ends while primary containment integrity is not required, the test interval may be extended provided all de-ferred testing is successfully completed prior to the plant requiring containment integrity."

9. Test Medium and Water Filled Systems. In 3.3.5(b) (page 6) the fol-lowing should be used in addition to the first sentence, after "... subsequent to an accident...", and in 6.4 (page 14) after "...for at least 30 aays...":

" ... assuming a single active failure in the affected system..."

10. Calibration. a. In paragraph 4.2.2 (page 7) the following should be used in place of the first sentence:

" Instrumentation used for Type A, B, or C tests shall be individually calibrated no more than six months prior to use. Primary test..."

. Draft 12/27/85 8 RG DIV TASK MS 021-5

t

{

1

b. In paragraph 4.2.3.2 (page 8) the following should be used in place of the last sentence:

"The check shall be done not more than one month prior to the ILRT."

c 7 In paragraph 4.2,.4 (page 8) the words " calibration checks," " checks,"

and "chicked," should not be used, and the words " calibration," calibrati$ns,"

or " calibrated" used in their place as appropriate. ,

. 11. Temperature Measurement. 3

a. Paragraph 5.5 (page 11) should be supplemented by:

"5.5.3 Volume Fractions. The temperature and humidity pre-test surveys shall be used to confirm assigned volume fractions. The calculated volume fractions combined with the placement of the sensors will ensure that each sensor represents the assigned volume.

The sum of the volume fractions is unity. If during the conduct of a test, a sensor fails or is deleted, that sensor's volume fraction i

should be re-assigned to'other sensors on the basis of the survey ,

results. For , subsequent tests the volume fractions shall be reviewed to determine their continued validity."  ;

b. In paragraph 5.5.1 the following should be used instiad of the i

existing te'xt: - ,

"... regional variations in temperature with the ventilation configura-tion planned for each test. Fans or other means of air circulation may be used..." t

c. In paragraph 5.5.2 the following should be used instead of the existing text:

"Psychrometric readings should be taken throughout the containment pHor to conducting the preoperational and initial periodic leakage rate tests to determine the correct location of the : ansors for the ventilation configuration plannbd for each '.en.'

4

12. Absolute Test Method. Paragraph 5.7.3 (page 12) should be supple-mented by an explicit definit' ion of the temperature term tc/be used in its equation as being derived from 1

Tg=

n "j m

i

): j=1 T 3

Draft 12/27/85 9 RG DIV TASK MS 021-5

. c, where w is a weighting factor or the percentage of overall containment volume , ,

assigned to a particular temperature sensor and n is the number of temperature sensors used-in the test.

13. Reporting of Results. The format and content of paragraph 5.8 (page 12) should be used for the submittal of_ reports required by Appendix J to 10 CFR Part 50, including the individual Type A, B, and C "as found" and "as left" leakage readings required by 10 CFR Part 50, Appendix J.

14. Flow Rate (Air, Water, Nitrocen). a. In paragraph 6.5.2 (page 14) the following should be used in place of the second sentence:

" Makeup fluid to the test volume required to maintain test pressure [er

, air-discharge-from-the-test-volume] shall be the same type or less viscous fluid as the test fluid and shall be measured using a flowmeter that directly measures valve leakage rate."

b. Paragraph 6.5.2 should be supplemented by:

"The air discharge method shall not be used."

15. Water Collection. Paragraph 6.5.3. should be supplemented by: .

"Where it is uncertain that all water is being collected, the water i

makeup to the test volume, provided to maintain test pressure, shall be measured in accordance with Section 6.5.2."

16. Vacuum Retention. In paragraph 6.5.4 (page 15), the following should not be used:

" Alternatively, a pressure (vacuum) gauge may be used to measure pressure buildup (vacuum decay) versus time," and the remainder of 6.5.4 fc11owing that sentence.

17. Recording of Leakage Rates. Paragraph 6.6 (page 15) should be supple-mented by:

"Any packing leakage that provides a leakage path outside the primary reactor containment shall be accounted for in Type C test results, but need not be quantified separately. Packing leakage that remains within the containecnt system need not be included in the Type C test results."

18. Containment Atmosphere Stabilization. The following supplementary recommendations should be used as noted:

"a. The upper confidence level of containment leakage should be equal to or greater than zero prior to declaring the start of the test.

(Supplements 5.2.1) ,

Draft 12/27/85 10 RG DIV TASK MS 021-5 i

. . -,, - ----- -..,,,..- , , - , - , . , +-- - , , , ,

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b. Containment air temperature is stabilized when:

(i) the slope of the temperature vs time curve is less than 0.5'F/

hour averaged over the last two hours, and (ii) the rate of change of the slope of the temperature vs time curve is less than 0.5'F/ hour / hour averaged over the last two hours. (Supplements 5.3.1.3)

c. Containment air temperature should remain stabilized over the entire test period, including the verification test, and the tests should be continued only so long as the temperature is stablized. If the temperature does not appear to be stable due to a problem with the

~

test procedures (such as a mechanical error / failure), rather than due to leakage, the test may be continued if the problem has been

. identified and corrected. (Supplements 5.3.1)"

19. Sensor Stability. Paragraph 3.2.6(c) (page 5,^1ast paragraph) should be supplemented by:

"For sensors rejected or suspected of being faulty, data should be recorded for the duration of the test (Type A test plus verification test) so that it is available if needed for post-test evaluation."

20. Data Recording and Analysis. The .following supplementary recommenda-tions should be used as noted:

"a. The start time of the containment integrated leak rate test should be declared following a determination that test conditions have stabilized and is not subject to change during or after data collection. If any test is restarted, the restart time should be selected as " time forward" not as " time backward." (Supplements 5.4)

b. Instantaneous (unaveraged) sensor readings should be recorded at approximately equal intervals but in no case at intervals greater than one hour. (Supplements 5.6)

c. The minimum test duration after the containment atmosphere and instrument readings have stabilized should be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a preoperational Type A test and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for periodic Type A tests.

(Supplements 5.4)

d. Additional criteria for data reduction are presented. These methods should be used to limit data scatter and nonlinearity when an integrated leak rate test is conducted. Use of alternative criteria would also be considered if demonstrated to be adequate.

i I

Draft 12/27/85 11 RG DIV TASK MS 021-5

._~ --- . - _ . - _ - , - - . - . - . - - . _ _ -

I l

ANSI Extended Criteria . .

I For data generated during the Type A and verification tests, the mass point method of ANSI /ANS 56.8 is utilized. In addition, the following two conditions must be met at the end of the test, at time t:

Condition (a)

(B'-B)IWj +(A'-A)IWgg t +C'IWg t 2

3.8414 (n 2-5.3n+8.0394)

(n-3) > (a.1)

IW9 2.B'IW9 -A'IWgjt -C'IW9 tj (n2-7.7098n+14.9069)

If condition (a.1) is satisfied, then condition (a.2) applies.

C't < 0.25 (a.2)

A' The notation of ANSI /ANS-56.8-1981 is used in these expressions. In addi-tion A', B', and C' are the solutions to the equations.

IW g = B'n + A' Itg+C'Itj IWg t, = B'Itg + A' It2+C'Itj IWj t2 = B'It2+A'Itj+C'Itj and are the coefficients for the least-squares parabola:

W = B' + A't + C'tz, The left side of (a.1) can also be written as: ,

2 I (W9 - A - Btg )2 - I (W - A' - B' tg - C't )2 g

(""3)

I (W9 - A' - B' tg - C't 2)2 g

and the right side of (a.1) is an approximation to the values of the F distri-bution corresponding to the 5% probability of exceeding F with 1 degree of Draft 12/27/85 12 RG DIV TASK MS 021-5

.s ..

. c, f'eedom r in the numerator and n-3 degrees of freedom in the denominator. This can then be recognized as a test used to measure if a parabolic term in the regression analysis is warranted.

The ancillary condition (a.2) sets a limit of 25% on the ratio of the average time varying leak rate component to the constant component. This ratio is estimated by taking the ti:ne derivative of the equation for the

'least-squares parabola and averaging it over the test duration.

Condition (b)

[n(IWgyt ) - (It )(IW g g)]2 L$m I(tg - f.)2

> (b.1)

[n(Itj)-(It )2][n(IW $ y2)-(Iw j)2] L2 am I(t -t)2 j + Ljt2(n-2)(n+1.33)(n+42.603) 112.87 (n-l.202)(n+28.155)

The left side of (b.1) is the coefficient of determination (or the square of the correlation coefficient) and has the equivalent expression:

I(hg - E)2 I (W - hj )2 + I(h 4 g - )2

^

Where W9 = B + Atj and 9 = IWg = B + A Itj=B+At n n Using the above and the fact that I(Wj -h)2=S2(n-2)thefollowing j expression can be shown to be equivalent to (b.1):

l a (n + 1.33) (n + 42.603) (b.2)

({S)2 I (L,,)2112.87 (n - 1.202) (n + 28.155)

This expression was derived by comparing20 , or the variance of the parent population (of which S2 is an estimate), to the instrumentation errors which are presumably the source. These errors are also used in computing ISG. For the sake of the derivation, the assumption was made that the instrumentat' ion errors Draft 12/27/85 13 RG DIV TASK MS 021-5

used in computing ISG are normally distributed and are at the 95% confidence , ,

level. This introduces a factor of 1.96.

e e e 32 (1.960)2=eg=Wj{(y)2+ ( V)2 + (q )

or 62 = j ( (b.3) 6 2400) where W j has been replaced by B.

To account for the fact that 52 is only an estimate of the variance of the parent population and this estimate improves with increasing number of data points the probability distribution of the statistic called the reduced chi-square was used. With 95% probability the following holds:

S2 p 5 1.08916 (n (b.4)

(n + 1.33) (n ++ 28.155)

- 1.202)(n 42.603)

The left side of the expression is called the reduced chi-square. The right side is an approximation to the values of the reduced chi-square corre-sponding to the 5% probability of exceeding chi-square with n-2 degrees of freedom.

Finally, expression (b.2) was derived by using the above expressions (b.3) and (b.4) and imposing the condition that ISG 1 0.25 L,.

Draft 12/27/85 14 RG DIV TASK MS 021-5

D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.

This draft guide has been released to encourage public participation in its development. Except in those cases in which an applicant proposes an acceptable alternative method, the method to be described in the active guide reflecting public comments will be used by the NRC staff in evaluating procedures for containment system leakage testing for compliance with the amended Appendix J to 10 CFR Part 50 if the amendments are promulgated as

' proposed in FR .

Draft 12/27/85 15 RG DIV TASK MS 021-5

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E ENCLOSURE 6 6

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4 DRAFT PUBLIC ANNOUNCEMENT e

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Encl. G

DRAFT i

NRC PROPOSES CHANGES .

TO CONTAINMENT LEAKAGE RATE TEST RULES The Nuclear Regulatory Commission is proposing to amend its regulations dealing with the leakage rate testing of comercial power reactor containment systems.

The proposed changes result from: experience in applying the existing requirements; advances in containment leakage testing methods; interpretations of the existing requirements made over time; simplification of the present text; application of alternative requirements reflecting variations in power reactor design; coments made on the existing

^

requirements over time; and requests for exemptions from the requirements received and approved over the years since the requirements went into effect in March 1973.

- . As proposed, the major changes would:

. (1) Make the containment system leakage criteria more general than specific;

- (2) Make editorial changes for improved clarity; p

L (3) Make provisions for consideration of alternative leakage test '

.\

requirements when necessary; i

(4) Make changes to resolve past questions of interpretation; (5) Eliminate an option to perform periodic reduced pressure testing in i lieu of testing at full calculated accident pressure; (6) Revise test frequency requirements; (7) Delete a requirement governing duration of testing;

> C3 (8) Put new emphasis on the requirement that containments be tested "as - .

is."

(9) Standardize reporting content and format.

(10) Focus test program on local, rather than overall leakage barriers, based on over a decade of test experience.

(11) Require that a corrective action plan be submitted when a reportable problem is identified; and (12) Require that an implementation schedule be submitted by a given date.

In addition to general comments on the proposed revisions to Appendix J of Part 50 of its regulations, the Commission is seeking connents on specific

. questions set forth in the Federal Register notice. A draft of a proposed regulatory guide endorsing a national standard on the same subject is being made available for public comment at the same time. Written comments should be submitted by (date). They should be addressed to the Secretary of the Commission, Nuclear Regulatory Commission, Washington, DC, 20555, Attention:

Docketing and Service Branch.

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EflCLOSURE 7 3

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• CJ e a CONGRESSIONAL LETTERlS)

N*emo-%,, ,.,,

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The Honorable Alan Simpson, Chairman Subconunittee on Nuclear Regulation .

Comittee on Environment and Public Works United States Senate Washington, DC 20510

Dear Mr. Chairman:

Enclosed for the infonnation of the Subcomittee are copies of a Notice of Proposed Rulemaking to be published in the FEDERAL REGISTER and a Notice of Availability for a related draft regulatory guide.

I

~

The amendment of 10 CFR Part 50 comprises a revision of Appendix J.

~ ~ ~ ~ ~ ~ ~ ~

" Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." It provides an extensive updating of the 1973 regulation, and reflects changes resulting from experience in applying the requirements, the issue of a national standard on procedures for such testing, interpretive questions, and exemption requests received and approved.

- The draft regulatory guide endorses National Standard ANSI /ANS 56.8,

" Containment System Leakage Testing Requirements," and provides guidance on procedures acceptable to the NRC staff for conducting leakage tests.

- Also enclosed are copies of a draft public announcement to be issued on this matter in the next few days.

Sincerely.

Robert B. Minogue Director Office of Nuclear Regulatory Research

Enclosures:

1. Federal Register Notice Notice of Regulatory Guide

2. l Availability

3. Draft Public Announcement l cc: Sen. Gary Hart DRAFT:GARNDT:fkm($1MPS,0N)

The Honorable Alan Simpson, Chairman Subcommittee on Nuclear Regulation Committee on Environment and Public Works United States Senate Washington, DC 20510 cc: Sen. Gary Hart The Honorable Morris K. Udall, Chairman Subcommittee on Energy and the Environment Committee on Interior and Insular Affairs United States House of Representatives Washington, DC 20510 cc: Rep. Manual Lujan The Henorable Richard L. Ottinger, Chairman Subcommittee on Energy. Conservation, and Power Committee on Energy and Commerce United States House of Representatives Washington, DC 20510 cc: Rep. Carlos Moorhead O

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ENCLOSURE 8 O

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fa w - 40 h

1 6 a ACRS 7/11/85 LETTER i

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? /g. ** "0g'g UNITED STATES e NUCLEAR REGULATORY COMMISSION ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS g- l mamwavow.o. c.nossa June 11, 1985

+.,,,,

Mr. William J. Dircks Executive Director for Operations U..$. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON THE PROPOSED REVISIONS TO APPENDIX J OF 10 CFR 50 AND THE RELATED REGULATORY GUIDE During its 302nd meeting, June 6-8, 1985, the ACRS considered the proposed revisions to Appendix J of 10 CFR Part 50, " Leak Tests' for Primary and Secondary Containments of Light-Water-Cooled (Task Nuclear Power No. MS 021-5),

Plants," and an associated draft Re

" Containment System Leakage WeTesting.gulatory Guide recomend that these documents be issued for public comment at this time.

Although we believe that the basis for limitini, leakage from contain-ments and the requirements for leak testing need to be reconsidered in the light of research on severe accidents and source terms, we believe that the development of significant and well-considered changes to Appendix J will require several years, and thus believe that publication of the currently proposed revisions need not be delayed.

~

We believe that the public coments on these proposals will provide a better basis for deciding whether this interim revision to the

~

regulations will indeed be a worthwhile and cost-effective improvement.

We wish to again review these proposed revisions once public coments have been received and responses to them have been prepared by the NRC Staff.

Sincerely, i ,I h David A. Ward

• Chairman

References:

1. Proposed Revisions to Appendix J of 10 CFR Part 50, " Leak Tests for Primary and Secondary Containments of Light-Water-Cooled Nuclear Power Plants," Draft E4, dated April 17, 1985

2. Proposed Regulatory Guide (Task No. MS 021-5), " Containment System Leakage Testing," Draft D4, dated May 31, 1985 cc: See page 2 ype --- O m 16

8 '

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June 11, 1985 <

Mr, William J. Dircks cc: 5. Chilk, SECY J. Roe, EDO R. Minogue, RES C. Bartlett, RES G. Arlotto, RES G. Arndt, RES R. Hernan, NRR 1

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ENCLOSURE 9 i

I 59-W "N ' * + +ey@ e- eh, G

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5 .s.

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. o 150.109 BACKFIT ANALYSIS FOR ,

. PROPOSED 10 CFR 50, APP. J AND PROPOSED RG MS 021-5 l

e n . _ _ . . .

As, ~s i. .

.J BACKFIT ANALYSIS AND CONCLUSION RELATING TO THE PROPOSED REVISION TO 10 CFR PART 50, APPENDIX J AND IT5 COMPANION REGULATORY GUIDE

~

10 CFR Part'50, Section 50.109, states that the Commission shall require a systematic and documented analysis pursuant to paragraph (c) of this same section for backfits which it seeks to impose.

This revision of 10 CFR 50, Appendix J is not being proposed by the NRC staff on the basis of any substantial increase in safety or decrease in costs.

Instead, it is being proposed as both safety and cost neutral. Justification for the revision is based on the need to conform present testing capabilities to the current state of the art, and to use the best available procedures, l

. ,thereby not freezing a stale (1972) technology. The revision will keep rule requirements unambiguous, technically current, uniform in application and usefulness, legally consistent, and flexible enough to accommodate differing plant designs.

l The following discussion and 150.109(c)analysisdescribehowtheseaspects, and the substantive elements of the backfit rule have been addressed in the review and oversight process that all rules and regulatory guides must go Lthrough prior to issue for public comment. Justifications for undertaking and completing such activities must be continually made throughout the development process. As a result, all of the issues and elements of interest under $50.109 have been scrutinized by a variety of reviewing bodies, and in public meetings, i

The conclusion presented is one believed to be supported through these previous reviews.

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The proposed rule is intended to be applied to the entire population of nuclear power reactors and it clearly constitutes a backfit.

Prior to the effective date of the backfit rule and its application to the rulemaking process, the NRC staff presented this proposed rulemaking activity, including its contents and the justification therefor, to the ACRS and the CRGR. After review and discussion of the proposed rulemaking activity, its relationship to other NRC activities related to containment integrity, a value-impact study, and related justifications for this updating activity, these review bodies recomended in favor of issuing the proposed rule revisions andcompanionregulatoryguide(MS021-5)forpubliccomment.

The regulatory analysis written for this proposed revision was considered by the ACRS and CRGR review bodies, and also placed on file in the Public Document Room. Included in this regulatory analysis package was a cost analysis by Science & Engineering Associates, Inc.; Mathtec Inc.; and S. Cohen &

Associates,..nc.

Tables 1.3 and 1.4 in the cost analysis estimated that the Appendix J revision can result in a potential total cost saving ranging from about $98 million (9 10% discount rate) to $164 million (9 5% discount rate) but with a potential increase in routine occupational exposure on the order of 10,000 person-rem over the assumed operating life of all existing and planned power reactors.

This projected increase in occupational exposures would on average equate to less than four person-rem per reactor year. It should be noted that current occupational exposure levels average annual collective doses of 753 person-rem per reactor year.

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o The analysis projected total costs to the NRC on the order of $4 million (@

10%) to $5 million (@ 5%), principally due to increased manpower efforts asscciated with technical specification revisions. Of this, about $3 million would be incurred over the next few years during implementation. The remainder represents the present worth of all NRC costs incurred over the operating life of the reactor population.

Implementation costs to the nuclear industry of about $4 million (@ 10% & 5%)

were projected due to preparation of technical specification changes minus the projected savings associated with reduced exemption requests necessitated by the current regulation. The major industry benefit would occur during the operating life of the power reactor population where present worth savings on the order of $106 million (@l0%) to $173 million (@ 5%) were projected.

Although the cost analysis also identified increased operating cosh, these costs would be outweighed by significant savings in replacement energy costs.

Savings in replacement energy costs would result because several of the changes to Appendix J will reduce the expected frequency of containment integrated leakage rate (Type A) tests. These tests currently require 3 to 5 days of reactor downtime per test.

A 10,000 person-rem increase in routine occupational exposure was estimated over the operating life of the power reactor population primarily due to an assumed increase in maintenance efforts for implementing Corrective Action Plans and in the industry's ability to substitute local penetration and valve (Type B and Type C) tests for Type A tests. On a per reactor-year basis, this represents an average projected increase in occupational exposure of approximately 0.4% relative to the 753 person-rem average from all other causes apart from Appendix J.

The analysis of the costs and benefits for the proposed Appendix J revision indicated a significantly favorable net cost benefit for the action when all tradeoffs and factors such as replacement energy savings are considered.

However, the NRC staff is aware that it may not be appropriate to factor the economic benefits of avoiding penalty replacement energy savings into its regulatory safety decision process. The NRC staff is therefore not factoring these particular savings into its conclusions regarding benefits and costs.

Hcwever, the NRC staff firmly believes that there exist regulatory and industry advantages that accrue from use of technically sourd and unambiguous regulations that minimize the need for exemptions. Therefore, even if the favorable economic benefits to industry are minimized in the balancing of the everall costs ard safety benefits involved, the staff estimates that, at worst, this revision should be considered neutral in its cost and safety effects.

The proposed revision of Appendix J includes the following considerations:

o This proposed revision of Appendix J is an administrative update due to changes in practice and replacement of a referenced ANSI standard. The revised regulation will provide general test criteria for testing leakage characteristics of the post-LOCA containment configuration. It will also standardize reporting requirements. The test method is basically the statistical evaluation of multiple pressure, temperature, and humidity readings needed to quantify a very small leakage rate from a very large volume. For example, a 0.1% per day leakage rate out of a congainment volume of 2,000,000 cu. ft. under a pressure of 55 psia at 150 F is 2

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roughly equivalent to that represented by a hole with a diameter of about 1/16 inch. The actual allowable leakage rate is defined for each plant in its technical specifications, based on analyses conducted pursuant to 10 CFR Part 100, whereas Appendix J establishes the criteria and tests to be used to verify the achievement of technical specification limits on leakage.

Relaxing to some degree the current leakage limits (if these are found to be overly restrictive through ongoing source term and risk profiling studies) would necessitate change to existing plant technical specifications and perhaps cause revision to the ANSI /ANS 56.8 standard that controls data error bands, instrument sensitivity, and test duration.

It would be unlikely to cause another significant revision to Appendix J, so long as the general test criteria contained in this proposed revision would not be affected. This should enhance the stability of this regulation, and allow greater flexibility for acceptance of alternative leak-test reqairements to accommodate variations in containment systems designs.

o The current leakage limits established by NRR for plant-specific siting are based on analyses pursuant to 10 CFR Part 100. These current leakage '

limits are expected to remain unchanged under this proposed Appendix J revision.

o Discussions between NRC staff, nuclear industry representatives, and ,

j professional and standards groups indicate that Appendix J to 10 CFR Part )

50 needs to be revised to update the criteria, clarify questions of l' interpretation, and delete references to an obsolete ANSI standard on l

leakage rate testing of containments of light-water-cooled nuclear power plants.

o This proposed revision of Appendix J would provide greater flexibility in applying alternative leakage test requirements taking into account the variations in plant design. It also reflects experience in appl}ing l

l existing requirements, advances in containment leak testing methods, and '

multiple requests (since 1973) for exemptions.

l o As proposed, Appendix J contains only the eneral requirements and acceptance criteria (no testing techniques for preoperational and subsequent periodic leak testing. Prescriptive and detailed testing techniques are not incorporated in this revision. Interested persons will be offered an op)ortunity to coment on specific guidance concerning leakage test metiods, procedures, and analyses that are acceptable to NRC staff to implement these requirements and criteria (draft Regulatory Guide MS021-5).

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Analysis of 50.109(c) Factors 50.109(c)

(1) Statement of the specific objectives that the proposed backfit is designed

' to achieve.

This revision of Appendix J will provide greater flexibility in applying alternative leakage test requirements due to variations in plant design, and reflects changes based on: (1) experience in applying the existing requirements; (2) advances in containment leak testing methods; (3) interpretive questions; (4) simplifying the text; (5) various external / internal comments since 1973; and (6) exemption requests received and approved. There is also the need to conform present testing capabilities to the current state of the art and to use the best available procedures, thereby not freezing a stale (1972) technology. The revision will keep rule requirements unambiguous, current, useful, consistent with practice, and flexible enough to accommodate differing plant designs.

Also, the publication of an expanded and updated national standard =on how to conduct such tests has now made it appropriate to generalize the, regulation by retaining test criteria and removing prescriptive testing details better left to the national standard.

(2) General description of the activity that would be required by the licensee or applicant in order to complete the backfit.

This action will require changes to the technical specifications, test procedures, data analyses, and test reports. In some cases it may entail modification of some systems to conform to all aspects of the revised leakage testing program, such as test taps to enable testing of some valve (s) not previously tested. With such minor exceptions, the activities required for compliance will be administrative and procedural, rather than physical or hardware changes. For plants that have been doing Type A tests at reduced pressure, an additional 3-10 hours pumping time may be needed when testing at full pressure. Those plants not reporting "as found" leakage results will be explicitly required to do so.

Licensees will have to review plant test procedures against the revised requirements and recomn.endations. This will determine the extent of changes needed to the technical specifications. Following this evaluation, licensees will submit to the NRC staff an implementation schedule for conforming to the new requirements. This schedule will take into account where the plant is in its testing timetable and the amount of work needed to change procedures, tech specs, etc.

I (3) Potential change in the risk to the public from the accidental off-site release of radioactive material.

Studies have indicated that containment systems of today's plants are strong and reliable against leakage of radioactivity for a spectrum of

postulated design basis accidents including the presence of large amounts i of radioactivity as is traditionally assumed for analyses pursudnt to 10 CFR Part 100. This reliability against leakage has been brought about oy i

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! NRC design requirements and use of industry codes and standards. The requirement to periodically test the containment system (Appendix J) is also an important way of assuring that this leaktight integrity is maintained over the plant's lifetime. The proposed revision to Appendix J is expected to continue this assurance of leaktight integrity of the containment system. However, experience over the past decade (since 1973) has revealed that the more likely leakage paths exist through penetrations and valves. Therefore, more focus is provided on penetrations and valve (Type B & C) leakage tests. This improved test focus is difficult to quantify because the available data from containment systems testing already indicates a high reliability for low leakage. Substantial safety benefits have derived from the existence of Appendix J itself. The proposed update and revision will at least continue these benefits, but will also produce greater confidence in the value of the test results, and do so, at worst, on an overall cost-neutral basis.

(4) Potential impact on radiological exposure of facility employees.

The changes to Appendix J are estimated to result in higher occupational radiation exposures than are presently experienced. The more frequent testing of individual containment penetrations does require additional time inside containment for test crews, resulting in higher occupational exposures. Data and derivations are provided in the Appendix to NUREG/CR-4398, " Cost Analysis of Revisions to 10 CFR Part 50, Appendix J, Leak Tests for Primary and Secondary Containment of Light-Water-Cooled Nuclear Power Plants." From these, average industry increases are about 3.0 person-rem per plant per year of operation. The high estimate is 5.6 .

person-rem per plant per year, and the low 0.5 perstn-rem. This com with an average annual collective dose of 753 person-rem per plant (paresfrom NUREG 0713, Vol. 5 " Occupational Radiation Exposure at Nuclear Power Reactors," 1983), and represents an average potential increase of 0.4%.

(5) Installation and continuing costs associated with the backfit, including the cost of facility downtime or the cost of construction delay.

A comprehensive cost analysis (NUREG/CR-4398) has been performed that indicates significant potential cost savings to the industry and public.

These have been estimated for the remaining life of all water-cooled nuclear power plants in this country, in operation or under construction, as ranging from $106 million to $173 million. Industry implementation costs are estimated to be about $3 million to $4 million, due to revision of technical specifications less savings associated with reduced exemption requests.

Although the cost analysis estimated large potential savings, the NRC staff has conservatively viewed the impact of this revision as cost-neutral on an industry-wide basis. This is because the savings are mostly replacement power costs for extra penalty Type A tests that could be avoided by changes proposed in the revision. However, these costs could also be viewed as currently avoidable for licensees that are maintaining their containment systems within technical specification leakage limits.

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.(6) The potential safety impact of changes in plant or' operational complexity, including the relationship to proposed and existing regulatory- ,

requirements.

As an updated inservice inspection program, no significant, quan.tifiable change is claimed to safety other than to occupational exposures, as previously noted. However, in return there will be indirect benefits of greater confidence in the reliability of the test results, better and more uniform tests and test reports, fewer exemption requests, and fewer interpretive debates. No changes in plant or operational complexity are foreseen. There is also no impact on other regulatory requirements.

(7) The estimated resource burden on the NRC associated with the proposed backfit and the availability of such resources.'

For the total population of all water-cooled power plants in this country, the estimated NRC resource burden is about $3 - 4 million for implementation and $1 million for operation over their remaining life.

This is due principally to increased manpower efforts associated with technical specification revisiens. The resources necessary to accomplish these tasks have been considered in the NRC budget.

(8) The potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit.

Uniformity in requirements, implementation, and reporting is being sought by the proposed rule revision. Although plants of different design and vintage are involved, it is believed that the net impact will not vary significantly. Major older (pre-Appendix J) designs plant problems have with been the existing handled rule that are unique to by granting exemptions where justified. Such exemptions, where still needed, will remain in force. NUREG/CR-4398 notes.that the net impact is not expected to vary significantly between BWR's and PWR's.

(9) Whether the proposed backfit is interim or final and, if interim, the justification for imposing the proposed backfit on the interim basis.

This proposed revision to Appendix J and its associated backfit will be issued, after the public coment period, as final, based on current regulatory approaches. Meanwhile, broader and more fundamental aspects of containment function will be subjected to review. These reviews could eventually result in future changes to Appendix J, but they are still some years away, and an immediate need exists to update Appendix J. Any resulting future changes to existing regulations would be made through normal rulemaking procedures, including ACRS review and public coment.

150.109(a)(3) CONCLUS10N There is no substantial increase in the overall protection of the public health and safety or the comon defense and security that can presently be quantified from the proposed backfit. 'However, the direct and indirect costs of.

implementation are justified due to better, more uniform tests and test repcrts, greater confidence in the reliability of the test results, fewer exemption requests, and fewer interpretive debates. For the benefit of the public, licensees, and the NRC staff, this proposed rule should be issued at this time for public comment.

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