ML20214W098

Forwards Listed Documents in Preparation for Publication of Notice of Proposed Rulemaking & Notice of Availability for Reg Guide,Task Ms 021-5 in Fr Re Containment Leakage Rate Testing Set Out in SECY-86-167

ML20214W098
Person / Time
Issue date:	10/16/1986
From:	Arlotto G NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Philips J NRC OFFICE OF ADMINISTRATION (ADM)
Shared Package
ML20214W102	List: ML20214W098 ML20214W140
References
RTR-REGGD-01.XXX, RTR-REGGD-1.XXX, TASK-MS-021-5, TASK-MS-21-5, TASK-RE NUDOCS 8612100081
Download: ML20214W098 (127)

Text

001 1 6 1986 MEMORANDUM FOR: John Philips, Chief Rules and Procedures Branch Division of Rules and Records Office of Administration FROM: Guy A. Arlotto, Director, Division of Engineering Safety Office of Nuclear Pegulatory Research

SUBJECT:

PP0 POSED REVISION TO 10 CFR 50, APPENDIX J & PROPOSED FEGULATORY GUIDE MS 021-5

~ By memorandum of September 18, 1986, the Secretary informed the NPC staff that the Comission (with all Comissioners agreeing) has approved publication of the proposed rule and regulatory guide on containment leakage rate testing set out in SECY 86-167, subject to the Comission coments as forwarded by the Secretary. These comments have been incorporated as shown in Enclosure 2.

Please arrange for publication of the notice of proposed rulemaking and the notice of availability for the regulatory guide in the Federal Register, allowing 90 days for public comment on each of them. If possible, arrange to have the rule and the guide published on the same day.

Enclosed are:

1. Federal Register notice for the proposed rule for submittal to the OFR for publication in the Federal Register (original + 6 copies).

2. Marked-up FPN with Comission changes for forwarding to SECY (1 copy).

3. Congressional letter and public announcement for the Chairman and ranking minority member of each oversight comittee for OCA (6 copies).

4. Draft public announcement for PA (2 copies).

5. Environmental Assessment and Finding of No Significant Impact for filing in the PDR (2 copies).

l 6. Backfit Analysis for filing in the PDR (2 copies).

I 7. Federal Register Notice of Availability for MS 021-5 together with regulatory guide MS 021-5 for submittal to the OFR for publication in the

, Federal Register (original + 6 copies of FRN; 2 copies of guide).

l l Note that a regulatory analysis was filed in the PDR as Enclosure 3 to the May 20, 1985 G. A. Arlotto memo to R. F. Fraley, ACRS (PDR Accession

1. 8506070342).

Finally, please request an extra r.eproduction quality copy of each published document from the OFR to facilitate copying for additional NRC distribution.

Orior.a1 siped by G. A. Arm g 21 g g e61016 Guy A. Arlotto, Director

01-XXX R PDR Division of Engineering Safety Office of Nuclear Regulatory Research

Enclosures:

As stated ',-

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/ OCT 161986 KEMORANDUM FOR: John Philips, Chief Rules and Procedures Branch Division of Rules and Records Office of Administration FROM: Guy A. Arlotto, Director Division of Engineering Safety Office of Nuclear Regulatory Research

SUBJECT:

PROPOSED REVISION TO 10 CFR 50, APPENDIX J & PROPOSED REGULATORY GUIDE MS 021-5 By memorandum of September 18, 1986, the Secretary informed the NRC staff that the Comission (with all Comissioners agreeing) has approved publication of the proposed rule and regulatory guide on containment leakage rate testing set out in 5ftY 86-167, subject to the Comission coments as forwarded by the Secretary. These coments have been incorporated as shown in Enclosure 2.

Please' arrange for publication of the notice of proposed rulemaking and the notice of availability for the regulatory guide in the Federal Register, allowing 90 days for public coment on ecch of them. If possible, arrange to have the rule and the guide published on the same day.

Enclosed are:

1. Federal Register notice for the proposed rule for submittal to the OFR for publication in the Federal Register (original + 6 copies).

2. Marked-up FRN with Comission changes for forwarding to SECY (1 copy).

3. Congressional letter and public announcement for the Chairman and ranking minority member of each oversight comittee for OCA (6 copies).

4. Draft public announcement for PA (2 copies).

5. Environmental Assessment and Finding of No Significant Impact for filing inthePDR(2 copies).

6. Backfit Analysis for filing in the PDR (2 copies).

7. Federal Register Notice of Availability for MS 021-5 together with regulatory guide MS 021-5 for submittal to the OFR for publication in the Federal Register (original + 6 copies of FRN; 2 copies of guide).

Note that a regulatory analysis was filed in the PCR as Enclosure 3 to the May 20, 1985 G. A. Arictto memo to R. F. Fraley, ACRS (PDP Accession

1. 8506070342).

Finally, please request an extra reproductior uality copy of each published dccument from the OFR to facilitate copying r ad tionalgCdistribution.

y A1 Arlotto, Director ivis' n of Engineering Safety Office of Nuclear Regulatory Research

Enclosures:

As stated l l

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ENCLOSURE 1 y . ., , , . - - - ,- - , - . . - - -- -- - - -

8 Draft H September 1986 Federal Register Notice I

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• NUCLEAR REGULATORY COMMISSION 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 testi' ng of containments of light-water-cooled nuclear power plants. The proposed rule would aid the lidensing and en-forcement staff by liminating conflicts, ambiguities, anu a lack of uni-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: Mail written comments to: U.S. Nuclear Regulatory Commission.

Washington, DC 20555, Attention: Docketing and Service Branch. Deliver comments to: Room 1121, 1717 H Street NW., Washington, DC, between 8:15 a.m. and 5:00 p.m. weekdays.

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

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

BACKGROUND SUPPLEMENTARY INFORMATION:

Appendix J of 10 CFR Part 50 was originally issued for public com-ment as a proposed rule on August 27, 1971 (36 FR 17053); published in final fonn on February 14, 1973 (38 FR 4385); and became effective on March 16, 1973. The only amandment to this appendix since, 1973 was a limited one, on Type B (pene' tion) test requirements that was published for coment on January 11,198u (45 FR 2330); published in final fonn 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.

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

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 integrity required, and validation of that integrity under conditions other than postulated in this rule. In order to better understand its 2

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, function and scope, assumptions inherent in Appendix J are presented 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 r 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 l

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.

8. A standard loss-of-coolant accident is assumed as the 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 In addition, pneumatic test pressure for the integrated leak rate test.

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

10. Shielding tests need not be perfonned.

11. Periodic testing provides adequate confidence in the level of g

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

The scope of this revision to Appendix J is limited to corrections and clarifications, and excludes new criteria. He ver, this notice a lso addresses related; broader, longer term activities. Following is informa-tion 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 containment leakage monitoring" (such as negative containment operating pressure) and "relatively frequent gross containment integrity check" (such as a low pressure..pumpup just prior to 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.

Coments 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.

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- 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 NRC regulatory requirements that have marginal importance to safety and to recommend appropriate actions to modify or to eliminate these unneces-October 3, sary requirements. A Federal Register notice was published 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.

INVITATION TO COMMENT l

Comments from all interested persons on all aspects of this revision and on the risk and cost effectiveness of containment leaktightness in

( general are requested by the comment expiration date in order that: (1) the final revision will reflect consideration of all points of view, and 5

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- (2) the staff's assessment of the risk importance of contair, ment leak-tightness can benefit from such coments. Especially requested are com-ments 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) whether present operating plants or plants under review shou' '

be given the opportunity to continue to meet the current Appan-dix J provisions if the proposed rule (reflecting consideration of public coments) becomes effective; (6) if the existing rule or its proposed revision were completely voluntary, how many licensees would adopt either version in its entirety and why; (7) whether (a) all or part of the proposed Appendix J revisions would constitute a "backfit" under the definition of that term in the Commission's Backfit Rule, and'(b) there are parts of the rule which do not constitute backfits, but which would aid the staff, licensees, or both; (8) since the NPC 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 exist-ing regulation; 6

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. (9) the advisability of referencing the testing standard (ANSI /ANS 56.8) in the regulatory guide (MS 021-5) instead of in the text of Appendix J; (10) the value of collecting data for the "as found" condition of valves and seals and the need for acceptance criteria for this condition; (11) whether the technical specification limits on allowable contain-

- ment leakage should be relaxed and if so, to what extent and why, or if not, why not; (12) what risk-important factors influence containment performance under severe accident conditions, to what degree these factors are cons'idered in the current containment testing requirements, and what approaches should be considered in addressing factors not presently covered; (13) 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; (14) what effect " leak-before-break" assumptions could have on the leakage rate test program. Current accident assumptions use instantaneous complete breaks in piping systems, resulting in f

a test program based on pneumatic testing of vented, drained lines. " Leak-before-break" assumptions presume that pipes will fail more gradually, leaking rather than instantly emptying.

(15) how to effectively adjust Type A test results to reflect indi-vidual Type B and C test results obtained from inspections, repairs, adjustments, or replacements of penetrations and valves I

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in the years in between Type A tests. Such an additional crite-rion, 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 B and C tests performed outside the outage in 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. Currently being dis-cussed by the NRC sta;ff 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 (nominally 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 penetra-tion nust be increased ur,til 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 compo-nent 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 2 or C "as found" test results (over the previous "as left" Type B or C test 8

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- results) shall be 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,, mea-sures 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,. Th,is 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.

Major Changes 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 regula-tion defines gereral containment systen 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.

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> E'4

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 g locktesting). As a result, provision has been made in this revi.eion for consideration by the NRC staff of alternative leakage test require-ments when necessary.

5. Type A test pressure. The option of perfonning p,eriodic 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 test results has turned out to be unsuccessful. Reasonable argument can be made for low pressure testing. However, the NRC staff believes that the peak cal-culated 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 accordance with Regulatory Guides 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 pro-vide 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 &

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• Pressure Vessel Code for mechanical systems. A different time base is used, but the frequency has remained es;entially 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 practic al." 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.

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

^

cent 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 leak-age rates are quantified. The regulation still refers to the individual plant technical specifications for these values. Debate continues, how-ever, 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.

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• 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 ir.plementation 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 test-ing schedules, and amount of technical specification changes r.eeded.

The separate views of Commissioner Frederic M. Bernthal follow:

The public should be aware of the fact that the Commission for over a year has attempted to adapt the Backfit Rule to all rulemaking, even rulemaking that has nothing to do with changes to powerplant hardware and the original intent of the Rule.

This rulemaking and the accompanying analysis illustrates the difficulty. When applied to human-factors rules, updating antiquated i

rules, and certain other rulemaking, the Backfit Rule continues to exact NRC resources wholly disporportionate to any conceivable benefit to the public. The record already shows cases where the Commission has been forced to sidestep a strict reading of the cost-benefit requirements and the "... substantial increase in overall protection..." threshold of the Backfit Rule, when it nevertheless 12

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finds broad agreement that a rulemaking is in the public interest (e.g. in the case of conversion of non-power reactors from HEU to LEU).

The public may therefore wish to comment directly on the question of whether the Commission should continue its attempts to apply the Backfit Rule to all rulemaking, or whether the Rule should be revoked as it applies to rulemaking activity per se.

s Alternatively, the public may wish to consider whether the Commission should amend the Backfit Rule to waive the " substantial, increase" provision, an. o indicate explicitly that non-monetary benefit; may be weighed by the Commission in the cost-benefit balance, when such considerations are found by the Commission to be in the public interest.

FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY The Commission has determined under the National Environmer.+al Policy Act of 1969, as amended, end 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 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 13

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significant impact on which this detemination 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, I

U.S. Nuclear Regulatory Comission, Washington, DC 20555, Telephone (301)443-7893.

PAPERWORK REDUCTION ACT STATEMENT ,

This proposed rule arre-ds 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 Comission. The draft analysis is available for inspection and copying in the NRC Public Document Room, 1717 H Street, NW., Washington, DC. The Comission requests public com-ment on the draft analysis. Coments 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 reactors. The analysis is available for inspection and copying in the NRC Public Document Room, 1717 H Street NW., Washington, DC. The Comission requests public coment on the analysis. Coments may be submitted to the NRC as indicated under the Addresses heading.

The analysis does not conclude that there is a substa,ntial increase in the overall protection of the publi 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 justi-fied due to better, more uniform tests and test reports, greater confid-ence in the reliability of the test results, fewer exemption requests.

and fewer interpretive debates. For these reasons, which are presented in greater detail in the backfit analysis, the Comission has decided to The proceed with publication of the proposed rule for coment.

Comission's decision regarding promulgation of the rule, even though it l

may not provide a substantial increase in the overall protection of the public health and safety or the comon defense and security, is tentative l

pending receipt of public coments on this issue.

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

605(b)), the Commission certifies that this rule will not, if 15

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> A'a promulgated, have a significant economic impact on a substantial r.uiu'uer 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 forth in the Regulatory Flexibility Act or the Small Business Size

~ 5tandards 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, Fi.e pre- 1 tion, Incorporation by reference, Intergovernmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Report-ing and recordkeeping requirements.

I RELATED PEGULATORY GUIDE

~

l The notice of availability of a draft regulatory guide on the same l

subject " Containment System Leakage Testing" (MS 021-5) is also being published in the notice section of the 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 I

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 I following amendments to 10 CFR Part 50.

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PART 50 -- DOMESTIL 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, 9ES, 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 50.58, 50.91, and 50.92 a' , issued

~

, under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C.'2152). Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42U.S.C.2234). Sections 50.100-50.102 also issued under sec. 186, 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)and50.54areissuedundersec. 1611, 68 Stat.

949,asamended(42U.S.C.2201(1));and50.55(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:

Leakage Tests for Containments of Light-Water-Cooled Nuclear Power Plants 17

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

1. Frequency

2. Pressure

3. Air Locks

. Acceptance Criteria C. Type C Test

1. Frequency I 2. Pressure / Medium

3. Acceptance Criteria l 4. Valves That Need Not Be Type C Tested 18

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i IV. SPECIAL LEAK TEST REQUIREMENTS A. Containment Modification or Maintenance 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 B. Effective Date I. INTRODUCTION One of the conditions of all operating licenses for light-water-cooled power reactors as specified in i 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-rents 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 19

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service life. This appendix identifies the general requirements and acceptance criteria for precperational and subsequent periodic leak testing.1 II. DEFINITIONS ACCEPTANCE CRITERIA Standards against which test results are to be compared for 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.

"A5 LEFT" LEAKAGE' RATE The leakage rate following any needed repairs or adjustments to the leakage barrier being tested.

CONTAlt' PENT INTEGRATED LEAK RATE TEST (CILRT)

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

CONTAINMENT ISOLATION SYSTEM FbNCTIONAL 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.

ISpecific guidance concerning acceptable leakage test method, 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 obt:'ned from the Nuclear Regulatory Commission, Document Management Branch, Washington, DC 20555 20

[7590-01] j CONTAINMENT ISOLATION VALVE Any valve defined in Gen,..al 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 priniary 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.

This definition does not include boiling water reactors' (BWR) reactor buildings or pressurized water reactors' (PWR) shield buildings.

l l Also excluded from the provisions of this appendix are the interior barriers such as the BWR Mark II 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.

21

[7590-01]

> A's 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).

g 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 leakage path (e.g., the smallest leakage of two valves in series). This 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-22

[7590-01]

• 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 conduct'ed 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 5 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 barrie".

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

l 23

~ ~

s f,

, , N [7590-01]

'N, ,

TYPE A TEST d,,

\

Atesttomeasurethecontainmentsystemb5erallintegratedleakage _s rate under conditions representing design basis 1pss-of-coolant accid'nt e containment pressure and systems alignments (1) aft ti the containmenk '

t; system has been completed and is ready for operation and (2) at >periodic - - \,

intervals thereafter. The verification test is not part of this N  :

definition - see CILRT. .

TYPE B TEST h .

A pneumatic test to detect,and measure local leakage throhgh the folicwing containment penetrations: '

c.

' ;. 3 -

V ,

4

) Those whose design incorporates resilient seals, gaskets,

. m l

sealant compounds, expansion bellows, or fitted with' flegible ; metal seal i i assemblies. L (2) Air locks, including coor seals and door o' para' ting mechan %h s'

i penetrations that are part of the containment presw re boundary.

s ..

e' i  ; "

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

1

\ .. , >

rates. VERIFICATION TEST '

, i.

Test to confirm the capability of the Type A test method.and equip- u.

s ment to measure L,.  ;

III. GENERAL LEAK TEST REQUIREMENTS ss i 3

A. Type A Test -f, , .

(1) Preoperational Test. A preoperationel Type.A. test must be  ;.

'\

conducted on the containirent system and must be preceded by:

(a) Type B and Type C tests.

(b) A struc' ural integrity test. [

E *w 24 3 s s g

\

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

[7590-01)

(2) Periodic Test. A periodic Type A test must be performed on the containment 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 between the first preoperational test and the first periodic test is exceeded, another preoperationa,1 test will be necessary. If such an additional preoperational Type A test or an additional Type A test required by Sec' ins 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 containment design pressure and must not fall more than I psi below P ac for the duration of the test, not including the verification test. The 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 l

l. 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 l Information on valve leakage that requires corrective action prior to, during, or after the test (see Section V.B.) must be included in the f

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

(

appendix.

25

[7590-01]

' (6) Verification Test. A leakage rate verification test must be perfcrmed after a Type A test in which the leakage rate meets the criterioninIII.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 verification test by more than 0.25 L,.

(7) Acceptance Criteria. ,

(a) For the pre ? rational Type A Test, the "as left" leakage rate must not exceed 0.75L,, as deterr.ined 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 (ii) the local leakage rates are measured before and after the isolation, repair, or adjustment and are reported under Section VI of this appendix.

26

[7590-01]

- (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) Retestifig.

(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 (it'.A.(3)) shall be submitted to the NRC staff for review and approval. An es left Type A test that meets the acceptance criterion of 0.75L, is required prior to plant startup.

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

(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 leakage test program is acceptable to the NRC staff on some 27 l

[7590-01]

other defined basis. This testing must be perfomed 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 Type A test failure might indicate that an alternative leakage test g program such as more frequent Type B or Type C testing may be more appro-priate than the perfomance of two consecutive successful Type A leakage tests. The licensee may then submit a Corrective Action Plan and an alternative leakage test program proposal for NRC staff review.

If this submittal is approved by the NRC staff, the lic< see 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)(1).

(9) Permissible periods for tasting. 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 l 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 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 I

the reactor co an operating mode requiring containment integrity.

t i 28 l

i

[7590-01]

I

+ (b) For containment penetrations employing a continuous leak-age monitoring system that is at a pressure not less than P,c, leakage readings of sufficient sensitivity to permit cor.parison with Type B test leak rates must be taken at intervals specified in the Technical Specifi-cations. 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 continuous leakage monitoring system cannot be isolated, such as inflatable air lock door seals, leakage into the containment must be accounted for and the Type A test results corrected accordingly.

(2) Pressure. Type B tests must be conducted, wheth" individually except as pro-or in groups, at 'a pneumatic pressure not less than P ac 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 I prior to initial fuel loading and at least once each 6-month interval Alternatively, if thereafter at an internal pressure not less than Pac.

there have been no air lock openings within 6 months of the last 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 that purpose of removing air lock testing equipment following an air lock test does not require further testing of the air lock.

(b) Intermediate tests must be conducted as follows:

l (i) Air locks opened during periods when containment l

l 29 i -

. [7590-01]

integrity is required by the plant's Technical Specifications must be tested within 3 days after being opened. For air lock doors opened more 4 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 opened during periods when containment integrity is not required by the 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 he done at Pac, the test pressure must be as stated in the Technical Specifications.

(ii) Whenever maintenance other than on door seals has been perfomed 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-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 E 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 obtained through component leakage surveillance systems (e.g., continuous pressurization of individual or clustered containment components) that maintain a pres-sure not less than P,c at individual test chambers of those same contain-30

[7590-01) ment penetrations during normal reactor operation. Similar penetrations not included in the component leakage surveillance system are still sub-ject to individual Type B tests.

(c) An air lock, penetration, or set of penetrations that fails to pass a Type B test must be retested following determination 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 isolation 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 l

! results must not exceed 0.60L, using maximum pathway leakage and including leakage rate readings from continuous leakage monitoring systems.

31

' [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 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 it can be shown that the valve does not constitute a potential containment atmosphere leak path during or following an accident, con-l l

sidering 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 trodification, repair, or replacement of a component that is part J of the containment system boundary and that may affect containment inte-grity 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 or Type C testing must also be preceded by a Type B or Type C test. The measured leakage from this test must be included in the report to the Comission required by Section VI of this appendix. Following structural changes or repairs that affect the pressure boundary, the licensee shall 32

=. . .. - . .

[7590-01]

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.

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-I 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.

i B. Multiple Leakage Barrier or Subatmospheric Containments The primary reactor containment barrier of a multiple 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 subatraospheric containments (e.g.,

i 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, B, and C Test Details Leak test methods, procedures, and analyses for a steel, concrete, 33

, , ~ , - - - - - - - - - - y -,,w, - , . . , - - , ,--. . . - _ . .,_n.._ ..- ._., - , . , , .,-,.,n_,,,,,,,.,___n, , - e -- . . , , . - , - - - . . - . , . -n- n,,.n

[7590-01]

D L's

- or combination steel and concrete containment and its penetrations and 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 s

with the periodic Type A test when perfomed during the same outage as s

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 1 eft 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.

VI. REFORTS 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 sumary technical report sent not later than 3 months after the conduct of each test to the Comission in 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 in the r.ianner specified in 5 50.4 and at the time of the next Type A test j

! submittal. Reports must be submitted to the NRC Regional Administrator within 30 days of completion of any Type B or C tests that fail to meet l

l their as found acceptance criteria.

I 34 ,

.e -. - - - - - - ____ - _ . _ _ _ _ -.- .- . - . - - _ _ - - _ - _ - - .

[7590-01]

o

~

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 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. Applicability The requirements of this appendix apply to all operating nuclear power reactor licensees as specified in 5 50.54(o) of this part unless it 4

can be demonstrated that alternative leak test requirements (e.g., for certain containment designs, leakage mitigation systems, or different test pressures not .ifically 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 4

Technical Specifications if approved by the NRC staff.

B. Effective Date Thisappendixiseffective(30daysafterpublication). 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 Huclear Reactor Regulation for implementing this appendix. This submittal must include an implementation schedule, with a final implementation no later than (insert a date 48 months after theeffectivedateofthisrevision). Until the licensee finally implements the provisions of this revisien, the licensee shall continue 35

[7590-01]

to use in their entirety the existing Technical Specifications and the Appendix J on which they are based. Thereafter, the licensee shall use  !

in their entirety this revision and the Technical Specifications conforming to this revision.

Dated at Washington, DC, this day of , 1986.

For the Nuclear Regulatory Commission.

Samuel J. Chilk Secretary of the Commission

.4 i

36

g..

e

> s. .

3 e

e ENCLOSURE 2 s

-' ' ~*" - - - , _ _ _ _ _ , , _ _ , _ _

l 1

Draft)ttT 08/25/0C Sepiemherl9%

Federal Register Notice

3 [7590-01] i l

1 NUCLEAR REGULATORY COMMISSION 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 ifcensing and en-forcement staff by eliminating conflicts, ambiguities, and a lack of uni-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 comment.s received on or before this date.

ADDRESSES: Mail written comments to: U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch. Deliver comments to: Room 1121, 1717 H Street NW, Washington, DC, between 8:15 a.m. and 5:00 p.m. weekdays.

1

' - - [7590-01]

1 Copies of draft 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:

t 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, f'

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.

l 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; i

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

2

,_ , _ . . - - . . . _ , -_r ,,,y.. ,,__,~_,___.n_. , _ ,,,,,y, 3 y_,,. r____ -,_ . m.___ _,w-,. -- _ _ . __ . . . .-- . ,._ _ _ _ _ _

• [7590-01] l l

l 1

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 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 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.

3

[7590-01)

T

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

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

tion, 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 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 Continuous monitoring of all individual containment system integrity.

isolation barriers is not necessary.

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

Following is also addresses related, broader, longer term activities.

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 containment leakage monitoring" (suc as negative containment operating pressure) and "relatively frequent gross containment integrity check" (such as a low pressure pumpup just prior to operation to check for openings) are under consideration by the NRC These would identify large breaches of the containment system boundary, It should be noted during, or just prior to, normal operating conditions.

4

~

(

__. [7590-01]

. c.

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 4

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 1

Appendix J by licensee's technical specifications, the work of the TSIP a

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 i

to recor. mend appropriate actions to modify or to eliminate these unneces-l

! 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.

5

FRN INSERTS INSERT (5), FAGE 6 (CARR)

(5) Whether present operating plants or plants under review should be given the opportunity to continue to meet the current Appendix J provisiens if the proposed rule (reflecting consideration of public comments) becomes effective.

[ Note: Old (5), now new (6), provides Commissioner Roberts' first requested commentsolicitation.]

INSERT (7), PAGE 6 (BERNTHAL, ROBERTS)

(7) Whether (a) all or part of the proposed Appendix J revisions would constitute a "backfit" under the definition of that term in the Commission's Backfit Rule, and (b) there are parts of the rule which do not constitute backfits, but which would aid the staff, licensees, or both.

INSERTS (9), (10) (IECH)

(9) The advisability of referencing the testing standard (ANSI /ANS 56.8) in the regulatory guide (MS 021-5) instead of in the text of Appendix J.

(10) The value of collecting data for the "as found" condition of valves and seals and the need for acceptance criteria for this condition.

[7590-01) i

' INVITATION TO COMMENT Comments from all interested persons on all aspects of this revision and on the risk and cost effectiveness of containment leaktightness in general are requested by the comment expiration date in order that: 1) 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 whici. 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 j yggg fy) whole and of its separate provisions; (g')[if the existing rule or its proposed revision were completely voluntary, how many licensees would adopt either version in its entirety and why; l Afsaff (7)

(g}([ 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 pro-

/3;$ cg,rg(g),fo) posed revision as an interim updating of the existing regulation; (11) % whether the technical specification limits on allowable containment leakage should be relaxed and if so, to what extent and why, or if not, why not;

(/1} ()d what risk-important factors influence containment performance under severe accident conditions, to what degree these factors are considered 6

j

' [7590-01) in the current containment testing requirements, and what approaches should be considered in addressing factors not presently covered; what other approaches to validating containment integrity could be

[/3 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; f/// d

) what effect " leak-before-break" assumptions could have on the leak-age rate test program. Current accident assumptions use instanta-neous complete breaks in piping systems, resulting in a test program based on pneumatic testing of vented, drained lines. " Leak-before-break" assumptions presume that pipes will fail more gradually, leaking rather than instantly emptying.

lI)(2'I)howtoeffectivelyadjustTypeAtestresultstoreflectindividual 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 criterion for Type B and C tests performed outside the outage in 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 factored into the determination of a Type A test "as found" condition.

7

' [7590-01]

O

b. If a particular penetration or valve fails two consecutive Type B or C tests, the frequency of testing that penetration must be 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 left" Type B or C test results) shall be 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 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 1 reportable condition.

The existing requirements that the sum of all Type B and C tests i

be no greater than 0.60 L,shall also remain in effect.

l i

8 i

e

- [7590-01]

O Maior Changes 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 provideo, 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

1

i

[7590-01) accordance with Regulatory Guides 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 s

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.

7

9. Type A test allowable leakage rate prorating. Seventy-five percent of 'the allowable leakage rate represents the "as left" Type A test acceptance criterion,. leaving 0.25 of the allowable leakage rate as i

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.

l

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 i .

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

INSERT SV The separate views of Commissioner Frederic M. Bernthal follow:

The public should be aware of the fact that the Commission for over a year

. has attempted to adapt the Backfit Rule to all rulemaking, even rulemaking i that has nothing to do with chances to powerpTant hardware and the original intent of the Rule.

This rulemaking and the accompanying analysis illustrates the difficulty. -

When applied to human-factors rules, updating antiquated rules, and certain other rulemaking, the Backfit Rule continues to exact NRC resources wholly [

disproportionate to any conceivable benefit to the public. The record (

already shows cases where the Commission has been forced to sidestep a -

strict reading of the cost-benefit requirements and the "... substantial increase in overall protection..." threshold of the Backfit Rule, when it nevertheless finds broad agreement that a rulemaking is in the public interest (e.g. in the case of conversion of non-power reactors from HEU to LEU).

The public.may therefore wish to comment directly on the question of whether the Commission should continue its attempts to apply the Backfit Rule to all rulemaking, or whether the Rule should be revoked as it applies to rulemaking activity per se.

Alternatively, the public may wish to consider whether the Commission should amend the Backfit Rule to waive the " substantial increase" provision, and to indicate explicitly that non-monetary benefits may be weighed by the Commission in the cost-benefit balance, when such considerations are found by the Commission to be in the public interest.

[7590-01]

l 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 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.

CV 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 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

• [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 found r.ot 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 Commission, 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 i

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

12 f

l

.: [7590-01)

BACKFIT ANALYSIS 1

The Commission 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 reactors.

The analysis is available for inspection and copying in the NRC Public Document Room, 1717 H Street NW, Wishington, DC. The Commission requests public comment on the analysis. Comaients may be 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 confidence in the reliability of the test results, fewer exemption requests, and fewer interpretive debates. For these reasons, which are presented in greater fg3,jg(q detail in the backfit analysis, the Commission has sr2nt:: ;.. .m...g t il'n J 13A

're- the re^rf-e eat 'er : deter i etica per0ernt t: $ SC.1C^(oji37 oni,4 roe a do+nemtn>+te- th t the -,,1 ut11 -;;rit 2, ; ;;d-Tw!: :::tte- _

• +'"ti31 #"Cr0000 #n th0 CV0r2I' preisw6svu vi Uic y2wl*C 5027t' Ond 00f0ty-Jer t': ;; ;n Jm'wuoe ;nd 0 :;rity te be d="4"ad '*e- tha be"'f 4 + -

REGULATORY FLEXIBILITY CERTIFICATION In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number 13

c r' ' f ' 3 INSERT BA (CARR)

--- For these reasons, which are presented in greater detail in the backfit analysis, the Comission. has decided to proceed with publication of the The Commission's decision regarding promulgation proposed rule for comment.of the rule, even though it may not provide a substantial incr overall protection of the public health and safety or the common defense and security, is tentative pending receipt of public comments on this issue.

s e

. - - - , , - - _ . , - - - - - - - . . . , - . . - - , _ , - .- . - .e. -- ., ,--. -- - ----.,,, -- - - . - -

[7590-01]

. c.

i,.

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

i LIST OF SUBJECTS IN 10 CFR PART 50 .

,l r

Antitrust, Classified information, Fire prevention, Incorporation by reference, Intergovernmental relations, Nuclear power plants and

+

reactors, Penalty, Radiation protection, Reactor siting criteria, i

Reporting and recordkeeping requirements.

RELATED REGULATORY GUIDE

+

The notice of availability of a draft regulatory guide on the same 1

subject " Containment System Leakage Testing" (MS 021-5) is also being published in the notice section of the Federal Register. The draft regulatory guide contains specific guidance on acceptable leakage test i

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 f following amendments to 10 CFR Part 50.

i 14

[7590-01)

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 Sections 50.58, 50.91, and 50.92 also issued under (42 U.S.C. 5851).

Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Sec-tions 50.100-50.102 also issued under sec. 186, 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:

15

[7590-01]

Leakage Tests for Containments of 4

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

6. Type B Test

1. Frequency l

2. Pressure

3. Air Locks

4. Acceptance Criteria l C. Type C Test 1

1. Frequency

2. Pressure / Medium 16

• [7590-01]

b --

3. Acceptance Criteria

4. Valves That Need Not Be Type C Tested IV. SPECIAL LEAK TEST REQUIREMENTS A. Containment Modification or Maintenance 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 B. Effective Date 17 i --- - - - . - , - , . . _ . . _ , . _ _ _ , _ _ _ _ _ , , _ . _ , , , , , , _ _

[7590-01] l l

I. 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 l 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 i

life. This appendix identifies the general requirements and acceptance criteria for pre..perational and subsequent periodic leak testing.2 II. DEFINITIONS ACCEPTANCE CRITERIA Standards against which test results are to be compared for 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 I The leakage rate following any needed repairs or adjustments to the leakage barrier being tested.

CONTAINMENT INTEGRATED LEAK RATE TEST (CILRT)

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

I 2 Specific guidance concerning acceptable leakage test methods, procedures, I

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 Copies of the regulatory guide i

~

public comment with the designation MS 021-5.

may be obtained from the Nuclear Regulatory Commission, Document Managem Branch, Washington, DC 20555.

18 J

,,-_-,.,,,r.,. . _ - . _ _ , , _ _ _ - . - . . . . - . ,

r_..

x -, - __.. __ ,. _ . , . __---_.,,_,,m .-_,,m, _ _ - _ _ _ _ _ _ _ _ . .

'. [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 s 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 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

[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,,(WE!GHT 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 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 series).

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, Salves, 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,e (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 5 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)

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 s

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.

VEkIFICATION TEST Test to confirm the capability of the Type A test method and equip-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

o 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 containment 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 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 test i

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 ncrmal 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 inc?uded 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 criterion 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 ai,- erification 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-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 s

left" containment leakage rates.

(d) The offects 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 test that meets the accept,ance 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 l

[7590-01)

D A a (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 leakage 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 critorion 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 alternative leakage test program proposal for NRC staff review. If this 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 refueling or at i

other convenient intervals but in no case at intervals greater than 26

[7590-013 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.

(b) For containment penetrations employing a continuous leakage monitoring system that is at a pressure not less than P,c, 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, leakage 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 pressuia not less than P,e except as pro-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 Alternatively, if thereafter at an internal pressure not less than P ac.

there have been no air lock openings within 6 months of the last successful test at P,c, 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 Opening of the air the air lock or its door seals at 6-month intervals.

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 t as follo (i) Air locks opened during periods when containm integrity is required by the plant's Technical Spec For air lock doors opened more tested within 3 days after being opened. be tested at least frequently than once every 3 days, the air t openings. lock must Air locks' once every 3 days during the period of frequenity is not requir opened during periods when containment integrduring t dly tested iring plant's such periods.

Technical Specifications need not For air lock doors having testable seals, test containment integrity. i ments of this paragraph.

the seals fulfills the intermediate test requdone re at P,c, the In the event that this intermediate testing cannot be i l Specifications.

test pressure must be as stated in the Techn cal has been (ii) Whenever maintenance other k test at a test pressure of than o re performed on an air lock, a complete air loc not less than P,e testing retaining boundary.

(iii) Air lock door seal testing or reduced-pressure iodic full-pressure test of may not be substituted for the initial or perII.B.(3)(a) of this S the entire air lock required in paragraph I Acceptance Criteria. t (4)

I The sum of the as found or as lef t Type B and

('a) thway leakage and including results must not exceed 0.60L, usingitoring maximumsystems.pa f leakage rate readings from continuous leakageh mon gh Leakage measurementstinuous are acceptable if obta f (b) pressurization f component leakage surveillance systems (e.g., con I

28 I

_ _ . . [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 that fails to pass a Type B test must be ratested following determination of cause and completion of coirective 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 P,c.

(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 B and C leakage rate if:

(i) The valves have been demonstrated to have leakage rates that do not exceed those specified in the Techr.ical 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 isoiation valve need not be Type C tested if 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 'eakage 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 affact the pressurc boundary, the licensee shall 30

. [7590-01) 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.

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 '

These shall include:

non-destructive examinations have been conducted.

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 The primary reactor containment barrier of a multiple barrier or subatmospheric containment shall be subjected to Type A tests to verify Other that its leakage rate meets the requirements of this appendix.

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. B. 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

, [7590-01) isolation valves for light-water-cooled power rea-tors 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 i

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.

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 the manner specified in S 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 in the manner specified in 5 50.4 and at the time of the next Type A test submit-tal. Reports 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.

32

[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 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. Applicability The requirements of this appendix apply to all operating nuclear power reactor licensees as specified in 5 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 J

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 By (insert This appendix is effective (30 days after publication).

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-This submittal must include an implementation i menting this appendix.

schedule, with a final implementation no later than (insert a date 48 Until the licensee months after the effective date of this revision).

finally implements the provisions of this revision, the licensee shall continue to use in their entirety the existing Technical Specifications 33

. [7590-01]

Q 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 , 1986.

For the Nuclear Regulatory Commission.

Samuel J. Chilk Secretary of the Commission 34

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e ENCLOSURE 3

P

• 5 .

CONGRESSIONAL LETTER (S)

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/fa aarg'o,, UNITED STATES j 3 o NUCLEAR REGULATORY COMMISSION j

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;  : a WASHINGTON,0. C. 20555 l

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The Honorable Alan Simpson, Chairman Subcommittee on Nuclear Regulation Committee on Environment and Public Works United States Senate Washington, DC 20510

Dear Mr. Chairman:

Enclosed for the information of the Subcommittee are copies of a Notice of Proposed Rulemaking to be published in the FEDEPAL REGISTER, a Notice of Availability for a related draft regulatory guide (MS 021-5), and a copy of the guide.

The amendment of 10 CFR.Part 50 comprises a revision of Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." It prov' ides 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, Eric S. Beckjord, )irector Office of Nuclear [llegulatory Research l

l

Enclosures:

1. Federal Register Notice

2. Notice of Regulatory Guide Availability & R.G. MS 021-5

3. Draft Public Announcement l

cc: Sen. Gary Hart

1. %, UNITED STATES

! o NUCLEAR REGULATORY COMMISSION g 3 -) WASHINGTON, D. C. 20555

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1 4

The Honorable Morris K. Udall, Chaiman Subcommittee on Energy and the Environment Committee on Interior and Insular Affairs United States House of Representatives Washington, DC 20510

Dear Mr. Chairman:

Enclosed for the information of the Subcommittee are copies of a Notice of Proposed Rulemaking to be published in the FEDERAL REGISTER, a Notice of Availability for a related draft regulatory guide (MS 021-5), and a copy of the guide.

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, S

5t*

Eric S. Beckjord, D ector Office of Nuclear R ulatory Research

Enclosures:

1. Federal Register Notice

2. Notice of Regulatory Guide Availability & R.G. MS 021 3. Draft Public Announcement cc: Rep. Manual Lujan

8

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UNITED STATES NUCLEAR REGULATORY COMMISSION

., y . ,E WASHINGTON D. C. 20555 t g The Honorable Edward J. Markey, Chairman Subcommittee on Energy Conservation and Power Committee on Energy and Commerce United States House of Representatives Washington, DC 20510

Dear Mr. Chairman:

Enclosed for the information of the Subcomittee are copies of a Notice of Proposed Rulemaking to be published in the FEDERAL REGISTER, a Notice of Availability for a related draft regulatory guide (MS 021-5), and a copy of the guide.

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, 3

Eric S. Beckjord, D i

ctor Office of Nuclear R latory Research

Enclosures:

1. Federal Register Notice

2. Notice of Regulatory Guide Availability & R.G. MS 021-5

3. Draft Public Announcement cc: Rep. Carlos Moorhead

' '~ .-.a--___ a.,_, -

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l ENCLOSURE 4 1

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f DRAFT PUBLIC ANNOUNCEMENT

DRAFT NRC PROPOSES CHANGES TO CONTAINMENT LEAKAGE RATE TEST RULES The Nuclear Regulatory Comission is proposing to amend its regulations dealing with the leakage rate testing of commercial 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; comments made on the existing requirements over time; and requests for exemptions from the requirements receiveo 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; (3) Make provisions for consideration of alternative leakage test requirements when necessary; (4) Make changes to resolve past questions of interpretation; (5) Eliminate an option to perform periodic reduced pressure testing in lieu of testing at full calculated accident pressure; (6) Revise test frequency requirements; (7) Delete a requirement governing duration of testing;

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(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 1eakage 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 comments 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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ENCLOSURE 5 i

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ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT

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5 ENVIRONMENTAL ASSESSMENT AND FINDING 0F N0'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 Identification of Proposed Action 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.

1 It This revision of Appendix J has been in preparation for some time.

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 leak testing methods; (3) interpretive questions; (4)

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

s 6

heed for the Prcp ged Action Changes in the state-of-the-art cf leakage testirig, experience with usir.g 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 revision cf Appendix J will have no radiological The prcposed environmental inipact offsite. However, if the rule is promulgated in final form as new proposed, there will be an average increase in occupational radiation exposure onsite of about 3.0 man-rem per year of plant operatiori fcr 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 arrendrrent does not affect ocn-radiological plant effluents and has no other environmental irrpact. Therefore, the Commission 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) El of I'EF A (42 U.S.C. A. 4332(2)(E)), the staff has ccnsidered possible alternatives to the proposed action. One 2

l l

s

' alternative was net to initiote 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 environnental impact change but problems incurred in using the present rule would not be resolved.

" 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-n.andatory and ccr.tainment integrity were not maintained.

The present approach of revising the existing rule was chosen as the best Revision cf Appendir J will be beneficial to all. Thc alternative.

public will benefit fron. in. proved reliability of containment leakege 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.

Alternative Use of Resources No alternative use of resources was considered, t

?

O

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

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.

4

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6 N

ENCLOSURE 6

O

. [.t *-/pg 650.109 BACKFIT ANALYSIS FOR PROPOSED 10 CFR 50, APP. J AND PROPOSED RG MS 021-5 I

BACKFIT ANALYSIS AND CONCLUSION RELATING TO THE PROPOSED REVISION TO 10 CFR PART 50, APPEtiDIX J AND IT5 COMPAN10h REGULATORY GUID_E 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 decreaseJustification in costs.

Instead, it is being proposed as both safety and cost neutral.

for the revision is cased on the need to conform present testing capabilities to the current state of the art, and to use the best availa requirements unambiguous, technically current, uniform in application and usefulness, legally consistent, and flexible enough to accomodate differing plant designs.

650.109(c) analysis describe how these aspects, The following discussion andand the substantive elements of the backfit review and oversight process that all rules and regulatory guides must goJustif through prior to issue for public comment.

completing such activities must be continually made throughout the 950.109 develo process.

As a result, all of the issues and elements of interest under have been scrutinized by a variety of reviewing bodies reviews.

The proposed rule is intended to be applied to the entire population of nuc power reactors and it clearly constitutes a backfit.

Prior to the effective date of the backfit rule and its app including CRGR.

its contents and the justification therefor, to the A relationship to other NRC activities related to containment integrity, a value-impact study, and related justifications for this up and companion regulatory guide (MS 021-5) for public coment.

The regulatory analysis written for this proposed revis Room.

Included in this regulatory analysis package was a cost analysis by Science & Engineering Associates, Inc.; Mathtec, Inc.; and S. Cohen &

Associates, Inc.

Tables 1.3 and 1.4 in the cost analysis estimated that the Appendix J revisio can result in a potential total cost saving ranging from about $98 million (

10% discount rate) to $164 million (0 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 less than four person-rem per reactor year. It should be noted that current occupational exposure levels average annual collective doses of 753 person per reactor year. 1

I o

The analysis projected total costs to the NRC on the order of $4 million (@

10%) to $5 million (9 5%), principally due to increased manpower efforts associated with technical specification revisions. Of this, about The $3 million remainder would be incurred over the next few years during implementation.

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 majcr industry benefit would cccur during the operating life of the power reactor population where present worth savings on the order of $106 million (010%) to $173 million (0 5%) were projected.

Although the cost analysis also identified increased operating costs, these costs.would be outweighed by significant savings in replace to Ap'pendix J will reduce the expected frequency of containment integra leakage rate (Type A) tests.

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 App 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 itsT regulatory safety decision process.

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 sound and unambiguous .Therefore, even regulations that minimize the need for exemptions. favorable economic cverall costs ar.d 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 Theto changes in practice and replacement of aItreferenced will also ANS characteristics of the post-LOCA containment configuration.The test method is b standardize reporting requirements.

statistical evaluation of multiple pressure, temperature, and humidity readings needed to quantify a very small leakage rate from a very large volume. Forexample,a0.1%perdayleakagerateoutofacongainment volume of 2,000,000 cu. ft. under a pressure of 55 psia at 150 F is 2

roughly equivalent to that represented by a hole with a dic7eter of about The actual allowable leakage rate is defined for each plant 1/16 inch.

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 This should enhance the stability of this would not be affected.

regulation, and allow greater flexibility for acceptance of alternative leak-test requirements to accommodate variations in containment systems designs.

o The current leakage limits established by NRR for plant-specific siting These current leakage are based on analyses pursuant to 10 CFR Part 100.

limits are expected to remain unchanged under this proposed Appendix J revision.

o Discussions between NRC staff, nuclear industry representatives, and professional and standards groups indicate that Appendix J to 10 CFR Part 50 needs to be revised to update the criteria, clarify questions of interpretation, and delete references to an obsolete ANSI standard on 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 appPfing existing requirements, advances in containment leak testing methods, and multiple requests (since 1973) for exemptions.

o As proposed, Appendix J contains only the general requirements and acceptance criteria (no testing techniques) for preoperational and subsequent periodic leak testing. Prescriptive and detailed testingInterested per techniques are not incorporated in this revision.

be offered an opportunity to comment on specific guidance concerning leakage test methods, procedures, and analyses that are acceptable to NRC staff to implement these requirements and criteria (draft Regulatory C/ide MS 021-5).

l 3

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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 Theavailable revision procedures, thereby not freezing a stale (1972) technology.

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 lef t 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, In some cases it maytest entail 4

procedures, data analyses, and test reports.

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 Those' plantshours notpumping reportingtime may be needed when testing at full pressure.

"as found" leakage results will be explicitly required to do so.

Licensees will have to review plant test procedures against the revised -

requirements and recomendations. This will detemine the extent of Following this changes needed to the technical specifications.

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.

(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 of radioactivity as is traditionally assumed for analyses pursuant to 10 CFR Part 100.

This reliability against leakage has been brought about by 4

, The NRC design requirements and use of industry codes and standards.

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 theHo containment system.

has revealed that the more likely leakage paths exist through penetrations and valves. Therefore, more This focus is provided on penetrations and valve improved test focus is difficult to (Type B 8 C) leakage tests.

quantify because the available data from containment systems testingsafety Substantial already indicates a high reliability for low leakage. The benefits have derived from the existence of Appendix J itself.

% 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 The more frequent radiation exposures than are presently experienced.

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 Theincreases are about high estimate is 5.6 3.0 person-rem per plant per year of operation. This compares person-rem per plant per year, and the low 0.5 person-rem.

with an average annual collective dose of 753 person-rem per plant (from NUREG 0713, Vol. 5, " Occupational Radiation Exposure at Nuclear Power Reactors," 1983), and represents an average potential increase of 0.4k.

(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 Industry construction, implementation as ranging from $106 million to $173 million.

costs are estimated to be about $3 million to $4 million, due to revisior.

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 Thisofisthis revision because theas savings are cost-neutral on an industry-wide basis.

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.

5

(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, quantifiable 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 populatici. of all water-cooled power plants in this country, the estimateo hRC resource burden is about $3 - 4 million for implenentation and $1 million for operation over their remaining life.

This is due principally to increased manpower efforts associated with technical specification revisions. 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 ar.d vintage are involved, it is believed that the net impact will not vary significantly. Major problems with the existing rule that are unique to older (pre-Appendix J) plant designs have been handled 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 comment 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 Any l

some years away, and an immediate need exists to update Appendix J.

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

950.109(a)(3), CONCLUSION There is no substantial increase in the overall protection of the public health and safety or the common 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 l reports, greater confidence in the reliability of the test rbsults, 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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EtlCLOSURE 7 l

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's [7590-01] i NUCLEAR REGULATORY COMMISSION Draft Regulatory Guide; Issuance, Availability The Nuclear Regulatory Commission has issued for public comment 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 Commission 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 conc 31ng certain of the information needed by the staff in its review of applications for permits 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 Divi-sion 1, " Power Reactors." It is being developed to provide guidance on procedures acceptable to the NRC staff for conducting containment leakage tests. This draft guide endorses American National Standard ANSI /ANS-56.8-1981, " Containment System Leakage Testing Requirements ~."

This draft guide, as issued for comment, proposes endorsement of the 1981 l version of ANSI /ANS 56.8. It should be noted that a revision to ANSI /ANS 56.8 l

is being completed. Roughly two-thirds of the positions in the draft guide are expected to parallel revisions made to ANSI /ANS 56.8. The current apparent large number of differences between the guide and the standard will therefore be greatly reduced to a relatively few actual differences upon publication of

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the new ANSI /ANS 56.8 standard. For information regarding the pending revision to ANSI /ANS 56.8-1981, contact the American Nuclear Society, 555 North Ken-sington Avenue, La Grange Park, Illinois 60525.

This draft guide is being issued to involve the public in the early 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 ana esis, has been prepared and made available in conjunction with the proposed revision to 10 CFR Part 50, Appendix J, that is also being published for public comment in the 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

! implementation schedule). Coments should be sent to the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Atten-tion: Docketing and Service Branch, by .

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.

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Regulatory guides are available for inspection at the Commission'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 dis-tribution list for single copies of future draft guides in specific divisions should be made in writing to the U.S. Nuclear Regulatory Commission, Washington, DC, 20555, Attention: Director, Division of Technical Informa-tion and Document Control. Telephone requests cannot be accommodated. Regula-tory guides are not copyrighted, and Commission approval is not required to reproduce them.

Dated at Rockville, Maryland, this day of 1986 Guy A. Arlotto, Director, Division of Engineering Safety, Office of Nuclear Regulatory Research 9

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