ML20236X694

Supplemental Safety Evaluation Supporting Seismic Qualification of Auxiliary Feedwater Sys (Multi-Plant Action C-14).SALP Input & 861029 Fr Re Proposed Rules Section Encl

ML20236X694
Person / Time
Site:	Oconee, 05000000
Issue date:	08/05/1986
From:	NRC
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References
FOIA-87-714, REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-81-14, NUDOCS 8712100101
Download: ML20236X694 (13)

Text

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9 l q, SUPPLEMENTAL SAFETY EVALUATION SEISMICQUALIFICATIONOFTHEAUXILIARYFEEDWATERSYSTEM(MPAC-14)

OCONEE NUCLEAR STATION, UNITS 1, 2 AND 3 I

By memorandum from L. S. Rubenstein to Gus Lainas dated October 22, 1985, the stafftransmitteditssafetyevaluationreport(SER)concerntng' compliance with the criteria of Generic Letter 81-14 " Seismic Qualification of Auxiliary Feedwater Systems" for Oconee, Units 1, 2 and 3.

Compliance with the criteria of the generic letter was to be demonstrated in order to assure that the requirements of GDC 2 and 34 were satisfied for assuring post seismic event shutdown decay heat removal capability. The staff SER identified three open items as follows:

4 1.

The capability of the auxiliary feedwater (AFW) system and/or safe shutdown facility to withstand a safe shutdown earthquake (SSE) concurrent with a single active failure.

Requirements for the isolation boundary between seismic emt non' seismic.

2.

portions of the AFW system.

3.

Walkdown of the currently nonseismically qualified areas of the AFW system.

The following is our evaluation of these issues.

1.

AFW System Single Failure Capability Following An SSE Generic Letter 81-14 states that licensees were to demonstrate that the AFW system could perform its shutdown decay heat removal safety function followinganSSE[maximumhypotheticalearthquake(MHE)forOconee)and concurrent single active failure. Alternatively, the licensee could demonstrate the availability of a seismically qualified alternative system for performing this function.

In the SER, the staff noted that the AFW system in each of the three Oconee units was located at the basement elevation of the turbine building and was. therefore, subject to a complete 8712100101 871204 PDR-FOIA WEISB87-714 PDR

V failure as a result of flooding caused by rupture of the nonseismic condenser circulating water line.

In such an event, the only identified means for shutdown decay heat removal for the three units would be the seismically qualified safe shutdown facility (SSF) auxiliary service water (ASW) pump. However, because the SSF consists of a single ASW pump for supplying feedwater flow to all three units, a single failure in it results in a loss of decay heat removal capability through the st'eam generators.

Consequently, we informed the licensee that we were pursuing a possible backfit of Oconee to correct this condition and satisfy the requirements of GDC 2 and 34 for decay heat removal capability following an earthquake.

By letters dated April 28 and May 7, 1986, the licensee provided addi-tional information regarding this concern in order to support their assertion that previously completed modifications $ 11 assure adequate post-seismic decay heat removal capability and no further backfit is cost-beneficial. The licensee indicated that penetration seals and water-proof doors have been installed between the turbine building and the auxiliary building in each unit to provide waterproofing to a height of 20 feet above the turbine building basement floor. Thus, the high pressure injection (HPI) system, low pressure injection (LPI) system, auxiliary service water system, and reactor building spray system pumps located in the auxiliary building would be available as an alternative to the AFW system and SSF ASW pump for shutdown' decay heat removal.

Further, the licensee indicated that revised operating procedures have improved the operator's ability to quickly respon'd to a turbine building

~

flood by providing guidance on means to isolate the circulating water system, initiating feed-and-bleed utilizing the HPI pumps and starting j

f the SSF ASW pump.

l TDn staff performed a quantitative probabilistic evaluation of the above information contained in the Oconee PRA.

Based on this review, a core melt frequency and an associated cost benefit were determined for the seismic flooding scenario accounting for the indicated plant improvements.

i J

_-______a

l

~~ I The results of this analysis indicated that adequate core melt protection has been provided and no further plant improvements to correct seismic flooding concerns are warranted. This determination is based on the flooding protection provided for the HPI and LPI pumps for use in the l

feed-and-bleed mode along with the SSF ASW pump which serve as a suitable i

redundant alternative decay heat removal means to the AFW system since the AFW system itself is unprotected from flooding and, there' fore, assumed unavailable following an SSE. The staff, therefore, concludes that the concern regarding post-seismic event decay heat removal capability and concurrent single failure is resolved, and the criterte of Generic Letter 81-14 regarding seismically qualified alternative decay heat removal means is satisfied.

2.

Isolation Between Seismic and Nonscismic-Portions of the AFW System By letters dated February 6,1986 and March 5,1986 (two letters), the licensee indicated that as a result of their continuing review of the seismic qualification of the AFW system in response to staff co,ncerns, a condition was identified outside the design basis for Oconee.

Speci-fically, a) certain manually operated boundary valves are not normally closed, b) certain valves do not have complete seismic qualification documentation, and c) some piping attached to the upper surge tanks is not seismically quhlified.

In the above letters, t'he licensee provided a safety evaluation which discussed the capability to safely shutdown the plant in the event of an SSE (MHE for Oconee) given the specifically indicated deficiencies and described proposed corrective actions for assuring AFW system seismic qualification in each case. The licensee also identified a schedule for implementation of required modifications in order to achieve AFW system seismic qualification in accordance with i

I the design basis.

l l

The licensee's safety evaluation provided a justification for continued operation based on the inherent seismic resistance of nonseismically qualified piping and valves, and on the diversity of seismically qualified alternative means of decay heat removal. The licensee indicated l

-4 L

that the results of the study of earthquake effects on power plants being perfomed by the Seismic Qualification Utilities Group-(SQUG) have shown nonseismically qualified piping and valves to generally remain functional in seismic events. This capability has been evaluated in depth by the staff in the resolution of USI A-46 Seismic Qualification of Equipment I

in Operating Nuclear Plants. More importantly however, as discussed in Item 1, above, the seismically qualified SSF ASW pump and' feed-and-bleed capability are available for decay heat removal should the AFW system fail following an SSE..These additional means for assuring shutdown are not only significant in the interim while the ider41fied AFW system seismic deficiencies are corrected, but also serve as additional defense-in-depth protection against core melt in the long tem given the seismically-l induced flooding vulnerability of the AFW system discusser veviously.

A Corrective actions identified by the licensee for AFW system seismic qualification deficiencies will be one of the following as appropriate:

a) Normally open boundary valves will be closed, or will be modified to be remotely operated, or analysis will demonstrate that failure of piping beyond these valves will have no impact on system function.

b) Seismically unqualified piping will be analyzed and supported to withstand an MHE.

c) Seismically unqualified valves will be shown capable of withstanding

)

an MHE, or will be replaced, or analysis will demonstrate that failure j

1 will have no impact on system function.

The staff finds the above identified corrective actions to be in accordance with the AFW system design basis and, therefore, acceptable for assuring its seismic qualification.

I

' The licensee's letter of March 5,1986 stated that a schedule for implementation of the required modifications will be provided to the staff by January 5,1987, with completion of the modifications estimated to be accom-plished by January 1990.

In the interim, the licensee indicated that plant l

procedures have been revised to instruct the operator to investigate those locations where nonnslly open valves exist in interfaces between seismic and nonseismic portions of the AFW system following an SSE in order that any l

necessary action can be taken to isolate the boundary. Because of the above indicated alternative decay heat removal means, the staff concurs with the 1

licensee's proposed schedule for implementation of corrective actions, however, any schedule slippage should be properly justified. We further conclude that adequate post-seismic event shutdown decay heat removal capability is provided for assuring continued plant safety, and the concern regarding isolation of the seismic /nonseismic boundary is resolved.

y 3.

Walkdown of Nonseismically Qualified Areas of the AFW System As indicated in the licensee's March 5,1986 letter which conta,ined LER 269/86-02, the licensee has reviewed the seismic /nonseismic interfaces f

in the AFW system for all three Oconee units. We, therefore, consider

)

the concern for a walkdown of nonseismically qualified areas to be

]

I resolved.

Based on the above, the staff concludes that the licensee has demonstrated adequate post-seismic event decay heat removal capability in accordance with j

the criteria of Generic Letter 81-14 by comitting to correct identified deficiencies in the seismic qualification of the AFW system itself, and by demonstrating adequate seismically qualified alternative capability utilizing the SSF ASW pump and HPI pump (feed-and-bleed) in the event of loss of the AFW system as a result of seismically induced flooding. We, therefore, conclude that Oconee meets the requirements of GDC 2 and 34 for post-seismic shutdown decay heat removal capability and is, therefore, acceptable. A schedule for implementation of required modifications should be provided by January 5, 1987 with actions completed by January 1990. We consider MPA C-14 Seismic Qualification of the AFW System to be complete for Oconee, Units 1, 2 and 3.

i 5-The licensee's letter of March 5,1986 stated that a schedule for implementation of the required modifications will be provided to the staff by

' January 5,1987, with completion of the modifications estimated to be accom-plished by January 1990.

In the interim, the~1icensee indicated that plant procedures have been revised to instruct the' operator to investigate those locations where nonna11y open valves exist in interfaces between seismic and nonseismic portions of_ the AFW system following an SSE in order that any necessary action can be taken to isolate the boundary. Because of the above indicated alternative decay heat removal means, the staff concurs with the licensee's proposed schedule for implementation of corrective actions, however, any schedule slippage should be properly justified. We further conclude that adequate post-seismic event shutdown decay heat removal capability is provided for assuring continued plant safety, and the concern regarding isolation of the seismic /nonseismic boundary is resc1ved.

y 3.

Walkdown of Nonseismically Qualified Areas of the AFW System As indicated in the licensee's March 5,1986 letter which contained LER 269/86-02, the liccaee has reviewed the seismic /nonseismic interfaces in the AFW system for all three Oconee units. We, therefore, consider l

the concern for a walkdown of nonseismically qualified areas to be resolved.

Based on the above, the staff concludes that the licensee has demonstrated adequate post-seismic event decay heat removal capability in accordance with the criteria of seneric Letter 81-14 by coninitting to correct identified deficiencies in the seismic qualification of the AFW system itself, and by demonstrating adequate seismically qualified alternative capability utilizing the SSF ASW pump and HPI pump (feed-and-bleed) in the event of loss of the AFW system as a result of seismically induced flooding. We, therefore, conclude that Oconee meets the requirements of GDC 2 and 34 for post-seismic shutdown decay heat removal capability and is, therefore, acceptable. A schedule for implementation of required modifications should be provided by January 5 1987 with actions completed by January 1990. We consider MPA C-14, Seismic j

Qualification of the AFW System to be complete for Oconee, Units 1, 2 and 3.

SALP INPUT Seismic Qualification of the Auxiliary Feedwater System (MPA C-14)

Plant: Oconee Nuclear Station, Units 1, 2 and 3 1.

Management Involvement and Control in Assuring Quality: Category 2 Once made aware of the importance the staff placed on resolving this issue, Duke Power Company management expressed strong interest in gaining final re. solution of the remaining concerns in the area of AFW system seismic qualification and provided the necessary control to assure quality However, management control did not expedite response to the responses.

staff.

2.

Approach to Resolution of Technical Issues from a Safety Standpoint:

Category 2 After additional discussions, the licensee finally recognized the significance of assuring post-seismic event decay heat removal capability, and demonstrated ade'quate knowledge for resolution of the remaining technical issues.

3.

Responsiveness to NRC Initiatives:

Category 3 4

Even after management involvement, the licensee was not timely in providing the necessary additional information for resolution of identified staff concerns. Final resolution of MPA C-14 took over four years to accomplish.

4.

Enforcement History:

Not Applicable 5.

Reporting and Analysis of Reportable Events:

Not Applicable 6.

Staff (Including management): Not Applicable 7.

Training and Qualification Effectiveness: Not Applicable Overall Rating:

Category 2 i

i l

39538 1

r d-i "+' '

Proposed Rules Vol. 51. No. 209 Wednesday. October 29, 1986 Trvs sect.on o' the FEDERAL REGISTEA FR 17053); published in final form on

7. NRC regulations should require l

t contans notees to the s@c of the February 14.1973 (38 FR 4385); and such an inspection program, and define Poposed asuance of rules and became effective on March 16.1973. The the test requirements and acceptance rigulasons. The purpose of these not.ces only amendment to this appendix since criteria.

as to gve interested persons an 1973 was a 1 mited one on Type B

8. A standard loss r/ccoolant accident (penetration) test requirements that was is assumed as the design basis accident.

published for comment on }anuary 11.

Since thi containment isolation system ma ng pr or to t f

final rules 1980 (45 FR 2330), published in final form is an engineered safety feature. only September 22.1980 (45 FR 62789); and safety grade systems and components l

l NUCLEAR REGULATO3Y became effective on October 22,1980.

are relied upon to define the COMMISSION This revision of Appendix l has been containment boundary tbst must be in preparation for some time. It will exposed to the containment pneumatic 1

10 CFR Part 50 provide greater flexibility in applying test pressure for the integrated leak rate i

alternative requirements due to test. In addition, all safety grade l

Lcakage Rate Testing of Containments variations in plant design and reflects systems are assumed to be subject to a cf Light Water-Cooled Nuclear Power changes based on:(1) Experience in potential single active failure, and must applying the existing requirements: (2) be locally leak rate tested accordingly.

Plants ActNcy; Nuclear Regulatory Advances in containment leak testing

9. Pneutratic testing to peak Commission.

methods; (3) Interpretive questions: (4) calculated accident pressure is adequate cetioN: Proposed rule.

Simphfying the text-(5) Various without testing for, or at, accident external / internal comments since 1973; temperatures or radiation levels.

suuuany:The Nuclear Regulatory and (6) Exemption requests received and

10. Shielding tests need not be Cnmmission is p posing to amend its approved.

performed.

regulations to up ate the criteria and The pro osed revision is for the

11. Periodic testing provides adequate clarify questions of interpretation in purpose o updating the existing confidence in the level of containment regard to leakage rate testing of regulation. Other related. longer term.

system integrity. Continuous monitoring containments of light. water-cooled and broader issues are currently under of allindividualisolation barriers is not nuclear power plants.The purposed rule review by the NRC staff, such as necessary, l

would aid the licensing and enforcement containment function, degree of integrity The scope of this revision to staff by eliminating conflicts, required, and validation of that integrity Appendix J is limited to corrections and ambiguities, and lack or uniformity in, under conditions other than postulated clarifications, and excludes new criteria.

the regulation of the inserdce inspechon in this rule. In order to better understand However, this notice also addresses pmgram its function and scope, assumptions related. broader, longer term activities.

CATE: Comment period expires January inherent in Appendix J are presented as Following is information of some of these other related activities that are not 26,1987. Comments received after this follow:

date will be considered if it is practical

1. Certain levels of radiation exposure reflet,ted in this proposed rulemaking.

to do so, but assurance of consideration at the plant site boundary shall not be in order to better identify the cannot be given except for comments exceeded under (a) operating or (b) availability of containment leakage integrity, concepts of" continuous received on or before this date.

design basis accident conditions, containment leakage monitoring" (such Aoomassts: Mail written comments to-

2. Certain levels of radistion exposure U.S. Nuclear Regulatory Commission.

to plant operating personnel shall not be as negative containment operating Washington.DC 20555. Attention:

exceeded under (a) operating or (b) pressure) and "relatively frequent gross Docketing and Service Branch. Deliver design basis accident conditions.

containment ir.tegrity check" (such as a comments to: Room 1121.1717 H Street

3. All four exposure levels (18.1b 2a.

low pressure pumpup just prior to NW., Washington.DC between 8:15 2b) may be different, but can be operation to check for openings) are m.m. end 5:00 p.m. weekde a calculated.

under consideration by the NRC staff.

Copies of draft regulatory guide MS 4 Defense.in-depth will be used for These would identify large breaches of 021-5 may be obtained from the Nulear protection assinst these levels of the containment system bounday.

Regulatory Commission. Document exposures. As the final barrier, a during or just prior to, normal operating l

Management Branch. Washington. DC containment system is required in order conditions. It should be noted they 20555 to maintain any or all of these exposure would only test the normal operating l

containment atmosphere boundary, not con FURTHER INF(WATION CO6.7ACT:

limits.

Mr. E. Cunter Arndt. Office of Nuclear

5. The required degree of containment the Appendix J post-accident boundary Regulatory Research. U.S. Nuclear system leaktightness for design basis including isolation valves. Comments on Regulatory Commission. Washington.

accider ts can be (a) calculated. (b) these or altemative concepts, and what DC W55, idephone (301) 443-7893 specified. (c) built. (d) maintained. (c) effeu,if any, they would have on the sumwentAny imponwateow:

inspected.

proposed Appendix J requirements, are

6. A generic inspection program can also being solicited in the following 8 N"#

be defined that verifies the required section of this preamble.

Appendix J of 10 CFR Part 50 was leaktightness of the containment Past practice has been to implement I

I onpndlly issued for public comment as following construction and periodically the provisions of Appendix l by means a proposed rule on. August 27.1971 (36 ihtoughout plant life.

of hcensees' technical specifications.

)

j i

Federal Register / Vol. 51. No. 209 / Wednesday, October 29, 1986 / Proposed Rules 39539 T

Currently, a Technical Specification (6)If the ex.isting rule or its proposed containment leaktightness on a more improvement Project (T5tp)is unde-way revision were completely voluntary, continuous basis than once every to reevaluate the NRC's philosophy and how many licensees would adopt either several years.The only existing or utilization of the technical version in its entirety and why:

proposed criterion for Type B and C specifications. While the proposed (7) Whether (a) all or part of the tests performed outside the outage in revision described herein assumes proposed Appendix l revisions would which a Type A tes!. is performed is that implementation of Appendix J by constitute a "backfit" under the the sum of Type B and Custs must not licensee's technical specifications, the definition of that term in the exceed 60% of the allowable work of the TSIp suo lead to some Commission's Backfit Rule, and (b) there containment leakage. Currently being changes in this form of implementation, are parts of the rule which do not discussed by the NRC staff are:

Another program is presently being constitute backfits, but which would aid

a. AllType B and C tests performed conducted to identify current NRC the staff. l'censees. or both; during the same outage as a Type A test, (6) Since the NRC is planning a or perforraed during a specified time regulatory requirements that have broader, more comprehensive review of period (nominally 12 months) prior to a

, arginalimportance to safety and to m

containment functional and testing Type A test, be factored into the recommend appropriate actions to modify or to eliminate these requirements in the next year or two.

determination of a Type A test "as j

whether it is thu still worthwhile to go found" condition.

unnecessary requirernents. A Federal R*gister notice was published on forward with this proposed revision as

b. If a particular penetration or valve October 3.1984, to announce the an interitu updating of the existin8 fa la two consective Type B or C tests, initiation of the program (49 FR 39066).

regulatkn:

the frequency of testing that penetration (9) The advisabihty of referencing the must be increased until two satisfactory As a part of the program. regulatory testing standard (ANSI /ANS 56.8) in the B or C tests are obtained at the nominal requirements associated with regulatory guide (MS 021-5) instead of in test frequency. Concurrently, existing cordainment leaktightness are being th tex of Appendi requirements to increase the frequency, evaluated. The risk and cost

{,

o Type A tests due to consecutive 'as effectiveness of containment g.*as found" condition of values and f und', failures are already being leaktightness requirements will,be.

seals and the need for acceptance relaxed in the proposed revision of examined to determine their value with respect to plant safety and possible Cdf%(e.r$e bical Appendix J. instead, attention would be (31 f cused on correcting cornponent alterna tive requirements.

spec fication limits on allowable degradation, r o matter when tested end Any resulting changes to existing containment leakage should be relaxed the "as found, Type A test would renect regulations will be made through normal and if so, to what extent and why, or if the actual condition of the overall rulemaking procedures.includmg ACRS not, why not; containment boundary.

review and public comment. Crmments (12) What risk-important factors

c. Increases or decreases in Type B or on the questions posed in this notice will influence containment performance C "as found" test results (over the also provide early, usefulinput to these under severe eccident conditions, to previous "as lef t Type B or C test associated activities.

what degree these factors are favitation To Comment considered in the current containment r;sults) shall be ag'ded to or subtracted from the previous as left' Type A test testing requirements, and what result.

Comments from allinterested persons arTroaches should be considered in if this sum exceeds 055 L but is less on all aspects of this revision and on the addressing factors not presently than 1.0 L. menures shall be taken to nsk and cost effectiveness of covered.

contairi:nent leaktightness in general are (13) Nhat other approaches to reduce the sum to no more than 0J5 L requested by the comment expiration validating containment integnty could

'!his will not be considered a reportable date in order that:(1) The final revision be used that might provide detection of condition.

will reflect consideration of all points of leakage paths as soon as they occur, If this sum exceeds 1.0 L. measurcs view; and (2)The staffs assessment of whether they would result in any shall be taken to reduce the sum to no the risk importance of cor.tainment adjustments to the Appendix l test more than 0J5 L This will be leaktightness can benefit from such program and why; considered a reportable condition.

comments. Especially requested arc (14) What effect " leak.before break" The existing requirements that the comments which address the following assumption could have on the leakage sum of allType B and C tests be no questions:

rate test program. Current accident greater than 0.00 L shall also remain in (1) The extent to which the assumptions use instantaneous complete effect.

positions in the proposed rule are breaks in piping systems, resulting in a mIw hen already in use; test program based on pneumatic testing (2) The extent to which those in use, of vented, drained lines. " Leak before.

The fo!!owing are the major changes and those not in use but proposed are break" assumptions presume that pipes proposed in this rulemaking.

desirable; will f ail more gradually, leaking rather

f. Levelof detoil. The level of detail (3) Whether there continues to be a than instantly emptyi.

addressed in the proposed revision of further need for this regulation:

(15) How to effective adjust Type A Appendix l has been limited.This i

(4) Estimates of the costs and benefits test results to reflect in ividualType B revision of the agulation defloes of this proposed revision, as a whole and C test results obtained from general containment system leakage test and of its separate provisions; inspections. repairs, adjustments, or criteria.

(5) Whether present operating plants replacements of penetrations and valves t Editoriol. For increased clarity, an l

or plants under review should be given in the years in between Type A tests.

expanded and revised Table of Contents the opportunity to continue to meet the Such an additional criterion. currently and set of definitions has been provided.

current Appendix ! provisions if the outside the scope of this propcsed conforming to current usate. The text proposed rule (reflecting consideration revision, would provide a more has also been revised to conform to of public comments) becomes effective:

meaningful tracking of overall

" plain English" objectives.

f

!L

39540 Federal Register / Vol. 51, No. 209 / Wednesday. Octobet 29. 1986 / Proposed Rules 9*

J /nterpretofions. Some changes have "as leIt" Type A test acceptance Rule, when it nevertheless finds broad

• been made to resolve past questions of criterien. leaving 0.25 of the allowable agreement that a rulemaking is in the interpretation (e.g., definitiore of leakage rate as a margin for public interest (e.g. in the case of

" containment isolation valves").

deterioration until the time of the next conversion of non-power reactors from

4. Createrflexibihty. A major regulatory scheduled Type A test, when HEU to IIU).

problem with Appendix j has been the the "as found" leakage rate criterion is lack of a provision for deahng with 1.0 of the allowable leakage rate.

The public may therefore wish to plants already built where design la Quontification of allowable romment directly on the question of

f. uiures are incompstib!c with,

/eoAage rotes. It should be noted that no whether the Commission should Appendix j requirements (e.g., air lock change has botn made to the way in continue its attempts to apply the y

t testing). As a result. provision has been which the allowable test leakage rates

,'heth r the mude in this revision for consideration are quantified. The regulation still refers ule s o. ld b voked as by the NRC staff of alternative leakage to the individual plant technical it applies to rulemaking activity per se.

test requirements when necessary.

specifications for these values. Debate

/. alternatively, the public may wish to

& Type A testpressure. The option of continues. however. on what these consider whether the Commission performing periodic reduced pressure values should be and whether they can should amend the Backfit Rule to waive testing in lieu of testing at full calculated be generically specified, rather than the "s ubstan tial increa se" provision.

accident pressure has been dropped.

individually specified for each site and and to indicate explicitly that non.

This change reflects the opinion that pla n t.

monetary benefits may be weighed by extrapolating low pressure leakage test

11. Refocusing of corrective actions.

the Commission in the cost-benefit results to full pressure leakage test When a reportable problem is identified. balance. when such considerations are tcsults has turned out to be a Corrective Action Plan is to be found by the Commission to be in the unsuccessful. Reasonable argument can submitted. It identifies the problem to public interest.

be made for low pressure testing.

the NRC staff. and notes the cause, what Howeser. the NRC staff beheves that was or will be done to correct it, and Finding Of No Significant Environmental the peak calculated accident pressure what will be done to prevent its impact: Availability (a) has always been the intended recurrence.

The Commission has determined reference test pressure. (b)is consistent increased local Fiage testm.g with the typical practice for NRC staff frequency may be necessary. Appendix l under the National Environmental pobey escluations of accident pressure for the originally addressed increased test.

Act of 1969, as amended. and the ferst 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with frequency only for Type A tests.This Comenission's regulations in Subpart A Regulatory Guides 1.3 and 1.4. (c) revision applies adjustment of test of 10 CFR Part St. that this rule, if provides at least a nominal check for frequency directly to identified problem adopted, would not be a major Federal gross low pressure leak paths that a low areas.

Action significantly affecting the quahty f'"

pressure leak does not provide for high

12. Thefino/porograph of the of the human erwironment and therefore

(

pressure leak paths. (d) directly proposedomendment speci/tes a date an environmentalimpact statement is rerresents technical specification by a hich on implementation schedule not required. There will be no leakage rate limits and (c) provides must be submitted, rather than by which radiological environmentalimpact greater confidence in contaminant it must be impien.ented This is because offsite, but there may be an system leak tight integrity. For these the ease with which licensees will be occupational radiation exposure onsite reasons. the full. rather than reduced, able to implement all the provisions of of about 3.0 man. rem per year of plant pressure has been retained as the test the amendment will be highly plar' operation for inspection personnel pressure.

specific depending on plant design.

(about 0 4% increase). A'ternatives to ti Tjpe A testfrequency. The test outage and testing schedules and issuing this revision were considered frequency has been uncoupled from the amount of technical specification and found not acceptable. The 10 year inservice inspection period used changes needed.

environmental assessment and finding by the ASME Boiler & Pressure Vessel The separate views of Commissioner of no significant impact on which this Code for mechanical systems. A Frederic M. Beenthal follow:

determination is based are available for different time base is used. but the The public should be aware of the fact inspection c' the NRC Public Document frequency has remained essentially the that the Commission for over a year has Room.1717 H Street NW., Washington, sa me.

attempted to adapt the Backfit Rule to Z T)pe A tcst duration. The duration o// rulemaking. even rulemaking that has DC. Single copies of the environmental hds been dropped from the test criteria Dolhing to do with changes to assessment and the finding of no m Appendix J !!is considered as part of powerplant hardware and the original significant impact are available from Mr.

the testing procedures. and is a function intent of the Rule.

E. Gunter Arndt. Office of Nuclear of the stateof the testing technology and This rulemaking and the Reguletory Research.

the level of corifidence in it.

accompanying analysis illustrates the U.S. Nuclear Regulatory Commission.

B T pe A test "os is"c/arification, difficulty. When apphed to human.

%,ashington. DC 20555. Telephone (301]

J Appendix l originally noted in Ill.A.1(a) factors rules. updating antiqua ted rules.

443-7893.

that the containment was to be ".. ~.

and certain other rulemaking. the Paperwork Reduction Act Statement inted in as close to the 'as is' condition Backfit Rule continues to exact NRC as practical." This is re-emphasized and resources wholly disproportionate to This proposed rule amends clarified by the esplicit requirements any conceivable benefit to the pubbc.

information collection requirements that that have been added to measure.

The record ahedy shows cases where are subject to the Paperwork Reduction record. and report "es found" and *as the Commission has been forced to Act of1980(44 U S.C. 3501 et segl This left" leakage rates sidestep a strict readmg of the cost.

rule bet han submitted to the Office of P Type A test al/oirable leoAope rote benefit em:!m.. tnt..nd the =..

Management and Budget for review and ptrmtmg Seventy.five percent of the substantialincrease in overall approval of the paperwork atiowable leekage rate represents the protection. " threshold of the Backfit requirements.

a Federal Register / Vol. 51, No. 209 / Wednesday. October 29,1980 / Proposed Rules 39541

' Regulatory Analysis the Small Business Size Standards set Appendix F14akage Tests for The Commis Mn has prepared a draft out in regulations issued by the Small Containments ofIJght Water-Cooled regulatory analyh en the proposed Business Administration at 13 CFR Part Nuclear Power Plants revision.The analysis examines the 121.

I. Introduction costs and benefits of the alternatives 1.t t of Subjects in 10 CFR Part 50

!!. Definitions s

con:Idered by the Commission. The Ill. General Leak Test Requirements Antitrust. Clas. ified information, Fire draft analysis is available for inspection s

A. Type A Test and copyina in the NRC Public prevention. Incorporation by reference.

1. Preoperational Test Document Room.1717 H Street. NW.,

Intergovernmental rela tions. Nuclear

2. periodic Test Washington DC.The Commission power plants and reactors. Penalty,

3. Test Frequency requests public comrnent on the draft Radiation protection. Reactor siting 4 Test Start and Finish analysis. Comments may be submitted criteria. Reporting and recordkeeping

5. Test Pressure to the NRC as indicated under the requirements.

8. Pretest Requirements AponEssts heading.

Relsted Regulatory Guide 7

cau n Tut

8. Acceptance Criteria BacMit Analysis

.!he notice of availability of a draft

9. Restesting The Commission has prepared a regulatory guide on the same subject to. Permissible Periods for Testing backfit analysis on the proposed

" Containment System Leakage Testing ~

B. Type B Test revision.The analysis is required under (MS 021-5)is also being published

1. Frequency to CFR Part 50. I 50.109. as of October elsewhere in this Federal Register, *Ihe L Pru sure L Air eks 21,1985, for the management of draft regulatory guide contains specific backfitting for power reactors. The guidance on acceptable leakage test 4 Acceptance Criteria analysis is available for inspection and methods, procedures, and e.nalyses that 1

copymg in the NRC Public Document may be used to implement these

2. PQ""e) Medium Room.1717 H Street NW., Washin6 ton, requirements and criteria..

3. Acceptance Cnieria DC. The Commission requests public For the reasons set out in the

4. Valves That Need Not Be Type C comment on the analysis. Comments preamble and under the. authority of the Tested may be submitted to the NRC as Atomic Energy Act of1954, as amended.

IV. SpecialI4ak Test Requirements inflicsted under the ADOnESSEF heading. the Energy Reorganization Act of1974, A. Containment Modification or The analysis does not conclude that as amended, and 5 U.S.C. 653, the NRC Maintenance there is a substantialincrease in the is proposing to adopt the following B. Multiple taakage Barriers or overall protection of the public health amendments to 10 CFR Part 50.

Substmospheric Containments and safety or the common defense and V. Test Method. Procedures and Analyses

/~,

security to be derived from the backfit.

PART 50--DOMESTIC LICENSING OF A Type A.R.and CTest Details t.

It does conclude, however, that the PRODUCTION AND UTILIZATION B. Corchination of Periodic Type A. B. and direct and indirect costs of FACILITIES C Tests implementation are justified due to V1. Reports better more uniform tests and test 1.The authority citation for Part 50 A. Submittal reports greater confidence in the continues to read as follows:

B. Content reliability of the teat results, fewer Authority Secs.103.104,161.181183.186-

.Ap bi ity exemption requests, and fewer 189,68 Stat. 936. 937,948. 953. 954, 955. 956, as B. Effective Date interpretive debates. For these reasons.

amended, sec. 234. 83 Stat.1244, as amended which are presented in greater detallin (42 U.S.C 2133. 2134,2201. 2232,2233,2236 1.latroduction the backfit analysis, the Commission 2239,2282); secs. 201, 202. 206. 88 Stat.1242.

One of the conditions of all operating has decided to proceed with publication 1246. se amended (42 U.S C 5841,5842. 5846).

licenses for light. water-cooled power of the proposed rule for comment.The unten otherwise noted.

reactors as specified in i So.54(o) of this part Commission's decision regarding Sr.ction 505 also issued under Pub. L 95-is that primary containment meet the leak promulgation of the rule, even though it 601. sec.10. 92 Stat. 2951 (42 U.S C 5851).

test requirements set forth in this appendix.

may not provide a substantialincrease Sections 50.58. 50 91, and 50.92 also issued The tests ensure that (a) leakage through the in the overall protection of the public under Pub. L 94-415,96 Stat. 2073 (42 U.S C primary containments or systems and health and safety or the common 2239). Section 5038 also luued under sec.

components penetrating these containments defense and security,is tentstive 122. 6a Sta t. 939 (42 U.S.C. 2152). Sectiona does not exceed s!!owable leanage rstea pending receipt of public commenta on 50.80-50.01 also inued under sec.184. es Stat.

specified in the Technical Specifications and this issue.

954. as a mended (42 U.S.C 2234). Sections (b) inservice inspection of penetrations snd 50.100-50.102 also issued vader sec.186. 68 isolation valves is performed so that proper Regulatory Flexibility Certifkation Stat. 955 (42 U.S.C. 2236).

maintenance and repairs are made during In accordance with the Regulatory For the purposes of sec. 223. 68 Stat. 958, as their service life. This appendix identifies the Flexibility Act of 1980. (5 U.S.C. 605(b)).

amended (42 U.S.C 2273); M.10 (a). (b). and generai requirements and acceptance criteria the Commission certifies that this rule (c). 50 44. So 46. 50.4a. 50.54, and 50.00(a) are for preoperational and subsequent periodic will not,if promulgated, have a issued under sec.161b. 68 Stat. 948. as leok testing.

significant economic impact on a amended (42 U.S.C 2201(b)); 50.10 (b) and (c) substantial rmrnber of small er,tities.

and 50.54 are inued under sec.1811,68 Stat.

a specine guidene, concemins accepiable gis propned rule affects only the 649, as amended (42 U.S C 2201(ij); and leakase test method. procedures. and analyses that i

licensine and operation of nuclear so.55(el. 50.59(b). 5050. 5031. 5032. 50J3. and may be used to implernent these requirements and

{

5018 are n.eued under uc. telo. Se Stat. 9so-cniens will be provided in a resulatory ruide that is j

pt,wer p an.s.The compan!es that own bems issued in draft form for pubbe commeni with

[

these plants do not fall w: thin the scope as ammded (42 U.S C 22et(o)).

th, de,,n.iion us ars-a. copies of the resulatory ei the defmition of"sma'l entities" set

2. Appendix J to Part 50 is revised to

'N,',#,$','",',d[ men,Y

c forth in the Regulatory Flexibility Act or read as follows:

wahansion, oc uss.

j t

i p*

39542 Federal Register / Vol. 51, No. 209 / Wednesday October 29. 1988 / Proposed Rules t

IL Definitions pressure p., obtained from testing the StructuralIntegrity Test e ntauunut systm in the state as clou as A pneumatic test that demonstrates the

=

Acceptante Cr/terio practical to that that would exist under capabihty of a primary containment to Standards against which test resulta are to design basis accident conditions (e g.. vented, withstand a specined internal design be compared for estabbshing the functaonal drsined, floodn!. or pressurized).

pressure load.

acceptabihty of the contamment system as a LeoA led age imut.ng bounda ry.

Type A Test AD opening that aUows the passage of a A test to n'eesure the containment eystem "A 9 Ted"/.rnAnga Rate

'" N I""8

' 8' " '*

The leakage rate prior to any needed con &uns npnsW dugn ksh W.

LeoAage repairs or adjustments to theleakage barrier coelant accident contairunent prepare and bems tested The quantity of fluid escaping from a leak.

cystems alignments (t) after the containment "As Left"LeoAage Rote LeoAege Bote system has been completed and is ready for The leakage raie follewing any needed The rate at which the contained fluid operation and (2) at periodic intetvals thereafter. The ver.fication test la r.ot part of repsirs or ediustments to the leakage barrier escapes from the test volume at a specified this definition-see CHRT.

t> ting tested test pressure.

Contornment Intepmted Leak Rote Test bloximum 1+thway Leakage Bote Type B Test (C/LRTl The maximum leskoge rate that can be A pneumatic test to detect and rnessure The combmation of a Type A test and its attributed to a penettstion leakage path (e s.,

local leakage through the following verifwat on test.

the larger, not total, leal. age of two valves in a20tainment penetrations:

series). T}us generally assumee a single O)Those whose design incorporates ontainment /solotson System fuochonal actin failure of the better of two leakage resilient seals, gaskets, sealant compounds, barriers in series when performing Type B or aKpansion beUows or Lttad with Cexible metal seal suembhes, A tal to venfy the proper perfomiance of C testa, (2) Air locks. includ ng door nah and door the isolahon sy tam by norrnal operation of Afinimum Pothway leologe Rote operating mechanism penetrations that are the vehes For automatic containroent isolaticn n> stems.a test of the automatic The minimum leakage rate that can be part of the containment pressure boundary.

isolation system performed by actuation s attributed to a penetrahon leakage path (eg,

Tjpe C Test the contamnient isolation signals.

the smallest leakage of two valves in aerses).

T}us is used when co'rectmg the sneasured A pneumatic teet to measure containment Containment / solation Volve value of containment leakage rate from the isolation valve leakage rates. Fenhecuoo Any v alve defined in General Design Type A !est (I.) to obtain the overau Test Cnteria 55. 56. or 57 of Append;x A " General integrated leakage rete and generally Test to conntm the capability of the Type Design Criteria for Nuclear power plants." to assumes no single active f ailure of redundant A test method and equipment to musure l this part leakage barriens under these test conditions.

IH. General 14ak Test Regidrements Containment LeoA Test Prvsmm OverallIntegrated Leakage Rate A

Tut The comprehensive testing of the The totalleakage rate through allleakage 0)pn perationalTest A preoperational containment system that includes Type A. B.

paths, includmg containment welds, valves, T)Te A test must be conducted on the and C tests.

fitt ngs, and components that penetrate the containment system. expressed in units of

[ ntainment system and must be preceded Containment Splem 8

a) Type B and Type C tests, The principal barrier. after the reactor pr ss r per 2 ho r (b) A structuralintegrity tot.

coolant pressure boundary, to prevent the

[2] Period c Test A peno&c Type A test wiease of quantities of rs&osebve matenal Pu (Ps/g/

must be performed on the antaintnent that would have a significant radiological The calculated peak containment intemal 8Fd**

effect on the health of the public.it includes; pressure related to the design basis loss.of-(3) Test Freqmcy. Mus a longer totmal 0)The prunary contairunent. including coolant accident as specified In the technical is specihcaUy approved by the,NRC stafL the access openings and penetrabons.

spec 16ca Uons, interval between the preopersuonal and Erst (2) Containment isolation v alves. pipes, Periodic LeoA Test periodic Type A tests must not exceed three closed systems. and other components used years, and the interval between subsequent to effect isolation of the containment Test conducted during plant cperating periodic Type A tests rnoet not exceed four atmenhere from the outside envfrons. and hfetime.

years. If the initial fuelload ng la deleyed so (3)Those systems of portions of systems Preoperationa/LeoA Test that the three yearinterval between the first that by their functions extend the primary containment boundary to include their Test conducted upon completion of preopera tional test and the first penodic test system boundary.

construchon of a primary or secondary is exceeded, anoth:r preoperational test wdl This definit.on does not include bolling contaitunent, including installa tion of be necessary. lf such an ad&uonal water reactors' (BWR) reactor buildtngs or mechanical Deid, electrical and preoperational Type A test or en additional pressunted water reactors'(pWR) shield instrutnentation systems penetrating these Type A lest required by Sechen Ifl. A.8 or buildmgs Also excluded from the provisions containment systems. and prior to the tirne fV.A. of this appen&x is performed. the Type of this appendix are the interiot barriers such containment integrity Is required by the A test interval may be restarted.

(4) Test Pressure.The Type A test preseure ce the BWR Mark 11 drywell floor and the Technical SpecjLcations.

must be equal to or greeter than p, at the drywell perimeters of the BWR Mark III and PN"C'68"IdlA*8"'

start of the test but mu.t not exceed the the PWR ice condenser.

The structure or wewel that encloses the containment design pressure and must not L (weight percent /24 hr) major components of the reactor coolant fall more than 1 psi below P. for the duration The maximum allowable Type A test pressure boundary as defined in 150.2(v) of of the test. not inclueng the varincation seat le kage rate in units of weight percent per 24 this part and is designed to contain accident The test pressure must be estabhahad relative bour period at pressore p, es specified in the presaure and sene as a leakage barner to the external pressure of the containment Technical Specifications.

against the uncontrolled release of This may be either atmospheric pressure or radioachvity to the environment The term the substmospheric pressure of a secondary L. (weight percent /24 hr)

"contamment" as used in this appendix refers contelnment.

The measured Type A testlaakage rate in to the primary containment structure and (5) Pretest Requirements. Closure of l

units of weight percent per 24-hour pened et associated leakage barriers.

containment isolation valves for the Type A L

Federal Register / Vol. 5L No. 209 / Wednesday. October 29. 1988 / Proposed Rules 39543 test must be accomplished by normal submitted to the NRC staff for review and (a) Initial and periodic tests. Air locks must operation and without any preliminary approval. An as left Type A test that meets be tested prior to initial fuelloading and at f

exercising or adjustments for the purpose of the acceptance criterion of 0.751,ls required least once each 6-month interval thereafter at i

improving performance le s. no tightening of prior to plant startup.

en intemal pressure not less than P.,

valve after closure by valve motor). Repairs (b)If two consecutive periodic as found Altematively. lf there have been no air lock of malfunctioning or leaking vabes must be Type A tests exceed the as found acceptance openings within 6 months of the last made as necessary. Infonnation on valve enterion of 1.014 successful test at P., this interval may be leakage that requires corrective action prior (i) Regardless of the pciodic retest extended to the next refuefmg outage or to. durmg, or after the test (see Section V.B.)

schedale ofIll.A.(3).a Type A test must be airlock opening (but in no case may the must be included in the report submitted to performed at least every 24 months (based on interval exceed 2 years). Reduced pressure the Commission as specified in Section VI of the refuelit.g cycle normally being about 18 tests must continue to be performed on the this appendix.

months) unless an attemative leakage test air lock or its door seals at 6-month intervals.

(6) Venfication Test. A leakage rate program is acceptable to the NRC staff on Opening of the air lock for the purpose of i

venfication test must be performed after a some other defined basis.This teskg must removing att lock testing equipment following Type A test in which the leakage rate meets be performed until two consecutive periodic an air lock test does not require further the criterion in Ill.A.(7)(b)(ii). He verification "as found" Type A tests meet the acceptance testing of the airlock.

test selected must he conducted for a criterion of 1.0L siter which the retest (b) intermediate tests must be conducted as duration sufficient to establish accurately the schedule specified in IILA.(3) may be change in leakage rate between the Type A resurned.

follows:

and verification tests.ne results of the Type (ii) Investigation as to the cause and nature (i) Air locks opened during periods when A test are accepteble if the sum of the of the Type A test failure might indicate that containment integrity is required by the venfication test imposed leadkage and the an attemative leakage test program such as plant's Technical Specifications must be containment leakage rate calculated from the more frequent Type B or Type C testing may tested within 3 days after being opened. For Type A test (b) does not diffet from the be more appropriate than the performance of air lock doors opened more frequently than leakage rate calculated from the venfication two consecutive successful Type A leakage once every 3 days the air lock must be tested test by more than 10.25 I,.

testa.The licensee may then submit a at least once every 3 days dunng the period (7) Acceptance Criteria.

Corrective Action Plan and an alternative of frequent openings. Airlocks opened during (a) For the preoperational Tme A Test, the leakage test program proposal for NRC staff periods when containment integnty is not "as left" leakage rate must not exceed 0.75!,.

nyiew. If this submittalis approved by the required by the plant's Techrucal as determined by a properly justified NRC staff. the licensee may implement the Specification need not be repeatedly tested statistical analysis. The "as found" leakage corrective action and alternative leakage test during such periods. However, they must be rate does not apply to the preoperational test.

program in lieu of one or both of the Type A tested prior to the plant requiring (b)For each peno&c Type A test, the leakage tests required by Section containment intYty. For air lock doors leakage rate, as determined by a properly III.A.(8)(b)(i).

having testable seals, testing the seals fgr;;;e jushfied statistical analysis, must not exceed:

(9) Permissible periods for testing The the intermediate test requirements of tlu. s (i)1,. for the "as found' condition.

performance of Type A tests must be limited paragraph. in the event that this intermediate (ii) 0.751,. for the "as lef t" cenition, to periods when the plant facility is secured testing cannot be done at p., the test (c)In meeting these Type A test acceptance in the shutdown condP. ion unoer the pressure must be as stated in the Technical criteria, isolation repalr.or adjustment to a administrative controls and safety I

Specifications.

leakage barrier that may affect the leakage procedures defined in the license.

(ii) Whenever maintenance other thar. on rate through that barrier is permitted prior to B Type B Test door seals has been performed on an air lock, or during the Type A hst provided:

a complete airlock test at a test pressure of li) Allpotentialleakage paths of the (i) Frequency, isolated, repaired, or ed usted leakage barrier (a) Type B tests, except tests for air locks.

not less than F is required. lf that are locally leak testable, and enust be performed or containment maintenance involved the peessure teteining i

(ii) the localleakage rates are measured penetrations during shutdown for refueling of boundary.

before and af er the isolation, repair, or at other convenient intervals but in no case at (iii) Air lock door seal testing or reduced-t adjustment and are reported under Section VI intervals greater than 2 years if opened pressure testing may not be substituted for of this appen&x.

following a Type A or B test. containment the initial or period 2e full. pressure test of the (iin All changes in leakage rates resulting penetrations subject to Type B testing must entire air lock required in paragraph from isolation, repair or adlustment of be Type B tested prior to returning the reactor Ill.B.(3)(a) of this Sectier..

leakage barners subject to Type B or Type C to en operating mode requiring containment (4) Acceptance Criteria.

aesting are determined using the minimum integnty.

(a)ne sum of the as found or as left Type pathway leakage method and added to the (b) For containment penetrations B and C test results must not exceed 0 601 Type A test result to obtain the "as found' etnploying a continuous leakage monitoring using maximum pathway leskage and ana "a s left" containtnent leakage rates.

system that is at a pressure not less than P.,

including leakage rate readmgs from (d) The effects of isolation, repair, or leakage readings of sufficient sensitivity to continuous leakage monitoring systems, adjustments to the containment boundary permit comparison with Type B test leak (b) taakage measurements are acceptable made after the start of the Type A test rates must be taken at intervals specified in if obtained through component leakage sequence on the Type A test results must be the Technical Specifications. These leakage surveillance systems (e.g continuous quantified and the appropnate analytical readings must be part of the Type B reporting pressurization of mdividual or clustered corrections made (this includes tightening of VI.A. When practical, continuous leakage containment components) that maintain a valve stem poching. additional tightening of monitoring systems must not be operating or h P" e indMdul ust P" " ""

manual valves, or any action taken that will pressurized during Type A tests.If the chambers of those same containment affect the leakage rates).

continuous leakage monitoring system cannot penetration during normal reactor cperation.

be f oolated, such as inflatable air lock door (a)If for any periodic Type A test, the as seals. leakage into the containment must be Similar penetrations not included in the (8) Retesting.

food leakage rate fails to meet the accounted for and the Type A test results component leakage surveillance system are J

still subject to individual Type B tests.

seceptance criterion of 1.0I.. a Corrective corrected accordmgly.

(c) An air lock, penetration, or, set of J

Action Plan that focuses attention on the (2) Pressure. Type B tests must be cause of the problem must be developed and conducted whether individually or in groups.

penetrailons that fails to pass a Type B test must be relested following determination of implemented by the licensee and then at a pneumatic pressure not less than P, cause and completion of corrective action.

submitted together with the Containment except a s provided in paregraph Ill B.(3)(b) of Corrective actiori te correct the leak and to leak Test Report as required by Section VI of this section or in the Technical prevent lis future recurrence must be this appendix.The test schedule apphcable to Specifications.

outssequent Type A tests (111.A (3)) shall be (3) Air locks.

developed and implemented v

l 1

39544 Federal Register / Vol. 5L No. 209 / Wednesday. October 29, 1986 / Proposed Rules f

(d)1ndividual acceptance criteria for all air attachments to the surface of the steel B and C tests included as a part of the Type i

lock teste sust be stated in the Technical pressure retaining boundary; Repair cavities A test sequence muel also be included in tha the depth of which does not penetrate the reqort.

Speerfications.

required design steel wall by more than tot VII. Application 1

1 C. Type W Welds attachms to the steal pressure (1) Frequency. Type C tests must be retaining boundary penetrations the nominal A. Applicability l

4 performed on containment isolation valves diameter of which does pot exceed one lach.

The mquinments of We oppendix appWo all operating nucleer power reactor licensees l

daring each reactor shutdowm for refueling or B. Ms/lipleicoAope Sorrr.er or as specified in i 50 54(o) of this part unless it

(

at other convenient intervals but in no case at Subormospheric Containments can be demonstrated that alternative leak intervals greater than 2 years.

The primary reactor containment banier of test nquinments (e g.for censin (t) Pressure / Medium.

(a) Containment isolation valves unless a multiple t.arrier or subatmospheric containment designs. leakage mitigation pressurized with a quahfied water seal containment shall be subjected to Type A systems, or different test prasures not system must be

-- ' arith air or test to verify that its laakage rate meets the specifically addressed in this appendix) are retropen at a pre'eame not less than P requirements of this appendix. Othat demonstrated to be adequate on some other (b) Containment isolation valves, that are s'ructures of multiple barrier or defined basis. Altemative, leak test essled with water from a qualifml seal subatmospheric containment (e.g.

requirements and the basis for them willbe syoem, mud be tested with wster at a secondary containment for boiling water made a part of the plant Technical pnesure not less than 1.10 P,.

reactors and shield buildings for pressurized Specifications if approved by the NRC staff.

water reactors that enclose the entire primary (3) Acceptance Criteria.

reactor containment or portions thenof) aball B. Effective Date (a)The um of the as found or as left Type B and C test results must not exceed 0.601, be subject to individual tests in accordance using maximum pathway leakage and with the procedures spedfied in the technical

• 'gogi'g* $'

• gI including leakage rate readings from specifications.

g g

inm7n is V.Tast Methodai-_ ' w.and Analyesa revision), each licensee and each app!! cant valves that are sealed with water from a seal A Type A B. and C Test Deloils

[*D e enjo on "p

een e a su mi (b) akage om co tsak tut methods, proceduna, and Regulation for implementmg this appendix.

e e m ined Type B e d en ge a e analyses for a steel, conesee, er combination This submittel must inclode an (i) The valves have been demonstrated to s' eel and o nente e ntainment and its implementation schedule, wtth a final have leakage rates that do not exmed those Penetrations and isolation valves for hght-impitmentation no later than (insert a date 48 specified in the Techrucal Specification, and water-c led power reactors must be months after the effective date of this i ) The installed isolati tv af nced or defined in the Technical revision). Until the licensee finally systern imentory is suffician o ensure the "I'"gicat;ons.

Implements the provisions of this revision.

Speci the licensee shall continue to use in thetr suling function for at leaat 30 days at a B. Combination ofPeriodic 7)pe A S. cad c entirety the existing Technical Specifications pressure of 1.10 P.,.

(4) Vahos That Need Not Be Type C Tests and the Appendix j on which they are based.

,[

Tested

!l\\

(a) A containment isolation valve need not co d t d in o unct th pe c

CII *

"8 '

f' does t constit te po n al

,'g

,"g,k,.hperio Dated at Washington, DC, this 22d day of A te n e containment atmosphere leak path during or perform. record. interpret. and report the tute October.1986.

following an accident, considering a single in such a manner that the containment For the Nuclear Regulatory Commisson.

active failure of a system component.

system leak t:ght status is determined en (b) Other valves may be excluded from both an as found basis and an as left basis.

Samuell. ChE Type C testing only when approved by the i.e Its leak status prior to this periodic Type

1. 1. 1. 1. #I& Co""i#

NRC staff under the provisions of paragraph A test together with the eetated Type B and C [FR Doc. 86-24496 Filed 10-28-ee. 8 45 am)

V!1.A.

tests and its status following the conclusion am Coot 76806 of these tests.

iv.SpacL114ak Test Requirements A. Containment Modification or Maintenance VI. Reporta Any modification, repair. or replacement of A Submitic/

DEPARTMENT OF TRANSPORTATION e component that is part of the containment

1. The preoperst onal and periodicType A system boundary and that may affect tests. including summaries of the results of Federal Aviation Administration containment integrity must be followed by Type B and C tests conducted in conjunction either a Tyle A. Type B.of Type C test. Any with the Type A test. miust be reported in a 14 CFR Part M modAcation. repair. or replacernent of a summary technical report aant not later than component subject to TWe B or Type C 3 months after the maduct of sedi tast to the

[ Docket No. BFCt4-AD) testing must also be preceded by a Type B or Commission in the manner spedfied in i 50.4.

Type C test.The mrasured leakage from this The report is to be titled " Containment Altworthiness Directives; 5eects M test must be included in the report to the g,,k og, tuty an

,y Commission required by Section VI of this

2. Reports of periodic Type B and C tests appendix. Following struceural changes or conducted at intervale intermediate to the Actsecy: Federal Aviation repairs that affed the pressure boundary. the Type A tuts must also be submitted to the Adtninistration (FAA). DOT.

licensee shall demonstrate whether or not a NRC in the manner specifwd in 1604 and at Notice of Proposed Rulesnaking structural integnty test is aseded prior to th*

the time of the next Type A test submittal.

ACTtON:

next Type A test.The earptance criteria of Reporte mut be submined to the NRC (NPRM).

paragrsph 11LA.[7). [11.844), or 111 C.(3) of this Regional Admuustrator within 30 days of sueMaAsm This Notice proposes to appenda, as appropnete, must be saet. Type complehan of any Type B or C tests that fail A testing of certain minor snoddications, to meet their as found acceptamoe antaria-adopt a new Airworthiness Drrective (AD). applicable to Beech 99 and100 repairs.or replacements may be deferred to 3 Content Series airplanes.He Proposal would

• be next regularly echaduled Type A tant if A Type A test Corrective Action Plan

• require inspecilon and repIacernent of locaj leakage lesimg is not posaihle and when required under paragraph 11LA.(4) of rivets which attach each elevator vaual(leakage) summaboas or nondestructive emananations have been this appendix, must be included ia the report.

outboard hinge to the alahilirer. Loose conducted These shall include; Welds of Any corrective action required for those Type

___