ML20202G480
| ML20202G480 | |
| Person / Time | |
|---|---|
| Issue date: | 03/06/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR ACRS-2402, NUDOCS 8607150401 | |
| Download: ML20202G480 (15) | |
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\lgj!LgO $ N UN O O( ~# ISSUED: MARCH 6, 1986
SUMMARY
/ MINUTES OF THE ACRS METAL COMPONENTS SUBCOMMITTEE MEETING FEBRUARY 27-28, 1986 WASHINGTON, D.C.
The ACRS Subcommittee on Metal Components met in Washington, D.C. on February 27-28, 1986 to review the 1) proposed broad scope rule change to GDC-4 (the application of leak-before-break concept to all high energy piping systems of nuclear power plants), 2) technical report on material selection and processing guidelines for BWR coolant pressure boundary piping (draft NUREG-0313, Rev. 2) and 3) format and content of plant-specific pressurized thermal shock safety analysis reports for pressurized water reactors (draft Regulatory Guide, Task SIO2-4).
Notice of the meeting was published in the Federal Register on February 7, 1986 (Attachment A). The schedule of items covered in the meeting is in Attachment B. A list of handouts kept with the offica f copy of the minutes is included in Attachment C. The meeting was entirely open to the public. There were no written or oral statements received or presented by members of the public at the meeting. E. Igne was the cognizant ACRS Staff member for the meeting.
Principal Attendees ACRS P. Shewmon, Chairman D. Ward, Member H. Etherington, Member W. Kerr, Member M. Bender, Consultant I. Catton, Consultant E. Rodabaugh, Consultant T. Kassner, Consultant NRC Presenters Other Presenters J. O'Brien R. Cloud, R. L. Cloud Assos., Inc.
R. Bosnak J. McInerny, Westinghouse W. Hazelton T. Chang, Westinghouse R. Woods S. Bernsen, AIF C. Johnson D. Norris, EPRI J. Reyes M. Vagins N. Randall c 3Ic"Arn 02:GI a 8607150401 860306 Certified By PDR ACRS 2402 PDR
METAL COMPONENTS 2 Feb. 27-28, 1986 Meeting Proposed GDC-4 Modification (Broad Scope Rule)
J. O'Brien, RES, discussed this matter. This proposed rule allows application of leak-before-break (LBB) technology to all high energy (275 psi or 200 F) fluid system piping to demonstrate that specific pipe ruptures need not be treated in the design basis. The rule depends on advanced fracture mechanics techniques which have been experimentally validated and include evaluations of water hammer, corrosion, leak detection and indirect sources of pipe rupture. All reactor piping in all reactor types which satisfy rigorous acceptance criteria can take advantage of the rule. Only dynamic effects associated with pipe ruptures are excluded from the design basis. Containment design, ECCS performance and environmental qualifications are not impacted. The removal of pipe whip restraints, jet impingement barriers and other related facility design changes in operating plants, plants under construction and future designs are permitted. Averted worker radiation exposures are measured in several (possibly many) 10,000's of man-rem.
Cost savings are measured in several (possibly many) $100 millions.
Public safety is believed to be enhanced.
It was stated by the NRC Staff that the detailed acceptance criteria for the selection of piping systems and their basis to which the broad scope rule will be applied is still being developed. The subcommittee, during the meeting, presented the following comments on application of the broad scope rule:
The rule should not be applied in the foreseeable future to piping systems operating at temperatures above 650 F~because of possible degradation due to creep mechanism.
In evaluating the piping systems the NRC Staff should consider the consequences of a crack opening along a weld. This is a credible failure mode and si. auld be considered.
B METAL COMPONENTS 3 Feb. 27-28, 1986 Meeting The material properties and fabrication processes of the weld in piping systems where LBB criteria are applicable should be given at least as much consideration as properties of the base metal of the piping system.
The LBB argument may also be usefully applied to matters related to ECCS, containment design and equipment qualification with similarly accompanying benefits.
The following are foreign practices with respect to the use of the broad scope rule.
UK: NII shows a strong inclination to reject LBB for Sizewell
! based on concerns with stress corrosion cracking and inadequate NDE of cast stainless steel piping and components. (CEGB at odds with NII on this.)
4 ,
France: Undecided, but weakly inclined to reject LBB at this time
- partly because of commitment to standardization, although research on LBB in progress.
FRG: Strong commitment to LBB in new PWR main coolant, main feed and main steam line inside of containment.
Italy: Close to FRG practices.
Japan: Inclined to accept LBB, even in BWRs. Heavy investment in LBB research.
Canada: Inclined to accept LBB for certain piping systems at the Darlington facility.
R. Bosnak, NRR, discussed plans for the implementation of the limited and broad scope proposed rule to GDC-4. With respect to the limited
METAL COMPONENTS 4 Feb. 27-28, 1986 Meeting scope rule, schedular exemptions have been granted to all recently licensed plants covering protection against dynamic effect, i.e., pipe whip restraints, jet impingement barriers, loads in unbroken portion of loop and branch lines, and pressurization transients. No change is permitted in containment design, ECCS and equipment environmental qualification. Recently, Crystal River-3 and Surry have requested that downsizing /or removal of large snubbers required to resist pipebreak loads is possible with an overall result of improved system performance and reliability. The Staff indicated that an independent design and fabrication assurance is a prerequisite.
The use of LBB applied to all high energy lines (broad scope rule) has been requested by lead plant Beaver Valley-2 in their Whipjet program.
Others are expected to submit their programs to use the broad scope rule. R. Bosnak stated that the acceptance criteria (Regulatory Guides, SRPs) will be developed following public comments, but should incorporate the following major items:
Pipe rupture probability should be extremely low a 10-6 ,
Alternatively, a deterministic evaluation with verified design and fabrication, and adequate ISI is necessary.
Leakage detection systems should be reliable, redundant, diverse and sensitive. (This requirement is more difficult to apply to pipingsystemsoutsideofcontainment.)
A margin of at least 10 on detection of leakage from through-wall flaws should exist.
In conclusion, R. Bosnak stated that 1) removal of dynamic pipe rupture protective devices and deletion of large dynamic pipe rupture loads is beneficial, 2) decoupling of SSE and LOCA is acceptable and 3) use of code allowable stresses is sufficient.
METAL COMPONENTS 5 Feb. 27-28, 1986 Meeting ACRS input and comment are sought by the NRC Staff in the following areas:
Guidance in the assurancr. of extremely low probability of pipe rupture, i.e., levels and failure modes.
If probability is above a given threshold level, what, if any, methods of protection should be used.
Need for inservice inspection augmentation.
Design load combinations and allowable limits for future plants.
Equipment qualification and the use of designated environmental profiles instead of bounding conditions.
Leak detection and reliable prediction of leakage through stable cracks.
T. Chang, Westinghouse, discussed their efforts in the application of the proposed broad scope rule. The technology used has been accepted for eliminating PWR reactor coolant loop breaks. Westinghouse has been the industry leader in the development and application of LBB technology. It was stated that an application to apply the broad scope rule was submitted to the NRC Staff in 1984. The NRC Staff has not reviewed the application because the broad scope rule has not yet been promulgated. The Westinghouse LBB methodology is similar to the proposed NRC rule under consideration.
Westinghouse has stated that the proposed broad scope rule should be expedited in order that plants under construction can fully realize the benefits.
e METAL COMP 0NENTS 6 Feb. 27-28, 1986 Meeting R. Cloud and members of Duquesne Light Co. presented an updated status report on the application of the LBB concept to high energy piping systems located inside and outside of containment. R. Cloud and Associates, Inc. is the major contractor for this program called "Whipjet." Briefly, Whipjet will do the following:
Satisfy DEGB postulation with engineering analysis showing a detectable leak-before-break is assured.
Reduce hardware in the plant in order to minimize plant cost, facilitate access for ISI and reduce time in performing inspections and maintenance to enhance ALARA position. ;
Increase plant safety through more complete knowledge of material properties and capabilities.
Whipjet is not intended to change the use of DEGB for establishing design criteria for ECCS, containment and equipment qualification.
A brief discussion to address the failure mode at Mohave and Monroe fossil plants was presented. It was stated that the catastrophic failure at the longitudinal weld of the reheat steam line was caused by creep rupture. The subcommittee agreed that if the upper temperature limit of the material is less than 650*F (as in LWRs) the creep rupture failure mechanism does not seem to occur.
l D. Norris, EPRI, briefly discussed their involvement in the area of the application of LBB to all high energy piping systems. He stated that enhanced plant safety and economic benefits to plant owners would occur if the proposed broad scope rule was promulgated. EPRI is involved with an ongoing program in this area with all NSSS vendors. Between 1975 and 1984 EPRI has spent $31 million in structural mechanics and NDE studies.
Between 1985 and 1989 they plan to spend about $18 million.
I
METAL COMP 0NENTS 7 Feb. 27-28, 1986 Meeting He stated that EPRI is supportive of NRC/ACRS initiatives in this area.
Some significant issues that need to be addressed are: leak detection methodology and its reliability outside containment, availability of weld toughness data, applicability to BWRs with 316NG with IHSI of the welds, and definition of acceptance criteria.
S. Bernsen, AIF, spoke briefly about AIF work in this area. He stated that the following criteria were proposed to the NRC and ACRS three years ago:
No intermediate breaks need to be postulated.
No breaks are assumed at terminal ends where leak-before-break criteria are satisfied, unless location is susceptible to unstable cracks from corrosion, thermal fatigue or water hammer.
Eliminate the SSE + LOCA loading combination for piping and support structures.
NUREG-0313, Rev. 2 (Implementation of BWR Pipe Crack Recommendation)
W. Hazelton, NRR, discussed the long-range approach for dealing with stress corrosion cracking in BWR piping as described in draft NUREG-0313, Rev. 2. Revision 2 expands Revision 1 coverage to include all stainless steel piping systems (class 1, 2, and 3), requires formal qualification of NDE examiners and procedures, provides guidelines for evaluation and repair of cracked welds, and upgrades leakage limits and monitoring. Revision 2 of the report generally follows the '
recommendations of the NRC Piping Review Committee as found in NUREG-1061, Vol.I.
Draft Regulatory Guide on PTS R. Woods, NRR, discussed the draft Regulatory Guide implementing the PTS rule. Briefly,thePTSruledefinesthescreeninglimit(270*F,300*F),
describes the calculation of RTPTS, discusses the flux reduction programs and determines when the PTS /PRA analysis needs to be performed.
g METAL COMP 0NENTS 8 Feb. 27-28, 1986 Meeting The draft Reg. Guide suggests the methodology of the PTS /PRA analysis.
The PTS /PRA analysis methodology is based on work done at H. B.
Robinson, Calvert Cliffs and Oconee plants.
ACRS Questions on PTS The Subcommittee next discussed the ACRS concerns with respect to PTS.
R. Woods, NRR, led the discussion of the NRC Staff's responses to the following ACRS questions.
Q.1: Is there any reason to believe that the issue of PTS cannot be treated generically, e.g., that some classes of plants or particu-lar designs are subject to a significantly higher frequency of severe vessel overcooling transients?
A.1: The NRC Staff based on B&W Report 1791, published in 1983, recommended to the Commission that the same screening limits should be applicable to B&W plants because, although the challenge frequency might be higher, the severity or risk rate, appears to be less. Hence, the NRC Staff was unable to justify a different screening limit for the B&W plant. [SECY-83-288, dated July 15, 1983 contains the NRC Staff's SER on this matter.]
With respect to pressure vessel material properties (copper / nickel content) this is properly accounted for automatically in the rule.
Q.2: How well justified is the crack distribution used in developing the NRC position on PTS? Is there a sufficient bases to justify the distribution used? Also, has allowance been made for crack growth?
Should there be?
A.2: The NRC Staff feels that the crack distribution assumed for the PTS study (modified Octavia and Marshall codes) is conservative, although the NRC Staff admits that it is one of the largest uncertainties that exist in the PTS study. It was also mentioned
METAL COMP 0NENTS 9 Feb. 27-28, 1986 Meeting that current pressure vessel inspection indicates a more conservative flaw distribution was assumed in the analysis. Crack growth was accounted for in the study by assuming flaws of various depths representing flaw size at the beginning and end of life of the pressure vessel.
Q.3: Throughout the transition temperature range, the Staff appears to permit use of unirradiated data with the upper shelf as a ceiling for fracture toughness. Is this so? If so, is this best estimate?
May it not be unconservative? If so, by how much?
A.3: Current studies are being performed to correlate the Charpy curve to the K Ic curve, specifically if the Charpy curve drops and hanges shape does the K Ic curve change shape? Information obtained thus far indicates that some margins have been eroded, but that the NRC Staff feels comfortable because the margins are still large.
Q.4: Dr. I. Catton, an ACRS consultant, questions whether HPI recovery following partial core uncovery is covered adequately under PTS (or,ifnot,elsewhere). Dr. Catton suggests that HPI following partial core uncovery will lead to low temperatures and possible water hammer. Can the Staff provide estimates on the frequency and severity of such an event? What are the major sources of uncer-tainty in these estimates, and their magnitudes? Is human error very important? Is plant design important? Is thermal hydraulic prediction adequate for the purpose?
A.4: R. Woods stated that the accident scenario of partial uncovering of the core followed by core recovery is a severe accident concern and is not regarded as a PTS issue.
Q.5: Are there any steam generator overfill scenarios which the Staff considers significant for PTS?
A.5: Yes, there are steam generator overfill scenarios that are important to PTS. Oak Ridge identified the sequence of a break in the steam line followed by overfeed by the aux 1111ary feedwater as an important, but not dominant, sequence for the Robinson analysis,
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METAL COMP 0NENTS 10 Feb. 27-28, 1986 Meeting although it is dominant for the Oconee analysis. The NRC Staff indicates that the Oak Ridge analysis supports the PTS rule.
With respect to loop flow stagnation concerns, J. Reyes stated that this problem is be.ing addressed by T. Theofanous.
Q.6: Are there any reactors for which the data on chemical composition of critical welds is not well determined? If so, how is a judgment made? Is the difference between the composition accepted and the worst possible significant? If so, how much less likely must the worst possible be? How is this judgment made?
A.6: N. Randall, RES, stated that updated chemical composition of critical welds were recently documented--on January 23, 1986. In general, the NRC Staff stated that they are reassured in the sense that the justification for numbers used in the PTS analysis is getting much better.
Four categories of weld chemistry data are available. The first category is the actual measured value of the critical welds, which is true for nearly all pressure vessels. The second category is called generic chemistry where a plant had only one or none of the measured values for their critical welds, but through searches have found other typical vessels. These are then sampled to determine its chemistry. B&W and Westinghouse Owners Groups have determined weld chemistry by this method. It was stated that about half of the plants approaching the screening criteria are in this category.
The third category is that of historical numbers. In this case, all they have is a statement indicating that for vessel welds fabricated in this time period a certain distribution of copper / nickel content is reported. From this data a conservative upper bound value is obtained. A quick look at the data in this category shows that ample margins exist before the screening value
e _
METAL COMPONENTS 11 Feb. 27-28, 1986 Meeting is reached. In the fourth category, if no data is available the NRC Staff dictates an upper bound value which is very conservative.
Q.7: What is the expected consequence of a through-wall crack? What is the likelihood of: (a) cora melt, (b) late containment failure, and (c) early containment failure?
A.7: In part, the answer relates to a severe accident scenario. But the NRC Staff did fund a study by Pacific Northwest Laboratory to develop a vessel failure model. This model was then applied to Oconee in order to determine the containment failure modes. The result is that 4/10-percent of the through-wall cracks would lead to early containment failure if we assume no containment spray and 1/10-percent would lead to early containment failure with containment sprays; these would lead to isolation containment failures. The study indicates that the " objective of individual risk of early and late fatalities are met." This study is reported in a paper by R. Barrett and E. Throm which was presented at the 12th Annual Water Reactor Safety Meeting at NBS in Germantown, Md.
Future Action The subcommittee decided that a subcommittee report on GDC-4 (Broad Scope Rule), NUREG-0313, Rev. 2 (Implementation of NRC Pipe Crack Review Committee on Long Ten: Fix of BWR Pipe Cracking) and Draft Reg. Guide on PTS (Implementation of PTS Rule) should be presented to the ACRS at its 311th Meeting in March 1986.
The meeting adjourned about 1:05 p.m. -
NOTE: A transcript of the meeting is available in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., or can be purchased from ACE-Federal Reports, 444 North Capitol Street, Washington, D.C. 20001, (202) 347-3700.
I
g i Fedil Register / Vol. 51. N'o.'26' / Pridiy, F;br'u'ar'y 7.19d / IVotides 5 4S33 ;
in addition, the licensee also was coritends that the development and the adequacy of B&W plant designs. ' '
, informed of the NRC concern about finalization of this long-term corrective including consideration of the severe procedural controls in high-radiation action program occurred in a prudent overcooling event atkancho Seco on areas via severalinformation notices and timely manner. October 2.1985.The Subcommittee may
- also review the NRC Staff's plans to and a circular (Information Notice 84-19 a uoh.on dated March 21,1984. Information , reassess the long-term safety of B&W -
Notice 82-51 dated December 26.1982. The NRC maintains that the long. term reactors. -
and Circular Notice 76-03 dated actions taken by the licensee were not Oral statements may be presented by September 13.1976).These notices particularly prompt in that some of the members of the public with the emphasized the importance of ensuring actions could and should have been in concurrence of the Subcommitte's that radiation protection procedures and place at the time of the enforcement Chatrman; written statements will be radiation protection training and conference.namely, an upgrade of the accepted and made available to the retraining programs specifically address procedures for entry into locked high- Committee. Recordings will be permitted the matter of control and access to such , radiation areas in general,and the TIP only during those portions of the areas and initiate appropriate retraining room in particular. These items were not meeting when a transcript is being kept.
of all plant personnel.They also - provided by the licensee at the and questions may be asked only by recomrhended that entry be allowed Enforcement Conference and appeared members of the Subcommittee.its only after appropriate management to have been considered only after the consultants, and Staff. Persons desiring review and approval. Further. they Enforcement Conference on September to make oral statements should notify recommended periodic audit of these 5.1985 and the Region I Confirmatory the ACRS staff member named below as actions to ensure their continued Action Letter (CAL) issued on far in advance as is practicable so that effectiseness.Many of the actions noted September 9.1985. In addition, four of appropriate arrangements can be made.
in the Notices are similar to tnose in the the six items in the licensee's PIR simply During the initial portion of the Confirmatory Action Letter issued by proposed evaluation of certain aspects meeting, the Subcommittee, along with the NRC to Vermont Yankee on of the program rather than describing any of its consultants wha may be Septemoer 9,1985. In additian, there specific actions taken or necessary to present, may exchange preliminary have been a number of es,calated correct deficiencies and improve the views regarding matters to be enforcement actions for similar program. It was not until September 21, considered during the balande o.'the violations at other plants of which the 1983 after the Enforcement Conference meeting licensee should have been aware. A and issuance of the Confirmatory Action The Subcommittee will then hear purpose of publishing escalated Letter (CAL) that the licensee committed presentations by and hold discussions enforcement actions in NUREG-0940 lh , sen tiv s o e C 9 R.
and Orders imposing Civil Penalties in toFortake thesethese actions' reasons, t he NRC maintains c the Federal Register is to give licensees that the licensee's long. term actions Persons regarding eis review.
notice of other enforcement actions were not unusually prompt and do not Further information regarding topics which niay bear on their own provide an adequate basis for mitigation to be discussed, whether the meetm, g operations. (See Vol. 4. No.1, p. I.A-94 of the civilpenalty. has been cancelled or rescheduled, the and Vol. 3. No. 2. p.1.A-1 of NUREG- .
Chairman's ruling on requests for the NRC Conclusion 0940.) opportunity to present oral statements Accordingly, the NRC maintains that After consideration of the answers the licensee had prior notice of potential received and the licensee's statements and the time allotted therefor can be obtained by a prepaid telephone call to problems associated with TIP rooms. of fact, explanation, and arguments for Therefore, a basis would have existed mitigation of the proposed civil penalty, the cognizant ACRS staff member,Mr.
the staff concludes that any adjustment Richard Major (telephone 202/634-1413) for an increase in the civil penalty '
amount had it not been for the licensee's to the civil penalty amount is between 8:15 A.M. and 5-00 P.M. Persons reporting of this event and prompt short- inappropriate. Therefore, the proposed planning to attend this meeting are
$50.000 civil penalty should be imposed. urged to contact the above named term corrective actions. individual one or two days before the Licensee's Assertion p Doc. 60-2740 Filed 2-6-86; 8.45 am] schedule meeting to be advised of any The licensee claims that at the time of e m ocoot m W changes in schedule, etc., which may the Enforcement Conference on have occurred.
September 5,1985, significant efforts Datedfebary q Advisory Committee on Reactor M*"*" "
had been taken to assess the specific causes of the incident and develop long- Safeguards Subcommittee and Wilcox Water Reactors; on Babcock , AssistantExecutiveDirectorforProject Meeting term proposed corrective actions. In Review.
particular, on the day (August 9) The ACRS Subcommittee on Babcock p Doc. 86-2761 Filed 2-6-46: 8.45 am]
following the event, the Plant Manager and Wilcox (B&W) Water Reactors will amo caos neo.ms ~
directed the Chemistry and HP hold a meeting on February 25,1986, _
1 technician to generate a Plant Room 1046,1717 H Street. NW, . /
VAdvisory Committee on Reactor l
j Inforination Report (PIR) so that the Washington, DC.
mnt could be analyzed and The entire meeting will be open to Safeguards Subcommittee on Metal j recommended long.long corrective public attendance. Components; Meeting i action could be provided.The final PIR. The agenda for the subject meeting which was issued approximately 6 shall be as follows: The ACRS Su.bcommittee on Metal Components will hold a meeting on weeks later on September 17.1985. Tuesday, February 25,19M30 A.Af. February 27 and 28. Room 1046,1717 H proposed six long. term corrective Untilthe Conclusion ofBusmess Street.NW Washington,DC.
actions. On September 21,1985 the Plant The Subcommittee will consider the To the extent practical thu meeting i Manager dispositioned the long. term will be open to public attendance.
recommendations.The licensee implications of operating experience on e wpf -
ATTACHMENT A
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e
--w-4834 Federal Register / Vol. 51, No. 28 / Friday. February 7.1186 / Notices ifowever, portions of the meeting may changes in schedule, etc., which may Emironmentallmpacts of the be closed to discups industry proprietary have occurred. Proposed Action-The environmental information. Dated February 3.1986. impacts associated with construction of The agenda forsubject meeting hall htorton W. Ubarkin, the facility have been previously be as follows- dis ussed and evaluated in the NRC Assistant Ex ecutive Diredarfor Projed E**i'* staff's Final Environmental Statement Thursday. Febmary 27. M86--dJo A.M (FES) issued in June 1974 for the untilthe Conclusion ofBusiness IFR Doc.as r6 Fded 8-6-88. 8 s5 am) construction permit stage which covered Friday. Februar>% 1986-4t:JO A3f. 8"**C *"* construction of two units. Unit 2 is not Untilthe Conchman o/Bassness
- affected by the proposed actior.
Since the proposed action insolves
%e Subcommittee will review, but IDocket No. 50-45f1 extending the construction permit, not necessarily be limited to, the Tadiologicalimpacts are not affected by following items:(1) NUREG-0313. Texas Utilities Electric Co. et al. this action. There are no radiological Revision 2. entitled. Technical Report Comanche Peak Steam Electric impacts associated with this action.The on Material Selection and Processing Station. Unit No.1; Environmental impacts that are involved are all non-Guidehnes for BWR Coolant Pressure Assessment and Finding of No radiological and are associated with Boundary Piping." E d (2) Regulatory Significant impact continued construction.
Guide XXX entitled. " Guide for license Since the construction of the facilityis Preparalion and NRC Staff Review of The Nuclear Regulatory Commission essentially 100% complete, most of the Plant Specific Analysis Required by PTS (the Commission)is considering construction impacts discussed in the Rule." he Subcommittee will also hear issuanm of an extension to the latest FES have already occurred:
c status report of the proposed broad construction completion date specified rule to modify CDC-4 of10 CFR Part 50 construction-related activities have in Construction Permit No. CPPR-126 disturbed about 400 acres of rangeland.
(the leak.before-break broad scope rule issued to Texas Utilities Electric the Squaw Creek Reservoir has been is applicable to allI.WR high energy Campany. Texas Municipal Power built, as have transmission lines and piping systemsj. Agency, Brazos Electric Power corridors, and a railroad spur.%ese Oralstatement may be presented by Cooperative. Inc. and Tex-La Electric activities and their impacts occurred members of the pubhc with concurrence Cooperative of Texas. Inc. (Applicants) earlier and are not affected by this of the Subcommittee Chairmaru written for the Comanche Peak Steam Electric proposed action.
statements wi!! be a'ccepted and made Station Unit No.1 (the facility) located The reinspection and rework that may cvailable to the Committee. Recordings on Applicants' site in Somervell County, be required will not have any signif; cant will be permitted only during those Texas. environmental impact. ne impacts portions of the meeting when a associated with the work are equivalcnt transcript is being kept, and questions EnvimnmentaW= =nat to those of a maintenance or repair may be as ed only by members of the Identification of PrceposedAction ne program. This activity will all tale place Subcomm)ttee,its i consultants, and Staff.
proposed action would amend the within the facuity and will not result in e son s t c struction permit by extending the impacts to previously undisturbed areas.
d II MSM latest construction completion date to There are no new s,gnificant i , impacts members as far in advance as Angust 1,1988. The proposed action is in associated with this extension.There practicable so that appropriate response to Applicants' request dated are however, impacts ' hat would arrangements can be made.
January 29,1986, as supplemented continue in order to complete plant During the initial portion of the February 4.1986. construction in addition to rework meeting. the Subcomcuttee, along with The Needfor the ProposedAction: discusse'd above.nese are community any of its consultants who may be The pro, osed action is needed because and traffic impacts. and continued present. may exchange preliminary the construction of the facility is not yet groundwater withdrawal.
views regarding matters to be fully completed.The Applicants state Community impacts from continued considered during the balance of the that, although construction on construction would be similar to those -
muting. Comanche Peak Unit 1 was essentially imp 4 cts previously assessed. ne total The Subcommittee will then hear completed early in 1935, major efforts to number of workers on-site for both units prisentations by and hold discussions reinspect and reanalyze various at the present time (about 5300) is about with representatives of the NRC Staff. structures, systems, and components is the same (although somewhat smaller) its consultants, and other interested currently underway. These efforts are as during earlier peak construction persons regarding this review. being conducted by the Applicants' periods.The number of workers Further information regarding topics Comanche Peak Response Team to specifically assigned to Unit 1 is small of be discussed. whether the meeting verify both design and construction compared with the number associated has been cancelled or rescheduled, the adequacy as weu as to respond to with the completion of Unit 2. The Chairman's ruling on requests for the numerous issues raised in the operatirs number of workers on-site will decline opportunity to present oral statements license proceeding. by the NRC's as the reinspection program for Unit 11s i and the time allotted therefor can be Technical Review Team, and by other completed during 1988. Continuing obtained by a prepaid telephone call to sources. This activity has been ongoing construction does not involve the cognizant ACRS staff member. Mr. since the fall of1984.The Applicants community impacts different from those Elpidio Igne (telephone 202/634-1414) anticipate that it wiU not be complete previously considered or significantly between 8:15a.m. and 5$0 p.m. Persons before the second quarter of'1986. In' greater than those previously considered planning to attend this meeting are addition. the operating license hearings or experienced.
urged to contact the aboire named are not yet completed and willinvolve The constructicn permits for individual one or two days before the additional trme for which the Comanche Peak Umts 1 and 2 limit scheduled meetmgto be advisedof any construction permit will be needed. groundwater usage to 40 gpm on an 1
1
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- REY. 1, 2/20/86 TENTATIVE' SCHEDULE ACRS METAL COMPONENTS SUBCOMMITTEE MEETING FEBRUARY 27-28, 1986 WASHINGTON, D. C.
February 27, 1986 I. Chairman's Opening Statement- P. G. Shewmon 15 min 8:30-8:45 II. Proposed Rule Change to GDC-4
- 1. Limited Scope Rule, status and update - J. O'Brien 15 min 8:45-9:00
- 2. Broad Scope Rule Presentation - J. O'Brien 120 min 9:00-11:15
- BREAK *** [witha15-min,breakat10:30)
NRR Plans - R. Bosnak 45 min 11:15-12:00
- LUNCH *** 12:00-1:00 Implementation
- Status at Beaver Valley R. Cloud I hr 1:00-2:00
- Westinghouse Plans - H. Clark 1 hr 2:00-3:00
- BREAK *** 15 min 3:00-3:15
- 3. Report of EPRI's Study in this area - D. Norris 1 hr 3:15-4:15
- 4. AIF CommentsSo$t.fnS6r)
- tid __rn:tci- 30 min 4:15-4:45 III. Subcommittee's Discussion 15 min 4:45-5:00 IV. RECESS February 28, 1986 I. Chairman's Statement-P. G. Shewmon 15 min 8:30-8:45 II. Presentation on NRC Technical Positions on BWR Pipe Crack, NUREG-0313, Rev. 2 -
B. D. Liaw 120 min 8:45-11:00
- BREAK *** [with a 15-min. break at about 10:00]
III. Presentation on Format and Content of Plant Specific PTS Safety Analysis Reports for PWRs, (Reg. Guide Task SI 502-4) - R. Woods 60 min 11:00-12:00
R. Woods /NRC Staff 90 min 1:00-2:30 V. Subcommittee Discussion 30 min 2:30-3:00 ADJOURNMENT 3:00 ATTACHMENT B
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ATTACHMENT C LIST OF HANDOUTS ACRS METAL COMP 0NENTS SUBC0KIITTEE MEETING FEBRUARY 27-28, 1986, WASHINGTON, D.C.
February 27, 1986
- 3. NRR Plans for the Implementation of the Limited and Broad Scope Rule Revisions to GDC 4, R. J. Bosnak, NRR
- 4. Alternative Pipe Break Criteria (Leak-Before-Break Concept),
John McInerny and Dr. T. Chang
- 5. Robert L. Cloud Associates:
Review of Whipjet, Swec Progress, R. L. Cloud Progress, EPRI Progress February 28, 1986
- 1. Long Range Approach for Dealing with Stress Corrosion Cracking in BWR Piping Draft NUREG-0313, Rev. 2, Gen. Issue 86, W. S. Hazelton, NRR
- 3. Pipe Break and Load Combinations, S. Bernsen, AIF Subcommittee on Load Combinations ATTACHMENT C