ML20151D233
| ML20151D233 | |
| Person / Time | |
|---|---|
| Issue date: | 02/10/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2549, NUDOCS 8804140059 | |
| Download: ML20151D233 (15) | |
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Ii CERTIFIED COPY DATE ISSUED: February 10, 1988
SUMMARY
/ MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON STRUCTURAL ENGINEERINE JANUARY 22, 1988 ALBUQUERQUE, NEW MEXICO The ACRS Subcommittee on Structural Engineering met on January 22, 1988 in the AMFAC Hotel in Albuquerque, New Mexico to review the test results of the 1:6 scale model reinforced concrete containment tests. A visit to the test site was made previous to the meeting.
Notice of the meetino was published in the Federal Register on January la, 1988 (Attachment A).
The schedule of items covered in the meeting is in Attachment B.
A list of handouts of handouts kept with the office copy of the minutes is included in Attachment C.
There were no written or orel statements received or presented from members of the public at the meetino.
E. Igne was cognizant ACRS Staff member for the meeting.
1 Principal Attendees ACR5 NRC Staff CTiess, Chairman J. Costello D. Ward, Member R. Kenneally P. Shewmon, Member J. Ebersole, Meniber C. Mark, Member M. Bender, ACRS Consultant Sandia National Laboratories (SNLA)
Others D. Horschel T GEards, NUTECH D. Clauss H. Ecker, NUS l
B. Parks i
l J. Weatherby l
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e Minutes / Structural Engineering 2
Meeting, January 22, 1988 Highlights 1.
J. Costello, RES, in his opening remarks stated that the objective of the containment integrity progran is to provide NRC with a suite of reliable methods i.e., computational, analytical and empirical which can be used to predict the response of containment system when subjected to loadings beyond the design basis i.e., severe accidents and earthquakes beyond the SSE. The methods will be qualified by performing experiments on scale podels of containment structures and full-size electrical penetration assemblies and a personnel airlock.
2.
D. Horschel, SNLA, presented an overview of the test and results of the test and post-test inspections. The 1:6 scale reinforced concrete containment is among the largest and most complex structures of its type tested. The containment was designed and built by United Engineers and Constructors to Section III Division 1 and 2 sections of the ASME code and has a design pressure of 46 psig. The containment is a right circular cylinder with an inside diameter of 22 feet and is capped by a hemispherical dome. With its 40-inch thic' basemat it has an overall height of over 37 feet.
Features of the containment include four major penetratiens (two equipment hatches and two personnel airlocks) as well i
i as several smaller penetrations that pass both separately and in clus-l l
ters through the containment wall.
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In order to properly assess the composite structure, several hundred transducers were embedded in the concrete wall. A total of about 1200 l
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Minutes / Structural Engineering 3
Meeting, January 22, 1988 tranducers i.e., strain gages, displacement transducers, thermocouples, pressure gages, inclinometers, were installed.
In addition, video cameras, still cameras, track transducers and acoustic detection systems were available.
As the containment was pressurized (nitrogen gas) many phenomena were observed. At 125 psig, some leakage was thought to be coming from equipment hatch 3.
The seal in this hatch was thermally aged before the test so that some leakage could be realized during the test. Leakage noise was monitored and did increase at higher pressure, it is unclear now whether leakage was coming through the seals or through the liner; during post-test inspection two small holes in the liner near equipment hatch B were found.
Leakage was noticed coming from the vicinity of ec,uipment hatch A at 138 psig and the ovalization of the sleeve of equipment hatch A was noticeable at 143 psig.
Displacement measurements indicated nearly i inch increase in the horizontal diameter.
During the final stages of pressurization, the uplift of the basemat from the concrete work mat was very noticeeble. The amount of uplift was about 3/8 inch. Cracks that were visible from the low pressure testing became much wider, some as wide as about 1/8 inch. The cracks followed seismic bars (angled at 45 degrees) as well as the hoop and meridonal directions.
In the lower sections of the containment few cracks were obvious. The test was terminated when leakage became large. Nitrogen gas was being delivered at a rate of 4000 cubic feet per minute and the pressure in the containment was decreasing slightly, maximum pressure was about 145 psig (leak rate was greater than 5000% mass / day). A data scan was initiated (actual model pressure was about 141 to 142 psig at
Minutes / Structural Engineering 4
Meeting, January 22, 1988 this time) and completed before the decision to discontinue testing was made.
i Post-test inspection inside the containment revealed a tear running in a vertical direction by the entire insert plate edge about one inch from it.
The tear was about 20 inches long. Upon closer inspection, many small tears on areas of distress were found on the liner near other insert plates.
Preliminary inspection indicate that the liner near the basemat cylinder wall junction shows no sign of severe distress.
D. Horschel than presented a brief discussion on possible future options with the further use of the 1:6 scale model containment, including NDT and sectioning, separate effects tests, and retesting of the containment.
3.
D. Clauss, SNLA, discussed the coordination of analyses of the 1:6 scale reinforced concrete containment indel. The pre-test analysis objectives were to obtain insights that could be used in planning instrurentation and conducting the test, to assess how accurately blind predictions matched actual performance, and to evaluate uncertainties in the analysis. Thepost-testanalysisobjectivesare1)tounderstand the limit states governing model failure, 2) recomend additional tests to investigate other potential limit states or to generate other data needed to validate analytical tools and 3) propose a comprehensive approach, which addresses all credible limit states for predicting the behavior of reinforced concrete containment buildings subject to severe accident loads. The results of the pretest analyses are reported in i
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Minutes / Structural Engineering 5
Meeting, January 22, 1988 NUREG/CR-4913, issued in May 1987 (which was completed before the high 4
pressure test).
Predicted capacities for the model varied from 130 to 190 psig.
Ten licensing and research oraanizations from the U.S. and Europe made the analyses on a voluntary basis (no cost to the NRC). A table of failure predictions for the 1:6 scale containment model for the participating organizations is attached. The table also shows the predicted limit mechanism.
Meetings of the participating organizatiens have taken place after the model containment test. The conclusions thus far are 1) global measures of containment response i.e., displacerents, free field strain, can be predicted accurately. 2) localized behavior, which may control containment failure, is more difficult to predict, and 3) it is not valid to assume that the liner is perfectly bonded to the concrete.
4 J. Weatherby, SNLA, discussed the pretest analysis and comparison of predictions and results conducted at SNLA.
SNLA employed both nonlinear mer.brane analysis and finite element (ABAQUS) code in their compu-tations.
Each of the following events was assumed to indicate a loss of l
containment integrity:
The stress level in the reinforcing steel exceeds 99ksi at any
- location, The equivalent plastic strain in the liner exceeds 15% at any location.
The transverse shear stress at any location exceeds the shear strength predicted by the empirical relation developed by l
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Minutes / Structural Engineering 6
o MGeting, January 22, 1988 Aoyagi et al.
(The shear-friction model was also used to estimateshearstrength).
t ShtA failure predictions were as follows:
Elastic-plastic analysis: -failure pressure of 170 psig and the failure mode is by bending failure at the cylinder /basemat juncture.
Cracking analysis:
failure pressure of 180 psig and the failure mode is that the ultimate strength of the hoop bars in the cylinder midsection is exceeded.
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ShlA stated that difficulties in the failure prediction are rnany. Among the major difficulties are the following:
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Axisymmetric finite element models neglected strain concen-trations in the liner caused by penetrations and the liner 1
anchorage system.
The 3-D equipment hatch model neglected the strain concen-trations caused by the liner anchorage system and slippage between the concrete wall and liner.
Uncertainty exists in the failure criterion for liner tearing.
1 Conclusions for the analysis of the piping penetration are as follows:
When the stud anchors are neglected, the liner strains are too i
snall to cause tears to develop.
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Minutes / Structural Engineering 7
Meeting, January 22, 1988 The primary mechanism that initiated the liner tear was the strain concentration caused by the stud anchors.
5.
D. Horschel, SNLA, discussed the CEGB's prestressed containment model. The 1:10 scale prestressed containment model test is sponsored-by the Central Electricity Generating Board (CEGB).
It is designed by Nuclear Design Associates and construction and testing will be done by Taylor Woodrow. This contairment will be a 1:10 scale model of the sizewell 'B' centainment structure. The model consists of one equipment hatch, two oersonnel airlocks, no liner and fabricated of mi-cro-concrete.
Scaled steel areas rather than 1:1 bar and tendon re-placement will be employed. The start of design originated on October 1987. Start of construction should begin in February 1988 and com-pletion of testing should be by September 1988.
6.
D. Clauss, SNLA, discussed the investigation of the leakage poten-tial of a personnel airlock. He stated that previous studies sponsored by the NRC have indicated that personnel airlocks have a high potential for leakage.
In order to address this concern, a full size personnel airlock from a cancelled nuclear power plant (Callaway) was purchased and tested to hypothetical severe accident environments. The seal was 0
aged in-place at 330 F for about one week to simulate the combination of radiation and thermal aging to IEEE spec.
Significant darrage to the seal was discovered after the aging process. Most obvious was a gross change in cross section of the seal. MARC finite element computer program was used to calculate sealing surface displacements and leak area. Test conclusions are as follows:
Minutes / Structural Engineering 8
Meeting, January 22, 1988 Relative displacement of the sealing surfaces was small, a prcdicted by analysis. As a result, leakage did not occur at moderate temperatures even at very high pressure.
With small deformatiens, t'he seal must be completely deteri-orated for leakage to occur.
The seal temperature is significantly less than the contain-ment atmosphere temperature, f
7.
M. Parks, SNLA, discussed the status of tests on seals and gaskets, inflatable seals and bellows used for containment penetrations.
Tests have been performed on seals and gaskets and test results obtained.
Tests en inflatable seals and bellows are planned.
The research budget has been reduced and the latter two tests have not been funded for FY 1988. The final report for SNLA tests, which also includes a summary of the INEL tests on seals and gaskets (INEL final report was published in July 1987) has been written and is under final review.
All tests on seals and gaskets performed at SNLA and INFL have been cort pleted.
Highlights of the test results are as follows:
All of the tested seals failed at more severe pressure and temperature conditions than predicted for the severe accidents i
of PWR and BWR MK III containments.
Some of the tested seals failed at lower temperatures than predicted for the severe accidents of BWR MK I and MK !!
contairments, i
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Minutes / Structural Engineering 9
Meeting, January 22, 1988 Aging had little effect on the failure temperature of the seals.
Flange rotation (0 to 12 degrees) had little effect on the failure temperature. However, the amount of seepage before failure we: graater for larger flange rotations.
(basedon the results of INEL test only.)
It was suggested that the planned tests on bellows need to be reviewed by subcommittee members to determine its usefulness.
Subcomittee Action Dr. Siess stated that the SNLA personnel did an excellent job on the 1:6 scale model containment tests. A subcommittee report to the full ACRS at the February meeting is planned.
It was proposed that the subcommittee meet again, sometime soon, tn discuss the following matters not discussed at this meeting because of lack of time.
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Proposed future options for the model and proposed separate effects test, Future work on pristressed containments, Completion of work on steel containments, and Seismic capacity of containments and results of Sargent &
Lundy study.
Minutes / Structural Engineering 10 Meeting, January 22, 1988 NOTE:
Additional reeting details ci.n be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H St., NW., Washington, D.C.C. or can be purchased from Heritage Reporting Corporation, 1220 L Street, NW.,
Washington, D.C. 20005,(202)628-4888 i
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Table 1.3 Failure Predictions for the 1:6-Scale Containment Model Org.
Capacity
)
Code psig (MPa)
Limit Mechanism SNL 168 (1.16)
Flexural fsilure at the cylinder-basemat juncture brought on by extensive crushing i
of concrete.
l ANL 180-190 (1.24-1.31)
Either (1) failure of a hoop rebar splice near midheight, (2) failure of a weld in the liner near the basemat, or (3) failure of the liner just above the knuckle.
EPRI 140-150 (0.97-1.03)
Liner tearing at
- ,..e connection of the liner to the wall-basemat juncture knuckle, triggered by basemat bending failure.
l Plasticity of rebars corresponding to CEA 138 (0.95) rapidly increasing displacements of the i
structure, i
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13D (0.90)
Transverse shear f ailure of cylinder just above the shear reinforcement (some probability for local tearing of liner around studs).
ENEA 161-164 (1.11-1.27)
Failure at the base of the cylinder caused by combined effect of bending, tension, and shear.
t SRD 164 (1.13)
Rebar at the pole of the basemat exceeds I
its ultimate strength.
l GRS 167-174 (1.15-1.20)
Failure at cylinder-basemat intersection.
174-189 (1.20-1.30)
Failure of hoop reinforcement or tearing of liner.
, 128 (0.88)
Flexural / shear failure at the well-basemat junction or gross yielding of hoop rebars.
I Flexural failure at the wall-basemat CEGB 160 (1.10) junction.
- towest bound of estimates (see page 251)
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ATTACHME@f u Fed:r'l Register / Yol. 53. No. 9 / Tbiirsday, lanuary 14, 1988 / Notices 971
-.~. _
and 2, we re dinussn.1 in NRds FES For further details with respect to this opportanity to present oest statements issued April l'N (N'JREGM4).
action. see the request for estetWon sad the time allowed therefor can be -
De pro;med ofension will not allow dated August 13.1987. which is
. Obtained by a peepaid telephone call to any work to be pi rfurna d that is not asailable for public inspection at the the cognizant ACRS Staff member.Mr.
already wiluwod by the nisting Commission's Public Document Room.
Sam Duraiswarny Delephone se2/6M-constiuttion permit. The exfension will 1717 11 St reet. N W., Wa shington. DC.
3287) between 815 a m. and 590 pm rmrr!3 grant the applicant mere time to and at the local public document room.
Persons planning to atter,I this meeting tumplete construction in accordance Pottstown Pubbe Library. 500 lligh are urged to contact the above named with the presiousl) apprmed Street. Pottstown. Pennsylvania 19464.
individual one oe two days before the construttiun permit.The probability of Dated at Bethesda. Maryland, this tith day scheduled meetir*g to be advised of any acidents has not becn increased and of] mar 3 was changes in schedule, etc. wNeh may have occurred.
post-acdJent rad. alogical releases will For the kleer Regulatory Commisuort not I e greater than praiously Walter R. Butler.
Date: January 11.1988 determined, nor does the proposed Dm u Preie" Deectorate 1-2. Dainen of Morton W. LJbarkin,
, rYfar the
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Commission conc!cded that there are no (FR Dac. 8MM Fited 1-1M8. 8 45 am]
1 oc.66-040 Faled 1-1Mn 8 45 em) significant radiulagkal enuronmental
- w "G Coot M * '-"
impacts associated with this proposed e s t c n don.
M ith reptd to potential non.
Advisory Committee on Reactor Advisory Committee on Reactor radiolopal impacts, the proposcd Safeguards, Subcommittee on Safety Safeguards Subcommittee on h"
[e','s [c$'d Research Program; Meeting Structural Ehgineering: Meeting s
e deimed in 10 CFR Part 20 It does not The ACRS Subcommittee on Safety The ACRS Subcommittee on affect non radiological plant (ffluents Researth Program will hold a meeting Swetural Engineering will hold e i.
and has no other environmental impact.
on Januaq 29,1958. Room 1NB,171711 meeting on January 22.1%8. at the t
This estension dets not allow aoy work Street NW.. % ashington. DC.
AMFAC llotel. 2910 Yale Blvd., SE.,
g to be performed of the t>pe no; The entire rneeting will be open t Albuquerque. NM.
preuously authorized by the esisting public attendance.
The entire meeting will be open to construction pt rmit. Therefore, the The agenda for subject meet.mg shafi public attendance.
Comm:ssion concludes that thc re are no be as follows:
%e agenda for the subject meeting p'
significa ni non-ra deecal Fridy /crwory 29. mN.30 om until shall be as follows:
I.
ensironmentalimpm.ts associated with i m p.m.
Friday. fonuary 2.2.196S-9'30 a.m. until this prepostd e s1ension The Subcorr.mittee will discuss the the conclusion of business
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Alternatiws la the Pivpowd At twn rnethodology to be used by the RES Staff The Subcommittee will review the t prioritize'NRC researth activities.
An attematne to the proposed actmn wou'd be to deny the request. Under this Oral statements may be presented by results of the concrete containment model test.
I-alternatn e, the app!icant wot.!d not be members of the public with concurrence of the Subcommittee Chairman: written Oral statements may be presented by able to complMe construction of the facility.This would result in denial of statements will be accepted and made members of the public with the as ailable to the Committee. Recordings concurrence of the Subcommittee
/ the benefit of power prodaction Furth(r. w rl! be permitted only during those Chairman; written statements will be this option would not chminate the
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ensironmentalimparts of construction portions of the meeting when a accepted and made available to the Committee. Recordings will be permitted already incurred.Therefore this transcript is being kept, and questions I
alternatne is rejected.
may be asked only by members of the only during those portions of the Subcorr.mittee,its consultants, and Staff. meeting when a transcript is being kept, Alicenotise Use o/Resourre Persons desiring to make oral and questions may be asked only by This action does not involse the use of statements should notify the ACRS Staff members of the Subcommittee,its resources not presiously considered in member named below as far en advance consultants. and Staff. persons desiring to make oral statements should notify the FES for the L.imenck Generating as practicable so that appropriate the aCRS staff member named below as Station.
arrangements can be made.
During the initialportion of the far in advance as is practicable so that Apocies codParsons Consulted meeting. the Subcommittee, along with appropnate stiangements can be made, The NRC staff reviewed the any of its consultants who may be During the initial portion of the applicant's request and apphcable present, may exchanPe preliminary meeting. the Subcommittee, along with documents referenced therein that views regarding matters to be any of its consultants who may be support this extension.The NRC dad not considered dunng the balance of the present. may exchange preffminary consult other esencies or persons.
meeting.
views rqartling matters to be
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The Subcommittee will then hear considered during the balance of the t
Finding of No Sign!ficant Impact presentations by and hold dieevssions meeting.
l The Commission has determined net with representstives of the NRC Staff.
The Subcommittee will then hear to prepare an environmentalimpact l's consultants. and other tnterested presentations by and hold discussions 6
statement for this acJion. Baeed upon persons regarding this review.
with representatives of the NRC Staff, 3r the environmental a ssessment. w e Further information regarding topics its consultants, and other interested sl conclude that this action will not base a to be discuesed, whether the meeting persons regardmg this review.
p significant eIfect on the quality of the has been canceded cr revherh:ted, the Further information regarding toples human ensironment.
Chairman's ruling on requests for the to be discussed, whether the meeting 4
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W UCT7KLCITUT7/TaOMYEVl/NEMW EhEts/wwd / mces w
his been c nc:lled et resch:duled. tho Since thsMfied infonrdtien m:y be (2) Replace the curves in Figures 3.2.1-Chairmin's ruling cn r: quests for th7 discussed during the briefing there ' vill 1 cnd 3.2.1-2 ts provide Maximum opportunity topresent oral statements be two sessient One open to the public.
Avsrage Planar Heat Generati:n limit and the time auotted therefor can be and one dosed.
curves for two new fuel types that will replace two existing fuel types during cbtained by a prepa!d telephone call to
'r ring. M ar> land, this 6th the next operating cycle (Cycle 2).
the cognizant ACRS staff member Mr.
day o fa'n"$
'I' 1 (3) Change TS 3/4.2.3 to provide new Elpidio C. Igne (telephone 202/634-1414)
For the Lclear Regulatory Commission.
hiCPR limits for Cycle 2 operation t e, tween 815 a.m. and 5 00 p.m. Persons planning to attend this meeting cre ilush L Thomp+on,lt.
providing limits for two exposure ranges urged to contact the above named Director, office of % clear Moteria/ Sofety rather than a single exposure range as in indisidual one er two day before the ond Solesords.
the existing TS. The two ranges are (a) scheduled meeting to be advised of any (FR Doc. 86 on Fded 1-n-ee, s 45 am) from Beginning of Cycle (BOC) to End-changes in schedule, etc., which may of. Cycle (EOC) minus 2000 hfWD/ST g,,
hsve occurred.
and (b) from EOC minus 2000 MWD /ST to EOC. The Aetion and Sun eil/ance D.te: lanuary 7,1968.
Requirements for TS 3/4 2.3 would also 3,,7,,y,6)barkin, Change of Address at Region 1 be revised to reflect this new option of Morton W t
,3, p,,,,,f,, p,gy, Re ne w.
Effective January 19,1988, NRC's using either of the two new exposure (FR Dx 8H91 Fded 1-1M8. 8 45 am)
Region 1 Office wtil be mosed to a new ranges and to delete the existing option location. The new addien will be U.S.
of operating at 400'F or less.
Nuclear Regulatory Commission. Region (4) Res!se existing Figure 3 2.3-1.
owwo coos tsoo.et as
- 1. 475 Allendale Road. King of Passia.
MCPR vs Tau, by providing the MCPR Pr: posed Amendments Regarding Penns>lunia 19406. The current vs Tau curves for the first exposure Safeguards Requirements for Fuel commercial telephone number (215-337 range discussed abos e and resise Facilities Possessing Formula 5000) remains unchanged. The l'rS existing Figure 3.2 3-2 K, Factor by Ouantities of Strategic Special Nuclear switchboard number has been changed deleting the K, Factor curve and Materlat; Meeting to 6-346-5000. Indisidual staff members replacing it with the MCPR vs Tau may be reached by dialing FTS 6-346 curves for the second exposure range discussed above, AitNCY: Nuclear Regulatory 5XXX. as the last three digits of their Commission.
cunent telephone numbers remain the (5) Add a new Figure 3 2 3-3 with a ACT60N: Notice of meeting.
same as listed in the NRC Telephone new K, Factor curve for C)cle 2 D. rectory (NUREC/BR-0046) opera tion.
suuuARY:The NRC staff will discuss (6) Delete Table 3 2 3-1 which draft guidance related to the proposed Dated at Bethesda. Mar > land, this lith day currently prosides MCPR Feedwater amendments for safeguards of January 1988.
l{ eating Capacity Adjustments for requirements for fuel facilities For the Lclear Regulatory Ccmmissicn.
operation below 400'F, possessing formula quantities of Donnie 11. Grimsley, (7) Resise the TSs to allow operation strategic special nuclear material which Dactor. D.usma o/RWes and Remeds, abose the 100v Load Line and up to were puthshed on December 31,1987, O'fece cf Administe:f!0n ondReso rces 105% Rated Core Flow by:
for public comment (52 FR 49418).
MancFed (a) Extending the K, Factor curse up to 110% cf Rated Core Flow (instead of the DAtts: January 19-20,1988.
[FV Dec. 8%rt ided 1-1M8 8 45 am) cunent 2004 AooRess: White Flint 1,11555 Rockville mo coos vs o4w (b) Clamping the Upscale Setpoints Pike. Rocksille Mar >lar.d. 2085~,
for the Rod Block Monitor in TS Table FoR FuRTHER INFORM AT1oM CONT ACT 3 3 6-2 at the 100% recirculation flow Kr:stina Z. Jamgochlan, Office of IDochet No. 60-3541 ulue.
Nulcar Malenal Safety and Safeguards.
(clincreasing the Motor Generator Set Public Service Electric & Gas Co. and il S. Nuclear Regulatory Commission, mechan! cal and electrical stops in TS Aa je C E
.ir ton, DC 20555, telephone (301) is uanh of 4 41.13 to physically allow for n
increased core flow, Amendment to Facility Operating Pri r to issuance of the prosposed suPPLtutNT AR7 INFORM AtloN: The Ucense and Opportunity for Prlor license amendment. the Commission purpose of this meeting is to further the Hearing will hase made findings required by the licensee's understanding of the proposed amendments, which strengthen The U.S. Nuclear Regulatory Atomic Energy Act of1954. as amnded safeguards at fuel facihties possessing Commission (the Commission) is (the Act), and the Commission's formula quantities of strategic special considering issuance of an amendment regulations, r.uclear material and which upgrade the to Facility Operating License No. NPF-By February 10,1988. the licensees facilnies to a lesel equivalent to the 57 issued to Public Service Electric &
may file a request for a hearing with protection in place at comparable Gas Company and Atlantic City Electric respect to1:suance of the amendment to Department of Energy facihties.The Company (the licensees) for operation of the subject facility operating license and meeting will elicit industry comments on the l{ ope Creek Generating Statior any person whose interest may be the associated guidance and answer located in Salem County, New lersey, affected by this proceeding and who ope =tians on the proposed phy sical The p ope'd amendment would:
wishes to participate as a party in the smrst) hquirements.
(1) increase the Minimum Critical proceeding must file a written petition The meeting will be divided into Power Ratio (MCPR) safety limit in for leas e to intervene. Requests for a arssions for the NRC presentations and Technical Specifications (TS) 2.12 and hearing and petitions for leave to for licensee questions on the current 3/4 4.1 and in Bases sections related to inten ene shall be filed in accordance draf t guidance.,
these TSs.
with the Commission's Rules of l
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s REVISION 2 January 14,1988 TENTATIVE AGENDA t.CRS SUBCOMMITTEE ON STRUCTURAL ENGINEERING AMFAC HOTEL, ALBUQUERQUE, NEW NEXICO JANUARY 22, 1988 I.
Site Visit StOO - && a.m.
Tour of Concrete Model (SNL) [ Meet in AMFAC WI w ti HotelLobby]
11.
Subcomittee Meeting t o;6 0 to.13 b30 - 4:45 a.m.
Intrtoduction - J. Costello (NRC)
( 0 *.W4M5 - 14430 a.m.
Discussion of Concrete Podel Tests t 2. ; w (W. von Riesenann, et al, SNL)
- v. How and why failed?
- i. Comparison to prediction v'. Post tests and inspectiors g u e4.f ) -
tL' W -\\'.40 e
f t ' C-v h8 HT30 - 1:00 p.m.
Future work on reinforced and prestressed (SNL) conteinments
- v. A look at other types of steel, prestressed, etc. containrrents hb u / British work, if any
. Applicability of model tests to prototypical full scale containments
)t90,s.2:00 p.m.
- ).UHCn ******
2:00 - 3:00 p.m.
Completion of work on steel centainments (SNL)
/ Lw d E ut O 3:00 - 3:45 p.m.
Future efforts on containment penetrations -
(SNL) and seismic issues
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'M 3:45 - 4:15 p.m.
Wrap-Up(SNL,NRC) 4:15 - 4:30 p.n.
Subcommittee Discussion and Guidance
--450 p.ih.
ADJOURNMENT G M5 y< A -
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, LIST OF HANDOUTS d
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Agenda, Revision 2 of Structural Engineering Subcommittee meeting, January 22, 1988.
2.
Federal Register Notice, January 14, 1988.
3.
List of Attendees 4.
The Reinforced-Concrete Containment Model - Testing and ResQlts Daniel S, Horschel, January 22, 1938. Sandia National Lab.
5.
Ltr. to James F. Costello from Daniel S. Horschel, Sandia National Lab, dated August 2,1987 6.
Quick look Report: The Low Pressure Testing of the 1/6-Scale Containment, Daniel S. Horschel, Sandia National Lab., dated July 27,1987 7.
Coordination of Analysis of the ]:6-Scale Reinforced Concrete Containment Model, D. B. Clauss, Containment Integrity Division, Sandia National Lab., January 22, 1988 8.
Analysis of the ]:6-Scale Containment Model, J. Randy Weatherby.
Applied Mechanics Division I, Sandia National Lab., Jan. 22, 1988 9.
The CEGB's Prestressed-Concrete Containment Model, Daniel S.
Horschel, Sandia National Lab., Jan. 22, 1988
- 10. Investigation of the Leakage Potential of a Personnel Airlock, D. B. Clauss, Sandia National Lab., Jan. 22, 1988
- 11. Status of the Following Research Topics for Containment Penetrations, Seals and Gaskets Inflataole Seals Bellows, M. Brad Parks, Jan. 22,1988
- 12. Containment Integrity Programs, Sandia National Lab., Dave Clauss.
Dsn Horschel, Dwight Lambert, Brad Parks, Randy Weatherby, Jan. 22, 1988.
- 13. Analytical Predictions for the Performance of Sequoyah Unit 1 Containment, D. B. Clauss, J. D. Miller (Not on the record)