ML20004G090
| ML20004G090 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 06/22/1981 |
| From: | Paulson W Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8106290165 | |
| Download: ML20004G090 (19) | |
Text
{{#Wiki_filter:. [pu% o,, UNITED STATES -{ g g NUCLEAR REGULATORY COMMISSION 3 ** .E WASHINGTON, D. C. 20555 k June 22,1981 Docket No. 50-213 i LICENSEE: CONNECTICUT YANKEE ATOMIC POWER COMPANY (CYAPC) FACILITY: Haddam Neck Plant
SUBJECT:
SUMMARY
OF MEETING HELD WITH CONNECTICUT YANKEE ATOMIC POWER COMPANY TO DISCUSS SEISMIC DESIGN CONSIDERATIONS (SEP TOPIC III-6) FOR THE HADDAM NECK PLANT On May 21,1981 a meeting was held in Bethesda, Maryland with representatives of Connecticut Yankee Atomic Power Company. The purpose of the meeting was (1) to discuss the licensee's SEP seismic reevaluation status, (2) discuss and schedule), and (3)pril 24,1981 (Seismic reevaluation program plan scope the NRC letter dated A (Lawrence Livermore Laboratory) questions raised by the NRC contractorA list of attend clarify Handouts provided by the NRC staff are also enclosed (Enclosure 2). Connecticut Yankee Atomic Power Company representatives discussed the scope and schedule of the seismic reevaluation program for Haddam Neck. Copies of the viewgraphs used during this presentation.are enclosed (Enclosure 3). The licensee's schedule indicates that the criteria document f,or the. piping analysis is scheduled to be submitted to the NRC in August 1981. The staff inotcated that we need this document before August 1981 to stay within the review senedule. During the discussions, the NRC staff indicated that the preliminary site specific spectra previously forwarded to the licensee will not change. The staff cautioned the licensee that in generating the in-structure response spectra using the site specific spectra, the use of only one time history may not be adequate because of inadequate frequency content, amplitude, and duration. The licensee should justify the adequacy of the time history or histories used in 'the analysis. n / .[46bs ((#db Walter A. Paulson, Project Manager. Operating Reactors Branch #5 Division of Licensing
Enclosures:
As stated cc w/ enclosures: See next page soo l$6 "m
. June 22,1981 cc Day; Berry & Howard Mr. W. G. Counsil, Vice President Counselors at Law . Nuclear Engineering and Operations -One Constitution Plaza Connecticut Yankee Atomic Power Company Hartford, Connacticut 06103 Post Office Box 270 Hartford, Connecticut 06101 Superintendent Haddam Neck Plant RF0 #1 Post Office Box 127E East Hampton, Connecticut 06424 Mr. James R. Himmelwright Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Middletown, Connecticut 06457 Board of S21ectmen Town Hall Haddam, Connecticut 06103 Connecticut Energy Agency ATTN: Ass ~1stant Director Research and Policy Development Department of Planning and Energy Policy 20 Grand Street Hartford, Ccnnecticut 06106 U. S. Environmental Protection Agency Region I Office ATTN: EIS COORDINATOR JFK Federal Building-Boston, Massachusetts 02203 Resident Inspector Haddam Neck Nuclear Power Station c/o U. S. NRC East Haddam Post Office East Haddam, Connecticut 06423
1 Enclosura 1 -l l ? ATTENDANCE LIST f W. PAULSON NRC R. SCHAFFSTALL KMC A. MAGID NUSCO E. DEBARBA NUSCO -C. GLADDING NUSCO i M. BAIN NUSCO T. Y. LO LLNL~ T. NELSON LLNL D. WESLEY SMA W. RUSSELL NRC P. Y. CHEN NRC K. HERRING NRC R. HERMANN NRC R. LAUDENAT' NUSCO e .I
1 GUIDELINES AND PROCEDURES FOR THE REVIEW 0F i SEISMIC QUALIFICATION OF SEP GROUP 11 PLANTS (San Onofre 1, Lacrosse, Big Rock P0 int, Yankee Rowe, Haddam Neck, and Dresden 1) I. BACrsGROUND In order to determine the margin of safety of the sekcted eleven operating nuclear power plants relative to those designed under current standards, criteria, and procedures, and to define the nature and extent of retrofitting to bring these plants to acceptable levels of capability if they are not already at such levels, the Office of Nuclear Reactor Regulation (NRR) has been proceed-ing with Phase II of the Systematic Evaluation Program (SEP) since October 1977 through the review of 137 safety topics developed in Phase I of the SEP. The seismic design considerations, Topics II-4. A. B, and C, III-6,111-11, and IX-1 are among the 137 safety topics to be reviewed. II. SCOPE AND OBJECTIVE The objective of Topic III-6 (including Topics III-11 and IX-1) is to review and evaluate the seismic resistance of facilities. As a mininum, the seismic progran should provide for an evaluation of: 1. The integrity of the reactor coolant pressure boundary, 2. The capability of systems and components to shutdown the reactor and maintain it in a safe s, hut'down condition (i.e., Topic VII-3), 3. The capability of systems and conponents necessary to mitigate the consequences of accidents (i.e., Topics XV on design basis events) including fuel storage (Topic IX-1), and 4. The integrity of structures containing the Items 1, 2 and 3 above. III. GENERAL CRITERIA AND REFERENCES The bases used by the staff for the review and evaluation will be the following: 1. NUREG/CR-0098, " Development of Criteria for Seismic Review of Selected Nuclear Power Plants," N. M. Newmark and W. J. Hall, May 1978. 2. The Final Ground Response Spectra from the NRC SEP Site Specific Spectra P roject. The interim spectra was fomarded by the August 4,1980 NRC. letter to the Group II owners. 3. "SSRT Guidelines for SEP Soil-Structure Interaction Review," N. M. newmark, December 8,1980. 4. Standard Review Plan, Sections 2.5, 3.7, 3.8, 3.9, and 3.10.
. 5. Regulatory Guides 1.26, 1.29, 1.60, 1.61, 1.92, 1.100, and 1.122. 6. For mechanical and electrical equipment not covered by NUREG/CR-0098 or Regulatory Guide 1.61, a damping value of up to 4 percent can be used for qualification by analysis. A higher damping value may be used if sufficient justification is provided and found acceptable. 7. In general, Items 1, 2 and 3 should be used as a group in cases where the criteria in Items 1, 2 and 3 differ from those in the Standard Review Plan or Regulatory Guides. IV. GENERAL PROCEDURES A) Licensees are to implement a program to perform the required analysis and to submit a Safety Analysis Report (SAR) addressing the analysis program, ~ scope of analysis performed and the results of the analysis. In accordance with 10 CFR 50.54(f) of the Commission Regulation, a letter was issued on August 4,1980 to Group II plants licensees requesting that the licensee:
- 1.. submit details of a seismic evaluation program plan addressing the scope of review, evaluation critaria and a schedule for co@!etion; and 2.
provide justification for continued operation in the interim until the program is coglete. The program plan should be acceptable to the NRC Staff. B) When the analysis is cogleted, the licensee will submit the Safety Analysis Report for staff review. In addition to the detailed descriptions of the methods and procedures used, and the results obtained, a summary of the evaluations of structure:, systems and components should be provided in the SAR. The licensee should provide the following information in tabular form for each equipment item: 1. Method of qualification used: (a) Analysis or test or a combination of test and analysis (indica'te the cogany that prepared the report, the reference report number and date of the publication).
j l (b) If by test, describe whether it was.a single or multi-frequency test and whether input was single axis or nulti-axis. (c) If by analysis, describe whether static or dynamic, single or nultiple-axis analysis was used. Provide natural frequency (or frequencies) and the danping value used in the inalysis. 2. Indicate whether the equipment has met the qualifiation requirements or.hether modification is required. 3. Indicate the system in which the equipment is located and whether the equipment is required for: a) hot stand-by b) cold shutdown c) both d) neither 4. Location of system or conponent, i.e., building and elevation. 5. Seismic input (indicate the spectra cr acceleration used). 6. Indicate whether it is within the scope of NSSS or B0P. V. STAFF REVIEW PROCEDURES A) Based on the licensee's SAR and the partinent reports referenced therein, the staff and available contractor or consultants will conduct the reviews. The staff review procedure will be different from that of Group I plants in that the review of Group II plants will be similar to an Operating License review. B) The staff will conduct a site review of the qualification methods, procedures, and results for a list of selected safety related structures, systems, and conponents and their supporting structures. The intention is also to observe the field conditions and installations of structures, systems, and conponents, based on which judgments can be made as to the validity of the mathematical modeling enplojed in the evaluation program, with respect to the configurations and the boundary conditions. C) In the evaluation of structures, the staff will review at least the follewing: 1. Adequacy of site ground response spectra and synthetic time histories input to structures for SSE. 2. Adequacy of structural mathematical model.
. 3. Possible effect of such parameters as damping, soil-structure interac-tion, torsion, and overturning. 4. Normal, seismic and seismic related loadings, load combinations, stresses, and deformations. 5. Adequacy of floor response spectra. 6. Relative notions which might influence piping entering buildings, or spanning between buildings, tilt, or interaction effects. D) In the evaluation of systens, components, and their supports, the staff will review at lesst the following: 1. Adequacy of the inputs to each system or component under SSE loading. 2. Adequacy of the analytical model and assumptions used to sinulate the field installation conditions. 3. Structural integrity of the systens, components, and their supports. E) In conducting the reviews, depending on the circumstances, the staff nay want te perform some independent evaluations such as confirmatory analyses or consultant's view to enhance the overall review evaluations. From among the list of selected safety related structures, systems.,and componnets to be reviewed, the staff will conduct an independent confirmatory analyses for the following: 1. Containment building and other buildings deemed desirable. 2. Some piping systems. 3. Some mechanical and electrical equipment. F) After each site visit, the staff will issue a trip report, which will identify findings, conclusions and follow-up items. G) At the conclusion of review, the staff will issue a Safety Evaluation Report (SER) for the Topic III-6. e l 1
o References ngp bhl',3 Ek i 1. "Haddam "eck Plant Systematic Evaluation Program, Seismic Reevaluation", Docket ho. 50-123, 810051, W. G. Counsil to D. M. Critchfield, August 5,1980. 2. "Haddam Neck - SEP topic III-6", " Seismic Design Consideration", D. M. Crutchfield to W. G. Counsil, Docket 50-123, LS05-81-04-008, April 8,1981. QUESTIONS 1) In Ref. 1 several ground response spectra are reference.1. Verify that the evaluation is still proceeding using the URS/J.R. Blume spectra which is r,eppetentative of the 84th percentile (+1 o ) for rock sites. ~.. Discuss the basis of using the 0.17g SSE peak ground acceleration rather than the 0.21g SSE developed by LLNL for the site specific spectra. If any u'se of the Weston Geophysical site specific spectra are con-tenplated, justify the low respon'se in these spectra in the 1 to 3 Hz range. US NRC Regulatory Guide 1.60 is included in Ref.1 as a general re-ference. Are these spectra expected to be used in the analysis? 2) The majority of the important Haddam Neck structures are founded on competent rock so that soil-structure interaction effects will be minimized for these structures. Describe the modelling procedure to be used in calculating soil compliance functions for structures founded partially on rock and partially on fill or lean concrete. ~ Describe the location for the seismic input to 'the structures which are partially founded on fill and partially founded on rock. Discuss the modelling of an,e embedment effects due to backfill. e.*
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- 3) Will the diesel generator building be reanalyzed for the 0.21g SSE peak pund acceleration?
4) for reinforced concrete structures with only slight cracking 5% of critical damping may be slightly nonconservative, thus in - creasing the input to equipment. Verify that 5% of critical damping will be used only for concrete structures with consider-able cracking.
- 5) Discuss the methods of developing equivalent static loads and the justification for using these methods if these methods ara to be used.
'7 Discussthecri$riatobe'usedindecidingwhethercoupledstruc-tural models are required or whether uncoupled models are adequate. Describe the generation and treatment of structure to structure loads if uncoupled models are used. 6) Depending on the dispersion in material properties, a + 5% broaden-ing of spectral peaks for the in-structure response spectra based on average properties may not be conservative enough to account for modelling uncertainties, particularly if aging is not considered. 4
- 7) Structural foundation are stated to have a factor of safety of 1,1 against sliding and overturning. What will be the basis of the loads used for this evaluation? (i.e.equivalentstatic. loads corresponding to base shears and moments or time history loads?)
- 8) Supply a list of computer programs to be used in the analysis and the basis of their verification.
- 9) The discussion of the equipment provided in the program plan (Ref.1) is limited to the reactor coolant system (RCS). Provide a list of systems.or portions of' systems to be evaluated as
~ \\
i bl f. ~ l tr t all of these systems are expected to be outlined in Ref 2, o Class I. Provide a discussion of methods of analyses and accep-tance criteria for these systems if they vary from the criteria used for the RCS.
- 10) The criteria proposed for evaluation cf the RCS is essentially consistent with current criteria which in some cases provides fcr higher stress allowables (although usually with increased load combinations) than were used in the design.
Implicit in the use of higher allowables is the assumption that current QA pro-cedures are in effect r P'rovide a discussion of the QA procedures in use during the construction of Haddam Neck including how the design QA procedures provide an adequate basis for the use of current and in some cases, higher stress allowables. Include a discussion of the program to be used to verify the as-built condition of piping and equipment installation to withstand ' seismic effects.
- 11) Tables 1, 3, 4 of Ref.1 are essentially consistent with current code allowables. However, in Table 3, a compressive axial load of up to 0.9 times the critical buckling load is proposed. This
~ should be justified. 121 No discussion of cable trays or the methods to be used to determine their seismic response or acceptance criteria is presented in Ref.l. This information should be provided. ~ 131 No discussion of qualification of electrical equipment is presented in Ref.1. A minimum program to verify anchorage adequacy is antic- ' ipated. The nethod of qualification of electrical equipment for nonstructural failure modes together with a list of components, test metho'ds, and acceptance criteria, etc. should be provided. e e eG* 4
A HADDAM NECK PLANT -- SEISMIC RE-EVALUATION PROSRAM SEP TOPIC III-6 MAY 21,' 1981 1 1. INTRODUCTI0ll -- OVERVIEW 0F MAJOR PROGRAl:S: A. WESTON 6EOPHYSICAL -- SITE SPECIFIC SPECTRA, 6EOLOGY B. WESTINGHOUSE -- RCS PRESSURE BOUNDARY C. URS BLUME -- STRUCTURES D. BLUME -- CABLE TRAYS E. WESTINGHOUSE -- ELECTRICAL EQUIPMENT F. SEISMIC ANCHORAGE G. MASONRY WALLS -- 80-11 H. PIPING ANALYSES -- 79-14 I. GENERIC LETTER 81-14 -- AUXILIARY FEEDWATER SYSTEM II. PROGRAM DESCRIPTION / STATUS III. LLL EVALUATION g g. IV. SCHEDULE En m V. APRIL 8,1981 LETTER 4 e e
i l ANALYSES OF STRUCTURES HOUSING SAFETY RELATED EQUIPMENT I. CRITERIA DOCUMENT TRANSMITTED TO NRC JANUARY 17)1980. II. SCOPE OF ANALYSES: A. Cr..:rAINMENT SHELL B. CONTAINMENT INTERIOR STRUCTURE C. SCREENWELL D. PRIMARY AUXILIARY BUILDING E. SERVICE / TURBINE BUILDINGS F. AUXILIARY FEEDWATER BUILDING i III. SEISMIC INPUT: A. SAFE SHUTDOWN EARTHQUAKE CHARACTERIZED BY HORIZONTAL RESPONSE SPECTRA TRANSMITTED TO NRC, AUGUST 5, 1980 B. VERTICAL SPECTRA 2/3 HORIZORTAL C. TIME HISTORY DEVELOPED MATCHING SSE RESPONSE SPECTRA USED AS INPUT FOR 6ENERATING FLOOR RESPONSE SPECTRA t
ATTACHMENT 1 HADDAM NECK \\ K;- ioe0.o ~ ~ l 1 4 / / a { 100.0 5 /\\ Haddam Neck Plant Spectra e J g In Use ~ % s e / b ?, b
- O NRC uggested l'"(
A /. l .21 g ZPA /
- O
{ NRC Suggested / Site Spectra 1 ~ g / H ,c / / / .17 g ZPA Haddam Neck ",, g, Spectra In Use c
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0.1 ..i-0.0i 0l IO IO** PERIOO-5EC. 5% damping . - + -. _.--,-,.-----.r-.
IV. CONDITIONS TO BE ANALYZED: A. DEADWEIGHT B. ACTUAL LIVE LOAD C. OPERATING TEMPERATUPE AND PRESSURE LOADS D. SAFE SHUTDOWN EARTHQ6?KE LOAD V. STATUS OF ANALYSES AND STRUCTURAL EVALUATIONS: A. CONTAINMENT SHELL ANALYSIS AND EVALUATION COMPLETE B. CONTAINMENT INTERIOR STRUCTURE ANALYSIS AND EVALUATION COMPLETE C. SCREENWELL ANALYSIS AND EVALUATION COMPLETE D. AUXILIARY FEEDWATER STRUCTURE --- ANALYSIS AND EVALUATION COMPLETE E. SERVICE / TURBINE BUILDING NNALYSISCOMPLETE;EVALUATIONINPROGRESS F. PRIMARY AUXILIARY BUILDING --- ANALYSIS IN PROGRESS; EXPECTED COMPLETION DATE JULY 1981 VI. GENERATIONGFIN-STRUCTURERESPONSESPECTRA IN-PROGRESS, EXPECTED COMPLETION DATE JULY 1981 4 4 't 9
~ VII. SUBMITTALOFFINALREPORTONSTRUCTbRALANALYSESANDEVALUATIONSCHEDULED FOR NOVEMBER 1981 VIII. IMPELMENTATION OF STRUCTURAL MODIFICATIONS: A. MINOR STRUCTURAL MODIFICATIONS.TO PRIMARY AUXILIARY BUILDING AND d SCREENWELL IN PROGRESS B. STRUCTURAL MODIFICATIONS TO IURBINE/ SERVICE BUILDING SCHEDULED TO BE STARTED IN JUNE 1981 C. FINAL COMPLETION OF hLL REQU, IRED STRJJCTURAL MODIFICATIONS DURING OR PRIOR TO 1983 REFUELING OUTAGE e I
l l 1 l l REACTOR COOLANT S! STEM ANALYSIS l I. CRITERIA DOCUMENT SUBMITTED TO NRC -- JANUARY 17, 1980, II. SCOPE OF ANALYSIS A. PIPING -- Il0T LEGS, COLD LEGS, CROSSOVER LEGS, PRESSURIZER SURGE LINE B. COMPONENTS -- CONTROL ROD DRIVE MECHANISM -- REACTOR VESSEL AND INTERNALS -- STEAM GENERATOR -- REACTOR COOLANT Pune -- PRESSURIZER C, COMPONENT SUPPORTS -- REACTOR PRESSURE VESSEL SUPPORTS -- STEAM 6ENERATOR SUPPORTS - REACTOR COOLANT PuMe SUPPORTS -- PRESSURIZER'8DPPORTS III. CONDITIONS TO BE ANALYZED DEADWEIGHT PRESSURE THERMAL SAFE SHUTDOWN EARTHQUAKE
A. Il0DELING.01 UVEU3ALL KECTOR LOOLANT bYSTEM LOMPLETE B. DETAILED STEAM GENERATOR MODEL COMPLETE C. DETAILED REACTOR COOLANT PUMe MODEL COMPLETE D. REACTOR PRESSURE VESSEL AND FUEL DETAILED MODELING IN PROGRESS V. SCHEDULED COMPLETION DATE A. ANALYSIS COMPLETE BY NOVEMBER 1981 B. ANY REQUIRED STRdCTURAL MODIFICATIONS INSTALLED DURING OR PRIOR TO THE SCHEDULED 1983 REFUELING OUTAGE + e S i e
II. SCOPE OF ANALYSES: A. SAFE SHUTDOWN PIPING GREATER THAN IWO INCHES NPS B. NON-ISOLABLE PORTIONS OF THE REACTOR COOLANT SYSTEM PRESSURE BOUNDARY (UP TO FIRST NORMALLY CLOSED ISOLATION VALVE, OR PIRST AUTOMATICALLY CLOSING ISOLATION VALVE) PIPING GREATER THAN IWO INCHES NPS Ill. STATUS: A. IS0 METRICS DEVELOPED DURING I&E BULLETIN 79-14 PROGRAM B. MODELING OF SEVERAL PIPING SYSTEMS IN-PROGRESS ~ IV. SCHEDULE: A. ANALYSES AND llACKFITS TO BE PERFORMED HAND-IN-HAND B. SUPPORT MODIFliCATIONS TO BE STARTED IN SCHEDULED 1981 REFUELING OUTAGE ~ C. FINAL COMPLETION OF ANALYSES AND BACKFITS SCHEDULED FOR JUNE 1983 ' x. -s l i l I
,/ ~, (. 4 MEETING
SUMMARY
DISTRIBUTION Oceket NRC PDR Local PDR ORB Reading J. Olshinski S. Varga R. Clark T. Ippolito JStolz D. Crutchfield H. Smith W. Pablson OELD OI&E(3) g ACRS (10) 'nsicy. TERA E. Adensam W. Russell D P. Y. Chen N K. Herring R. Hermann G) , [ D 6 LULL l L , 84 Jutt 2 5198F 9.5. @ u 9 b /mi e -f R t . - ~ .-}}