ML19312C879
| ML19312C879 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/20/1972 |
| From: | Peltier I US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001130049 | |
| Download: ML19312C879 (19) | |
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Docket Mos. 50-270-M 3 01972 and 50-287 R. C. DeYoung, Assintr.nt Director for Pressurized Water Reactors, Licensing Ti!RU:
A. Schwencer, Chief, Pressurized Water Reactors Branch No. 4. L MEETI13G VIT11 DUKE POWER COMPANY, BABCOCK AND WILCOX AND EECHTEL CONCERNING REVIEW OF Ti1E OCONEE 2/3 FOR OPERATING LICENSE, JUNE 16, 1972
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Enclosed is a su: mary of the meeting held on June l6,1972 with Duke Power Company, Babcock and Wilcox and Bechtel to discuss outstanding i
i concerns of the Machanical Engineering Brcoch, Technical Review. An attendance list is also enclosed.
k OHgtnal sigacalix trying A.Pdtiec j I. A. Paltier, Project Leader Pressurized Water Reactors Branch No. 4 Directorate of Licensing Enclosures 1.
Meeting Su:: nary 2.
Attendance List DISTRIBWTION_
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Docket (2)
R. S. Boyd RP Reading D. Skovholt PWR-4 Reading D. Knuth R. Haecary R. Tedesco H. Denton PWR Branch Chiefs i
R. W. Klecker n n p rp n RO I. A. Peltier
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Licensing Assistant (2)
E. L. Branumer S. H. Eau K. Kapur u
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1 ENCLOSURE I MEETING WITil DUKE POWER COMPANY, BABCOCK AND WILCOX AND BECHTEL June 16, 1972 1
SUMMARY
Attached as Enclosure IA and Enclosure IB are the concerns expressed by Technical Review, Mechanical Engineering Branch and a summary of the resolutions respectively.
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ENC 1.OSURE 1A REQUEST F0F. ADDITIONAL I2TIORMATIM OCONEE NUCLEAR STATION 1l NITS 2 & 3 DOCKET NOS. 50-270/287 3.6 CRITERIA FOR l'ROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED k'ITH A IICA 1.
Provide a more detailed description of the neasures that have been used to assure that the containment liner and all casential equip-cent within the containeent, including co=ponents of the prieary and secondary coolant systems, engineered safety features, and equipment saports, have been adequately protected against blow-
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dova jet forces, and pipe whip resulting from a loss-of-coolant I
accident. The description should include:
Pipe restraint design require ents to prevent pipe whip impa:t.
a.
b.
The features provided to shield vital equipment from pipe whip.
The measures taken to physically separate piping and other c.
co=ponents of redundant engineered safety features.
d.
A description of the type of pipe whip restraints and the location of all restraints.
2 Describe the dynamic system analysis ecthods and procedures that were used to confirm the structural design adequacy of the reactor coolant system (unaf fected loop) and the reactor internal lcadings.
The following infore.2 tion should be included:
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tha locations of the postulated double c.nded pipa rupture on which dynamic analysea vore based.
b.
the rupture type (s), such as circumferential and/or longitudinal break (s), for each postulated rupture location.
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the description of the forcing functions used for the pipe whip dynamic analyses. The function should include direction, rise time, magnitude, duration and initial conditions. The forcing function should adequately represent the jet strea:n dynamics
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and the syste:n pressure dif ferences.
i d.
a description of the =athe atical codel used for the dynamic analysis.
e.
analyses perfor=ed to demonstrate that unrestrained cotion of ruptured lines vill not sever adjacent impacted piping or pierce impacted areas of containment liner.
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- 3. [. 2 SEISMIC SYSTEM MIALYSIS 1.
Confirm the validity of a fixed base assumption in the mathe=atical models for the dynamic system analyses by providing su==ary analyt-ical results that indicate that the rockine and translational
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responnes are insignificant. A brief description should be included of the method, mathe=atical =odel and damping values (rocking verti-cal, translation and torsion) that have been used to consider the soil-structure interaction.
i 2
Describe the method employed to consider the torsional modes of vibration in the seismic analysis c f the Category I building structures.
If static factors are used to account for torsional accelerations in the scismic design of Category I structures, i
justify this procedure in lieu of a co=bined vertical, horizontal, and torsional cultimass system dynamic analysis.
3.
The use of both the modal analysis response spectrum and time history methods provides a check on the responses at selected points in the station structure.
Sub=it the responses obtained from both of these methods at selected points in the Category I structure to provide the basis for checking the seismic system analysis.
4 Provide the dynamic methods and procedures used.to determine Category I structure overturning momcats.
Include a description i
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4-g veedures used to account of the pro-for soil reactions and vertical hqua h --
5, Provide tT '
analysis procedure followed to account for the danping in diffe ele =ents of the =odel of a coupled system.
Include
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3.7.3 5"E15:11C SUBSYSTCt ANALYSIS-1 1.
Proutde the criteria for ec=hining modal responses (shears, moments, i
l stresses, deflectioas. and/or accelerations) when codal frequencies 1
are closely spaced and a response spectrum nodal analysis method is 7
used.
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With respect to Category I piping buried or otherwise located out-3ide of the containment structure, describe the seismic design criteria employed to assure that allowable piping and structural l
I stresses are not exceeded due to differential movement at support points, at containment penetrations, and at entry points into other atructures.
1 3.
Describe the evaluation performed to deternine seismic induced i
l ef fects of Category II piping systems on Category I piping.
t 4.
Provide the criteria employed to determine the field location bf seismic supports and restraints for Category I piping, piping system components, and equipment, including placement of snubbers and dampers. Describe the procedures followed to assure that the field location and characteristics of these supports and restraining devices are consistent with the assumptions made in the dynanic analyses of the system.
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3.7.4 CRI11.RI A FOR SEISMIC INSTRLTEWNf1CN Provide the following infornation with respect to the use of the sciamic instrumentation for the facility:
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reiscuss the seismic instrumentation provided and co= pare the proposed seismic instrumentation program with that described in AEC Safety Guide 12. " Inst ru=entation for Earthquakes." Submit the basis and justification for elements of the proposed program l
that differ substantially from Safety Guide 12.
2.
Provide a description of the seismic instrumentation such as peak 4
i recording accelerographs and peak deflection recorders, that will be installed in selected Category I (Class 1 Seis=ic) structures and on celected Category I (Class 1 Seismic) components.
Include the basis for selection of these structures and conponents, the j
basis for location of the instru=entation, and the extent to which
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this instru=entation vill be employed to verify the seis=le analyses j
following a seismic event.
3.
Describe the provisions that will be used to signal the control room operator the value of the peak acceleration level experienced in the tendon access gallery of the reactor containment structure to the control room operator within a few minutes af ter the earth-j quake.
Include the basis for establishing the predetermined values for activating the readout of the accelerozraph to the control room operator.
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Provide the criteria cnd procedures that vill be used to compare measured renponsca of Category I (Class 1 Seismic) structures and the selected comp ~onents in the event of an carthquake with th i
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results of the systes dynamic analyses.
Include consideration of
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different underlying soil conditions or unique structural dynamic j
characteristics that may produce different dynamic responses of Category I (Class 1 Seismic) structures at the site.
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3.7.5 SEISMIC 1&:s.CN CONTROL MEAS'u?ES Describe the design control 'easures (as specified in Appendix B -
of 10 CFR Part 50 "Quali Assurance Criteria for Nuclear Power Plants") implemented to assure that appropriate scismic input data (as derived from se.
.it system a'nd subsystem analyses including any necessary feedback from such analyses) are correctly specified i
1 to the manufacturer of Category I components and equipment to constructors of other Category I structures and systems.
The responsible design groups or organizations that will verify the t
l adequacy and validity of the analyses and tosts employed by manu-facturers of Category I co=ponents and equipment and constructors of Category I structures and systems should be identified.
A description I
of the review procedures e= ployed by each group or organization should be included.
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9 3.9.1 DYNAMIC SYSTD4 ANALYSIS AND TESTING 1.
Paragraph 1701.5.4 of the ANSI B31.7 Nuclear Power Piping Code excessive requires that piping shall be supported to prevent Submit vibration under startup and initial operating conditions.
y a discussion of your vibration operational test program which will be used to verify that the piping and pipine, restraints within the reactor coolant pressure boundary have been desianed to withstand Provide a dynamic effects due to valve closures, pump trips, etc.
3 list of the transient conditions and the associated actions (pump J
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trips, valve actuations, etc.) that will be used in the vibration l
operational test program to verify the intec,rity of the system.
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Include those transients introduced in systers other than the reactor coolant pressure boundary that will result in significant vibration response of reactor coolant pressure boundary systems and components.
Discuss the testing procedures used in the design of Category I 2.
mechanical equipeent such as fans, pumps, drives, valve operators t
and heat exchanger tube bundles to withstand scissic, accident and i
operational vibratory loading conditions, including the r.anner in which the methods and procedures erployed will consider the frequency spectra and amplitudes calculated to exist at the equiprent supports.
Where tests or analyses do not include evaluation of the equipnent in the operating mode, describe the bases for assuring that this e
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loadings.
3.
Provide a brief description of the dynamic system analysis methods and procedures used to determine dynamic responses of reactor internals and associated Class I components of the reactor coolant pressure i
boundary (e.g., analyses and tests).
The discussion should include the preoperational test program elecents described in Safety Guide 20, Vibration Measurecents on Reactor Internals.
In the event j
clements of the program differ substantially from the require-
=ents of Safety Guide 20, the basis and justification for these dif fercnces should be presented.
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Provide a discussion of the preoperational analysis and testing i
results that vill be used to aug=ent the LOCA dynamic. analysis I
methods and procedures, i.e., barrel ring and beam modes, guide i
tube responses, water mass and compliance ef fects, damping factor selection, etc.
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~ AA 3.9.2 ASHE CODE CIASS 2 AND 3 COM?cNENTS 1
The FSAR states that faulted operating condition cat egories have been applied to certain reactor coolant system components.
Iden tify any other components or systc=s that are not a part of the reactor i
coolant pressure boundary for which the design stress limit s
associated with faulted conditions were applied If faulted conditions are used for such cases, then provide justification for applying auch conditions, including the bases for the loading con-ditions and combinations, and associated design stres s limits which vere applied.
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In addition, for all components and systems comparabic t i
o ASME Code L
Class 2 and 3, provide the design condition categories (
nornal, upse.t or ccergency),
the associated design loading combinations and th e
design stress limits which will be applied for each l oading combination.
This inforr.ation may be submitted in tabular form as s uegested below:
System Design Loading Combinations Desir.n Condition and/or
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Desien Component Categories (Normal, Stress Upset, or E=crgency)
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2 If any design stress limits allow inelastic defor lli mation (or are comparable to the faulted condition limits defined in ASME S ection III for Class I components) 1:
then provide the bases for the use of inclastic design linits by demonstratine that the co nonent l
will naintain its functional or structural integrity under th 4
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design loading conbination.
Include a brief description of the methots and design procedures that were used in such cases.
3.
Describe the design and installatica criteria applicable to the mountinr, of the pressure-relievine, devices (safety valves and relief valves) for the overpressure protection of systems with Class 2 components.
In particular, specify the desica criteria used to take into account full dischar,;c loads (i.e., thrust,
- ending, torsion) imposed on valves and on ccnnected piping in the event all the valves are required to discharge.
Indicate the j
provisions made to acconnodate these loads.
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5.2.1 DESIGN CRITERIA, METHODS AND PROCEDURES (REACTOR COOI. ANT PRESSURE EOUNDAT-?)
1.
Categorite all transients or combinations of transients listed in Table 4-8 of the FSAR vith rc pect to the conditions identified as "nortul", " upset", "enerEency", or " faulted" as defined in the ASME Section III Nuclear Component Code.
In addition, provide the design loading combinations and the associated stress or deformation criteria.
2.
Table 4-20 of the FSAR includes faulted condition stress li=its for pu=p casings.
Describe the criteria c= ployed to assure that active
- I conponents vill function as designed in the event of a postulated piping rupture (Faulted Conditien) within the reactor coolant pressure beundary (c.a., 'allevable stress limits established at or near the yield stress calculcted on an clastic basis).
Where e pirical rethod.
(tests) are c= ployed, provide a su==ary description of test nethods, loading techniques and results obtained including the bases for extrapolations to co:ponents larger or s= aller than those tested.
3.
The design criteria which was used to account for full discharte loads (i.e., thrust, bendine, torsion) i= posed on safety and relief valves and connected piping in the event all valves are required to discharre, including the provision ride to acco nodate these loads should be specified.
- Active components or a fluid system (e.c., valves, pueos) are those whose oper-aoility is reli:2 upon to perforn a safety functica sach as safe shu:covn of tae reactor or cititation of the con:cquences of a postulated piec bre k in the reartcr 1
coolant presauce boundary.
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s 5.2.2 0Vi:9PFl:S"UR17.ATION PROTECTION 1.
To facilitate review of the basca for the pressure relieving capacity of the reactor coolant pressure boundary, subnit (as an appendix to the FSAR) the " Report on Overpressure Protection" that has been prepared in accordance with the require ents of ie AS"E Section III Nuc1 car Po er Plant Cenponents Ccde or, if the report is not available, indicate the app oximate date for submission.
In the event the report is not expected to be available until either the Operating License review or late in i
the construction schedule for the plant, provide in the FSAR the bases and analytical approach (e.g., prelininary analyses) being utilized to establish the overpressure relieving capacity required for the reactor coolant pressure boundary.
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I ENCLOSURE IB i
RESOLUTION OF MECHANICAL ENGINEERING BRANCH CONCERNS 3.6 Information previously provided in the PSAR is sufficient review.
to complete 3.7.2 Seismic System Analysis j
1.
The applicant will docucent that addition of soil springs did not contribute significantly to the floor responses and structural
- stresses, j
2.
The applicant will demonstrate that the consideration of torsional modes does not contribute significantly to structural responses and stresses based on analyses of similar containments.
l 3.
This question is withdrawn.
4.
The applicant will provide the factor of safety against seismic overturning moments for the containment.
5.
This question is withdrawn.
3.7.3 1.
Question deleted.
2.
Duke will =ake a statement that there is no Category I piping in the Oconee Station.
3.
Applicant will describe the interaction between Category I and Category II piping.
4.
Duke will document Q/A procedures to show the field location of piping supports and restraining devices are consistent with assumptions made in the design analysis.
3.7.4 Applicant will install additional strong motion accelerometer and an additional peak recording accelerograph on Unit 1.
4 Provide a state =ent concerning criteria and procedures that will be i
used to compare maacured responses to their analysis, 3.7.5 Question withdraen.
i 3.9.1 1.
Duke will su=marize surveillance done on piping systems for i
vibration and thermal. movements.
i 2.
Duke will summarize pieces of equipment which actually received
. qualification tests and what the tests were.
1 3 & 4. B&W will submit a topical report relating to Oconee 1 internal redesign. Duke will commit to meeting Safety Guide 20.
(Reference g
appropriate B&W document) 4; i
3.9.2 1&2 Applicant will make a statement that faulted operating conditions were not applied to any components that are not a part of the reactor coolant pressure boundary.
Design stress limits for components comparable to ASME code class 2 and 3 did not allow inelastic deformation.
3.
MEB will review Amendment 32 for safety valve analysis.
Duke will do analysis on closed system including monentum effects and present in future amendments.
5.2.1 1.
Applicant will submit a revised Table 4-8 in the FSAR categorizing each transient with respect to normal, upset, energency or fau]ted conditions.
2.
Question deleted.
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3.
Same as 3.9.2 Question 3, with respect to reactor coolant pressure boundary.
5.2.2 1.
Applicant will reference topical report BAW-10043.
Explanations:
3.6 A separate meeting was held between AEC Licensing and Babcock and Wilcox to review this itcm and the conclusion was that the record (FSAR) contained sufficient information for MIB to complete its review.
3.7.2 - 2 The conclusion is based on Nev= ark's paper regarding 10% eccentricity, identified at the 4th World Conference on Earthquake Engineering, Santiago, Chile, "Torcions in Sy:etrical Buildingc."
3.7.2 - 3 It was concluded at the meeting that there is no need for this conparisca.
3.7.2 - 5 A constant damping factor was used in the analysis.
3.7.3 - 1 The square root of the sum of the squarea cethod was used in the analysis for combining model responses. No other cethod was used for closely spaced model frequencies.
3.7.3 - 2 Duke will take into consideration the condenser cooling water lines in its response.
3.7.5 This concern is the same as 3.7.3 and will be answered with 3.7.3.
5.2.1 - 2 Table 4-20 of the FSAR does not address faulted conditions.
GENERAL - Duke has agreed to voluntarily submit the above information for the record (future amendment).
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l ENCLOSL*RE II MEETING VITH DUKE PO* DER COMPAW AND bat 00CK AND WILCOX COMPAW_
~ ATTENDANCE LIST June 16, 1972 NAME ORGANIZATION TITLE OR FUNCTION
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l I. Peltier AEC, Licensing PWR-4, Project Leader
- A. Schwencer AEC, Licensing PWR-4, Chief
- E. O. Hooker Babcock & Wilcox Unit Manager Systems Design
- R. R. S teinke Babcock & Wilcox Licensing l
R. E. Miller Duke Power Ct=pany Senior Engineer, Mechanical Design R. J. Annoll Duke Power Company Steam Products, Licensing i
A. A. Vizzi Bechtel Engineering Super. Civil D. S. Mehta Bechtel Senior Engineer, Bechtel (Civil)
H. L. Brancor AEC, Licensing Mechanien1 Engineering Branch
- S. H. Hou AEC, Licensing Mechanical Engineering Branch K. Kcpur AEC, Licensing Mechanicci Eneineering Branch N. Davison AEC, Licensing Machanical Engincaring 3 ranch D. Lange AEC, Licensing Mechanical Engineering Branch Chief
- L. R. Allen Babcock & Wilcox Associate Project ManaBer i
- H. J. Fortuns Babcock & Wilcox Principal Engineer l
R. V. Straub Babcock & Wilcox Project Manager i
W. H. Owen Duke Power Cocpeny Vice President -
i Engineering i
W. H. Scheffler Duke Power Company Design Engineering E. J. Keith EOS D. A. Codfrey Duke Power Co:peny ' Design Engineering W. O. Parker Duke Powar Company Assistant Manager Steam Production K. S. Canady Duke Power Company System Nuclear Engineer S. B. Hager Duke Power Co=pany Design Engineer J. W. Childro Bechtel Project Engineer W. E. Ridgeway Bechtel Engincar Spec., Moch.
W. P. Brittle Bechtel Engineer, Civil Paul H. Barton Duke Power Cocpany Mgr. Tech & Nuclear Services
- Attended separate meeting on concern 3.6 and reported back to main meeting on findings.
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