ML20134P031
| ML20134P031 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 08/30/1985 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20134P025 | List: |
| References | |
| NUDOCS 8509060112 | |
| Download: ML20134P031 (20) | |
Text
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EVALUATION OF REACTOR COOLANT SYSTEM LOADS AND COMPONENT SUPPORT MARGINS RESULTING FROM OPTIMIZED REACTOR COOLANT PUMP SUPPORT CONFIGURATION CRYSTAL RIVER 3 GENERATING PLANT Prepared for Florida Power Corporation by Babcock and Wilcox, Lynchburg, Virginia 4
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Florida Power Corporation (FPC), Babcock & Wilcox (B&W) and the NRC staff have discussed on several occasions the application of advanced fracture mechanics techniques to certain postulated pipe
' breaks in the Reactor Coolant System (RCS) main loop piping.
PPC, based on these discussions and information provided has proposed to utilize those techniques to eliminate mechanical and structural load ef fects associated with postulated RCS main loop pipe breaks.
In Reference (1), the NRC indicated that advanced fracture mechanics 'could be employed as a basis for an alternate approach these postulated pipe breaks.
FPC in Reference (2) submitted a request for pa r tial exemption from General Design Criteria 4 (GDC-4).
Specifically in Reference (2), FPC requested a partial exemption from those portions which require protection of structures, systems, and components against certain dynamic (including mechanical and structural loading) effects associated with postulated RCS main loop pipe breaks.
This exemption pertains to all postulated breaks specified in the Crystal River-3 RCL piping.
FPC has not requested exemption from GDC-4 for other postulated breaks.
The request does not affect the CR-3 Nuclear Generating Station design basis for environmental, containment, equipment qualification or ECCS analysis.
The following information is provided as additional justification for the exemption:
1.
In Reference (3), Babcock and Wilcox submitted for NRC staff review a fracture mechanics analysis to validate the
" Leak-Before-Break" (LBB) failure scenario for their Nuclear Steam System (NSS) designs.
Staff review of this submittal is complete and no significant deficiencies have been identified.
Approval is pending based on completion of materials testing and minor revisions.
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2.
Reference (3) demonstrates that for the NSS-177 FA RCS main loop piping:
a)
A substantial sized flaw in the piping would not grow through the wall nor extend significantly in length during the plant design lifetime, b)
If a flaw were to grow through the wall of the piping, it would open sufficiently to leak many times in maximum allowable leakage before extending anywhere near critical crack length.
c)
A very long through wall crack (many times longer than a leak detectable longitudinal or circumferential crack length) would remain stable under normal operation plus SSE loadings.
This demonstration provides sufficient justificat. ion for elimination of large postulated breaks from the design basis for the CR-3 Unit's RCS main loop piping.
3.
The ACRS in Reference 4 has approved the application of the aforementioned fracture mechanics techniques to the analysis of asymetric blowdown loads.
Reference (4) states "That is, there is no known mechanism in PWR primary piping material.-
for developing a large break without going through an extended period during which the crack would leak copiously."
4.
Appendix A presents an assessment of the new RCS pump supports configuration loadings with respect to the loadings evaluated in the LBB Report, Reference 3.
The new RCS pump support configuration was outlined in the GDC-4 exemption request, Reference 2.
The new support configuration loadings are enveloped by the generic analysis performed in support of the LBB report.
The generic loadings exhibit a margin of 6.6% over the new FPC loadings at the maximum load location.
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At other locations, the margin is significantly larger.
5.
Appendix B presents an assessment of the RCS Components Supports seismic margin using loadings from the new RCS pump support configuration analysis.
The new support configuration margins are calculated using techniques employed in the LLNL probalistic evaluation of B&W plants, Reference 5.
The new and old margins are listed along with the minimums margin of any B&W plant evaluated.
In conclusion, there would be no adverse effect on safety resulting from the exemption.
If the exemption is granted, it would have no effect on the potential for occurrence or the severity of accidents previously considered by the staff.
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I References 1.
-Generic Letter 84-04, D.
G.
Eisenhut to PWR Licensees, Construction Permit Holders and Appliennts for Construction I
Permits, dated February 1, 1984.
2.
Letter 3F0285-02, G. R. Westafer to H. R. Denton, " Request i
for Exemption From a Portion of 10CFR50, Appendix A, General Design Criteria 4" dated February 1, 1985.
4 3.
Babcock and Wilcox Owners Group Report, " Leak-Before-Break Evaluation of Margins Against Full Break for RCS Pri:aary Piping of B&W Designed NSS," B&W Topical Report BAW-1847, dated September 1984.
4.
ACRS Letter, J. J. Ray to W. J. Dircks, " Fracture Mechanics Approach to Pipe Failure," dated June 14, 1983.
5.
- Ravindra, M. K. et. al.,
Probability of Pipe Failure in the Reactor Coolant Loop of Babcock and Wilcox PWR Plants, Volume 2:
G u i ll o tine Break IndirectIV Induced by Earthauakes, Lawrence Livermore National Laboratory, UCRL-53644, NUREG/CR-4290, Vol. 2/ (1985).
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Appendix A The LBB evaluation for B&W nuclear plants was performed using piping loads at various locations on the hot and cold legs of the RCS piping.
These loadings were obtained from existing stress reports for each of the B&W Owners Group plants.
From these loads a single set of generic loads was evaluated.
Tables-4-1 and 4-6 of Reference 3 list the centro 11ing generic load sets.
Other load sets were evaluated; but these are limiting.
The next page is a listing of the Crystal River 3 piping loadings resulting from the reconfiguration of the RC pump supports vers s the generic load set evaluated in the LBB report.
The loadings are considered to be enveloped if at the limiting locations the new Crystal River 3 maximum is less and the new Crystal River 3 minimum is greater than those loading evaluated in the LBB report.
It can be seen that the moments resulting from the reconfiguration of the RC pump supports are within the envelope of moments that have been justified relative to Leak Before Break criteria.
Therefore, the moments resulting from the reconfiguration of the RC pump supports are also acceptable for the Leak Before Break criteria.
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CRYSTAL RIVER VERSUS GENERIC EVALUATION:
PIPE SIZE GENERIC MAXIMUM MOMENT F?C MAXIMUM MOMENT
( FT-KIPS )
( FT-KIPS )
28" 1.D.
STRAIGHT 3098.0 AT Jt.12 2004.4 AT Jt. 21.5 A ROW 2822.8 AT Jt. 24 1617.0 AT Jt. 27.5 36" 1.D.
STRAIGHT 2376.5 AT Jt. 9 2230.1 AT Jt. 9 ELS0W 2376.5 AT Jt. 9 2230.1 AT ' Jt. 9 1
i PIP.E SIZE GENERIC MINIMUM MOMENT-FPC MINIMUM MOMENT
( FT-KIPS )
( FT-KIPS )
28" 1.D.
STRAIGHT 560 AT Jt.12 1556.1 AT Jt. 21.5 ELBOW 1246 AT Jt. 24 1294.3 AT Jt. 27.5 36" 1.D.
STRAIGHT 1010 AT Jt. 5 1738.0 AT Jt. 9 ELBOW 1010 AT Jt. 9 1738.0 AT Jt. 9 6-
Appendix B The requirement to design the Crystal River
-3 Nuclear Power Plant (CR-3) for the effects of an instantaneous double-ended guillotine break (DEGB) of. the reactor coolant loop (RCL) piping has led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel.
NRC/ Lawrence Livermore National
' Laboratory (LLNL) sponsored a research program aimed at exploring whether the probability of DEGB in RCL Piping of nuclear power plants is acceptability small and if the requirements to design for DEGB effects (e.g., provision of pipe whip restraints) may be removed.
Reference 5 describes the study performed for ten Babcock & Wilcox (B&W) plants of which CR-3 was included.
Reference 5 estimates the probability of indirect DEGB,in RCL piping as a consequence of seismic-induced structural failures within the containment of B&W supplied pressurized water reactor nuclear power plants in the United States.
The Reference 5 study used actual component support loadings to determine the seismic capacity factors for the component supports.
Below is a listing of the new capacity factors (calculated in a manner similar to LLNL methodogy) f o r CR-3 and the factors calculated in the LLNL study for CR-3 and the minimum calculated for any B&W plant.
Capacity Pactors of Safetv for CR-3 New Qld Min. B&W Reactor Pressure Vessel 12.7 5.56 3.76 Steam Generator 34.2 118.8 2.57 Reactor Coolant Pump 4.6 N/A 9.91 Only the strength factor F was recalculated in the above. The s
probabity of indirect DEGB at CR-3 due to new capacity factors and methods of analysis-was not recalculated.
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However, from Figure 2-8 of Reference 5 it can be inferred that a
50% reduction in capacity would result in a probability reduction of approximately one order of magnitude.
Therefore, the new support-configuration at CR-3 has more than sufficient margin when compared to PWR nuclear: plants in-total.
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ATTAClllENT 1 Description of Computer Codes STALUM and RESPECT vso,sa.
_hant--
12129:1 DansLiktiga
_Classifissilta.
STALuM Ceneral static. thermal, and STALuM enalyses three-4tmomatonal, flaise segment h6W propaletary dyneale analyale for llevas systema comaletlag of unitosa os acaualfosa piplag elastic and gap stauctuses segmente, closed-leep assengemente. end suppostlag eleneste. STAL.um postosas both static and draanic essuctural analyene undesgolas naall llaeas, elas.
Elc Jeformatione. The static emelyene le based on the satste eleplacement method.
The static lead-inen age static mechanical forces. thessal, and/or support displacement leadings. The dynamic analyele le based on lumped mass and nosaan modo eatsactica techalques. The draanic laput leadlage can be a seaponse spectse et fosse llam hietesy.
The eeneatlal laput to t he psoggas comelete of t he Eliyelcal psopestles of the eyesen. the bounJasy condittene, anJ/es the loadthg lafossation; the eenential output cunelete of the soeulteht potat displacement. sotations, fescen. and monaute et both ende of each esgaeat and esteense et westeus lusatleaa la each segment.
DEsptCT To calculate amplifted seaponse DEspECT calculates ehe maalaus acceleration B&W psoprietasy spectse seaponse of a slagte degree of tseedom ISport oscillates subjected to an loput accolocatlea time bletory at the bene.
An SDor le descstbed by a seceae ordes diffesential aquellen which costelas a coefficleat descalbed as the Eigen value or natusal (Pbguency equered. When tale equation le selved fos.laput accenega.
tima Line bletery ulth vasytag Eigen walues, the teaultlag samlaua accelesatlon geoponse and natusal fsequency fosa an adpeleratloa seaponse spectra (ARS).
This progsam also calculates an ASE due to situctusal emplification between a knous pelat and an attachment potat.
Thle techalgue sequises a etsuctural sweponse spectra solution laccelegattene and laestle tosces) and the associated accelesatloa llae bletosy.
ATTACHMENT 2 1
OUTLINE OF METHODOLOGY USED FOR CR-3 PLANT SPECIFIC SEISMIC AND THERMAL ANALYSIS WITH ALTERNATE SUPPORT CONFIGURATION PERMITTED BY LBB CONCEPT Developed primary loop engineering mathematical model for seismic and thermal analysis. Revised an existing model to reflect CR-3 design for
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component support locations, stiffness (es), configuration, shield wall des!gn and thermal properties. Seismic and thermal analyses were performed by B&W with their computer program STALUM; a fully certified, finite element piping analysis program designed for seismic applications via the response spectrum technique.
Performed deadweight, thermal and Operating Basis Earthquake (OBE) seismic analysis for the revised support configuration, using increased damping.
The combination of structural vibration modes was by the closely spaced modes technique as described in NRC Regulatory Guide 1.92.
Amplified response spectra (for attachments to the RCS piping and pumps) were generated by B&W with their computer program RESPECT, which is a fully certified analysis program which calculates acceleration response spectra due to structural vibration amplification between a known point and an attached point.
Revised nozzle spectra were compared to existing design spectra for spray, HPI, drain and letdown lines.
Assured that all new piping loads were bounded by the loads utilized in the-B&W Owners Group LBB Evaluation, BAW 1847 Damping considerations included:
Original damping for components, shield walls, and vital piping systems is 0.5% of critical damping (as listed in FSAR and Reg. Guide 1.61); RCS analysis for optimized. support configuration utilizes variable damping from 2-5% of critical damping for piping per ASME Code Case N-411 (a value of 2% was utilized for components and 4%
for shield walls using Regulatory Guide 1.61).
Combination of directional earthquake response was per the absolute sum method-maximum of x + y or y + z earthquakes (FSAR 5.4.5.2); pump loadings were combined by SRSS of the x, y, z direction earthquake loads.
Design codes utilized:
Power Piping; USAS B31.1.0 (1967) with erratum dated March 1969.
This is the original construction code for the existing RCS pump supports.
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Nuclear Power Piping; USAS B31.7 (February 1968 with erratum dated
. June 1968). This is the original construction code for the existing RCS
. piping.
ASME Section III, Class A (1965 Edition with Addenda through Summer 1967). This is the original construction code for the existing RCS pressure vessels.
ASME Nuclear Code Case N-411 (September 17, 1984).
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L ATTACHMENT 3 DESCRIPTION OF SNUBBER ASSEMBLIES, OVERALL SNUBBER CONFIGURATION, AND PUMP SUPPORT CONFIGURATION PLANNED FOR CR-3 ALTERNATE SUPPORT CONFIGURATION PERMITTED BY LBB CONCEPT t
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5?CEEER ASED'BLY DESCEITTION i
RC PUMP 3Al
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Assemoly Estimated Length Wei gnt Tac No.
Tyoe (inches)
(counds)
.RCHS -1A 2000K 64.58 7320 RCHS-2A 1200K 77.68 4454 RCHS-3A 2000K 72,82 7468 RCHS 4A 1200K 89.38 4620 i
RCHS-5A 1000K 151.38 4455 RCHS-6A 1000K 116.38 4016 1
RCHS-7A 1200K 123.31 5104 RCHS-BA 1000K 54.50 3242
^
RC PUMD 3A2 1
Assembly Estimated Lengtn-Wei gnt Tae NC.
Tvoe (inches)
(Dounes)
RCHS-1B 1600K 61.00 5471 RCHS-2B 1600K 63.12 5505 RCHS-3B 2000K 68.78 7396 RCHS 4B 1000K 216.00 5815 RCHS-5B 2000K 167.34 9171 t
RCHS-6B 1000K 185.50 4882 RCHS-7B 1200K 260.75 8588 1
RCHS-8B 1600K 211.12 8441 i
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S!UEBER ASSEMELY DESCkIPTION RC PUMD 381 Ass embly Estimated Lencth Wei9nt Ta; No.
TyDe (inCbes)
(DounOSI RCHS-1C 1200K 57.75 4140 RCHS-2C 1000K 58.18 3288 RCHS-3C 1000K 134.25 4241 RCHS-4C 1600K 155'75 6905 RCHS-SC 1200K 98.04 4744 RCHS-60 2000K 99.50 7949 RCHS-7C 1000K 117.50 4031 RCHS-8C 1600K 102.43 6138 RC PUMP 382-Assembly Estimated Length Wei gn Tac Nc.
TyDe finenesI (counes)
RCHS-10 2000K 79.86 7595 1
RCHS-20 1600K 117.18 6375 RCHS-3D 1000K 126.75 4147 RCHS-4D 1600K 65.68 5547 RCHS-50 1600K 103.75 6159 RCHS-60 2000K 95.78 7882 i
RCHS-70 1000K 160.00 4563 RCHS-BD 1000K 101.00 3824 I
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OVERALL SNUBBER CONFIGURATION ST*WI GENERATOR
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3B RCHS-4C
-I RCHS-3C RCHS-6D RCHS-5D RCHS-5C '
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RCHS-6C ACES-4D RCP 331 RCP 33Z N~HS-7C 3 N T
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m x ROS-SC RCES-7,
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RCHS-8D RCES-IC d RCHS-2C y
RCHS-lD RCHS-2A R HS-3A VHS-4A RCHS-2B RCHS-43 RCHS 1A RCHS-5A RCHS-1B RCHS-3B RCP JAl ECP 3A2 gg.dS-55
..r*
RCHS-6A RCHS-48 RCES-6A r.5; L-RCliS-8B r RCHS-7A STEA:!
RCHS-7B CE.;ERATOR 3A 3 -3
PUP CUFFCET N
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I R C HS-E A.
PUMP NO bht TA G N * -
RCP.-1A acus-sA@
Rest-5 RCH3-4A R.H3-1 SITUEEER ELIMIllATED
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C H S - (. D 3 2 - RIN G 4
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