ML20198G005
| ML20198G005 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 11/05/1975 |
| From: | Cox T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1101 NUDOCS 8605290183 | |
| Download: ML20198G005 (28) | |
Text
{{#Wiki_filter:O s NOV. 5 1975 DOCU:7 f0S.: 50-460 AND 50-513 APPLICANT : WASHING' ION PUBLIC POWER SUPPLY SYSTBt FACILI'IY
- WNP-1,4 SIM!ARY OF f.EETING IIELD ON OCI0BER 24,1975 A meeting was held in Bethesda, Maryland to discuss applicant's method of calculating reactor vessel internal pressure differentials resulting from the IDCA, and the method of applying these loads to determine reactor pressure vessel (RPV) support systems. 'Ihe meeting was essentially a presentation by BSI. describing their reactor vessel support analysis and use of the CRAFT-2. code to predict RPV intemal pressures as a function of time after a pipe failure, with subsequent questions by NRC staff. The capability und validity of CRAIT-2 relative to WIW.t: was discussed. Infomation presented by B4W will be presented fomally in report fom at a future date which was not specified at the meeting. 'Ihe staff indicated that, in view of the WNP-1,4 safety hearing scheduled to start on 11/11/75,any significant change in the already filed staff testimony on the RPV support issue was impmbable. B4W agreed to pmvide infomation within a few days concerning (1) NSS stiffness relative to vessel support stiffness, (2) maxin.tn displacement of the vessel fmm the nominal centerline, and (3) justification of the selection of 10 milliseconds as the design basis break opening time.
Applicant's representative agreed to provide additional data within a few days concerning the design criteria for guard pipes. Criteria specified will include stress levels, materials, and frequency of inspection. Visual aids used by B5W in the meeting are reproduced here as Attachment 1. is an attendance list. DISCUSSION l
- 0. Ilooker (BGV) gave introductory remarks and described the B6W organization for accomplishing the reactor vessel internals and NSS structural design.
Ile also reviewed the mechanical design layout and features of B6W's NSS. L. Smalec then described the calculation of IDCA loads on 205 F.A. reactor internals using CRAFT-2. lie noted that a break opening time of 10 milliseconds was assuned, which they felt pmperly represented the system. l V y g5290183751105 ADOCK 05000460 A PDR I l
Meeting Steraary for WNP-1,4 on October 24, 1975, In support of the selected 10 millisecond break opening time, BGi cited experimental work done at BNL on rupture and crack pmpagation times, and agreed to pmvide a reference on this work. Le relative capability of CRAFT-2 and MWf, an older code, was discussed at length. J. Cudlin (B5W) discussed comparisons of CRAFT-2 calculations with 1*W! data and presented some copies of these comparisons to NRC staff members. He point offered by B5W was that the CRAFT-2 calculations appeared to predict IDFT semi-scale experimental data from selected tests at least as well as the MMI model. E. Throm of the staff stated that at least two features of CRAFT-2 would require careful evaluation; (1) the development of the " control voltzae" network, and (2) the iterative adjustment of inertial (L/A) factors to achieve asstaned conditions regarding wave propagation in the vessel. P. Norian pointed out that the staff was certainly willing to consider in detail the CRAFT-2 capability to calculate the vessel internal pressure distribution, and would do so when material was presented for review. S. Bmwn described how B4W uses the force components calculated with CRAFT-2 to analyze the dynamic response of the NSS relative to the system supports. Computations are made using ST3DS and Ltr.S which are linear elastic computer models developed by Franklin Institute. BGW stated that most of the load in this system is reacted by the vessel supports because the support stiffness is significantly greater than that of the NSS piping and other system supports. V. Noonan of the staff requested additional infomation on guard pipe design criteria for WNP-1,4. A. !!osler agreed to supply this infomation in a letter to the staff prior to the WNP-1,4 hearing. e ssiaal sic m a g m 2.;., n T. Cox, Project Manager Light Water Reactors Branch 2-3 Division of Reactor Licensing Attachments: 1. BG1 Visual Aids 2. Attendance List '-1/ R. A. Hedrick, J. J. Cudlin, and R. C. Folt, " CRAFT 2 - Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss-of-Coolant," BAW-10092, Rev. 2, Babcock 6 Wilcox, April 1975. 2] Stanislav Fabic, " Computer Program NW1 for Calculation of Pressure, Velocity, and Force Transients in Liquid Filled Piping Networks," Kaiser Engineers Report No. 67-49-R (November 1967). x7886/13WR2. :i. orric= *- ( TCox:myyt., .v. l 11/f/75 Foran AEC.)ls (Rev. 9 5)) AICM 0240 W v. e, sovsamusat Paentine orreces es?..eae.see
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~ LOCA LOAD CALCULATION - 205 REACTOR IflTERNALS I. CRAFT 2 MATHEMATICAL MODEL 11. MODEL INPUT 111. BLOWDntIN ANALYSIS CRITERIA IV. RESULTS e S e
1, CPAFT 2 MATHEMATICAL MnDEL A. TRAtlSIEllT SOLUTin!I 0F MASS, ENERGY, a MnMEflTUM EntlATIONS B. PRIMARY DF.W LOCA CODE 1, MEETS FAC REQUIREMENTS FOR ECC-RFLATED MATTERS, 2, MASS & ENERGY RELEASE, 3. SUBC00 LED BLOWDOWN LOADS, C. SPECIALIZED FEATURES FOR LOADS CALCULATIONS 1, 2 0 CAPABILITY 2. FORCE ON CORE 3. PIPING REACTION FORCES 4. FLUID JET INTENSITY 5. VENT VALVES
11. f10 DEL INPilT A. " MULTI-N0DE" REPRESENTATIn.'l 0F REACTOR INTERNALS. 1. SilFFICIENT DETAIL TO DEFlf!E REQUIRED LOADINGS ON INTERNALS. 2. INPUTS ADJUSTED TO PROMOTE PPnPER WAVE "TRACffl!!G" B. PARAMETRIC STUDIES TO DEMONSTRATE CONSERVATISM IN RESULTS. 1. DISCHARGE MODELS 2. SPATIAL DETAIL 3. BREAK SIZE & LOCATION 4. INERTIAL FACTORS, L/A k e s
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- III, ANALYSIS CRITERIA
-A. 100% POWER OPERATION B. FULL-AREA BREAKS 1, 2-AREA CIRCUMFERENTIAL BREAKS AT VESSEL HOT F, COLD LEGS, l l 2, 1-AREA LONGITUDlHAL BREAK AT VESSEL COLD LEG. l C. MOODY DISCHARGE MODEL l 1. CD 1,21 = l D. BREAK OPENING TI'E OF 0,01 SECONDS E. RUN DURATION OF 0,5 SECONDS i i l l l l l t
6 IV. RESULTS A. PEAK 6P ACROSS UPPER CORE SUPPORT CYLINDER, COLD LEG BREAK 1. AT N0ZZLES 540 PSI 2. AT BOTTOM 0F ANNULUS 280 PSI B. PEAK AXIAL FORCE ON CORE COLD LEG BREAK 9 X 105 Lpp 1 5 2. HOT LEG BREAK 4 X 10 LEF C. PEAK 6P ACROSS PLENUM CYLINDER 1. HOT LEG BREAK 120 PSI D. MAXIMUM JET INTENSITY AT BREAK 1. COLD LEG SPLIT 1,5 X 106 LBF E. MAXIMUM INTEGRATED LATERAL LOADS & M0MEllTS Off REACTOR IflTER!!ALS 1. COLD LEG GUILLOTINE A) Fx = 9,5 x 106 LBF 6 B) Fz = 7,9 x 10 LBF d) M = 57 x 107 IN-LBF x 64 x 107 o) M IN-LBF = z l l L
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SUMMARY
4 1. LOCA LOADS CALCULATED WITH CRAFT ARE REPRESENTATIVE. II, LOCA LOADS USED FOR 205 PLANT INTERNALS DESIG AND VESSEL SUPPORTS & RESTRAINTS DESIGN ARE CONSERVATIVE, G
A ATTAONENT 2 MEETING WITH B4W October 24, 1975 ATTENDANCE LIST l hRC T. Cox, RL V. Noonan, TR P. Chen, TR R. Bosnak, TR R. Major, ACRS staff E. 'Ihrom, TR P. Norian, TR N. Anderson, RL J. Slider, RL K. Desai, TR P. Baranowsky, 1R B6W / 4 R. Borsun E. Hooker S. Brown B. Dunn L. Smalec J. Cudlin H. Baker J. Happell J. Agar WPPSS A. Ibsler i __}}