ML19273A992
| ML19273A992 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 10/10/1978 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML19273A984 | List: |
| References | |
| TAC-08843, TAC-8843, NUDOCS 7902010178 | |
| Download: ML19273A992 (9) | |
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- w. r. stew ^a T. oin e CTOn POWER PRODUCTION January 26, 1979 Mr. Victor Stello, Director Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C. 20555
References:
(a) License No. DPR-72 (Docket No. 50-302)
(b) Letter, Victor Stello, Jr., NRC, to all PWR Licensees, January 25, 1978.
(c) Letter Florida Power Corp., to Victor Stello May 4, 1978, Asymetric LOCA Loads, Phase 1.
Dear Mr. Stello:
Phase 1 of our asymmetric LOCA loads program that was outlined in our previous letter to you (Reference (c)) has been completed on Crystal River Unit 3.
That phase of our work has identified the need to proceed with less conservative approaches and, therefore, this letter submits our plan for proceeding with Phase 2.
The attached report entitled "B&W 177 FA Owners Group Asymmetric LOCA Loads Evaluation Program, Phase 2" identifies the content and schedule.
Please note that we are still engaged in the requested evaluation as a participant of the B&W 177 Fuel Assembly Owners Group and, where permissi-ble, plan to take advantage of generic analyses. When doing so, we will provide justification for the generic grouping.
At this time, the attached program addresses LOCA loads only.
How-ever, based on direction received f rom your Mr. Steve Hosford in a Decem-ber 12, 1978 meeting, we are now evaluating inclusion of seismic into the program.
Until we have had time to evaluate the available seismic loads and develop a satisfactory seismic-LOCA load combination approach, we can-not define the seismic portion of the program.
These investigations should be completed by mid-1979 and when a program is determined, additional information will be submitted.
If you have any further questions regarding this matter, please con-tact this office.
Very truly yours, FLORIDA POWER CORPORATION i
W. suee W.
P. Stewart ECSekcRll(D4) 7902010/78
/
File 3-0-3-a-3 Genera:l Office 32o1 Thirty-fourtn street soutn. P O. Box 14042, st. Petersburg, Ficnca 33733 813-866 5151
/
STATE OF FLORIDA J
COUNTY OF PINELLAS W.P.
Stewart states that he is the Director, Power Production, of Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information and belief.
6
" iLO.CM W. P.' Stewart Subscribed and sworn to before me, a Notary Public in and for the S ta te and County above named, this 26th day of January, 1979.
/fd NotaryPubly Notary Public, State of Florida at Large, My Commission Expires:
July 25, i980
B&W 177 FA OWNER'S GROUP ASYMMETRIC LOCA LOADS EVALUATIONS PROGRAM, PHASE 2 Arkansas Power & Light -- ANO-1 Duke Power Company -- Oconee 1, 2, 3 Florida Power Corporation - Crystal River 3 Metropolitan Edison Company -- Three Mile Isl-
.i,2 Sacramento Municipal Utility District - Rant _ 3eco Toledo Edison Company -- Davis-Besse 1 November 10, 1978 Y
CCNTENTS 1.0 Introduction 2.0 Evaluation Bases 3.0 Work Plan (Phase 2) 4.0 Computer Codes 5.0 Applicable B&W Topical Reports 6.0 Phase 2 Schedule
- ii -
1.0 INTRODUCTION
This report summarizes Phase 2 of the detailed plan prepared by the B&W 177 FA Owner's Group in response to the NRC Division of Operating Reactors letter dated January 25, 1978.
Phase 1 of the program was catlined in the reports, "B&W 177 FA Owner's Group Asyninetric LOCA Loads Program," datc April 10, 1978, and " Toledo Edison Company Asymmetric LOCA Loads Evaluation Program for Davis-Besse 1 Plant," dated Apri' 21, 1978.
Whereas, Phase 1 performed preliminary investigations using estimates and generalities to bracket the evaluation requirements, Phase 2 will go into greater details to determine more specific load; ?nd component /
structural evaluations.
The need to proceed with Phase 3 is still unresolved at this time, and will be addressed at a later date after discussions with the NRC.
2.0 EVALUATION BASES 2.1 The components to be evaluated during Phase 2 for the LOCA breaks analyzed include:
a.
Fuel Assemblies, Including Grid Structures c.
Control Rod Drives d.
ECCS Piping that is Attached to the Primary Coolant Piping e.
Primary Coolant Piping in Close Proximity to the Reactor Vessel f.
Reactor Vessel Supports g.
Reactor Internals h.
Biological Shield Wall and Neutron Shield Tank (where applicable) i.
Core Flooding Piping J.
Related Building Structures 2.2 LOCA analyses will be performed for breaks rendering the worst loadings for the Reactor Vessel supports and Reactor Internals.
For these breaks, all components listed in Paragraph 2.1 will be evaluated to assure (1) maintaining core coolable geometry and (2) mitigating the conse-quences of an accident.
2.3 Jet impingement effects will be evaluated for breaks analyzed.
This evaluation was not explicitly stated in the NRC letter, but was identified as a requirement in a previous meeting (March 31,1978) with the NRC.
2.4 As appropriate, the evaluation will consider:
a.
Limited displacement break areas where applicable b.
Use of actual time-dependent forcing function c.
Reactor support stiffness d.
Break opening times e.
Break location utilizing stress criteria 2.5 Where justifiable, a generic review of the B&W Owner's Group plants will be used.
3.0 WORK PLAN (PH, E 2) 3.1 The objective of this task is to define the resultant forces and moments which would act externally on the reactor vessel in the event of a reactor coolant system (RCS) pipe rupture inside of the reactor subcompartment.
The CRAFT 2 computer code will be used to calculate the transient, asymmetric pressure distributions inside the subcompartment for a spectrum of break cases.
Analysis guidelines established in "tandard Review Plant (SRP) 6.2.1.2 for subcompartment pressurization.alculaticns and in SRP 6.2.l.3 for mass and energy release calculacions wi'. be folicwed.
Three different reactor cavity designs will be evaluated. Design selection will be based upon the following considerations:
1.
Cavity volume between reactor vessel and primary shield wall.
2.
Insulation design.
3.
Vent areas of primary piping penetrations.
4 Shield plugs or blow-out devices.
5.
Flow obetructinns.
The 177 plants (0wner's Group) will be categorized by B&W into three groups, each group being represented by one of the cavity designs.
3.2 Develop a mass and energy release calculation model for a generic 177 lowered-loop RCS operating at a power level of 1.02 X 2772 MWt.
Generate mass and energy release data for the initial two (2) seconds of blowdown for hot and cold leg breaks of the following sizes:
a.
2.0A b.
0.6A c.
0.3A d.
To be determine' e.
To be determi ac 3.3 Develop a mass and energy release calculation model for the Davis-Besse 1 plant at a power level of 1.02 X 2772 MWt.
Generate mass and energy release data for the initial two (2) seconds of blowdown for hot and cold leg breaks of the follonicg sizes:
a.
.5 ft 2
b.
1.0 ft c.
.5A d.
1.0A e.
2.0A 3.4 The core flood line will be treated as a cold leg break of appropriate size.
3.5 For evaluation of the mass and energy data generated in Paragraph 3.2 develop three (3) reactor cavity models.
Calculate the reactor cavity pressurization rates for the spectrum of hot and cold leg breaks identified.
Calculate the time-histories of the lateral and vertical forces and moments acting on the reactor vessel out to a blowdown time which is sufficient to define the peak magnitudes of these forces and moments.
3.6 For evaluation of the mass and energy data generated in Paragraph 3.3, develop a reactor cavity model.
Caica.ute the reactor cavity pres-surization rates for the spectrum of hot and cold leg breaks identified.
2.culate the time-histories of the lateral and vertical forces and moments acting on the reactor vessel out to a blowdown time which is sufficient to define the peak magnitudes of the forces and moments.
3.7 Calculate the loss-of-coolant-accident (LOCA) loadings on the reactor internals structures of 177 FA plants.
The calculations will be performed using procedures documented in BAW Topical Report 10132.
A spectrum of break sizes will be considered in the hot and cold leg piping of the reactor coolant system (RCS).
Break locations insid; the reactor cavity and outside the primary shield wall in the steam generator compartment will be considered.
3.7.1 A 3eneric model will be developed for calculating LOCA loads on 177 FA lowered-loop plants.
The modeling criteria established in Topical Report BAW-10132 will be used in the development.
Initial reactor fluid conditions which encompass all 177 plants for purposes of LOCA load calculations will be specified in the model.
Design LOCA load calculations will be performed for eight (8) break cases inside the reactor cavity and for two (2) break cases in the steam generator compartment, wi'a a contingency for analyzing up to four (4) additional breau anywhere in the RCS.
The break sizes and the corresponding break opening times will be selected on the basis of the results of the Phase 1 program.
The analyses will be conducted out to 0.3 sec of the blowdown.
The following parameters will be recorded as a function of time for each design case calculation:
a.
Control volume pressures.
b.
Major component aP's.
c.
Vertical force on the core.
d.
Vessel head AP.
e.
Integrated lateral forces on pressure vessel and core support cylinder.
f.
Integrated lateral load on plenum cyclinder.
g.
Mass and energy release to containment.
h.
Jet intensity at breax plane.
3.7.2 The reactor internals LOCA load model used for analysis on Davis-Besse 2 and 3 will be converted over for Davis-Besse 1.
This conversion will include a change to a 15 X 15 fuel assembly model, a change to Davis-Besse i specific fluid conditions, and changes brought acout by considerations of BAW-10132 methods.
Reactor internals LOCA loads for the initial 0.3 sec of blow-down will be generated for up to ten (10) break cases.
The break locations, sizes, and opening times will be determined on the basis of the results of the Phase 1 program.
The following parameters will be recorded as a function of time for each design case calculation:
a.
Control volume pressures.
b.
Major component 2P's.
c.
Vertical force on the core.
d.
Vessel head aP.
e.
Integrated lateral forces on pressure vessal and core support cylinder, f.
Integrated lateral load on plenum cylinder.
g.
Mass and energy release to containment.
h.
Jet intensity at break plane.
3.8 The fuel assembly model parameters (mass, spring rate, and damping halves) will be calculated for use in the fuel assembly model.
These parameters will be envelope values for the Mark B fuel assembly and will be used as inputs in the development of the Core Bounce Model.
3.9 An existing core bounce model will be modified to reflect the 177 Mark B fuel assembly.
The vertical cavity pressure will be applied to the fuel assembly model with a spectra of breaks previously identified, and the resultant load impact at the upper and lower grids will be calculated.
These loads will be presented in the form of time-histories and will be used as input into the Reactor Vessel Isolated Model.
The core bounce model is non linear in nature due to the springs and gaps.
Thus, amplified forces supplied to the linear model include the dynamic impact of the fuel assemblies in the vertical direction.
3.10 The bending and extensional stiffnesses of the reactor vessel inter-nals will be calculated for input into the isolated dynamic model.
For the non-redundant structures such as the core barrel, thermal shield, and core support shield the stiffnesses will be calculated using classical methods.
For the pl'em assembly, the apparent differences of two redundant load pat., will be calculated:
the plenum cylinder path and the column weldment path.
This is accom-plished with a three-dimensional finite element model of the plenum assembly.
The apparent stiffness o' the column weldments will be calculated from their average disp N;ement while the plenum cylinder average displacement at its base will be used to calculate its stiff-ness.
This method accounts for the effects of the plenum cover and upper grid which will not be included in the isolated model.
3.11 Existing 177 fuel assembly reactor vessel internals model (TEC0, Davis-Besse 2 and 3) will be modified to reflect the RV skirt sup-port or the TECO Davis-Besse 1 supports.
This model will include the reactor vessel internals as beam elements obtained earlier, and the fuel assembly model also obtained earlier.
The model will include service support structure, CRDM, cold leg piping, and hot leg piping to the extent feasible.
The TECO Davis-Besse 1 model will reflect the internals design for Davis-Eesse 1 instead of Davis-Besse 2.
3.12 Dynamic LOCA analysis (linear elastic) will be performed on the model generated above.
This analysis will include the following as input forcing functions:
a.
Horizontal delta pressures integrated over the wetted surfaces of the internals and the inside of the vessel shell to describe the horizontal forcing functions on the vessel and internals.
b.
Vertical delta pressures integrated over the RV heads ta describe the vertical force on the vessel.
c.
Vertical core bounce forcing functions are applied at the plenum cover ledge and include all vertical delta pressure integrations across the internals and core, and all the vertical dynamics of the internals.
d.
Asymmetric cavity pressures are integrated ovc,' the outside sur-face of the vessel and applied to the vessel.
fiOTE:
1)
The " thrust force" is included in (a) above the area of the broken pipe is excluded from the integration and the area of the urbroken pipe is included.
- 2) This task will utilize the results of a hydrodynamic mass coupling developed separately.
3.13 flon-linear pipe whip analysis will be constructed representing the non-linear material properties and existing gaps in each of the plants.
These models will reflect the as-designed status and the as-built gaps that can be obtained from Phase 1. Calculations will be performed to determine the break discharge area for each of the breaks identified in the spectrum of the breaks outlined earlier.
An iterative approach will be required to obtain the final non-linear pipe break area.
3.14 To obtain the reaction forces <
the fluid on the primary coolant boundary, the actual break art. must be represented.
The reaction forces are a function of area changes and direction changes of the fluid along its flow path due to a leak path from the system boundary.
The model for ' ew volumes will require the definition of the break area (leak a,ea) for a time-history calculation of the forcing functions.
The solution for the forcing functions will require iteration to obtain a solution based on consistent conditions (break area and dynamic response).
The results will be obtained in the form of area versus time.
3.15 Fuel assembly deformation limits will be established based upon allowable grid deforma' ions as determined by ECCS requirements.
These established requ1 cements will be confirmed by analysis to assure that peak cladding temperatures do not exceed those allow:d by 10CFR50.46.
Additional analyses may 12 necessary if the actual deformations exceed the established values.
3.16 A core evaluation model will be developed to simulate the fuel assembly interaction during dynamic excitation.
The model will consider gaps that exist between inner assemblies and between outer assemblies and the baffle wall.
Available experimental test information such as spacer grid dynamic properties, damping, and fuel assembly frequencies will be used as input to the core evaluation model.
This model in conjunction with the coolable geometry criteria will be used in evaluating the fuels coolable geometry.
3.17 Using the loading generated (as outlined earlier), the components identified in Paragraph 2.1 will be evaluated.
In addition, integrity of the cavity walls will be evaluated when subjected to the effects of asymmetric pressures.
Due to the very limited time available for Component Evaluations, complete structural analyses for these components will not be per-formed and stress reports will not be prepared.
However, the components will be evaluated using applied loads and the resultant stresses compared to material capabilities in critical areas of the structures.
Based upon the results, conclusions will be drawn with respect to the structural integrity of the affected components, structures, and the coolable geometry of the Fuels and Core.
4.0 COMPUTER CODES In the performance of the analyses, several different computer codes will be used. The following list identifies the major codes to be employed:
a.
ANSYS b.
ADINA c.
ST3DS d.
LUMS e.
STARS f.
CRAFT 2 g.
SUPERPIPE h.
GDSGAP 1.
PWHIP 5.0 APPLICABLE B&W TOPICAL REPORTS Techniques described in topical reports submitted to the NRC by the B&W Company will be used in the evaluation.
These topical reports are:
a.
BAW-10131 - Reactor Coolant System Structural Analysis b.
BAW-10127 -- LOCA Pipe Break Criteria for the Design of Babcock &
Wilcox Nuclear Steam Systems c.
BAW-10132 -- Analytical Methods Description ~ Reactor Coolant System Hydrodynamic Loadings During a Loss-of-Coolant Accident d.
BAW-10133 - Mark C Fuel Assembly - LOCA-Seismic Analyses e.
BAW-10060 -- Reactor Internals Design / Analysis for Normal, Upset
.ind Faulted Conditions li. C PHASE _2 SCHEDULE No detailed schedules are included with this report because of the complex interactions required for the analyses described herein.
However, at this time, it is projected the conclusion reports will be available by March, 1980.