ML19224A826
| ML19224A826 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Midland, Crane  |
| Issue date: | 04/10/1979 |
| From: | Arkansas Power & Light Co, Consumers Power Co, Duke Power Co, Florida Power Corp, Metropolitan Edison Co, Sacramento Municipal Utility District (SMUD), Toledo Edison Co |
| To: | |
| Shared Package | |
| ML19224A825 | List: |
| References | |
| NUDOCS 7906010218 | |
| Download: ML19224A826 (9) | |
Text
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B&W 177 FA OWNERS GROUP ASYMMETRIC LOCA LOADS EVALUATIONS PROGRAM, PHASE 2 Arkansas Power & Li ht - ANO-1 9
Duke Power Company - Oconee 1, 2, 3 Florida Power Corporation - Crystal River 3 Metropolitan Edison Company - Three Mile Island 1, 2 Sacramento Municipal Utility District - Rancho Seco Toledo Edison Company - Davis-Besse 1 Consumers Power Company - Midland I and 2 7 9 0 6 010 21E" 16B 303 April 10, 1979
a CONTENTS 1.0 Introduction 2.0 Evaluation Bases 3.0 Work Plan (Phase 2) 4.0 Ccmputer Codes 5.0 Applicable B&W Topical Reports 6.0 Phase 2 Schedule 168 304
- ii -
1.0 INTRODUCTION
This report summarizes Phase 2 of the detailed plan prepared by the B&W 177 FA Owners Group in response to the NRC Division of Operating Reactors letter dated January 25, 1978.
Whereas, Phase 1 performed preliminary investigations using estimates and generalities to uracket the evaluation requirements, Phase 2 will go into greater details to determine more specific loads and 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 Drive Mechanisms d.
ECCS Fiping that is Attached to the Reactor Coolant Piping Reactor Coolant Piping in Close Proximity to the Reactor Vessel e.
f.
Reactor Vessel Supports g.
Reactcr Internals h.
Reactor Cavity Wall i.
Core Flood Piping j.
Building Structures Associated With the Above Components 2.2 LOCA analyses will be performed for breaks rendering the worst load-ings for the Reactor Vessel Supports and Reactor Internals.
For these breaks, all components listed in Paragraph 2.1 will be eva-luated to assure that core coolable geometry is maintained mitigating the consequences of a loss of coolant accident.
2.3 Jet impingement effects will be evaluated for the breaks analyze.d.
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-deperdent forcing function Reactor support stiffness d.
Break opening times e.
Break location utilizing stress criteria 2.5 Wnere possible, generic evaluations of the B&W Owners Group plant components will be performed.
168 305 3.0 WORK PLAN (PHASE 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 s7ectrum of break cases. Analysis guidelines established in Standard Review Plant (SRP) 6.2.1.2 for subcompartment pressurization calculations and in SRP 6.2.1.3 for mass and energy release calculations wi?l be followed.
The reactor ca.
'y designs will be evaluated based upon the following conswrations:
1.
Cavity vol% between reu. tor vessel and primary shield wall.
2.
Insulation design.
3.
Vent areas of primary piping penetrations.
4.
Shield plugs or blow-out devices.
5.
Flow obstructions.
3.2 Develop a mass and energy release calculation model for a generic 177 lowered-loop RCS operating at a power level of 1.C2 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.
1.5A c.
1.0A d.
.6A e.
.3A 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 following sizes:
a.
.5 ft2 b.
1.0 ft2 c.
.5A d.
1.0A e.
2.0A 3.4 The core flood line will be treatt.d as a cold leg break of appro-priate size.
3.5 For evaluation of the mass and energy data generated in Paragraph 3.2 develop five (5) 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 cf these forces 168 306
and moments. The development of five (5) reactor cavity models will enable the determination of forces and moments on a plant specific basis.
3.6 For evaluation of the mass and energy data generated in Paragraph 3.3, develop a reactor cavity model.
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 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 inside the reactor cavity and outside the primary shield wall in the steam generator compartment will be considered.
3.7.1 A generic model will be developed for calculating LOCA loads on 177 FA lowered-loop plants.
The 'odeling criteria established in Topical Report BAW.
32 will be used in the development.
Initial reactor fluid conditions which encom-pass all 177 plants for purposes of LOCA load calculations will be specified in the F.odel.
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, with a contingency for analyzing up to four (4) additional breaks 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 APs.
c.
Vertical force on the core.
d.
Vessel head AP.
e.
Integrated lateral forces on pressure vessel and core support cylinder.
f.
Int? grated lateral load on plenum cylinder.
g.
Mass and energy release to containment.
h.
Jet iatensity at break plane.
16B 307 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 identifind, and the resul# load impact at the upper and lower grids will be~ calculated.
Enese loads will be presented in the form of time-histories and will be used as input into che 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 tha dynamic impact of the fuel assemblies in the vertical di6 ection.
3.10 The bending and extensional s+.iffnesses of the reacter vessel internals 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 tha plenum assembly, the apparent differences of two redundant load paths will be calculated:
the plenum cylinder path and the column weldment path.
This is accomplished with a three-dimensional finite elemen* model of the plenum assembly.
The apparent stiffness of the column weldments will be calculated from their average displacement while the plenum cylinder average displacement at its base will be used to calculate its stiffness.
This method accounts for the effects of the plenum cover and upper grid which will not be included in the isolated model.
- 3. P Existing 177 fuel assembly reactor vessel internals medel (TEC0, Davis-Besse 2 and 3) will be modifieJ 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 thc fuel assembly model also obtained earlier.
The model will include service support structure, CRCM, cold leg piping, and hot leg piping to the extent feasible.
The TECO Davis-Bene 1 model will reflect the internals design for Davis-Sesse 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:
Horizontal delta pressures integrated over the wetted surfaces a.
of the internals and che inside of "s vessel shell to describe the horizcntal forcino functions on the vessel and internals.
b.
Vertical delta pressures integrated over the RV heads to describe the vertical force on the vessel.
Vertical core bounce forcing functions are applied at the c.
plenum cover ledge and include all vertical delta pressure integrations across the internals and core, and all the gg }h vertical dynamics of tha internals.
d.
Asymmetric cavity pressures are integrated over the outside surface of the vessel and applied to the vessel.
NOTE:
- 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 unbroken pipe is included.
- 2) This task will utilize the results of a hydrodynamic mass coupling developed separately.
3.13 Non-linear pipe whip analysis model 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 detemine the break discharge area for each of the breaks identified in tne spectrum of the breaks outlined earlier.
An iterative approach will be reauf red to obtain the final non-linear pipe break area.
3.14 To obtain the reaction forces of the fluid in the primary coolant boundary, the actual break area must be re) ented.
The reaction forces are a function of area changes and airection changes of the fluid along its flow path due to a 'eak path fro 1 the system boundary. The model for flow valumes will require 9 dc'inition of the break area (leak area) for a time-history caiculation J the forcing functions.
The solution for the forcing fr actions will require iteration to obtain a solution based on '.unsi- :ent conditions (break area and dynamic response).
The results will be obtained in the form of area versus time.
3.15 Fuel assembly defomation litaits will be established based upon allowable grid defomations as detemined by ECCS requirements.
These established requirc.1ents will be confimed by analysis to assure that peak cladding temper:turos do not exceed those allowed by 10CFR50.46.
Additional analyses may be 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, N
and fuel assembly frequencies will be used as input to the core
[ff evaluation model.
fj m
This model in conjunction aitn the coolable geometry criteria will be used in evaluating the fuels coolable geometry.
} Q jh.
~~~
l 3.17 Using the loading generated (as outlined earlier), the components identified in Paragraph 2.1 will be evaluated.
In addition, integ-ity of the cavity walls will be evaluated when subjected to the e#fects of asymmetric pressures, tue 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.
"; wever, the components will be evaluated using applied Icads and the resultant stresses ccr cared 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 i.
PWHIP j.
STALUT1 k.
FESAP 1
HYDROE n
NASTRAN 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 evaluatior.
These topical reports are:
BAW-10131 - Reactor Ccolant System Structural Analysis a.
b.
BAW-10127 - LOCA Pipe Break Criteria for the Design of Babcock &
Wilcox Nuclear Steam Systems BAW-10132 -- Analytical Methods Description - Reactor Coolant System c.
Hydrodynantic Loadings During a Loss-of-Coolant Accident d.
FAW-10133 - Mark C Fuel Assembly - LCCA-Seismic Analyses BAW-iOC60 - Reactor Internals Design / Analysis for Normal. Upset e.
and Faulted Conditions f.
BAW-10104 - B&W's ECCS Evaluation Model
)68 310 6.0 PHASE 2 SCHEDULE No detailed schedules are included with this report because of the complex interactions requited for the analyses described herein.
However, at this time, it is projected the conclusion reports will be available by April, 1980.
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