ML19326D078
| ML19326D078 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 11/10/1978 |
| From: | ARKANSAS POWER & LIGHT CO. |
| To: | |
| References | |
| NUDOCS 8006040471 | |
| Download: ML19326D078 (9) | |
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B&W 177 FA OWNER'S GROUP ASYMMETRIC LOCA LOADS EVALUATIONS PROGRAM, PHASE 2
'I IHIS DOCUMENT CONTAINS
'I POOR QUALITY PAGES Arkansas "ower & Light - AMO-1 Duke' Power Company - Oconee 1, 2, 3 Florica Power Corporation - Crystal River 3 Metropolitan Edison Company - Three Mile Island 1, 2
,e Sacramento Municipal Utility District - Rancho Seco Toledo Edison Company - Davis-Besse 1 L
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l November 10, 1978
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4 COCEl:TS i-I 1.0 Introduction
.2.0 Evaluation Bases 3.0 Werk Plan (Phase 2) r 4.0 Com3L:er Codes 5.0 Applicable B&W Topical Reports i
5.0 Phase 2 Schedule y.
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3 i.0. INTR 000CTION This report summarizes _ Phase 2 Of tne 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 outlined in the reports, "B&W 177 FA Owner's Group Asymetric LOCA ~ Loads Program," dated April 10, 1978, and " Toledo Edison Company Asymmetric LOCA Loads Evaluation Program for Davis-Besse 1 Plant," dated April 21, 1978.
Whereas, Phase 1 performed oreliminary investigations using estimates and generalities to bracket the evaluation recuirements, Phase 2 will go into greater detafis to determine more specific loads and component /
. structural evaluaticns.
The need to proceed with Phase 3 is still unresolved at this time, and
-will be addressed at a later date cfter discussions with the NRC.
2.0 EVALUiTION EASES-2.1 Tne components to be evaluated curing Fhase 2 for the LOCA breaks analyzed include.
a.
Reactor Pressurc Vessel b.
Fuel Assemblies, Incluaing Grid Structures c.
Centrol Rod Crives
- d. ~ ECCS Pipi. g tnat is Atticud to the Primary Coolant Piping
. Primary Coolant Piping f a Cicse Proximity to the Reactor Vessel e.
f.
Reactor Vessel Supports
- g.
Reactor Internals
- h.. Biological Snield Wall ana Neutron Shield Tank (where applicable) 1.
Core Ficcdi% Piping f.
Related B;iicing Structures 2.2 LOCA analyses will je performed for creaks rendering the worst loadings for the Reacte,r Vessel supsorts and Reactor Internals.
For these breaks, all components listad in Parsgraph 2.1 will be evalua'ted to assure (1) maintain:ng core coolable geometry and (2) mitigating the conse-quences of an accident.
2.-3' Jet finpingement effects.will be evaluated for breaks i,nalyzed. This
-evaluation was nct explicitly stated'in the NRC letter, but was
-identified as a requirement in a previous meeting (March 31,1978) n
.with~the NRC.
- 2.4 ' As-appropriate, tne evaluaticn will consider:
a.
Limited displacement-break areas wnere applicable 5.
Use of actual time-dependent forcing function
. c.
Reactor se? cort stiffness c.
Break cpning tices e.
Break;1ocatier otDizing s:ress criteria 6-l 6
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2.5.Where justifiable, a generic review of the S&W Owner's Group plants will be used.
s 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 en the reactor vessel in the ever,t.of a reactor ccolcr.t-system (RCS) pipe rupture inside of the reactor subccmpar;. Tan:.
ThE ~. RAFT 2 computer ccde will be used to calculate the trar.sient, asymmetric oressure. distributions inside
.tne subccmpartment for a spectrum of break cases. Analysis guidelines established in Stancard Review ?lant (SRP) 6.2.1.2 for subcompartment pressurization calculations and in SRP 6.2.1.3 for mass and energy release calculaticns will be followed.
Three different reactor caeity designs will be evaluated. Design selection will te ba;ed upon the following considerations:
1.
Cavity voltme t.etween reactor vessel and primary shield wall.
2.
Insulation des :gn.
3.
Vent areas'ef primary piping penetrations.
4.
Shield plugs or blow-out aevices.
5.
Flow cbstructions.
.The 177. plants (0.anar's 8 000) aill be categorized by B&W into three grcups, each'g-otp being represented by one of the cavity designs.
3.2 Develop a mass and energy release calculation model for a generic 177 lovered-locp R:: cparcting at a power level of 1.02 X 2772 MWt.
Generate r.. ass and snerg; re!etse data _for the initial two (2) seconds
-of blowdown for hot and ccid leg breaks of the following sizes:
a.
2.0A b.
0.6A c.
0.3A d.
To be determined e.
To be determined 3.3 Develcp e 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 cata for the initial two (2) seconds of blowdown for hot and cold leg breaks of tne following sizes:
-2 a.
.o it 2
b.
1.0 ft.
c.
.5A d.
1.0A e.
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~ 314 ~TI.e core ' loco line N C ce irerted as a cold leg break of appropriate
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i3.5 For evaluation of the mass and energy data generated in Paragraph.3.2 develop tnree (3) reactor cavity models.. Calculate the reactor cavity
' pressurization rates.for -tne spectrum of hot and cold leg breaks identified. -Calculate the timo-histories of the lateral and vertical forces and_ moments actir.g on tae reactor vessellout'to a blowdown time which 'is1 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.
Calculate the reactor cavity pres-
-surization rates for tne spectrum of hot and cold leg breaks identified.
Calculate the time-histories of the lateral and vertical forces and moments acting on :ne reactor vessel out to a blowdown time which is suf ficient -to. define the peak magnitudes Of the forces and moments.
3.7~~ Calculate the itss-cf-coclant-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 tha reactor coolant system (RCS).
Break locations inside the reactor cavity and outside the primary shield wall in the steam
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generator compartment will be considered.
3.7.1 A generic model will be developed for calculating LOCA loads on 177-FA lowerec-loop plants. The modeling criteria established in Tcpical' Report EA'J-10132 will be used in the development.
Initial reactor fluid conditions which enccmpass.al' 177 plants fo~r purposes of LCCA 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 casesJin the steam generator compartment, with a contingency for;anaiy ing un ce four.(4) additional breaks anywhere in the RC5. The brial siras and tne corrasponding break opening times will be seleited on the basis of the results of the Phase 1 program.
The analyses will be conducted out to 0.3 sec of the bi7:!cown.
-The following pcrameters will be recorded as a function of time for each. design case _ calculation:
a.
Control volume pressures.
b.
Msjor couponsnt LF's.
c.
Vertical force on tne core.
d.
Vessel hs:d LP.
e.
Integrated lateral forces on pressure vessel and core support cylinder,
- f. JIntegrated_16teral icad on plenum cyclinder,
- g. : Mass and' energy release to containment.
n.
Jet intensity at bre&k plane.
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3.7.2 The rea cor internals LOCA. load model us d for cnalysis on Davis-Besse 2 and 3 will be converted over for Davis-Besse.l.
This' conversion will include a change.to a 15 X 15 fuel assembly model, a change to Davis-Besse 1 specific fluid conditions, and changes = brought about 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 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 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 mooified 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 supolied to the linear model include the dynamic impact of the fuel assemblies in the vertical direction.
3.10. The b~ending 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 plenum assembly, the apparent differences of two redundant load paths will be calculated:
the plenum cylinder path and the column weldment path.
This is accom-i plished with a three-dimensional finite element 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 stiff-ness. This method accounts for the effects of the plenum cover and upper grid which will not be included in the isolated model.,
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- i 3.11 Existing 177 furl assembly reactor vessel internals model (TECO, 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 I
the reactor vessel internals as beam elements obtained earlier, and the fuel assembly model also obtained earlier. The model will
.inclade service support structure, CRDM, cold leg piping, and hot
-leg piping to the extent' feasible.
The TEC0 Davis-Besse 1 model will reflect the internals design for Davis-Besse 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 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 to 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 ar,a core, and all the vertical dyn6mics of the internals.
d.
Asymmetric cavity pressures are integrated over the outside sur-face 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 e
mass coupling developed separately.
3.13 Non-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 identifiad 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 of the fluid on the primary coolant boundary, the actual break area 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
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boundary. The.model for flow volumes will require the definition of the break area (leak area) for a time-history calculation of the forcing functions. L
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The solution tur tht forcing functions will require iteration.to -
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obtain a solution based on consistent conditions (break area and dynamic response).
The results will be obtained in the form of area versus time.
3.151 Fuel assembly deformation limits will be-established based upon allowable grid deformations as determined by ECCS requirements.
These established. requirements will be confirmed by analysis to assure that peak cladding temperatures 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, 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 asymetric pressures.
Due to the very limited time available for Component Evaluations, complete structural ana?yses 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
~. LUMS d
e.
STARS-f.
CRAFT 2 g.
SUPEnPIPE' h.
GDSGAP
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'5 0' APPLICABLE B&W TOPl xL REPORTS l
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. 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 BAW-10132 -- Analytical Methods Description -- Reactor Coolant System c.
Hydrodynamic Loadings During a Loss-of-Coolant Accioent
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d.
BAW-10133 -- Mark C Fuel Assembly -- LOCA-Seismic Analyses BAW-10060 -- Reactor Internals Design / Analysis for Normal, Upset e.
and Faulted Conditions ti; 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 coinclusion reports will be available by March, 1980. w_