ML19319D395
| ML19319D395 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 04/10/1978 |
| From: | Arkansas Power & Light Co, Duke Power Co, Florida Power Corp, Metropolitan Edison Co, Sacramento Municipal Utility District (SMUD) |
| To: | |
| Shared Package | |
| ML19319D392 | List: |
| References | |
| TAC 08843 NUDOCS 8003160230 | |
| Download: ML19319D395 (7) | |
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B&W 3 77FA OWE"S GROUP ASYMMETRIC LOCA LOADS EVALUATIONS PROGRAM l
Arkansas Power & Light - ANO 1 Duke Power Company - Oconee 1, 2, 3 Florida Power Corporation - Crystal River 3 miopolitan Edison Conpany - Three-Mile 141and 1, 2 Sacramento Municipal Utility District - Epncho Seco l
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CONTENTS 1.0 Iutroduction 2.0 Evaluation Bases 3.0 Work Plan (Phases) 4.0 Computer Codes
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5.0 Applicable B&W Topical Reports 6.0 Schedules e
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1.0 INTRODUCTION
This report summarizes the detailed plan prepared by the B&W 177EA Owners Group in response to the NRC Division of Operating Reactors letter dated January 25, 1978.
The plan described herein is separated into three phases.
Each phase is described to the level of detail possible at this time. The phasing is intended to allow progression toward a completed assessment by providing for intermediate evaluations as the program proceeds.
This. plan is based upon th. 2nderstandings achieved in a meeting between the B&W Owners Group and NRC/ DOR on March 31, 1978.
2.0 EVALUATION BASES 2.1 All components listed in Enclosure 2 of the NRC letter will be addressed for the LOCA breaks evaluated. This includes:
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 f.
Reactor Vessel, Steam Generator and Pump Supports g.
Reactor Internals h.
Biological Shield Wall and Neutron Shield Tank (where applicable) 1.
Steam Generator Compartment Wall-2.2 LOCA analysis vill 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 vill be evaluated to assure (1) maintaining core coolable geometry and (2) mitigating the consequences 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 ss a requirement in the March 31, 1978, meeting mentioned in paragraph 1.0.
2.4 As appropriate, the evaluation vill consider:
limited displacement breaP areas where applicable a.
b.
use of actual time-dependent forcing function c.
reactor support stiffness d.
break opening times e.
break location utilizing stress criteria 2.5 If results of the evaluation indicate loads
- 1.. ding to inelastic action or displacements exceeding previous design limits, then inelastic be-havior (including strain hardening) of the material analyzed and the 4
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2.0 EVPLUATION BASES (continued) 2.5 effect on the load transmitted to the backup structures to which the-component is at' ached will be included.
t 2.6 Where justifiable, a ger tric review of the B&W Owners Group plants will be used. The categorization timing and extent will be discussed later in this report.
L.0 WORK PLAN (PHASES) 3.1 Phase 1 will be a short duration (7 month) preliminary assessment.
The specific plant drawings will be. reviewed to assess the similarity of the various plants to assure that asymmetric pressures will be similar for all plants in each category.
3.1.1 A preliminary scoping study of each plant's restraint design will be performed. The results of this study will be esti-mated maximum pipe break opening areas for each of four breaks (upper cold leg and hot leg guillotine at the Reactor Vessel nozzle and upper cold leg and hot leg guillotine outside the primary shield wall). The location of the break outside the primary shield wall vill be determined with acceptable break location criteria and from these, design cases will be chosen based on parametric studies performed by B&W on their 205EA piknts and a results comparison for these plants under evaluation.
3.1.2 The peak magnitudes of the major LOCA load components acting on the reactor internals will be estimated as a function of breaksize.
Sensitivity study results which are available for B&W 205EA plants will be used to develop scaling factors for estimating loads on the 177EA plants. The particular loads which will be considered are (1) total lateral force on the core support cylinder; (2) total vertical force on the reactor vessel due to head differential pressure; and (3) vertical force on the core. These loads will be estimated for the four breaks described in paragraph 3.1.1.
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3.1.3 Estimates for the magnitude of peak lateral force which acts externally on the reac*or vessel due to asymmetric pressures within the reactor cavity will be made. These estimates will be extrapolations made from existing 177 cavity pressure data to include a consideration of break size.
3.1.4 Using the estimated, asymmetric cavity and internals pres-sures determined in paragraphs 2.1.3 and 3.1.3, a comparison
- between the applied loadings and the load carrying capability of the Reactor Internals and the Reactpr Vessel support for
- each plsnt will be made.
Based cn1 this co=parison, additional analysis '.nd hardware mor'.ifications ill be recommended.
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3.0 WORK PLAN (PHASES)
(. continued) 3.2 Phase 2 analysis will be initiated-if results of Phase 1 indicates a need for more detailed review and/or a need to review some of '
the plants on a specific case basis.. The extent of analysis cannot be specified until the results of Phase 1 are known.
During this phase,one, or a combination, of the following three action paths will be pursued:
a.
Detailed Analyses b.
Hardware Modifications c.
Licensing Actions As du Phase'1, this phase will focus,on the Reactor Vessel and structures / components in close proximity.
If the results of Phase 1 are acceptable, conclusive and defendable, this phase will not be executed.
If it is required to progress on to this phase, an additional detailed plan with schedules will be submitted to the NRC.
3.3 Phase 3 analysis will also only be initiated if the results of Phase 1 indicate a need for a more detailed review. Whereas Phase 2 concentrates on the Reactor Vessel area, this phase will focus on the Steam Generator and R.C. Pump areas.
Here again, there exists the possibility of three courses of action, as outlined in paragraph 3.2, and until the specific needs are identified from Phase 1 efforts, the details of this phase cannot be identified. If it is required to execute this phase, an addi-j tional detailed plan with schedules will be submitted to the NRC.
i 4.0 COMPUTER CODES in t'.ne performance of the analyses, several different computer codes will 3
be used. The following list identifies those codes:
a.
ANSYS'
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b.
ADINA c.
ST3DS d.
LUMS e.
STARS f.
CRAFT 2 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:
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.5.0 APPLICABLE B&W TOPICAL REPORTS (continued)
BAW-10131 - Reactor Coolant 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.
Hydrodynamic Loadings During a Loss-of-Coolant Accident
_,d.
BAW-10133 - Mark C Fuel Assembly - LOCA - Seismic Analyses BAW-10060 - Reactor Internals Design / Analysis for Normal, Upset and i
e.
Faulted Conditions 6.0 PLAN SCHEDULES 6.1 Phase 1 schedule is as follows:
Activity 1977 Description April May June July August September October
- 1. Preliminary Scoping i
Study (Paragraph 1
3.1.1)
- 2. Reactor Internals LOCA Pressure o
I Analysis (Paragraph 3.1.2)
- 3. Reactor Cavity
-Asymmetric Pressure o
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~ Analysis (Paragraph I
3.1.3)
- 4. Results Assessment' o
(Paragraph 3.1.4) 6.2 Phases 2 and 3 schedules cannot be firmed up until specific detail needs are known. However, the overall program schedule is as follows:
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6.2 '(continued) 1978 1979 1980 MA AP }.A J1'JLAU SE OC N0lDEJA FE MA @ MA JU JL AU SE OC NCDE JA FE MA AP JU JU AU SE OC LICENSING
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PRELIMINARY HARDWARE 1 MODIFICATION
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ASSESSMENT j
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I DETAILED ANALYSES 6.3 As shown in paragraph 6.2, all analysis can probably be completed within the two year time frame discussed in the NRC letter. How-ever, if hardware fixes are required, full implementation of all fixes would exceed the two year time frame allowing for caterial procurement, fabrication, scheduled shutdowns and erection. The NRC will be kept advised of firm dates as they are determined.
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