ML20077J680

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Nonproprietary Rev 1 to Final Assessment of Waterford-3 Fuel Structural Integrity Under Faulted Conditions. Info Deleted
ML20077J680
Person / Time
Site: Waterford Entergy icon.png
Issue date: 07/15/1983
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML19277D687 List:
References
CEN-159(C)-NP, CEN-159(C)-NP-R01-NP, CEN-159(C)-NP-R1-NP, NUDOCS 8308180085
Download: ML20077J680 (5)
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t WATERFORD-3 DOCKET 50-382 9:

,3 CEN-159(C)-NPRev.1-NP FINAL ASSESSMENT OF WATERFORD-3 FUEL STRUCTURAL INTEGRITY UNDER FAULTED CONDITIONS JULY 15, 1983 A

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, COMBUSTION ENGINEERING, INC.

NUCLEAR POWER SYSTEMS POWER SYSTEMS GROUP WINDSOP., CONNECTICUT 06095 c

. , 8308180085 830812

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LEGAL NOTICE This report was prepared as an account of work sponsorad by Combustion Engineering, Inc. Neither Combustion Engineering nor any person acting on its behalf:

i A. Makes any warranty or representation, express or

  • implied including the warranties of fitness for a particular

. purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.

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  • INTRODUCTION i

This document formally transmits.the final assessment of the Waterford-3 fuel structural integrity under faulted condi.tions. The anlytical tech-niques, models, and acceptance criteria used for the seismic and LOCA analyses follow the methodology described in CENPD-178, Rev.1 (Ref.1)..

REFERENCES:

1) CENPD-178-P, Rev. 1-P, " Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading", C-E Proprietary Report, August 1981
2) Final Safety Analysis Report, San Onofre Nuclear Generating Stations Units 2 and 3, NRC Docket Nos. 50-361 and 50-362, Response to NRC Question 231.26
3) WSES-3 FSAR l

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  • V FINAL SEISMIC AND LOCA FUEL ANALYSES The seismic and LOCA analyses of the fuel followed a step-by-step procedure.

In the first step, the input excitations to be used for the coupled reactor internals and core models were developed. These consisted of the horizontal, vertical, and rotational (rocking) time-history responses of the reactor vessel determined from the reactor coolant system analyses. The LOCA analysis also included the fluid pressure transient forces resulting from each postulated pipe break. These excitations were input into separate horizontal and vertical models of the reactor internals and core.

In the horizontal direction, the motions of the core plates and core shroud,

, which were used as input to detailed models of the reactor core, were obtained from the analyses of the coupled internals and core models. Separate core models were used to analyze rows of the core with different numbers of

. fuel assemblies. The seismic and LOCA horizontal direction core analyses were performed using models representing 17, 9, and 4 fuel assemblies in a row.

For the vertical direction, the core response was obtained directly from the coupled internals and core model analyses.

The LOCA analyses were performed for a postulated 125 in2 break (see Ref. 3, Section 3.6.2.1.1) at the reactor vessel inlet nozzle and a 600 in2 break (Ref. 3, Table 3.1) at the steam generator inlet nozzle. The seismic analyses were performed for the Safe Shutdown Earthquake (SSE). These analyses provided the peak spacer grid one-sided and through-grid impact loads which were used in the final assessment of the Waterford-3 fuel.

The longest row case (17 fuel assemblies) provided the most severe impact loads in the seismic analyses, whereas, the highest LOCA loads were found in the shortest row case (4 fuel assemblies). Peak impact loads from both the seismic and LOCA analyses are summarized in Table 1. The resulting loads do not exceed the spacer grid strength, except in isolated locations in the four-assembly rows of the core.

An Emergency Core Cooling System (ECCS) analysis was performed- to determine the coolability of these peripheral assemblies with reduced flow area due to grid deformation. The methodology and acceptance criteria used in this ECCS evaluation were identical to previous analyses of this type (Ref. 2). As documented in Ref. 3 (Section 4.2.3.1.3), the ECCS acceptance criteria were satisfied for the Waterford-3 spacer grids.

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.. TABLE 1

, 1 COMPARISON OF SPACER GRID IMPACT LOADS WITH Pcrit VALUES NUREG-0609, " Asymmetric Blowdown Loads on PWR Primary Systems, Resolution of G:neric Task Action Plan A-2, January 1981", states that it is a sufficient LOCA acceptance criterion to show that combined loads on the grids remain below Pcrit. It is our position that it is not necessary to combine seismic and LOCA loads on the grids. For the convenience of the NRC review, the seismic and LOCA loads on the grids have been combined (by SRSSI ) and compared to values of Pcrit.

2 2 l Peak Peak Peak Pcrit i

LOCA SSE Combined Loads Load Load Case . Description (Lbs) (Lbs) (Lbs) (Lbs) 17 Fuel Assembly Row One-side load .

1 Through-grid load 9 Fuel Asse'mbiy Row .

One-sided load -

Through-grid load .

4 Fuel Assembly Row .

One-sided load E Through-grid load 9

1 = The combined loads are.obtained from the square root of the l sum of the squares of the SSE and LOCA loads for the same grid (function of grid location).

2 = ' Designates the peak loads independently of grid location.

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