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September 18, 1981

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1 UNITED STATES OF AMERICANUCLEAR REGULATORY COMMISSION

2BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

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In the Matter 01 S4

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HOUSTON LIGHTING & "OWER COMPANY S Docket No. 50-466.

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(Allens Creek Nuclear Generating 56 Station, Unit 1) S

7TESTIMONY OF ROBERT R. HOBSON AND

8 THOMAS G. DUNLAP REGARDING DOHERTYCONTENTION NO. 45 - CORE LATERAL SUPPORT

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10 Q. Mr. Hobson, would you please state your full name,

11 your position and describe your educ '.ional and professional

12 background?

A. My name is Robert R. Hobson. I am employed at13

General Electric Company (GE) as a Principal Design Engineer,74

R1 actor Pressu*e VessL1 and Internal Design. My educational,_.c

and professional background is described in Attachment RRH-1.

Q. Mr. Dr.tlap, would you please state your full name,17

your position and describe your educational and professional18

background?19

A. My name is Thomas G. Dunlap. I am employed at GE

20 as a principal engineer in fuel assembly design and analysis.2^1 My educational and professional background is described in22 Attachmenc TGD-l.23 Q. Mr. Hobson, are you responsible for the core

24 structure design of the BWP-6 reactor?

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A. Yes.1

Q. How do you define the core structure?3

A. Basically, the core structure consists of the

shroud, core plate, control rod guide tubes, fuel supports,

top guide, and shroud head and separators assembly.5 lC. Do you have to design the core structure to with-6

stand flow forces in a lateral direction?7

A. The configuration of the core structure provides8 axial coolant flow into, through, and out of the core area.

9 The only exception to axial coolant flow is when a lod pressure10 coolant injection event occurs which gives relatively small

*1 radial reaction forces which are accounted for in the shroud9

analyses. Therefore, there are no appreciable unbalanced12

13 unbalanced flcws or flow forces in a lateral direction in14 the core area.

15 Q. Is this also true for a simultaneous Loss-of-

16 Coolant Accident (LOCA) event and seismic event?

17 A. For t LOCA event, it ir assumed that a recircula-

tion outlet line is severed as it is the LOCA that most18

highly loads the core structure. In the core area describedyg

above, this LOCA will cause a change in a p across various20

components due to outside pressures Cropping and flashing

occurring. This change in A p is included in analyses,

such as, loc, stress in the shroud. There are no resulting

23unbalanced lateral loads in the core area. Prior to this

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1 change in a p, there are loadings from this LOCA event

2 which will cause lateral loads outside the shroud. Any

simultanecus dynamic loading such as seismic will cause3

lateral loads both inside and outside the shroud. The4

various core structure components are analyzed for this-

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OVent.gt

Q. Would you please describe how the core structure.

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components are analyzed for such an event?

A. The following impulsive loads are included in the9 '

design of the shroud and shroud support. These loads apply

10 in addition to the steady state operc. ting condition loads11 which precede the changes in pressure differentials which12

result from the pipe rupture. These impulsive leads would-

9~3 occur as a result of the acoustical propagation of the reduced1 pressure caused by a recirculation outlet pipeline rupture15 which is a faulted condition.

a. A unit impalsive load of 650,000 pounds for 516

17 millisecords duration is acting uniformly over

18 the shroud rul grid cylinder surf ace toward the

19 reactor recirculation outlet nozzle to which thebroken line is attached. The effects of the20

1 calized loads of item "b", below, are not added21

to this load as they occur at a diffcrent time.22

b. L aliz d unit impulsive loads of:23

(1) 200 P i applied outwardly across the shroud over an24

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1 area equal to the area of the recirculation line

2 for 5 milliseconds. This pressure includes the

3 effect of the steady state operating pressure

differentials..*

(2) Shrc ad support loading resulting from 75,000-

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POLnds acting for 5 milliseconds uniformly over the6

length of the jet pump dif fuser toward the,

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recirculation oatlet nozzle. This may be treated

as a 75,000 pouad static load applied to the jet9

pump diffuser 54 inches above the shroud support10

plate and the diffuser is treated as cantilevered11

from the shroud support plate. These loads apply

12to the jet pump dif fusers adjacent to the reci rcu-

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lation outlet nossle.

9These impulsive loads are then combined with othe- loads-

15 that occur simultaneously. In accounting for the impulsive

16 loads we consider flashing both inside and outside the core

17 structure and combine these loads with other appropriate

18 loads. Using these loads in our stress analyses we have

19 determined that the core structur< does comply with AEME Code

20 allowables. Therefore, the core structure retains its

21 physical integrity under all design loading conditions.

Q. Mr. Dunlap, are you responsible for the design of22

tests and analyses which determine the adequacy of the BWR/623

fuel assembly fcr withstanding mechanical loadings?,.

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1 A. Yes.

2 Q. I;a s the Allens Creek fuel assembly been designed

to withstand simultaneous scismic and LOCA loads?3

A. Yes. The ACNGS reactor core is composed of BWR/64

YF<^ fuel assemblies. These fuel assemblies have been5

designed to withstand the combined loadings from a safe

shutdown earthquake and a Loss-of-Coolant Accident (LOCA).,

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Documentation of the capabilicy of the fuel assemblies to8

withstand this loadir.g is contained in NEDE-21175-P , "BWR/69

Fuel Assembly Evaluation of Combined Safe Snundoen Earthquake10

(SSE) and Loss-of-Coolant Accident (LOCA) Loadinns." The

11 documented evaluations demonstrate that the BWR/6 fuel

assemblies are capable ci withstancing combined seismic

and LOCA loads which are in excess of the ACNGS seismic14 and LOCA loads.

15 Specifically, this report documents how the load

16 Enths are determined, the magnitude of loading on 'ach fuel

17 assembly component part, the lower bound material property

18 limits and the tests and analyses which demonstrate the

19 capability of the fuel asse:.bly to withstand these combined

20 loadings.

Since the above mentioned document was published,21

the NRC has published NUREG/CR-1018, " Review of LWR Fuel22

System Mechanical Response with Recommendations for Component3

Acceptance Criteria." This document provide; recommendations

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1 for component acceptance criteria. Specifically, recommenda'

tions are made for determining allowable loads on spacer grids2

and for fuel assembly ,omponents other than spacer grids.3

The ACNGS fuel assemblies are in compliance with these4

recommendations.3

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Attachment RRH-1

ROBERT R. HOBSON

COMPANY: General Electric Co.

TIfLE: Principcl Design EngineerReactor Precsure Vessel and Internals Design

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DEGREE: B.S.M.E.

LICENSES: Professional Engineer-Mechanical - CaliforniaProfessional Engineer-Mechanical - Pennsylvania

EXPERIENC :: 40 years with General Electric Co. asMechanical Engineer.

1956 to date Nuclear Power Plants - designed BWR/6 core plate,shroud, top guide, and shroud head andseparators assembly. Conceptual design ofhardware for joining the above, and theconceptual design of tha steam dryer, corespray spargers, aad LPCI coupling. Wrotecore suppc-t structure and shroud head designspecificat.ons, functional specification,purchase specification, and handling specifica-tion for BWR/6 core structures. Assisted ASMECode working group in preparation of initialSubsection NG.

Provided technical leadership or consultingrole in resolution of fabrication and installa-tion problems, resolution of impact of newdynamic loads and special tests, such as,core spray, LPCI, and vibration.

Prior to BWR/6, had design responsibilityfor various reactor components includingreactor pressure vessels, control rod drives,core structures, and steam dryers on most ofGeneral Electric reactors including nuclearthermionic reactor for space application.

1941 to 1956 Had technical responsibility for development,design, fabrication and testing of 20 mmVulcan machine gun now used by UST.F and other,

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forces. This involved the resolution ofvery large dynamic and vibratory loads undera large temperature ranga. Had similartechnical responsibility for remote controlaircraft machine gun turrets. Variousresponsibilities on products from 5" Navy gunmounts to refrigerators to steam turbines.

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Attachment TGD-1

THOMAS G. DUNLAP

Mr. Dunlap is a Principal Engineer in the Fuel AssemblyDesign Unit at General Electric. His primary responsibilityis direction of .nnalysis and testing of fuel assemblycomponents for assurance of structural integrity for various

Mr.' loading conditions including LOCA and seismic loading.Dunlap has hald this position since 1974.From 1972 to 1974, Mr. Dunlap was responsible engineer forfuel rod mechanical design and analysis, also as a member ofthe Fuel Assembly Design Unit.

Dunlap is a Professional Engineer and has been a majorMr.contributor to the development of American National Standard:ANSI /ANS-5 7. 5 - 198_ " Light Water Reactors Fuel AssemblyMechanical Design and Evaluation."

From 1966 to 1972, Mr. Dunlap was a mechanical engineerassociated wath the design of aircraft engines for GeneralElectric's Flight Propulsion Division.

Mr. Dunlap received his BSME degree from the University ofTennessee in 1964 and was employed as an instrumentationoesign engineer by United Aircraft from 19C4 to 1966.

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