ML20206F886

Amend 4 to RESAR-SP/90 Pda Module 5, Reactor Sys

ML20206F886
Person / Time
Site:	05000601
Issue date:	11/30/1988
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19295G691	List: ML20206F868 ML20206F886 NRC-88-3382, Forwards Proprietary & Nonproprietary Amend 4 to RESAR-SP/90 Pda Module 5, Reactor Sys. Proprietary Rept Withheld (Ref 10CFR2.790)
References
RESAR-SP-90, NUDOCS 8811210384
Download: ML20206F886 (26)

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WESTINGHOUSE CLASS 3 1

h AMENDMENT 4 TO RESAR-SP/90 PDA MODULE 5 REACTOR SYSTEN

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1, WAPWR RS AMENDMENT 4 i

5379e

1d NOVEMBER, 1988

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AMENDMENT 4 TO RESAR-SP/90 PDA MODULE 5 REACTOR SYSTEM O INSTRUCTION SHEET i

Replace current pages 11/111 with revised pages 11/i11.

Replace current page 1.6-4 with revised page 1.6-4. *

. Replace current page 1.6-6 with revised page 1.6-6.  :

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Replace corrent pages 3.2-1 through 3.2-3 with revised pages l 3.2-1 through 3.2-3.

ll I l Replace current page 17.0-1 with revised page 17.0-1.

l Replace current page 3.9-1/3.9-2 with revised pages 3.9-1

', through 3.9-2.

i insert pages A4-1 through A4-8 in Question / Answer section, following Amndment 1.

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! WAPWR-RS AMENDMENT 4 i 5379e:1d NOVEMBER, 1988

h TABLE OF CONTENTS Reference O Section Title Pace SAR Section Status

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF 1.1-1  !!

PLANT

1.1 INTRODUCTION

1.1-1  !!

1.2 GENERAL PLANT DESCRIPTION 1.2-1  !!

1.2.2 Principal Design Criteria 1.2-1  !!

1.2.3 Plant Description 1,2-1 II 1.2.3.1 Reactor System 1.2-1 II 1.3 COMPARISON TABLES 1.3-1 II 1.3.1 Comparison With Similar Facility Designs 1.3-1 II 1.5 REQUIREMENTS FOR FURTHER TECHNICAL 1.5-1  !!

INFORMATION O 1.5.1 Fuel System Tests 1.5-1 I 1.5.1.1 Fuel Assembly Tests 1.5-1 1 1.5.1.1.1 Fuel Assembly Structural Tests 1.5-1 I f 1.5.1.1.2 Fuel Assembly Hydraulic Flow Tests 1.5-2 I 1.5.1.2 Core Components Tests 1.5-3 I  ;

1.5.1.2.1 Red Cluster Control and Gray Red 1.5-3 I i Assembly Tests' i 1.5.1.2.2 Water Displacer Rod Assembly Tests 1.5-3 I  ;

1.5.1.3 Drive Mechanism Tests 1.5-4 I i l 1.5.1.3.1 Control Rod Drive Mechanism Tests 1.5-4 I [

l 1.5.1.3.2 Water Displacer Rod Drive Mechanism 1.5-5 I l

! Testa 1.5.2 Reactor Internals Design Verification 1.5 6 I L Tests l 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6-1  !!

I 1.8 CONFORMANCE WITH THE STANDARD REVIEW 1.8 1  !! (

PLAN i O

i WAPWR RS 11 ,

l 5379e:1d JULY. 1984

TABLE OF CONTENTS (cont)

Reference O

SAR Section Section Title Page a Status 2.0 SITE CHARACTERISTICS 2.0-1 NA O

3.0 DESIGN OF STRUCTURES, COMPONENTS, 3.1-1 11 EQUIPMENT AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN 3.1-1  !!

CRITERIA 3.2 CLASS!FICATION OF STRUCTURES, COMPONENTS, 3.2-1  !!

SYSTEMS 3.2.1 Seismic Classification 3.2-2  !!

3.2.2 System Quality Group Classification 3.2-2  !!

3.2.3 Safety Classes 3.2-2 11 3.2.4 References 3.2-2 Il 3.9 NECHANICAL SYSTEMS AND COMPONENTS 3.9-1  !!

3.9.2 3.9.2.3 Dynamic Testing and Analysis Dynamic Response Analysis of Reactor 3.9-1 3.9-1 1

g Internals Under Operational Flow Transients and Steady-State Conditions 4 3.9.2.4 Preoperational Flow-Induced Vibration 3.9-le I t Testing of Reactor Internals 3.9.2.5 Dynamic System Analysis of the Reactor 3.9-2 I Internals Under Faulted Conditions 3.9.2.6 Correlations of Reactor Internals 3.9-3 I 3.9.4 Vibration Tests With Analytical Results Rod Drive Systems 3.9-3 1 g

3.9.4.1 Descriptive Inforr.ation of the Control and 3.9-3 1 ,

Gray Rod Drive Systems 3.9.4.1.1 3.9-3 Control Rod Drive Wechanism (CRDN) and Gray Rod Drive Wechanism (GRDM)

I gi 3.9.4.1.1.1 RCCA and GRA Withdrawal 3.9-7 1 3.9.4.1.1.2 RCCA and GRA Insertion 3.9 9 1 O

WAPWR-RS iii AMENDHENT 4 5379e:1d NOVEMBER, 1988 l

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MATERIAL RbAE EFERENCE Westinghouse SAR Topical Revision Section Submitted Review g Report No. Title Nueber Reference to the NRC Status WCAP-7267-L(P) Core Power Capability Rev 0 4.3 10/69 0 WCAP-7809 in Westinghouse PWP.s WCG-1308-L(P) Evaluation of Nuclear Hot Rev 0 4.3 7/9/70 V WCAP-7810 Channel Factor Uncertainties 12/16/71 7

WCAP-7359-L(P) Application of THINC Rev 0 4.4 9/8/69 0 WCAP-7838 Program to PWR Design 1/17/72 WCAP-7588 Evaluation of Rod Ejection Rev 1A 15.4 1/7/75 A Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods WCAP-7667-P- Interchannel Thermal Mixing Rev 0 4.4 1/27/75 A A(P) With Mixing Vane Grids WCAP-7755-A WCAP-7695-P- DNB Tests Results for Rev 0 4.4 1/21/75 A O- A(P) NewMixir.gVaneGrids(R)

WCAP-7958-A WCAP-7706-L(P) An Evaluation of Solid State Rev 0 4.6 9/2/71 0 l WCAP-7706 Logic Reactor Protection t in Anticipated Transients WCAP-7800 Nuclear Fuel Business Unit Rev 7 4.2 10/6/88 U

' E Ouality Assurance Prograai Plan WCAP-7907-P-A LOFTRAN Code Description Rev 0 15.0, 15.4 10/11/72 1.

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15.0, 15.4 O WCAP-7908 FACTRAN - A FORTRAN-IV Code for Thermal Transients in a U02Fuel Red Rev 0 9/20/72 U WCAP-7912- Power Peaking Factors Rev 0 4.3 1/16/75 A P-A(P)

WCAP-7912-A O WAPWR RS 1.6-4 AMENDMENT 4 8379e:1d NOVEMBER, 1983

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TA8LE 1.6-1 (cont)

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MATERIAL INCORPORATED BY REFERENCE 1 l

Westinghouse SAR Topical Revision Section Submitted Review, Report No. Title Nuter Reference to the NRC Status V WCAP-8306 SATAN-VI Program: Compre- Rev 0 15.0 7/12/74 AE hensive Space-Time Depen-dont Analysis of Loss-of-Coolan',

WCAP-8330 Wstinghouse Anticipated Rev 0 4.3. 4.6, 9/25/74 U s Transients Without Trip 15.4 ,

Analysis WCAP-8359 Effects of Fuel Densifi- Rev 0 4.3 7/2/74 AE cation Power Spikas on Clad Thermal Transients WCAP-8370 Westinghouse Energy Systems Rev 11 17 10/6/88 U 4 Business Unit Quality Assurance Plan WCAP-8377(P) Revised Clad Flattening Rev 0 4.2 8/7/74 A WCAP-8381 Model 8/6/74 WCAP-8385(P) Power Distribution Control Rev 0 4.3 10/9/74 A WCAP-8403 and Lead Following Procedures WCAP-8453-A(P) Analysis of Data from Rev 0 4.4 5/10/76 A l

WCAP-8454 Zion (Unit 1) THINC Veri-fication Test l

WCAP-8498 Incore Power Distribution Rev 0 4.3 7/22/75 U l Determination in Westing-house Pressurized Water Reactors. Program Sumaries -

1 Fall 1974 '

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WCAP-8567-P(P) Improved Thermal Design Rev 0 4.4, 15.0 7/75 A WCAP-8558 Procedure WCAP-8575(P) Augmented Startup and Rev 0 4.3 6/76 U WCAP-8576 Cycle 1 Physics Program Supplement 1 WCAP8584(P) Failure Mode and Effects Rev 0 4.6 4/23/76 U WCAP-8760 Analysis FMEA) of Engi- Rev 1 2/80 neered Sa eguard Features Actuation System O WAPWR-RS 1.6-6 AMENMENT 4 5379e:1d NOVEMBER, 1988 I

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3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, MD SYSTEMS l

Certain structures, components, and systems of the reactor system are l

s important to safety because they:

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a. Assure the integrity of the reactor coolant pressure boundary,

b. Assure the capability to shut down the reactor and maintain it in a safe condition.

c. Assure the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10CFR 100.

d. Contain or may contain radioactive material. ,

1 The purpose of this section is to classify structures, systems, and components according to the importance of the item in order to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. Table 3.2-1 of RESAR-SP/90 PDA Module 7, "Structural / Equipment Design', delineates each of the items in the plant which fall under the above-mentioned categories and the respective associated classification that the NRC, MS and industrial codes coernittees have developed. Each of the classification categories in Table 3.2-1 is addressed in the following sections.

1 The classification of specific piping runs and valves in these runs is pro-vided in the system flow diagrams contained in this module. Instrumentation and electrical equipment required to shutdown the plant or mitigate an acci-dont which is associated with the reactor system will be classified as IE (or

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! Safety Class 3 per MS 51.1) and identified in the appropriate module.

4 3.2.1 Seismic Classification Seismic classification criteria are set forth in 10CFR 100 and supplementee by Regulatory Guide 1.29.

WAPWR-RS 3.2-1 MENDMENT 4 5379e:1d NOVEMBER, 1988

All components classified as Safety Class 1, 2, or 3 (classifications are as defined by Reference 1), are seismic Category 1.

1 Seismic Category I structures, components, and systems are designed to with-stand the Safe Shutdown Earthquake (SSE) and other applicable lead combina-tiens, as discussed in RESAR-SP/90 PDA Module 7, "Structural / Equipment gj Design". Seismic Category I structures are sufficiently isolated or protected I from the other structures to ensure that their integrity h maintained.

3.2.2 System Ovality Group Classification Gl l l The Quality Assurance Program described in Subsection 17.1 is applied to all l l 4 Safety Class 1, Safety Class 2 and Safety Class 3 structures, systems and corponent s.

l The components are classified according to their importance to safety, as I dictated by service and functional requirements and by the consequences of j their failure. The quality assurance requirements and code requiree4nts for the reactor system meet the intent of Regulatory Guide 1.26.  ;

3.2.3 Safety Classes Table 3.2-1 lists the safety class assigned to applicable systems and com-ponents in accordance with MS 51.1 (Reference 1). The criteria (of Reference

1) are used in the plant design to provide an added degree of assurance that the plant is designed, constructed, and operated without undue risk to the health and safety of the public, 3.2.4 References

1. ' Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", MS 51.1, November 1983.

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l O WAFWR RS 3.2-2 AuENDMENT 4 i

3375e:1d NOVEMSER, 1983 l

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TABLE 3.2-1 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMP 0NENIS FOR THE REACTOR SYSTEM Principal Quality Safety Code Construction Seismic Quality System / Component Location Group Class Classification Codes & Stds Category Assurance Reactor Vessel Internals Int ated Head ac age Cooling Shroud Missile Shield Lif t Rod Assedly Lift Rig Asscably (See Table 3.2-1 of RESAR-SP/90 PDA Cable Bridge Assembly 16dula 7, "Structure / Equipment Design")

, Cable Assembly 4 Seismic Support System

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CRDM Control Rods G ay Rods Housing j Displacee Rod Drive

) Mechanism Housing j Reactor Core fuel assembly Water displacer i Rod assembly

, Gray rod assembly

Control rod assembly i

k WAPkit-RS 3.2-3 AMEN 06 TENT 4 B379e:Id D EMER, 1968

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. i a 6 O 3.9 MECHANICAL SYSTEMS AND COMPONENTS l

3.9.2 Dynamic Testing and Analysis 3.9.2.3 Dynamic 0 sponst Analysis of Reactor Internals Under Operational l Flow Trr .ients and Steady-State Conditions l Flow Induced Vibration (FIV) and resulting wear have been recognized throughout the SP/90 design ef fort as important issues in reactor internals i design. In f act, consideration of potentially damaging ef fects of FIV have l

had a significant influence on the basic reactor internals design as evidenced I by the selection of an upper calandria configuration. The following sections outline offorts performed in specific areas. I l l i

o ACSTIC Evaluations ACSTIC is a Westinghouse Proprietary mainf rame computer code written l

in FORTRAN which is used to determine the characteristics Cf the l standing waves and wJve propagation in the primary system reactor 4!

j coolant loop. Given a model of nodes and flow paths, the coolant l naturai frequencies can be determined. The loop response at the characteristic reactor coolant pump excitation frequencies can also be Wettferined. Timetmating pressure gradients across structural components can thei be estimated. An evaluation of various fluctuating pressure affects on the reactor internals can ther be made.

1 An ACSTIC analysis performed for the SP/90 reactor coolant system indicates the following primary frequencies:

I f = 7 Hz  !

f = 10 Hz f f 3= 13 HZ f4 = 18 HZ j f, . 23 Hz  :

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These frequencies are disparate from the expected pump shaft rotational frequencies, and consequently the pump rotational induced response in the reactor coolant loop is relatively small.

o Upper Core Plate Axial Vibration The upper core plate axial vibratory response to turbulence excit lion has been analyzed. A reasonable pressure fluctuation was determined f rom a sv.'vey of available data and the response calculated using structural parameters developed in a finite element analysis. The resulting calculated petk fluctuating displacement has been deternined to be acceptable from a fatigue and wear viewpoint.

o Lower Core Plats Axial Vibration The lower core plate axial vibratory response to turbulence excitation 4 nas been analyzed in a similar manner. A reasonable pressure fluctuation was detet.ained from a survey of available data and the response calculated using structural parameters deve'soped in a finite element analysis. The resulting calculated peak fluctuating displ:cem:nt h:s been deter: tined to M scceptsbic f rz. & fatig66 a rid fuel assembly lift-off viewpoint.

o Core Barrel Vibration The important core barrel vibrations consist of the cantilever beam mode and the lower numbered shell modts. These have been traditionally of interest in the internals design. The SP/g0 internals design also incorporates an inner barrel that is in close proximity to the inside of the cor- barrel for the upper two thir- of the core barrel lenstii. This configuration introduces some impt . ant barrel interactions that are not present in past Westinghouse designs.

The core barrel cantilever mode has been investigated using a P .ial simplified configuration of the Reactor Equipment System Modo SM)

WAPWR-RS 3.9-la O

AMENDMENT 4 Tille:Id NOVEMBER,1988

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O finite element model. Normal operation and hot functional test lateral system modes have been calculated for var'ous support conditions at the upper core plate pin interface between the inner O. barrel and the core barrel and the lower radial suppo t interface l between the core barrel and the vessel. l The lower racial support normal operation vibratory loads wera ,

determined using the above RESM model and an appropriate downcomer )

turbulence forcing function. The resulting peak vibratory loads on the lower radial supports are dependent upon the boundary conditions ,

at the upper core plate pin and lower radial support interfaces.

Finite element calculations were performed to investigate these effects.

The core barrel shell modes intere:t with the inner barrel shell modes I

through the hydraulic coupling in the gap between the barrels. This  !

interaction adds substantial hydrodynamic mass to both the inner barrel and the core barrel. The amount of tateraction or coupling for a particular configuration is dependent upon the mode numbers of the two cylinders. For the circumferential modes, only modes with the same number can couple.

f For the axial modes, specific relationships involving cylinder lengths, axial mode number, and end boundary conditions must bt met

, before complete coupling can occur. For modes that do not meet the

! specified conditions only partial coupling occurs. When the coupling is nearly zero, the cylinders vibrate independently. For that case, the of fective hydrodynamic mass is equivalent to that resulting when the other cylinder is rigid.

1 For the case of the partial coupling, one of the coupled frequencies 16 lower than either of the above adjacent rigid wall frequencies, and the other coupled frecuency is higher than either of the above

adjacent rigid wall frequencies. For the SP/go, the lower mode shell j frequencies then fall in the 15 Hz to 25 Hz range.

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e o Rod Guide Vibration The reactivity control cluster (RCC) and water displacer rod cluster (WDRC) rod guide designs are arranged 4tfferently from existing Westinghouse PWR ciesigns . The SP/90 is a close packed arrangement with a relatively small. 0.24 inch, nominal gap between adjacent rod guide enclosure walls. This arrangen 't introduces considerable hydrodynamic lateral coupling between the vibratory responses of the rod guides.

An analytical fluid /ttrut6ure interaction study was performed on a st,bs et of the rod guide region. These studies indicated that the water coupled natural f recuencies of the rod guides are considerably lower than the in-air natural frequencies. Also, complex interactions in the norizontal plane exist between the rod guides which result in 4 many system frequencies. The in-air natural f requency of both the RCC and the WDRC rod guide is approximately 36 Hz. The coupled system frequencies are as low as 7.5 Hz.

The fluid / structure interaction studies have been used to develop WDR and RCC rod guide models for use in the RESM system model.

The RCC and WDRC rod guides are bolted to the core plate at the bottom. The upper lateral support for these guides has been an area of extensive investigation. An important parameter af fecting the upper support design is the flow induced vibratory reaction load.

Calculations have been performed which address the expected FIV reaction 104/s. Considerable uncertainty exists concerning the exact excitation mechanism of previously measured internal responses.

Consequently, relatively conservative assumptions were made in perforning the above cited calculations. This resulted in calculated l reaction loads that are conservatively high, but which have been shown l to be acceptable for the design of the rod guide top end support.

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MAPWR-RS 1117P:1d 3.9-ic AMENDMENT 4 e

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o RCC Vibration and Wear Significant ef forts were expended to demonstrate a RCC war life of 20 O years.

r RCC rodlet motions as a result of flow induced vibration were measured in a full scale RCC Axial Flow Test factitty. This facility incorporated the capability to adjust the intermediate support i locations in the RCC rod guide to achieve the maximum RCC rodlet motion. Separately, impact-f retting wear coef ficients were measured

' under simulated reactor conditions of pressure, temperature, and chemistry. With these data, it was possible to calculate wear as a result of flow induced vibration. ,

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! Sliding wear as a result of CRDM stepping was measured in the r Westinghouse 0-loop f acility which contained a complete full-scope RCC  !

(Reactivity Control Cluster) driveline and associated equipment. The

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driveline included the RCC itself, drive rod and couplings, and the  !

l CRDM. Associated equipment included the pressure vessel, simulated lower internals with a prototype fuel assembly, and simulated upper l internals with a prototypic RCC rod guide and an upper calandria l i mock-up. The tests were conducted at reactor conditions of pressure, j temperature and chemistry. The CROM was exercised for 7.8 million i steps in a pattern which included simulated baseload operation, load l i

i follow, f requency control and rod drops.

Combining the ficw induced vibration and stepping wear led to the contiusion that the RCC's would be able to operate well in excess of l

' 20 years without clad wear through.

o Scale Model Testing In addition to the above described analytical activities, a one-fif th scale test of the full SP/90 reactor was performed. This was a low O

WAPWR-R$ 3.g -1 d AMENDMENT 4 T177e:1d NOVEMBER,1988

pressure, low temperature test eerb *as executed in 198s and 1987 at the Takasago Research and Develsyncrc Center of Mitsubishi Heavy Industries in Japan.

The test incorporated three phases .ns folhws:

(1) Vibration characteristics test Datural frequencies, vibration snodes and damping ratios) in atrr and in water. ,

(ii) Pressure loss test.

(iii) Flow induced vibration test Main conclusions f rom the latter phase of JestJAg are:

4 (i) The bottom inounted instrumentation, calandria tubes, and flow shrouds were the smly components that showed vibration near their natural frequencies; however, the level of vibration was low. .

(ii) Thr retpense of the other reactor internals showed random vibration in a frequency range of 0-1000 Hz, and neither structural resonance mrr vnstable vibration (e.g. of the hydroelastic type) ens observed.

In general, acceleration and stress levels were significantly below i values that would cause f atigue rencerns.

3.9.2.4 Preoperational Flow-Induced Vibration Testing of Reactor Internals Every $P/90 plant to be built will undergo a Act functional testing program.

A part of that program is devotet to assuring the structural integrity of the reactor internals by the successful completion of a full flow, high I temperature, high pressure test, Additionally, the first SP/90 plant to l O

WAPWR-RS 3 9-le AMENDMENT 4 T177e:1d NOVEMBER, 1988 l

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o become operational will also be u.assified as a "prototype" according to Regulatory Guide 1.20 of the United States Nuclear Regulatory Comission.

Thus, that plant will also need to be instrumented for the monitoring of structural vibration responses during preoperational testing.

A scoping study addressing this general issue has been undertaken as a part of the SP/90 reactor internals design process. The main conclusion o' the study i is that preoperational vibrational assessment testing of the present reactor internals design is f**.:ible with only detailed hardware changes needed for transducer nunting, transducer protection, and transducer lead routing.

Considerattel was also given to the possible need to perform the preoperational testing with a dumy core in place. The conclusion of the study is the t an adequate preoperational vibrational assessment test program could be conducted without the presence of a dumy core.

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O 3.9.2.5 Dynamic System Analysis of the Reactor Internals Under Faulted Conditions Analysts of the reactor internals for loads resulting from postulated pipe breaks which result in loss-of-coolant accidents (LOCAs) are typically based on the time-history response of the reactor internals to hydraulic forcing functions applied simultaneously. The forcing functions are defined at points in the system wnere changes in cross-section or direction of flow may occur such that differential loads may be generated as a consequence of the pipe break (s). Because of the complexity of the system and the components. it may be necessary to use finite element stress analysis codes to provide not e detatted information at various points.

(*) See RESAR-5P/90 PDA Module 7 ' Structural / Equipment Design

• for the detailed application of Westinghouse revised pipe break criteria to the MAPWR design. 1 1

EAPWR-R$

Ill7e:1d 3.9-2 AMEN 0 MENT 4 NOVEMBER, 1988 ei  ;

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17.0 QUALITY ASSURANCE 17.1 Quality Assurance During Design and construction

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i The Westinghouse Energy Systems Business Unit / Nuclear Fuel Business Unit  !

Quality Assurance Program is described in Reference 1. I i

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! 17.1.1 References  ;

i l 1. "Westinghouse Energy Systems Business Unit / Nuclear Fuel Business Unit  !

Quality Assurance Plan," WCAP-8370, Revision 11, October 1988. {

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l WADWR RS 17.1-1 AWENDu.!NT 4 3379e:1d NOVEMBER, 1988 l i f

<s REQUEST FOR ADDITIONAL INFORMATION WESTINGHOUSE ADVANCED PRESSURIZED WATER REACTOR (RESAR-SP/90)

DOCKET NO. 50-601 O

The following Questions / Responses were fortrally transmitted in Addendum 1 to RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3304, dated January 7, 1988, 252.1 Verify that the aging and tempering temperatures of heat treatable materials used in the control rod drive mechanisms are specified to eliminate susceptibilit to stress corrosion cracking in reactor coolant (4.5.1, Wodule 5)y

Response

The CRDH heat treatable materials are 410 SST tubing for the drive rod. 403 modified SST bar for the coupling, and Inconel X-750 spring wire. The 410 SST and 403 SST are tempered to provide minimum yield strengths of 80 XS! and 90 KS! respectively. This tempering is well O ,

below the threshold where susceptibility to stress corrosion cracking (SCC) becomes a concern -

yields greater than 120 KSI. The Inconel X-750 is manufactured to MIL-5 23192 which offers the most favorable conditions for precluding SCC.

252.2 What materials, other than austenitic stainless steels of limited cold ork (maximum yield stren of 90 ksi) are used for reactor internals? (4.5.2, Module 5) gth

Response

O Other materials used for reactor internals are: 1) Inconel X 750, for guide tube support pins and guide tube flexures (where applicable) and; 2) Stellite Hardfacing, for the radial support keys. Stellite Hardfacing is principally co6 posed of Cobalt (Co) Chromium (Cr) and O Tungsten (W).

O WAPWR-RS A4-1 AMENDHENT 4 5379e:1d NOVEMBER, 1988

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- l O The following Questions / Responses were formally transmitted in Addendum 5 to j

RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3338, dated May 13, 1988.  ;

O 440.2 How is the Improved Thermal Design Procedure (ITDP) factored in the 2% power as well as the allowances on pressure e.nd temperature 1 l

RESPONSE

O The Improved Thermal Design Procedure (ITDP) was used for most

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DNB related transients. Consistent with the methodology l presented in WCAP-8567, Reference 3 in Section 4.4 of l RESAR-SP/90 PDA Nodule 5 "Reactor System," allowances for l power, pressure, temperature and flow are included. These l

uncertainties were calculated specifically for the APWR design.  !

The following Questions / Responses were formally transmitted in Addendum 6 to i RESAR-SP/90 PDA in Westinghouse letter NS-NRC-88-3354, dated July 7, 1988.

210.29 The staff's conenents in 0210.25 also apply to portions of

{ Section 3.2 and Table 3.2-1 of Module 5. These sections should 1 be revised to agree with the response to 0210.35.

1 j RESPONSE:

i l Please refer to our original response to Staff 0210.1.

! Westinghouse believes that the initiative taken to design the

! $P/90 plant to the latest industry codes and standards, includ-l ing ANS!/ANS 51.1, provides additional assurance that this plant l design will operate more safety and with better reliability than 1

current nuclear power plant designs. If this issue is not

settled prior to final design submittal, Westinghouse will l reexamine the manner in which Safety classifications are l assigned for systems, conponents, and structures for the SP/90 plant.

O WAPWR-RS A4 2 AMENDMENT 4 3379e:1d NOVEM5ER,1968

G 210.30 Section 1.5.1.3 of Module 5 briefly discusses proposed tests of the Cortrol Rod Drive Mechanism (CRDM) and the Water Displacer Rod Ne hanism. These tests were scheduled to be conducted in 1985 anc' 1985. Provide a detailed description of the test program and a summary of the results. If applicable, provide a O comparison of tests which were conducted on existing Westinghouse CRDM's with the WAPWR CRDMS's.

RESPONSE

Following is additional information on Control Red Drive Mechanism (CRDM)andWaterDisplacerRod Drive Nechanism (DRDM) testing.

CRDM 1 The CRDM test program was performed at the D-Loop test facility at the Westinghouse Forest Hills site. Previous CRDM testing has also taken place in this facility, i

D-Loep i. a high temperature, high flow rate test facility which l can test full size corponents under simulated conditions of

( chemistry, temperature, pressure and flow. The test section is

) isothermal; no heat is generated and only depleted nuclear futi t

is used. The loop piping is designed for flow rates up to 4500 l gpm, a maximum operating pressure of 2000 psig, and a temperature of 600'F. The loop flow rate is measured using a l square-edge orifice plate and is adjustable by means of flow l control valves in the main loop piping. The flow through the model is measured using four venturi flow meters built into the lower core plate. At 4000 gpm, the canned motor pump is capable I

of developing a head of 300 feet of water. In actual practice.

l the maximum loop flow attainable with the APWR model was approximately 2800 gpm.

O WAPWR-RS A4-3 AMENDMENT 4 5379e:1d NOVEMBER, 1958

All piping in the primary loop is Type 304 and Type 316 stainless steel. Loop pressure is automatically controlled by a constantly operating makeup and letdown system. Makeup is maintained by 3 Aldrich positive displacement pumps. Letdown is O controlled by Grove Nitymite regulators. A rupture disk, set to relieve at 2400 psia, is provided for overpressure protection.

Loop temperature during steady-state operation is maintained by O controlling bypass flow through the loop coolers.

13e KW of heat input is available for startup and for A total of steady-state operation through strip heaters mounted on the loop piping. The heaters are controlled by monitoring loop piping thermocouples and automatically regulating cycle timing of the respective heaters. An additional 235 KW is obtained from the main circulation pump at high flow rates. A Pan-Alarm system provides audible and visual indication of potential system malfunctions.

The D-Loop Facility also includes a cooling air system for the Control Rod Drive Mechanism (CRDM). The CRDM is an electro-mechanically operated device which relies on forced air cooling to maintain the magnet coils at a safe operating

]

temperature. The cooling system consists of a 1000 cfm

centrifugal blower, a throttling damper, an airflow measurement section, and a full length cooling baffle. Thermocouples were installed to permit calculation of total heat rejection.

O The D-Loop test facility contained a complete full-scale RCC l (Reactivity Control Cluster) drivaline and associated equipment.

The driveline included ' m RCC itself, drive rod and couplings,

and the CRDN. Associated equipment included the pressure vessel, simulated lovi,r internals with a prototype fuel

]

d assembly, and simulated upper internals with a prototypic RCC rod guide and an ;pper calandria mock-up. An additional vessel

!O i

j WAPWR-RS A4-4 AMENDHENT 4 B379e:1d NOVEM5ER, 1988 a

spool piece was required to acconnodate the extra upper internals length caused by the calandria.

The CRDH was exercised for 7.8 million steps in a pattern which included simulated baseload operation, load follow, frequency control and rod drops.

Throughout the test period of approximately 6 months, the CROM operated without problems. Nessured heat rejection was similar to that found on previous D-Loop tests. Post-test inspection revealed that from a wear point of view, the latch arms could probai.*y have operated up to approximately 10 million steps.

As noted previously. CRDM's have been tested in D-L:op in the past. However, maxir.um stepping duty in previous testing has not exceeded 3.5 to 4.0 million steps, primary because load follow operation and frequency control were not considered.

DRDM The DRDM prototype hydraulic test program was performed at the M-Loop Test Facility at Westinghouse Electro-Mechanical Division located in Cheswick, PA. This loop has traditionally served as the production test facility for Control Rod Drive Mechanism and was modified to ' provide the adequate pressure control and make-up water capacity required by the DRDN in order to operate.

The loop pressure source is a 120 gallon pressurizer with a maximum available heat input of 80 KW. Pressure control settings ar e variable up to 2500 psig. The loop is capable of temperatures up to 650'F, by heat input as provided by O electrical resistance strip heaters rated at 103.5 KW. A Westinghouse Model 150-0 canned motor puep maintains loop water flow through the heaters. The high pressurt makeup pump which O

WAPWR RS A4 5 AMEN 0u!NT 4 5379e:Id NOVEM3ER, 1955

I I

maintains liquid level in the pressurizer during DRDM withdrawal, is a variable speed triplex plunger pur.p capable of flows to 9.2 GPM at 2500 psig. The control for this pump is  !

from the pressurizer liquid level controller.

The facility is designed in accordance with the ASME Boiler & l 1

Pressure Vessel Ccde and is licensed to operate as a i Pennsylvania Special Boiler (Pa. Spc. 3020) in the state of j Pennsylvania.

The DRDM test unit consisted of five major subassemblies: the pressure housing, the internal cylinder, the drive rod, the weight assembly ar.d the rod position indicator / cooling box.

1) The DRDM perfortned in ar. acceptable manner for the duration i i

cf the test. No withdrawal or insertion failures occurred.

2) No significant wear was found on the latch and spear; j j tentact marks were visible, but no measurable material l l re

noval took place.

l

t

3) h significant wear was found on the cylinder with the ID i bore unchanged. The ID surface showed light scratch marks  ;

f which are considered normal for this application.  ;

I I

L l

4) Piston ricq amar of the radial wall thickness and the re-sulting charges in the ring end gap and spring tension were O

such that the rings are considered to be worn out. This was expected considering the duty imposed on these rings.

l p 5) Addition of crud to the system did not result in noticeable i

(/ degradation of DROM operating performance. i I i l I

6) Heat rejection from the DRDH was higher than calculated. l I

! WADWR-RS A4 6 AWENOWINT 4 i i 537Se:1d NOVEMBER, 195:

l t

.e O The DRDN was also tested in the Hot Single Channel Test #2 in Japan, which was funded by the Japanese government. The results of this test have not yet been made public. They are expected to be available at the FDA stage.

O The DRDM is a first-of a-kind component, therefore, no comparison to previous tests can be made.

210.31 In Section 3.9.2.4 of Module 5, it is stated that the reconnendations of Wegulatory Guide 1.20 will be satisfied by conducting examinations of the reactor internals both before and after confirmatory hot functional testing of the internals.

Based on the staff's understanding of the WAPWR reactor internals design, this is not an acceptable com.itment. Section 3.9.5 of Nodule 5 describes a design which is "significantly different from existing Westinghouse designs.' The prototype plant for existing Westinghouse four loep plants is Indian Point Unit 2. The data from the Indian Point 2 reactor internals

, verification test program has been supplemented by data from J

tests conducted at the Trojan and Sequoyah plants. This i supplemental data was provided at the staff's request to verify that design changes to reactor internals in Westinghouse four i loop plants subsequent to the Indian Point 2 design did not j result in a significant difference from the Indian Point 2

verification data.

1 It is not apparent to the staff that the WAPWR reactor internals l response to flow induced vibration is enveloped by the above

prototype verification data for four loop plants. Therefore, the staff will require a commitment that the first WAPWR plant

, will be identified as the prototype plant and will meet all of i

the applicable Regulatory Guide 1.20 guidelines. Revise Section l 3.9.2.4 in Module 5 to provide this coesitment or provide

justification for not doing so.

O RESPONSE:

i j in response to this question, Subsection 3.9.2.4 has been modified.

! 210.32 As stated in 0210.31, the staff position is that the first WAPWR plant be designated as the prototype as defined in REG Guide

1.20. The information in Section 3.9.2.3 of Module 5 relative lO l

l WAPWR-RS A4-7 AMINDMINT 4

{ 5379e:1d NOVEMBER. 1963 i

i

1 i

! to the dynamic response analysis of reactor internals under 1 operational flow transients and steady-state conditions is not

acceptable for a prototype plant. Revise this section to be 4

consistent with the guidelines in Standard Review Plan, Sections

! 3.9.2.1.3, 3.9.2.!!.3 and 3.9.2.!!!.3.

i O RESPONSE:

In response to this question. Subsection 3.9.2.3 of RESAR SP/90 i PDA Module 5. "Reactor System' has been revised to reflect our comitment to meeting the staff guidelines as outlined in $RP l Section 3.9.2.

l f 210.33 Section 3.9.5.1.3.4. of the Nodule 5 discusses the bottom

, mounted instrumentation (BMI) thimbles. A problem of

- unacceptable accelerated wear of Westinghouse designed BMI l thimbles in a European 14 foot core plant was identified in 1985. Subsequently. Westinghouse modtfied the thimble design to reduce flow velocity in the gap between the thimble anc the BMI j column. However, this modification did not resolve the problem.

, but instead increased the rate of wear. The same modified

' design has been incorporated into the South Texas, Unit 2

! pressure vessel which also contains a 14-foot core. Since this j potential problem could be applicable to the WAPWR, provide a

! comitment in Section 3.9.5.1.3.4 that the final resolution of this problem for South Tomas and the European plants will be incorporated into the WAPWR design, or provide justification for not doing so. A failure of one or more of these thimbles could i result 'n a small break in the reactor coolant pressure boundary l which cannot be isolated.

RESPONSE

The $P/90 and South Texas designs are different in the area of botton mounted "instrumentation (BMI) such that it may not be l

necessary to incorporate the final resolution of the current BMI l

problems into the APWR. Prior to FDA submittal the APWR BMI i design will be reviewed to determine if design changes are warranted.

l l

O WAPWR-R$ A4 B AMENONENT 4 5379e:1d NOVEMBER, 1988