ML20116E278

Nonproprietary Amend 2 to RESAR-SP/90 Module 5, Reactor Sys

ML20116E278
Person / Time
Site:	05000601
Issue date:	04/30/1985
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19269B453	List: ML20116E269 ML20116E278 NRC-85-3030, Forwards Proprietary & Nonproprietary Amend 2 to Westinghouse Advanced PWR RESAR-SP/90 Module 5, Reactor Sys, Section 4.5, Reactor Matls. Ltr Requesting Withholding Encl.Proprietary Rept Withheld (Ref 10CFR2.790)
References
NUDOCS 8504300244
Download: ML20116E278 (17)

Text

. .. . - . . . . . - - . . - . . . . - . - . - - . . _ - . - - . . - . - - . _ _ - - . . . . - . - - _ . . . - . . - - - - - .

i-I v.

1 AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 5, " REACTOR SYSTEM" 4

s 1

f i l l l

\

i f

i f

j i

i

(

l i

L@

I

(

l.

1 i

i l@

l l

l I

W3ggf g l@

! AMENDMENT 2 l MAPWR-RS APRIL, 1985

1476e

1d i

4

.-..- -.-~. -. .__ ......- . - - -_- -. -.-_ _ -

N I

I i

AMENDMENT 2 TO RESAR-SP/90 PDA MODULE 5, " REACTOR SYSTEM" l

I INSTRUCTION SHEET O .

Remove Insert

, _ - = -

Pgs xii thru xx . - -~~~' Pgs xii thru xx Pgs 4.5-1 thru 4.5-3 Pgs 4.5-1 thru 4.5-6 6

O O

O WAPWR-RS AMENDMENT 2 1476e:1d APRIL,1985

s ,

p, t

O TABLE OF CONTENTS (cont) l Reference

g- ' SAR Section Section Title Pigg Status l

4.4.3.2 Operating Restrictions on Pumps 4.4-31 I 1 4.4.3.3 Power-Flow Operating Map (BWR) 4.4-31 I

, O 4.4.3.4 Temperature - Power Operating Map Load Following Characteristics 4.4-31 I 4.4.3.5 4.4-32 I i

4.4.3.6 Thermal and Hydraulic Characteristics 4.4-32 I I

Summary Table 4.4.4 Evaluation 4.4-32 I 4.L.4.1 Critical Heat Flux 4.4-32 I i

4.4.4.2 Core Hydraulics 4.4-32 I 4.4.4.2.1 Flow Paths Considered in Core Pressure 4.4-32 I Drop and Thermal Design O 4.4.4.2.2 4.4.4.2.3 Inlet Flow Distributions 4.4-33 I Empirical Friction Factor Correlations 4.4-34 I 4.4.4.3 Influence of Power Distribution 4.4-35 1 4.4.4.3.1 Nuclear Enthalpy Rise Hot Channel 4.4-35 I Factor, F H

4.4.4.3.2 Axial Heat Flux Distributions 4.4-37 I 4.4.4.4 Core Thernal Response 4.4-38 I 4.4.4.5 Analytical Techniques 4.4-39 I 4.4.4.5.1 Core Analysis 4.4-39 I

() 4.4.4.5.2 4.4.4.5.3 Steady-State Analysis Experimental Verification 4.4-39 4.4-40 I

I 4.4.4.5.4 Transient Analysis 4.4-40 I L

4.4.4.6 Hydrodynamic and Flow Power Coupled 4.4-41 I

() 4.4.4.7 Instability Fuel Rod Behavior Effects f rom Coolant 4.4-44 I

Flow Blockage t

4.4.5 Testing and Verification 4.4-46 I 4

O WAPWR-RS xit JULY, 1984 1476e:ld w-, ow,~-- wn, wn, v

s TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 4.4.5.1 Tests Prior to Initial Criticality 4 ,o I 4.4.5.2 Initial Power and Plant Operation 4.4-46 1 4.4.5.3 Component and Fuel Inspections 4.4-46 I 4.4.6 Instrumentation Requirements 4.4-46 I 4.4.6.1 Incore Instrumentation 4.4-47 I 4.4.6.2 Overtemperature and Overpower AT 4.4-47 I Instrumentation 4.4.6.3 Instrumentation to Limit Maximum Power 4.4-48 Output I 4.4.7 References 4.4-49 I 4.5 REACTOR MATERIALS 4. 5-1 1 4.5.1 Drive System Structural Materials 4. 5-1 I 4.5.1.1 Control Rod and Gray Rod Drive 4. 5-1 I System Structural Materials 4.5.1.1.1 Control Rod Drive Mechanism (CROM) 4. 5-1 I and Gray Rod Drive Mechanism (GRDM)

Materials Specifications 4 . 5 .1.1. 2 Fabrication and Processing of Austenitic 4.5-3 I Stainless Steel Components 2 4.5.1.1.3 Contamination Protection and Cleaning of 4.5-3 I Austenitic Stainless Steel 4.5.1.1.4 Other Materials 4.5-3 I 4.5.1.2 Displacer Rod Drive System Structural 4.5-4 I Materials 4.5.1.2.1 Materials Specifications 4.5-4 I 4.5.1.2.2 Fabrication and Processing of Austenitic 4.5-5 I Stainless Steel Components 4.5.1.2.3 Contamination Protection and Cleaning 4.5-5 I l of Austenitic Stainless Steel WAPWR-RS xiii AMENDMENT 2 1476e:1d APRIL, 1985

e

? . .

lY TA8LE OF CONTENTS (Cont)

Reference SAR Section d Section Title Page Status 1 4.5.2 Reactor Internals Materials 4.5-5 I 4.5.2.1 Materials Specifications 4.5-5 I -

, (,/ 4.5.2.2 Controls on Welding 4.5-6 I 4.5.2.3 Nondestructive Examination of Tubular 4.5-6 I 2 N Products and Fittings 4.5.2.4 Fabrication and Processing of Austenitic 4.5-6 I Stainless Steel Components 4.5.2.5 Contamination Protection and Cleaning 4.5-6 I of Austenitic Stainless Steel 4.6 FUNCTIONAL DESIGN OF REACTIVITY CONTROL 4 . 6-1 1 SYSTEMS P

4.6.1 Information for the Control Rod Drive 4 . 6-1 I System (CRDS) 4.6.2 Evaluation of the CRDS 4.6-1 I 4.6.3 Testing and Verification of the CRDS 4.6-2 I 4.6.4 Information for Combined Performance 4.6-2 I of Reactivity Systems 4.6.5 Evaluation of Combined Performance 4.6-3 1 4.6.6 References 4.6-4 I 5.0 REACTOR COOLANT SYSTEM AND CONNECTED 5.0-1 NA k SYSTEMS 6.0 ENGINEERED SAFETY FEATURES 4 6.0-1 NA 7.0 INSTRUMENTATION AND CONTROLS 7.0-1 NA 8.0 ELECTRIC POWER

, 8. 0-1 i NA Ox 9.0 AUXILIARY SYSTEMS 9. 0-1 NA 10.0 STEAM AND POWER CONVERSION SYSTEM 10.0-1 NA 11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.0-1 NA 12.0 RADIATION PROTECTION 12.0-1 ,

NA v 13.0 CONDUCT OF OPERATIONS 13.0-1 NA MAPWR-RS xiv AMENDMENT 2 1476e:ld APRIL, 1985

4 TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 14.0 INITIAL TEST PROGRAM 14.0-1 NA 15.0 ACCIDENT ANALYSES 15.0-1 II 15.0.1 General 15.0-1 II 15.0.2 Classification of Plant Conditions 15.0-1 II 15.0.2.1 Condition I - Normal Operation and 15.0-2 II Operational Transients 15.0.2.2 Condition II - Faults of Moderate 15.0-4 II Frquency 15.0.2.3 Condition III - Infrequent Faults 15.0-6 II 15.0.2.4 Condition IV - Limiting Faults 15.0-7 II 15.0.3 Optimization of Control Systems 15.0-8 II 15.0.4 Plant Characteristics and Initial 15.0-9 II Conditions Assumed in the Accident Analyses 15.0.4.1 Design Plant Conditions 15.0-9 II 15.0.4.2 Initial Conditions 15.0-9 II 15.0.4.3 Power Distribution 15.0-10 II 15.0.5 Reactivity Coefficients Assumed 15.0-11 II in the Accident Analyses 15.0.6 Rod Cluster Control Assembly 15.0-12 II Insertion Characteristics 15.0.7 Trip Points and Time Delays 15.0-13 II to Trip Assumed in Accident Analyses 15.0.8 Instrumentation Drift and Calorimetric 15.0-14 II Errors - Power Range Neutron Flux 15.0.9 Plant Systems and Components Available 15.0-15 II for Mitigation of Accident Effects O

HAPWR-RS xv AMENDMENT 2 1476e:1d APRIL, 1985

.. . . . - =

a' l l

,  %- l l

TABLE OF CONTENTS (cont)

Reference SAR Section O. Section Title Page Status 15.0.10 Fission Product Inventories 15.0-16 11 p 15.0.10.1 Inventory in the Core 15.0-16 II d 15.0.10.2 Inventory in the Fuel Pellet 15.0-16 II Clad Gap 15.0.10.3 Inventory in the Reactor Coolant 15.0-16 II 15.0.11 Residual Decay Heat 15.0-17 II 15.0.11.1 Total Residual Heat 15.0-17 II 15.0.12 Computer Codes Utilized 15.0-17 II 15.0.12.1 FACTRAN 15.0-17 II 15.0.12.2 LOFTRAN 15.0-18 II

~

15.0.12.3 O

v 15.0.12.4 TWINKLE THINC 15.0-19 15.0-19 II II 15.0.13 References 15.0-19 II

,15.4 REACTIVITY AND POWER DISTRIBUTION 15.4-1 II ANOMALIES 15.4.1 Uncontrolled Rod Cluster Control Assembly 15.4-2 I Bank Withdrawal f rom a Subcritical or Low-Power Startup Condition 15.4.1.1 Identification of Causes and Accident 15.4-2 I Description 15.4.1.2 Analysis of Effects and Consequences 15.4-4 I 15.4.1.2.1 Method of Analysis 15.4-4 I i

15.4.1.2.2 Results 15.4-7 I 1

, 15.4.1.3 Conclusions 15.4-7 1 15.4.2 Uncontrolled Rod Cluster Control Assembly 15.4-8 I Bank Withdrawal at Power 4

15.4.2.1 Identification of Causes and Accident 15.4-8 I Description WAPWR-RS xvi AMENDMENT 2 1476e:1d APRIL, 1985

TABLE OF CONTENTS (cont)

Reference SAR Section Section Title Page Status 15.4.2.2 Analysis of Effects and Consequences 15.4-10 I 15.4.2.2.1 Method of Analysis 15.4-10 I 15.4.2.2.2 Results 15.4-12 1 15.4.2.3 Radiological Consequences 15.4-14 I 15.4.2.4 Conclusions 15.4-14 1 15.4.3 Rod Cluster Control Assembly Misoperation 15.4-15 II (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident 15.4-15 II Description 15.4.3.2 Analysis of Effects and Consequences 15.4-17 II 15.4.3.2.1 Method of Analysis for Dropped or 15.4-18 II Misaligned RCCA 15.4.3.2.2 Statically Misaligned R';CA Results 15.4-18 I 15.4.3.2.3 Single RCCA Withdrawal Method of Analysis 15.4-19 I 15.4-20

~

15.4.3.2.4 Single RCCA Withdrawal Results 1 15.4.3.3 Radiological Consequences 15.4-20 I 15.4.3.4 Conclusions 1 5.4-21 I 15.4.7 Inadvertent Loading and Operation of a 15.4-21 I Fuel Assembly in an Improper Position 15.4.7.1 Identification of Causes and Accident 15.4-21 I Description 15.4.7.2 Analysis of Effects and Consequences 15.4-22 I 15.4.7.2.1 Method of Analysis 15.4-22 I 15.4.7.2.2 Results 15.4-23 1 15.4.7.3 Radiological Consequences 15.4-24 I 15.4.7.4 Conclusions 15.4-24 I O

WAPWR-RS xvii AMENDMENT 2 1476e:1d APRIL, 1985

t TABLE OF CONTENTS (cont)

Reference j SAR Section V Section Title Pace Status 1

15.4.8 Spectrum of Rod Cluster Control Assembly 15.4-25 I Ejection Accident 15.4.8.1 Identification of Causes and Accident 15.4-25 I Description 15.4.8.1.1 Design Precautions and Protection 15.4-25 I 15.4.8.1.1.1 Mechanical Dttign 15.4-25 I 15.4.8.1.1.2 Nuclear Design 15.4-26 I ,

15.4.8.1.1.3 Reactor Protection 15.4-27 I 15.4.8.1.1.4 Effects on Adjacent Housings 15.4-27 I 15.4.8.1.1.5 Effects of Rod Travel Housing Longitudinal 15.4-27 I Failures O

Q 15.4.8.1.1.6 Effects of Rod Travel Housing 15.4-28 I Circumferential Failures )

15.4.8.1.1.7 Possible Consequences 15.4-28 I 15.4.8.1.1.8 Summary 15.4-28 I 15.4.8.1.2 Limiting Criteria 15.4-29 I >

15.4.8.2 Analysis of Effects and Consequences 15.4-30 I 15.4.8.2.1 Calculation of Basic Parameters 15.4-32 I 15.4.8.2.1.1 Ejected Rod Worths and Hot Channel 15.4-32 I Factors Ok 15.4.8.2.1.2 Reactivity Feedback Weighting Factors 15.4-33 1 15.4.8.2.1.3 Moderator and Doppler Coefficients 15.4-34 I 15.4.8.2.1.4 Delayed Neutron Fraction, 8,ff M.4-34 I 15.4.8.2.1.5 Trip Reactivity Insertion 15.4-34 I v 415.4.8.2.1.6 Reactor Protection 15.4-35 I 15.4.8.2.1.7 Results 15.4-36 I 15.4.8.2.1.8 Fission Product Release 15.4-39 I 15.4.8.2.1.9 Pressure Surge 15.4-39 I 15.4.8.2.1.10 Lattice Deformation 15.4-39 I WAPWR-RS xviii AMEN 0 MENT 2 1476e:1d APRIL, 1985

TABLE OF CONTENTS (cont)

Reference SAR Section ,

Section

• Title Page Status 15.4.8.3 Radiological Consequences 15.4-40 1 14.4.8.3.1 Analytical Assumptions 15.4-40 1 15.4.8.3.1.1 Source Term Calculations 15.4-40 1 15.4.8.3.1.2 Mathematical Models Used in the Analysis 15.4-41 I 15.4.8.3.1.3 Identification of Leakage Pathways and 15.4-41 I Resultant Leakage Activity 15.4.8.3.2 Identification of Uncertainties and 15.4-42 I Conservative Elements in the Analysis 15.4.8.3.3 Conclusions 15.4-43 1 15.4.8.3.3.1 Filter Loadings 15.4-43 I 15.4.8.3.3.2 Dose to Receptor at the Exclusion Area 15.4-43 I Boundary and Low Population Zone Outer Boundary 15.4.9 References 15.4-44 II 15A ACCIDENT, ANALYSIS RADIOLOGICAL 15.A-1 II CONSEQUENCES EVALUATION MODELS AND PARAMETERS 15A.1 General Accident Parameters 15.A-1 II 15A.2 Offsite Radiological Consequences 15.A-1 II Calculational Models 15A.2.1 Accident Release Pathways 15.A-2 II 15A.2.2 Single-Region Release Model 15.A-2 11 15A.2.3 Two-Region Spray Model in Containment 15.A-4 II (LOCA) 15A.2.4 Offsite Thyroid Dose Calculation Model 15.A-5 II 15A.2.5 Offsite Beta - Skin Dose Calculational 15.A-6 II Model 15A.2.6 Offsit9 Gamma-Body Dose Calculational 15.A-6 II ,

Model WAPWR-RS xix AMEN 0HENT 2 1476e:1d APRIL, 1985 l i

A O TABLE OF CONTENTS (cont)

Reference SAR Section j O Section Title Pace Status 15A.3 Control Room Radiological Consequences 15.A-7 II t Calculational Models 15A.3.1 Integrated Activity in Control Room 15.A-7 II

, -15A.3.2 Integrated Activity Concentration in 15.A-8 II Control Room From Single-Region System

, 15A.3.3 Control Room Thyroid Dose Calculational 15.A-9 11 Model 15A.3.4 Control Room Beta - Skin Dose 15. A-10 II I

Calculational Model

! 15A.3.5 Control Room Gamma - Body Dose 15. A-11 II

, Calculation 15A.3.5.1 Model for Radiological Consequences Due 15. A-11 II to Radioactive Cloud External to the Control Room 15A.4 References 15. A-12 II i 16.0 TECHNICAL SPECIFICATIONS 16.0-1 N/A

17.0 QUALITY ASSURANCE 17.1-1 II 17.1 Quality Assurance Ouring Design and 17.1-1 II

] Construction b 17.1.1 References 17.1-1 II l

io

-t i WAPWR-RS xx AMENDMENT 2 i

1476e

1d APRIL, 1985

Y 4.5 REACTOR MATERIALS 4.5.1 Orive System Structural Materials 4

4.5.1.1 Control Rod and Gray Rod Drive System Structural Materials 2

4.5.1.1.1 Control Rod Drive Mechanism (CRDM) and Gray Rod Drive Mechanism (GRDM) Materials Specifications O All parts _ exposed to reactor coolant are made of metals which resist the corrosive action of the water. Three types of metals are used exclusively:

stainless steels, nickel-chromium-iron, and cobalt-based alloys. In the case of stainless steels, only austenitic and martensitic stainless steels are used. For pressure boundary parts, martensitic stainless steels are not used in the heat-treated conditions which cause susceptibility to stress-corrosion cracking or accelerated corrosion in Westinghouse pressurized water reactor chemistry. Pressure boundary parts and components are made of type 304 stainless steel, or Inconel 600.

Internal latch assembly, drive rod assembly and hub extension assembly parts are f abricated of heat-treated martensitic stainless steel. Heat treatment is such that susceptibility to stress-corrosion cracking is not initiated.

a. CROM/GRDM Pressure Vessel Assembly All pressure retaining materials comply with Section III of the

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code and are fabricated from austenitic (type 304) stainless steel or Inconel 600.

I b. CRDM/GRDM Coil Stack Assembly i

J The coil housings require a magnetic material. Both low carbon cast 1 steel and ductile iron have been successfully tested for this

O WAPWR-RS

_ 4.5-1 AMEN 0 MENT 2 1253e:1d APRIL, 1985

application. The choice, made on the basis of cost, indicates that ductile iron will be specified on the CROMs and GRDMs. The finished housings are zinc flame sprayed to provide corrosion resistance.

Coils are wound on bobbins of glass reinforced silicon thermoset molding material, with double glass insulated copper wire. Coils are then vacuum impregnated with silicon resin. A wrapping of mica sheet is secured to the coil outside diameter. The result is a well insulated coil capable of sustained operation at 200*C.

c. CROM/GROM Latt.h Assembly Magnetic pole pieces are fabricated f rom type 410 stainless steel.

All ncnmagnetic parts, except pins and springs, are f abricated f rom

> type 304 stainless steel. Haynes-25 is used to fabricate latch / link pins. Springs are made f rom a nickel-chromium-iron alloy (Inconel X-750). Latch arm tips are clad with Stellite-6 to provide improved g wearability. Hard chrome plate and Stellite-6 are used selectively W for bearing and wear surf aces.

d. CRDM/GRDM Drive Rod Assembly The drive rod assembly utilizes a type 410 stainless steel drive rod.

The coupling is machined f rom type 403 stainless steel. Other parts are type 304 stainless steel with the exception of the springs, which are nickel-chromium-iron alloy and the locking button, which is Haynes-25.

e. CROM/GRDM Hub Extension Assembly The hub extension assembly utilizes a type 410 stainless steel hub extension rod. The coupling hub and semi-permanent coupling are machined from type 403 stainless steel. The locking sleeve is machined from Inconel 600 material.

O WAPWR-RS 4.5-2 AMEN 0 MENT 2 1253e:ld APRIL, 1985

t 4.5.1.1.2 Fabrication and Processing of Austenitic Stainless Steel Components 2 The discussions provided in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4, p " Reactor Coolant System", concerning the processes, inspections, and tests on V austenitic stainless steel components to ensure freedom from increased susceptibility to intergranular corrosion caused by sensitization; and the control of welding of austenitic stainless steels (especially control of delta f errite) , are applicable to the austenitic stainless steel pressure-housing s components of the CROM/GRDM.

4. 5.1.1. 3 Contamination Protection and Cleaning of Austenitic Stainless Steel 2 The CROM/GR0M are cleaned prior to delivery in accordance with Westinghouse process specifications. Process specifications in packaging and shipment are discussed in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System". Westinghouse personnel conduct surveillance of these operations to ensure that manuf acturers and installers adhere to appropriate requirements as D discussed in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System".

4.5.1.1.4 Other Materials g

Haynes-25 is used in small quantities to fabricate latch / link pins and locking buttons. The material is ordered in the solution-treated, cold-worked condition. Stress-corrosion cracking has not been observed in this application over the last 20 years in the environment similar to the WAPWR.

The CROM/GROM springs are made from a nickel-chromium-iron alloy (Inconel X-750) ordered to MIL-S-23192 Class A No.1 temper cold-drawn wire or Class D, spring temper cold drawn wire. Operating experience of similar designs has shown that springs made of this material are not subject to stress-corrosion cracking.

WAPWR-RS

_ 4.5-3 AMEN 0 MENT 2 1253e:1d APRIL, 1985

4. 5 .1. 2 Displacer Rod Drive System Structural Materials 4.5.1.2.1 Materials Specifications All DROM parts exposed to the reactor coolant are made of metals which resists the corrosive action of the reactor coolant water chemistry. The following metals are used exclusively - 300 and 400 series austenitic and martensitic stainless steels, nickel-chromium-iron, and cobalt based alloys.

A. Pressure Boundary Pressure retaining parts forming the primary pressure boundary are f abricated solely from nickel-chromium-iron alloy and the austenitic types 304 and 316 stainless steels where additional processes for property enhancement are permissible but not mandatory. All pressure boundary materials comply with the ASME Boiler and Pressure Vessel Code Section III.

2 B. Internals Cylinder Assembly The internals cylinder assembly is comprised of 304 or 316 materials for the cylinder, upper and lower support pieces, latch pins, and lock pins.

(a,b,c) Alternate nickel based gall resistant alloys [ ] may be utilized for latch and support hardware. Springs are made from nickel-chromium-iron alloy (Inconel X750). Hard chrome plate is used selectively for bearing and wear surfaces.

C. Drive Rod Assembly The drive rod assembly is made up of two drive rods. One is type 410 stainless steel, the second is type 304 SST. The joining coupling is machined f rom type 403 stainless steel. The remaining drive rod assembly parts are made from 304 stainless, nickel-chromium-iron alloys for springs, screws, and pins, cobalt based alloys for seals and high wear -

locking devices, the selective use of hard chrome on threads and wear (a,b,c) surfaces, and the nickel based anti-gall alloy [ ] for the piston.

((APWR-RS 4.5-4 AMEN 0 MENT 2 1253e:1d APRIL, 1985

D. RPI Support Assembly The RPI Support Assembly is external to the primary coolant however due to the functional . requirements imposed by the detection system the support assembly parts are fabricated solely of austenitic stainless steels. The top plate which serves as a seismic support is fabricated from material complying with the ASME Boiler and Pressure Vessel Code Section III, Sub-Section NF.

4. 5.1. 2. 2 Fabrication and Processing of Austenitic Stainless Steel Components The discussions provided in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4

" Reactor Coolant System", concerning the processes, inspections, and tests on austenitic stainless steel components to ensure freedom from increased susceptibility to intergranular corrosion caused by sensitization, and the discussions provided in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4. " Reactor Coolant System", concerning the control of welding of austenitic stainless 2

steels especially control of delta ferrite, are applicable to the austenitic stainless steel pressure-housing coer.ponents of the DROM.

4. 5.1. 2. 3 Contamination Protection and Cleaning of Austenitic Stainless Steel The DROMs are cleaned prior to delivery in accordance with the guidance of American National Standards Institute (ANSI) discussed in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4 " Reactor Coolant ' System". Westinghouse personnel conduct surveillance of these operations to ensure that manuf acturers and installers adhere to appropriate requirements as discussed in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4 " Reactor Coolant System".

4.5.2 Reactor Internals Materials 4.5.2.1 Materials Specifications All the major materials for the reactor internals are type 304 stainless steel. Parts not fabricated f rom type 304 stainless steel include bolts and WAPWR-RS 4.5-5 AMEN 0 MENT 2 1253e:1d APRIL, 1985

(

dowel pins, which are f abricated f rom type 316 stainless steel, and radial support key bolts, which are fabricated of Inconel-750. Radial support clevis inserts are Inconel-600 with Stellite-6 hardfacing, and the holddown spring is type 304 stainless steel. There are no other materials used in the reactor internals or core support structures which are not otherwise included in the ASME Code,Section III, Appendix I.

4.5.2.2 Controls on Welding The discussions provided in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4 O

" Reactor Coolant System" are applicable to the welding of reactor internals and core support components.

4.5.2.3 Nondestructive Examination of Tubular Products and Fittings 2

The nondestructive examination of wrought seamless tubular products and fittings is in accordance with Section III of the ASME Code.

O 4.5.2.4 Fabrication and Processing of Austenitic Stainless Steel Components Subsection 5.2.3 of RESAR-SP/90 PDA Module 4 " Reactor Coolant System" discusses the level of conformance of reactor internals and core support structures with Regulatory Guides 1.31, 1.34, 1.44 and 1.71.

4.5.2.5 Contamination Protection and Cleaning of Austenitic Stainless Steel The discussions provided in Subsection 5.2.3 of RESAR-SP/90 PDA Module 4,

" Reactor Coolant System" are applicable to the reactor internals and core support structures and verify conformance with ANSI 45 specifications and Regulatory Guide 1.37.

O MAPWR-RS 4.5-6 AMENDMENT 2 1253e:ld APRIL, 1985