ML20090D295

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Nonproprietary Chapter 1 of RESAR-SP/90 Westinghouse Advanced PWR Module 4, RCS, Introduction & General Description of Plant
ML20090D295
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Site: 05000601
Issue date: 06/30/1984
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WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML19273A237 List:
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NUDOCS 8407180182
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1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

The Westinghouse Electric Corporation (hereinafter referred to as Westing- j house) has developed this Reference Safety Analysis Report (RESAR-SP/90) for ,

the Westinghouse Advanced Pressurized Water Reactor (WAPWR) as part of its l continuing efforts toward design and licensing standardization. of nuclear l power plants. RESAR-SP/90 is a standard safety analysis report submitted i initially for Preliminary Design Approval (PDA) in accordance with Appendix 0, l

" Standardization of Design; Staff Review of Standard Designs," to Part 50 of l Title 10 of the Code of Federal Regulation (hereinafter referred to as f

10CFR). The ultimate objective is to obtain a Final Design Approval (FDA) of [

RESAR-SP/90 followed by a rulemaking proceeding and design certification.

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I JUNE, 1984 WAPWR-RCS 1.1 -1 1086e:1d l

I i 1.2 GENERAL PLANT DESCRIPTION 1.2.2 Princioal Desian Criteria l

RESAR-SP/90 is designed to comply with 10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants." The specific applications of

-General Design Criteria to RESAR-SP/90 are discussed in Section 3.1 of PDA  !

Module 7 Structural / Equipment Design'. Those General Design Criteria applicable to this module are listed in Section 3.1 of this module.

o i 1.2.3 Plant Description 1.2.3.2 Reactor Coolant System i The MAPWR Reactor Coolant System (RCS) consists of four closed heat transfer i loops connected in parallel to the reactor vessel. Each loop contains a steam '(a,c) generator and a . reactor coolant pump. 'In addition, the system includes a 3

  • ft pressurizer, a pressurizer relief tank, and ?.5e valves and instrumenta-

. tion necessary for operational control and safeguards actuation. Also  !

included in the RCS is a reactor vessel head vent and a reactor vessel level  !

l instrumentation system (RVLIS), as well as a displacer rod drive mechanism l -' I l

(DROM) vent system, used in connection with operation of the displacer rod drive mechanisms. The DROM's are fully described in RESAR-SP/90 PDA Module 5, I l ' Reactor System". All system equipment is located in the reactor containment,

  • except certain containment isolation and process-actuated valves located in i

the lines connected to the pressurizer relief tank. A simplified flow diagram  ;

l O of the system is shown in Figure 1.2-1.

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.During operation, the reactor coolant pumps circulate pressurized water through the *eactor vessel and the coolant loops. The water, which serves as l

Ge : reactor coolant, the moderator and the solvent for boric acid (chemical  ;

shim control) is heated as it passes through the core. It then flows to the r

I l steam generators, where the heat is transferred to the steam system, and returns to the reactor coolant pumps to repeat the cycle. Af ter the reactor l-l I. 1. 2-1 JUNE, 1984

'MAPWR-RCS ,

l 1086e:ld  !

i has been shut down, the reactor coolant is circulated by the residual heat removal subsystem equipment of the primary side safeguards system (sea RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System") to hve the heat generated in the fuel f rom fission product decay. The RCS also serves as the i second barrier against fission product release to the environment.

c RCS pressure is controlled by operation of the pressurizer, where water and  ;

steam are maintained in equilibrium (saturated conditions) by electrical  !

heaters and a water spray. Steam can De formed by the heaters and condensed ,

by the pressurizer spray to control pressure variations due to contraction and expansion of the reactor coolant. Spring loaded safety valves and power l operated relief valves are connected to the pressurizer and discharge to the pressurizer relief tank. In the event of safety valve or PORV operation, the j steam discharged is condensed and cooled ir, the pressurizer relief tank (PRT) by mixing with the water normally present in the tank. j l

The RCS -is also serviced by a number of auxiliary systems, including the  !

l chemical and volume control system (CVCS), the integrated safeguards system f (ISS), the main steam and feedwater systems (MSS and MFWS), and the secondary I I

side safeguards system (SSSS).

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i 1.3 COMPARISON TABLES i

1.3.1 Comparison With Similar Facility Desians i

Table 1.3-1 presents a design comparison of the major parameters and features of the WAPWR reactor coolant system (RCS) with RESAR-414 (Docket No.

STN-50-572; PDA-13), RESAR-3S (Docket No. STN-50-545; PDA-7), and RESAR-41 i (Docket No. STN-50-480; PDA-3). i LO lI i-l l

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TABLE 1.3-1 OESIGN COMPARISONS T; ,

RESAR- RESAR- RESAR- RESAR-( Parametar or Feature SP/90 414 3S 41 i (a,c) Total thermal flow rate (106 lb/hr) 150.5 140.3 144.7 O Reactor coolant system temperatures (*F)

1. Vessel outlet 624.9 618.3 624.0
2. Vessel average 596.1 588.2 591.8 ,
3. Vessel inlet 563.8 558.1 559.8  !

i Number of coolant loops 4 4 4 6

Total steam flow (10 lb/hr) 17.35 15.14 16.96 O Reactor vessel

! 1. Inside diameter (in.) 173 173 173 l

l 2. Inlet nozzle inside diameter (in.) 27.5 27.5 27.5

3. Outlet nozzle inside diameter (in.) 29 29 29
4. Number of reactor closure head studs 54 54 36 Reactor coolant pumps
1. Model 100 93Al 100
2. Horsepower 9000 7000 8000
3. Capacity (gpm) 107,600 99,000 99,700 I

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TABLE 1.3-1 (Continued)

DESIGN COMPARISONS j 1 [

RESAR- RESAR- RESAR- RESAR- l

' Parameter or Feature SP/90 414 35 41 Steam generators 4 1. Model WAPWR H D D

2. Feedwater entrance Feedring Feedring Preheater Preheater

.3. Heat transfer area (ft ) 82,500 48,300 67,000

4. Number of U-tubes 6650 4674 4864
5. Tube wall thickness, nominal (in.) 0.040 0.043 0.043 Pressurizer 3 1800 2100
1. Internal volume (ft ) 2100 Pressurizer safety valves 3 3 3
1. Number Maximum relieving capacity (lb/hr) 501,700 420,000 420,000
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JUNE, 1984 WPWR-RCS 1.3-3 1086e:1d

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1.6 MATERIAL INCORPORATED BY REFERENCE

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The WAPWR RCS module incorporates, by reference, certain topicat- reports. The

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topical reports, . listed in Table 1.6-1, have been filed previously in support f l O of other Westinghouse applications.  !

l The legend for the review status code letters used in Table 1. 6-1 is as  !

follows:  ;

O A -

U.S. Nuclear Regulatory Commission review complete; USNRC acceptance I letter issued -

AE -

U.S. Nuclear Regulatory Commission accepted as part of the f Westinghousa emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for  ;

ECCS analyses.

B -

Submitted to USNRC as background information; not undergoing formal l USNRC review 0 -

On file with USNRC; older generation report with current validity; not actively under formal USNRC review i

N - Not applicable; that is, open literature, etc.  !

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Actively under formal USNRC review ,

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MP MATERIAL INCORPORATED BY REFERENCE EM "

4 Westinghouse

' Topical SAR

_ Report No. Revision Section Title

_ Number Submitted Review Reference to the NRC Status WCAP-8301(P) LOCA-IV Program:

WCAP-8305 Loss-of-Coolant Rev 0 15.0. 15.6 Transtent Analysis 7/12/74 AE

' WCAP-8302(P) SATAN-IV Program: Comprehensive WCAP-8306 Rev 0 15.0, 15.6 7/12/74 Space-Time Dependent Analysis of AE Loss-of-Coolant WCAP-8324-A l Control of Delta Ferrite in Rev 0 5.2

! Austenitic Stainless Steel Weldments 6/23/75 A WCAP-8370 Westinghouse Water Reactor Divisions Rev 9A

. Quality Assurance Plan 1.9. 178 11/14/77 A 1

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" WCAP-8424 I

Evaluation of Loss-of-Flow Accidents Rev 1 15.3 5/30/75 Caused by Power System Frequency U Transients in Westinghouse PWRs

} WCAP-8510 Method for Fracture Mechanics Rev 0 5.3 Analysis of Nuclear Reactor Vessels 7/76 U l

Under Severe Thermal Transients WCAP-8567-P(P) Improved Thermal Design WCAP-3568 Procedure Rev 0 15.0 7/75 A WCAP-6693 Delta Ferrite in Production Rev 0 5.2 Austenitic Stainless Steel Weldments 3/16/76 8 WCAP-8768 Safety-Related Research and Rev 2 5.4 10.78 Development for Westinghouse Press- 8 a urized Water Reactors Program s Sumaries - Winter 1977 f through Summer 1978 G

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, TABLE 1.6-1. (cont)

I y% MATERIAL INCORPORATED BY REFERENCE E '

.6 l "'O Westinghouse . SAR-l Topical Revisi*n Section Submitted Review Report No. Title Number Reference to the NRC_ Status

! WCAP-8846-A Hybrid B4C Absorber Control Rod Rev 0 15.0 10n? A Evaluation Report WCAP-9600(P) Report on Small Break Accidents for Rev 0 5.4 6D9 A WCAP-9601 Westinghouse NSSS Systems 1

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t 1.7 DRAWINGS AND OTHER DETAILED INFORMATION  !

1.7.1 Pipit. and Instrumentation Diaarams i i i

Table 1.7-1 contains a list of each piping and instrumentation diagram and the corresponding RCS figure numoer. Figure'1.7-1 illustrates and defines symbols and abbreviations used in the diagrams.

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TABLE 1.7-1 .

,. PIPING AND INSTRUMENTATION DIAGRAMS ,

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1.8 .CONFORMANCE WITH THE STANDARD REVIEW PLAN I

t In accordance with 10CFR50.34(g), Table 1. 8-1 identifies a'nd evaluates deviations f rom the acceptance criteria of those sections of the NRC Standard  :

Review Plan (NUREG-0800) pertinent to the RCS.

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6 l Atilt 1. t$- 1 l STANDARD REVIEW PLAN DEVIATIONS  !

SRP Acceptance Criteria Deviation _Section i s  !

i BTP MEB 3-1, Westinghouse does not assume (1)

G l U paragraph B.1.C.(1) the arbitrary intermediate pipe f break stated in B.1.C.(1)(d)  ;

BTP MEB 3-1, Westinghouse does not assume the (1)  !

O paragraph B.1.C.(2) arbitrary intermediate pipe break l

i stated in B.1.C.(2)(b)(ii)

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i O (1) See RESAR-SP/90 PDA Module 7, " Structural / Equipment Design" for justifica-tion for this deviation.

I O c WAPWR-RCS 1.8-2 JUNE, 1984 T086e:1d  ;

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TABLE 1.8-2 l CONFORMANCE TO US NRC 9EGULATORY GUIDES i

APPLICABLE TO THE WAPWR RCS _4  !

O REGULATORY GUIDE 1.14. REVISION 1. AUGUST 1975, REACTOR COOLANT PUMP FLYWHEEL INTEGRITY  !

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' Since the issuance of Regulatory Guide 1.14. Revision 1, the NRC has provided l

to Westinghouse a copy of draft 2, Revision 2, of Regulatory Guide 1.14. This i draf t was formulated from industry and concerned parties' conments . It is j l significant that the draf t 2 version ' incorporates several of the Westinghouse consents on Revision 1. Since draf t 2 has not been formally published as l Revision 2 of Regulatory Guide 1.14, the exceptions and clarifications (f rom i I - the original. Westinghouse comments) are provided below:

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Cross rollina ratio of 1 to 3 ,

O Westinghouse's position is that specification of a cross rolling ratio is  ;

unnecessary since past evaluations have shown that ASME SA-533, Grade B, Class ,

1 materials produced without this requirement have suitable toughness for l l

typical flywheel applications. Proper material selection ard specification of >

minimum material properties in the transverse direction adequately ensure .

l flywheel integrity. ~ An attempt to gain isotropy in the flywheel materials by  ;

! means of cross rolling is unnecessary since adequate margins of safety are j provided- by both flywheel material selection ( ASME 5A-533, Grade B, Class 1) I and - by specifying minimum yield and tensile levels and toughness test values [

taken in the direction perpendicular to the maximum working direction of the l material.  :

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l O The requirements for the vacuum melting and degassing process or

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l Q 1. C.1-the electroslag process are not essential in meeting the balance of r

I the regulatory position or do they, in themselves, ensure conformance ,

with the overall regulatory position. The initial Saf ety Guide 14 ,

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WAPWR-RCS 1.8-3 JUNE, 1984  :

I T086e:1d t 6

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l TABLE 1.8-2 (continued) I O i (October 27, 1971) stated that the " flywheel matgrial should be produced by a process that minimized. flaws in the material and l improved its fracture toughness properties." This is accomplished by using ASME SA-533 material including vacuum treatment.

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Specification of a cross rolling ratio is considered unnecessary since i past evaluations have shown that ASME SA-533, Grade B, Class 1 materials produced without this requirement have suitable toughness for typical flywheel applications. Proper material selection and h specification of minimum material properties in the transverse direction adequately ensure flywheel integrity. An attempt to gain isotropy in the flywheel material by means of cross rolling is I unnecessary since adequate margins of safety are provided by both f

flywheel material selection ( ASME SA-533, Grade B, Class 1) and by  :

specifying minimum yield and tensile levels and toughness test values ,

taken in the direction perpendicular to the maximum working direction [

of the material.

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2. C.2 Because the MAPWR design specifies a light interference fit i between the flywheel and the shaf t, at zero speed the hoop stresses l and radial stresses at the flywheel bore are negligible. Centering of ,

l the flywheel relative to the shaf t is accomplished by means of keys [

t l and/or centering devices attached to the shaf t, and at normal speed, the flywheel is not in contact with the shaft in the sense intended by  !

Revision 1. Hence, the definition of " excessive deformation," as ,

I defined in this guide, is not applicable to the design since the enlargement of the bore and subsequent partial separation of the l flywheel f rom the shaft does not cause unbalance of the flywheel.

Extensive experience with reactor coolant pump flywheels installed in 1

O this fashion has verified the adequacy of the design.

The combined primary stress levels, as defined in Revision 0 of I Regulatory Guide 1.14 (regulatory positions C.2.a and C.2.c) are both f 5

WAPWR-RCS 1.8-4 JUNE, 1984 1086e:1d i

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TABLE 1.8-2 (continued)

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4 . conservative and proven and no changes to these st6ess levels are j

[ necessary. Westinghouse designs to these stress limits and thus does l not have permanent distortion of the flywheel bore a,t normal or spin test conditions.  :

Paragraph 2.b is considered as delineated. The interpretation removes the ambiguous reference to an undefined overspeed transient.

C.3 Conform.

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4. C.4 WCAP-8163 shows that the flywh'el e would not fail at 290 percent j of normal speed for a flywheel flaw of 1.15 in, or less in length.  !

Results for a double-ended guillotine break at the pump discharge with i full separation of pipe ends assumed, show the maximum overspeed to be L

1ess than 110 percent of normal speed. The maximum overspeed was calculated to be about 280 percent of normal speed for the same postulated break, and an assumed instantanewc: loss of power to the l reactor coolant pump. _

In comparison with the overspeed presented  !

above, the flywheel is tested at 125 percent of normal speed at the factory. Thus, the flywheel could withstand a speed up to 2.3 times s greater than the flywheel spin test speed of 125 percent provided that l no flaws greater than 1.15 in. are present. If the maximum speed were 125 percent of normal speed or less, the critical flaw size for l

failure would exceed 6 in. in length. Nondestructive tests and l

! critical dimension examinations are all performed before the spin tests. The inspection methods employed provide assurance that flaws l significantly smaller than the critical flaw size of 1.15 in, for 290 l I

percent of normal speed would be detected. Flaws in the flywheel will i be recorded in the prespin inspection program. Flaw growth attribu-table to the spin test (i.e., from a single reversal of stress, up to j I

speed and back), under the most adverse conditions, is about three orders of magnitude smaller than that which nondestructive inspection F WAPWR-RCS 1.8-5 JUNE, 1984 5086e:1d ,

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techniques are capable of denctirg. For these reason,s, no post-spin l

}- inspections are performedfsince the prespin'. test inspections are l m i considered adequate. .1 L l

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Refer to Subsection 5.4.h5 for further discussion? l l

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REGULATORY GUIDE 1.26, REVISION 3, FEBRUARY 1976, O QUALITY GROUP CLASSIFICATIONS AND STANDARDS FOR l WATER , STEAM , AND RADI0 ACTIVE-WASTE-CONTAINING .

3 ;' .

t COMPONENTS OF NUCLEAR POWER PLANTS j

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Quality group classification's, . and standards for water , steam , and l ib

! radioactive-waste-containing components for the WAPWR meet the intent of .

s Regulatory Guide 1.26 with the following alternat M s:

s The safety class tenninology of ANSI /ANS 51.1-1983 is used instead of the O quality group terminology. Thus, the terms Safety Class 1, Safety Class 2, l Safety Class 3, and Non-nuclear Safety (NNS) Class are used instead of Quality f Groups A, B, C, and D, respectively, and are consistent with present nuclear f

i industry practice. l L  :

Paragraph NB-7153 of the ASME Code Section III requires that there be no f valves between a code safety valve and its relief point unless special  !

l interlocks prevent shutof f without other protection capacity. Therefore, as I L

an alternative to Paragraphs C.1.e and C.2.c a single safety valve designed, O manufactured, and tested in accordance with ASME III Division 1 is considered

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acceptable as the boundary between the reactor coolant pressure boundary and a i lower safety class or NNS class line.

Lo i Each component which is required to mitigate the consequences of an accident, i

j as defined in ANSI /ANS 51.1, shall be classified Seismic Category I. In addition. all components classified as Safety Class 1, 2, or 3 shall be j l

designated Seismic Category I. Seismic Category I components, structures, and l

i l i l WAPWR-RCS 1.8-6 JUNE,1984 [

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1d l

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R ,

systems shall be designed to remain functional in the event of the safe ,

shutdown earthquake (SSE). All Seismic Catego.y I components are designed and V constructed to Quality Assurance (0A) Category I requirements. j i Portions of structures, systems, and components which are not required for safety functions, but whose failures could affect safety-related components, l shall be designed such that the SSE will not result in adverse effects on safety-related components.  ;

i  !

Seismic Category I design requirements shall extend to the first seismic  !

l restraint beyond the seismic boundary and shall include the interface portion i of the boundary itself (that is, for piping systems, the isolation valve at a boundary between Seismic Category I and nonseismic porti>ns shall be designated Seismic Category I. The piping up to, and including, the first seismic restraint beyond the valve shall be designed to Seismic Category I requirements (but shall not be designated Seismic Category I). By this means,  ;

the Seismic Category I boundary is defined with respect to safety-related (

function, and the interfacing portions meet the seismic design requirements in ,

order to ensure the integrity of the boundary.

( REGULATORY GUIDE 1.29, REVISION 3. SEPTEMBER 1978, ,

SEISMIC DESIGN CLASSIFICATION s

The WAPWR conforms with this regulatory guide as shown in Table 3.2-1 of l RESAR-SP/90 PDA Module 7. " Structural / Equipment Design." j With regard to regulatory position C .1, each nuclear steam supply system

(~ .(NSSS) component important to safety is classified as Safety Class 1, 2, or 3; i

l

\ these classes are qualified to remain functional in the event of the SSE, ,

except where exempted by meeting all of the below requirements. Portions of I l

systems required to perform the same safety function as required of a safety r l class component which is part of that system shall be likewise qualified or

' granted exemption. Conditions to be met for exemption are:

l i

WAPWR-RCS 1.8-7 JUNE, 1984 T086e:1d

.., < ~ , , , .. . . . _ . - ,, , ~ . , _ . _ .

TABLE 1.8-2 (continued) o Failure would not directly cause a Condition Ill o r. IV event (as defined in ANSI N18.2-1973) .

9 o (here is no safety function to mitigate, nor could failure prevent mitigation of, the consequence of a Condition III or IV event.

o Failure during or following any Condition IV event would result in consequences no more severe than allowed for a Condition III ./ent.

o Routine post-seismic procedures would disclose loss of the safety function.

REGb ..ORY GUIDE 1.31, REVISION 3, APRIL 1978, CONTROL OF Fi.RRITE CONTENT IN STAINLESS STEEL WELD METAL The WAPWR confonns to the regulatory position of this regulatory guide. Refer to Subsection 5.2.3.4.6.

( REGULAT3RY GUIDE 1.34, DECEMBER 1972, CUNTROL OF ELECTROSLA' WELD PROPERTIES

.N-The WAPWR conforms to the regulatory position of this regulatory guide. Refer i

'\ to Subsection 5.2.3.4.6.

REGULATORY GUIDE 1.26, FEBRUARY 373, NON-METALLIC I. ,

THERMAL INSULA ~10N FOR AUSTENITIC STAINLESS STEEL f

\\f The WAPWR conforms to the regulatory positior of this regulatory guide. Refer

\ to Subsections 5.2.3.2.3 and 6.1.1.i.3.

s

^

k .\

\

\\e o a

~

1.8-8 JUNE, 1984 WAPWR-kCS 1086e : h'

\s' asaamm

l TABLE 1.B-2 (continued) f REGULATORY GUIDE 1.37, MARCH 1973, QUALITY ASSURANCE REQUIREMEllTS

[

FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER - l COOLED NUCLEAR POWER PLANTS i

The Westinghouse quality assurance is provided in WCAP-8370. See the RESAR-SP/90 integrated PDA document for a description of the complete quality  :

assurance program, including modifications to reflect the expanded s: ope of

(

the WAPWR Nuclear Power Block.

REGULATORY GUIDE 1.43, MAY 1973, CONTROL OF STAINLESS ,

STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS Qualification testing is performed on any high-heat input welding process

{

(such as the submerged-arc wide-strip welding process or the submerged arc t j 6-wire process) used to clad coarse or fine grained SA-508 Class 2 material.  !

l This test follows the recommendations of this guide. Production welding is monitored by the f abricator to ensure that essential variables remain within the limits establi-hed by the qualification. If the essential variables exceed the qualification limits, an evaluation is performed to determine if l

the cladding is acceptable for use. Where Westinghouse permits the use of L submerged-arc strip process on SA-508 Class 2 material, a two-layer technique is used to minimize intergranular cracking. Refer to Subsection 5.2.3.3.2.

l REGULATORY GUIDE 1.44, MAY 1973, CONTROL OF THE USE OF I SENSITIZED STAINLESS STEEL l

l The WAPWR conforms to the regulatory position of this regulatory guide.

Refer {

to Subsection 5.2.3.4.

V REGULATORY GUIDE 1.45, MAY 1973, REACTOR COOLANT PRESSURE BOL 10ARY LEAKAGE DETECTION SYSTEM ,

p. The WAPWR conforms to the regulatory portion of this regulatory guide. Refer to Subsection 5.2.5.  !

t!APWR-RCS 1.8-9 JUNE, 1984 1086e:1d j

. - - ., , - .- . . - - - - . . . .-_ _ _ _ _ _ _ _ _ -L

TABLE 1.8-2 (continued)

O REGULATORY GUIDE 1.46, MAY 1973, PROTECTION AGAINST PIPE WHIP INSIDE l

CONTAINMEl- i This guide presents bases for identifying postulated rupture locations and orientations for Code Class 1, 2, and 3 piping inside containment with the '

intent of providing guidance for satisfying General Design Criterion 4, i

" Environmental and Missile Design Basis," of Appendix A to 10 CFR Part 50. l GDC-4 requires, in part, that structures, systems, and components important to l l safety be appropriately protected against dynamic effects that may result from  ;

equipment failures, including the effects of pipe whipping. -

Subsequent to the issuance of Regulaton/ Guide 1.46 in 1973, the NRC released Branch Technical Position MEB 3-1, " Postulated Rupture Locations in Fluid {

System Piping Inside and Outside Containment," (Revised, July 1981). Although j both sets of criteria address Code Class 1, 2, and 3 piping.and are compatible in intent, they contain quantitative differences which will havi~a significant i effect on plant design. Standard Review Plan 3.6.2, which addresses j protection against pipe rupture, allows the use of MEB 3-1 criteria inside l

containment if it is impractical to implement the criteria of Regulatory Guide t

1.46.  !

[ i i

The criteria implemented in the evaluation of the main reactor coolant loop is ,

based on draf t ANS Standard 20.2, " Design Basis for Protection Against Sipe Whip," and is documented in WCAP-B172-A, " Pipe Breal s for the LOCA Analysis of

- the Westinghouse Primary Coolant Loop." WC'AP-8172-A has received NRC approval l I ' as providing an equivalent degree of protection as would be obtained by applying the criteria of Regulatory Guide 1.46.

i On February 1, 1984 the NRC issued letters to Westinghouse and all PWR j I \ Licensees on the resolution of Generic Issue A-2 (Asymmetric LOCA loads) ed  ;

the use of alternate pipe break criteria for the reactor coolrnt system j l

primary loop. In these letters, the NRC accepted the justification provided l l

0 l WAPWR-RCS 1.8-10 JUNE, 1984 1086e:ld

t t

TABLE 1.8-2 (continued)

O l I

by Westinghouse for the elimination of the pipe greak locatians defined in NUREG-0609 as a basis for resolving Generic Issue A-2 and stated that the I g conclusions are also applicable to all other PWR's. The NRC letter further concludes that since the pipe breaks considered in the A-2 evaluation represent the most limiting condition, pipe whip restraints and jet j impingement devices can be eliminated for any pipe breaks in the RCS primary loop provided the applicability of WCAP-9558 Rev. 2 and WCAP 9787 Rev. O is O demonstrated and the conditions defined in the NRC letters are met.

Refer to RESAR-SP/90 PDA Module 7, " Structural / Equipment Design" (Section 3.0) for the alternative pipe break criteria for the RCS primary loop piping as f well as other Class I portions of the RCS, i

REGULATORY GUIDE 1.50, MAY 1973, CONTROL OF PREHEAT TEMPERATURES FOR WELDING 0F LOW-ALLOY STEEL i

! (  :

I

1. C.1 Paragraph 1.a is conformed with when impact testing, in accordance with ASME Boiler and Pressure Vessel Code,Section III, Subarticle 2300, is required. When impact testing is not required,  ;

l

! specification of a maximum interpass temperature in the welding i procedures is not necessary in order to assure that the other required

! mechanical properties of the weld are met.

I  ;

Paragraph 1.b conforms. '

l

2. C.2 Conforms for pressure vessels with nominal thicknesses greater t

than l'in. Maintenance of preheat beyond completion of welding until postweld.

l: O '

The NSSS Class -1 components are in conformance with Regulatory Guide 1.50

~

l except for regulatory positions 1.b and 2. For Class 2 and 3 components, l

Westinghouse does not apply Regulatory Guide 1.50 recommendations.  ;

O r V ,

! r WAPWR-RCS 1,8-11 JUNE, 1984 1086e:ld ,

P

TABLE 1.8-2 (continued) ,

.O In the case of regulatory position 1.b, the welding proceduresy are qualified f within the preheat temperature ranges required by Section IX of the ASME p Code. Experience has shown excellent quality of welds using the ASME h qualification procedures. '

In the case of regulatory position 2, it is felt that this position is both  !

unnecessary and impractical. Code acceptante low-alloy steel welds have been l b and are being made under present Westinghouse specified procedures. It is not necessary to maintain the preheat temperature until a post-weld heat treatment j has been performed as rec;uired by the guide, in the case of large components.

In the case of reactor vessel main structural welds, the practice of l maintaining preheat until the intermediate or final post-weld heat treatment has been followed by Westinghouse. In either case, the welds have shown high ,

integrity. Westinghouse practices are documented in WCAP-8577, The Application of Preheat Temperature After Welding of Pressure Vessel Steel, l g which has been accepted by the NRC. Refer to Subsection 5.2.3 for further l

C/ discussion. l l REGULATORY GUIDE 1.54, JUNE 1973, QUALITY ASSURANCE .

REQUIREMENTS FOR PROTECTIVE C0ATINGS APPLIED TO j WATER-COOLED NUCLEAR POWER PLANTS Equipment located in the containment building is separated into four categories to identify the applicability of this regulatory guide to various ,

types of equipment. These categories of equipment are as follows:

O i i

Category 1 - Large equipment  !

Category 2 - Intermediate equipment  !

O Category 3 - Small equipment >

Category 4 - Insulated / stainless steel equipment l

l A discussion of each equipment category follows:

O G

l WAPWR-RCS 1.8-12 JUNE, 1984 l 1086e:ld  !

i l

I i i

l l

1 l

i O TABLE 1.8-2 (continued) i

a. Catecory 1 - Larae Eauioment 1; ,

l The Category 1 equipment consist of the following:

Reactor coolant system supports.  :

Reactor :oolant pumps (motor and motor stand).

Accumulator tanks. J i

Manipulator crane.  ;

. t Since this equipment has a large surface area and is procured f rom

, only a few vendors, it is possible to implement tight controls over these items. ,

Stringent requirements have been specified for protective coatings on

[ equipment through the use of a painting specification in our c l

l procurement documents. This specification defines requirements for:

1. Preparation of vendor procedures. f f
2. Use of specific coating systems which are qualified to ANSI NiO1.2. t r

i

3. Surface preparation. j i
4. Application of the coating systems in accordance with the paint  !

L manufacturer's instructions.

5. Inspections and nondestructive examinations. l l
6. Exclusion of certain materials. f
7. Identification of all nonconformances.

O D 8. Certifications of compliance.

[

FPWR-RCS 1.8-13 JUNE, 1984 !

1086e:1d  ;

-.~_._,-._-__._.._.,.____.___._.._.___.-,___.___.._._...__J

i I

i  ;

TABLE 1.8-2 (continued)

\

The vendor's procedures are subject to review by engineering personnel, and the vendor's implementation of the specification l

requirements is monitored during the Westinghouse quality assurance

! surveillance activities. (

- i This ' system of controls provides assurance that the protective  !

coatir.gs wil's properly adhere to the base metal during prolonged exposure to a post-accident environment present with the containment v  ;

building. .

f i

i

b. Cateaorv 2 - Intermediate Eauipment The Category 2 equipment consists of the following: j I

Seismic platform and tie rods. l Reactor internals lifting rig.

O' Head lifting rig.

Electrical cabinets.

t Since these items are procured f rom a large number of vendors, and I individually have very small surf ace areas, it is not practical to ,

enforce the complete set of stringent requirements which are applied {

to Category 1 items. However, another specification has been [

implemented in . the NSSS procurement documents. This specification l defines to the vendors the requirements for: [

1. Use of specific coating systems which are qualified to ANSI N {

101.2.

2. Surface preparation.
3. Application of the coating systems in accordance with the paint l manufacturer's instructions. 6 WAPWR-RCS 1.8-14 JUNE, 1984 1086e:1d t
l. -

l L  :

l

[' TABLE 1.8-2 (continued)

The vendor's compliance with the requirements is also thecked during i the quality assurance surveillance activities in the vendor's plant. l

-These measures of control provide a high degree of assurance that the l protective coatings will adhere properly to the base metal and l withstand the postulated accident environment within the containment i building.- i O c. Cateaorv 3 - Small Eauipment i

, Category 3 equipment consists of the following: l Transmitters l Alarm horns- f

, i a Small instruments  !

Valves l

Heat exchanger supports  ;

i These items are procured from several different vendors and are j painted .by - the vendor in accordance with conventional. industry  ;

practices. Because the total exposed surface area is very small, no further requirements are specified.

d. .Cateaorv 4 - Insulated or Stainless Steel Equipment  !

t i-  ;

j, Category 4 equipment consists of the following:

Steam generators - covered with blanket insulation.

Pressurizer - covered with blanket insulation. j e Reactor pressure vessel - covered with rigid reflective insulation. l Reactor cooling piping - stainless steel. ,

i Reactor coolant pump casings - stainless steel. l r

i O  :

i JAPWR-RCS 1.8-15 JUNE, 1984

-1086e:ld  ;

i

. 4

i TABLE 1.8-2 (continued)

Since Category 4 equipment is insulated or is stainYess steel, no painted surface areas are exposed within the containment. Therefore, I this regulatory guide is not applicable for Category 4 equipment. I For further information, refer to Section 6.1.

REGULATORY GUIDE 1.65, OCTOBER 1973, MATERIALS AND f INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS  ;

, l Ine WAPWR conforms 4th this guide except for two points. The use of modified {

SA-540 Grade 824 material as specified in ASME Boiler and Pressure Vessel Code Case 1605 is not specified in the guide but is used by Westinghouse. The use I of this Code Case has been approved by the NRC via Regulatory Guide 1.85.

i t i The maximum limit of 170 ksi ultimate tensile strength is not explicitly specified by Westinghouse as required by the guide. Westinghouse does specify l fracture toughness of 45 ft/lb and 25 mils lateral expansion as required t,y j the ASME Code and 10 CFR 50, Appendix G. These requirements also result in '

strength levels below the maximam limit, as demonstrated by the actual stud ,

material properties for the WAPWR. l REGULATORY GUIDE 1.70, REVISION 3, NOVEMBER 1978, STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS l FOR NUCLEAR POWER PLANTS L

O The format and content of the RESAR-SP/90 PDA Modules meet the intent of this regulatory guide. ,

i REGULATORY GUIDE,1.71, DECEMBER 1973, WELDER j QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY (

i This guide provides guidelines above and beyond requirements of ASME Section

IX. All welder qualification for the WAPWR will be in conformance with ASME l

WAPWR-RCS 1.8-16 JUNE, 1984 1086e:1d

t TABLE 1.8-2 (continued) l l

Section IX. Few welds of limited accessibility are eipected to be encountered. Reasonable engineering judgment will be used to determine if  ;

performance qualification is necessary under simulated access conditions for '

any specific case.

f Westinghouse prac.tice does not require qualification or requalification of  ;

welders for areas of limited accessibility as described by the guide and has  !

provided welds of high quality. Limited accessibility qualification or l requalification, which are additional to ASME Section III and IX requirements,  !

is an unduly restrictive requirement for shop fabrication, where the welders' physical position relative to the welds is controlled and does not present any f significant problems. In addition, J 70 welds of limited accessibility are

{

repetitive due to multiple production of similar components, and such welding  ;

is closely supervised. ,

Refer to Subsection 5.2.3 for further discussion.

REGULATORY GUIDE 1.83, REVISION 1. JULY 1975, INSERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR TUBES The MAPWR conforms to the regulatory position of this regulatory guide. The ,

steam generators are designed to permit access to tubes for inspection and/or repair or plugging (if necessary). The inservice inspection program is discussed in Subsection 5.4.2 and the Technical Specifications.

REGULATORY GUIDE 1.84, REVISION 20, NOVEMBER 1982, DESIGN AND FABRICATION CODE CASE ACCEPTABILITY - ASME I SECTION III, DIVISION 1 The MAPWR conforms to the regulatory position of this regulatory guide. The l NSSS Code Cases are discussed in Subsection 5.2.1.2.

i 1.8-17 JUNE , '.')84 e MAPWR-RCS 1086e:1d

1 N

TABLE 1.8-2 (continued) i\

( REGULATORY GUIDE 1.85, REVISION 20, NOVEMBER 1982, 5 f

MATERIALS CODE CASE ACCEPTABILITY - ASME SECTION III DIVISION 1 v

The WAPWR conforms with the regulatory position of this regulatory guide. The  ;

NSSS ode Cases are discussed in Subsection 5.2.1.2. i REGULATORY GUIDE 1.99, REVISION 1 APRIL 1977 EFFECTS 0F RESIDUAL ELEMENTS ON PREDICTED RADIATION DAMAGE TO i REACTOR VESSEL MATERIALS -

f There are two primary issues with the guide:

i

1. The guide provides a procedure and curves for predicting radiation damage (as relating to the shift of the reference temperature, RTNOT), in terms of chemistry (Cu and P) and fluence. This guide's procedure differs significantly from the one used by the WAPWR.  !

Since the adjustments in reference temperature obtained from the radiation damage curves are used in developing heatup and cooldown limits for plant operation, the use of the curves in the guide could result in over-conservative heatup and cooldown limits during plant life.

2. The guide restricts the end of life transition temperature to 200*F maximum. Control of r.'sidual elements such as :opper, phosphorus, l

sulfur, and vanadium in the reactor vessel beltline materials of new j plants to levels that result in a predicted adjusted reference l temperature of less than 200*F at end of life is considered technically unnecessary and could lead to unnecessary changes in ,

chemistry (Cu and P) requirements with corresponding adverse impact on cost and materials availability. +

.V j WAPWR-RCS 1,8-18 JUNE, 1984 1086e:ld  ;

i F

O V

TABtr 1.8-2 (continued) I i

l

~

One additional feature of the guide constitutes a lesser bu't nevertheless j important issue: l

1. Figure 2 of the guide presents a curve which gives the decrease of upper shelf impact energy with fluence as a function of Cu content. l Although it appears that the prescribed relationship does not predict l unacceptable drops in upper thelf toughness for vessels with f

controlled chemistry the curves are nevertheless over conservative.  !

l The MAPWR position with respect to each of the guide positions is as follows: r 1

1. The basis as well as the scope of the guide for predicting adjustment l of reference temperature as given in regulatcry position C.1 are f
inappropriate since the data base used was incomplete and included l some data which were not applicable.  !

I l 2. The MAPWR design is consistent with the guide position C.2a. However, with respect to guide position C.2b, Westinghouse believes that f Figure 2 of the guide is incarrect since the upper shelf energy for f

6-in.-thick American Society of Testing Materials (ASTM) A3028 1 reference correlation monitor material reported by Hawthorne indicates t 1 x 10 I9 l l

essent..ily a constant upper shelf at fluences above -

n/cm2 ,(a) {

s

3. The Westinghouse position with reference to the guide position C.3, .f

! controlling residual elements to levels that result in a predicted ,

6 adjusted reference temperature of less than 200*F at end of life, is  ;

that the stresses in the vessel can be limited' during operation in I order to comply with the requirements of Appendix G to 10 CFR 50 even though the end of life adjusted reference temperature may exceed l r

i -a. Hawthorne, J. R., " Radiation Effects Information Generated on the ASTM f Reference. Correlation-Monitor Steels," ASTM, Philadelphia, 1974.  !

MAPWR-RCS 1,8-19 JUNE,1904 1086e:1d [

i I

TABLE 1.8-2 (continued) j I I

, 200*F. By applying the procedures of Appendix G to ASqE Section III, the stress limits including apprcpriate. Code safety margin can be met.

4. Recent surveillance capsule data indicate a steady-state condition of radiation damage well below that predicted by current trend i curves.5"I This effect is believed to be due to the annealing of f n

v the vessels at the operating temperature. As an alternative to Regulatory Guide 1.99, operating limits will be determined using the I

current radiation damage curves developed by Westinghouse.I ) It is expected that as future surveillance specimens are evaluated it will become increasingly evident that both the Regulatory Guide 1.99 and i

Westinghouse trend curves are very conservative.  !

Refer to Section 5.3 for further discussion.

?

i n

v REGULATORY GUIDE 1.121, AUGUST 1976, 8ASES FOR PLUGGING DEGRADED PWR STEAM GENERATOR TUBES i

l

! Conform, with the following exceptions:

1. C . '. Westinghouse interprets the term " unacceptable defects" to apply  !

to those imperfections resulting from service induced mechanical or  :

chemical degradation of the tube walls which have penetrated to a j i depth in excess of the plugging limit.  ;

1  ;

1 O# 2. C.2a(2) and C.2.a(4) Westinghouse will use a 200-percent margin of I safety based on the following definition of tube f ailure. Westing- l house defines tube failure as plastic deformation of a crack to the extent that the sides of the crack open to a non-parallel, elliptical l >

a. Letter NS-TMA-1843 to the Secretary of the Commission, T. M. Anderson,  !

I June 23,1978.  !

b. Westinahouse RESAR-35, chapter 16, figure B/3/4.2, page 83/4 4-8. I

(

~

WAPWR-RCS 1.8-20 JUNE, 1984 !

1086e:1d  ;

i r

%,. m....,,..._...,_.v..,_ _,,,,,_w,,,,, _

t l

TABLE 1.8-2 (continued) l O configuration. This 200-percent margin of safety compares f avorably l with the 300 percent margin requested by the NRC against(gross failure.  !

e

3. C.2.b In cases where suf ficient inspection data exist to establish O- degradation allowance, the rate used will be an average time-rate

, determined from the mean of the test data. f Where requirements for minimum wall are markedly different for different areas of the tube bundle, e.g., U-bend area versus straight length in Westinghouse designs, two plugging limits may be established  ;

! to address the varying requirements in a manner which will not require i unnecessary plugging of tubes. i i

j 4. C.3.d(1) and C.3.d(3) The combined effect of these requirements would  !

be to establish a maximum permissible primary-to-secondary leak rate which may be below the threshold of detection with current methods of  !

l measurement. Westinghouse has determined the maximum acceptable j length of a through-wall crack based on secondary pipe break accident j loadings' which are typically twice the magnitude of normal operating j pressure loads. Westinghouse will use a leak rate associated with the j crack size determined on the basis of accident loadings. j

5. C.3.e(6) Westinghouse will supply computer code names and references f l rather than the actual codes. l p

i

6. C.3.f(1) Westinghouse will establish a minimum acceptable tube wall i

thickness (plugging limit) based on structural requirements and  !

consideration of loadings, measurement accuracy and, where applicable. l l

a degradation allowance as discussed in this position and in accord-  !

f ance with tht general intent of this guide. Analyses to determine the j maximum acceptable number of tube failures during a postulated '

I condition are nonna11y done to entirely dif ferent bases and criteria  !

are not within the scope of this guide. [

Refer to . Subsection 5.4.2 for f urther discussion.

WAPWR-RCS 1.8-21 JUNE, 1984 1086e:1d  :

.h y

TABLE 1.8-2 (continued) .

REGULATORY GUIDE 1.124, REVISION 1 JANUARY 1978, i SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 [

LINEAR-TYPE COMPONENT SUPPORTS In general., the WAPWR conforms.

For the NSSS portions, the following j exceptions are taken.

Paragraph C.2 of the regulatory guide presents two methods of estimating the I

ultimate tensile strength S, at temperature. It is believed that method No. 2 is not conservative at elevated metal temperature (in excess of 800*F).  !

In Westinghouse's judgment, values of S at these elevated temperatures should be determined by test rather than via the method given in C.2(b).  !

Paragraph C.4 of the regulatory guide states: "However, all increases, i.e., 8 those allowed by NF-3231.1(a), XVII-2110(a), and F-1370(a), should always be i n)

(J. limited by XVII-2116(b) of Section III." Paragraph XVII-2110(b) specifies that member compressive axial loads shall be limited to two-thirds of critical buckling. Satisfaction of this criteria for the faulted condition is  :

unnecessarily restrictive.

The most significant faulted condition loads on equipment supports result from seismic disturbances and postulated LOCAs, both of which are dynamic events.

The allowable faulted condition compressive load should not be limited to  ;

two-thirds of critical buckling because these f aulted dynamic lo2ds are of extremely short duration, and support members can take impulsive loads that l exceed static critical buckling load. Westinghouse will use a compressive ,

axial load of 0.9 of critical buckling since the dynamic buckling capacity of the member is greater than the static buckling capacity.  !

Paragraph C.6(a) of the regulatory guide appears to erroneously allow the use of faulted stress limits for the emergency condition. Westinghouse will

, interpret this paragraph as follows: "The stress limits of XVII-2000 of

Section III and regulatory position 3, increased according to the provisions f WAPWR-RCS 1.8-22 JUNE, 1984 '

1086e:1d i

?

i ,

TABLE 1.8-2 (continued)

, i of XVII-2100(a) of Section III, should not be exceeded for comp'onent supports l designed by the linear elastic analysis method." f O Westinghouse will use the provisions of F-1370(d) to determine faulted r _

condition allowable loads for supports designed by the load rating method. I The method described in paragraph C.7(b) of the regulatory guide is very

conservative and inconsistent with the remainder of the faulted stress limits.

In paragraphs 8.5 and C.8 of the regulatory guide, Westinghouse takes f

exception to the requirement that systems whose safety-related function occurs

  • during emergency or f aulted plant conditions must meet upset limits. The reduction of allowable stress to no greater than upset limits (whia.h in reality are only design limits since design, normal, and upset limits are the i same for linear supports) for support structures in those systems with normal f i safety-related functions occurring during emergency or faulted plant conditions is overly conservative for components which are not required to mechanically function (inactive components). In addition. Westinghouse r i

i believes that emergency and faulted condition criteria are acceptable for {

active components. However, when these criteria are invoked for active I

( components, any significant deformation that might occur is considered in the i f

r i evaluation of equipment operability. j i

REGULATORY GUIDE 1.130, REVISION 1, OCTOBER 1978,  :

SERVICE LIMITS AND LOADING COMP'. NATIONS FOR CLASS 1 PLATE- AND SHELL-TYPE COMPONENT SUPPORTS j Conform, except as indicated below.

Paragraphs C.3, C.4(a), and C.6(a) of the regulatory guide state that the l allewable buckling strength should be calculated using a design margin of- 2 f i

for flat plates and 3 for shells for normal, upset, and emergency conditions.  !

l  !

O l U  !

~

LdAPWR-RCS 1.8-23 JUNE, 1984 1086e:1d

. f TABLE 1.8-2 (continued) ,

In the design of plate- and shell-type supports, member combressive axial loads shall be limited per the requirements of paragraph C.3 .for normal, O

V upset, and emergency conditions.

i In paragraph C.6 of the regulatory guide, inclusion of the upset plant ,

condition is inappropriate in the load combination under discussion.

Westinghouse does not include the upset plant condition in this combination, i

In paragraphs C.6(a) and B.1 of the regulatory guide, the stress limits of  ;

F-1370(c) are discussed. The criterion states in F-1370(c), "

. . . loads should  ;

not exceed 0.67 times the critical buckling strength of the support..." In  !

the design of plate- and shell-type component supports, member compressive axial loads shall be limited to 0.67 times the critical buckling strength. ,.

- If, as a result of a more detailed evaluation of the supports the member ,

i compressive axial loads can be shown to safety exceed 0.67 times the critical buckling for the faulted condition, verification of the support function  ;

adequacy will be documented and submitted to the NRC for review. The member compressive axial loads will not exceed 0.67 times the critical buckling strength without NRC acceptance.

In paragraph C.6(b) of the regulatory guide, the limit based on the test load given in the regulatory guide T.L. x 0.7 S'/S, is overly conservative ard is  ;

inconsistent with AiME Code requirements presented in Appendix F.

l The provisions of F-1370(c) to determine service level 0 allowable loads are  !

used for supports designed by the .oad rating method. l REGULATORY GUIDE 1.133. SEPTEMBER 1977, LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF ,

LIGHT-WATER-COOLED REACTORS j Westinghouse has taken a position which takes exception to any need for i regulatory guidance relative to loose parts monitoring. This position is t

i WAPWR-RCS 1.8-24 JUNE, 1984 1086e:1d  :

. .-- ...-_---.-.----.---.--.-_---D

- , . . =- - .- - _- -

l TABLE 1.8-2 (continued)  !

contained in a letter to the secretary of the Consnission datgd December 20,  !

1977. In response to the Westinghouse arguments, echoed by the majority of  !

NSSS suppliers, the NRC draf ted major revisions to Regulatory Guide 1.133, one f of which deleted Class lE requirements for loose parts monitoring systems j (LPMS): Westinghouse prefers metal impact monitoring systems (MIMS).  !

However, until Regulatory Guide 1.133 is issued in revised form, our position f is as follows:  :

i '

Westinghouse concurs with the desirability of implementing a loose parts l

detection program in its nuclear power plants. Experience has shown the  ;

system's merits, functional adequacy, and reliable perfonnance in operating ,

t. plants. Operating history has proven that there can be early warning benefits f from experienced laterpretation of system data. It is believed that nuclear i plant owners and ' operators ' derive an additional increment of reassurance I attributable to Westinghouse MIMS. However, Westinghouse takes exception to f

requirements which go beyond the need for a reliable system which provides C

( anything more than this basic reassurance. These exceptions are delineated as i

[

- follows: I i,

o The functional performance requirements for this system serve solely  !

to provide an alert, by impact monitoring, for circumstances which could result fro.n loose metallic objects in the primary system. The system does not maintain the RCPB, serves no automatic reactor protection functions, and does not classify as a Class lE system as  !

defined in IEEE Std. 308. Consequently, the application of Class lE O criteria is unjustified.  ;

j o Based on the first exception above, the requirements of Regulatory [

Guide 1.100, Seismic Qualification of Electric Equipment for Nuclear l O Power Plants, are not applicable.

[

o Based on the first exception above, the redundancy and separation  ;

requirements for the system are not applicable, except for [

in-containment hardware where good engineering practice prevails.

WAPWR-RCS 1.8-25 JUNE, 1984 1086e:ld j t

. i

~ , - --v.- .~.-,-,,.,n-,-n,n.-, - _ . , . . . - - - ,

i e

TABLE 1.8-2 (continued) o The use of system noise to provide functional tests is 4dequate. The j need for calibrated simulated signals or in-containment calibration is l l unduly restrictive and unnecessary. Impact energy sensitivity is  ;

verifiable from system noise signatures.  :

o. Based on the first exception above, the limiting condition for operation imposed by regulatory position C.S.b is not applicable. The availability of a non-Class 1E system is not essential for continued }

l safe operation. .

1 o The implementation of the method described in Regulatory Guide 1.133, j i Section D, is a backfit as defined by 10 CFR 50.109. Based on the first exception above, the backfit requirement cannot be justified. l l l t  ;

o Westinghouse also take's exception to technical requirements not i L

{

enumerated here but which include, as an example, the restriction of i

the background noise level of the primary system in a nuclear plant to j 20 percent of an arbitrary system sensitivity level simultaneously l established by the guido. [

f I

! Refer to Subsection 4.4.6.4 for further description.

i l b lO t

I I

t O

MAPWR-RCS 1.8-26 JUNE,1984 f 1086e:1d f

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