ML20094D187

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Nonproprietary Chapter 1 of RESAR-SP/90 Westinghouse Advanced PWR Module 5, Reactor Sys, Introduction & General Description of Plant
ML20094D187
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Site: 05000601
Issue date: 07/31/1984
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WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

The Westinghouse Electric Corporation (hereinafter referred to as Westinghouse) has developed this Reference Safety Analysis' Report (RESAR-SP/90) for the Westinghouse Advanced Pressurized Water Reactor (WAPWR) as part of its continuing ef forts toward design and licensing standardization of nuclear power plants.

RESAR-SP/90 is a standard safety analysis report submitted initially for Preliminary Design Approval (PDA) in accordance with Appendix 0, " Standardization of Design; Staf f Review of Standard Designs," to Part 50 of Title 10 of the Code of Federal Regulations (hereinaf ter referred i

to as 10CFR).

The ultimate objective is to obtain a Final Design Approval i

(FDA) of RESAR-SP/90 followed by a

rulemaking proceeding and design certification.

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O 1.2 GENERAL PLANT DESCRIPTION 1.2.2 Prir.cipal Desian Criteria O

RESAR-SP/90 is designed to comply with 10CFR Part 50, Appendix A.

" General Design Criteria for Nuclear Power Plants."

The spt.c i f ic applications of General Design Criteria to RESAR-SP/90 are discussed in Section 3.1 of PDA Module i,

" Structural / Equipment Design."

Those General Dasign Criteria j

applicable to this module are listed in Section 3.1 of this module.

1.2.3 Plant Description 1.2.3.1 Reactor System The WAPWR reactor system consists of the equipment and components constituting the operating nuclear reactor.

It includes the reactor vessel, integrated head, reactor internals, control rod drive mechanisms, displacer rod drive O

mechanisms and the reactor core; including fuel assemblies, water displacer rod assemblies, gray rod assemblies, and rod cluster control assemblies.

The primary features of the WAPWR reactor design are:

1)

LOW POWER DENSITY - Low power density refers to the significantly reduced power density in this core design compared to other contemporary PWR core designs.

The WAPWR core is increased in diameter, contains more fuel rods (19x19 fuel array), and has more weight of fuel.

The additional fuel O

V loading results in significant reductions in specific power (kw/kg),

~

2 average linear power, and average rod heat flux (Btu /hr-f t ).

The lower average linear power reduces peak clad temperature in a large break Loss of Coolant Accident (LOCA) significantly.

The low average rod heat flux d

provides additional DNB margin.

For a given burnup, the increase in f uel loading reduces the f raction of the total core loading which must be replaced at the end of a fuel cycle.

WAPWR-RS 1.2-1 JULY, 1984 1476e:1d

The result for the same energy extraction is a reduction in the required f eed enrichment.

The low power density results in a lower cycle burnup (MWD /MTU) because of the additional fuel loading, which increases the number of zones or reduces the fraction of the core replaced.

This O_

results in a lower core average burnup at end of the cycle which reduces the required feed region enrichment.

2) MODERATOR CONTROL SYSTEM - The moderator control concept controls excess reactivity by varying the amount of moderator in the core instead of using control p# sons for neutron absorption.

This control of reactivity is achieved by displacing water volume in the fue.1 lattice during the first part of the fuel cycle and returning it later in the cycle as needed.

With less water in the lattice, less neutron nederation occurs and neutrons remain at resonant energies for a longer period of time, thus increasing neutron absorotion in the fertile naterial, U-238, and producing more plutonium.

When additional reactivity is require'd later in the cycle, displacer rods are removed, thereby increasing the water content of the fuel lattice, increasing neutron moderation, and reducing the proiability of fertile capture which results in the depletion of the plutonium produced earlier in the cycle.

The end result is that the amount of fissile uranium and plutonium remaining at end of life is about the same as in a poison-controlled core; however, the initial core feed enrichment is much lower, which results in an additional savings in ore and enrichment (separative work) requirements.

Physically, the core water content is varied by inserting or withdrawing banks of Zircaloy-clad rods called water displacer rods which contain (a,c) [

]

The primary ef fect of these rods on core reactivity is the displacement of water, as they have a very low neutron absorption probability.

3)

RADIAL NEUTRON REFLECTOR - The radial neutron reflector consists of a close-packed array of stainless steel rods assembled in enclosures and located on the core periphery.

It replaces the current baffle-former O

WAPWR-RS 1.2-2 JULY, 1984 1476e:1d e.

structure located between the barrel and the fuel.

Its benefit is a reduction 'in net neutron leakage which increases core reactivity and reduces feed enrichment requirements.

The result is a substantial savings o

in ore use with a potential for increased benefit with a low leakage fuel management scheme.

The reflector design also helps to reduce reactor vessel fluence levels.

The advanced reactor core utilized U-238 enriched with approximately [

(a,c) v

] weight percent of U-235.

Fuel rods are comprised of stacked ceramic UO pcliets clad in Zircaloy tubing, with an active fuel length of [

] (a,c) 2 inches.

The fuel rods are arranged in a 19x19 array to make up the fuel assembly as shown in Figure 1,2-1.

1 The WAPWR reactor vessel (which is. described in detail in Section 5.3 of RESAR-SP/90 PDA Module 4,

" Reactor Coolant System") is a large cylindrical pressure vessel with a welded hemispherical bottom head and removable flanged and gasket upper head (see Figure 1.2-2) which functions to contain and O

support the operating reactor core, and to provide for insertion and removal of the components and instrumentation used to control reactor power level and monitor reactor core operation.

Specifically, it houses the core, core support structures, rod cluster control assemblies, displacer rod assemblies and other components directly associated with the core.

The rod cluster control assemblies, displacer rod assemblies and gray rod assemblies are operated by sealed drive rod mechanisms mounted on the vessel head.

The vessel head has 185 penetrations arranged in a square pattern to acconnodate the two types of drive rod mechanisms.

Sixty-one nozzles penetrating the q

bottom head provide for connection of the bottom mounted in-core l

instrumentation conduits.

The integrated head package (IHP) is a system that combines the head lifting rig, mechanism seismic supports, lift rods, reactor vessel missile shield, CRDM cooling system, and the power and instrumentation cabling into the efficient package.

Mounted directly on the reactor vessel head, the system minimizes the time, manpower, and radiation dosage associated with head removal and replacement during a refueling (see Figure 1.2-3).

WAPWR-RS 1.2-3 JULY, 1984 1476e:ld

The WAPWR reactor utilizes a control element (either a rod control cluster, gray rod cluster, or water displacer rod cluster) over 185 of the 193 fuel assemblies.

Therefore, a rod drive mechanism is required to move each of O

those 185 control elements.

The control rod clusters and gray rod clusters which are used in control of reactor power and for shutdowns are positioned using the conventional, magnetic jack type drive mechanism.

They' provide stepwise movement of the control rods.

The 88 water displacer rod clusters are positioned either f ully inserted or fully withdrawn f rom the core by means of a hydraulic mechanism called a displacer rod drive mechanism (DRDM).

The DRDM is composed of a pressure housing, a hydraulic cylinder, the mechan' cal i

latching device and a vent system (see Figure 1.2-4).

The reactor internals for the WAPWR perf orm functions similar to those in conventional pressurized water reactors:

core support, flow direction, and guidance and protection of control rods.

The WAPWR reactor internals however have additional functions since gray rods and water displacer rods are employed in addition to control rods.

The similarity of the WAPWR and the 4XL Models is illustrated in Figure 1.2-5, which shows the equivalent inlet nozzle, downcomer, lower plenum, and upper plenum or calandria regions.

The two designs are similar except for changes in the region from the upper core plate to the outlet nozzle.

Because of the increased number of control elements that must be moved in the rod travel space, a new calanaria structure is provided above the rod guide region to turn the flow to the outlet nozzles.

This approach provides for axial flow in the rod guide region thereby minimizing the potential for flow induced O

vibration.

The addition of a calandria at the outlet nozzle elevation results in a longer reactor vessel.

The ' upper core plate is much thicker to accept axial loading permitting the elimination of support columns.

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Figure 1.2-1 Fuel Assembly Outline (PROPRIETARY)

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i 1.3 COMPARISON TABLES t

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1. 3.1 Comparison With Similar Facility Desians 1

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Table 1.3-1 presents a design comparison of the major parameters and features of the WAPWR reactor system with RESAR-414 (Docket No. STN-50-572; 'PDA-13),

l RESAR-35 (Docket No. 'STN-50-545; POA-7), and RESAR-41 (Docket No. STN-50-480; I

PDA-3).

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TA8LE 1.3-1 I

%Is DESIGN COMPARISON i

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P. P l

E" E Parameter or Feature RESAR-SP/90 RESAR-414 RESAR-35 RESAR-41 l

Reactor core heat output (MWt) 3800 3000 3411 3800 1

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l Total thermal flow rate (10 lb/hr) 150.6 150.5 140.3 144.7

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Reactor coolant system temperatures (*F) 1 1.

Core outlet 625.4 627.4 620.8 626.6 i

2.

vessel outlet 621.1 624.9 618.3 624.0 3.

Core average 596.2 597.5 589.4 593.2 4.

Vessel average 590.1 596.1 588.2 591.8 5.

Core inlet 559.1 563.8 558.1 559.8 1

1 6.

Vessel inlet 559.1 563.8 558.1 559.8 i

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Average linear power (kW/f t) 5.06 5.20 5.44 5.33 i

Peak linear power for normal operation (kW/ft) 13.2 14.0 12.6 13.3 f

i Heat flux hot channel factor F 2.60 2.70 2.32 2.50

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1 Fuel assembly array 19 x'19 17 x 17 17 x 17 17 x 17

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Number of fuel assemblies 193 193 193 193 2

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TABLE 1.3-1 (cont) h DESIGN COMPARISON

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Parameter or Feature RESAR-SP/90 RESAR-414 RESAR-35 RESAR-41 i

Uranium dioxide rods per assembly 296 264 264 264 1

Fuel weight as uranium dioxide (lb)

_(a,c) 259,860 222,739 253,675 I

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Number of grids per assembly 10 9-Type R 8-Type R 9-Type R (8 Zircaloy, i

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2 Inconel)

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Rod cluster control assemblies i

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Number of full /part length 69/0 57/0 53/8 61/8 i

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Absorber material l~

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Clad material Stainless Stainless Stainless Stainless l

Steel Steel Steel Steel

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Clad thickness (in.)

0.063 0.0385 0.0185 0.0185 l

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Equivalent core diameter (in.)

156.7 132.7 132.7 132.7

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Active fuel length (in.)

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'l Fuel enrichments (weight percent)

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Region 1 1.60 2.10 2.10 cp 2.

Region 2 2.40 2.60 2.60 l

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Region 3 3.10 3.10 3.10 i

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1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION

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The purpose of this section is to present a description of the safety related I

research and development programs which are being carried out for, or by, or in conjunction with, Westinghouse and which are applicable to the RESAR-SP/90 scope.

Each of the research and development programs applicable to the reactor system are described below.

This description includes a sumary of the program i

purpose, pertinent results to date, the facility in which the testing is l

performed (if applicable), and the status of the research and development effort.

j The technical information generated by these research and development programs I

will be used either to demonstrate the safety of the design and more sharply i

define margins of conservatism, or will lead to design improvements.

Progress in these development programs is reported on a timely basis.

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safety related research and development programs will also be described in future amendments to RESAR-SP/90, as appropriate.

1.5.1 Fuel System Tests i

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1. 5.1.1 Fuel Assembly Tests l

l 1.5.1.1.1 Fuel Assembly Structural Tests Test Purpose The purpose of these tests is to provide lateral and axial stiffnesses, O

vibratory characteristics and impact strengths of the prototype WAPWR fuel assembly.

O WAPWR-RS 1.5-1 JULY, 1984 1476e:1d

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Results (Later)

O Test Facility These tests will be conducted at the Mitsubishi Atomic Power Industries (MAPI)

Nuclear Development Facility.

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Status The fuel assembly is being manufactured and the test fixtures are being

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constructed.

1. 5.1.1. 2 Fuel Assembly Hydraulic Flow Test I

Test _PurDose The purpose of this test is to obtain fuel assembly pressure drops, lift forces, and fuel rod and assembly vibration and f retting wear data with a prototype WAPWR fuel assembly in a full flow loop.

I Results (later)

Test Facility This test will be conducted at the Westinghouse Fuel Assembly Test System (FATS) hydraulic loop at the Forest Hills Laboratory; Forest Hills, PA.

Status The fuel assembly and loop internals are currently being manufactured.

MAPWR-d5 1.5-2 JULY, 1984 1476e:ld

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1.5.1.2 Core Components Tests 1.5.1.2.1 Rod Cluster Control and Gray Rod Assembly Tests i

Test PurDose I

The purpose of these tests is to submit prototype WAPWR drive line components to simulated plant conditions to determine hydraulic, vibratory and wear characteristics.

Results (Later)

I Test Facility i

These tests will be conducted at the Westinghouse D-Loop test f acility at the O

Forest Hills Laboratory; Forest Hills, PA.

Status The rod clusters and loop internals are currently being manuf actured.

1.5.1.2.2 Water Displacer Rod Assembly Tests Test Purpose The purpose of these tests is to submit prototype drive line components to simulated plant conditions to determine hydraulic,

vibratory and wear characteristics.

Results l

(Later)

WAPWR-RS 1.5-3 JULY, 1984 1476e:1d

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Test Facility These tests will be conducted at the Mitsubishi Atomic Power Industries (MAPI)

Nuclear Development Facility.

Status The test loop is currently being constructed.

1.5.1.3 Orive Mechanism Tests i

1.5.1.3.1 Control Rod Drive Mechanism Tests Test PurDose The purpose of these tests is to measure the electrical, mechanical, hydraulic l

and thermal performance of the electro-mechanical stepping mechanism and its rod position indicator system in drive line operation representing accelerated design lifetime in an operating plant.

The testing program includes 6

approximately 230 control rod drops and 7.5 x 10 mechanism steps.

Rod position readout, rod drop time, electrical and mechanical operation of latching and unlatching, mechanism operating temperature and heat load measurements are made during cycling followed by disassembly for inspection and wear measurements.

Results (later)

Test Facility O

i These tests will be conducted at the Westinghouse 0-Loop test facility at the Forest Hills Laboratory, Forest Hills, PA.

Prototype hardware of an entire control rod drive line will be tested in the loop at operating plant pressure,

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temperature, drive line flow rate and water chemistry.

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Testing is to be conducted in 1985.

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t Test Purcose i

O The purpose of these tests is to measure the mechanical and hydraulic i

performance of the hydraulic mechanism and its rod position indicator system in drive line operation representing accelerated design life time in an i

operating plant.

The testing program includes approximately 200 insertion and withdrawal cycles.

Rod position readout, insertion and withdrawal velocity, and piston ring pressure drop and leakage rate will be measured during cycling

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followed by disassembly for inspection and wear measurements.

Results (Later) l Test Facility These tests will be perf ormed at the D-Loop facility at Mitsubishi, Takasago Institute in Japan.

Prototype hardware of an entire displacer rod drive line will be tested in the loop at operating plant pressure, temperature, drive line flow rate and water chemistry.

Status Testing is to be conducted in 1986.

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1.5.2 Reactor Internals Design Verification Tests Test Purpose r

i The purpose of these tests is to measure mechanical and hydraulic performance and operating characteristics of the interaction of internals compo'nents at representative plant operating conditions.

The primary areas of investigation l

include overall drive line performance, hydraulic losses, flow and pressure distribution, heat transfer, flow induced vibration and hydraulic loads, operating natural f requencies and component wear.

Testing programs include full scale, partial and complete driveline assemblies and a sub-scale model of the reactor internals.

Results (later)

Test Facility Test loops at Westinghouse, Forest Hills Laboratory and Mi_tsubishi, Takasago Institute in Japan will be used.

Prototype and sub-scale model components will be tested at representative operating plant pressure, temperature, flow rate and water chemistry.

i Status l

Testing to be conducted in 1985 and 1986.

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i lO WAPWR-RS 1.5-6 JULY, 1984 l

1476e:ld 1

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l.6 MATERIAL INCORPORATED BY REFERENCE I

The WAPWR reactor system module incorporates, by reference, certain topical reports.

The topical

reports, listed in Table 1.6-1, have been filed previously in support of other Westinghouse applications.

The legend for the review status code letter follows:

A U.S.

Nuclear Regulatory Commission review complete; USNRC acceptance letter issued.

AE -

U.S.

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

Submitted to UNRC as background information; not undergoing formal B

USNRC review.

l On file with UShRC; older generation report with current validity; not 0

I actively under formal USNRC review.

Actively under formal USNRC review.

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WAPWR-RS 1.6-1 JULY, 1984 1476e:1d

O TABLE 1.6-1 V

MATERIAL INCORPORATED BY REFERENCE 4

Westinghouse SAR Tcpical Revision Section Submitted Review Report No.

Title Number Reference to the NRC Status WCAP-2048 The Doppler Effect for a Rev 0 4.3 7/62 0

Non-Uniform Temperature Distribution in Reactor Fuel Elements WCAP-2850-L(P) Single-Phase Local Boiling Rev 0 4.4 5/66 0

WCAP-7916 and Bulk Boiling Pressure Drop Correlations WCAP-2923 In-Pile Measurement of Rev 0 4.4 3/66 0

U02 Thermal Conductivity WCAP-3269-8 Hydraulic Tests of the Rev 0 4.4 6/64 0

San Onofre Reactor Model WCAP-3269-26 LEOPARD - A Spectrum Rev 0 4.3, 15.4 9/63 0

Dependent Non-Spatial Depletion Code for the O-18M-7094 WCAP-3385-56 Saxton Core II Fuel Rev 0 4.3 7/70 0

Performance Evaluation i

WCAP-3385-56, Part II, Evaluation of Mass Spectrometric and Radiochemical Materials Saxton Plutonium Fuel WCAP-3680-20 Xenon-Induced Spatial Rev 0 4.3 3/68 0

i Instabilities in Large O

Pressurized Water Reactors (EURAEC-1974)

WCAP-3680-21 Control Procedures for Rev 0 4.3 2/69 0

Xenon-Induced X-Y Instabilities in Large O

Pressurized Water Reactors (EURAEC-2111)

WCAP-3680-22 Xenon-Induced Spatial Rev 0 4.3 9/69 0

Instabilities in Three Dimensions (EURAEC-2116)

WAPWP-RS 1.6-2

, JULY, 1984 1476e:ld

TABLE 1.6-1 (cont)

MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review Report No.

Title Number Reference to the NRC Status WCAP-3966-B Pressurized Water Rev 0 4.3 10/68 0

Reactor PH Reactivity Effect Final Report (EURAEC-2074)

WCAP-3726-1 PUO2 -UO2 Fueled Critical Rev 0 4.3 7/67 0

Experiments WCAP-6065 Melting Point of Irra-Rev 0 4.4 2/65 0

diated UO2 WCAP-6069 Burnup Physics of Rev 0 4.4 6/65 0

Heterogeneous Reactor Lattices WCAP-6073 LASER - Depletion Rev 0 4.3 4/66 0

Program for Lattice Calculations Based on MUFT and THERMOS WCAP-6086 -

Supplementary Report Rev 0 4.3 8/69 0

on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, including Isotopes of Elements Thorium through Curium WCAP-7015 Subchannel Thermal Rev 1 4.4 2/14/69 0

Analysis of Rod Bundle 4

Cores WCAP-7048-PANDA Code Rev 0 4.3 1/9/75 A

P-A(P)

WCAP-7757-A 4

p WCAP-7208(P)

Power Distribution Control Rev 0 4.3 9/68 0

(

WCAP-7811 of Westinghouse Pressurized Water Reactors WCAP-7213-TURTLE 24.0 Diffusion Rev 0 4.3 1/9/75 A

P-A(P)

Depletion Code WCAP-7758-A

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WAPWR-RS 1.6-3 JULY, 1984 1476e:1d

l TABLE 1.6-1 (cont) d MATERIAL INCORPORATED BY REFERENCE -

Westinghouse SAR Topical Revision Section~

Submitted Review ReDort No.

Title Number Reference to the NRC Status WCAP-7267-L(P) Core Power Capability Rev 0 4.3 10"?

O WCAP-7809 in Westinghouse PWRs WCAP-7308-L(P) Evaluation of Nuclear Hot Rev 0 4.3 1/9/70 V

WCAP-7810 Channel Factor Uncertainties 12/16/71 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

A(P)

New Mixing Vane Grids (R)

WCAP-7958-A WCAP-7706-L(P) An Evaluation of Solid State Rev 0 4.6

,S/2/71

.O WCAP-7706 Logic Reactor Protection in Anticipated Transients WCAP-7800 Nuclear Fuel Division Rev SA 4.2 11/20/79 A

Quality Assurance Program Plan

/O WCAP-1907-P-A LOFTRAN Code Description Rev 0 15.0, 15.4 10/11/72 A

V WCAP-7908 FACTRAN - A FORTRAN-IV Rev 0 15.0, 15.4 9/20/72 U

Code for Thermal Transients in a U02 Fuel Rod Power Peaking Factors Rev 0 4.3 1/16/75 A

O WCAP-7912-P-A(P)

WCAP-7912-A O

WAPWR-RS 1.6-4 JULY, 1984 1476e:1d

1 1

TABLE 1.6-1 (cont)

O MATERIAL INCORPORATED BY REFERENCE 5

Westinghouse SAR Topical Revision Section Submitted Review Report No.

Title Number Reference to the NRC Status WCAP-7941-Effect of Axial Spacing On Rev 0 4.4 1427/75 A

P-A(P)

Interchannel Thermal Mixing WCAP-7959-A with the R Mixing Vane Grid

]

WCAP-7956 THINC-IV - An Improved Rev 0 4.4 10/22/73 A

Program f or Thermal-Hydraulic Analysis of Rod Bundle Cores WCAP-7964 Axial Xenon Transient Rev 0 4.3 6/15/71 0

Tests at Rochester Gas and Electric Reactor WCAP-7979-TWINKLE - A Multidimen-Rev 0 15.0, 1/7/75 A

P-A(P) sional Neutron Kinetics 15.4 WCAP-8028-A Computer Code WCAP-7988-Application of Modified Rev 0 4.4 1/75 A~

/

P-A(P)

Spacer Factor to L-Grid WCAP-8030-A Typical and Cold Wall Cell DNB WCAP-8054(P)

Application of THINC-IV Rev 0 4.4 12/7/73 A

WCAP-8195 Program to PWR Design 1/11/74 WCAP-8174 Effect of Local Heat Flux Rev 0 4.4 2/75 A

P-A(P) on DNB in Nonuniformly WCAP-8202-A Heated Rod Bundles WCAP-8183 Operational Experience Rev 12 4.2 8/83 B

with Westinghouse Cores q

(up to December 31, 1982)

WCAP-8218 Fuel Densification Experi-Rev 0 4.2, 4.3, 3/6/75 A

P-A(P) mental Results and Model 4.4 WCAP-8219-A for Reactor Application n

WCAP-8298-The Effect of 17 x 17 Rev 0 4.4 1/28/75 A

f]

P-A(P)

Fuel Assembly Geometry on WCAP-8299-A Interchannel Thermal Mixing WCAP-8305 LOCTA-IV Program:

Rev 0 15.0 6/74 AE Loss of Coolant Transient Analysis 1476e:1d

,,__-------.._,..,r-..,

p TABLE 1.6-1 (cont)

V MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review Report No.

Title Number Refereng to the NRC Status WCAP-8306 SATAN-VI Program:

Compre-Rev 0 15.0 7/12/74 AE hensive Space-Time Depen-dent Analysis of Loss-of-Coolant WCAP-8330 Westinghouse Anticipated Rev 0 4.3, 4.6 9/25/74 U

Transients Without Trip 15.4 Analysis WCAP-8359 Effects of Fuel Densifi-Rev 0 4.3 7/2/74 AE cation Power Spikes on Clad Thermal Transients WCAP-8370 Westinghouse Water Reactor Rev 9A 17 11/14/77 A

Divisions 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 Load Following Procedures WCAP-8453-A(P)

Analysis of Data from Rev 0 4.4 5/10/76 A

WCAP-8454 Zion (Unit 1) THINC Veri-fication 7est WCAP-8498 Incore Power Distribution Rev 0 4.3 7/22/75 U

Determination in Westing-house Pressurized Water Reactors, Program Summaries -

Fall 1974 WCAP-8567-P(P)

Improved Thermal Design Rev 0 4.4, 15.0 7/75 A

WCAP-8568 Procedure WCAP-8575(P)

Augmented Startup and Rev 0 4.3 6/76 U

WCAP-8576 Cycle 1 Physics Program Supplement 1 WCAP-8584(P)

Failure Mode and Effects Rev 0 4.6 4/23/76 U

WCAP-8760 Analysis (FMEA) of Engi-Rev 1 2/80 neered Safeguard Features WAPWR-RS 1.6-6 JULY, 1984 1476e:ld

.. - -. ~. -.

l l

O TABLE 1.6-1 (cont)

MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review ReDort No.

Title Number Reference to the NRC Status WCAP-8682(P)

Experimental Verification Rev 0 4.3 3/18/76 B

WCAP-8683 of Wet Fuel Storage Criticality Analyses l

WCAP-8691(P)

Fuel Rod Bowing Rev 1 4.2, 4.4 7/79 A

WCAP-8692 WCAP-8708-P-A MULTIFLEX - FORTRAN-IV Rev 0 3.9 9/16/77

-A (P), Volumes Computer Program for Ana-I and 11 lyzing Thermal-Hydraulic WCAP-8709-A, Structure System Dynamics Volumes I & II WCAP-8720(P)

Improved Analytical Models Rev 0 4.2 11/2/76 A

WCAP-8785 Used in Westinghouse Fuel Rod Design Computations WCAP-8762(P)

New Westinghouse Correlation Rev 0 4.4 7/76 A

O WCAP-8763 WR8-1 for Predicting i

Critical Heat Flux in Rod l

Bundles with Mixing Vane Grids WCAP-8768 Safety-Related Research Rev 2 4.2, 4.3 9/28/78 B

and Development for West-inghouse Pressurized Water l

Reactors, Program Summaries -

Winter 1977 through Summer 1978 WCAP-8768 Safety-Related Research Rev 0 4.3 6/17/76 B

's and Development for Westinghouse Pressurized x

Water Reactors, Program Summaries - Spring 1976 Hybrid 8 C Absorber Rev 0 4.2, 15.0 10/77 A

WCAP-8846-A 4

Control Rod Evaluation O

Report WCAP-8963(P)

Safety Analysis for the Rev 0 4.2 3/31/71 A

WCAP-8964 Revised Fuel Rod Internal 8/11/77 Pressure Design Basis WAPWR-RS 1.6-7 JULY, 1984 1476e:1d

TABLE 1.6-1 (cont)

MATERIAL INCORPORATED BY REFERENCE Westinghouse SAR Topical Revision Section Submitted Review ReDort No.

Title Number Reference to the NRC Status WCAP-8976 Failure Mode and Effects Rev 0 4.6 10/26/77 U

Analysis (FMEA) of Solid State Full-Length Rod Control System WCAP-9000-L(P) Nuclear Design of Rev i 4.3 7/69 0

Westinghouse Pressurized Water Reactors with Burnable Poison Rods WCAP-9004(P)

Inlet Orificing of Rev 0 4.4 1/17/72 B

WCAP-7836 Open PWR Cores WCAP-9105(P)

Axial Power Distribution Rev 0 4.3 7/77 U

Monitoring Using Four-Section Ex-Core Detectors WCAP-9179(P)

Properties of Fuel and Core Rev 1 4.2 8/2/78 A

O WCAP-9224 Component Materials WCAP-9485(P)

Paladon - Westinghouse Rev 0 4.3 12/78 A

WCAP-9486 Nodal Computer Code.

WCAP-10444(P)

Westinghouse Reference Rev 0 4.2, 4.3, 12/83 U

Core Report - VANTAGE 5 4.4 Fuel Assembly O

O O

WAPWR-RS 1.6-8 JULY, 1984 1476e:1d

1.8 CONFORMANCE WITH THE STANDARD REVIEW PLAN In accordance with 10CFR50.34(g),

Table

1. 8-1 identifies and evaluates deviations f rom the acceptance criteria of those sections of the NRC Standard Review Plan (NUREG-0800) pertinent to the Reactor System.

In addition a listing of NRC Division 1 Regulatory Guides pertinent to the WAPWR Re4ctor System, and a discussion of the extent to which the WAPWR complies with the regulatory positions of these Regulatory Guides is given in Table 1.8-2.

O i

O i

l O

O

!O WAPWR-RS 1.8-1 JULY, 1984 1476e:ld l

l l

f TABLE 1.8-1 STANDARD REVIEW PLAN DEVIATIONS I

SRP AcceDtance Criteria Deviation Section (To date, no deviations f rom the acceptance criteria of those SRP, sections i

applicable to the WAPWR Reactor System have been identified).

f i

i i

l O

i l

i i

i i

t'!O l

O 1O i

WAPWR-RS 1.8-2 JULY, 1984 l

1476e:1d

TABLE 1.8-2 CONFORMANCE TO US NRC REGULATORY GUIDES APPLICABLE TO THE WAPWR REACTOR SYSTEM O

REGULATORY GUIDE 1.13, REVISION 1, DECEMBER 1975, SPENT-FUEL STORAGE FACILITY DESIGN BASIS Westinghouse conforms to the regulatory position of this regulatory guide.

REGULATORY GUIDE 1.20, REVISION 2, MAY 1976, COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM FOR REACTOR INTERNALS DURING PREOPERATIONAL AND INITIAL STARTUP TESTING The comprehensive vibration assessment program for the WAPWR internals during preoperational and initial startup testing will confonn with the recommendations of this guide.

REGULATORY GUIDE 1.29, REVISION 3. SEPTEMBER 1978, SEISMIC DESIGN CLASSIFICATION The WAPWR conforms with this regulatory guide as shown in Table 3. 2 -1 of RESAR-SP/90 PDA Module 7, " Structural / Equipment Design",

l 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; 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 systems required to perfonn the same saf ety f unction as required of a safety class component which is part of that system shall be likewise qualified or granted exemption.

Conditions to be met for exemption are:

o Failure would not directly cause a Condition III or IV event (as defined in ANSI N18.2-1973).

WAPWR-RS 1.8-3 JULY, 1984 1476e:1d

,..,_._,_.__.m-

O TABLE 1.8-2 (Continued)

V o

There is no safety function to mitigate, nor could f ailure prevent mitigation of, the conseauence of a Condition III or IV event.

OV o

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

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

REGULATORY GUIDE 1.65, OCTOBER 1973, MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS The WAPWR conforms with 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 of this Code Case has been approved by the NRC via Regulatory Guide 1.85.

The maximum limit of 170 ksi ultimate tensile strength is not explicitly specified by Westinghouse as required by the guide.

Westinghouse does specify f racture toughness of 45 f t/lb and 25 mils lateral expansion as required by the ASME Code and 10 CFR 50, Appendix G.

These requirements also result in strength levels below the maximum limit, as demonstrated by the actual stud material properties for the WAPWR.

i l

REGULATORY GUIDE 1.70, REVISION 3, NOVEMBER 1978, STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS p

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

l l

\\

WAPWR-RS 1.8-4 JULY, 1984 1476e:1d

TABLE 1.8-2 (Continued)

REGULATORY GUIDE 1.77, MAY 1974, ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD EJECTION ACCIDENT FOR PRESSURIZED WATER REACTORS 9

Westinghouse conforms as discussed in Subsection 15.4.8.

REGULATORY GUIDE 1.99, REVISION 1, APRIL 1977, EFFECTS OF RESIDUAL ELEMENTS ON PREDICTED RADIATION DAMAGE TO REACTOR VESSEL MATERIALS There are two primary issues with the guide:

1.

The guide provides a procedure and curves for predicting radiation damage (as relating to the shift of the reference temperature, RTNDT), 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 lif e transition temperature to 200'F maximum.

Control of residual elements such as copper, phosphorus, sulfur, and vanadium in the reactor vessel beltline materials of new l

plants to levels that result in a predicted adjusted reference 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.

O WAPWR-RS 1.8-5 JULY, 1984 l

l 1476e:1d 4

e.

,,.,,r.,

. - - -, -.,.,, - -, - - - - - -, -, - - - - - -. - - - - - - - - - -.. ~ -, - - - - -, - - - -,. - -, -,

TABLE 1.8-2 (Continued)

One additional feature of the guide constitutes a lesser but r.evertheless important issue:

O 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.

Although -it appears that the prescribed relationship does not predict unacceptable drops in upper shelf toughness for vessels with controlled chemistry the curves are nevertheless over conservative.

The WAPWR position with respect to each of the guide positions is as follows:

1.

The basis as well as the scope of the guide for predicting adjustment of reference temperature as given in regulatory positi5n C.1 are inappropriate since the data base used was incomplete and included I

some data which were not applicable.

O 2.

The WAPWR design is consistent with the guide position C.2a.

liowever, with respect to guide position C.2b, Westinghouse believes that Figure 2 of the guide is incorrect since the upper shelf energy for 6-in.-thick American Society of Testing Materials (ASTM)

A302B l

reference correlation monitor material reported by Hawthorne indicates 1 x 10 essentially a constant upper shelf at fluences above n/cm,(a) 2 3.

The Westinghouse position with reference to the guide position C.3, controlling residual elements to levels that result in a predicted l

l adjusted reference temperature of less than 200*F at e' ' of lif e, is that the stresses in the vessel can be limited duri', operation in order to comply with the requirements of Appendix G to 10 CFR 50 even though the end-of life adjusted reference temperature may exceed 200*F.

By applying the procedures of Appendix G to ASME Section III, the stress limits including appropriate Code saf ety margin can be met.

O WAPWR-RS 1.8-6 JULY, 1984 1476e:ld

TABLE 1.8-2 (Continued) 4.

Recent surveillance capsule data indicate a steady-state condition of radiation damage well below that predicted by current trend curves.I")

This effect is believed to be due to the annealing of the vessels at the operating temperature.

As an alternative to Regulatory Guide 1.99, operating limits will be determined using the current radiation damage curves developed by Westinghouse.

It is expected that as f uture surveillance specimens are evaluated it will become increasingly evident that both the Regulatory Guide 1.99 and Westinghouse trend curves are very conservative.

REGULATORY GUIDE 1.126. REVISION 1, MARCH 1978, AN ACCEPTABLE MODEL AND RELATED STATISTICAL METHODS FOR THE ANALYSIS OF FUEL DENSIFICATION This guide states that the model presented in this guide is not intended to supersede NRC approved vendor models.

The WAPWR uses the Westinghouse model which has been approved by the NRC.

Refe'r to Subsection 4.2.3.2 for further i

discussion.

a.

Hawthorne, J.

R.,

" Radiation Effects Information Generated on the ASTM Reference.

Correlation-Monitor Steels," ASTM, Philadelphia 1974.

O O

WAPWR-RS 1.8-7 JULY, 1984 1476e:ld 1

i

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