Text

Summary of 980407 Meeting W/Util in Rockville,Md to Discuss Fuel Design Changes.List of Attendees & Presentation Slides Encl
	ML18106A570
Person / Time
Site:	Salem, Hope Creek
Issue date:	04/23/1998
From:	Milano P; NRC (Affiliation Not Assigned)
To:	; NRC (Affiliation Not Assigned)
References
	NUDOCS 9805040135
	Download: ML18106A570 (44)
	v • d • e

UNITED STt1:rES 67J-o27~P,11

~-357' NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 April 23, 1998 LICENSEE:

Public Service Electric and Gas Company Hope Creek Generating Station FACILITIES:

Salem.Nuclear Generating Station, Unit Nos. 1 and 2.

SUBJECT:

SUMMARY

OF APRIL 7, 1998, MEETING REGARDING FUEL DESIGN CHANGES This summary refers to the meeting with Public Service Electric and Gas Company (PSE&G) conducted on April 7, 1998, at the U.S. Nuclear Regulatory Commission (NRC) office in Rockville, Maryland. The meeting was held at the request of PSE&G to discuss fuel design changes for the Hope Creek Generating Station and Salem Nuclear Generating Station, Unit

Nos. 1 and 2. Enclosure 1 is a list of the attendees, and Enclosure 2 and 3 are copies of the slides presented by PSE&G.

Summary for Hope Creek Generating Station.

The PSE&G presentation closely followed the material in their slides (Enclosure 2). The presentation provided an overview of PSE&G's plans to change fuel vendors from General Electric (GE) to Asea Brown Boveri (ABB). The following major topics were discussed:

1.

Hope Creek plant status including the planned schedules and fuel design for Cycles 8, 9,

2.

3.

4.

and 10; Overview of the project including discussion of the PSE&G/ABB project team, replacement of the NSSS Process Computer, upgrade of the ABB Core Monitoring System, and application of lessons learned on fuel vendor transitions at other sites; Overview of the transition plan including current project status, near term activities, and transfer of data from PSE&G to ABB; Design of the ABB SVEA-96+ fuel that will be used including similarities to the SVEA-96 fuel being used at WNP-2, mechanical compatibility with existing plant design, thermal-

. hydraulic considerations, and nuclear design considerations; and

5.

Licensing of ABB Topical Reports including applicable topical reports that are already approved by the NRC and topical reports that will be submitted for NRC approval.

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PSE&G stated that they believe th.e replacement of the NSSS Process Computer and upgrade of the ABB Core Monitoring System will still be within the existing design basis and should not require any changes to the license (i.e., changes could be accomplished under the 10 CFR 50.59 process.)

ABB discussed their plans to submit four licensing topical reports for NRC approval. One of the topical reports (CENPD-389-P) is specific to Hope Creek and the other three topical reports (CENPD-287-P Supplement 1, CENPD-390-P, and CENPD-391-P) are applicable to both Hope Creek and WNP-2. The NRC staff expressed concerns as to whether these reviews are included in the fiscal year 1999 budget since the approvals could be affected by the priorities set in the budget planning process. A non-plant-specific meeting between ABB and the NRC is tentatively scheduled for mid-May. It was proposed that this meeting be used to set priorities on the review of the four topical reports. The NRC staff also noted* that ABB should be careful in the proprietary markings of the topical reports in order to expedite the review process.

Summary for Salem Nuclear Generating Station Unit Nos. 1 and 2 The PSE&G presentation closely followed the material in their slides (Enclosure 3). The presentation provided an overview of PSE&G's plans to begin using a new robust fuel assembly (RFA) design from Westinghouse. The following major topics were discussed:

1.

Current Salem design and performance of the Westinghouse Vantage+ fuel assemblies.

2.

Salem RFA design features incorporating Zirlo MV5H mid-grids and intermediate flow mixing grids, thicker guide tube thimbles, a protective bottom grid, and annular axial blankets. These features will improve performance with regard to flow induced vibration, corrosion and debris protection, rod internal pressure, departure from nucleate boiling margins, control rod insertion, and axial offset anomaly.

3.

Salem licensing basis and future submittal of a request to use the WRB-2 departure from nucleate boiling correlation with RFAs with intermediate flow mixing grids.

The NRC staff described the supporting information that would be required to complete its review of an application to use the WRB-2 correlation. The supporting information should include:

(a) references to previously approved topical reports, (b) the background regarding the use of intermediate flow mixing grids at Salem, (c) a statement or evidence that supports the application of WRB-2 to the Salem RF As and intermediate flow mixing grids, (d) a discussion regarding RFA features conformance with the Westinghouse topical report on fuel criteria evaluation, (e) assurance that the Salem design is within the range of parameters for WRB-2,

3 -

and (f) a specific discussion that addresses each of the conditions imposed when WRB-2 was approved generically by the NRC. The licensee also stated that it would request NRC staff approval for WRB-2 in time to support the September 1998 fuel fabrication schedule.

/SI Patrick D. Milano, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1. Attend~nce List

2. PSE&G Slides for Hope Creek

3. PSE&G Slides for Salem Docket Nos. 50-272, 50-311, and 50-354 cc w/encls: See next page DISTRIBUTION:

. Docket Fiie"-::i PUBLIC E-MAIL:

SCollins/FMiraglia BBoger JZwolinski (JAZ)

RCapra PDl-2 Reading File JMoore, OGC TClark (TLC2)

MChatterton AAttard HRichings EKendrick OFFICE PDl-2/PM PDl-3/LA NAME REnnis ~lj:J3 c..

TClaridJ\\C..,

DATE 04 h.\\ /98 040--0 /98 04/J,{)/98 OFFICIAL RECORD COPY DOCUMENT NAME: HC&SA4-7.MTS 04t:J./t98 ACRS PMilano REnnis Jlinville, RGN-1 SPindale, Hope Creek SRI SMorris, Salem SRI PDl-2/D * \\

RCa ra RO-> \\

04/2! /98

3 -

and (f) a specific discussion that addresses each of the conditions imposed when WRB-2 was approved generically by the NRC. The licensee also stated that it would request NRC staff approval for WRB-2 in time to support the September 1998 fuel fabrication schedule.

Enclosures:

1. Attendance List

2. PSE&G Slides for Hope Creek

3. PSE&G Slides for Salem Docket Nos. 50-272, 50-311, and 50-354

cc w/encls: See next page Patrick D. Milano, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects - 1111 Office of Nuclear Reactor Regulation

l Public Service Electric & e Company cc:

Jeffrie J. Keenan, Esquire Nuclear Business Unit - N21 P.O. Box236 Hancocks Bridge, NJ 08038 General Manager - Salem Operations Salem Nuclear Generating Station P.O. Box236 Hancocks Bridge, NJ 08038 Mr. Louis Storz Sr. Vice President - Nuclear Operations Nuclear Department P.O. Box236 Hancocks Bridge, NJ 08038 Senior Resident Inspector Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN415 Trenton, NJ 08625-0415 Maryland Office of People's Counsel 6 St. Paul Street, 21st "Floor Suite 2102 Baltimore, MD 21202 Ms. R. A. Kankus Joint Owner Affairs PECO Energy Company 965 Chesterbrook Blvd., 63C-5 Wayne, PA 19087 Mr. Elbert Simpson Senior Vice President-Nuclear Engineering Nuclear Department P.O. Box236 Hancocks Bridge, NJ 08038 Hope Creek Resident Inspector U.S. Nuclear Regulatory Commission Drawer0509 Hancocks Bridge, NJ 08038 Salem NuclAenerating Station, Units 1 and!";nd Hope Creek Generating Station Richard Hartung Electric Service Evaluation Board of Regulatory Commissioners 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Lower Alloways Creek Township.

c/o Mary 0. Henderson, Clerk

Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Manager-Licensing and Regulation Nuclear Busienss Unit - N21 P.O. Box236 Hancocks Bridge, NJ 08038 Mr. David Wersan Assistant Consumer Advocate Office of Consumer Advocate 1425 Strawberry Square Harrisburg, PA 17120 Manager - Joint Generation Atlantic Energy 6801 Black Horse Pike Egg Harbor Twp., NJ 08234-4130 Carl D. Schaefer External Operations - Nuclear Delmarva Power & Light Company P.O. Box231 Wilmington, DE 19899 Public Service Commission of Maryland Engineering Division Chief Engineer 6 St. Paul Centre Baltimore, MD 21202-6806 General Manager - Hope Creek Operations Hope Creek Generating Station P.O. Box236 Hancocks Bridge, NJ 08038 Mr. Harold W. Keiser Executive Vice President-Nuclear Business Unit Public Service Electric and Gas Company Post Office Box 236

MEETING ATTENDANCE LIST Licensee:

Plant(s):

Public Service Electric and Gas Company Hope Creek: Salem. Units 1 and 2

Subject:

Fuel Design Changes Date: April 7. 1998 Time: 10:00 a.m.

Location: NRC Offices. OWFN Room 1-F-5 NAME NRC STAFF P. Milano R. Ennis M. Chatterton A. Attard H. Richings E. Kendrick PSE&G D. Notigan F. Safin G. Salamon T. Ross C. Smyth G. Schwartz D. Ebeling-Koning W. Harris C. Brinkman Westinghouse W. Rinkacs P. Schueren R. Buechel Sr. Project Manager Project Manager Acting Section Chief Reactor Engineer Sr. Reactor Engineer Reactor Engineer Supervisor Principal Engineer Manager Supervisor Manager Senior Engineer Supervisor, BWR Reload Analysis.

Engineer, BWR Reload Analysis Director, Nuclear Licensing Fuel Project Engineer Fuel Licensing Engineer Product Design Manager ORGANIZATION NRR/DRPE/PD1-2 NRR/DRPE/PD1-2 NRR/SRXB NRR/SRXB NRR/SRXB NRR/SRXB Hope Creek Fuels Hope Creek Fuels Hope Creek Licensing Salem Fuels Salem Licensing Salem Fuels

I~*

PSE&G Reload Fuel Transition Presentation

1 Don Notigan - PSE&G Derek Ebeling-Koning - ABB Bill Harris - ABB NRC Meeting April 7, 1998

Meeting Agenda

!* Purpose

- Present PSE&G/ABB Plans for Reload Fuel Transition

- Receive NRG Staff Comments Early in Project

!* Presentation Outline

Public Service Gas & Electric \\fl 2

- Hope Creek Plant Status

- Reload Fuel Project Overview

- Vendor Transition Plan

- Open Discussion I NRG Feedback PStiG A

. '~~

II 1111

Hope Creek Generating Station Operating Schedule

- Current Plant Operating Plan

- Achieve Equilibrium 18-month Cycles of 513 Operating Days

- 97°/o Rated Power Operating Capacity Factor assumed in Design

- Power Coastdowns of up to 3 weeks may be utilized Cycle Start End months Cycle Exposure (MWd/StU) 8 Dec. 13, 1997 Feb. 21, 1999 14 9,337

9 Apr. 11, 1999 Apr. 22, 2000 12 9,028 10 May30, 2000 Oct.6, 2001 16 11,508 *

Includes 3 week coastdown

- Fuel Loading Plan Cycle 8 Core

1 OOo/o GE9 Fuel Design (8x8 Barrier Fuel)

GE9 Equilibrium Core batch avg. w/o U-235 target of 3.25°/o to 3.40°/o

Split feed batch with 3.25°/o and 2.98°/o w/o U-235 bundle avg. to maximize GE9 fuel design margins during fuel transition.

Cycle 9 Core

100% GE9 Fuel Design (8x8 Barrier Fuel)

Batch avg. w/o U-235 lowered from 3.25°/o to 2.80°/o to maximize GE9 fuel design margins during fuel transition.

Cycle 1 O Core

First transition core SVEA-96+ and GE9.

Combustion Engineering, Inc. 0 Public Service Gas & Electric PSfiG l 1111 4

Hope Creek Generating Station Current Status

!* Plant Monitoring System Changes Cycle 9 Core

Year 2000 Concerns

Replace NSSS Process Computer

Upgrade to ABB Core Monitoring System

!* Three Reloads - Cycle 10, 11, and 12

!* Reload Engineering Services Reload Design Reload Licensing and Safety Analyses

. *.*~

Combustion Engineering, Inc. 0 PSLiG Public Service Gas & Electric E

,,1111.,.,

6

Project Team for Vendor Transition

!* Overall Contract Project Managers

- PSE&G - Don Notigan

- ABB - Dick Matheny

!* Supporting Project Managers and Teams for

-- Fuel Engineering Fuel Supply

- Fuel Fabrication

- Quality Assurance

- Core Monitoring

- Operations

- Licensing Combustion Engineering, Inc. 0 Public Service Gas & Electric PS~G 7

j,11111

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~*

Project Team Communication Plan

!* Close, frequent communications between PSE&G and ABB.

!* Periodic communications with NRC on Transition Plan progress.

!* Raise potential issues early for proactive resolution.

!* Apply Lessons Learned on fuel vendor transitions.

Combustion Engineering, Inc. 0 Public Service Gas & Electric PS~G Al 8

- - *~

Transition Plan - Overview

!* Current Status {where we are today)

Exchange of Information (Establish "Conditions for Design")

- Near Term Activities {where we are headed)

Constructing Lattice Physics Models Benchmarking Nuclear Methods to Cycles 1-7 of Hope Creek Constructing Thermal-Hydraulics Models for GE9 Co-resident Fuel

Hydraulic Compatibility

CPR Performance Constructing Licensing Analysis Models

AOOs, LOCA, CADA, Misleading, Fuel Handling

Combustion Engineering, Inc. 0 Public Service Gas & Electric PS~G jlRll 9

- Plant Specific Data *.

- Systems, Structures, and Components (SSC)

- NSSS Design

- Plant Performance

- Plant Design Bases

!* Procedures

- Tech. Spec., Operating procedures, EOP

- Nuclear Data

- Allowable Plant Operating Domain

- Resident Fuel Design Data Combustion Engineering, Inc. 0 Public Service Gas & Electric PS~G 10

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lO 1998 SVEA-96+ Fuel Design

- Similar to SVEA-96 design operating in WNP-2

- Main difference SVEA-96+ has seven spacer grids vs. six for SVEA-96.

!* Similar to SVEA-96+ Lead Use Assemblies operating in Susquehanna Unit 2

- Main difference second generation debris filter design following proven in-core applications in Europe.

Combustion Engineering, Inc. 0 Public Service Gas & Electric PS~G 12 l 1111

- Compatibility with core internals, components

!* Compatibility with resident fuel

!* Compatibility with plant fuel handling equipment

- Compatibility with fuel storage facility Combustion Engineering, Inc. 0 Public Service Electric & Gas PStiG jl 1111 lllllJID 13

!* ABB methodology specifically

- Sizes bypass flow hole to maintain interassembly flow fraction

- Accounts for all bundle type hydraulic designs

!* Cycle-specific SVEA-96+ SLMCPR

!* Resident fuel SLMCPR is established from the previous cycles

!* Resident fuel CPR performance established in accordance with ABB NRC-approved methodology Combustion Engineering, Inc. ~

Public Service Electric & Gas V PSfiG II 19 14

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Fuel Transition - Nuclear Design Compatibility

!* Nuclear design codes benchmarked over several cycles

- Hot, cold KEffective

-Tips

- Thermal limits comparisons

!* Reference Core specifically describes all fuel in mixed core

!* Transient and accident analyses use Reference Core nuclear input Fast/slow transients

- CADA HGAPS Reactivity coefficients Combustion Engineering, Inc. 0 Public Service Electric & Gas PSfiG l 11111 15

!* Apply NRC-Approved Generic Reload Licensing Methodology CENPD-300-P-A approved May 24, 1996

!* Requirement for Implementing the SVEA-96+ Fuel Design

- SVEA-96+ CPR Correlation (CENPD-389-P)

- - ~

Combustion Engineering, Inc. 0 Public Service Electric & Gas PStiG jl 1111 16

ABB Current Approved BWR Reload Licensing Topical Reports

(CENPD-287-P-A)

Mechanical Design "ABB Seismic/LOCA Methodology for BWRs" (CENPD-288-P-A)

Mechanical Design "Fuel Rod Design Methods for BWRs" (CENPD-285-P*A)

Accidents - LOCA "BWR ECCS Evaluation Model:

Description and Qualification" (RPB-90-93-P-A, CENPD-293-P*A)

Accidents - LOCA "BWR ECCS Evaluation Model:

Code* Sensitivity" (RPB-90-94-P-A, CENPD-283-P*A)

Combustion Engineering, Inc. 0 PSL)G Public Service Electric & Gas

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17 Nuclear Design "ABB Atom Nuclear Design and Analysis Methods for BWRs" (BR-91*402-P*A)

Reload Analysis Methodology

. "Reference Safety Report for BWR Reload Fuel" (CENPD-300-P*A)

Accidents - CRDA "Control Rod Drop Accident Analysis Methodology for BWRs "

(CENPD-284-P-A)

Thermal-Hydraulic Design "CONDOR: A Thermal Hydraulic Pertormance Code or BWRs "

(BR-91-255-P*A, Rev.1)

Thermal-Hydraulic Design "SVEA-96 Critical Power Exper.

on a Full Scale 24-Rod Sub-Bundle "

(BR*91 *255*P*A, Rev.1)

AOO - Fast Transient Analysis "Dynamic Analysis Code for BWRs Description and Qualification "

(RPA-90-90-P-A, CENPD-292-P*A)

Special Events

Stablllty "ABB Advanced Stability Methodology for BWRs" (CENPD*295-P-A)

Special Events

Stability "ABB Advanced Stability Methods for BWRs" (CENPD-294-P*A) l 1111

Amendments to ABB Reload Licensing Topical Reports Mechanlcal Design Nuclear Design Therrnal-Hydraullc Design "Fuel Assembly Mechanical "ABB Atom Nuclear Design and "CONDOR: A Thermal Hydraulic Design Methodology for BWRs "

Analysis Methods for BWRs" Performance Code or BWRs "

(CENPD-287-P*A, Supl. 1)

(BR-91-402-P*A)

(BR-91-255-P*A, Rev.1)

Mechanlcal Design Nuclear Design Thermal-Hydraulic Design "Advanced PHOENIX and POLCA "SVEA-96 Critical Power Exper.

"ABB Seismic/LOCA Methodology for BWRs" Codes for Nuclear Design of BWRs" on a Full Scale 24-Rod Sub-Bundle "

(CENPD-288-P*A)

(CENPD-390-P*A) 7 (BR-91-255-P-A, Rev.1)

Mechanlcal Design

~

a Thermal-Hydraullc Design "Fuel Rod Design Methods "SVEA-96 and SVEA-96+ Critical for BWRs" Power Ratio Correlation" (CENPD-285-P*A)

Reload Analysis Methodology (CENPD-389-P)

"Reference Safety Report Mechanical Design

for BWR Reload Fuel"

~

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Fast Transient Analysis "STAV7 Fuel Rod Performance (CENPD-300-P*A)

"Dynamic Analysis Code for BWRs Methods" Description and Qualification "

(CENPD-391 *P*A)

(RPA-90-90-P*A, CENPD-292-P*A)

Special Events - Stablltty Accidents - LOCA "ABB Advanced Stability "BWR ECCS Evaluation Model:

Methodology for BWRs" Description and Qualification" (CENPD-295-P*A)

(RPB-90-93-P*A, CENPD-293-P*A)

, r

CADA Special Events - Stablltty "BWR ECCS Evaluation Model:

"Control Rod Drop Accident Analysis "ABB Advanced Stability Code Sensitivity" Methodology for BWRs" Methods tor BWRs" (RPB-90-94-P*A, CENPD-283-P*A)

(CENPD-284-P*A)

(CENPD-294-P*A)

Combustion Engineering, Inc. 0 PSL)G Public Service Electric & Gas E

18

- - . ~-

Licensing Topical Reports for NRC Review for Hope Creek Application

!* CENPD-389-P "SVEA-96 and SVEA-96+ Critical Power Correlations"

- Submittal Date: By May 30, 1998

- Requested Approval Date: By September 30, 1999

- U.S. Applications:

SVEA-96+ for Reload Licensing Applications for Hope Creek

Description:

Critical power correlation for SVEA-96+ fuel design with expanded experimental data base and correlation implementation. Also improved correlation for SVEA-96 fuel based on expanded data base and revised correlation formulation.

Combustion Engineering, Inc. 0 Public Service Electric & Gas PS~G

,,1111.,.,

19

!* CENPD-390-P "Advanced PHOENIX and POLCA Codes for Nuclear Design of BWRs"

- Submittal Date: By September 30, 1998 U.S. Application:

- '~.

Used in Core Monitoring System for Hope Creek Cycle 9

Approval Requested for Reload Licensing Applications

- Hope Creek Cycle 10 - 12

-WNP-2 Cycle 16-21

==

Description:==

PHOENIX/4 lattice physics code and POLCA/7 three dimensional nodal simulator code description and qualification for numerous plants.

Combustion Engineering, Inc. IJ!'lt..

Public Service Electric & Gas V PS~G lDll 20

Other Licensing Topical Reports for NRC Review

!* Improved Fuel *Rod Per_formance Methodology

- CENPD-391-P "STAV7 Fuel Rod Performance Methods"

Description:

Upgraded fuel performance code to remove current NRC SER restrictions.

- CENPD-287-P, Supl. 1 "Fuel Assembly Mechanical Design Methodology for BWRs"

Description:

Revise fuel rod integrity evaluation and input to licensing analysis methodologies for change in fuel performance code.

- Submittal Date: To be established at planned May 14, 1998 ABB/N.RC Status Meeting.

- U.S. Application:

Reload Licensing Application for WNP-2 and Hope Creek

May Use For Future CE PWR Fuel Performance Applications Combustion Engineering, Inc. 0 Public Service Electric & Gas PS~G j,1111 JI.,.,

21

- Hope Creek Strategic Planning

- Upgrading NSSS Process Computer for Cycle 9 Replacing Core Monitoring System for Cycle 9

- Optimization of Cycle 9 Core for introduction of new Fuel Design in Cycle 10

!* Reload Fuel Project Management in Place

!* Vendor Transition Activities in Progress

- NRC Comments on Transition Plans

!* NRC Comments on Proposed Milestones for Licensing Topical Reports*

Combustion Engineering, Inc. 0 Public Service Electric & Gas PS~G j,1111 11.,.,

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8 Ill 8 8 i181ii8iiiii iiiili~I""1i:oi"Siill8i:lElllT" ;i!iiii'888W1i1188 Fuel Design Changes Salem Units 1 & 2 Presentation to Nuclear Regulatory Commission April 7, 1998

Meeting Agenda

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New Salem Fuel Design: Robust Fuel

Current Salem Fuel Design

Fuel Performance Issues

Salem RFA Fuel Design Features

Implementation of Intermediate Flow Mixers

IFM Benefits

Salem Licensing Basis

IFM Implementation

Current Fuel Design

!ii 181111 18

Salem Vantage+ Fuel Assembly

17 x 17 Array with 0.374" OD Fuel Rod

ZIRLO' Cladding & Mid-Grids

lnconel Top & Bottom Grids

Removable Top Nozzle

Debris Resistant Bottom Nozzle

Features Not Currently Implemented

IFM's

Annular Axial Blankets

Protective Bottom Grid

I Current Salem Fuel Assembly 17X17 VANTAGE SH (with some V+/P+

features)

ZIRLQTM Guide Thimbles ZIRLO Claddin ZIRLO Mid-Grids RTN

-- Zirconium Diboride IFBAs DFBN

FuelPeriormancelssues 1111 I ass :E

Flow Induced Vibration (F~V)

DNB Performance

IFBA I Rod Internal Pressure (RIP)

Incomplete Rod Insertion (IRI)

Debris Potential

Axial Offset Anomaly (AOA) Potential

New Salem Fuel Design u;~;;;;;:i~~iiili,,

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Salem Robust Fuel Assembly (RFA)

Vantage+ Design: 17 x 17, 0.374" Rod, ZIRLO'

Modified Mid-Grid Design

Thicker Guide Thimble

Protective Bottom Grid *

Annular Axial Blankets

Desired Additional Design Improvement

IFM's

Salem Robust Fuel Assembly 17X17 MV5H IFMS Thick GT P-Grid ZIRLO' Guide Thimbl Thick-Walled I

ZIRLO Claddin 1------

ZIRLO MV5H Mid-Grids Axial Blankets Axial Blankets 1---~, ZIRLO MV5H IF Ms Zirconium Diboride IFBAs Protective Grid DFBN

Modified Mid-Grid Design

Design Changes

Vane Pattern

Vane Geometry

Spring I Dimple Geometry

Comprehensive Testing

Delta P

Assembly I Rod Vibration

Mechanical (Bulge Joint, Grid Crush, etc.)

DNB

,.:_~~~le * ',_

Modified Mid-Grid.Design (Cont'd) '~,~J*rrrn~ :

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Eliminates Flow Induced Vibration

Removes Source of Grid to Rod Fretting

Eliminates Rotated Grid Requirement

Restores IFM DNB Performance

Compatible with WRB-1 & 2 DNB Correlations

Currently In Use (Wolf Creek)

Thicker Guide Thimble

Larger OD

ID - No Change

OD - Used in Standard lnconel GT

Significant Strength Increase

Resists Thimble Tube Distortion

Enhances RCCA lnsertability (IRI Issue)

Within Experience -sase

Curre_ntly in Use at Wolf Creek

Protective Bottom Grid

Welded lnconel Grid Design

Used with Longer Fuel Rod End Plug

Positions Fuel Rods Near to Bottom Nozzle

Transmits Rod Weight to Nozzle (Unloads GT)

Provides Rod Support in High Crossflow Region

Enhances Debris Protection

Enhances RCCA lnsertability (IRI Issue)

Considerable Operating Experience

Already in Use at 19 Plants

Annular Axial Blankets i

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Annular Pellets: Used in Top & Bottom 6"

Slight Increase in Axial Peaking

No Tech Spec Peaking Factor ~ncrease Required

Provides Significant RIP Margin

Assists in Mitigating IFBA I RIP Issue

Considerable Operating Experience

Already in Use at 19 Plants

I. *.*

Salem RFA Summary.

- - - - -*************************************** ***********************************************************~****************************************~*-****************************************************

1 *Salem RFA Feature Enhanced Margin I I

ZIRLO Corrosion I

RIP I

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M1a=Gr1a oes1gn FiV DNB i

Thicker GT IRI Protective Grid Debris Protection IRI

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Annular Blanket RIP

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Intermediate Flow Mixers I

IFM Additionally Enhances Margins

DNB Performance

PCT Performance

IRI

FIV

AOA

Considerable Operating Experience

Already in Use at 21 Plants 9,

~*:~

PSE&G Plans to Implement IFM's Next Reload

Current Salem Licensing Basis

Margin Recovery LCR 94-41 (NRC Approved)

Performed Analyses for Fuel With & Wit~out IFM

IFM Use Delayed (FIV I Rotated Grid Issue)

LCR Submitted Analyses for Fuel Without IFM

WRB-2 Correlation not *Included in Submittal

~*

IFM Implementation

IFM's Provide Margin Enhancements

Implement IFM's on Next Reload Core

IFM's Require WRB-2 DNB Correlation

No Tech Spec Changes Required

Licensing Submittal

Letter Request for Approval of WRB-2 at Salem

Approval to Support 9/98 Fuel Fabrication i

I

.*'i--

'l'

ML18106A570

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ML18106A570

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