ML20128F686

Summary of 930111 Meeting w/ABB-CE in Windsor,Ct Re Review Status of CE Sys 80+ Design

ML20128F686
Person / Time
Site:	05200002
Issue date:	02/08/1993
From:	Mike Franovich Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9302110505
Download: ML20128F686 (67)

Text

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Docket No.52-002 APPLICANT: ABB-Combustion Engineering, Inc. (ABB-CE)

PROJECT: CE System 80+

SUBJECT:

PUBLIC MEETING OF JANUARY 11, 1993, TO DISCUSS THE REVIEW STATUS OF THE CE SYSTEM 80+ DESIGN WITH SENIOR MANAGEMENT On January 11, 1993, a public nieeting was held at the ABB-CE facilities in .

Windsor, Connecticut, between senior management representatives of ABB-CE and the U.S. Nuclear Regulatory Commission (NRC). Enclosure 1 provides a -list of attendees. Enclosure 2 is the material presented by ABB-CE.

ABB-CE opened the meeting with a status of the project and an overview of the System 80+ design process, ABB-CE also provided a progress report on ABB-CE responses to items from the CE System 80+ draft safety evaluation report (DSER) and an update of progress achieved frr closure of issues from-shutdown risk and severe accidents areas.

ABB-CE expressed concern over severe accident closure for the issue of core-concrete interaction (CCI) and containment performance goals under this condition. ABB-CE stated that they were attempting to quantify the appropriate heat transfer coefficient. ABB-CE was concerned that assumptions of very low heat transfer coefficients could result in the System 80+

containment not meeting the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> containment performance goal. . System 80+

design modifications were under consideration as a contingency. The-staff reiterated that severe accidents are not a design-basis accident requirement for evolutionary advanced light water reactors, and the 24-hour criterion for containment integrity is a performance goal. ABB-CE should analyze what the CE System 80+ containment and reactor cavity can accommodate with respect to CCI.

During the presentation on the System 80+ probabilistic risk assessment (PRA),

the staff commented that the results of the system importance studies must be analyzed and provided as input into the design reliability assurance program (D-RAF) and the operational reliability assurance program (0-RAP). Results from the importance studies would also be appropriate for roadmapping purposes.

The morning session concluded with a demonstration of the PASCE computer aided design (CAD) system. ABB-CE also demonstrated PASCE application for System 80+ design work and potential use for future plant owners /operatars.

9302110505 930208 PDR ADDCK 05200C02 , ..

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s February 8, 1993 The afternoon session consisted of discussions on several technical issues.

ABB-CE made presentations on issues concerning reactor coolant pump (RCP) seal integrity and instrumentation and controls (l&C) diversity with the associated common mode failure analysis (CMF) (see Enclosure 2). The afternoon session concluded with a tour and demonstration of the Nuplex 80+ control room complex.

The following commitments were made during the meeting:

(1) ABB-CE committed to provide an agenda for the CE System 80+ inspections, tests, analyses, and acceptance criteria (ITAAC) industry review for the first week at the ABB-CE facilities in Windsor, Connecticut and the second week at Duke Engineering & Services, Inc. (DESI) in Charlotte, North Carolina.

(2) ABB-CE committed to address the potential for cracking in the reactor vessel upper head including the CRDs. The staff noted this due to operating 0xperience review and the System 80+ vessel outlet temperature of 615 'F and the potential for high residual stress in the vessel material due to the fabrication process. ABB-CE should also address use of lower stiess fabrication processec., alternate materials, temperature adjustments, and leakage detection equipment.

(3) ABB-CE committed to evaluate that additional pathways for interfacing system loss-of-coolant-accident (ISLOCA) (beyond the shutdown cooling system scope) are designed to the ultimate rupture strength criteria.

Pathways such as the direct vessel injection lines, the emergency core cooling system (ECCS) lines outside containments, containment-spray pumps, valves, etc. should be addressed.

(4) The staff will evaluate if the use of relief valves on the suction of the shutdown cooling system (SCS) is sufficient to meet Branch Technical Position 5.1 on low temperature overpressure protection (LTOP).

(5) For operating experience review, NRC and ABB-CE management both need to determine what process was used to give confidence that operating experience has been comprehensively addressed. ABB-CE will provide a copy of the operations and maintenance study report-for System 80+.

(6) ABB-CE should address both TID [ technical information document) source term and the new source term in the power upgrade analysis.

(7) ABB-CE committed to provide an insights document that provides:

(a) interfaces with CESSAR-DC Chapter 18 and the associated task analysis review for the Nuplex 80+ control complex (i.e., emergency procedure guidelines (EPGs) PRA, shutdown risk, and severe accident input to the control room).

(b) roadmap input document for ITAAC, D-RAP, 0-RAP, EPGs, etc.

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'I February 8, 1993 (8) For the NRC structural audit, ABB-CE committed to define loads and load combinations, identify limiting regions and members, and prioritize-critical plant areas for CESSAR-DC. This action should be. completed in the February / March time frame. ABB-CE committed to evaluate improvement of the schedule for identification and prioritization of critical areas in the System 80+ structures prior to the current- schedule for identification of April 30, 1993. This action should identify and prioritize critical areas for which detailed analysis and audit are r needed for CESSAR-DC and the final safety evaluation report (FSER).

(9) for Generic issue 23, " Reactor Coolant Pump Seal Failures," the NRC committed to evaluate why the staff required a diverse-safety grade seal injection system in lieu of a diverse but reliable seal cooling system.

The staff will determine what will be needed for diversity.

(10) For CMF analysis of the I&C systems, the NRC and ABB-CE managemen',

committed that their respective staff's conduct an integrated morting with representativcs from reactor systems, PRA, and the I&C disciplines.

Both parties will need to identify potential hardspots for this issue in conjunction with the following CESSAR-DC accidents:

(a) reactor trip on main steamline break or feedwater line break accident.

(b) automatic safety. injection (SI) on large break LOCA.

(11) For ccmmon mode position 4, ABB-CE proposed use of LCl level output.

LCl is a digital switch which has a binary output (on/off) for a specific set of inputs- from either group control- or manual. The staff-will evaluate -that the use of a digital switch versus use of a relay is'-

appropriate for this issue.

(12) ABB-CE committed to evaluate if there is an adequate set of diverse ~-

displays. ABB-CE stated during the meeting that for.the diversity issue, the main control room will provide diverse display of 15 parameters, and the set was' based on Regulatory Guide 1.97. ABB-CE should evaluate if this is the sufficient set of parameters with input from the CHF/ diversity analysis.

(13) ABB-CE should be prepared to provide their definition of "to the extent practicable" for the ISLOCA issue during the January 21, 1993, meeting.

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(14) The stat f will provide ABB-CE a copy of the outline for development of the~PRA portion of the advanced boiling water reactor FSER.

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,4_ February 8, 1993-(15) The staff will contact ABB-CE for a discussion on leak-before-break methodology and application on the CE System 80+ design.

(16) ABB-CE will provide slides for an overview of the System 80+ program.

Sincerely, (Original signed by T.V. Wambach for)

Michael X. Franovich, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/o enclosures:

See next page DISTRIBUTION w/ enclosures:

Docket File PDST R/F MFranovich PDR RPerch, 8H7 PShea DISTRIBUTION w/o enclosure 1:

TMurley/FMiraglia DCrutchfield WTravers RPierson RBorchardt WRussell ACRS (11) JMoore, 15B18 GGrant, EDO- JPartlow, 12G18 TWambach EJordan, MNBB 3701 BGrimes, llE4 AThadani, 8E2 RBarrett, 801 TBoyce CMcCracken, 801 MMalloy BBoger, 10H5 WBeckner,10E4 BLiaw, 7D26 MWaterman, 8H3 DTerao, 7H15 GBagchi, 7H15 MRubin, 8E23 JWermiel, 10024 JRichardson, 7D26 JLyons,.8D1 RJones, 8E23 CRossi

/J)(u i D1 FConge , 10E2 0FC: LA:PDS PM:PUST:ADAR PM:PDST:ADAR SC- T:ADAR NAME: PShe ,

Franovich:tz TVWambach . RB hardt DATE:g/:i .

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/j'793 J /g793 J/g/93 0FFICIAL RECORD COPY: MSUM0ll1.MXF a

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7 1ABB-Combustion Engineeririg, Inc. Docket No.52-002 .

cc w/o enclosure:-

Mr. C. B. Brinkman, Acting Director Nuclear Systems: Licensing Combustion Engineering, Inc.

-1000 Prospect Hill Road Windsor, Connecticut 06095-0500 ,

Mr. C. B. Brinkman, Manager-Washington Nuclear Operations Combustion Engineering, Inc.

12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan-Ritterbusch Nuclear Systems Licensing Combustion Engineering, Inc.

1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500

-Mr. Daniel F. Giessing V. S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Steve Goldberg

' Budget Examiner 'i 725 17th Street, N.W.

Washington, D.C. 20503 Mr. Raymond Ng 1776" Eye Street, N.W. '

Suite 300 Washington, D.C, 20006-k 1

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4 MEETING ATTENDEES January-ll,.1993 RaME- ORGANIZATION M. Franovich NRR/PDST-T. Wambach NRR/PDSTL W. Borchardt K'9/ PDST --

D. Crutchfield NRR/ADAR W. Russell NRR/ADT T. Murley NRR/DO:

B. D. Liaw NRR/DE A. Thadani NRR R. Newman ABB-CE R. Matzie ABB-CE C. Brinkman ABB-CE S. Ritterbusch ABB-CE J. Longo ABB-CE F. Carpentino AUB-CE -

G. Davis ABB-CE L. Gerdes ABB-CE D. Finnicum ABB-CE R. Turk ABB-CE K. Scarola* ABB-CE A. Hyde

• ABB-CE M. Cross * .-ABB-CE Mr. Windsor * .ABB-CE D. Harmon* ABB-CE R. Schneider* ABB-CE-P. Hansen ABB-CE P. Lang- -U.S. DOE T. Crom -DE & S G. Hedrick DE & S

-S. Stamm Stone & Webster-

• Afternoon session only Enclosure 1:

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Proposed Agenda NRC Senior Management Meeting January 11,1993 - Windsor, CT 8:30 Welcome- R. Newman 8:40 Agenda Review C. Brinkman 8:45 Introductory Remarks T. Murley 1 R. Matzie 3

9:00 Design Summary - R. Turk j 9:45 Chapter 15' Analysis i Technical Approach F. Carpentino l 10:00 Process:for Road-Mapping PRA Insights. D. Finnicum Break 10:30 10:45 Submittal Schedule S. Ritterbusch l 11:15- Documentation of issue Closure S. Ritterbusch -

11:30 . Plant Information_ System Demonstration - Building 12 T. Crom .

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Proposed Agenda { cont'd) 12:15 Lunch in Conference Room 1:00 Technical Issues Discussions Structural Design Details L . Gerdes RCP Seal Coolability M. Cross

-l&C Diversity A. Hyde 1:45 Status of-Other Technical issues S. Ritterbusch 2:45 Depart for Nuplex 80+

Demonstration 3:00 Nuplex 80+ Demonstration D. Harmon 4:45 Wrap-up 5:00 ~ Adjourn d , . - . ,,...u e_

4 Chapter 15 Analysis Approach k

Overview:

7

! - Resolved Open items Regarding Methodology  ;

-Incorporate Additional Changes to improve Margin and/or to Resolve Other issues (e.g., LBB? a

- Adjust for 3% Increased Core Power ,

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4 Safety Analyses Changes q The Following Changes are Being incorporated per j i NRC Staff Agreement: .

- Delete 3 Seconds Time Delay for Loss of Offsite Power  :

- Retain DNBR Convolution for any Event Which Fails the.DNBR SAFDL

- Calculate Doses Using tid-14844 and '

NUREG-1465 Source Terms-i

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Safety Analyses Changes { cont'd)

The Following Additional Design input Changes Will be included in the Safety Analysis as Well:

- Erbium Burnable Poison

- DVI Line Size increase

- PZR Surge Line L/D increase

- Decrease Maximum Charging Flow

- Letdown Line K-Factor increase

- X/Q Increases (EPRI URD)

1 Reanalysis Scope The Following Scope was Accepted by the Staff.

It Excludes Repeat of Previous Sensitivity Studies Since They Remain Valid ,

- Containment P&T (MSLB only) (6.2)

- Large LOCA ECCS (Worst Break) (6.3) R

- Small LOCA ECCS (Worst Break) (6.3)

- Excess Load (15.1)

- Main Steam Line Break (15.1) 1

- Loss of Condenser Vacuum (15.2)

- Feedwater Line Break (15.2)

- Locked Rotor (15.3) j

-- L'oss of Flow (15.3). u i

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Reanalysis Scope (Cont'd)

- CEA Drop (15.4)

- CEA Withdrawal (15.4)

- CEA Ejection;(15.4)

- Pressurizer Level Malfunction (15.5)

- Letdown Line Break (15.6)

- Steam Generator Tube Rupture (15.6)

- LOCA Offsite Doses (15.6)

- Fuel Handling Accident (15.7)

.In. Addition to These Reanalyses, a RSB 5-1 Natural.

Circulation Cooldown Analysis is Being Done for-l Chapter 5

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Process for PRA Road Mapping

.i PRA is Integral Part of Design Process. t I - PRA Insights-from Previous PRAs Factored into initial Design a

- Baseline System 80 PRA Reviewed by Engineering

- Baseline PRA Modified During Design Process  ;

. PRA Staff Attend Weekly Design Meetings- i 1

PRA ReviewLof Design Changes 1

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Process for PRA Road Mapping (Cont'd) i 3

' PRA Comparative Analysis for Selected issues .

- 4 DG vs 2 DG and CT

- Bus Arrangement- .

CCWS/SSWS System Design

- Cavity Flood System PRA Reflects Design Evolutions  ;

Revision 0, System 80+ PRA and Comparison  ;

Assumptions-Document

.- Reviewed by Engineering

- Submitted to NRC for Review.

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. t Process for PRA Road Mapping (Cont'd) 1 Revision 1, System 80+ PRA in Progress

- Address-NRC Review Comments a I - Reflect. Latest Design Charges

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Review by Engineering All PRA Assumptions and Results Fully Documented in PRA Report Stand-alone Assumptions Document for Revision 0 l

D-RAP Program Plan Prepared and Submitted to NRC

- Covers Reliability Assurance During Detailed l L

Design Phase i

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l L Process for PRA Road Mapping (Cont'd) r PRA Assumptions and insights Document Being I Prepared'for Final PRA-

- Replace Rev. O Assumption Document

- By System  :

< Design Assumptions  ;

Operator Actions Operation Assumptions Risk Significant Components  !

- Submit to Engineering for Review.  :

InputLto ITAACE 3 Cross Reference Items to.lTAAC/CESSAR-DC q

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. 'k ABBIMS - DSER PROGRESS

SUMMARY

NUMBER OF ITEMS ,

see  ;

459 453 I,. TOTAL ITEMS: 936 c,, _

TOTAL SUBMITTED: 630  !

PERCENT COMPLETE: 67.3%

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DUE NOVEMBER DUE DECEMBER DUE ANUARY M' ' '

ITEMS DUE ITEMS SUBMITTED r

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Pag 31 DSER PROGRESS BY CHAPTER JANUAHY 1 SIH. TOIAL !6MAINING i'ERCEtfl 10IAL NOVEMBEH 1STH. DECEMBEH 1 SlH. COMPltle CHAPIEH DUE l SUBMi l lEC SUBMITTED DUE l SUBMil LEU DUE l SUBMil LEU 5 0.0 %

O O 1 0 1 1 1 2 1 1 3 3 50 G 100.0 %

41 41 6 6 2 50 37 10 1 01 ' 44 69.7 %

61 61 47 30 3 145 0 22 0 100.0 %

5 5 17 17 0 4 22 4 59' 19 75.6 %

39 23 16 16 5 78 39 25 5 44 20 68.8%

32 32 7 7 6 64 16 0 9t 201 31.0 %

9 9 4 0 7 29 39 0 6 39 13.3%

6 6 0 0 8 45 27 25 5 143 23: 86.1 %

9 166 112 111 29 29 61.7 %

29 28 3 0 14 1 18l 10 47 i 15 2 12 13 48.0%

10 10 0 0 11 25 12 12 12 50.0%

12 11 0 0 1 12 24 -.

06-Jan-93 ABB-CENP

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DSER PROGRESS BY CHAPTER :Page2-CHAPTER TOTAL NOVEMT el 15IH. DECEMBER 15IH. JANUARY 151H. TOTAL FEMAININC PERCENT DUc SUBMIiiEU DUE SUBM i IEU DUE ISUBM1IEC SUBMTED COMPLETE 13 12 0 0 4 3 8 0 3 9! 25.0%

14 35 30 30 2 0 3 0 30 5 85.7%

15 22 11 10 10 10 1 0 20 2 90.9 %

16 2 2 2 0 0 1 0 2 0 100.0 %

17 .23 6 6 6 6 11 6 18 5 78.3 %

(+2 repeats) 18 17 5 .. 4 - 2 'O 10 0 4 13 ' 23.5%

19 -72 27 26 5 2 40 2 30 .42 41.7 %

-20 56 21 21 28 14 7 0' 35 21 62.5%

. TOTALS 936 459l 453 - 193 138 284 39 630 306 67.3 %

NO. ITEMS 936 NO. SUBM.' 630-TOTAL 630. U *.31% COMPLETE 1

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SUBMITTALS BY BRANCH OF NRC

- l BRANCH l TOTAL l SUBM. I TO GO l PERCENT l ECGB 242 179 63 74.0 %

EMCB 63 47 16 74.6 %

EMEB 20 14 .6 70.0 %

EElB 48 10 38 20.8 %

HtcB 32 .11 21 34.4 %

HWB 31 13 18 41.9 %

OTSB 3 3 0 100.0%

POST 3 3 0 100.0 %

PEPB 5 1 4 20.0 %

PHPB. .31 19 12 61.3 %

RPEB' 55 45 10 . 81.8 %

PSGB- [ 29 21 8 72.4 %

-SPts 252 195 57 77.4 %

SPSB 75 31 44 41.3 %

sax 8 47 38 9 80.9%

. TOTALS 936 . 630 306 ~ 67.3 %

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W Submittal Status and Schedule

-November 15 DSER Responses (48%)

l . December 15 DSER Responses (63%)

January 15 New Source Term Report

-January 21. DSER Responses (100%)

January 28 ITAAC.(~60%)

February;15 Severe. Accident RAI Responses 4 February 15 Road Map Draft

. February 28

. Safety Analysis Results

. February 28 !TAAC (100%)

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4 Submittal Status and Schedule (Cont'd)

March 15 Severe Accident RAI Revisions I

, March 15 Road Map - Complete March 31 Safety Analysis SAR.Writeups March 31 ITAAC Industry Review Revisions

- April 1 Human Factors Engineering Submittal April 15 CESSAR-DC Amendment.

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Documentation of Issue Closure Issue:

- Closure of Review After Submittal of Material Proposed Approach:

- Document Review of Major Submittals via '

Internal . Status Reports 1

- Maintairi Explicit Listings of Confirmatory and Open ltem Status l Compile Status Reports and Confirmatory Item j

! List and Release in a Mid-Summer Progress

-Report l

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Structural Design Detail I 1.

1 Issue:  ;

- Present Sufficient Design Detail to Provide Staff Confidence'in Adequacy of Structural Design in Their Safety Evaluations Objectives:

, -identify Critical Areas for Design I - Provide Detened Design for Critical Areas t

. Status: .;

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- Meeting Held with Staff to Reach Agreement on Requirements, Objectives, General Approach / Methodology, Preliminary List of ' ]

Critical Areas and Schedule  ;

- Development of Governing ' Loads / Load Combinations and 1 Preliminary Detailed Design of Selected Critical Design Areas-  !!

are Proceeding in Parallel to Minimize Schedule i

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4 Design Methodology ,

+ Design Details Will be Developed for the Following Critical Design Areas:

-Containment Vessel

-Nuclear Island Basemat i

-Shear Walls Including Connections to the Basemat

-Side Walls to include Dynamic Soil Pressures

-Floor Slab Connections to Shield Building

-Freestanding Portion of the Shield Building, Particularly in the Dome  :

Region

-Containment Vessel Evaluation and Reactor Cavity Walls, including Static and Dynamic Pressure Capacity, for Severe Accident Loading ,

o Conditions  :

-Steel Containment Vessel Embedment '

-Steam Generator Cavity' r

-Non-Nuclear Island Safety Related Structures

k; I Design Methodology J

Determine Critical Structural Members from Analysis Results. i Determine Critical Local Loadings That May Govern Design of Members (Wind, Tornado, Missile impact, Equipment Loads, Hydrodynamic Loads, etc.) o

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Develop Preliminary Reinforcing Schemes for Walls, Floors, Basemat and Connections for the Existing Member Sizes Apply Enveloped Global and Any Local Loads to Members From Analysis'Results and Determine Maximum Shears, Forces and Moments in Members Check Adequacy.of Members and Connections per Codes end.

Standards identified in CESSAR-DC Table 3.8-4 Redesign Members That do not Meet Code and Standard.

Requirements l

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Analysis Methodology l

- Develop a Detailed 3-D Finite Element Model of the Nuclear Island l- and the Annex Structures for Global Load Analyses .

! - Use Shell Elements to Model the Shear Walls and the Floor i L Diaphragms L - Use Solid Elements to Model the Basemat

- Use Spring Elements to Model the Soil ]

i. The Containment Analysis Will be Based on a Separate Dynamic  !

[ Response Spectrum Analysis of the SCV Alone

[ The Finite Element Mode? Will be Used for Global Load Distribution ]

of all Other Structures-L The Behavior of the Composite Static 3-D Model Will be Cross-Checked Versus That of the Stick Model Used in the Dynamic SSI Aralysis

- Local Loads Will be Developed for Critical Design Areas t

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1/30/93 Complete the 3-D Model of the Nuclear Island Structures j 1/30/93. Identify Global Loads and Combinations i

2/15/92 Reactor Cavity Capacity and Design for Severe Accidents l

2/26/93 Complete Loading Analyses in Support of Wall / Floor

Design 3/ /93 Meeting With NRC Staff to Discuss and Resolve

- Comments on Previously Submitted Material 3/15/93 Preliminary Details for Walls, Floors, Basemat and Connections for Critical Areas i

L3/31/93 Complete Loading Analysis in Support of Basemat l Design 3

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j Schedule (cont'd) l 4/30/93 Confirm Critical Areas

[ 5/ /93 Meeting With NRC Staff .

L - Discussions on and Resolution of Questions on i: Previously Submitted Material i- Concurrence on Final Definition of Areas for l Which Critical Design Details are Provided I 6/30/93 Check Adequacy of Design for Critical Areas I 6/30/93 Complete Criteria and Required Design Detail for

[ Non-Nuclear Island Safety Related Structures i: 7/: /93 NRC Audit of Structural Design i-e l

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t Design Philosophy .

oMaintain the advantages of ABB's well proven

! NSSS: System 80 i

. An " Evolutionary" ALWR  :

e Assess design against ALWR goals eincorporate enhancements

. E?RI Utility Rec uirements Document l . Reg u atory rec uiremen~:s

.Si: art-ua & oaera:iona experience e

? eEnsure integration of design

• Standardize design  !

i System 80+

l Design Approach oStart with System 80 NSSS & Duke Power's BOP oConsider Changes Based on:

. EP RI ALW R Rec uirements

. N RC Mandated Changes 03 era":iona Exaerience o Assess impact of Changes on:

1

. Safe ~y 3erformance, Oaerations & Vaintenance

. Cost eincorporate Changes and Certify with NRC ,

System 80+

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EPRI ALWR Requirements Conformance oSystem 80+ is firmly based on the EPRI ALWR Utility Requirements Document .

o Currently only 23 rionconformances identified

' oWorking towards resolution of nonconformances by:

o Resolving NRC open items e Revising System 80+ design l' o Proposing revisions to URD i

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System 80+ Desig n Objectives

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Design Objectives Major Changes from System 80 Area f.1aintain Proven Design Very Few Changes Reactor Part-strength Rods for f.teet Utility load follow Performance Needs Increased Core f.targin Lower Operating Temperatures Reactor Improve Plant f.1argins Increased System Volumes Coolant Improved f.1aterials l Increased Redundancy Safeguards Reduce Core - Added Safety Depressurization System Systems ?1elt Frequency Redesign in Very Close Conformance l

with EPRI ALWR Requirements i

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System 80+ Design Objectives Design Objectives Major Changes from System 80 Area

- Simplify Design - Non-safety CVCS Auxiliary Systems

. Address Severe Accidents - Use Dual, Spherical Steel Design Containment Large Maintenance Access Areas and Nuclear - Meet Utility Maintenance Needs - Specific Radiation Protection Annex Features

- Nuplex 80+ Advanced Instrumentation . Provide State of the Art Control Complex and Control Human Factors Engineered Control Complex Improve Reliability Greater Redundance Electric and Diversity Distribution Consistent with Safeguards Systems and Support Systems

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i a e g n t hd n o .

n n gvga t i i t

i ag o nonh r a r

min s e riC t ey trPet n r

l s v u n u o o e o eDl a s' et S c +

r 'A n l v i r n 0 i

nt ng

. t t ce e nPCMwBC- aEat i o o. 8 c

ev s a m

a b. o r sh - t e

e OPEC s y

R = = S 4 il4l;ll

l -.

i

'i L

Reactor Coolant System ,

- O'a'ec:ive: Increase ,

- - - ~ ~ . , e, -

Vlarcins , c-- t - -

-Lower Ooera :inc -

(" = =

j-

=_=,,

/

! em aera:ure / = "" ~~

J ncreasec. Sys:em 4- w {saj(( '

,p

'g Vo.umes -

W, i

i - RinC Forcec

'i m

~

in 4

! Vesse s ___ -

iut marovec S:eam -- - D -.,_ _1

_ m -, _ - c,,_

t

Genera":or Jesicn

. System 80+

f e

Improved Steam Generator Design i

Increasec . . _ . . .

Aq Jowncomer Vo.ume =_.Qsrgw4

= Incone 690 Tu aes -~~iAeo I. I" 0% 1uae Plugg.ing

=~ g s=r- '

g.. ._

V arcin

== \% ivML

%{w m --

= mprovec ,> 'i '.I j =-

, (p'. o . qi n i V.airr:enance Access ': 6 L i!! ij o o ii

= Imorovec Dryers W TM =- '

= :;; 2 l:::h

_.,QT r

^

3 / = =""

A ...~ ,. .

=gy. ..... /- .

= . a ...

4 System 80+

e ,va w . , e- - -- w-

Safety Systems .

o Objective: Meet PRA Goals

. AccecL Safe":y Jearessuriza: ion Sys"cem

. maroved Safe:y .nlecilon l

'oerformance

. 4 ~~ rain Emerg ency =eecwa~:er l

. A ~:erna~:e AC ?ower Source (Gas

~~

uraine)

System 80+

l

Safety System Approach eTwo division redundancy & separation at system l

level e Four train redundancy & separation at component level

.Two Emergency Diesei Generators is e"non-safety grade" diversity for all safety functi

. e.g . , Gas ~:urbine g enera:or System 80+

Safety injection System e4 Train redundancy for S"*' '"#$'1" S"*"

I small break e Direct Vessel Injection eIRWST - no automatic e

! recirculation actuation g g v,-

• e No fuel damage for g. ~ , . ...

-" r !!_, ~

_a breaks less than 10 in. *euW.:.' -

e r ull flow testing

, g Il ,,

• l l

System 80+

L

f .

Shutdown Cooling & Containment Spray c

Increased design sm. .. c..,, s 3fessUre Containment Spray System o o' 23 Jecicated system -

configurations

-2 accitional 7 eat - ,

• 1 exclangers +

N o rea ic nmen~: ~ Tom c.

\ tw II

- Imi ea g 1" i > " ' ""' v"=.

& g ii in ec':lon L K e' i?o"I

"~"

Jum as & lea': ..

exc7ang ers intercnangea ale

= Zu low tes":ing System 80+

Safety Depressurization System

= Venting noncondensibles soie,y yo,,,,

= Alternative if Pressurizer A spray is unavailable ,,,,y' ,

eei e ei ,

= Depressurization to initiate feed & bleed r' ~ 3 g,,,,,,,,,

= Depressurization during E E% s 1 Gas Vent

/

ei ie seactor l severe accident ette V Ell'*;'n,

, /

ei f;e ib b

Reactor T nk WI I Reactor <

Vessei V System 80+

Emergency Feedwater System e Dedicated safety. system o Cavitating venturi h eliminates isolation logic e Motor & turbine driven e Redundant storage tanks pump for each SG NCH-S AFE T Y g

! CONDENSATE l

n = a ~ *" 2; s':: f 'xl J

T

+

+

T.

I T hl 1

Insis!  ! M

-[I "~'

J LL~_ I c== - -

• d1T g- I CDN A M MIl COM AINMENI

'EufftCENCY g fEEDwaTER

==

• J.  ?: '-

L ,T,:" O+ ; T l l  %

ro9 l

L ~ j=AJ l T

{+'"it

+

R; c==n i '

System 80+

.b

, t Electrical Distribution System ,

RESERVEN l: MAIN E!

[

= G Brea<er ,

3 ~ier onsi~:e mauxa; ,

u i aOwer ii, ,

l Non1E sTG

\On F Q~g

,I I

PNS W i 3ermenan.:

o nOn-Sace u1

~

I e CT i

i .l E

-F , ,E i

s mm

= JV /o increase in m Onsi:e caoaci:y o Q2 System 80+

_ _ _ _ -.____.____-...___.______..__t

r o, d

l Containment .

= 200 F:. Diame':er .

S":ee Sahere ~_

Swa Cesw

-3 ~: Taic< Slie c .: ., -

' ' , - .m ./ .

~ * * -

3u.. .llcIna

" ~s JL gi-

,i,

/  :'s

= nCreasec S3 ace or

~/wrg

, auuum m ' t 7=p TEgTj ug Tywretehm:Lru> ,em-, e Vain:enance Access sa w k m C il % i/ L J - ,

V bl

= Jesigned ~:0 Mi':ig a':e 3rf 3' R s, g

Core Damage ~ ~ - w - =. , a  ;

= Su as 37ere -ouses i

l ,

Sa=e"y Sys~: ems ,

i l

System 80+ l

-[

~

I 1  !

l {

l Plant Layout i

! e 4 Quadrant Separation e-- e- sm - ---

x i eCommon Basemat C h $ j ; p k I L il=_1y T ms.L, s!IIN $.-l1 ."

!, e Safety S Y stems in D -

Sub-sphere == l N[E , m _ II1%* g :

l

@N -T.

/T s.rP

%v' UZT-'f

. ~~_

< _ , _ - m ~i t

~

1.y.,[ +

6

=~

i s",lJ,J cs 5 !; f !i i b - Q =

l

" ' " . C8._ ,_

F 1 Lgj W = s. -

.. . =

wS ja

==

. 1 .= / S , s. . m

,_. / t grds Der--

! System 80+

i L Severe, .ccident Mitigation e System 80+ includes features to provide additional Safety Margin Beyond Deterministic Licensing Bases  ;

• Severe Accident Features  !

~

. Cavity Design to Permit Core Spreading & Retention t

. Provisions for Cavity Flooding

. Hydrogen Control System

. Containment Performance Evaluation e Reliability Assessment Program will Provide Monitoring i

of Systems Important to Safeaty

System 80+

3 i

s Severe Accident Features

)

To H2 Recombiner(Typ.of 4) 7-- *'~ '

~8 "a d a

'?*T g ,:

1 Reactor

^

n if**& Y n .. f

c- m, WMI I

~y

} g ) hef- f noia,, ...

=--

_L',_f--j a 3 oden t yog , ..

j p

\ g EL84 3'I.4 \

L ,3 7 t.-= l

),

IRWST (A n, p

d

.M j

s..k.:

(Trp.or s>

. v. sow av n"

• C""

• 1 - ' an-=

Spillway '

+

5 h c

y n "m** Reactor Cavity spmway l j .,.s,,

nn-.

/ . !J1 na-* -

l l} \ i'i$':-[S.0?:Mi**'-

' St Pump Room

)

/ - Holdup Volume Spillway CORE DEBRIS CHAMBER

-;y; qy.+;-

System 80+

s is RCP SEAL COOLING f o NRC DOES NOT AGREE WITH ABB-CE DESIGN .

APPROACH o ABB-CE DESIGN APPROACH HAS REDUNDANT MEANS OF COOLING THE REACTOR COOLANT PUMP SEALS o NRC WANTS A SEAL ASSEMBLY THAT WOULD NOT LEAK EXCESSIVELY DukiNG A STATION BLACK OUT. A SATISFACTORY RESOLUTION INVOLVES:

A DEMONSTRATION TEST OR DIVERSE, SAFETY GRADE, SEAL COOLING o ABB-CE IS INVESTIGATING A DEMONSTRATION-TEST, IF AGREEMENT TO THE TEST ACCEPTANCE CRITERIA CAN BE ACHIEVED WITH THE NRC o JANUARY 21, 1993, ABB-CE WILL BE HEETING WITH THE NRC-STAFF TO DEFINE THE DEMONSTRATION TEST CRITERIA F

-* ._,...,~-dO.,-, _

i ILC P S_EAL INJE_CTION

- / CVC_S D_ESIG1 <

SEAL INJECTION (SI) PROVIDED BY TWO INDEPENDENT &

REDUNDANT CVCS DIVISIONS TO ASSURE RELIABILITY, REDUNDANCY AND AVAILABILITY. .

o CHARGING /SI PORTION OF CVCS ASME III - SAFETY CLASS 3 DESIGN o TWO CENTRIFUGAL CHARGING PUMPS - SAFETY CLASS 3 DESIGN t o CHARGING PUMPS POWERED FROM NON-SAFETY RELATED  :

BUSES o EACH DIVISION CAN PROVIDE COMPLETE CHARGING FLOW RANGE (44-132 GPM) o FOR STATION BLACK 0UT (sso). EVENT CHARGING PUMPS POWERED FROM ONSITE ALTERNATE AC (AAC) POWER SUPPLY '

o FOR SB0 EVENT CONTINUED SEAL COOLING. ASSURED BY SI AND AAC o CVCS/SI SYSTEM DESIGN MEETS DRAFT RG 1008-REQUIREMENTS FOR AN INDEPENDENT! POWERED SYSTEM

I r

li I

L ILC P SEAL C_0_Ol_ING / C_C3 S DESIAN-COOLING WATER IS PROVIDED BY THE NON-ESSENTIAL COOLING WATER HEADER.

o CCWS CONSISTS OF TWO TRAINS  !

o EACH TRAIN SUPPLIES ESSENTIAL AND NON-ESSENTIAL COOLING WATER o EACH TRAIN INCLUDES REDUNDANT CCW PUMPS o ESSENTIAL CCW LOOP IS COMPOSED OF SAFETY CLASS 3 PIPING AND COMPONENTS o THE NON-ESSENTIAL CCW LOOP IS COMPOSED OF NON-NUCLEAR SAFETY PIPING AND COMP 0NENTS l

l 3 ---ev -,-,w w- - , , , , , -4,. -+w-- ,%--7 :n-ng-- -,y- p o,w- -,-i,--

l

)

I l

S_EAL C_0_0_L_ING.

o SEAL COOLING PROVIDED BY INDEPENDENT AND REDUNDANT COOLING SYSTEMS SEAL INJECTION (SI) WATER (6.6 GPM EACH PUHP AT 120*F) INTRODUCED INTO SEAL COOLING CIRCUIT.

COMPONENT COOLING WATER (CCW) WHICH COOLS SEAL WATER BY HIGN PRESSURE SEAL COOLER (HPSC) AND THROTTLE SEAL COOLERS (TSC).

o SEALS CAN OPERATE INDEFINITELY WITH:

LOSS OF SEAL INJECTION (SI) WATER WITH COMP 0NENT COOLING WATER AVAILABLE.

LOSS OF CCW WITH SI AVAILABLE.

. . - . m , . _ . J. :. - _ , . . . _ . , , _. , ,

-lSEAL LEAKAGE Pol = SEAL CAVITY PRESSURE P01 = INTERMEDIATE PRESSURE l CB0 P03 = BACKUP PRESSURE '

Ft.OW T

RESTRICTOR CCW RIGID COUPUNG - - -

HIGH PRESSURE L COOLER I~ --

H --- -MS THIRD SEAL ~J#

(8ACKUP SEAL) '[ """"

THROTTLE a A_ COOLER H .-

I '

M4 SECOND SEAL p02 d ccw

'[ ~ THROTTLE H a --cOOun FIRST SEAL r-*i'

~* I~~l _

Pol l

T02 CYCt.ONE AUXlLIARY *""l T~ 7 ,

FILTER IMPELLER i 1,,,,,,.; L JOURNAL BEARING JET 1

PUMP u

-~_

- p-nn IMPELLER _ JSEAL +

INJECTION '

I Figure 2 Flow Diagram for Hydrodynamic Shaft Seal System.

Normal Operation - with CCW & SI 9

_4.'

RCP S_EAL INTEGRITY

, o PROVEN BY MULTIPLE SHOP TESTS ON PALO VERDE RCP's o -TEST SIMULATED ALL POSSIBLE LOSS OF-SEAL C001.ING EVENTS LOSS OF CCW AND SEAL INJECTION WITH PUMP OPERATING

~

STATION BLACK 0UT -(SB0) CONDITIONS (LOSS' 0F CCW AND SEAL INJECTION WITH PUMP' IDLE IN .H0T TEST LOOP) o TEST RESULTS SHOWED THAT SEAL TEMPERATURE LIMITS WERE NOT EXCEEDED o INTEGRITY PROVEN BY-ACTUAL PLANT OPERATING EXPERIENCE AT PALO VERDE SEVERAL L G Gr SEAL COOLING EVENTS INCLUDING 3 HOUR EQU11ALENT SB0 EVENT-SEALS LAST FOR AT LEAST TM0: FUEL- CYCLES, THIk0 CYCLE.BEING CONSIDERED ,

y .

+

t i

TECHNICAL ISSUE: I&C DIVERSITY RECOMMENDATIONS FOR ACHIEVING COMPLIANCE WITH NRC POSITION ON .

DIVERSITY AND DEFENSE IN DEPTH.

PRESENTED AT NRC-CE SENIOR MANAGEMENT MEETING

?

JANUARY 11, 1993

)

i i

i l

I i

l SYSTEM 80 + "*'

u

Nuplex 80+ DEFENSE-IN-DEPTH Success Path Critical Function Non-safety Safety b l inlection System, Reactor Trip Reactivity Control Rod Control, CVCS (Boration)

B a cr

^ " " ' " " ' ^ "" "

Vital Auxl!!arles DC Station Battery Station Battery RCS Inventory Control CVCS (Charging / Letdown) Safety injection System RCS Pressure Controi lY l " S ste Safety Heaters / Spray,CVCS (Charging) lS De ress riz Ion y, Core Heat Removal Forced Circulation Natural Circulation Emergency Feed, Shutdown Cooling RCS Heat Removal Main Feed & Safety injection System Containment Isolation Control Valves isolation Valves Containment Environment Fan Coolers,H3 Ignitors Containment Spray, Recombiners Radiation Emission

"' " n I. Radiation isolation of Release Paths elc Pa s A ERIT

1. 455E9 ASE A BROW 4 BOYE8tl SYSTEM 80+"

RESULTS_OF DIVERSITY & DEFENSE IN DEPTH EVALUATION FOR-THE 28 CHAPTER 15 EVENT INITIATORS WHICH APPLY.

TO SYSTEM 80+:

l

1) FOR 19 EVENTS, THE ABB EVAlilATION DETERMINED THE DIVERSITY IN NUPLEX 80+ TO BE ADEQUATE: I 16 -

MODERATE FREQUENCY  !

2 -

INFREQUENT 1 -

LIMITING FAULT

2) FOR 4 EVENTS, THE ABB EVALUATION DETERMINED THAT ADDITIONAL ANALYSIS WOULD BE HEEDED TC DEMONSTRATE THAT THE DIVERSE SYSTEMS WOULD MEET CHAPTER 15 CRITERIA:

LOSS OF 0FFSITE POWER -

MODERATE FREQUENCY.

i RCP SHAFT SEIZURE -

LIMITING FAULT l

RCP SHAFT SHEAR - -

CEA EJECTION -

(CONTINUED) l SYSTEM 80+ .

l RESULTS OF DIVERSITY & DEFENSE IN DEPTH EVALUATION

( CONTINUED )

IT IS LIKELY THAT FURTHER DISCUSSION WITH THE -STAFF WILL DETERMINE TilAT ADDITIONAL ANALYSIS IS NOT REQUIRED TO ESTABLISH THAT THE DIVERSE SYSTEMS ADEQUATELY ADDRESS THESE EVENTS (BASED ON RELAXED ACCEPTANCE-CRITERIA).

3) FOR 5 LIMITING FAULT 1 & 3 EVENT INITIATORS, ABB'S ANALYSIS DETERMINED THAT THE DIVERSE SYSTEMS DO NOT PROVIDE ADEQUATE PROTECTION:

LIMITING FAULT EVENTS ADDITIONAL PROTECTION REQUIRED i CLOSURE OF LETDOWN LETDOWN LINE BREAK ISOLATION VALVE.

STEAM GENERATOR TUBE RUPTURE CLOSURE 0F THE MSIVs.

MAIN STEAM LINE BREAKS ACTUATION OF REACTOR TRIP AND MSIVs.

FEEDWATER PIPE BREAKS ACTUATION OF REACTOR TRIP AND MSIVs.

LOSS OF COOLANT ACCIDENT ACTUATION OF REACTOR TRIP,-SAFETY INJEC-TION, CONTAINMENT-SPRAY, AND CONTAIN-MENT ISOLATION.

l l

SYSTEM 80+"A

i ABB RECOMMENDATIONS

1) WORK WITH THE NRC STAFF TO ESTABLISH ADEQUACY OF DIVERSE SYSTEMS FOR ALL MODERATE AND INFREQUENT EVENT INITIATORS.

2) MITIGATION OF LIMITING FAULT EVENTS SHOULD BE VIA MANUAL ACTUATION OF ESF SYSTEMS USING CONTROLS IMPLEMENTED TO COMPLY WITH DEFENSE IN DEPTH POSITION 4.

JUSTIFICATION o MANUAL ACTUATION SHOULD PROVIDE ADEQUATE PROTECTION RELATIVE TO SITE RELEASE CRITERIA, GIVEN THE IMPROBABILITY OF A COINCIDENT COMMON MODE FAILURE.

o LIMITING FAULT EVENTS SHOULD NOT BE CONSIDERED IN THE DESIGN BASES FOR THE DIVERSE SYSTEMS BECAUSE THE PROBABILITY OF A COINCIDENT COMMON MODE SOFTWARE FAILURE IS NOT CREDIBLE.

o MODIFYING THE DIVERSE SYSTEMS TO INCLUDE LIMITING FAULT EVENTS IN THE DESIGN BASIS WILL ADD COMPLEXITY, (THEREFORE REDUCE RELIABILITY) AND INCREASE SUSCEPTABILITY TO SPURIOUS ACTUATION (THEREFORE INCREASED CHALLENGES TO PLANT SAFETY).

SYSTEM 80 +"^

1 ABB RECOMMENDATIOR_ EOR C.0MPLIAiCE WITH DEFENSE IN  :

DEPTH POSITI0b 4

1. MANUAL CONTROLS o MANUAL ACTUATION PROVIDED FOR 1 TRAIN OF EACH ESF PRIMARY FLOW PATH o CONTROL IMPLEMENTATION BYPASSES ALL MULTIPLEXING AND ALL COMPUTERS WITH LARGE

• SOFTWARE APPLICATIONS (I.E., HARDWIRED)-

THE LAST PLC PROVIDING DIRECT CONTROL'T0 THE CONTROLLED COMPONENTS IS RETAINED.

JUSTIFICATION o SMALL PLCs ARE MORE RELIABLE THAN EQUIVALENT MECHANICAL RELAYS.

, o THE SOFTWARE APPLICATION IS SUFFICIENTLY ~

SMALL.AND-SIMPLE, SUCH THAT IT-IS;100%-

TESTABLE.

(I . E . , NO COMMON MODE FAILURES)

( CONTINUED )

-summm.

SYSTEM 80+#

DIVERSE MANUAL ESF ACTUATION INTERFACE TO ESF COMPONENTS HCP BtVER$C MANUAL Csr aCiu-flow Dtv A Div C O II O l

C CS C

^ KEY *

^

V Crt V LC

• Loop Coatroner MCC - Motor Coatrae Center O '&

N$l NCP CDMTr0L3 esp CONTFDLS MCP CONTROLS RSP CDefROLS & IN91Catt0N3

& INDICafl@S & INDICAf f041

& IN9tCAfl0N3 C$r.CCS DIV13tt> C Cir-CC5 DtV15 foe a Gpour Ct>tPOL GeOJP CONIROL GROUP CDeT*0L GROUP CONTROL C#0VP CONTROL GROJP CONTROL 2 3 0

i 2 3 wwww. s'

. tC e l H tC e l ,

i tC 11 .

LC I w -

, sw sv ,v /

LCIl LCIl LCtl LC4l l I 1I <C l act 1 I act I I *C i1 *C l l acC l I <C I l <C 1I =C l

$1 CS 30 SC -

$8 CS $GI sG2 mstV fem:N A TRA;w 3 TRalN A featu A Teagu 3 teatN A feaIN C fangw a featw a PtPF & ' DfV C W V DTV C CTV DIV C PUMP & D1V A CTV DIV A CrW Div A VMVC VALVC VALVC M AND V AL VC PUMP P W AND VALVC VALVC VALVC

[

SYSTEM 80+#

ABB RECOMMENDATION FOR COMPLIANCE-WITH DEFENSE IN DEPTH POSITION 4 (CONTINUED)

2. INDICATORS o INDICATORS ARE PROVIDED FOR REG GUIDE 1.97 CATEGORY 1 PARAMETERS.

o CONTINUE TO USE COMPUTERS FOR INSTRUMENTATION WHICH IS IMPLEMENTED USING COMPUTERS ON ALL CURRENT PLANTS, E.G.:

CORE EXIT THERM 0 COUPLES REACTOR VESSEL LEVEL MONITORS SUBC00 LED MARGIN MONITOR o ONE CHANNEL OF OTHER INSTRUMENTS ARE HARDWIRED TO A DEDICATED DISPLAY DEVICE, WHICH USES A SMALL COMPUTER.

JUSTIFICATION o COMPUTER BASED DISPLAYS HAVE DEMONSTRATED HIGHER RELIABILITY THAN THEIR ANALOG PREDECESSORS.

l

! o THE SOFTWARE APPLICATION IS SUFFICIENTLY-SMALL AND SIMPLE, SUCH THAT IT IS 100%

TESTABLE.

(I.E., NO COMMON MODE FAILURES)

SYSTEM 80+#

4

. v.

4 .

JO.

4 CHRONOLOGY OF DIGITAL DISPLAY- IMPLEMENTATION -

FOR REG. GUIDE - 1.97 CATEGORY 1- PARAMETERS IN SAFETY RELATED DISPLAY INSTRUMENTATION P-CCS OR PC1 i.. CURRENT...........,.........i PLANTS. Esr-CCs OR PPs I i mALOG 01HCR SC , MCTCR S i PARATTCR$

P RCS 1-RCS COMPUTER (Sm) DICf1AL

~ - - ~ " - - - ~ " " ~ " DISPL AY CC1s I I (CC t >

HJit s 1 gg I --

(RvtM)

DP$

l~

P-CC5 DI A$ *N

@ PC$

i...PRE.VID.US

. . . . . . . . . . NUPLEX

. . . .. .... 8.0+ i Csr-CCS OR PPS DI A$.P OTHCR i PARAMCICR$

$ WSC ' .

CATCGORY p l CATCGORY l PARANCTCR$ j R INCLUDING P-RCS L 1-RCS CMPUTER' - - - - - - - - - - - - - -> P""$Ct fo g

DI$ PLAYS (sm) - - - - - - - - - - - - - +

CCis I gg i~ (CCT) -

m1Cs wJ mem. ,__J . - - - - - - - - - - - - - - +

DP5-l

.CC, . - - _ - - - - - - _ ., ,,As. ,

M PCS i.RE.VISCD NUPLEX 80+ i

. ... .............. ... ....... Cgr-Cs ,___,_______,,,

CATCGDRY l PARAMCICRS - DIAIf DiWR I gg i I V PARAMCICRS J p CATCGORY P-RCS PARA CTCR COMPUTER -----------~' DCDICatCD t-RCs -

Cl% PLAYS is-> __.-._ - - - _ - - -._ - ,,

Ctts I i (CCi > __

mtCs i Sc i (Rvtm --------------+

SYSTEM 80+# '

, . , - . . y -

n

s Safety System Software Development Program

. For the 1/15/93 submittal, ABB-CE will provide the following for implementation in Tier 2:

. Software Safety Plan Description

. Software Program Manual, consisting of

. Software Quality Assurance Plan

. Software Verification & Validation Plan

. Software Configuration Management Plan

. Software Operations & Maintenance Plan

. ABB-CE has identified how safety software should be treated in the ITAAC:

Software in a safety system is satisfactorily developed according to the software plans, as evidenced by successful audits of the software development programs against those plans.

. Some reviewers are not satisfied that safety software will be adequately assured through this ITAAC approach.

. Although ABB-CE is reviewing other approaches, no satisfactory alternative has been identified by either NRC or ABB-CE.

. ABB-CE and NRC are working constructively to resolve this problem.

l SYSTEM 80+