ML20215E653

Trip Rept of 861120-21 Visit to GE Facilities in San Jose,Ca Re General Capabilities Which Support Licensees & NRC in Technical Areas & Numac Equipment Marketed to Licensees.List of Attendees & Handouts Encl

ML20215E653
Person / Time
Issue date:	12/12/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:	Bernero R Office of Nuclear Reactor Regulation
References
NUDOCS 8612230048
Download: ML20215E653 (127)

Text

{{#Wiki_filter:- ._4 UNITED STATES J ' 't, NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 k.... ,o December 12, 1986 MEMORANDUM FOR: Robert M. Bernero, Director Division of BWR Licensing FROM: John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing

SUBJECT:

NOVEMBER 20-21, 1986, TRIP TO GENERAL ELECTRIC On November 20 and 21, 1986, I. Villaiva, resident inspector for la Crosse, Region III; J. Donohew, Project Manager for Oyster Creek; and myself visited the General Electric (GE) facilities at San Jose, California. The purpose was to discuss (1) the general capabilities GE has at san Jose which support licensees and NRC in nuclear power plant technical areas (November 20thi and (2) the Nuclear Measurement Analysis and Control (NUMAC) equipment GE is marketing to licensees (November 21st). Attachment 1 is a list of the individuals attending the meetings. Attachment 2 contains the handouts from the meeting on November 20th. Attachment 3 has the handcuts on the NUMAC equipment from the meeting on November 21st. The fol. lowing is a summary of the significant items discussed during the two-day trip. 1.0 Meeting on November 20th This meeting was held at GE's main facilities in San Jose at 175 Curtner Avenue. The agenda for the meeting is in Attachment 2. The individuals attending the meeting are listed in Attachment 1. 4 An introduction to its Nuclear Energy Business Operations (NERO) and to its Family of Coordinated Utility Services (FOCUS) was made by GE. ' Handouts are in Attachment 2. These were discussed briefly. Several of GE's Nuclear Systns and Services Operations were also discussed. These are underlined in the associated handout enclosed in Attachment 2. GE reactor performance improvement programs to improve plant operating flexibility,(maneuverability, power output and capacity factor were discussed. GE maintains files and records on performance with their performance indicator base being well over 100 items and probably close i to 200.) These programs include upgrading the unit's licensed generating power level, expanding the core operating region, improving hardware reliability and reducing lost capacity due to out-of-service eouipment, i Two handouts are enclosed in Attachment 2. tl71 In the discussion on hydrogen water chemistry, GE explained its system /b to monitor crack growth in the reactor coolant system. This is the ,7 This is being used by BWR licensees in plant tests to deter crack verification system (CVSI. It monitors growth in an existing crack. 2 hP the effectiveness of hydrogen water chemistry. An extensive discussion took place on GESSAR, GE's approved Topical for reloads. 8612230048 861212 / PDR TOPRP ElWOENE B PDR. j

s 2 There was a tour of the pipe test lab, the Atlas SRY/ Fuel Design Testing facility and the maintenance training center. The pipe test lab can test 1500 welds at a time, and.4" and 12" diameter pipes. The Atlas facility tests the thermal hydraulics of new fuel designs and safety relief valves (SRV). The Maintenance Training Center has a mockup of a complete operating floor, including the core, spent fuel pool, and fuel handling equipment, and the lower part of a BWR reactor vessel. These facilities have been used by licensees to prepare for work in radiation areas in their plants. 2.0 Meeting on November 21, 1986 This meeting was held at a GE facility downtown in San Jose on its NUMAC equipment. Handouts on th2 equipment are in Attachment 3. The individuals attending the meeting are listed in Attachment 1. The NRC stated that it wanted tc be more familiar with the GE NUMAC' product line because at least two licensees (La Crosse and Big Rock Point) are interested in this equipment. GE stated that its NUMAC equipment was the following: microprocessor based with one for equipment operation and one for the user interface, as the visual display; modular design with redundant power supplies built-in; ful.ly automatic with self testing on-line; digital data processing after an analog-to-digital data transformation; user friendly including the use of the microprocessor in the user interface; and common spares for Class IE and non-Class IE uses. GE stated the equipment is of high quality and high reliability and the benefits to the licensee are reduced surveillance time, reduced spare inventory and improved performance. NRC and GE discussed the training offered a licensee in the use of the NUMAC equipment. This training by GE stressed the training of the licensee's e'gineers to understand how the equipment functioned and displayed data. There was a tour of the testing facility, where different NUMAC equipment was operated with dummy signals. The equipment demonstrated to NRC was the rod worth minimizer, log count rate meter, process radiation monitor and hydrogen water chemistry control monitor. 3.0 Conclusions } The GE facility has capabilities that licensees and NRC could use in the future, These include the pipe testing facility Atlas SRV/ Fuel Design Testing facility, Maintenance Training Center and Crack Verification System, GE is working to assist licensees to (1) have better nuclear measurement capabilities with NUMAC and (2) improve power generation of existing nuclear power plants through upgrading the licensed power level, expanding the core operating region, improving hardware reliability and reducing generation time lost to out-of-service equipment. j i j i

[ " The NUMAC equipment demonstrated to NRC did appear to exhibit improvements over existing equipment in nuclear plants. The professed higher reliability, self testing on-line,~ modular design with built-in redundant power supplies, common spares for Class 1E and non-Class 1E uses, and being user friendly are qualities that appear to be focused to meet the needs of the nuclear industry. Original signed by John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing

Enclosures:

1. List of Attendees 2. GE Handouts from Meeting on 11/20/86 3. GE Handouts from Meeting on'11/21/86 cc w/ enclosures: R. Houston G. Lainas B. D. Liaw - W. Hodges E. Adensam-. W. Butler D. Muller DISTRIBUTION:w/ enclosures w/ list of enclosures 2&3 only

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L ENCLOSURE 1 MEETINGS BETWEEN NRC AND GENERAL ELECTRIC (GE) NOVEMBER 20-21, 1986 November 20, 1986 Company J. Zwolinski NRC/NRR/ DBL I. Villalva NRC/ Region III J. Donohew NRC/NRR/ DBL R. Hill GE B. Smith GE G. Nortin GE J. Embley GE L. Younghorg GE J. Klapproth GE J. Post GE T. Lee GE C. Van Damm GE J. Charnley GE November 21, 1986 Company J. Zwolinski NRC/NRR/ DBL I. Villaiva NRC/ Region III J. Donohew NRC/NRR/ DBL C. Van Damm GE D. Reigel GE H. Tellsley Dairyland Power Cooperative R. Rowe GE F. Chao GE orw---r-. .---,m --w---. -,t --w-- r-----yr--- =+---w y-w ,---w ewr+ -.ms W - w .,w.--w e. w-1r-. -w

I ~ ) i l ENCLDEURE 2 MEETING ON NOVEMBER 20, 1986, WITH GE Enclosures Agenda Nuclear Energy Business Operation Organization chart Introduction to General Electric (GE) NEB 0 Services FOCUS, Family of Coordinated Utility Services General Electric Nuclear Systems and Services Operations Reactor Performance Improvements Plant Operation Performance Improvements by GE BWRs General Electric BWR/6 Thermal Power Upgrade Capability Assessrent Power Uprate / j \\

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. WRC/GE MIE!ING 7 'i. November 20, 1936 r

• 8:30 Introduction i

Agenda GE Organisation 8:45 Services Overview 4 t Services Procedage Executive Summary ? 9:15 ' Focus (Family of Coord4==+=4 Utility Services) 9:45 Selected specific Services YSAR/TS Updating 'sua rever Uprate Water Quality / Hydrogen Water c h iatry* Instrumentation W.. - - -- t Services Core Perfommence Improvements 12:00 Lunch Containment Vent Purge and Repressurisation Emergency Procedure Guidelines 1:00 Bernero Meeting summary 1:15 GESTAR Review 2:00 Open Discussion

• 3:00 Test Facilities Tour (San Jose) i i

Pipe Test Lab Atlas i Training Center i I b l l \\. -,, -. -, - ~, -... ~. _ -. _ _ - _ _ _ _ _ _ - _ _. _..,. _.., _ _ _ -. -. - - - - --.-. ---.

i t NUCLEAR ENERGY BUSINESS OPERATION i NUCLEAR TECHNOLOGY NUCLEAR PRODUCTS NUCLEAR PROJECTS NUCLEAR FIELD AND ML CUSTOM.E kE. RV.I.CES. ENGINEERI SERVICES SERVICES -.UE.L ENGINEERIN.G ELECTRICAL DESIGN FIELD SERVICES F _CUSTO_MER SERVI.C_ES AND MANUFACTURINC' ENGINEERING RECIONS TERRITORIES FUEL MARKETING ENGI_N.EE. RING MARKETING OPERATION AND AND PROJECTS SERVICES SERVICES [NSTALLATION SERVICE,. TRAINING EUROPEAN NUCLEAR FIELD TECHNICAL SERVICES SERVICES TERVICES l ELECTRONIC -.OJ.EC..TS - - - PR - - - - - A.ND- - COMPUTER SERVICES MATERIALS SERVICES ~ N - M N - MM 4 o i 4

~ ,,., q. .. n +, e p INTRODUCTION TO GENERAL ELECTRIC NEB 0 SERVICES 9 GUS NORTIN MARKETING SERVICES NOVEMBER 1986

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ENHANCE NEB 0 PRACTICES OF IMPROVED CUSTOMER ' LISTENING / FEEDBACK TO GUIDE GE DECISIONS e DEVELOP PACKAGED SOLUTIONS TO INDIVIDUAL CUSTOMER CONCERNS / PROBLEMS /NEEDS I PROVIDE SERVICES THAT CUSTOMER WANTS/NEEDS 07GN2:11/86 l

INDUSTRY AND PLANT LIFE 5 INDUSTRY LIFE i. EXTENSION OPERATION. CONSTRUCTION DESIGN I 1960 1970 1980 ~1990 E PLANT LIFE j j CONSTRUCTION SHAKEDOWN OPERATION EXTENSION -10' 0 5 40 a 80 ~ OASSEMBLE BOPERATIONS EPERFORMANCE E PROTECT KEY COMPONENTS j OCHECK OUT BOUTAGES E REGULATORY E REPLACE FOR BACKFITS PERFORMANCE { i & LONGEVITY CDOCUMENT EDOCUMENT E TECHNOLOGICAL R ESTABLISH CONTROL UPGRADE RELIABILITY OF SAFETY SYSTEMS CINITIAL WPLAN MODS E MAINTAIN / REPLACE SPARES EMONITOR/ INSPECT l 07GN2:11/8.6,I a .---._..w~_-_____..~,n,ec-- .n.,-.-----

~ w, SERVICES FROM GENERAL ELECTRIC GENERAL ELECTRIC WILL PROVIDE EQUIPMENT, SERVICES, AND CONSULTING TO HELP KEEP GE BWR's OPERATING SAFELY AND ECONOMICALLY. GEN,ERAL ELECTRIC WILL SUPPORT OPERATING BWR's BY PROVIDING - SERVICES AIMED AT SERVING MAINTENANCE NEEDS, RESOLVING REGULATORY CONCERNS, AND ULTIMATELY EXTENDING OPERATING PLANT LIFE. GE SERVICES FORM THE FRAMEWORK TO HELP ENSURE THAT BWRs MAINTAIN GOOD PERFORMANCE AND OPERATE IN SUCH A WAY THAT LONG TERM AGING EFFECTS ARE MINIMIZED. l 07GN2:11/86 __

G.. J.. OUR SCOPE e MECHANICAL a ELECTRICAL PRODUCTS e ENGINEERING SERVICES e ELECTRONIC & COMPUTER SERVICES e REUTER-ST0KES (NUCLEAR INSTRUMENTATION) e MATERIALS SERVICES e QUALITY ASSURANCE SERVICES e SAFETY AND LICENSING SERVICES e FIELD SERVICES ~ e TRAINING SERVICES e OPERATOR TRAINING SERVICES GE SERVICES ARE UNIQUE, ARE TAILORED TO CUSTOMER NEEDS, 07GN2:11/86 :

-r. -,,--e OUR SCOPE IDENTIFIED N0, 0F SERVICES e MECHANICAL & ELECTRICAL PRODUCTS 50 e ENGINEERING SERVICES ELECTRICAL DESIGN 28 ENGINEERING ANALYSES, DESIGN, 34 AND CONSULTING e ELECTRONIC & COMPUTER SERVICES 38 e REUTER-STOKES 15 e MATERIALS SERVICES 8 e QA SERVICES 14 e SAFETY AND LICENSING 24 e FIELD SERVICES 40 e TRAINING SERVICES (EliGINEERING, 116 MAINTENANCE, AND I&C) e OPERATOR TRAINING SERVICES 75 TOTAL: 442 07GN2:11/86 Q..... ~ ' ~ MECHANICAL 8 ELECTRICAL PRODUCTS: e RETROFITS FOR BETTER OPERATION e REGULATORY BACKFITS e TECHNOLOGICAL UPGRADES e SOLVE PROBLEMS e IMPROVE OUTAGES 9 SOME EXAMPLES ARE: REFUELING PLATFORM UPGRADES, IMPROVED JET PUMP BEAMS, LASERTRAC, FEEDWATER N0ZZLE SURVEILLANCE INSTRUMENTATION SYSTEM, AND REMOTE-0PERATED CRD HANDLING EQUIPMENT, 07GN2:11/86 7-

e ENGINEERING SERVICES: e DESIGN / DRAFTING SERVICES e UPDATES e DEDICATED SITE ENGINEERS e SYSTEM IMPROVEMENTS e OPERATIONAL ANALYSES e REALISTIC / IMPROVED COMPUTER MODELS e PLANT LIFE EXTENSION PROGRAMS e SCRAM REDUCTION PROGRAMS e IGSCC PROGRAM e VARIOUS CONSULTING SOME EXAMPLES ARE: l IMPROVED SETPOINT METHODOLOGY,

FOCUS, SAFER /GESTR, AND DECONTAMINATION CONSULTING.

07GN2:11/8C -_ I

_ -[ ,c ELECTRONIC AND COMPUTER SERVICES: e OMNIBUS (FAMILY OF COMPUTER-BASED PRODUCTS) GEDAC GETARS ERIS GEPAC PLUS SPDS MONICORE & MONIC0RE PLUS 3-D MONICORE e DIGITAL CONTROLS e IMPROVED EQUIPMENT. e INSTRUMENTATION FOR MEETING WATER LEVEL REQUIREMENTS e IMPROVED NUCLEAR INSTRUMENTATION DESIGNS i IMPROVED AVAILABILITY, RELIABILITY, AND PERFORMANCE, REDUCED SURVEILLANCE TIME, AND MINIMIZED HALF-SCRAMS AND SPARE PARTS INVENTORY. 07GN2:11/86,

~ l: p, 6 i i REUTER-ST0KES e IMPROVED NUCLEAR INSTRUMENTATION GAMMA TIP l TIP SYSTEM, BWR/6 COMP 0NENTS AUTO TIP l e IMPROVED LPRM e WIDE RANGE NEUTRON MONITOR SYSTEM MEETS REG. GUIDE 1.97 REQUIREMENTS LPRM PROGRAM PACKAGES INCLUDE BOTTOM ENTRY LPRM, DISPOSAL SYSTEMS, CABLE REPLACEMENT, AND MANAGEMENT SERVICES. 07GN2:11/86 _ _ -

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l MATERIALS SERVICES: e INSTRUMENTATION REPAIR AND RETURN l e RENEWAL SPARES PROGRAM l e ENVIRONMENTALLY QUALIFIED HARDWARE - BASED ON UTI TY NEEDS PROGRAM MANAGEMENT, LICENSING RESPONSES AND ASSISTANCE ANALYSES t l TESTING SPECIFICATIONS 1 I 1 l i l l l l GE PROVIDES BOTH l QUALIFIED NEW PRODUCTS AND SERVICES TO L QUALIFY EQUIPMENT, INCLUDING PREDICTING OPERATING LIFE. l i l 1 l 07GN2:11/86. -,r,

7,,.,. QUALITY ASSURANCE SERVICES: e SOURCE INSPECTION - SUPPLIER SURVEILLANCE o QUALITY AUDITS e QA SEMINARS l e QA AND QC TRAINING e UPDATING OaM MANUALS l I e CONFIGURATION MANAGEMENT e INSPECTION SYSTEMS e DOCUMENT SYSTEMS DRAWS ON GE PARTICIPATION IN INSTALLATION AND STARTUP 0F EQUIPMENT IN OVER 50 BWR PLANTS. 07GN2:11/86, -,,.., _.....,,...... _..

Li i LICENSING / REGULATORY SERVICES: '

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TECH SPEC UPDATES e REG GUIDE ASSESSMENTS e FSAR UPDATES e LIVING SCHEDULES e RADIOLOGICAL EVALVATIONS e REG COMPLIANCE ASSESSMENTS e CONSULTING SERVICES o e EMERGENCY OPERATING PROCEDURES AND TRAINING GE HAS EXTENSIVE EXPERIENCE IN INTERACTING WITH REGULATORY AGENCIES. = 07GN2:11/86 - -.

FIELD SERVICES: e VISUAL INSPECTIONS e UT SYSTEMS e OUTAGE PLANNING AND SCHEDULING SUPPCRT e REFUELING FLOOR C0ORDINATOR AND TECHNICAL DIRECTOR e PROJECT MANAGEMENT e PLANT / SYSTEMS OPERATIONS SUPPORT ENGINEERS e OPERATIONS ENGINEER PROGRAM o PREPARATION OF PROCEDURES e SPECIALTY TOOLING / EQUIPMENT e ALARA CONSULTATION FOR MORE THAN THREE DECADES, GE HAS BEEN A l SUPPLIER OF ENGINEERING AND TECHNICAL SERVICES FOR BWR'S, ALSO, GE HAS COMMITTED SIGNIFICANT RESOURCES TO THE DEVELOPMENT OF SPECIAllZED BWR FIELD SUPPORT SKILLS, 07GN2:11/86.j

~ V; . ~.. w i TRAININ.G SERVICES: e MAINTENANCE TRAINING CRD MSIV REFUELING FLOOR RELIEF VALVE DIVER RECIRC PUMP SEAL 3 e ENGINEERING SNE STA BWR CHEMISTRY COMPUTER PRA HP ALARA DEGRADED CORE e INSTRUMENTATION & CONTROL NUCLEAR INST. PROCESS MINI-COURSES REACTOR CONTROLS GE INSTRUCTORS HAVE FULL ACCESS TO THE ENGINEERING RESOURCES OF GENERAL ELECTRIC'S NEB 0 STAFF. 07GN2:11/86, t a m. mmm

g. ,.Z ~ - OPERATOR TRAINING SERVICES: e SIMULATOR QUALIFICATION e TRAINING SYSTEM DEVELOPMENT e ACADEMIC AND INTRODUCTORY LEVEL COURSES e R0/SR0 BWR SYSTEMS TRAINING e NRC EXAM PREPARATION e INSTRUCTOR TRAINING o SR0 SUPERVISORY SKILL PROGRAMS o BWR OPERATING FUNDAMENTALS e R0/SR0 HOT LICENSE TRAINING e POST-LICENSE R0/SR0 TRAINING e STA OPERATIONS TRAINING i i 4 l l SINCE 1968, GE HAS PROVIDED OPERATOR TRAINING TO OVER 2200 UTILITY OPERATORS ON BWR PRODUCT LINES BWR/1 THRU BWR/6. 07GN2:11/86 ;

~,, PACKAGED SERVICES: EXAMPLES: e SPECIAL OUTAGE MANAGEMENT SERVICES PIPE REPLACEMENT DRYWELL SURVEYS ULTRASONIC TESTING VESSEL UT SPECIALTY TOOLING lHSI 9 1 l I r l l l l 07GN2:11/86 17 - s- .,,---.....,.--,..--.n.-,.-,,,-..,,--,,,-..---n.,.

ec m, m.- i l PACKAGED SERVICES: l EXAMPLES: j l e PLANT UPGRADE i OMNIBUS NUMAC DIGITAL CONTROLS WRNM/ IMPROVED LPRM l TIP UPGRADES l i l l l l l I L c l l i i t i h 07GN2:11/86 ! l

r L r L i PACKAGED SERVICES: I EXAMPLES: l e FUELS MANAGEMENT SERVICES -CONTROL CELL CORE BARRIER FUEL i FEEDWATER TEMPERATURE REDUCTION IMPROVED CONTROL BLADES [ 07GN2:11/86. 4m s

' SEEERALOitic1Bic anw n samt SYSTERAS A85 SERVKlES OPERMMNGS Ester Reesemasasseen ch Expedsmos 15 m -3BS IJWE e General Electric has developed a new, longer W.W. Phelan life M -300 LPRM. Its advantages are: Reuter Stokes a. Diffused sensor coating ubich provides (216) my to 201 longer life. 581-9400 h. New seal / purge ubich eliminates sensor / cable failuras. c. Incorporates a quick-disconnect connector. d. Uses low carbon and low cobalt materials to reduce Intergranular Stress Corrosion Cracking and Ram esposure concerns. e. Directly replaces the E-200 LPSII with-out electronic changes. 15 2 u-350 IJet e Bottom entry LPaN are standard on BM/6 W.W. Phelan plaats. General Electric has developed the Reuter Stokes bettse entry E-350 LPRM for backfit into (216) the earlier am plaats. 581-9400 The MA-350 LPtet includes the same longer e life features incorporated into the MA-300 LPRIE (see Recommendation Number 152A). e Individual Sensors can be replaced in the MA-350 LPElt. this reduces costs and Rae esposure. e A Bottom Intry Disposal System (BEDS) is required (see Reca====dation Mumber 152C). S 9 j ~

I ~ D SEufBAL $ ELECTRIC -,.. Sm SERVICES OPERMIWGS h homenneemdmena e-Expedance 15 30510M WERY e A Bottom Entry Disposal System (BEDS) is W.W. Phelan DES N EAL STST M required to dispose of spent E-354 LPRM Reuter Stokes sensors. (216) 581-9400 BEDS quickly and effkciently disposes of ~ o the LPRM lead cable and sensors. Ceneral Electric is developing a two step W.W. Phelam 152 LPau CAM.E e RErtAcamur LPRM cable replacement program. The Reuter first step is te replace the present Stokes RC59 (soft) lead cable with mineral (216) 581-9400 insulated (hard) cable. The seceed step is to route the mineral insulated cable above the Centrol Red Drive flanges. This could eliminate discommeeting, rolling up, and receanecting all of the LPRM lead cables each refueling outage. These steps would greatly reduce the LPRM cable failures. It would also greatly reduce the undervessel cable discom-mection/recomaection effort each refuel-ing outage. m e

p n. 3 1 l GENERAL $ ELECTRIC i DR) CLEAR SN A85 i SGIVICES OPERATIONS i l h m comument Empertence ANTO-TIP 5T573E e Ceneral Electric has designed an Automated W.W. Phelan Traversing Incore Probe (A-TIP) System. Reuter The system is a modification to the exist-Stokes ing TIP System. The existing drive control (216) units are replaced by A-TIP Control Units 581-94C0 (ATCU). A-TIP has three modes of operation: e automatic, semi-automatic, and manual. [ e In automatic, all TIP are operated sinultaneous by the master ATCU control room microprocessor. e Flux data is stored in individual ATCU-memory and later forwarded to the host i computer. e Redundant functions and diagnostic self-checking capability eliminate problems inherent in the old system designs. i m g A-37 m

1 pW [* FOCUS FAMILY OF C0ORDINATED UTIIII"I SERVICES i a 6' 4 i i PRESENTED TO: REPRESENTAINES OF US NUCLEAR RECULATORY COMMSSION STAFF ' NOG 8ER 20,1986 / B.W.SWTH GENEiul. aECTRIC COMPANY NUO. EAR ENERGY OPERATIONS

I FOCUS PROCESS UTILITY EEDS MO GE SYSTEMS AHD PROCESS PLANT EXPERTISE EXPERTISE WITH DATA BASE MULTIDISCIPLINED UTILITY /CE WORKDC CROLP ASSESS AND DOCUMENT r-------------- mImY m-GOALS I I e p .l l IDENTIFY IDENTIFY IDENTIFY EXISTDC CURRENT POTENTIAL l PROBLEMS LIMITATIONS PROBLEMS 1 I q u u IDENTIFY #0 IDENTIFY AHD IDENTIFY flHD l PRIORITIZE PRIORITIZE PRIORITIZE SOLUTIONS ERMNCEMENTS CONTDGENT ACTIOlG I l~ o I INTEGRATE l I o I l IMPLEMENT I m L_______________ nEASuRE SUCCESS BWS3.35 1 11/86

~ FOCUS ELEMENTS PERFORMANCE e IMPROVE ON PAST PERFORMANCE (CAPACITY IMPROVEMENT FACTOR, ALARA, 0&M COST, ETC.) PLANT e FORESTALL FUTURE PROBLEMS AGING LIFE e PREPARE FOR RELICENSING /RECERTIFICATION EXTENSION PRODUCTIVITY e ADDRESS ORGANIZATIONAL / ADMINISTRATIVE IMPROVEMENT CONCERNS SAFETY e EVALUATE EFFECTS OF MODIFICATIONS ASSESSMENT REGULATORY e IMPROVE REGULATORY PERFORMANCE COMPLIANCE INTEGRATED e PRIORITIZE AND PLAN FOR ACTI0flS COMMON LONG RANGE PLAN TO ALL FOCUS ELEMENTS BWS3.35 2 11/86 s

f. ~. FOCUS RESULTS e EFFECTIVELY COMBINES THE UNIQUE TALENTS OF THE UTILITY AND GE e PROVIDES A DISCIPLINED, SYSTEMATIC APPROACH TO UTILITY G0AL ACHIEVEMENT e IDENTIFIES AND PRIORITIZES ACTIONS NEEDED TO MEET G0ALS e FACILITATES UTILITY BUDGET PLANNING e PROVIDES JUSTIFIABLE, DOCUMENTED BASIS FOR UTILITY SPENDING POLICY AND IMPLEMENTATION SCHEDULES BWS3.35 3 11/86

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h_ DEFINE "0PTIMUM" PERFORMANCE s ACKNOWLEDGE PLANT AND UTILITY DIFFERENCES OPERATING VS. CONSTRUCTION PLANT MIX PLANT AGE STATE REGULATORY BODIES l e ESTABLISH PLANT UNIQUE DEFINITION i a b DEVELOP INDEPENDENT PARAMETER LISTING INDUSTRYWIDE MEASURABLE PARAMETERS J k = OTHER IMPORTANT PARAMETERS ASSIGN PARAMETER WEIGHTING FACTORS BWS3.35 7 11/86

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~ ~ ~ ASSESS CURRENT PERFORMANCE e GATHER NEEDED PARAMETRIC DATA SPECIFIC PLANT INDUSTRY e EVALUATE PERFORMANCE EACH PARAMETER AND PLANT ~ TWO OR THREE TIME PERI 0DS e ESTABLISH INDUSTRY RANK BY PARAMETER OVERALL e PERFORM MORE EXTENSIVE REVIEW IN KEY AREAS CONSIDER USE OF MAJOR SUPPLIER EQUIPMENT DATA BASES ASSESS ROOT CAUSES OF PROBLEMS BWS3.35 9 11/86 = - - - - - -

q L-PARAMETER RANKINGS FOR ONE PLANT (FROM POPULATION 0F 22 DOMESTIC OPERATING BWR'S) ? i I PARAMETER RANK 1 REGULATORY CONFORMANCE 2 NRC VIOLATIONS 6 NRC FINES 11 NRC OPEN ITEMS 8 EXPOSURE 1 l LIQUID DISCHARGES 4 l GAS DISCHARGES 7 RADWASTE 3 EMERGENCY-PREPAREDNESS 9 PLANT PERFORMANCE 18 i CAPACITY FACTOR 19 SCHEDULED OUTAGE TIME 8 ACTUAL / PLANNED OUTAGE 1 HEAT RATE 21 OVERALL 14 sWS3.35 10 11/86

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Si F9 i = 1 ..w.

..;T ~ '+- CAPACITY FACTOR LOSS COMMENTS s INDEX DATE I LOSS (1) COPMENTS 1 4/80 1.3861 FW NOZZLE CLAD REMOVAL. I 1 5/80 8.4699 NO COMMENT 1 6/80 1.0929 NO COMMENT I 2 5/77 0.2740 INSPECT CRD RETURN LINE NOZZLE. MANY INDICATIONS FOUND 2 5/77 2.1139 GRIND OUT CRACK INDICATIONS IN CRD RETURN LINE NOZZLE j 2 10/78 0.1598 CRD NOZZLE INSPECTISH 2 10/78 0.5137 CRD NOZZLE REPAIR i 5 7/77 0.9589 VESSEL IN-SERVICE INSPECTION. } 5 3/85 0.1370 RPV FLANGE ISI 8 5/77 0.0571 INSPECT RX VESSEL FLANGE '~ 6 4/80 0.1386 INSTALL SERVICE PLATFORM FOR FW NOZZLE WORK. 8 4/80 2.7322 DRAIN /DECON RX CAVITY IN PREPARATION FORFW NOZ7LE CLAD REMOVAL. 8 12/75 0.0913 RX SCRAM DURING SURVEILLANCE TESTING. MSIV CLOSUREON APRM HIGH FLUX SIGNAL. FLUX SPIKE RESULTED FROM PRESSURE TRANSIENT DURINGMSIV FAST CLOSURE TEST. ONLY ONE MSIV WAS BEING TESTED WHEN REACTOR SCRAMMED. i 8 6/76 1.4458 NO COMMENT 8 2/84 0.0228 F DECREASING POWER FOR MAINTENANCE OUTAGE 8 2/84 0.8083 MSIV LEAK RATE TESTS PER TECH SPEC REQUIREMENTS 8 2/84 0.0378 F TESTING FOLLOulNG MAINTENANCE OUTAGE 9 10/74 1.7130 S/R VALVE 71E OPEN SPONTANEOUSLY & COULD NOT BE CLOSED. WORN SECOND STAGE 4 P _3 ILOT SEATS ALLOWED LEAKAGE. LAPPED SEAT. 9 11/74 2.5894 S/R VALVE 71L OPENED SPONTANEOUSLY S COULD NOT BE CLOSE. RX WAS MANUALLY SCRAMM ED. VALVE 71J HAD LEAK ALARM. ALL VALVES INSPECTED. FIVE VALVES REQUIRED Rg EPAIRS.VARING FROM LAPPING OF PILOT VAL-VE 10 REPLACEMENT OF SECOND STAGE DISC. 9 8/73 0.1598 RV-2-71K K RELIEF VLV HAS EXCESS PILOT SEAT LEAKAGE.VLV OPENED DEPRESS RX. PILOT l VLV LAPPED. 4 9 6/78 0.3529 RX SCRAM WHEN SAFFTY RELIEF VALVE 71A LIFTED SPON-TANEOUSLY AND STUCK OPEN. BLOW DOWN TO 150 PSI OCCURRED. REMAINED SHUTDOWN TO INVESTIGATE AND REPAIR VAL VE. VALVE SHOWED STEAM CUTTING OF PILOT SEAT. 9 7/78 0.5806 NO COMMENT 9 11/76 0.546d RX MANUALLY SCHAMMED VdEM A SAFETY-RE8.IEF 71F VALVE LIFTED. THE DEFECTIVED V ALVE WAS REPLACED ALONG WITH VALVES 718, 71J, AND 71L. WHICH EXHIBITED HIGH TEMPERATURES VALVE 7tF STUCK OPEN. REACTOR BLEW DOWN-TO SOOPSI 9 1/77 0.2511 INSPECT TARGET ROCK RELIEF VALVE DIAPHRAGMS. 9 1/77 0.0219 INSPECT AND TEST SAFETY-RELIEF VALVE WHICH LIFTED SPONTANEOUSLY DURING RESTART. REACTOR BLEW DOWN. RF.SULTANT FLUCTUATING WATER LEVEL SCRAPMED REACTORPRIOR TO M : ANUAL SCRAM. 9 2/78 0.6849 REMOVE 5 SUSPECTED LEAKING PILOT VALVES ON S/R VALVES C. D, E. F 4 Lt. REPLAC E WITH REWORKED ASSEMBLIES. ONE COMPLETE ASSEfBLY (L), 4 BASE ASSEMBLIES .ONLY (C, D. E & L). 9 1/80 0.4098 REPLACEMENT OF 71G ADS SAFETY / RELIEF VALVE DIA-PHRAGM. i .j (1) F INDICATES A FORCED POWER REDUCTION COMMENT BWS3.35 12 11/86

ses, i

= .a ..-4 i-ESTABLISH IMPROVEMENT G0ALS e REVIEW RANKING RESULTS e NOTE P0OR SCORES IN HIGHLY WEIGHTED AREAS e SET IMPROVEMENT G0ALS NEXT MEASUREMENT PERIOD LONGER RANGE BWS3.35 13 ~11/86

L.. SET IMPROVEMENT PRIORITIZATION METHOD e REVIEW DEFINITION OF OPTIMUM PERFORMANCE AND PERFORMANCE RANKINGS e CONSIDER ADJUSTMENTS TO PARAMETER WEIGHTS e ASSESS NET BENEFIT OF EACH IMPROVEMENT PREDICT IMPACT ON EACH PARAMETER NET IMPACT IS NET BENEFIT e CALCULATE BENEFIT / COST RATIO e UTILIZE BENEFIT / COST RATIO AS PRIORITY SCORE o 0BTAIN EXECUTIVE LEVEL ENDORSEMENT e CONSIDER INSTALLATION ON SPREADSHEET-TYPE SOFTWARE BWS3.35 14 11/86 1 e -.

y- =.. DEFINE. INTEGRATE AND PRIORITIZE IMPROVEMENTS e DEVELOP SOLUTIONS WITH COGNIZANT PLANT PERSONNEL e SEEK COMBINATION OPPORTUNITIES TO' ADDRESS MULTIPLE ISSUES' e PRIORITIZE COMBINED IMPROVEMENTS 6 a BWS3.35 15 11/86

g: RECONCILE BUDGET, GOALS AND SCHEDULE e MATCH NEEDED VS AVAILABLE RESOURCES e ESTABLISH TENTATIVE SCHEDULE e PROJECT IMPROVEMENT WITH AVAILABLE RESOURCES I e ITERATE ON BUDGET, G0ALS AND SCHEDULE FOR COMPATIBILITY I J BWS3.35 16 11/86

^ ~ y...

SUMMARY

e UTILIZED PLANT-UNIQUE DEFINITION OF PERFORMANCE e BALANCED REGULATORY, ECONOMIC AND OTHER G0ALS f e MEASURED KEY AREAS OF PLANT PERFORMANCE e IDENTIFIED IMPROVEMENT NEEDS l e PRIORITIZED IMPROVEMENTS TO BEST MEET G0ALS e PROVIDED CHECKS AN'D BALANCES ON BUDGET, G0ALS AND SCHEDULE e OPTIMIZED ACHIEVEMENT OF G0ALS l t l BWS3.35 17 11/86 l I. - m._, -_, m _ ___.,,,_......_Z",_..

'GENERALOrtscraic NUCLEAR SYSTEMS AND SERVICES OPERATIONS 1 Number RecommendsNon Comment Experience h ION Cut 0M&focaAPHY e General Electric has developed a range of D.N. More than SERVICES services and products for on-line water purity Rodgers 30,000 hours j monitoring based on ion-chromatography (IC). 152977 of related experience Services are designed to supplement a plant's logged at e protective water chemistry program by provid-PWRs ing specific and cuntinuous information on ,j, contaminate intrusion. More than 1 15,000 hours Includes both diagnostic and consul'tation ser-of relate'd I e vices utilizing portable IC equipment and/or experience specifiestion of analysis systems, training of logged at operators, reduction of data, procedures ex-BWRs. pension of existing capabilities, and sample conditioning. Sold Units to Hatch, Applicable to BWR, PWR, and fossil-fired Savannah e steam generation systems. River, Peach

Bottom, Also, see Recommendation No. 108.

Cenkai e (Japan). ~ A highly-reliable, direct replacement for F.F. Witt Instal).ed on IMPROVED F1.UID CDUPLINC a CouTR0rna the existing fluid coupling controller (scoop 152881 six reactors j tube positioner controller) is available. in Japan, h with up to e A Technical discussion between General four years Electric and BWR plant personnel'is recom-operating mended. The. discussion would cover the experience. plant experience with their present scoop ti tube controls. Based on this discussion, 1 a specific recommendation on control j replacement. could be made. Il peg, A-2 s

== i

l~ P ^- 9 i f GENERAL $ ELECTRIC L NUCLEAR SYSTEMS AND SERVICES OPERATIONS i Number Recommendeuon Corr.mont Experience i New neutron flux scram circuitry is available F.E. Hatch, Browns 104 SIMULATED THERNAL e i POWER MONITot to replace the present flow-referenced neutron Holland

Ferry, l

flux scram circuits. X54340 FitzPatrick, Chinshan e Will increase the margin to scram, improve plant availability. e Has been licensed at ten BWR plants and is standard equipment on all BWR/5 and 6 plants. 105 SERVICES FOR MEETINC e Ceneral Electric is available to assist F.E. RPV WATER LEVEL utilities in whatever course of action is Holland INSTRIDIENTATION taken to meet the NRC requiremente of Regu-154340 i REQUIREMENTS 1 story cuide 1.97 and Ceneric Letter No. 84-23. This includes rerouting of the reference legs, design of systems to ' ,I prevent reference leg overheating, and determination of whether the utility's present system or an alternate fix is l adequate. + 105 CONPEMSATED WATER e The Compensated Water Level Instrumentation F.E. Nine Mine i LEVEL IMSTRISIENTATION System was designed to meet the Category I Holland Pt. 1, SYSTEM (CWLIS) requirements of Regulatory Guide 1.97 and 154340 Oyster Creek, addresses the WRC's Inadequate Core Cooling Chinshan Concern of high drywell temperature. i e Use of the CWLIS with some e:erator action may be acceptable to the NRC requirements specified,in Generic Letter No. 84-23. gg cm,

l GENERAL O nticTRic NUCLEAR SYSTEMS AND SERVICES OPERATIONS Number RecommandoNon Comment Experience The Analog Trip System is a direct replacement C.V. Dain All BWR/6, A M M TRIP SYSTEM e for the various pressure, level, flow, and X51205 Fit: Patrick, temperature switches in the plant. Nine Mile Pt. 1, Chinshan e Units require less maintenance. 1&2,

• Tokai 2, Will reduce scrans since unita have superior Hatch 1 & 2 e

drift characteristics and less scasitivity to mechanical shock.

• e Are seicnically qualified to IEEE 344-1975 and environmentally qualified to IEEE 323-1974.

e e ( I pop A-4 90740

F O Q i l GENER AL $ ELECTRIC l NUCLEAR SYSTEMS AND i SERVICES OPEstATIONS 1 ~ Number Recommendegon Comment Emperience CatedA TIP SYSTEM e Canna sensitive probes are being used at W.W. seven operating BWR plants and have been Phelan r l purchased by sir other plants. Reuter l

Stokes, i

e Data from these plants confirm that the (216) { Camma TIP System monitors core power more 581-9400 l accurately than the thermal TIP System. More accurate core monitoring results in additional thermal margin. This can be used to simplify reactor operation and to attain more efficient, lower-cost core loadings. e Reduced maintenance time at one plant has resulted in an annual reduction ir,the radiation exposure to the maintenance ~ i l personnel. BWR/6 components can be retrofitted into W.W. Phelan TIP SYSTIDI UPDATE e l existing BWR/2-5 operaging plant TIP Systems Rueter to improve operation and reduce both mainte-Stokes nance and personnel exposure. (216) 581-9400 e The major equipment items ares a. Purge air control cabinet. b. Indexing mechanism. c. TIP control unit. d. Drive mechanism. i 1 p.g. A-7

I GENEaAL O nticinic i l NUCLEAR SYSTEMS AND l SERVICES OPERATIONS . - l i Progrent r j w Menegst i The General Electric open/close monitor F.E. [ l BAFETY/ RELIEF VALVE e OPEN/CLOSE MnsfT10E provides a positive indication (not Holland susceptible to ambiguous readings from 154340 leaking safety / relief valves). i e Switches are qualified for 40 years in-containment. j i Switches are not susceptible to cross-talk g from other valves and S/RV. A new control rod design (a hybrid design) K.W. Most BWRs 1 MYBRID CDIfrBOL EDO e l with improved lifetime is available. Brayman X56587 t Hybrid control rods, on the average, have i e l '40 percent improvement in lifetime over the B C control rod. 4 i Is compatible with existing core designs, is e a direct replacement (same nuclear worth). i Contract can be negotiated with the utility ) e j to provide cost incentives to the utility 1 in return for agreeing to a long-term supply contract with NEBO. Individual terms and i conditions can vary, price is indexed to labor and material indexes to provide J escalation protection. 1 l T 1 8"*. b g. g ) l

l 'GENEaAL O ntscraic NUCLEAR SYSTEMS AND + SERVICES OPERATIONS Experience Number RecommendsNon Comment Plant-specific analyses have demonstrated C.A. Walker

Cooper, l

IAW-LOW SET /IAMERED e MSIV WATER 12 VEL TRIP the capability of Low-Low' Set (LLS) relief 154277 Duane Arnold, logic and Lowered MSIV Water I4 vel Trip to Hatch 1 & 2 mitigate postula'ted thrust and torus loads associated with subsequent actuations of a safety / relief valve in a BWR plant with a Mark I containment. e Advantages of installing LLS/Imvered MSIV , Water Level Trip include a reduction of safety / relief valve challenges and cycles, decreased fatigue duty on the containment, reduced safety / relief valve maintenance, j reduced probability of " sticking open" l safety / relief valves, and improved plant availability by reducing spurious MSIV trips ) and scrams. j FEEDWATEg marr u e General Electric has developed a thermal K.W. Hess j SLRVEILLANCE INSTEU-detection system which compares outside 151343 j MENTATION SYSTEM nossle wall temperature with supply pipe j (FMSIS) outside wall temperature. 1 i e The system accurately monitors and records leakage rates,past the inboard seals of the ) double piston-ring feedwater sparger. l e Installation of the system should relieve i in-vessel test requirements imposed by NLIREC-0619. pope A-70 1 serne y O l

i GEN ER AL O rtscraic NUCLEAR SYSTEMS AND SERVICES OPERATIONS Number RecommendeNon Comment Experience An improved mechanical overspeed trip P.F. Kachel HPCI 1URRIME e OVER-SPEED TRIP assembly is available. This assembly includes 151307 a new improved tappet assembly, a new pis-ton, and modification to (or replacement of) the value body. e For additional information see SIL Nambers 353 and 392. O14 An overspeed trip often occurs when the RCIC M.S.

Hatch, RCIC STARTUP e

turbine is started. Laurent Fermi,, TRAMSIENT IMPROVEMENT 151325 Crand Culf, e General Electric has developed a steam bypass Kuosheng modification which greatly aids RCIC turbine 1&2 startup. General Electric has available's supply of S.N. Patel Hatch 1 & 2, CEMAC 5000 e REPIACEMENTS plug-compatible CEMAC 5000 replacement units. 156688 Browns Ferry, (105MAC-500) These instruments would be dimensionally and Millstone functionally replaceable with the icxisting plant instruments. BWR plant reliability would be improved by e use of the replacement product line. LPRM Ma m 2 W NT e General Electric has developed an LPRM Manage-W.W. Phelan SERVICES ment Service that provides BWR plant opera-Reuter Stokes tors with the necessary level of data plus (216) i special analytical techniques to develop a 581-9400 l replacement schedule that maximizes the life l o'f LPRM. ll Page.A=26_ 1 Q

m,e O I e m Ae V i. W ws . 'h Nkh ,'"'97 ., l,- .IV,'h f sXa - REACTOR PERFORMANCE IMPROVEMENTS e e (T. LEE) 1 1 (. r. l

- REACTOR PERFORMANCE IMPROVEMENTS The BWR Reactor Performance improvement Package is a family of programs that provides improved plant operating flexibility, maneuverability and capacity factor to the BWR owners. It includes features to uprate the unit's generating power output, expand the operating region, improve system hardware reliability and provide insurance against lost capacity due to out of service equipment. These are achieved via design and licensing evaluations, technical specification modifications and minor setpoint and hardware changes. .. These increased plant operating states have been demonstrated by analysis, testing and operation in various plants. Most of these improvements have been licensed in both the U.S. and overseas. k Significant insight was gained as a consequence of performing this analysis and i i testing for the Leibstadt plant. Leibstadt is the lead BWR 6 Mark 111 plant in j Switzerland, which has implemented many of these improvement programs. l The ability of Leibstadt to operate in these extended operating states was established by performing extensive analytical studies and was confirmed via special tests + ) performed during the startup test program and by monitoring plant performance during normal cycle operation. Plant operating experience during Cycle 1 and particularly in Cycle 2 has demonstrated that these programs have resulted in significant j improvements in plant operating flexibility and capacity factor. 1 Based on the performance to date, the reactor performance improvement pograms are i j achieving the expected gains in operating maneuverability and simplicity and capacity i factor at the plants which have implemented them. l 2

1; AGENDA PROGRAM ELEMENTS LICENSING STATUS LElBSTADT EXPERIENCE

SUMMARY

1 3

/...,v. s

t. J,

e f. NOMENCLATURE / ACRONYM - POWER UPRATE (PU) MAX' MUM EXTENDED OPERATING DOMAIN (MEOD) EXTENDED LOAD LINE LIMIT ANALYSIS (ELLLA) INCHEASED CORE FLOW ANALYSIS (ICFA) APRM/RBM TECH SPEC MODIFICATION (ARTS) ,) . (q. - FINAL FEEDWATER TEMPERATURE REDUCTION (FFWTR) c. OUT OF SERVICE (OS) EQUIPMENT FEEDWATER HEATER (S) (FWH) MAIN STEAM ISOLATION VALVE (S) (MSIV) i SAFRTY RELIEF VALVE (S)(SRV) 1 L V RECIRCULATION LOOP (SLO) i. ~ )- i i t s e e - a n-se-c -e, wa s-e--- ee-a. ,,en- .-- - -w-m-ne-,,_ n,-----

f i i CORE THERMAL POWER l (% OF RATED) (% OF UPRATED) 13 l 1" UPRATED ~ 5 IN CONDITION l KKL 1042/o POWER UPRATE N ;;;;;;c;ct;; c; 188 g l OPERATING MAP MEOD BOUNDARY / max / CORE-90 t' ~ j FLOW l-gg ~ \\ [/ RATED ROD / 70 i M ~ LINE // 60 ~ RATED ui ptow \\ / \\ h n ~ / j SLO 40 NAT CIRC / I Significant increase in AIIOwed ~ Operating States Achieved via CAWTADON N N M LOW SPEED o Design & Licensing Evaluations max valve POSITION. 10 10 o Technical Specification Changes o Minor Setpoint and Hardware d ), [ [ 3 8 Adjustments ,3 CORE FLOW (% OF RATED)

CAPACITY FACTOR IMPROVNMENT DUE TO PROPOSED 250*F FFWTR=0.7% 1%/ WEEK FFWTR 100 % / 94 % ,/ 2-1/2% WEEK / COASTDOWN 85 % i ( MWE 6 WEEKS FOR PROPOSED FFWTR FUEL CYCLE LENGTH 6

e 9 TYPICAL REACTOR PERFORMANCE IMPROVEMENT PROGRAM PACKAGE ELEMENTS POWER UPRATE OPERATION AT 104.2%' THERMAL POWER ? 5% INCREASE IN STEAM FLOW PRODUCTION MAJOR UPRATE BEYOND 5%IN THE FUTURE MAXIMUM EXTENDED OPERATING DOMAIN RATED POWER ALLOWED FROM 75% TO 105% CORE FLOW APRM AND RBM HARDWARE AND TECH SPEC MODIFICATIONS FINAL FEEDWATER TEMPERATURE REDUCTION RATED THERMAL POWER ALLOWED FROM 420*F TO 250*F FEEDWATER TEMPERATURE y OUT OF SERVICE EQUIPMENT RATED POWER ALLOWED WITH FWHOS INCREASE ALLOWABLE OUT OF SERVICE TIMES WITH SRVOS APPROXIMATELY 95% POWER ALLOWED WITH MSIVOS APPROXIMATELY 70% POWER ALLOWED WITH SLO OPERATING STATES EXTENDED TO ACHIEVE MAXIMUM OPERATING FLEXIBILITY WITHOUT MAJOR HARDWARE MODIFICATION.

e ) r POWER UPRATE 1 OBJECTIVE l UPRATE THE UNIT POWER TO THE DESIGN RATING OF 105% STEAM FLOW (104.2% THERMAL POWER) ALLOW OPERATION OF OTHER IMPROVEMENT PROGRAMS AT THE UPRATED POWER ', j o, BENEFITS PROVIDE ADDITIONAL GENERATING CAPACITY I SIGNIFICANT UPRATE BEYOND 5% IS FEASIBLE IN THE FUTURE WITH FUEL AND PLANT MODIFICATIONS AND TESTING CONFIRMATION l I 7s

a c o o ~ 9 i MAXIMUM EXTENDED OPERATING DOMAIN If a a DEFINITION. ~ MEOD FEATURES ELLLA + ICFA + ARTS OBJECTIVE RATED POWER OPERATION OVER TARGET FLOW RANGE OF 75% - 105% FLO t ENHANCE THE ROD BLOCK MONITOR (RBM) SYSTEM (FOR PRE-BWR 6 PLANTS) INCREASE MARGINS TO THERMAL LIMITS = SIMPLIFY OPERATIONS FEATURES EXTENDED LOAD LINE REGION (ELLR) ~ INCREASED CORE FLOW REGION (ICFR) ELIMINATION OF ARPM TRIP SETDOWN NEW RBM SYSTEM AND CORRESPONDING TECH SPNC MODIFICATIONS i

1 s i i i i MAXIMUM EXTENDED OPERATION DOMAIN BENEFITS INCREASED MANEUVERABILITY, FLEXIBILITY AND CAPACITY FACTORS REACTIVIT/ COMPENSATION FASTER STARTUP ELIMINATE APRM SETDOWN LOAD FOLLOWING f l ELIMINATE UNNECESSARY RBM ROD BLOCKS i FLOW CONTROL SPECTRAL SHIFT r i PUMPlNG POWER BENEFITS l OTHER FEATURES OPTION - AUTOMATIC FLOW DEMAND LIMITER COMBINED WITH FFWTR TO MAXIMlZE FUEL CYCLE ECONOMICS i I

~ ~ i. e i. FINAL FEEDWATER TEMPERATURE REDUCTION OBJECTIVE ALLOW RATED THERMAL POWER OPERATION WITH REDUCED FEEDWATER TEMPERA TO EXTEND AN OPERATING CYCLE TARGET RATED FEEDWATER TEMPERATURE'OF 250*F M BENEFIT IMPROVED CAPACITY FACTOR RELATIVE TO EOC COASTDOWN IMPROVED' FUEL CYCLE ECONOMICS COMBINED WITH MEOD TO MAXIMlZE BENEFIT

.{ c i. OUT OF SERVICE EQUIPMENT 1 ? OBJECTIVE k ALLOW POWER OPERATION WITH VARIOUS EQUIPMENT OUT OF SERVOCE i OUTOF SERVICE EQUIPMENT INCLUDES: FEEDWATER HEATER (S) MAIN STEAM ISOLATION VALVE (S) t SAFETY RELIEF VALVE (S) RECIRCULATION LOOP BENEFIT IMPROVED OPERATING AND MAINTENANCE FLEXIBILITY INCREASED AVAILABILITY t PROTECTION AGAINST COSTLY PLANT SHUTDOWN - INSURANCE POLICY

... ~ REACTOR PERFORMANCE IMPROVEMENTS LICENSING SERVICE UCENSING ANALYSIS REPORT SUGGESTED TECH SPEC MODIFICATIONS ( LICENSING SUPPORT PRESENTATION / DISCUSSION QUESTIONS / ANSWERS 10CFR 50-59 EVALUATIONS SIGNIFICANT HAZARD CONSIDERATION l FSAR UPDATE l 13 o

e a < i; i CURRENT LICENSING STATUS l 1042%POWERUPRATE DUAEARNOLDLICENSEDANDOPERATNG LEBSTADT PROVISIONALLY UCENSED AND OPERATING SEVERAL BWR PLANTS COMPLETED FEASIBlU1Y STUDY MAXNUMEXTE!OEDOPERATIONDOMAN LElBSTADT, GRAND GULF, APO KUO SWNG UCENSED APO OPERATING PERRYFSARAPPROVED MANY BWR PLANTS LICENSED AND OPERATNG WITH ELLLA AND ICFA HATCH, DUANE ARNOLD, MONTICELLO AND KKM LICENSED AND OPERATING WITH ARTS FNAL FEEDWATER TEl#'ERATUE FEDUCTION SEVERAL BWR PLANTS UCENSED AND OPERATING OUT OF SERVICE EQUIPMENT - FWH, MSIV, SRV, SLO PERRYAND GRAND GULF FOR FWHOS SEVERAL BWR PLANTS LICENSED AND OPERATED WITH SLO WORK IN PROGRESS FOR SEVERAL BWR PLANTS

4 i i I c PLANT LICENSING STATUS PLANT PRCDUCT PU MID ELLLA ICFA ARTS FFWTR FWHOS SLO

• L_.

l UNE W/ICFA DOGIESTIC PLANTS DESMN283 3 X MILLSTONE 3 X X MONTICELLO 3 X X X PEGRIM 3 X X X X OUAD Cmts 1&2 3 X BROWNS FEW 1-3 4 X X

s BRUNSWICK 182 4

X COC?-Zu-- 4 ~ X DU?.5 ,")LD 4 X X X X - FERM12 4 X isizrATRICK 4 A HATCH 1&2 4 X X X X X [ HOPEChetK1 4 X PEACHButiUM 243 4 X X X s SUSOUC T3rr.1&2 4 X X VERunNT Vf_""EE 4 j LIMERICK 1&2 4 X X .. = uriu 2 5 l LA SALLE 152 5 X NhP2 5 X X GRAND GRF 1 ti X X X PtMMY1 6 X X X X RNEFMEND 1 5 X CLINTON 1 6 sesI ri n IL Ra ia X NUGLtnutt 3 X GAursau 4 X G ,O "7152 4 X X FUKUSHNA 2 4 MUHLEBERG (MMM) 4 X X X X Cut MtMits 6 X KUD SHtNG 152 6 X LtIB5I AU T (MKL) 6 X X i X X {

• ANALYSIS COMPLETED-LICENSING EITHER COMPLETED OR UNDER IMPLEMENTATION 1

~

~ BWR/3 MAJOR NO-12%) POWER UPRATE LICENSING STATUS FEASIBILITY STUDY COMPLETED FOR TWO BWR/3 PLANTS GE-NRC MEETING FEBRUARY,1986 PROVIDED OVERVIEW TO STAFF ~ STAFF GENERALLY RECEPTIVE OBTAINED NRC OPINIONS ON LICENSING ISSUES MEETING WITH MR. BERNERO MARCH,1986 . jpENTIFIED NECESSARY LICENSING STEPS AND ANALYSES DEVE. LOPED ACTION TO ESTABLISH LICENSING PATH AND PROCESS LICENSING ACTION PLAN DEVELOPED FOR IMPLEMENTATION f i 16 I

.. - : : :-.-:.z =.

.- e --===.:=:=_-.;===- . :. u.. u. = : = = - = -

i l t i l LElBSTADT OPERATING EXPERIENCE WITH REACTOR PERFORMANCE IMPROVEMENT PROGRAMS I ( l i i i l 17 i ~r - ~~; -' ;;- t::i'? - ~ - - -

- - in- ~ r -. ~

- - ~ e ce e

-,-=*em--~==~--~~----------~^'-------

p LElBSTADT PLANT DEFINITION UTILITY: KERNKRAFTWERK LElBSTADT 4 LOCATION: SWITZERLAND BWR PRODUCT LINE: BWR6 MARK lli POWER RATING: 3012 MWTH,950 MWE RATED CORE FLOW: 88.5 MLBM/HR ( VESSEL SIZE: 238 IN. DIAMETER NUMBER OF BUNDLES: 648 GE-6 FUEL DESIGN CONVENTIONAL FUEL LOADING (NON-CCC) POWER DENSITY: 54.1 KW/l STATUS: COMPLETED CYCLE 1 & CYCLE 2 ' ~ i l 18 -_r~. ,, -., ~ -.. -.,. _,. .,.,,-3 .. ~.

i LElBSTADT REACTOR PERFORMANCE IMPROVEMENT PROGRAMS l I r MAXIMUM EXTENDED OPERATING DOMAIN (MEOD) OPERATION WITH EXPANDED POWER / FLOW REGION RATED POWER ALLOWED FROM 75% TO 105% CORE FLOW SINGLE LOOP OPERATION (SLO) i EXTENDED OPERATION WITH ONE RECIRC LOOP UP TO 70% POWER 3 FINAL FEEDWATER TEMPERATURE REDUCTION OPERATION WITH REDUCED FEEDWATER HEATING TO EXTEND THE FUEL CYCLE RATED THERMAL POWER ALLOWED FROM 420*F TO 250*F FEEDWATER POWER UPRATE ~ OPERATION AT 104.2% THERMAL POWER OPERATING STATES EXTENDED TO ACHIEVE MAXIMUM OPERATING FLEXIBILITY WITHOUT MAJOR HARDWARE MODIFICATION

i r + i CORE THERMAL POWER (% OF RATED) (% OF UPRATED) t 120 110 CONDITION i KKL 104.2% POWER UPRATE N ion gg OPERATING MAP g e / MAX MEOD BOUNDARY So / CORE' 80 % FLOW / ~ n i \\ I RATED ROD / 7e 3 LINE / 1 / ~ so RATED S FLOW / l 50 50- / NSLO / a a / NAT CIRC / \\ Significant increase in Allowed 38 Operating States Achieved Via CAVITATION N 20 20 i LOW SPEED ~ o Design & Licensing Evaluations MAX VALVE I 10 POSITION 9* o Technical Specification Changes 8 o o Minor Setpoint and Hardware t 1" 128 Adjustments CORE FLOW (% OF RATED) w

(; KKL LICENSING STATUS MEOD--LICENSED SLO--LICENSED FFWTR--LICENSED POWER UPRATE 102% POWER LICENSED 104.2% POWER TESTING b INITIAL DEMONSTRATION AT BEGINNING OF CYCLE-2 APPROVED EXTENDED DEMONSTRATION TEST FOR REMAINDER OF CYCLE-2 APPROVED LICENSING REVIEW AND DISCUSSION IN PROGRESS FOR CYCLE-3 UPRATE OPERATION PROVISIONAL LICENSE FOR CYCLE-3104.2% POWER OPERATION J 21 ._..,.7_.._ . ;-...,, _ __. ~ ,g_ 3 n,

~ KKL OPERATING EXPERIENCE ~ STARTUP TEST DEMONSTRATION CYCLE 1 OPERATING PARAMETER 98% AVAILABILITY; 96% CAPACITY FACTOR 12% MARGIN ON MCPR 2% MARGIN ON KW/FT ~ 3% MARGIN ON MAPRAT 90% TO 102% CORE FLOW RANGE AT RATED POWER ( CYCLE 2 OPERATING PARAMETER 99% AVAILABILITY; 98% CAPACITY FACTOR 6% MARGIN ON MCPR 2% MARGIN ON MLHGR 4% MARGIN ON MAPRAT 75% TO 104% CORE FLJW RANGE AT RATED POWER 104.2% OPERATING POWER I-FULLY UTILIZED EXPANDED OPERATING MAP FOR l STARTUP AND RESTART 22

• --------I'

' - - - ~ ^

N' e KKL ROD ADJUSTMENT TO REACH 104.2% POWER FOR TESTING unow 48 45 30 M 90 05 M 75 N 05 W 95 100 c i I I I I I I I i i L l $/ a suntserone ~" v DOWNPOWER MANEUVER 8 ] noW 5 85 l commx. i nene a t i ? l n-i 1 l l se 'e i

=

a i eawe j a l noos l MOVED i / ~ 55 i

H e l; j, POWER OUTPUT FOR STANDARD vs MEOD L,. FLOW CONTROL OFERATING STRATEGY SCHEMATIC i. + g -- OPERATION WITH FLOW i< CONTROL MEOO (80-105%) E.xC.] STANDARD OPERATION CAPACITY. i FACTOR LOSS,(1+%) 1" i-NT ~~"M M mij T f\\7 l. RODSEQUENCEEXCHANGE i lll l L I i

I l

3 = l;l c j l m ^ 1 1 Il j1 n ROO 4osuST=ENT l i i; MANEUVER ~. Il l; e s = n b 80 li I i 1 70 i $ BEG ] CYCLE, END OF CYCLE 03 10 20 30 40 230 240 250 260 270 l TIME (DAYS) )-

O hi KKL HIGH XENON RESTART LOAD LINE D l % POWER (100% = 3012 MWI) o l! m 100 n.Ow I. CONTROL EOUll.18RfuM 90 XENON CONDITION u* d'Q 80 r% b h 5 70 I BEFORE SCRAM ROD PATTERN ROOS MOVE l I I I I I 40 50 80 70 80 90 100 1 % FLOW l

c 1! g LElBSTADT EXPERIENCE

SUMMARY

L ANALYSIS, TESTING, & OPERATING EXPERIENCE HAS DEMONSTRATED l VIABILITY OF OPERATION IN EXTENDED MODES -ANALYSIS VERIFIED VIA TESTING / OPERATIONS F -EXPECTED MARGINS DEMONSTRATED M PLANT OPERATIONS HAVE BEEN SIGNIFICANTLY ENHANCED -OPERATING FLEXIBILITY / SIMPLICITY -CAPACITY FACTORS 4 SIGNIFICANT LICENSING EXPERIENCES HAVE BEEN OBTAINED em e 9 f I i

l. ~~j ,y e

SUMMARY

MANY BWR PLANTS HAVE BEEN SUCCESSFULLY LIC8NSED WITH THESE REACTOR PERFORMANCE IMPROVEMENT PROGRAMS MOST OF THESE PLANTS HAVE OPERATED WITH THEM RESULTING IN' SIGNIFICANT ECONOMIC AND OPERATING FLEXIBILITY BENEFITS ( 27 / y,em.._,, ;-. ~, -, ., -.. - ~ s

{f ~ i 't PLANTOPERAT!ditPERFORMANCEINE0yEEENTSOFTRE GENERAL ELECI1LIC (GE) BOILING MATER REACTORS (BWR'S) T. C.12I General Electric Ceapany R. C. STIEN, General Electric Caspany 175 Curtaer Avenue, M/C 740 175 Curtser Avenue, II/C 740 San Jose, California 95125 saa Jose, Califorata 95125 { (408) 925-6136 (408) 925-4139 Ass m er This paper summastises seen of the plant contro111ag roeirculaties f1sw and esatrol rode i operation performance tapreveneet techniques la the allevable licensed thermal power / moderator developed by the General Electric Company flew operation resten. N BUR's lead fellowing Ihiclear Energy Busiasse Operation for the potential and capability have been successfully i Seaeral Electric Boiling Water Reactore (CE demonstrated. Enaspies of these discussione and BWR's). h ough the use of both thermal and demonstratione are presented in References 1 and plaat hardware operettas astgias, substantial 2. additional fleetbility la plant operation can be achieved resultlag la significant taprove-2., Beacter Vater Becirculaties Sye' ten: monts in plant capacity and availability facter and poteattel fuel cycle oceaemics for the The GE BWRa are designed with the reactor eartently operating er requisition GI BWR plaats. unter restreslation systen to circulate the Thia list of techniques facludes esponding the required seelaat threagh the reacter core. For BWR thermal power / moderator flew operettas deasta the GE BWR 5 and 6 product lines, the synten i to the maniana achievable region, operation with ceanists of two leopa enternal to the reactor a stacle recirculation leep out of service and vessel each contatalas a pump with a directly operation at rated theresi power with reduced coupled water-cooled (air-weter) aster, a flew feedvetor temperatures. These plant improve-eentrol valve and two shutoff valves. Bigh per-mente and t,perettaa techniques can potentially formance jet pumps' located withis the reactor t sacrosse plant capacity factor by II to 21 and vessel are used la the EUR rectremistion system. i provide addittomal fuel cycle acomonica savings h jet fieupe, which have se arving parte, to the E BWE's evaars. provide a,sentismene internal circulaties path for a asjer pertion of the core coolant flow. 2NT3000Cf!M 3. Therasi power coastdown capability: The design of the current operettas General Electric Boiling Water Beactors (CE Cycle atension initiated at the end of a BWRs) la based en the proper seabinaties of assy aeraal feel sycle seuld be obtatand by permitting design variables and operettas esperience. the reacter thermal power level to decrease 4 i These factors contribute to the achievement of gradually af ter all control rede have been fully reliability, performance and fuel cycle econsey. withdrena at the end of as operating eycle. As j gone of the esseples that accounted for the the core reactivity decreases due to fuel deple-signif tsamt fleatbility that antated la the tion, positive reactivity is introduced into the a I sporettaa of a GE BWR to date are givest below: eere by reducing the sere therent power level. l Thermal power seestdown utillees the power co-l 1. Lead Following Capability afficient to obtain the reactivity gain to estead the fuel eyele. h and-of-cycle seastdeva method r As the difference between peak and stain e le being used esteme19ely by EUR evnere to stretch lead densed sacreases and the pereestage of out more emergy est of the asisting fuel to improve j suelear generation capacity la utility system the fuel eyele oceaany. 1 grows, it beceans more important to perfota ) lead followise operetten and to aska the grid Due to the flexibility and eeneervative margia i j systen more flanible. Lead following to a term is the GISWR design, further enhancement la opera-describing a power plaat whose power le raised tien performance can be ande to achieve higher and lowered to meet the day-to-day demand of pleet capacity factor. This paper suasarises its electrical grid. GE SWR power plante ese three plant operation perferasece improvement heetcally change power easily and rapidly by techniques that are developed la relationship to t l E4-1 i " rz: u :.==.

w. *

= .. = = = - - - -

h, ~ the three examples given above to further cation allowable thermal power / moderator f1ow enhance the performance and flexibility of the operation region for the operation of the CE CE BWRs. BWR is defined by: DESCRIpT1CN a) The 1002 rated power condition b) h e 1001 rated steam flow power In addition to the performance and flex 1 constant control rod line bility demonstrated by the saisting BWR opera-c) The 1002 rated moderator core flow ting plants, the current lines of the CE EWRs d) Recirculation system component have been designed with significant conserve. cavitation at low power tisas and margine. Based on operating experi. e) Itinima core flow resulting from once established by CEEUR's over the years, the pump speed or re*:freulation flow control feasibility has been demonstrated that by valve position utilising some of these design and operating margine (both thermal and plant hardware), The expanded power flow map *2nder the further enhancements in BWR plant operation maximum extended operating domain improvement flexibility can be achieved. This is ac-progran can be divided into 3 regions. comp 11shed by defining new modes of operatione and performing safety and impact analyses and/or a) Espanded operation in the lower than providing hardware modifications to justify that 1001 rated core flow region which is all requirements ander the Code of Federal termed eatended load line region. Regulations are met for these operation tech. sigues. b) Expanded operation in the higher than 100% core flow region which is his section summarises these three termed increased core flow region. opecial modes of operation developed for the Gs3WR's by use of the above discussed opera. c) Espended operation above 100% rated ting margins. They are Power which is termed power uprate region. 1. Operation in a modified reactor operating domain by eatending the existing A typical CE BWR power / flow operating thermal power /moderster flow operation region. amp with the esisting allowable operating region This is termed " Maxima Estended Operating and the typical maaisua estended operating domain Domain" in this paper. This mode of operation is presented in Figure 1. The eatended lead line enhances the SWR to reach rated power faster, region is operationally feasible due to the maintain rated power longer and further esisting operational thermal limit margins at enhances the load following capability. these operation conditions as the result of improved plant hardware protection designs. The 2. Operation with only one recirculation increased core flow region is available due te system flow loop running. This is called the reactor water recirculation systen escess " Single loop Operation" in this paper. This design capacity. The power uprate region is mode of operation provides insurance to available due to the additional design margin protect against recirculation system equip. utilised at 100% rated design power. ment malfunction to maintain the CE BWR reliability. Safety and impact evaluation consistent with the analyses documented in the esisting 3. Operation with reduced rated feedwater Standard Final Safety Analysia Report (TSAR) are temperature at rated thermal power. His is required to justify operation in this extended called "Feedweter Temperature Reduction Opera. domain. Reformating of a few Technical Specifi. tion" in the paper. H is mode of operation cation and minor plant hardware modifications provides an alternative to thermal power are also necessary. coastdown to extend an operating fuel cycle as well as provides protection against Several of the currently operating gWRa feedwater heating systes equipment f ailures. have demonstrated rated power operatien over a core flow range of 871 to 10$1 of rated and i A. Maximum Extended Operatina Dossin operation under the corresponding constant control rod position load line. For the CEgVR 6 product 1. Description line, fossibility studies and licensing evalue. tiene using Nuclear Regulatory Commise1on (NRC) h e existing BWR Technical Specif1 approved methods and codes have shown that rated i K4 2 _y.. ~~..

~ G power operation over a core flow range of 751 to In addition, the extended load line and 105% of rated and operation under the correspond. increased core flow region provides the required ing constant control rod position load line are flow range for optimised flow control spectral both safe and operationally feasible. Several shift operation. This* node of operation offers of the currently operating W R's have been uprated fuel cycle advantages over standard operation at from their initial thermal power rating to a new rated power and flow. reting ranging from 102% to 105% of rated. Flow control spectral shift is a method 2. Benefits used to increase fuel ut111:stion thereby reducing fuel cycle costs. The concept of flow In addition to the added load following control spectral shif t is basically to operate capability with the espar.ded operating region, at rated cora power and the lowest core flow the primary benefits of the maximum estended possible throughout the cycle until the "all operating domain fall into two main areas: rode out" condition is reached. Core flow is Reactivity compensation beyond midpoint of the then gradually increased to maintain rated power operating cycle and plant startop with either until maximum core flow is attained. Operation zenon or zenon free condition. at rated power and reduced core flow results in a neutron energy spectrum shif t due to the In order to ensure that 81* or more excess higher core average void content. The higher flow capability exists at rated power for exposure energy spectrum achieved during reduced core flow compensation, additional operating room is re. generates more plutonium and results in a more quired above the rated rod line in order to cou. bottom peaked power distribution.. In addition, , pensate for power reduction during plant startups uranius ut111:stion la improved since a larger with transient menon. Croes power reduction prior fraction of the fission power is generated from to re-establishment of equilibrium menon condition U233 fast fissions. As core flow is increased at rated power have been observed, from plant toward and of cycle, the power shif ts to the top operating experience, to be as great as 101-12g of the core. This results in an increase in during startups with peak zenon and 8-10% during cycle exposure length because power is shifted zonen free startups. In either case, excess flow to a region of lower esposure and higher fissile capability of approximately 25% would be required plutonius content and the void fraction is in order to ensure that subsequent to attainment decreased resulting in increased reactivity. If of menon equilibrius, the plant would have an constant cycle length is desired, initial U235 additional 8% of excess flow available for loading can be decreased. reactivity compensation. Therefore, in order to achieve rated power in the shortest time possible It met be noted that optimited spectral and to maintain 100% power thereaf ter, it is shif t mode of operation is no,t, part of this advantageous to allow for operation above the improvement program. This program merely standard allowed 100% rod line during power provides the ground work and necessary flow range ascensions. The entended load line region required to accomplish, spectral-shif t operation. provides this operational flexibility. CE estimates a capacity factor improve-Once rated power is achieved, the addi-ment ranges from 0.31 to 11 from the entended tional flow range ar. rated power as provided by load line and increased core flow region. plant both the extended load line region and the in. benefits very between GE BWR product lines and creased core flow region can be used to compensate existing plant modifications in place (e.g., for reactivity reduction due to fuel burn-up pellet-Clad-Interaction PCI resistant fuel). The during the operating cycle. Rated power can be additional potential fuel cycle economic improve-maintained longer without control rod movements. monts due to spectral shif t operation ranges from This is especially beneficial late in the opera-2 to 3% depending on flow range and operating ting cy' le if, doey . control rod manipulation strategy utilised. i c is not recousanded. The obvious benefit of the power uprate region is increased power production. The actual

• 8% is the estimated flow change required to com-
• * "** II"8 # *
• I* '8 "'"

I

• • • • • "" I' * ****
• pensate for reactivity reduction for about two weeks of exposure. 1he time represents the nominal interval between power reductions for 3.

supporting Analysis surveillance tests and power shaping rod In order to implement operation in the K 4-3 . q _ ._.c__

( 4.... maximum extended operating domain, safety and practice. Ultimately, these limits translate impact evaluations consistent with the analyses to a maximum

• core thermal power in the range of documented in the existing Final Safety Analysis 651 to 75% of the rated thermal power level Report (FSAR) are required. The software dependent on the diffirent reactor water re-package includes the analyses to support safe circulation system designs for various product plant operation in the proposed entended region lines of the GE BWRa.

as depicted in Fig.1 including a licensing report for subelttal to the appropriate Nuclear Almost all currently operating CERWRs Eagulatory Agency / Commission and the recommended have esopleted the licensing evaluations for changes to plant Technical Specifications. The single loop operation. A few of these plants hardware portion includes the modification to have successfully operated with only one re-the Neutron Monitoring System to accousodate circulation loop running up to as long as the extended operating region. The detail scope two months to prepare for the maintenance work of the required sof tware and hardware analyses required to repair the malfunction component and modifications has been developed by the at a planned outage. General Electric Company. In general, the highest feasible rod line (spper bound of the 2. Benefits entended load line region), the highest fossible The asjor benefit for single recircula-core flow and the highest sprate power level will be determined from thermal, reactivity and tion loop operation is protection against costly stability consideration as well as plant hardware plant shutdown due to recirculation system capability, vibration and internal pressere dif-egotyment failures resulting in improvements forence evaluations. General Electric Theresi in plant availability and corresponding Analysis Basis (CEft.B) evaluation, core wide capacity f actor. A recirculation loop can transients loss of coolant accidents and con-become inoperable due to malfunction of one of tai ment responses are some esemples of the several component items (e.g., recirculation required analyses. Detail analysis scope is to pup, pump actors, valves, etc.). The actual be provided to the BUR stility owners interested capacity factor improvement realised will in this plant improvement progras. depend on the achievable power level during operation with single recirculation loop. 3. Sinale Recirculation Loop Operation Based on operating plant asperience and data. General Electric estimates single loop opera-1. Description tion will result in a capacity f actor improve-ment of appreminately 0.1%. This is estiasted The standard US EUR plant Technical based on the average number of days downtime Specification specifies a 12 hour operation limit due to recirculation loop problems and the when one recirculation loop is inoperable. This probability that it will happen during a BWR results from the fact that analyses were not plant lifetime. Another benefit of single originally performed to justify extended single retiresistion loop operation is that it allows loop operation se part of the GESWR standard for maintenance schedule flexibility. Unplan-plant offering. For single loop operation, the ned recirculation loop maintenance can be post-core pressure-drop is reduced, the total discharge poned until a planned outage or a period when flow from the active bank of jet ymps increases power demande are low. at raced (2 loop) drive flow, flow thronsh the inactive loop reverses direction, and the jet The ability to operate and produce pump flow pattern in the reactor lower piene power with single recirculation loop with no becomes highly asy m etric relative to rated time limitation is a sound inst.rance policy for conditions with balanced two loop inlet flow. any GE BWR owner to have in place. Bowever, the core power distribution remaine l unchanged because the core coolant flow distri-3. Supporting Analysis bution is not altered during single recirculation loop operation. In 6tder to tuplement single recircula-tion loop operation, safety and impact evalue-Since single loop operation was not a tions consistent with the analyses documented plant design basis operating condition, it is in the aristing Final Safety Analysis Report necessary to establish a proper technical beste (FSAR) are required. This software package for operating in this mode. This implice that includes a licensing report for submittal to a broadly based analysis suet be performed to the appropriate Nuclear Regulatory Agency /Coe-establish the limiting factors that still satisfy mission and the recomended changes to plant i i the many criteria applied in normal design Technical Specificatione. The detail scope of E4-4 i l vm

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~

hm sS, ,'r.s-3: j r / g r-t i \\( \\ p. the required analyses to esta611eh the bases equilibrium.menos yield reactivity gains which i and limite justifying pale operation h single compensate for reactivity losses due to deple -

• '/

leep operating mode has been developed by the tion of flesionable anterial. This method of General ylectric Company. In general, a new cycle estension results la relatively rapid i break spectrum analyets for the loss of coolant rate of decrease of electrical output resulting . I accident during eingle loop operation will be from reduced core thermal power. performed.' General hectric Thermal Analyste l, f Baete (CgTHA evoluuson, core wide transiente, Final Feedvator Tamperature Reduction 2 contaiment tasponses and stability considera-extends a fuel cycle by asistaining rated core ties are soep of the asamples of the required thermal power. This is accomplished by (. analysee., harious 'other safety and operational systematically volving out the feedwater J. impact eva.'uatione snich address vessel heaters. Reductica of feedwater heating causes vibration.ireverse flow phenomenon, flew colder water to flow f ato the reactor vessel. ',/ macertainties, wate chneistry, cavitation. Colder feedwater sizes with the reactor water i recirculation flow coattel, neutron noise and sacreases core subcooling. thereby decreases i / 'teutron monitoring system setpoints will be ste m voide in the core which in turn increases j performed. Detail naalyste scope is to be pro-core reactivity as more moderator is available. i I vided to the BWR utility owners interested la This sede of operation requires moderator i this program. To optimise the benefit and minimise perturbe-control to maintata rated core thermal power. {? - C. Feedwater Temeerature Reduction tion to plant operation feedvet,er temperature

{i

• 023rgion, should be reduced in several gradual steps.

[ These step changes occur periodically to com- ,f 1.gdescription pensate for the continually decreasing core

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reactivity due to fuel depletion. This gradual l[ The standste US BWR license constrains reduction in feedwater temperature extends the j g plant rated thermal power operation at feedwater fuel cycle while maintaining core therust power i' ,?, temperatures corresponding to the rated design at or near rated throughout the entire extended i f ( values. Operation with reduced feedwater operating time. Electrical power output de-l temperature at rated power can be separated into creases during Final Feedwater Temperature two categottee., One le operation during an opera-Reduction operation, but the rate of decrease ting cycle and cLe other le operation at the end is less than that of 10C coastdown. As core l of the fuel cyc'e. inlet subcooling incrosses, more thermal nega- \\ watts are used to heat up the moderator and j Opetstton at reduced feedwater tempera-therefore less steam to produced. As steam i i I ture during an operating cycle is termed " Feed-production decreases, electrical power output water Bester(s) Out of Service Operation" in thie in tura decreases. The reduction in electrical t paper. D ie. mode of operation is used in the power output for Final Feedwater Temperature l \\ event that a malfunction of the feedwater heating Reduction le approsisately 1-1 1/21 per week I system occurs reducing feedwater temperature and whereas and of cycle coastdown results in a j eentinued operation of the plant la highly 21/23 redettion per week. This results in a desirable to maintain power outpet. Operation five to tea percent increase in station with reduced feedwater temperature at the end of electrical power output ever coastdown capabi-1 an operating cycle is termed " Final Feedwater lity through the period of extended operation. Temperature Reduction" in this paper. Die mode t of operation to sfeed toward entending the opera-These modes of operation with reduced ( ties cycle to improve capacity factor (relative feedwater temperatures duries the operating T1 # to the thsenal power coastdown method) and fuel cycle as well as at the end of the cycle is t cycle scenamics. feasible due to esisting design margine in the i. BWR plant. One example of this le the margins l Final Feedwater Temperature Reduction to thermal fatigue usage on the feedwater L to as alternattve to the traditional method of mossles and sparsers. Although reduced feed-and of cycle (BOC) thermal power coast-down used water temperature operation increases fatigue antensively by,3WR ownere to entend se operating on the feedwater nossles and spargere, the i } / fuel cycle. The coast-down method of entended design margins os both of these components / b operation is accomplished at redeced thermal allow the operation in these reduced feedwater f power but with cornal feedwater temperature and temperature modes with only minor time limita-steen pressure. Continued operation with thermal tion at reduced feedveter temperature operation power coastdown is possible because reduced steen during an operation cycle. velds, reduced fuel temperature and reduced R4-3 i / )l -e A w.. ,..n w,

y 9 Proper technical bases need to be temperature reduction modes of operation, established for these modes of operation. safety and impact evaluations consistent wi,th the analyses documenthd in the Final Safety Safety and equipment impact evaluations are Analysis Report (FSAR) are required. This therefora required to justify operation in these feedwater temperature reduction modes. Minor sof tware package includes a licensing report technical specification modifications are also for submittal to the appropriate Nuclear required. Regulatory Agency /Cousission and the reconsended changes to the plant Technical Specifications. Several of the currently operating The detail scope of the required analyses to BURS have used Final Feedwater Temperature justify operation in these feedwater temperature Reduction to extend their operating cycle. For reduction operation modes has been developed by I the CE3WR6 product line, impact and licensing the Ceneral Electric Company. General Electric evaluations have shown that end of cycle rated Therail Analysis Basis (CETAB) evaluations, core thermal power operation with,F (down from 420,F wida transients, loss of coolant accident, a feedwater tempergture reduction of 170 contaissent responses and stability considera-f to 250 F ratsd) is safe and operationally tion are some examples of the required analyses. feasible. Evaluations have also been completed Yarious other safety and operational impact for some BWRa for Feedwater Heater (s) out of evaluations which address feedwater nossle and service operation during an operating cycle at sparser fatigue, reactor protection system rated theras1 peger with e feedwafer tempegature setpoints, tapact on internals as well as reduction of 100 F (down from 420 F to 320 F optimised feedwater temperature reduction steps of rated for the CE3WR6 product line). from both heat rate efficiency and compliance with fuel design standpoint are to be performed. 2. Benefit Detail analysis scope is to be provided to the BWR utility owners interested in this program. =- Feedwater Heater (s) out of service operation during an operating cycle protects Q)NC1.USION against costly plant shutdown due to feedwater heating system equipment malfunctions, thus pro-The current lines of operating CE BWRs vidi::3 availability and capacity factor improve-have demonstrated excellent operation flexibi-ments as well as operational flexibility by lity and performance. In addition, the CE BWRs allowing maintenance schedule to be arranged at have also been designed with significant con-a planned outage or a period when power demands servatisas and margina. By use of some of these are low. Final feedwater temperature reduction thermal and plant hardware operation margins, i offers significant fuel cycle economic advantages flazibility in CE BWR plant operation can be through lower fuel cost resulting from cycle further enhanced resulting in substantial extension while at the same time reducing the capacity factor improvemente, protection against l capacity factor losses incurred by the more costly plant shutdown due to equipment failures traditional end of cycle thermal power ccastdown and subsequently economic benefits to the GE method of extending fuel cycles. For some BUR owners. These plant operation improvement product lines of the CE BWR's, capabilities exist technigoes can potentially increase plant thagareductioninfeedwatertemperatureupto capacity f actor by 12 to 22 and* provide 170. F is feasible. This reduction in feedwater additional fuel cycle economy savings. The temperature can extend a cycle by 1000 75tD/T capacity factor improvements could roughly be exposure or up to 2 months while maintaining translated to a cost saving of about 2 to 4 ]001 thermal power. Although electrical output million US dollars per year (1934 value) to a will fall off during the extended period, the CE BWR owner stility based on an approximate decrease is only about half of what occurs during average equivalence of 12 capacity factor equals an end of cycle coastdown. to 2 million US dollars per year to a CE BWR owner utility. General Electric estimates these feed-water temperature reduction modes of operation Plant safety and impact evaluations are will result in a capacity factor improvement of required to justify these special modes of about 13 per cycle of BWR plant operation over operation. Reformating a few Technical Speci-the thermal power coastdown method of cycle fications and minor plant hardware modifications extension with additional fuel cycle economic are also required. The Plant Operation improvements. Forformance Improvement Programs described in this paper have been proven to be operationally 3. Supporting Analysis feasible and have been successfully demonstrated by several operating CE SUR owners throughout In order to implement these feedwater the United States and the rest of the world. K4-6 m _, _ _.

( ~ ~ RETERENCES 1. F. N. Ontko, General Electric Company. " lead Following at Brunswick" presented at the American Nuclear Society Winter Meeting. November 12-16, 1978 Washington, D.C. l' 2. D. C. Carroll, R. C. Serenka. E. R. Propst, General Electric Company. "3WR Maneuvering Capability", presented at the American Power Conference. April 23-25, 1979, Chicago, Illinois. REGION 1: Extended Lead Line Region REGION 2: Increased Core Flow Region i REGION 3: Power Uprate Reg' ion RATED ] POWER FLOW FOINTb 4 1 M Rated Bated 2 Rod Core w Line Flow I / l Natural g Circulation j E E Low Recirc. / Speed Valve / Max. Foeition Cavitation Region I l e CORE Flow (I RATED) FICURE 1 TYFICAL CE BWR F0VER/ FLOW MAP E4-7 l ) 1 --~-.I -y;. ..y .-- - - 7 r ---- --, - -,, - - -

h"e s ,n e SENERAL ELECTRIC WR/6 TRERMAL p0WER M' RATE CAFABILITY ASSESSIENT T. C. Lee, General Electric Company 175 Qartner Avenue, M/C 740 gen Jose, California, 95125 (40s) 925-6519 ABSTRACT Nost of the CE W R/2 and 3 prodi,et line pisats have already uprated their initial power An assessment study was performed to determine ratings to their design rating of approximately the feasibility of tacreasing the core thermal 2041. N CE W R/4 product line plants have power to approximately 104.2% of the original also been coming on-line la the past ten years licensed rated value is order to obtata an increase and their performance at the initial power level in the core steam flev rate of 31 for a typical is also well established. For these plants, General Electric WR/6 power plant. Areas which efforts have been oderway to uprate their are espected to be affected by the increase in thermal power level to the designed 1041 rating. the rated thermal power were evaluated in the study. Results of this assessment study show that, h GE WR/6 product line is the sixth overall, the 5% increase is steam flow and the generation of the design evolution of the CE associated 4.21 increase in core thermal power WRs. N W R/6 product line is capable of are feasible for a typical W R/6 plant. Bowever, producing up to 201 acre power from the same reduced fuel therus1 margins any constrain size pressure vessels as used in the W R/5 operation at the 104.2% uprated power levet d: ring product line without increasing the size of the part of the operating cycle with current CE6/7 respective buildings or supporting systems. fuel designs. Improved CEg8 fuel design with a Therefore, the WR/6 product line represents the 14.4 kw/f t linear heat generation rate limit, highest core aver m power density plants of all coupled with recently IBNRC approved more realistic of the CE EWR product lines. less-of-coolant accident licensing models and assumptions, should allow operation at the uprated The design basis thermal level of BWR/6 l power level during the entire operating cycle. plants also represents an approximately 4.2% This assessment coeclusion for a typical W R/6 margin over the commercial and licesing basis plant is derived aminly based on the reactor core rated power level. Recently improved calcula-performance e,apability of the plant. por specific tional and test data reduction / verification plant capability assessment application, detailed methods have demonstrated considerable extra systems and epipment tapact, as mipely applied safety margin. These existing analytical, to each individual plant, are to be evaluated. It system and epipment margins can be used to t is also felt that, based on assessment results of increase the operating thermal power above the this study, significant anjor power uprates beyond originally licensed rated power level. With the 4E level is potentially feasible for WRs in sufficient agineering assessments, licensing the future with fuel and plant modifications and analyses and confirmatory testing, it is possible testing verifications. to sacrease and license the WR/6 at 4.2% higher thermal power reealting in a 5I facrease INTRODUCT10ll da core steem flow production and a corresponding l increase in electrical power generation. The h General Electric (CE) Boiling Water obvious benefit for the thermal power uprate is Beactors (WRs) have bem designed with significant increased power proudction, resulting la a esaservatism and agineering design margins. Based economic benefits to the WR owners. en operating experience established by the 3WRs over the years, it has bem demonstrated that some General Electric has developed a two-step I of the existing analytical, systems and epipment power Uprete program for the WR Owners to j margins that are a part of the CE BWR plant design implement the power uprate for their WR plants. een be utilised to increase the maxima operating N analytical assessments and epipment power above the current commercial basis rated evaluation, including reviewing esisting plant reactor power rating.. and/or startup test data or definition of i + =..

~ required testing to confirm the plant is capable some adjustments in some of the cooling systems of the uprate, constitute Phase 1 of this program. prior to operation at the uprated condition. Phase 2 covers the plant-specific safety Some NSS system setpaints modifications will licensing analyses and documentation necessary aho be necessary to implement the uprate. to assure long term safe operation of the plant Overall, the conclusion is that these typical and would moraally begia only af ter the successful WR/6 plant systems and equipment have enough completion, with positive findings, of the Phase design margin to accommmodate the 4.21 thermal 1 feasibility study. This two-phase Power power uprate. Details of each of the systems Uprate Program is used for all CE BWR product evaluated are not presented in this paper lines. A Phase 1 power uprate assessment study (except the Turbine Control Valve) as they are was recently performed for a typical SWR /6 very plant-dependent. A discussion of the product line plant. This paper summaarises the Turbine Control valve performance is provided results of a few major subj ects from this capa-because this ia one of the asjor components bility, assessment study for a 5% steam flow impacted by the uprate. '(approximately 4.21 thermal power) uprate for this WR/6 plant. The tapact of the 4.21 power uprate on the CETAB, transient performance and containment FEASIBILITT EVALUATION capability are also not presented in this paper. The overall conclusion of this assessment is The areas that are espected to be affected that there is minimal impact on their perfor-by the 4.2% increase in operating power were mance and capability except some system set-assessed in this typical BVR/6 plant feasibility points will need to be revised. Detailed study. The areas assessed includes technical specification andifications are not presented due to the plant-specific nature. 1. Nuclear Steas Supply (NSS) and Balance of Flant (BOP) systems and equipment. The remaining paper summarises assessment This assessment also included the results of the following major performance Turbine Island Systems, subjects from this feasibility study. They ares 2. Reactor Core Performance. 3. General Electric Thermal Analysis Basis

1) Turbine control Valve Forformance; (CETAB) and Transient Performance.

2) Reactor Core Operatie3 Limits; 4.

Emergency Core Cooling System (Ecci)

3) ECCS 14CA Forformance; loss-of-Coolant Accident (LOCA)

4) Reactor Vessel Internals Vibration.

Performance. 5. Containment capability. This typical BWR/6 feasbility study is per-6. Vessel Internals vibration. formed for the entire power flow operating map 7. Technical Specification mdifications. as defined by the Maximum Estended Operating Domain (MEOD)I program available for the WR/6 These affected areas were assessed both product line. A typical BWR/6 power flow qualitatively and quantitatively. Existing operating map with MEOD is illustrated in startup test

• data were used to assess the Figure 1.

A typical 104.2% power uprate available margins and the tapact of power uprate operating map with MEOD is illustrated in Figure on these margins for the typical BWR/6 plant. 2. In addition, the impact of the power uprate TURBINE CONT 30L VALVE on the various plant performance taprovement undes of operation 1 available to the SWR /6, The Turbine Control Valve (TCV) of a BWR/6 such as the Maximum Extended Operating Ibmain, plant la used to control the steam flow generated Single Recirculation loop Operation and Feedwater in the vessel to the turbine. Beat balance data Temperature Reduction Operation, was also aammined and the valve flow capacity data were used to in this study. determine if the valve can accomodate the 5% incrnnse in steam flow for the uprate. When the j l The various NSS and 30F systems and equip-TCV has enough capacity to handle the additional ment, as well as the Turbine Island impact, are steam flow, operation at the uprated power will l plant-specific. Plant-unique heat load informa-' result in the TCv to be operating higher in the tion, design practice and startup or operating flat region of the TCv lif t curve. Operation at data are required to evaluate the impact of power the uper and of the TCV position will be sub-I uprate on these systems and equipment. jected to a slightly larger travel distance when responding to small variation in steam flow, l These specific systems and equipment were Therefore, in order to limit duty on the TCV as evaluated in this typical BVR/6 plant feasibility well as provide good TCV operating characteris-study. The evaluations show that some adjustments tics, it is desirable to avoid operating the to the turbine protection and control system arv valves at the upper end of their position by ( be required. It may also be necessary to make increasing the pressure regulator setpoint, i A, _ _ _ ~.x

t-Increasing the pressure regulator setpoint to limits examined were: Minima Critical Power lower the TCY operating position will, in turn, natio OtCPR), Linear coat Generation aate (uCR), increase the core operating pressure of the plant. Maxima Average Planar Linear Beat Geeration Rate OIAPGCR), and hot' escess reactivity. The .The BWR/6 assessment study included a review assessment was performed on a typical W R/6 of a typical SWR /6 plant startup data. This plant for an early fuel cycle and an equilibrium review concluded that the TCTs of this typical cycle. WR/6 plant have sough capacity to accamodate the 51 increase in turbine steam flow, thereby For this theruni margin assessment, a not limiting the uprate capability of the plant. coetrol rod pattern study was performed using The data indicated that the capacity of the TCT the CE 3-Dimensional core simulator at intervals at the valve wide open (WC) positics when of 1000 It'd/t fuel esposure through the projected operating at the current 975 pais design TCT cycles to assess reactivity and thermal margins. Anlet pressure, is espected to be about 1051 of Initial attempts were performed at the uprated the rated steam flow. The FCVs are espected to esmditions and various flow conditions to allow be capable of providing rasponsive pressure con-for flow control reactivity coupesation I trol at a position corresponding to about 971 utilising the MEOD. Rated power conditions of the Wo steam flow. Therefore, when operating were utilised for the part of the operating at 1053 steam flow condition, it would be cycle when design thermal margins were less than necessary to provide a higher steam pressure at desired. Bot excesa reactivity was determined the valve in order to pass the higher staan slow by back-burning from the appropriate Maling at the same valve position. An approximation of calculation to the desired cycle esposure and an this effect is 10 poi for each 11 estra steam all Bods Out (ARC) 3-D simulation. flow desired. Thus, a 30 poi incrasse would be needed to achieve the required 31 flow margin. The assesrent results cf the reactor core oper.ging limits are presented it the following A review of startup data of this typical subsections. I WR/6 plant indicated that the steam line pressure drop (at rated steam flow) from the A. Minimum Crici:s1 Powir 1,tio (MCPR) vessel dome to the TCT is about 45 poi which is Operating Limit 20 poi less than the rated power design basis of 65 poi (1040 psi-975 psi). The operating pressure A 4.21 power increase would be espected to drop is aspected to increase to 50 poi at the result in approximately 4% decrease in actual 1053 steam flow condition. Since the present operating MCPR valuu. Projected ar.d exist 2ng WR/6 design limitation on operating reactor dome WR/6 operating data indicare that signifDant pressure is 1060 psia, this means that a nazism margin exists between the MCPR operating limit of 1010 pois steam pressure will exist at the TCY. and the actual operating value at the current This 35 psi increase over the rated turbine inlet rated power level. Therefore, sufficient NCPR pressure of 975 psia should be sufficient tcf pass margin is w.ticipated at the 104.2% uprate con-the 1051 steam flow at a TCV position which is dition. The results of the rod pattern atady equivalent to about 96.51 of WO flow. conclude that MCPR is not the limiting factor for the uprate capability. It is concluded, based on the review of this typical WR/6 plant data and normal design prac-3. Maxima Linear Best Generation Rate l tice, that an operating pressure increase of (MCR) Limit and Maximus Average somewhere between 10 to 20 poi is expected to be Planar Linear Heat Generation Rate acceptable for the 104.2Z power uprate (from 1040 (MAPLHCR) Limit psi to 1060 psi at the done). TCV valve flow capacity, heat balance and steam line pressure Available and predicted SWR /6 data indicate drop data are strongly plant-dependent. The that the margin to the 13.4 Rw/f t MCR limit and 1 assessment and conclusion presented above are the MAPMCR limit for WR/6 plants may be ] determined by reviewing a single WR/6 plant. limiting the ability to uprate power during part If TCV data are not available, pressure control of the operating cycle. Qarrently, the MAPMCR j testing is recomended to determine and confire limit is constrained by the 13.4 Kv/ft M CR 1 the acceptability of operation at this condition, limit but is also very close to being determined and to determine the desired operating and TCv by the 2200*P Peak Cladding Temperature (PCI) ) pressures. limit at some exposures. With a 4.21 power sprate, the MAPLHCR limit may be set by the EEACTOR Q)RE OPERATING LIMIT PERPORMANCE Amersecy Core Cooling System (ECCS) PCT con-straints. The subsequent section on ECCS impact Increasing the core thermal power by 4.21 provides mors discussion on this MAPMCR limit will affect the operating thermal margins for the restriction. reactor core. The W R/6 assessment study was performed to examine the impact of the uprated The results of the 3-D simlator rod power on core thermal limit margin. The thermal pattern study show that the limiting parameters

h t for both the early and equilibrim cycles were IMERGENCY CORE C00LINC *TSTEM FERFORMANCE peak Rw/f t and MAFMCR. The study indicates tlut uprated power could be maintained for Two separate Energency Core Cooling Systems approzinstely 75% of the time for early cycles, (ECCS) performance analyses can spraally be whereas a larger botton-peaked Baling power dis-performed for BWR plants at normal operating tribution in equilibrium cycle potentially legal requirements of 10CFR50 Appendia E model conditions. The first analysis uses standard reduced uprated capability to about 50% of the cycle. If the specific plant exhibits signifi-assumptions and assees one single active com-cant margin to thermal limits through knowledge-ponent failure (N-1 assumption). This analysis ment core annagement, 1041 power could potentially is normally used for most BWR/6 plants. The he usintained throughout the cycle. second analysis assumes one component out-of-service plus one single active component failure It must be noted that this power uprate (N-2 assumption). This analysis normally takes study result is based on a typical EUR/6 plant credit for the Special Ebergency Beat Removal with current CE6 fuel design. With improvement (SEHR) system plus various ECCS andel improve-in a fuel design to a peak UCR limit of 14.4 seats. Ew/ft (i.e., the CE85 fuel), it is espected that MCR would not limit the power uprate capability. Ifbether N-1 assumption or N-2 assumption (or both) is to be used as the licensing basis C. Bot Racess Reactivity is dependent on the regulatory requirements set forth by the regulators of each individual i The desired minima bot excese reactivity country. The calculated PCT of the required st beginning of cycle used for CE core design is ECG analysis is to be below the legal require-II. This 11 bot excess reactivity is espected ment 2200*F FCT limit of 10CyR50.46. to be achievable with 104.21 power sprate condi-tion with the normal core design and fuel cycle The FCTs for both current ECCS analyses for evaluation. The rod pattern study results show the typical BWR/6 giant assessed in this study that, in general, the uprated power reduces hot are below the 2200 F limit. No MAF M CR restric-escess reactivity by about 0.21 AK. tions have been placed on the CE6 fuel design. D. Summary Discussion The ECCS analysis for jower uprate must be performed at 102% of the uprated power level in i The core operating limit discussion accordance with legal requirement 10CFR50 Appen-presented above concluded that the most limiting diz K. This assessment study results show that thermal limit parameters are found to be MAPMCR the increase in power at the uprated condition and MLHCR 11atts. The rod pattern study per-vill produce a slightly longer period of core formed indicates that the typical BWR/6 plant uncovery following a loss-of-coolant accident has sufficient margin to meet all core thermal (14CA). As a result, the calculated FCT will limits for a 4.2Z power uprate for approximately jacrease for power uprate. Therefore, it may 75% of an early fuel cycle and 50% of the time be necessary to apply slightly reduced MAFLHCR for an equilibrium fuel cycle. To increase the limits in some exposure range to keep the cal-likelihood of meeting these core thermal limits culated PCT below the limit for the CE6 fuel and provide operating margins comparable to that design. The MAP 1JICR mitiplier is a plant of the current power level. CE vould recomend specific value dependent on the asom t of FCT the improved CEBB fuel design to increase the margin esisting currently and whether N-1 or N-2 MCR limit to 14.4 Ew/f t. la a(dition, new assumption is required for the evaluation. N-2 ECCS LOCA licensing methodology' would also be analysis normally results in a higher FCT value. suggested to restore MAFMCR margin with the This typical BWR/6 assessment study showed that increased UCR (see later sectbn en ECCS), a MAFMCR reduction of 11 is expected in the 10,000 to 20,0001stD/ST esposure range for an It is also importane to note the importance N-2 IACA calculation. of power shaping during uprate operation, Figure 3 shows the predicted typical 104.2% power uprate The MAPLHCR restrictions which may result early cycle and equilibriima cycle end-of-cycle from the ECCS power uprate analysis can be optimun Baling power profiles. The modest eliminated by the use of more realistic analysis increase in the bottom peak of the equilibrim assumptions. Use of the ANS 5.1-1979 decay heat cycle optimm Baling power distribution resulted model is espected to result in a calculated FCT in a reduction in uprated power availability fos below 2200'T even if a 20% adder is applied to that cycle in comparison to that of the early the decay power term. Also, General Electric cycle. Since the Reling pouer profile is depen-has obtained USNRC approval for a more rplistic dont upon the power shaping performance in prior set of 14CA licensing methodology models. The cycles, it is imperative that knowledgeable core application of these models in place of the annagement be performed in implementing uprated current models will gypically result in reduc-i l power operation. tions of 600 to 1000 F in the calculated FCr for i f i , - - ~._

'l < - r O the limiting case. For this uprate v.ibration assassment, only balanced flow responses for a few major sensors parthemore, in order to sacreose the were evaluated in detail. A qualitative and likelihood of meeting the LBCR thermal limit to quantitative review of the other sensors' provide full sprate capability, CEug fuel design responses were performed which showed that they with 14.4 Ev/ft is recommended. The higher are not of concern. 14.4Ew/JeLBCRlimitwillresultinsaceeding the 222rF PCT limit with the current ECCS thCA To predict the vibration responses antici-model. These more realistic 1ACA models are pated in the power uprate region of the power essential for meeting the 2200*F FCT limit with flow asp, data on the ME0D rod line and the ao MAP 1JICR restrictica. Therefore, it is con-two-loop anximum flow rate line were esamined. cluded that these realistic 14CA analysis In addition, the vibration aloag the constant assumptions and models coupled with the CEUS 50% core flow rate line were also investigated. fuel design are m yocted to provide full 104.2% All vibration responses are estiasted from power uprate capability for the BWR/6 plants. response spectra using the first level data reduction technique (conservative estimates REACTOR INTERNAL 3 YIBRATION ASSES 5MNT from INE values). Fatigue strength is the primary asterial The vibration responses vers azaained as property of concern for reactor pressure vessel a function of core flow on the MEOD rod line, components subjected to high-cycle alternating as a function of power at amminum core flo'v, stresses resulting from flow-induced vibration. and as a function of various rod lines at WOZ Vibration measurements are made during plant constant flow rate line. Results of the startup testing by means of instrumentation evaluation show that two sensors responded lo'cated on components inside the reactor vessel. maziam vibration amplitude of close to 100Z The purpose of vibration monitoring is to confirm of acceptance criteria at 1001 power at anximum the structural integrity of major components in core flow. All other sensors and conditions the reactor with respect to flow-induced vibra-examined are well below the 100% allowable tion in accordance with requirements of USNRC limits. Regulatory cuide 1.20. This is done by comparing the measured vibration amplitudes (strain or Second level data reduction technique displacement) against a set of acceptance criteria. (filtering) was then used for these two sensors. The acceptance criteria are basically a set of Filters are designed to pass signals in some frequencies and corresponding allowable amplitudes frequency band and reject signals in all others. (for each sensor) derived from an analytical Band pass filters pass signals that fall between model. This ensures that stresses everywhere are two cutoff frequencies. Bence, the vibration below the asterial allowable stress. at a specific mode can be separated from the total vibration. This andal response can then An assessment was performed based on the be campared to the criteria at that mode. The vibration tist data collected during startup present criteria for each mode is then added testing of this typical BWR/6 plant. Reactor by absolute samation. When filtering, care is internal vibration tests were conducted during takan to assure that filter characteristics and preoperational testing, 75% power flow rod line their possible attenuation effects are properly testing, 100% tod line test considered. This is achieved by asking a ded Operating Domain (MOD)pg and Maxianas Exten-testing. Operating spectrum of the filtered signal and superimposing conditions during each test period include steady-it on the spectrum of the unfiltered signal, and operation (514){ low, unbalanced flow, single loop correcting for attenuation affects, if any. state balanced , and transient testo consisting of one pump and two pump trips, and load reject The filtered andal responses for these two from rated flow conditions. sensors ara compared to the first level data reduction evaluation results. The comparisons There are different levels of data reduction show that filtering eliminated substantial con-techniques. In the first level, a conservative servatism. m dal roeponses for these two eatieste is obtained. She percent criteria for sensors are plotted as a function of power. The each ande is conservatively estimated from the data are then entrapolated to 104.21 power. Root Mean Square (RIS) value from a peak spectrum The extrapolation indicates vibration ampli-and all modes are combined by absolute summation. tudes of less than 80% of acceptance criteria If the sum of the percent criteria exceed 100, a using the second level of data reduction. more realistic estimate of the stresses is made by using the second level of data reduction, It is concluded, based on the assessment filt e ring. Again the contribution of the modes results from the typical BWE/6 plant startup are added by absolute summationt Righer levels test data, that the vibration of vessel inter-of data reduction are used only if necessary. nais is not expected to exceed allowable limits ,-.---,-._.)-.-...-.-~.-- - - -.. ~ 1 ^

\\ 6 g i .e at the 104.2% urrate power condition.' to. CDitCLUSION ensemsm The results of this assessment study show I 8' w5 ) that a typical ENR/6 plant is capable of a 51 m. g increase in steam flow (equivalent to a 4.21 increase in core thermal power). This conclusion / is mainly based on the reactor core performance m-capability of the plant. Detailed systes equip- /. ment and Balance-of-Flant impact, as applied to m ac @ each indivianal plant, are to be evaluated on a 'unn plant-by-plant basis for any specific WR/6 m. /suam plant application. While the typical ElfR/6 plant may not be s< able to operate at this uprated power all of tt.e time due to thermal operating limit constraints e a e a e un us with current CE6 fuel design and current ma nab ayaum licensing andel assumptions, it is expected that Fig.,1 Typical BWR/6 NEOD Operating Map the improved CE8B fuel design, coupled with recently IENRC approved more realistic 14CA 8 licensing models and assumptions, can achieve g in-full uprate poser operation during the entire is cyrie. J e< 88 b m in ceder to taplement plant operation at t,,, eso, E at the uprated power level, detailed plant-specific e , g safety engineering licensing analyses and safety r o analysis report updates are necessary to assure I a< j["# ] long-tera safe operation of the plant. Plant / o 6 technical specification ardifications and plant g ( es ca: operation setpoint changes are also required. susta /emaco m. A significant major therani power uprate l' { beyond the 41 level is judged to be technically 5 8' u feasible in the future. This major uprate will e. w probably involve fuel design and plant modifica-tions as well as testing verifications. A 41 f, 4 5 5 in se uprate is fasible now and is reistively easy to implement. This should pave the way and provide verification experience for an ultimate thermal Fig. 2 Typical FWR/6 Fower Uprate MEOD power sprate for EUR owners in the future. Operating Map u 1. T. C. Lee, R. C. Stirn, Ca eral Electric g Opapany. " Plant Operation Performance Improvements of the General Electric (CE) Boiling Water Reactors (BWRs)", presented in the International Nuclear Power Plant Thermal Eydraulits and Operations Topical Meeting on October 22-24, 1984, Taipei, Taiven. es < 2.

3. S. Shiralkar, J. C. M. Anderson, l **<

A. B. Burgess, S. A. Wilson, General Electric Company " Evolution of IACA Analysis at Cameral as. Electric", presented in the International Nuclear Power Plant Thermal Bydraulics and Operations ar. Topical Meeting on October 22-24, 1954 Taipei, Taiwan. es.... una 3 s t e si u e it a ti 31 v am ust Fig. 3 Typical BWR/6 Fower Upra:e ECC Power Distribution Comparison l ? *: a _ :_ :_:__r. 2:._; ~=~ ~ ~ ~ " "

• ~

POWER UPRATE o 0BJECTIVE ~ PROVIDE ADDITIONAL POWER FROM OPERATING PLANT AT MINIMUM COST o ALTERNATIVES-STRETCH TO 105% STEAM FLOW INCREASE BY 10% OR MORE o PROGRAM APPROACH TWO PHASES o FEASIBILI.TY EVALUATION o DETAILED ANALYSES, MODIFICATIONS SCHEDULE o FEASIBILITY -6 MONTHS o DETAILED ANALYSIS - 10 MONTHS o PLANT MODIFICATIONS - DEPENDS UPON POWER LEVEL o BENEFITS EXTRA CAPABILITY AT LOW COST REPLACEMENT ENERGY SAVINGS MARGIN TO SCRAM SETPOINTS CAVD 1 a _a* . yes. ~%" h"* P. C

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FEASIBILITY EVALUATION 0 ESTABLISH UPRATED POWER ASSESS SAFETY AND CAPABILITY REACTOR SYSTEMS TURBINE GENERATOR POWER CONVERSION AND AUXILIARY SYSTEMS DESIGN REVIEW o DETERMINE EFFECT ON SYSTEMS REACTOR INTERNALS PRESSURE RELIEF RESIDUAL HEAT REMOVAL PRESSURE REGULATOR TURBINE GENERATOR POWER CONVERSION AND AUX SYSTEMS 6 o EVALUATE IMPACT ON FUEL OPERATING MARGINS REACTIVITY STABILITY FUEL CYCLE o ACCESS IMPACT WITH OTHER CORE PERFORMANCE IMPROVEMENTS o REPORT CONCLUSIONS RECOMMENDATIONS FOR PHASE 2 REVIEW CAVD 2 m.-_m..,__,..._,,______..

E, DETAILED ANALYSES o HEAT BALANCE AND SETPOINTS O REACTOR FLUID CONDITIONS o TRANSIENT ANALYSES OVERPRESSURE MCPR STABILITY o ACCIDENT EVALUATIONS CONTAINMENT ECCS RADIOLOGICAL = 0 REACTOR INTERNALS DIFFERENTIAL PRESSURES FATIGUE USAGE RADIATION 0 REVISED OPERATING PARAMETERS TECH SPECS POWER FLOW MAP INSTRUMENT SETPOINTS PROCESS COMPUTER DATA BANK 0 LICENSING REPORT CONCLUSIONS NRC MEETING SUPPORT CAVD 3

~' y, _e.c ~ BENEFITS REPLACEMENT ENERGY SAVINGS o o CAPACITY BENEFITS O TOTAL SAVINGS FOR 5% UPRATE $30 - 150M o OTHER CONSIDERATIONS PAYS BACK COST IN FIRST YEAR COST OF ADDED CAPACITY MUCH LESS THAN NEW PLANT /KW CAVD 4 l

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, ~ - L, EQUIPMENT MODIFICATIONS ' (MONTICELLO: 10% UPRATE) O MAIN TURBINE NEW N0ZZLE PLATE, BUCKETS NEW OR MODIFIED 2ND, 3RD STAGE DIAPHRAGMS NEW PACKING RINGS, THRUST BEARINGS ~ LARGER RELIEF VALVE 0 CONDENSATE DEMINERALIZER NEW BEDS o ESTIMATED COST (INCL. DETAILED ANALYSIS) $30M CAVD 5 .. = -.

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}.y l ,a ~,.. ..., +,,. - e,- ..~. ENCLOSURE 3 MEETING OF NOVEMBER 21, 1986, WITH GE - Enclosures Agen'da NUMAC Log Count Rate Meter NUMAC Area Radiation Monitor NUMAC Process Radiation Monitor i NUMAC Log Radiation Monitor NUMAC Rod Worth Minimizer Microprocessor Based Nuclear Power Plant Instrumentation Microprocessor Based Rod Worth Minimizer e s a.-, --.,y,y. .c_ m e a-en

V,.,,' - n. + ~ ~. NRC NUMAC REVIEW 11/21/86 Room 6A, 111 North Market ~ i 8:00 - 8:15 AGENDA REVIEW Dave Reigel 8:15 - 9:00 OVERVIEW OF NUMAC PRODUCT LINE Dave Reigel 9:00 - 9:30 DAIRYLAND/ BIG ROCK POINT FOCUS Dave Reigel 9:30 - 11:00 LAB TOUR AND HANDS ON DEMO Eng. Staff 11:00 - 11:45 GETRAM Bill Rowe Fred Chao 11:45 - 1:00 LUNCH 1:00 - 2:00 DISCUSSIONS / ADDITIONAL ITEMS i All i l .m ,,2,-.M ._4&. g*," g-+--e*e===.=t-m --+---t- . - ~ s

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y ,4.m-3 A g ;.5. S g 5p.g%v.QypStq; \\ 9 . The Prtx:ess Radiation Monitor (PRM), member of the Four trip circuits are provided: two upscale (Hi and GE Nuclear Measurement Analysis and Control Hi-Hi Radiatioa), one downscale (Lo Radiation), and one (NUMAC) family of microcomputer based instruments inop (instrument in calibration, instrument failed, etc.). replCces 194X900 which meets the requirements (less Each trip circuit has two outputs: one logic (to trip the period circuit) of the Source Range Monitor Per-auxiliary units) and one alarm. Since the trip function is j formance Specification 167A2210. The instrument is handled digitally by the microcomputer, there is no trip used with the NA-07 scintillation detector 11781681G001 circuit drift. Trip accuracy depends only on measurement and remote pulse preamplifier 11C2276. accuracy. The input signal-conditioning module is the NUMAC The instrument contains a polarization power supply discriminator which interfaces to the NUMAC micro-for one detec:or. This supply is under microcomputer computer. Discriminator adjustment is accomplished control and is adjustable from the operator panel. Power 3 through the front panel operational control. The supply gross failures are annunciated via the Inop trips. logarithmic count rate function, compensation, calibra. Each detector power supply is provided with a hard-tion, alarm and trip functions are performed by the wired, over-voltage protection circuit. j NUMAC microcomputer. An intemal calibration check facility is provided, and The standard instrument range is seven decades automatic calibration of the instrument is a feature of the from 1E-1 counts per second to 1E+ 6 counts per NUMAC design. second. l GENER AL @ ELECTRIC _ ['l,.,- ...-,,$~- ,. - -,)., 5-, ^~ ~ * ' " ..L- ?? T

\\l......;. .w L,, T',K""'.o ~.j~hy.&L;;;.K' ~~ Each PRM channel contains a voltage source to power its The Inop trip monitors gross failure of the detector power associated detector. supplies, microcomputer " watchdog" timer, card-out-of-file and Voltage Range: 100 VDC -1200 VDC instrument-out-of-operate mode. Self-test detected failures are Maximum Current: 3 mA 100 VDC-333 VDC also reported via the Inop trips.- 1 mA 333 VDC-1200 VDC Maximum Voltage Ripple: 1% RMS Each PRM channel has provisions for driving a recorder Alarm: When +I-10% of setpoint (1.0 volt full scale), process computer input (160 mV full scale), is exceeded and remote meter (1.0 mA full scale). Noise and ripple are less than 0.5% of full scale. Each instrument prendes three independent adjustable trip circuits (Hi-Hi, Hi and Downscale). Trip hysteresis is adjustable from the front panel over the range of 0% to 25%. Logic: 12 VDC,25 mA Alarms: 20 VDC,50 mA a. I d M r k,* d iti Y C ? * : ;

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The instrument will discriminate negative current pulses with amphtudes between 1.6 and 24 mA (as measured at the instrument input). The log integration time constant'(0-63% et step change) will not exceed the limits shown below for a factor of two herease in count rate at the discriminator input. All accuracy measurements are referred to the front panel display which is also the basis for trip adjustments. At design Change le Standard oenter CCidGis, the observed Count rate will not deviate from fg,,7 the true count rate by more than 21% of equivalent linear full y scale over all seven decades. Coincidence counting losses are 0.1 to 0.2 40 10 40 12 excluded from this specification statement. 1.0 to 2.0 40210 18 6 The coinodence counting losses at 1XE + 6 counts per second 10 to 20 35 + 10 1.8 + *4 random input signal are less than 1.2% of equivalent linear full scale for the preamplifier and PRM instrument interconnected 1E+ 2 to 2E+2 13 2 3 0.1R .03 with triple-shielded 75 ohm cable (drawing number 167A2510). 1E+3 to 2E+3 1.02.3 0.018 2.002 Y 1E+4 to 2E + 4 0.1 .03 <.018 At design limit conditions, the observed count rate will not 1E + 5 to 2E + 5 0.01 .003 <.018 deviate from the true count rate by more than 2 3% of I fwgher <.01 <.018 equivalent linear full scale over the top six decades and by more than 2 5% of equivalent linear full scale for the bottom i r decade. Coincidence counting losses are excluded from this i The instrument will produce a count at the log count rate output spooficaton statement. for a negative current pulse equal to or greater than 1.6 mA n when the discriminator is set for maximum senssiivity i L_ L' G I N W W ' 1 1% of equivalent linear full scale for a thirty-day interval after 2701 UN : ' P! F . ' cc - initial warmup of at least two hours. 7 For the combination of the pulse preamplifier and the PRM, a 2.0 mV negative step through a 10pf capacitor to the pulse preamplifier input will produce a count at the log count rate output. (This is equivalent to an instantaneous charge of 0.02 pico-coulombs.) GENERAL $ ELECTRIC

...-... _ -.c'- -~ -. ~.... ... +. (M Area Radiation Monitor (304A3704) y -__., -. y n. -y l l U--

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l l 1 4 / r [ / / / / ~ ' ^ ,2 .w ~~ .- a s The Area Radiation Monitor (ARM), member of the GE The NUMAC microcomputer performs the Nuclear Measurement Analysis and Control (NUMAC) logarithmic count rate function, compensations, 4 family of microcomputer based instruments, replaces up calibration, alarm and trip functions, detector voltage to eight indicator and trip units (12982802). Each of the setting, digital filteri and required mathematical eight detector inputs is polled every 50 ms. The operations. Instrument can be used with ionization chambers, The standard instrument includes two remote (trip) Geiger-Mueller (G-M) tubes or scintillation detectors. outputs per chassis: one for Downscale/Hi radiation trip New auxiliary units are available on request, or the ARM indications and one for self-test detected faults. Each trip can be interfaced to present 237X892 auxiliary unita and output is a 12-volt logic level output. An additional output 194X927 sensors and converters. The system is to the trouble light on the ARM cabinet is also provided. available safety related. It will light on any Downscale/Hi radiation level trip and Two NUMAC four-channel femtoammeter models on any self-test detected fault. Up to three additional trip serve as input amplifiers when used with lonization outputs per chassis can be supplied to satisfy customer chambers in the direct current (dc) mode. An analog-to-requirements. The microcomputer digitally handles the digital converter interfaces amplifier output to the trip function, eliminating trip circuit drift. Trip accuracy NUMAC microcomputer. depends only on measurement accuracy. When used with ionization chambers, G-M tubes or The instrument contains a redundant (auctioned) scintillation detectors in the pulse-counting mode, two polarization power supply. A single voltage, adjustable NUMAC four-channel discriminators which interface from the operator panel, is under microcomputer control. directly to the NUMAC microcomputer act as input The self-test trip output annunciates power supply signal-conditioning modules. Front panel operational failures (voltage output errors and current overioad). controls accomplish discriminator adjustment. Each detector power supply includes a hardwired The maximum instrument range is eight decadas overvoltage protection circuit. The NUMAC design from 10 mR/hr to 10,000 R/hr. Any number of decades features an automatic intemal calibration check facility. up to a maximum of eight within this range can be provided for application with a particular detector. GENERAL $ ELECTRIC

- -w. ynum.;ww. ++.n..e,,.s. -+.;m.?w.m -+m.w.-. .. m,. %%,.l,ww.n. - m..,. &n~ ;m1.v...wm. .w m ~ , : m n:. ,x m. ,qt ...-??W ,.ex.rm m?.. ' .. v.n..: ? -c c l. e fg' ?. w Q '. &. py.p.h.,.p}V[x.~.;.p...I. .'&q}-w g'< ',pl*., =.n p, .~ n' . _. w.. ',:k(*:.y W%..,f: a :.. ~ s 3 ...m._ m_ & i d - P.w c.- Su.v r. '?a-Setf-Test Trip Each of the eight ARM channels contains a voltage The f., elf-test trip monitors gross failures of the detector source to power its associated detector., power supplies, the microcomputer "welchdog" timer, Voltage Range: 100-1250 VDC redundant and the card-out-of-file and instrument-out-of-operate (auctioned) mode Maximum Current:1 mA maximum Self-test failures on the module level are also Voltage Ripple:1% RMS reported. Self-Test Alarm: When 210% of setpoint is exceeded or Other OutEuts cunent overload is detected Each channel (up to eight per chassis) fumishes outputs "rn,p Ocfput to the auxiliary units which include radiation level Each of the eight ARM channels has two independently indication, Wsual and sonic alarms and sonic alann atirantahia trip setpoints (Downscale and Hi). Trip reset. As an opton, each channel can provide &20 mA, hysteresas is adjustable from the front panel over the 22% linear full-scale remote outputs. Extemal supplies range of 0 to 25%. to the ARM electronics cabinet provide the power for the 4-20 mA outputs. ,--.-w n, ----- g- ? -- .pp-a.--- m - g. g ..a=-- um n. ' e, a .2,.,. =.,. m - r,y p; x.. p i. n.. ;;.p a.,. s,.::. x.q.mga,gw%gse wp.i.m...e" w.w.g;m..a-u.....,. . ~., -t n- - c. ...y, a, .ma..c, ; . w. g...% ,. am m. a m ..,,2 y M.,...-: a n 1 n!?;. & - u y. a H n U-n u e t 1W 5 5...J.3'C. '.* W L &u,. & f. g ;;yt %., _ w. a ,,.,.x;. a fgwyN m -_ %...Q W...._ _ p..= N.n ' S?GQ:...., .A i A _.m O o. ,y -.,g...u.. ; xm ww _ ~ -. -.w Input Signci Discriminator and Window Rangos From either current-type (ion chamber) or pulse-type For G-M detectors (G-M tube) detectors Discriminator: Adjustable from 25 mV-2.5 V Char:nci ncceracV window: Adjustable from 10 mV-2.5 V Forion chamber detectors Settings are digital and do not drift. Design center conditions are tabulated below: Response Times d$ Digitally filtered as a function of either count rate or w cunent M. 1E-13 to 1E-11 1.2 25 1E-11 2 0.76 15 All other decades 2 0.52 210 O e GENERAL $ ELECTRIC

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n, The Log Count Rate Meter (LCRM), member of the GE The range of the instrument is 10 to 1E6 counts / Nucl:ar Measurement Analysis and Control (NUMAC) minute. t:mily of microcomputer based instruments, replaces up Each of the three LCRM channels contains three to three existing BWR Log Count Rate Meters Form-C contact trip outputs. Since the trip function is (145C3284AA). handled digitally by the NUMAC microcomputer, there is Each of the three NUMAC Log Count Rate Meter no trip circuit drift. Trip accuracy depends only on channels receives input voltage pulses from the existing measurement accuracy, associated pulse preamplifier. These pulses are routed to the NUMAC input signal-coiGdoriing module the The instrument contains polarization power supplies channel discriminator which interfaces with the for three detectors. These supplies are adjustable from Instruments functional microcomputer. The logarithmic the operator panel and are under microcomputer control. count rate function, compensations, calibration, alarm Power supply gross failures are annunciated via the inop and trip functions, detector voltage settings and required trips. Each detector power supply is provided with a rnith matical operations are made by this computer. hard-wired, over-voltage prctection circuit. Discriminator adjustments and calibrations are made An intemal calibration check facility is provided, and through the front panelinterface. automatic calibration of the instrument is a feature of the NUMAC design. i 1 i l GENER AL h ELECTRIC

y }- ,w-e c. +.,,-,,.ps *.,:ny.:- l., e w.,n ..,.. n Each nf the three LCRM channels contains a voltage source to The Inop trip monitors gross failure of the detector power power its associated de%ctor. supplies, microcomputer " watchdog" timer, card-out-of-file and Voltage Range 450 VDC - 1200 VDC instrument out-of-operate mode. ] Maximum Current: 1 mA maximum Self-test detected failums are also reported via the Inop Voltage Ripple: 1% RMS trips. The Inop trip outputs apoear in parallel with each of the Alarm: When +/- 10% of setpoint three Hi-Hi trip outputs. is exceeded Each of the three Log Count Rate Meter channels has pro-Each of the three Log Count Rate Meters provide three inde-veons for driving a recorder (1.0 V full scale), process pondently adjustable trip circuits (Hi-Hi, Hi and Downscale). computer input (160 mV full scale), and remote meter (1.0 mA Trip hysteresis is adjustable from the front panel over the range full scale). Noise and ripple are less than 0.5% of full scale. of 0% to 25%. .J. _. s. Posstive, double differentiated pulses,25 mV to 4 V,

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Window: Adjustable from 10 mV-2.5 V Settings are digital and do not drift. Response time constants wry with count rate as follows:

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l 1 The Main Steam Une or Log Radiation Monitor (LRM), &ci,oiTipiished by firmware change and are available as I member of GEh Nuclear Measurement Analysis and options. Control (NUMAC) family of microcomputer based instru-Trip outputs are configured to be exact replacements ments, is a direct replacement for existing GE LRMs for the existing instruments. The standard instrument (194X629,238X660 or 368X428AA). includes four trips. Each trip output includes a 12-volt Economic analysis of conversion to the NUMAC logic level output. Additional trip output modules (up to LRM indicates that savings in 1) instrument technician su total) can be provided to satisfy customer require-calibration time,2) scram avoidance and 3) spare parts ments and will be quoted on request. The first trip output inventory reduction can achieve a payback period.o' used is reserved for the chassis Inop function. Since the three years. trip function is handled digitally by the microcomputer, The input amplifier is the standard NUMAC femto, there is no trip circuit drift. Trip accuracy depends only on ammeter input module. The amplifier has an input span measurement accuracy. from 10 femtoamperes to 0.3 milli-amperes. The ampli ' The' instrument contains a polarization power supply e l fier output is interfaced to the NUMAC rnicrocomputer by for one detector. This supply is adjustable from the oper-an Enalog-to-digital converter. ator panel and is under microcomputer control. Power. Temperature compensation, calibration and supply gross failures are annunciated via the Inop trips. mathematical operations are accomplished by the Each detector power supply is provided with a hard-l Instiament microcomputer. wired, over-voltage protection circuit. The standard instrument range is six decades in the An intemal calibration check facility is provided, and range from 3.33XE-13 to 3.33XE-7 ampere for the Main automatic calibration of the instrument is a feature of the Steam Line Mon;;or. Altemate instrument ranges are NUMAC design. GENERAL $ ELECTRIC Ll

,,,. _... a .-. n ~ e.; 3, - - - n.,.,. ~ 3.. Each LRM channel contains a voltage source to power its The Inop trip monitors gross failure of the detector power associated detector (5467870G016). supplies, microcomputer " watchdog" timer, card-out-of-file and Voltage Range: 100 VDC - 350 VDC instrument-out-of-operate mode. Self-test detected failures are Maximum Current: 3 mA also reported via the Inop trips.. Maximum Voltage Fbpple 1% RMS 4 ;' ' T I' P '" . Over-Range Umit: 450 VDC Alarm: Each LRM channel has provisions for ng driving a recorder When +/- 10% of setpointis (1.0 V or 10.0 V full scale) and process computer input (160 - exceeded mV full scale) Noise and ripple are less than 0.5% of full scale. Each LRM provides three independent adjustable trip circuits (Hi-Hi, Hi and Downscale). Trip hysteresis is adjustable from the front panel over the range of 0% to 25%. n. -n,..., i ~ m m l *.. - '.... ?. ; ' ' '.. ' *.l kE' i .aL- ^ iiC C U.'D C y N G :. Positive susser % et. Feu had cut (Aspen) Puha seats ' g,,,, q 3.33E 3mE-12 25 21.6 time constants wry 2 input et M as follows 3.33E 3mE-10 z15 1.0 chases m Time 3mE 3.33E-7 10 20.7 Cemet Level (Amesm). Cemeteet Fies It (ses) 3.33E-12 6 3.33E-13 %ueser 3.33E-12 3.33E-11 1 % el Fue hystcomme(Ameen) runs seeie 3.33E-11 3.33E 10 0.5 31.3E 13 - 3.33E-10 2 50 22.9 3.33E-10 3mE-9 0.5 3.33E 3mE-7 z20 m 1.3 3.33E-9 3.33E-8 0.5 3.33E-8 3.33E 7 0.5 3.33E-7 1 3.33E-13 1 l o GENERAL ELECTRIC

Rott Worth Minimizer (304A3702) I Wll.J.$- ,D 2 $ ;%.l ~ t$ l Q 5 s 3/ pr ... -Q-Q - .- *~ j ' ' W4?[ g' ..z. ..x W b y qt. o ~ f* l. F w i . ;~ it - c.v mig b m W.- .. ~ _ n > }[sfyk < !g aw

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h, h$ ,5 kN & h.. M.yW $& y, o3y.v.y., : .v 1 .y ~. .m-n. The NUMAC rod worth minimizer (RWM) is a member of the dependent. The BPWS logic is desenbed in document Nuclear Measurement Analysis and Control (NUMAC) family of NEDO-21231. An option is available which allows the sequence trucroprocessor-based instruments currently being offered for a to be loaded, then disp!ay the sequence to an operator as wriety of applicahons in nuclear power plants. The NUMAC requested. RWM replaces the present BWR rod worth minimizer (RWM) The NUMAC RWM has the following system implemented on stand-alone GEPAC computers or as part of characteristics: GEPAC 4010 or 4020 computers or Mcs,q;;;:: 4400/4500 m Four sequences with operator-selectable names are process computers on most BWRs. It also replaces the available. hardwired group-notch Rod Sequence Control System (RSCS) a Selecuon of normal / bypass mode. For a dual-channel and associated RWMs used on most of the BWR/4s and /5s in system, only one channel is bypassable. the U.S. m Substitute rod position capability. The NUMAC RWM function rnonitors and enforces a Selection of shutdown margin test mode. adherence to established low power level control rod insert and a Rod coordinate indication of insert / withdrawal errors. withdramel sequences. This function prevents the operator from a Status indication of select, withdraw and insert blocks. _ 7#aig control rod pattoms that are not consistent with a Keylock bypass of up to eight inoperable control rods tie predesenbod sequence by initiating appropriate rod select (controlled by Tech Specs 4 i error werrung, rod withdramel block and rod insert block signals. m NUMAC RWM failure indication. The instrument enforces control rod sequencing procedures a Display of control rod identification and position. designed to limit (and thereby minimize) individual control rod u Identification of rod group from which rod selection is made. worths to acceptable levels as determined by the rod drop u Error messages. accident design basis. Its function does not interfere with The NUMAC RWM is designed to interface with the data normal reactor operation, and in the event of a failure, does not acquisition equipment at existing plants. The following input itself cause rod pattems to be established which wnuld violate interface modules are available: the rod drop accident design basis, a Module for inputs from an existing relay multiplexer used to The generic NUMAC RWM functional diagram is shown in provide position data to the instrument. the accompanying illustration. It uses the Banked Position a Module for inputs from the serial encoded words used in Withdrawal Sequence (BPWS) algorithm which is based on RSCS. stored Boolean logic in read-only firmware and is not cycle e Module with a fiber optic serial input from a replacement RPIS. GENER AL @ ELECTHIC

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~ - d -'- -- Each NUMAC RWM is constructed of standard plug-in modules The dual-channel NUMAC RWM is required for all plants in a chassis designed for rack mounting. All components are with an RSCS and is an option for single-system RWM plants. mounted in funcbonal modules. Maintenance is accomplished The dual design supplies direct rod position data and rod by module replacement. motion information to either NUMAC RWM channel. Each The universal chassis is 16 inches deep x 19 inches wide channelis powered by separate, divisions and electrically ] x 7 inches high. Chassis slides provide access to the top-isolated from the posrtion information source to minimize the i loaded printed circuit modules for maintenance with the possibility of a failure in one channel causing an interruption of ) instrument in the extended position. Instrument interfacing the RPIS signal to the other channel. j connections are mounted on the top rear sechon of the chassis deck. A complement of mating connections and a retract 8 SM @Ff!C fW' E mechanism for mounting the interfacing cables to the slide-mounted chassis are provided in a package of accessory parts, For safety-related applications, the NUMAC RWM is aveilable i if re@ ired. qualified to the requirements of IEEE 323,1974 and IEEE-344, The instrument is microcomputer based. This permits the. 1975. Other options include operation and maintenance capability for auto-ranging, automatic calibration and on-line training, spare parts and a process computer link wherein one computation of operational and maintenance information. All of the RWMh two serial data links can be used to provide data logical functions, arithmetical functions or control algorithms transmission to and from the plant computer. A next rod are impleri,tnted by the microcomputer module. prompting option will provide rod motion prompting to the The instrument design includes a self-test capability which operator as prescribed by either utility-loaded startup and identifies and annunciates failed modules, if any, to a shutdown sequences or by the Banked Position Withdrawal replaceable module level permitting repair within a one-hour Sequence. period. The primary function of the self-test feature is to Additional options are available such as extended maximize instrument channel availability. With the self-test application engineering and site documentation updating, as feature, instrument availability is designed to be greater than well as options fcr power descent and data logging and 0.9997 for an estimated mean time to repair of 30 minutes. reporting. Without self-test, assuming a weekly surveillance interval. An optional " inoperative" trip signal can be provided instrument availability is designed to be greater than 0.989. The which monitors the instrument for a gross failure of the self-test feature verifies instrument repair and minimizes microcomputer, card out of file and the instrument not being in maintenance time. the operational mode. For a dual-channel system, the NUMAC = = --m--g RWM can provide cross-channel checks using fiber optic

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===a-- position data. ~,:ll - Desamma fuses amas am, anse aamama .f .s .r . +.. -.4. %,gg L' ~ mas,suse Ceem .e j:lllll,_ 'allllll" + - ~.. Single-ChanneYNUMAC Rod % tth Minimizer Most RWM modules are common to other NUMAC instruments. This leads to ready availability of spare modules 1..---,. art the potential for substantial spares inventory reductions. '"~ ^ " - ~ The standard NUMAC RWM instrument has a menu-C l.. - 4 driven electroluminescent display. The display provides ' ~ ~ ' ' ' ~ - ' " ~ " ' operationaf and maintenance interface data display information along with operational switches necessary to operate and "*I*'*I" 'h* I"****" TypicalNUMAC RWM Displays The instrument uses standard AC power,120 Vac, 50/60Hz. Other available options are 220 Vac,125Vde,24Vdc and 20vdc. Future Upgrados It la designed for normal operation in the range of 5'C and 10% to 90% relative humidity noncondensing. It meets GE The NUMAC RWM is designed to be compatible with future requiremena for EMI-RFI per GE's EMI susceptibility test upgrade products such as improved rod position / status display guide. Dual instrument power supplies are provided to enhance and hardware and analysis to permit relaxation of Banked availability. Position Withdrawal requirements. GENERAL ELECTRIC

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+. s q3 i 1 4 o f 6.'%/. y g ( Microprocessor Based Nucleaf Posler Plant Instrumentation t t, i L! s. f

• Ease of calibration, test and repair, INTRODUCTION in the past decade, microprocessord and associated in-
• Qualifiable designs for safety related applications, tegrated circuits have revolution zed the design of elec-tronic instrumentation. Irr. proved performance, At General Electric's Nuclear Energy Business Operations, cepability and reliability can be achieved whik at.the same time reducing size and weight, power consu'.iption, and the pursuit of the above objectives has lead to the cre-t ation of NUMAC, a new, environmentally qualified m:intenance effort. In the design ofJustrumentation for family of products for the measurement and analysis of both old and new nuclear power p6nes the incorpora-radianon and process signals, the interchange of data with tion of microprocessors appests a' tractive.

host computers, and the imtiation of control actions based on these measurements and data exchanges. The(eplacement of existing instrumentation is ually motiv' ted by one or more of the following: improved per-a f:rmance, be.t'er user 0nterface, wearout of installed in-HARDWARE DESIGN struments, and additional requirements. Whatever the reasons, present day digital technology can provide new Mechanical Configuration capibilities ana benefits. Since modern electronics is a rapidly evolving field,'t xlay's technology is several gener- ^ ". A instmments utilize a common chass.. is designed for slide-mounting m a standard 19' instrument ations beyond that which was available at the time rrost rack. They occupy 7" of rack space and vary between 16' original plant equipment was'd: signed. i to 20" in depth. A typical NUMAC instrument, the Log Count Rate Meter, is shown in Figure 1. Thr paper presents a set of design objectives for microprocessor based instruments and discusses Gemr-al E!ettric's response, the NUMAC (Nuclear Measure-mint Analysis and Control) family of instruments. e DESIGN OBJECTIVES goe design of instrumentation for retrofit applications R 4! fords many opportunities for providing added value [ r,nj performance. In order to best obtain these benefits, a set g design objectives should first be formulated. m These objectives might vary from plant to plant, but ""m , woulgerts aly include many of the following: j - Jj Mitnility for both direct mechanical and functional, ,k i j(, -replacement of old instruments and new applications.1 D l ' / = Modularity of design for ease of configuration and l + adaptability to changes. l Figure 1. NUMAC Instrument

• Commonality of coniponent modules and instrumer:t (Log Count Rate Meter) operation to simplify training, maintenance, spares provisioning.

The standard chassis contains a card file and mother

• Ability to commti.tieake, with other instruraents anf board capable of accommodating a complement of up to 15 pnnted circuit modules; a front panel with an elec-f.

'I5' '* ' ' ,' p 7 troluminescent display, functional softkeys, cursor keys, a Hiht performance, reliability and availabi'ity, data entry keys, and a key-lock switch; and redundant t J \\'I e s g f'

f

1 A 1. j DIGITAL F/O F/O DIGITAL INPUTS INTERFACE INTERFACE OUTPUTS ANALOG CONDITION b_.

• CONVERT AfD D/A ANALOG INPUTS CONVERT CONDITION OUTPUTS i

i i I f///// DATA BUS V/////g PULSE COUNTER / HV PWR ~ INPUTS DISCRIM. SUPPLIES i i I I CONTACT CONDITION INPUTS CONDITION CONTACT FUNCTIONAL OUTPUTS i CPU / MEMORY i W/SELF-TEST SERIAh LINK REDUNDANT 120 VAC POWER - SUPPLIES i DISPLAY CPU / MEMORY I' f M

• * * * < Eli].

.~ Figure 2. NUMAC Block Diagram power supplies. Instrument application will determine the tions as well as automatic self-test and calibration. They number and mix of modules provided. are based on the use of an 80C86 microprocessor (16 bit) and an associated data bus. Most power and signal input / output connections are made at a connector bracket mounted at the rear of the included in each instrument is a functional computer chass,s. Its design varies with application. For added reh,- i module containing the microprocessor, random access ability, very sensitive signals bypass this bracket and con-memory (RAM), a program installed in read only nect directly to appropriate modules. memory (ROM;, and electrically alterable read only memory (LAR+ M). It directs instrument operation and Electrical Configuration per'& ad. nctions as temperature and linearitycor- [ A NUMAC instrument's electrical circuits are divided regio" w/ r, 2: ared data, parameter and control calcu. into three sections: the functional microprocessor and its lations, trip comparisons and event sequencing. l related modules which perform the instrument's prime Additionally, it controls self-test and evaluates results, i functions; the front panel and its display controller to directs autocalibration, and communicates with the front

handle user interface; and the power supplies which pro-panel controller via a serial data interface.

vide the operating voltages for b-th the above. A basic block diagram illustrating this arrangement is provided Process input signals are received by appropriate signal in Figure 2. conditioning modules. Analog input modules is'olate sen-Functional Circuits sor and transmitter signals and convert them to digital form, contact input modules sense switch and relay sig-Functional circuits are those which perform the instru-nals, ahd pulse input modules perform pulse height dis-r l ment's primary measurement, analysis and control func-crimination and counting. 2

u... ~.. Process output modules convert digital signals to forms usable by devices being driven by the NUMAC instru-7 g -.. g p .-.- ggy ment. They include analog output modules which per-lipunune - 4, h,Z '[ O M ^ "i form digital to analog conversion and signal buffering, 7 .i and contact output modules which provide relay or' solid state contact closures. ~i;: M ,oo s s.

g,m..

. y..< g. e.- High voltage power supply modules provide the voltages needed to polarize any associated radiation detectors. These supplies have their own regulating circuits whose setpoints are controlled by the computer and whose out-g-'",*,,,,,,,_,L[ i, put levels are monitored by the self test system. t,', ww - ".." y External data communications (e.g., high speed serial D' '" k data messages) are handled by interface modules cus-O y=='. ;--=== 47_, p ] ~ tomized for signal type, speed, level and protocol com-patibility. High speed fiber optic as well as RS-232 interfaces can be accommodated. Figure 3. TypicalNUMACDisplays User Interface i User interface is through a front panel containing key-board and display units, and a keylock switch to set the Room is provided in each chassis for two redundant low instrument in either the Operate or Inoperative (main. voltage power supply modules. The specific modules used ' tenance) mode. The panel is the same for all NUMAC will depend on the power source specified (120VAC, j instruments and is customized for different applications 220VAC,125VDC, etc.) by the installation of appropriate software. Environmental Qualification An electroluminescent screen with a resolution of 512 x NUMAC instruments find use in both safety-related and 128 addressable pixels provides both alphanumeric and .non-safety-related applications, but a re all built in the same graphic displays, it is driven by a display control mod-manner using the same components. Extensive qualifica-J ule whose memory contains the logic and formats re-tion testing has been performed to date. Additionally, all l quired for displaying data provided b,y the functional safety-related modules undergo testing to their individual computer. The module is mounted in the chassis cardfile performance specifications. As a result, all safety-related but does not interface with the functional microproces-instruments delivered have been qualified to their specified sor it contains an NSC800 microprocessor, RAM, ROM, control room environments in accordance with IEEE and EAROM. Standards 323-1974 and 344-1975. The environmental and seismic profiles used in these tests envelop all known Four "softkeys" below the display are used to control power reactor applications the instrument's operation. The specific functions as-sociated with these keys vary under software control. In SOITWARE DESIGN addition to the softkeys, four keys for controlling the dis-play's cursor and a keypad for numeric data entry are A NUM AC instrument includes modularly designed soft-provided. ware (firmware) for both its functional computer and its display controller. Such design facilitates the preparation in the " Operate" mode, the computer sends all neces-and validation of codes for each specific instrument as sary data to the control module on a periodic basis. Com-well as the addition or deletion of functions for unique munication in the oppo.ite direction is minimized so that applications. controller failures have little or no effect on operation. With the switch in the " Inoperative" position the instru-An event-driven, real-time, multitasking operating sys-ment can be calibrated, a more thorough self-test can be tem kernal is common to all functional computer mod-performed, and, under password control, the more criti-ules in the NUMAC product line. Its use allows NUMAC cal operating parameters can be changed. software to be partitioned into various functional tasks which execute in pseudo-concurrent fashion. The oper-Displays may contain status highlights, data tabulations, ating system controls the resources of the functional com- ) graphs and operating instructions. Typical front panel dis-puter including the microprocessor and certain memory plays are shown in Figure 3. pools. Typical tasks running under its control include i l

t th:se for process I/O and computations, display con.

2. Logarithmic Count Rate Afeter (LCRAf) - Log troller interface, communication with external computers, Count Rate Meters are one to three channel instru-self-calibration, and self-test. This software is written in ments (configurable from the front panel) that dis-PL/M-86 and ASM-86 languages, developed on a VAX criminate and count output pulses from radiation 11-780, and tested using intel in-circuit emulators.

detectors over 5 to 7 decades in an overall range of 0.1 to IE6 cps. They also provide independent polariz-The software design used for the Display Controller uti. ing voltages for associated detectors. lizes an executive loop which calls on lower level proce-dures to perform its major functions. These procedures

3. Rod Worth Afinimizer (RWAf) - Rod Worth Minimizers monitor control rod movements and en-melude an executive, a functional computer interface, a display handler, and an instrument status keeper. This force control rod insertion withdrawal sequences when software is written in both PASCAL and 280 assembly a reactor is below its low power set point. They ob-language, and developed on an HP 64000 Logic Develop-tain sequences from and return status to plant com-ment System.

puters, provide rod blocks when a sequence is violated, interface with a remote operator's display, "" F MAINTENANCE trol benchboard. Maintenance of a NUMAC Instrument is simplified through use of several built-in features. They include: To be Delivered in 1986

• Self-Testing - A built-in self-test system functions

1. Neutron Afonitoring System (NAfS) - Two instru-continuously while the instrument is operating on line.

ments make up this version on an out-of-core neutron This testing is done via the functional computer by per-monitoring system. A Source Range Monitor dis- ~ forming RAM, ROM and internal microprocess'or criminates and counts pulses from a proportional checks, by interrogation of special data resisters counter, and calculates count rate and reactor peri-on the various circuit modules, and through voltage od. A DC Wide Range Monitor measures current measurements made by the analog modules. Faults are from a compensated ion chamber over a ten decade traced internally to the replaceable module level and span. The top three decades are measured linearly announced via trip output circuits. A user can inter-while the remaining seven (plus overlap) are handled rogate the instrument via the front panel to determine logarithmically. Power level and margin are cal-status. culated.

• Self-Calibration - Each instrument contains appro-

2. Process Radiation Afonitor (FRAf) - The Process priate reference standards (e.g., precision resistors and Rad Monitor is similar to a single channel Log Count voltgge sources) that allow self-calibration. Under Rate Meter except for the input pulse rate rartge, the direction of the functional processor these standards number of trip output contacts, and contact ratings.

may be verified against external ones.

3. Hydrogen Water Chemistry (HWC) Alain Control-

• Help messages - Throughout operation and main-This instrument controls both the injection of hydro-tenance, the user is guided by explanatory and diag-gen into the recirculation flow to prevent stress cor-nostic messages on the front panel display. If further rosion cracking of pipes due to oxygen enriched water assistance is required, a system of explanatory " Help" and the addition of oxygen elsewhere to allow burn-messages can be called on at any time.

ing of excess hydrogen in the recombiner. NUMAC INSTRUMENTS

4. TIP Control Unit (TCU) - This instrument controls the msertion and withdrawal of a flux probing moni-The power and versatility of microprocessor technology tor within the reactor core. Position and flux data are in the design of instrumentation for nuclear power reac-sent to an external x-y recorder. Discrete control sig-tor application is demonstrated by the list of instruments nals to the TIP drive, indexer and ball valve are gener-that are either currently under active design or have al-ated. Data is also sent to a host computer. Probe ready been delivered to the field.

movement is controlled from the front panel. Instruments Delivered

5. Automated TIP Control Unit (A TCU) - This inst ru-ment is the same as the TCU except that its operation

1. Logarithmic Radiation Afonitor (LRAf) - Log Rad is controlled by an external computer.

Monitors measure the de current from gamma sensi-tive ion chambers, typically for 6 to 7 decades in an overall range of 3E-13 to 3E 4 amperes, and provide polarizing voltages for associated detectors. 1 'S

Proposed Instruments

3. Area Radiation Afonitor (AR31) - The Area Rad M nit r is an eight-channel instrument that can be I here are currently several other applications of NUM AC technology under investigation. They include:

C nfigured to accept a combination of de(picoampereL pulse and mean square analoginputs from various rad,- i ation detectors, to furnish polarizing voltages for these i.1/SV IVide Range Afonitor - This instrument com-detectors,and to providetrip alarmand recorder /com-bines th'e functions of a source range neutron monitor Puter outputs. (pulse counter) with those of an intermediate range

4. Compensated Waterlevellndicator-Thisindicatoris monitor based on mean square voltage technology. The a safety-related instrument that determines reactor ves-resulting Il-decade monitor covers a neutron range sel water level by utilizing both vessel pressure and from startup to ten percent power with a single incore or reference leg fluid temperatures to compensate the dif-out-of< ore detector. Options include continuous rang-ferential pressure between two reactor water level ing rather than manual and the addition of a linear instrument taps. Flashpoint indication and remote power range.

meter and recorder outputs are provided.

2. Xeutron Afonitoring System - This incore neutron

5. Temperature Afonitor - This instrument, used in monitoring system consists of a MSV wide range moni-safety-related applications, monitors the outputs of tor and an appropriate number of 24-channel power thermocouples and RTDs, detects open thermocouples, range monitors for each reactor safety division. The performs compensations and other calculations (c. g.,

power range monitors measure currents from local taking temperature differences), provides trip outputs power range monitoring neutron detectors and provide when user-set limits are exceeded and transmits results the Average Power Range Monitoring function. to external meters, recorders and computers. 1 l l l l / h

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y, a =. i A Microprocessor-Based Rod Worth Minimizer B Introduction Control rods are withdrawn from the reactor core in a in order to prevent undesirable core flux pattems from predetermined sequence of steps. For each step, a developing during rod motion, General Electric has gr up of rods and their insert and withdraw limits are Identified. The reactor operator selects and drives always. teorporated a Rod Worth Minimizer (RWM) function in its boiling water reactors (BWRs), either in these rods to their withdraw limits. When all rods in a the forrt, Of a freestanding computer or as part of a lar-current step are withdrawn to their limits, the operator ger process computer system. Maintenance of these proceeds with the next step. This process continues RWMs is being increasingly hindered by the reluctance until the Low Power Set Point (LPSP) (about 20% of of computer vendors to support old equipment designs. rated power) is reached. Above the LPSP, no single Hence, utilities. have begun to replat,e their near-rod's worth is great enough to cause fuel damage if it obsolete computers with new ones, thereby forcing a were to drop out of the reactor core. reimplementation of the RWM function. If the reactor operator does not properly follow the rod Based on reliability, efficiency and economic consid-motion secuence, the Rod Worth Minimizer will remove erations, General Electric has designed a new Rod the rod motion permissive signals which it sends to the Worth IVlinimizer as part of the company's line of Reactor Manual Control System (RMCS) thereby microcomputer-based Nuclear Measurement Analysis blocking the operator from any further motion of the end Control (NUMAC") instrumentation. This RWM selected rod. For example, if the operator selects a controlgodwhich,snotamemberof thegroupof rods i can work with new, concurrently installed process computers while using existing rod position indicating to be mov,ed in the current step, both insert and with-and control systems. In addition to providing a direct draw permissives are removed, and the operator is r;plicement for existing RWM functions, the new blocked from moving the selected rod. Once a rod instrument has incorporated capabilities not previous-reachesjts step limit, the permissive for that direction of ly found, rod moton,s removed. i g This paper will discuss the functions of a Rod Worth Though rod motion sequences define the movements Minimizer, describe the configuration of a new RWM of all rods from their fully-inserted to the,r fully-w,th-i i system as well as the NUMAC instrument hardware drawn positions, the operator need not follow them exacW once power ems N N b mat case, and software designs, and present the rpsults of the instrument's first application in an operating plant. the RWM serves only in an advisory capacity and allows for "out of sequence rod motions. Above the LPSP, the reactor operator may change control rod positions, N, adjust power level, or modify the power distribution in g SW Me the core without incurring rod blocks. K Wh;n moving control rods in a BWR during low-power operation it is important to avoid control rod configura-When the reactor operator brings the plant down in tions which would give a single control rod a high power, rod worth minimization again becomes impor-re ctivity worth (ability to affect the power distribution tant. Prior to reaching the LPSP, the operator must within the reactor core). This rod worth minimization is configure the control rod pattem to match some step of int;nded to mitigate the effect of a postulated control the prescribed rod insertion sequence. Once the LPSP rod drop accident. Rod Worth Minimizer instruments has been reached, the Rod Worth Minimizer again are designed to enforce predetermined sequences of enforces the rod motion sequence and any attempt to it control rod motion in order to minimize the worth of move control rods out of sequence will result in the cny one control rod in the reactor core at all times. removal of rod motion permissives. i ..y

System Configurction The configuration of a given RWM system will depend sequences are loaded into the RWM, the instrument on both existing and replacement hardware, on fea-can perform its sequence enforcement function with-tures and options desired and on other plant-specific out the aid of the process computer. matters. Described below is just one of many possible configurations. Control rod position and other rod status information is obtained from th'e RPIS and the RMCS by a data The major elements composing the RWM system are acquisition system which may either be incorporated in shown in Figure 1. The system includes the RWM the RWM or be part of a separate device. instrument chassis (located in a backrow panel), the RWM operator's chassis (located on the reactor opera-When an operator selects a rod for either insertion or tor's main control console) and portions of the plant's withdrawal, the RWM will perform an evaluation based process computer data acquisition system, Rod Posi-on the current rod motion sequence, the selected rod's tion and Information System (RPIS) and Reactor Man-identificationandposition,andthepositionof allother ual Control System (RMCS). rods in the core. lf it is determined that the selected rod may be moved, a permissive signal is sent to the RMCS. Control rod motion sequences which assure rod worth if not, the RWM will block the rod's motion by removing minimization are normally developed and modified on the permissive and will send an annunciation to the the process computer and validated by computer operator's console. cnecking against various sequence constraints. Vali-dated sequences may then be downloaded from the The RWM has four basic modes of operation, OPER-computer to the RWM instrument chassis. All commun-ATE, BYPASS, TEST, and INOR When the instrument ication between the RWM and the process com puter is is in the OPERATE mode, rod position and status performed via a data formatter (i.e., a buffer device) and information is received and processed continuously. sent over fiber-optic serial communication lines. Once The instrument performs a data integrity check to test bATA Y E DATA / / / FORMATTER f f rod POSITIONS RPl$ } Q PROCESS y- _/ COMPUTER Roc- /1 STAW8 f I 0 WORTH MINIMIZER / 4 c7 o o o o b, , / E. <C 3 o mer RWM OPERATOR'S CHASS48 $f / = = -ce-i i ecco aceo M o oga EMI illi oa o a o o o o o nX ./F F 5 9 9 9 9 9 4 %, -O-FIBER OFTIC CA8LE

• MuLTS CONDUCTOR CABLE COAX 1AL CABLE REACTOR OPERATOR'S CONSOLE Figure 1. F1WM System Configuration 2

the validity of all input data and other intemal data. identifies the minimum number of rods which Sequence enforcement calculations a's then performed must be moved to get the control rod pattem to determine if rod motion pertr%sives need to be into sequence (for any step). r: moved and if annunciation is required. The opera-tofs display provides a variety of information such as (4) Next Rod Prom,oting. The RWM identifies current RWM mode of operation, sequence identifica-the next rod to be moved, its current position tion, step in the sequence, power range, self-test status, and its limits of motion. status of the permissive outputs. selected rod and its redt:en and more. RWM status information is continu-(5) Rod Scram Timing-The positions of all rods cusly transmitted to the process computer for logging. as a function of time are recorded imme-diately following a scram event. The reactor operator may bypass the RWM with the y k;ylock switch on the operators chassis. While in the (6) Confirmation of Shutdown - The RWM con-BYPASS mode the RWM will continue to perform all firms that, af ter a shutdown, all rods have been sequence enforcement calculations as in the OPER-inserted past predetermined limits. ATE mode, and the operator will still be provided the same RWM status information. However, the output to The RWM system configuration described above may, the RMCS will be locked in the permissive state, and no of course, be modified to meet the applicati,on require-(nnunciations to the operators console will be made. ments of specific power plants. For exampl3, the RWM can bemadeindependentof theprocesscomputerby The keylock switch on the operators chassis may also having its interface with a microcomputer instead. The be used to place the RWM in the TEST mode. While in data acquisition function can be incorporated directly th) TEST mode the RWM can enforte either a shut-into the RWM through the addition of appropriate down r.largin test sequence or a single rod test. The hardware modules. shutdown margin test sequence identifies two control rods which may be withdrawn from the reactor core whil] all other rods remain fully inserted. During the Hardware Design test, the RWM ensures that only the two identified rods ara withdrawn from their fully-inserted positions and The Rod Worth Minimizer is a member of General that they are not withdrawn past their assigned limits. Electric's NUMAC (Nuclear Measurement Analysis The single rod test is used to evaluate the motion of a and Control) line of instrumentation. NUMAC products giv:n control rod while all other rods are fully inserted. can replace obsolete or near-obsolete nuclear instru-During this test, the RWM ensures that only the ments with an integrated product line offering both selected rod is moved. better performance and simplicity of installation. 1 NUMAC products benefit from a combination of mod-ll A keylock switch on the instrument's chassis may be ern technology and modular design which allows for used to place the RWM in its INOP mode. When in this improved reliability, reduced maintenance and surveil-I mode, the permissive signal to the RMCS is removed, lance costs and reduced spares inventory (Ref.1]. h and an annunciation is made at the operators console. They also utilize a self-test function which continu-3 Tha INOP mode is used to perform sequence down-ously checks hardware integrity and alerts operators to y loads from the process comnuter place individuat rods any self-test detected failures. This reduction in time j in bypass (remove from the sequence enforcement needed to detact hardware failures greatly increases d logic) and perform more exhaustive self-tests. instrument availability. f r l The RWM design also provides for the following The NUMAC RWM consists of both an operators h optional features: chassis and an instrument chassis. The operators ( chassis (shown in Figure 2) is a 10" x 8.4" x 10.4" 4 (1) Bypass of ControIRods - Under certain con-enclosure mounted on the reactor operators main con-ditions, up to eight control rods may be removed trol console. it contains a 7.3" x 1.7" electroluminescent from the sequence enforcement logic. This display having 512 x 128 add ressable pixels, four" soft-bypass is under keylock control. keys" directly below this display whose functions are under sof tware control, a keylock switch to control the (2) Substitute Rod Position - If the RPIS indi-operationalmodeof theinstrumentandadisplaycontrol cates an unknown position for a rod, the opera-module (printed circuit board) to control the display, tor may, (under certain conditions), provide scan the soft keys and communicate with its counter-l a substituted position for the rod. part in the instrument chassis. This module uses an l l NSC800 microprocessor and onboard random access l (3) Operating Sequence Alignment - The RWM memory (RAM), electrically alterable read-only memory j l 3

w.vx... I ~! O h.D,- k < INkkf)kMkh -, ~ - $ ~ N..o ---m..hNMh8., f5 12 + l-. t!- NI.) t. g linnus >72 f< nan 9-7 i i;l-j l in a n i iry; i e r Figure 2. RWM Operator's Chassis Figure 3. RWMInstrument Chassis DATA ROD MOTION i ACQUISITION PERMISSIVES l SYSTEM PROCESS COMPUTER RMCS I l h ANNUNCIATION SERIAL i RPIS FIBER OPTIC DATA DATA LINK p p i# e <W g GEDAC 1 GEDAC 2 HIGH LEVEL h COMMUNICATIONS COMMUNICATIONS CONTACT AL DU MODULE MODULE INPUT / OUTPUT MODULE 0 0 0 0 l NuMAC 8uS l O 4 7 7 1 + DISPLAY C M E hE UL t $, j OOOO 3 d RWM INSTRUMENT DISPLAY RS232 p DATA LINK I I 66,66 T RWM OPERATOR'S DISPLAY Figure 4. RWM Hardware Block Diagram (EAROM) and programmable read-only memory redundant power supplies for improved availability, a (PROM). front panel consisting of the same electrolurninescent display as the operator's chassis as well as soft keys, The instrument chassis (shown in Figure 3) is a stan-cursor keys, data entry keys, a keylock switch and, dard T* x 19" slide-mounted instrument which may typically, six electronic modules (see Figure 4). The be located some distance from the operator's chassis, number and types of these modules may vary in order it contains a card file and mother board capable of to accommodate I/O requirements of specific plant accommodating up to 15 printed circuit modules, system configurations. 4 i i u ?-,

7.F C..-.L- ~.' s ~ w, _ m - The GEDAC'" 1 module is used to receive control rod operating system (NM86) used in the software design - position and status information from the data acquisi-of all functional CPU modules in the NUMAC product tion system and is equipped with both electrical and line. NM86 is an event-driven, real-time, multitasking fiber-optic receivers and transmitters. Data reception operating system kernel. Its use allows NUMAC soft-and transmission is performed under Direct Memory ware to be partitioned into various functional tasks Access (DMA) control to minimize processor over-which may then execute in pBeudo-cor. current fashion. hud. The module also contains on-board, battery-Figure 5 depicts the software design for both the func-backed R AM which is fuily accessible by both the func-tional CPU and the display controller. The NM86 oper-tional processor and the DMA. Positions for all control ating system controls the resources of the functional. rods in the core are stored in this card module. In some computer including the CPU and certain memory RWM system configurations, the GEDAC 1 is replaced pools. The tasks running under its control in the RWM by modules capable of direct interface with the RPIS design include: i and RMCS. (1) RWM Coordinator Task - Performs data integ-The GEDAC 2 module is used for process computer rity checking, the sequence enforcement func-communication. The rod motion sequences, which are tion, the substitute rod position function, the downloaded from the process computer, are stored in shutdown margin test function, the rod test this module. Up to four motion sequences and two test function and other essential RWM functions. sequences may be stored. The module also handles tr.-_nsmission of " alarm" messages to the process com-(2) Data Acquisition Receiver Task - Processes 4 puter for logging. raw control rod data from the data acquisition system and rearms the DMA on the GEDAC 1 Th3 High Level Contact l/O Module is used to output module for the next data reception. the rod motion permissives and annunciation signals. Thb module contains five Form C contacts and is (3) Process Computer //O Task - Performs all i equipped with a hardware failsafe timer which will solicited process computer communications cause all contacts on the card to go to their normal including sequence download and computer (power off) condition if the functional CPU fails to requests for information, periodically reset it. l (4) Unsolicited Message Task - Handles all unsoli-The analog module reads the operator's keyIock cited status messages to the process computer, switch position. It is also used during self-testing to monitor the redundant power supplies, the instrument (5) Display Message Input Task - Used for the bus voltage levels and the status of the output contacts reception of messages from the display on the I/O module. controller. 3 - The functional CPU module performs the instrument's (6) Display Message Output Task - Transmits system and self-test functions. It controls the GEDAC messages to the display controller. communication modules and the output contacts on the High Level I/O Contact Module. It also communi-(7)Self. Test Task - Performs the self-test func-cates with the instrument's display control module. The tion. This task is the lowest priority task and is module contains a CMOS 8086 microprocessor and always running "in the background" when no on-board RAM, EAROM and PROM memory. other task is ready to run. The display control module is identical (both in hard-The software for the functional CPU module was writ-w:re and software) to the one used in the operator's ten in PL/M-86 and ASM-86 languages, developed on a chassis. It is used to control the electroluminescent VAX 11-780 and debugged using the Intel 121CE 8086 display, scan the four soft keys, data entry keypad and

emulator, cursor control keys, communicate with the functional l

CPU module and communicate with the display con-The software design used for the display controller i trol module in the operator's chassis. (also shown in Figure 5) utilizes an executive loop which calls on lower-level procedures to perform its major functions. These procedures include: (1) Executive - The display system controller. j The RWM design includes software (firmware) for both The executive loops continuously, checking the functional CPU and display controller (s). Func-for new messages from the functional CPU, for tional CPU module software utilizes a multitasking user inputs from the keypad, for expired timers, 5

I ( .-, =.. l puncTioNAL cpu DISPLAY CONTROLLER g Process C TER cooRDenAroR TASK TASK angssAGE NANDLER Raoussi MANoLER UNSOLICITED i Nesse oPaRATese ! Executivs l g o Juis'T isassA 1 Tasa TASK DesPLA status KEEPER g n DasE ^ saLF.TasT TASK output 1 TASK 2 h Figure 5. RWM Software Design f-I l and for messages to relay from or to the opera-self-test status, power level, the present rod sequence i tofs display chassis. and step, and any rod blocks. The middle section dis-plays information appropriate to the operators current (2) Message Handler-The messages processor use of the instrument. The lower section labels the for the display system. It includes communica-functions of the soft keys.These functions (and labels) tion with the functional CPU and the other change during the course of RWM operation, thereby display controller. prompting and guiding an operator in the instrument's use. (3) Request Handler-Responds to user requests as enter' d from the data entry keypad or The following user options are available in the standard e soft keys. versson of the RWM: j (4) Status Keeper - Keeps track of the overall (1) Show " Help" Messages Instrument status. (2) Show " Error" Messages l (3) Show Self-Test (5) Display Handler - Controls the screen seen (4) Run Self-Test l on the electroluminescent display. (5) Show Bypassed Rods (6) Bypass /Unbypass Rod The software developed for the display controller (7) Show Substitute Rods module used in the RWM operators chassis is the same (8) Substitute a Position as that used in the instrument chassis (making the (9) Explain Blocks i modules completely interchangeable). The software (10) Download a Sequence was written in both PASCAL and Z80 assembly lan-(11) Set Parameters guage and was developed on an HP 64000 Logic (12) Show Sequence C.;:W, eat System (13) Check Display (14) Check Pushbuttons i The software design for both computers utilizes NUMAC (15) Select Sequence product generic software wherever possible. The (16) Turn Display Off modular design facilitates the addition or deletion of (17) Shutdown Margin Test functions for unique applications. (18) Rod Test Figure 6 shows some typical RWM user displays. All The RWM has undergone many hours of simulation displays are partitioned into an upper, a middle, and a testing. Its current version of software has been vali-lower portion. The upper section contains general dated both at General Electric's San Jose facility and instrument information including instrument mode, on site. 6 = = - - - - -m

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%~ :. - M EST = - - -- t lin=4. it T h. 4.6,"^- .E.,, N ... A. 4 " U * " N.,. '-, .n. ~-9 I %'lh',.7..,WW,,.,,.. _ 5..h.; Q..!3.g,..Ag,j&,... sLf. 4.1 t Figure 6. Typical RWM Displays j Field Experience Tne RWM hardware was installed in a European eper-RWM was next used during shutdown margin testing i at:ng boiling water reactor in June 1985 (see Figures 7 and performed properly. and 8). Partial software (firmware) installation was made on July 1,1985. The plant subsequently used the, Finally, on September 11, the plant started up using the RWM to reduce power dunng shutdown on August 6. RWM which performed flawlessly. It enabled the plant 1985 before scramming the plant in order to obtain operator to cleanly move through his sequence of scram timing hformation. Final software installation, steps and to keep control rods within their withdrE.w allowing process computer interface to be fully opera-limits. Through successful use of the instrument's g tional, was made in September. The plant's rod motion many displays he quickly developed confidence in the sequences were developed and validated on the pro-functioning of the NUMAC RWM. i cess computer and then downloaded to the RWM. The ? " 8"E M ~1 _ll g( ). 3., e y.. h mm. J j 4 8 9 e .,T m,-,. %. i. = 1 .c ^ ~ ~ ~ ~ ~ l.$:&h>a.;.Q l

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Figure 8. Installation of Operator's Display 2 I i l Conclusions References i The successful design and application of the NUMAC

1. M. R. Benson and S. D. Sawyer, "A New Startup Range NmWn Wnitor,"lEEE Transactions on Nuclear science, Rod Worth Minimizer has demonstrated that hi9 -

h Vol. NS-31, No.1, pp. 868-871. February 1984. quality microprocessor-based hardware can: i

2. U. E. Dennis, Digitallnstrumentaton for Retrofit Applica.

t (1) Serve as the vehicle for the modular design of tions." EPRI seminar: Power Plant Digital Control & Fault. l an integrated family of nuclear instrumenta. werent emcompum, Apnl N 2. 25. l tion and control products. S 1 j (2) Be used to design flexible, plant-specific RWM I systems. 1 (3) Be retrofitted to older nuclear power p' ants to provide enhanced performance. (4) Provide distributed control functions, includ-ing those that depend on the exchange of data with plant process computers. l I i i g i / ~ --}}