ML20236X206
| ML20236X206 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1987 |
| From: | Vissing G Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-675A NUDOCS 8712090049 | |
| Download: ML20236X206 (164) | |
Text
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November 24, 1987 Project No. 675 MEMORANDUM FOR:
The Record FROM:
Guy S. Vissing, Project Manager i.
Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation
SUBJECT:
SUMMARY
OF MEETING WITH COMBUSTION ENGINEERING TO DISCUSS THE SPECIAL FEATURES OF THE SYSTEM 80- DESIGN FOR DESIGN CERTIFICATION INTRODUCTION A meeting of the technical review staff with tha representatives of Combustion Engineering (CE) was held at the NRC offices in Bethesda, Maryland, on November 19, 1987. The purpose of the meeting was for CE to introduce the System 80+ design to the staff, to identify the schedules of the program, to identify the special features of the System 80+ design and to identify materials to aid in the review. Enclosure 1 provides a list of those in attendance. Enclosure 2 provides the viewgraphs which CE used to discuss their program. Enclosure 3 provides a corrected diagram of the proposed emergency feedwater system.
DISCUSSION j
The System 80+ design will include changes to the current System 80 design which CE desire to improve safety and reliability and to accomodate the EPRI ALWR requirements and which NRC mandate through the Safety Goal Policy, the Severe Accident Policy and the Standardization Policy. CE has characterized the System 80+ design to be a revision of the System 80 design.
The scope of the System 80+ design consists of the NSSS, containment, control room, amergency feedwater system and standardized functional descriptions for the remainder of the plant.
CE does not intend to identify deviations from the current Standard Review Plan in those areas which will not be changed from the System 80 design.
However, the areas of proposed changes will follow the guidance of the SRP.
They do not intent to address IE Bulletins but will review the design with respect to the bulletins. Through the request for information process, CE intends to address all staff concerns. There will be no specific information presented concerning any particular site; however, CE will identify those things which will be needed to be provided by an aplicant for site review.
Reviewers will be provided copies of the current System 80 CESSAR-F.
Submittals for each section of the System 80+ CESSAR-DC will be provided in total with the revisions indicated by bars in the right margin. The project 8712090049 871t24 pm
manager will also provide each reviewer copies of the System 80 SER with j
supplements to aid in the review.
There was a concern expressed if consideration was given to designing the l
System 80+ with inservice inspection requirements in mind with such details as providing space and access to accommodate ISI, to reduce exposures, and to meet ISI requirements. CE indicated that ISI and maintenance was a consideration in the EPRI requirements and that CE would follow the EPRI requirements.
Concern was expressed if security will be considered in the design - for instance, if vital areas and equipment would be identified, and if equipment would be separated and isolated. A security plan would not be provided but CE would take under consideration specific security concerns in the design.
No particular detatic have been censidered yet. No design criteria has been identified yet for designing for sabotage. CE indicated that they will develop criteria for designing for sabotoge and then implement it in the design.
Indication was that the design will address fire protection.
Indication was that separation would be used. The staff indicated that fire protection should be considered and the design should provide for fire barriers as the prime means of' protection.
CE intends to provide a safety grade depressurization system consisting of a manual operated valved line from the top of the pressurizer to the refueling water storage tank.
It will be a feed-and-bleed system. There will be no PORV. The size of the valves would be approximately 3 inches.
The proposed emergency feedwater system consist of a four pump system consisting of two motor operated pumps and two turbine driven pumps and two 300,000 gal. storage tanks. For station blackout considerations the system would provide 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of operation. CE is considering expanding this to accommodate 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation which is the EPRI requirement. This wou'd add another 8.0,000 gal. to each tank.
CE was appraised of the status of the proposed leek-before-brake rule change and given a copy of the Federal Register Notice.
Technical Specifications will be provided and will follow the guidance of the current CE Owners Group activity on Technical Specifications.
Y m lVissing, GW S.
ject Manager Standardization nd Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects
Attendance List 1
for Meeting with CE l
on System 80+ Design i
November 19, 1987 i
Name Organization l
Stan Ritterbusch C-E Licensing Guy Vissing NRC Rick Turk C-E ALWR Engr.
Michael Green C-E, Manager Standard Plant Licensing Barry Mendelsohn NRC/NRR/RSGB B. K. McQuoid C-E Regis Matzie C-E, ALWR Project Director George Davis C-E, DOE Project Manager Tom Crom Duke Power John McConaghy Duke Power Philip Hepner CE-Fluid System Ken Scarola CE-Aouanceo-I&C Supu.
Charles Brinkman CE-Bethesda Brad Hardin NRC/RES/DRA/ARGIB Horace Shaw NRC/NRR/ DEST /EMEB Martin Hum NRC/EMTB Frank Witt NRC/ECEB Changwoo Kim NRC/ICSB Jim Stewart NRC/ICSB Norman Romney NRC/ESGB Jonak Revel NRC/SPLB Tim Collins NRC/SRXB L. Rubenstein NRC/PDSNP i
l
November 24, 1987 i i manager will also provide each reviewer ccpies of the System 80 SER with l
supplements to aid in the review.
There was a concern expressed if consideration was given to designing the System 80+ with inservice inspection requirements in mind with such details as providing space and access to accommodate ISI, to reduce exposures, and to
{
1 meet ISI requirements. CE indicated that ISI and maintenance was a consideration i
in the EPRI requirements and that CE would follow the EPRI requirements.
i Concern was expressed if security will be considered in the design - for instance, if vital areas and equipment would be identified, and if equipment J
would be separated and isolated. A security plan would not be provided but l
CE would take under consideration specific security concerns in the design.
No particular details have been considered yet.
No design criteria has been identified yet for designing for sabotage.
CE indicated that they will develop criteria for designing for sabotoge and then implement it in the design.
Indication was that the design will address fire protection.
Indication was that seaaration would be used.
The staff indicated that fire protection should 3e considered and the design should provide for fire barriers as the prime means of protection.
CE intends to provide a safety grade depressurization system consisting of a manual operated valved line from the too of the pressurizer to the refueling water storage tank.
It will be a fee M -bleed system. There will be no PORV.
The size of the valves would te a, w dmately 3 inches.
The proposed emergency feedwater system coi ist of a four pump system consisting of two motor operated pumps and Wo turbine driven pumps and two 300,000 gal. storage tanks.
For station blackout considerations the system would provide 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of operation.
CE is considering expanding this to accommodate 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation which is the EPRI requirement.
This would add another 80,000 gal, to each tank.
CE was appraised of the status of the proposed leak-before-brake rule change and given a copy of the Federal Register Notice.
Technical Specifications will be provided and will follow the guidance of the l
current CE Owners Group activity on Technical Specifications.
original signed by Guy S. Vissing, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects DISTRIBUTION:
Docket File LRubenstein EJordan ACRS (10)
NRC PDR GVissing JPartlow HBClayton, RI i
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Encisoure 2 mm0@6 NUCLEAR POWER MODULE CESSAR 8!!nncemon Kickoff Meeting for Review of CESSAR-DC November 19,1987 COMBUSTION ENGINEERING
COMBUSTION ENGINEERING DESIGN CERTIFICATION PROGRAM KICK 0FF MEETING FOR REVIEW 0F CESSAR-DC NOVEMBER 19, 1987 COMBUSTION ENGINEERING, INC.
3YM1 h h
KICKOFF MEETING FOR REVIEW OF CESSAR-DC PURPOSE o
SUMMARIZE THE SYSTEM 80+ DESIGN O
DESCRIBE EXPECTED PROCESS FOR REVIEW OF CESSAR-DC I
o OBTAIN NRC COMMENTS AND FEEDBACK
KICKOFF MEETING FOR CESSAR-DC REVIEW INTRODUCTION 1
DR. M. D. GREEN
- MANAGER, STANDARD PLANT LICENSING SY472M$hh
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AGENDA Combustion Engineering Design Certification Program -
Kickoff Meeting for CESSAR-DC Review November 19, 1987 Room P118, Phillips Building Bethesda, MD 9:30 - 9:40 Introduction M. D. Green 9:40 - 10:00 Program Overview G. A. Davis 10:00 CESSAR-DC Review R. A. Matzie 10:00 - 10:30 o Summary of System 80+ Design o Materials Engineering Branch o Mechanical Engineering Branch o Electrical Systems Branch o Chemical Engineering Branch o Structural and Geosciences Branch 10:30 - 11:30 o Reactor Systems Branch
- Reactor R. A. Matzie Reactor Ccolant System R. S. Turk 11:30 - 12:30 Lunch 12:30 - 1:30 o Reactor Systems Branch (cont'd) l Safety Systems R. S. Turk P. H. Hepner 1:30 - 2:30 o Plant Systems Branch B. K. McQuoid T. Crom (Duke) 2:30 - 4:00 o Instrumentation and Controls K. Scarola Systems Branch 4:00 Closing M. D. Green
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PROGRAM GOALS 0
SIGNIFICANTLY IMPROVE SAFETY AND RELIABILITY 0
BUILD UPON SYSTEM 80 DESIGN, CONSTRUCTION, AND OPERATING EXPERIENCE o
CERTIFY SYSTEM 80+
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ACTIVITIES T0-DATE AUGUST 30, 1985
- REQUEST FOR FORWARD REFERENCEABILITY MARCH 24, 1987
- PROGRAM INTRODUCTION TO NRC MANAGEMENT JUNE 2, 1987
- KICKOFF MEETING FOR SYSTEM 80+ PRA JULY 2, 1987
- SUBMITTAL OF DRAFT LRB DOCUMENT AUGUST 12, 1987
- KICK 0FF MEETING FOR SYSTEM 80+
DEGRADED CORE ANALYSIS AUGUST 25, 1987
- MEETING ON PRA METHODOLOGY SEPTEMBER 17, 1987 - KICKOFF MEETING FOR RESOLUTION OF USIs/Gis SEPTEMBER 18, 1987 SUBMITTAL OF CESSAR-DC CHAPTER 1 SEPTEMBER 18, 1987 - C-E PROPOSAL FOR DUAL DOCKETS REQUEST FOR CLOSE0VT OF ATWS ISSUE SEPTEMBER 18, 1987 SEPTEMBER 18, 1987 - REQUEST FOR CLOSE00T OF AUXILIARY PRESSURIZER SPRAY ISSUE SEPTEMBER 18, 1987 - REQUEST FOR,.CLOSE05.,~ OF SG TUBE VISRATION ISSUE OCTOBER 13, 1987
- NRC RESPONSE ON DUAL DOCKETS NRC MEETING - EXECUTIVE SUP9tARY OCTOBER 29, 1987 NOVEMBER 11, 1987 SUBMITTAL OF COP 94ENTS ON RULEMAKING FOR STANDARDIZATION 3K5ft
'i OVERVIEW 0F C-E's DESIGN CERTIFICATION PROGRAM G. A. DAVIS PROGRAM MANAGER, DOE ALWR PROJECT
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DESIGN CERTIFICATION PROGRAM APPROACH 0
USE SYSTEM 80/CESSAR AS STARTING POINT PROVEN, STANDARD DESIGN SHOWN TO MEET CURRENT NRC REQUIREMENTS COMPLETE DESIGN DETAIL AVAILABLE FOCUSES ATTENTION ON DESIGN IMPROVEMENTS AND NEW NRC REQUIREMENTS 0
ADDRESS EPRI ALWR REQUIREMENTS DOCUMENT 4
0 ADDRESS NRC SEVERE ACCIDENT POLICY CURRENT REGULATIONS PROBABILISTIC RISK ASSESSMENT AND EVALUATION RESOLVE NRC's UNRESOLVED SAFETY ISSUES EVALUATE DEGRADED CORE ISSUES o
APPLY FOR DESIGN CERTIFICATION UNDER NEW NRC STANDARDIZATION POLICY L
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DESIGN CERTIFICATION PROGRAM EXPANDED SCOPE o
CURRENT SYSTEM 80 DESIGN INCLUDES NUCLEAR l
STEAM SUPPLY SYSTEM (NSSS) o SYSTEM 80+ DESIGN INCLUDES NSSS CONTAINMENT EMERGENCY FEEDWATER SYSTEM J
ADVANCED CONTROL CENTER (NUPLEX 80+)
STANDARDIZED FUNCTIONAL DESCRIPTIONS FOR REMAINDER OF PLANT l
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DESIGN CERTIFICATION PROGRAM STANDARDIZED FUNCTIONAL DESCRIPTIONS o
CURRENT VERSION OF CESSAR (FOR SYSTEM 80)
INCLUDES INTERFACE REQUIREMENTS FROM NSSS SYSTEMS o
NEW VERSION OF CESSAR (FOR SYSTEM 80+) INCLUDES DETAILED FUNCTIONAL DESCRIPTIONS FOR SYSTEMS AND STRUCTURES OUTSIDE OF NUCLEAR POWER MODULE'S (NPM) SCOPE o
FUNCTIONAL DESCRIPTIONS WILL INCLUDE DESCRIPTION OF SYSTEM / STRUCTURE INTERFACE REQUIREMENTS FROM NPM ASSUMPTIONS FROM SAFETY ANALYSES AND PRA, INCLUDING RELIABILITY EPRI ALWR REQUIREMENTS ALL INFORMATION NECESSARY FOR NRC TO CLOSE OUT REVIEW 0F NPM o
DUKE POWER C0. IS PREPARING STANDARDI7_ED FUNCTIONAL DESCRIPTIONS I
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l 10.0 STEAM POWER CONVERSION SYSTEM 10.1
SUMMARY
DESCRIPTION
~
See Applicant's SAR.
J 10.2 TURBINE GENERATOR See Applicant's SAR.
10.3 MAIN STEAM SUPPLY SYSTEM 10.3.1 DESIGN BASES The Main Steam Supply System is discussed in the Applicant's SAR.
See Section 5.1.4 for interface criteria.
10.3.1.1 Environmental Design Bases The environmental design requirements are contained in the Applicant's SAR.
10.3.1.2 Inservice Inspection Requirements Inservice inspection requirements are contained in the Applicant's SAR.
10.3.2 EVALUATION See Applicant's SAR.
10.3.3 INSPECTION AND TESTING REQUIREMENTS See Applicant's SAR.
10.3.4 SECONDARY WATER CHEMISTRY 10.3.4.1 Chemistry Control Basis Steam generator secondary side water chemistry control is accomplished by:
a.
Close control of the feedwater to limit the amount of impurities which can be introduced into the steam generator, b.
Continuous blowdown of the steam generator to reduce the concentrating effects of the steam generator, c.
Chemical addition to establish and maintain an environment which minimizes system corrosion.
d.
Preoperational cleaning of the feedwater system.
Minimizing feedwater oxygen content prior to entry into the steam e.
generator.
10.3-1 1
)
f)
Isolate the NSSS steam generators from the remainder of ' the main 1
steam system. when necessary (including containment isolation-post i
LOCA).
g)
Provide adequate overpressure protection for the NSSS steam j
generators and main steam system.
h)
Conform to applicable design codes.
1)
Permit visual inservice inspection.
j)
Adequately cover the environmental operating conditions for the system and be thermally insulated to protect adjacent equipment and conserve energy.
2.
The safety related portion of the main steam system is that portion between the steam generators and main steam isolation valves.
10.3.2 SYSTEM DESCRIPTION Steam is generated in two steam generators by heat transferred from the Reactor Coolant System to the feedwater.
Steam for the turbine driven emergency feedwater pumps is taken from either of the two steam generators via two of the four main steam lines at a point outside of containment and upstream of the main steam isolation valves (MSIVs).
Five ASME Code spring loaded secondary safety valves are provided for each individual main steam line for protection against over-pressurization of the shell side of the steam generators and the main steam line piping up to the inlet of the turbine stop valve.
An atmospheric steam dump valve is provided on each of the two main steam lines from each steam generator downstream of the safety valves.
Each main steam line is provided with an isolation valve for positive isolation against forward steam flow and adequate isolation against reverse fl ow.
The MSIV on one of the two lines from each steam generator is provided with a bypass around,it for warm-up of the steam lines downstream of the isolation valves and pressure equalization prior to admitting steam to the turbine.
Downstream of the MSIVs are eight power-operated bypass valves to bypass steam to the condenser.
These valves comprise the Turbine Bypass System, which is discussed in Subsection 10.4.4.
Each of the four main steam lines is provided with a turbine stop valve and a turbine control valve to shutdown and control the turbine.
The following standardized functional descriptions and requirements present the system configuration necessary to meet the ilPM licensing, safety, and reliability requirements.
The final detailed design and layout of the main steam supply system is described in the site-specific SAR supplement.
l 10-3 i
I
-l CESS AR-DC 10.3.2.1 System Performance 1)
The main steam piping, its isolation valves, all steam branches, their isolation valves, and all associated supports from the steam generators up to and including the required isolation valves are Seismic Category I, and are designed in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code, Class 2.
The remaining steam piping is in accordance with ANSI-831.1.
The steam piping and supports are designed so that any single adverse event, such as a ruptured main steam line or a closed isolation valve, can occur without:
a.
Initiating a loss-of-coolant incident; a
b.
Causing failure of any steam lines, Main Steam Isolation Valves (MSIV), Main Feed Isolation Valves (MFIV), Safety Valves, Atmospheric Dump Valves, or any feedwater line required for controlled cooldown of the unaffected steam generator; c.
Preventing the Reactor Protective System, and Engineered Safety Features Actuation systems from initiating proper safety actions.
d.
Transmitting excessive loads to the containment pressure boundary.
e.
Compromising the function of the plant control room.
f.
Precluding an orderly cooldown of the RCS.
2)
The design pressure, temperature and flow rating of the main steam piping and valves are greater than or equal to the design pressure, temperature and flow rating of the steam generator secondary side.
10.3.2.2 System Arrangement 1)
All valves in the main steam lines outside of containment up to and l
including the MSIVs are located as close to the containment wall as practical.
2)
The main steam lines are headered together prior to the Turbine Stop Valves but not upstream of the MSIVs, and a cross-connect line is provided which will maintain steam generator pressure differences I
(considering arrangement effects only) within the stated limits for all 1
normal conditions (including exercise of the Turbine Stop Valves):
1.
0-15% power operation pressure difference to be 1 psi 2.
15%-100% power operation pressure difference to be 3 psi 3.
For transient conditions of no greater than 1 minute duration, pressure difference to be less than 30 psi.
This cross-connection is sized to permit a full-closure test at 90% power of one of the HP turbine stop valves to be performed during operation without imposing a severe pressure / load transient on one of the steam generators.
10-4 4
GESSAR-DC 3)
There are no isolation valves in the main steam lines between the steam generators and the Secondary Safety Valves.
4)
The MSIVs, the Secondary Safety Valves, the Atmospheric Dump Valves, and the MSIV Bypass Valves are environmentally protected against the effects of missiles, steam, pipe whip, etc., such that these events cannot l
prevent the valves from performing their requisite safety function.
l 5)
The Secondary Safety valves are installed in accordance with the l
applicable provisions of the ASME Boiler and Pressure Vessel Code Section i
III.
6)
The Secondary Safety Valve discharge piping is arranged and supported such that the limiting loads are not exceeded for normal and relieving conditions.
l 7)
In tre combined event of a steam line break and the loss of power or a steam generator tube rupture and loss of power, personnel access to P' manual operator of the intact Atmospheric Dump Valves on the intact steam generator is possible.
8)
No automatically actuated valves are located upstream of the Main Steam Isolation Valves except as required for automatic startup of steam driven emergency feedwater pumps.
l 9)
The system piping is designed to allow cleaning for the removal of foreign material and rust prior to operation and to prevent introduction of this material into the turbine.
Chemical cleaning or hand cleaning may be employed. During chemical cleaning, no fluid can enter the steam generators. Suitable bypass piping is provided where applicable.
- 10) Emergency feedwater pump turbine steam supplies are taken off the main steam lines upstream of the Main Steam Isolation Valves.
- 11) Main feedwater pump turbine steam supplies are located on the downstream side of the Main Steam Isolation Valves.
- 12) Following a secondary line break, either all steam paths downstream of the MSIVs are shown to be isolated by their respective control systems following an MSIV Actuation Signal, or the results of a blowdown through a non-isolated path are shown to be acceptable.
An acceptable maximum steam flow from a non-isolated steam path is 10", of Maximum Steaming Rate *. It is not required that the control systems for downstream valves nor the downstream valves themselves be designed to ASME Code,Section III, Seismic Category I, IEEE 279 or IEEE 280 Criteria.
- 13) The Main Steam Safety */alves are arranged such that any condensate in the line between the safety valves and main steam line drains back to the main steam line.
6
- 19 x 10 lb/hr 9 1000 psia saturated steam.
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IMPROVE MAN-MACHINE INTERFACE PLANT OVERVIEW COMPREHENSION ALARM HANDLING FRIENDLY TRANSITION TO ACR OPERATION WITHOUT CRTS REDUCE CONSTRUCTION AND MAINTENANCE COSTS MULTIPLEXING i
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i MICR0 PROCESSOR TECHNOLOGY UTILIZATION OF 0FF-THE-SHELF MICR0 PROCESSOR TECHNOLOGY TO
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REDUCE THE NUMBER OF PHYSICAL DISPLAY DEVICES:
GROUPED ALARM WINDOWS WITH DYNAMIC MESSAGE DISPLAYS DYNAMIC INDICATORS CRT'S o
REDUCE THE QUANTITY OF DATA:
LIKE SENSOR VALIDATION PARAMETER PRIORITIZATION ALARM ELIMINATION, FILTERING AND PRIORITIZATION O
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NUPLEX 80 PLUS INTEGRATED PROCESS STATUS OVERVIEW (IPS0) 5 HIGHEST LEVEL IN DISPLAY HIERARCHY TO IMPROVE INTEGRATION OF DISCRETE INDICATORS, ALARMS AND CRT'S, PROVIDES SINGLE LOCATION FOR CONTINU0US DISPLAY OF THE FOLLOWING STATUS CONDITIONS FOR MAJOR PLANT PROCESSES AND COMPONENTS:
EXISTENCE OF PRIORITY 1 ALARMS:
TO AID OPERATOR PRIORITIZATION OF ALARMS BETWEEN MAJOR SYSTEMS AND COMPONENTS DEVIATIONS FROM CONTROL SETPOINTS AND IDENTIFICATION OF IMPROVING OR DEGRADING TRENDS:
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ALARMS INDICATE A PROBLEM ONLY - STATUS INFORMATION l
INDEPENDENT OF ALARM SYSTEM, NUMBER OF DISCRETE ALARM WINDOWS SIGNIFICANTLY REDUCED BY ALARM GROUPING WITH DYNAMIC MESSAGE DISPLAYS, NUMBER OF ALARMS SIGNIFICANTLY REDUCED THROUGH SIGNAL VALIDATION AND DYNAMIC ALARM PROCESSING.
DIRECT ACCESS TO CRT PAGES PROVIDED FOR OBTAINING DIAGNOSTIC INFORMATION.
FIRST-0VT INDICATION PROVIDED FOR REACTOR AND TURBINE TRIPS,
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PROVIDE REDUCTION OF INDICATIONS THROUGH AUTO RANGING AND SIGNAL VALIDATION TECHNIQUES.
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l PROVIDE ACCESS TO PLANT DATA REQUIRED FOR OPERATION WITHOUT THE PLANT COMPUTER; FUNCTIONALLY INDEPENDENT FROM PLANT l
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TOTAL 73 INDICATORS 13 INDICATORS TOTAL OVERALL REDUCTION 82%
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CHEMICAL VOLUME CONTROL SYSTEM SYSTEM 80 NUPLEX 80+
SYSTEM INDICATORS INDICATORS CHARGING 11 3
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" UTILITY DISPLAY CONTAINS INFREQUENTLY ACCESSED PARAMETERS IN PREVIOUS CATEGORIES.
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SLIDE 065 CRT DISPLAYS PROVIDE ACCESS TO ESSENTIALLY ALL PLANT INFORMATION FROM ANY LOCATION WITHIN CONTROL ROOM OR REMOTE LOCATIONS:
DYNAMIC COLOR GRAPHIC PLANT MIMIC DISPLAYS, INCLUDING IPS0 ALARMS AND OPERATOR AID INFORMATION PLANT COMPUTER CALCULATED INFORMATION TRENDS AND REPORTS IMPLEMENTATION THROUGH NON-SAFETY PLANT COMPUTER.
DUPLICATES AND VERIFIES ALL DISCRETE ALARM AND DISPLAY SYSTEM PROCESSING AND IDENTIFIES SIGNIFICANT DIFFERENCES.
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SLIDE 065 NUPLEX 80+ CONTROL ROOM CONFIGURATION DESIGNED FOR ONE PERSON OPERATION FROM HOT SHUTDOWN TO FULL POWER.
DESIGNED TO ACCOMMODATE ADDITIONAL STAFFING CONTINUOUSLY AND FOR SHUTDOWN OR EMERGENCIES.
DESIGNED TO ACCOMMODATE A CONTROL ROOM SUPERVISOR.
DESIGNED TO MAINTAIN A " CLEAN CONTROL ROOM";
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DESIGNED TO PROVIDE CONTINUOUS PLANT OVERVIEW VISIBLE FROM
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UNWARRANTED CPC TRIPS ALSO OCCUR DUE TO SINGLE FAILURES IN THE ONE-0UT-0F-TWO CEA POSITION / CALCULATOR DESIGN, FOR INSERTION DEVIATIONS (REAL OR SPURIOUS) DELAY CPC TRIP AND INITIATE REACTOR POWER CUTBACK.
IF CUTBACK FAILS OR DOES NOT PROVIDE SUFFICIENT MARGIN, THEN CPC TRIP WILL OCCUR.
FOR WITHDRAWAL DEVIATIONS (REAL OR SPURIOUS), INITIATE CEA MOTION BLOCK USING REDUNDANT 1/2 LOGIC, CHANGE CPC TRIP TO 2/2 LOGIC.
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SLIDE 065 NUPLEX 80 PLUS FOUR-CHANNEL ESFAS CURRENT PPS DESIGN HAS FOUR CHANNEL ACTUATION FOR SOME ESF FUNCTIONS (EFW, RAS) TO PRECLUDE SINGLE FAILURES FROM SPURIOUSLY ACTUATING.
SPURIOUS ESF ACTUATIONS HAVE OCCURRED WHERE FOUR-CHANNEL ACTUATION IS NOT UTILIZED (CSAS MOST SIGNIFICANT).
5 NUPLEX 80 PLUS PPS DESIGN WILL INCLUDE GENERIC CAPABILITY OF FOUR CHANNELS FOR ALL ESF ACTUATION SIGNALS.
SYSTEM 80 PLUS DESIGN STRIVING TO RELAX SAFETY NEED FOR j
FOUR-CHANNEL DESIGNS.
I C-E WILL MAINTAIN FOUR-CHANNEL DESIGN TO ENHANCE INVESTMENT PROTECTION AND AVOID UNWARRANTED PLANT TRANSIENTS.
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l NUPLEX 80 PLUS ALTERNATE PROTECTION SYSTEM ALTERNATE REACTOR TRIP SYSTEM (ARTS)
TRIPS CEDM POWER SOURCE ON HIGH PRESSURIZER PRESSURE.
ALTERNATE FEEDWATER ACTUATION SYSTEM (AFAS)
ACTUATES EMERGENCY FEEDWATER SYSTEM ON LOW STEAM GENERATOR LEVEL.
DIVERSE FROM PPS NON-CLASS 1E TWO-CHANNEL DESIGN WITH 2/2 ACTUATION.
TO ACCOMMODATE SPURIOUS AFAS ACTUATION, REACTOR TRIP SIGNAL WILL BLOCK / TERMINATE AFAS (N0 ATWS).
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ALL SYSTEM ELECTRONICS GE0 GRAPHICALLY SEPARABLE BY CHANNEL.
ELECTRICAL INDEPENDENCE OF CHANNELS MAINTAINED IN ALL AREAS.
HOWEVER, PHYSICAL SEPARATION AND BARRIERS ELIMINATED WITHIN CONTROL PANELS TO AVOID COMPROMISING GOOD HUMAN FACTORS IN CONTROL PANEL LAYOUTS.
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SLIDE 065 NUPLEX 80 PLUS COMPONENT CONTROL SYSTEM PROVIDES CONTROL LOGIC FOR ALL COMPONENTS CONTROLLED FROM THE MAIN CONTROL ROOM (= 1050).
FOUR CLASS 1E, TWO NON-CLASS 1E TRAINS.
PROVIDES CENTRAL INTERFACE FOR ALL COMPONENT RELATED CONTROL AND STATUS SIGNALS FROM/TO ALL SYSTEMS.
MAXIMIZES USE OF DISTRIBUTED MULTIPLEXER AND INTER-SYSTEM DATA LINKS TO REDUCE CONSTRUCTION COSTS.
3 IMPLEMENTATION USING 0FF-THE-SHELF PROGRAMMABLE LOGIC CONTROLLERS.
t COs1PONENTS ASSIGNED TO PLC'S TO ENSURE MINIMIZED AND PREDICTABLE DEGRADATION.
INCLUDES DIESEL LOADING SEQUENCERS THAT ADAPT THE LOAD SEQUENCE BASED ON ESFAS SIGNALS TO MINIMIZE LOADING TIMES AND ELIMINATE SUCCESSIVE LOAD SHEDDING.
DIVERSITY WITHIN THE DESIGN TO ENSURE:
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SLIDE 065 NUPLEX 80 PLUS DISCRETE INDICATION AND ALARM SYSTEM i
GATHERS AND PROCESSES DATA FOR DISCRETE INDICATORS AND ALARM 3
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SYSTEM 1/0, PROCESSING AND INDICATION ARE SEGMENTED BY FLUID SYSTEM (E.G., RCS, CVCS, ETC.).
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l PLANT DATA ACQUISITION SYSTEM REMOTE DISTRIBUTED MULTIPLEXER TO GATHER FIELD INPUTS DATA LINKS TO OTHER ISC SYSTEMS REDUNDANT COMMUNICATIONS AND DATA PROCESSING PLANT MONITORING AND DISPLAY SYSTEM PROCESSES ALL DATA FOR DYNAMIC CCLOR CRT PRESENTATIONS NSSS AND B0P OPERATING LIMIT MONITORING PERFORMANCE CALCULATIONS HISTORICAL DATA STORAGE AND RETRIEVAL REDUNDANT CPU'S DISTRIBUTED CRT'S EACH CAPABLE OF ACCESSING ANY INFORMATION JYArt h h
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NUPLEX 80+ SOFTWARE QUALIFICATION 0
VERIFICATION AND VALIDATION OF USER DEVELOPED APPLICATIONS SOFTWARE IN ACCORDANCE WITH:
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SLIDE 065 NUPLEX 80 PLUS
SUMMARY
l ADDITIONAL STANDARDIZATION THROUGH GENERIC APPLICATION OF CCS MICROPROCESSOR TECHNOLOGY, REDUCED CONSTRUCTION COSTS THROUGH ADDITIONAL MULTIPLEXING.
BETTER INFORMATION PRESENTATION AND FRIENDLY TRANSITION TO AN ADVANCED CONTROL ROOM THROUGH ENHANCED DATA PROCESSING AND INTEGRATION OF CRT'S WITH CONVENTIONAL DISPLAYS.
INCREASED PLANT AVAILABILITY THROUGH RPCS, 4-CHANNEL ESFAS AND IMPROVED FIRE PROTECTION, IMPROVED LOAD FOLLOWING THROUGH ENHANCED MDS, '
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