ML13317A015

Rept in Support of Simulation Facility Certification,San Onofre Nuclear Generating Station,Unit 1,Control Room Simulator

ML13317A015
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Site:	San Onofre
Issue date:	02/29/1992
From:	Southern California Edison Co
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ML13309A185	List: ML13309A184 ML13317A015
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NUDOCS 9202120216
Download: ML13317A015 (92)

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REPORT IN SUPPORT OF THE SIMULATION FACILITY CERTIFICATION FOR THE SAN ONOFRE NUCLEAR GENERATING STATION UNIT 1 CONTROL ROOM TRAINING SIMULATOR February 1992 9202120216 920205 POR ADOCK 05000206 P

PDR

SUMMARY

Title 10, Code of Federal Regulations, Part 55.45 (10 CFR 55.45) requires that either a certified or an NRC approved simulation facility must be used to administer operating tests after May 26, 1991.

To comply with this requirement, Southern CalifornIa Edison (SCE) purchased, in April 1990, a Control Room Training Simulator for the San Onofre Nuclear Generating Station Unit 1 (SONGS 1).

The purchase agreement of the simulator along with, the establishment of a certification program for the simulator was created to certify that the facility meets the requirements of 10 CFR 55.45, and conforms to the guidance contained in Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for Use in Operator License Examinations," and ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training."

By letters dated March 23, 1990, and November 2, 1990, SCE requested an exemption from the schedular requirements of 10 CFR 55.45(b) regarding certification submittal. The NRC, in a March 22, 1991 letter, granted an extension for submittal of NRC Form 474, "Simulation Facility Certification."

With this report, SCE is submitting NRC Form 474, which together with this report provides the required supporting documentation. The facility reviews, baseline data, configuration control,. and performance tests conducted are summarized in the report.

I

TABLE OF CONTENTS

SUMMARY

i TABLE OF CONTENTS.............................

LIST OF APPENDICES..............................

1 LIST OF TABLES....................................

iv LIST OF FIGURES.......................................

iv CHAPTER

1.0 INTRODUCTION

1 1.1 REGULATORY REQUIREMENTS 1

1.2 CERTIFICATION PROGRAM 2

1.3

SUMMARY

OF EXCEPTIONS TO ANSI/ANS 3.5-1985.

2Property "ANSI code" (as page type) with input value "ANSI/ANS 3.5-1985.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 1.4 SIMULATOR REVIEW BOARD.

4 CHAPTER 2.0 FACILITY DESCRIPTION.

7 2.1 SIMULATOR CONTROL ROOM.

7 2.2 INSTRUCTOR STATION.

9 2.3 OPERATING PROCEDURES AND OPERATOR AIDS 15 2.4 SIMULATOR BASELINE DATA 15 2.5 HARDWARE VERIFICATION AND OVERALL FIDELITY.

15 2.6 ACCEPTABILITY FOR OPERATOR TESTING....

16 CHAPTER 3.0 SCOPE AND LIMITS OF SIMULATION..

21 CHAPTER 4.0 PERFORMANCE TEST ABSTRACTS..........

22 4.1 REAL-TIME TEST ABSTRACT 23 4.2 STEADY-STATE TESTS ABSTRACT 25 4.3 CORE PHYSICS TESTS ABSTRACT 31 4.4 SURVEILLANCE TESTS ABSTRACT 37 4.5 NORMAL OPERATIONS TESTS ABSTRACT.

41 4.6 TRANSIENT TESTS ABSTRACT 44 4.7 MALFUNCTION TESTS ABSTRACT.

48 4.8 DEFICIENCY STATUS 49 CHAPTER 5.0 CONFIGURATION MANAGEMENT SYSTEM 57 5.1 CONFIGURATION CONTROL IN MONTREAL 60 5.2 REPORTS AND RECORDS........

61 CHAPTER

6.0 CONCLUSION

S.

63 (IIi

TABLE OF CONTENTS (continued)

LIST OF APPENDICES Appendix A Simulator Verification ProcedureSO123-XXI-3.1.1 Appendix B Simulator Verification ProcedureSO123-XXI-3.1.2 Appendix C Simulator Verification ProcedureSO123-XXI-3.1.3 Appendix D Simulator Verification ProcedureSO123-XXI-3.1.4 Appendix E Simulator Verification ProcedureSO123-XXI-3.2.1 Appendix F Simulator Verification ProcedureSO123-XXI-3.2.2 Appendix G Simulator Verification ProcedureSO123-XXI-3.2.3 Appendix H Simulator Verification ProcedureSO123-XXI-3.2.4 Appendix I Simulator Verification ProcedureSO123-XXI-3.2.6 Appendix J Simulator Verification ProcedureSO123-XXI-3.3.1 Appendix K Simulator Verification ProcedureSO123-XXI-3.3.2 Appendix L Simulator Verification ProcedureSO123-XXI-3.3.3 Appendix M Simulator Verification ProcedureSO123-XXI-3.3.4 Appendix N Simulator Verification ProcedureSO123-XXI-3.3.5 Appendix 0 Simulator Verification ProcedureSO123-XXI-3.3.6 Appendix P Simulator Verification ProcedureSO123-XXI-3.3.7 Appendix Q Simulator Verification ProcedureSO123-XXI-3.4.1 Appendix R Surveillance Tests Data Sheets Appendix S

......... Normal Operations Tests Data Sheets Appendix T

............. Transient Tests Data Sheets (1)i

LIST OF TABLES TABLE I List of Simulator Verification Procedures...... 6 TABLE II Panels and Cabinets within Main Control Room 12 TABLE III ANSI 3.5 Requirement vs Malfunction Number 17 TABLE IV Baseline Data...................

20 TABLE V Simulator Malfunctions 50 LIST OF FIGURES Figure No. 1.

SONGS 1 Simulator Layout...

10 Figure No. 2.

SONGS 1 Control Room Layout.....

11 iv

1.0 INTRODUCTION

Southern California Edison's (SCE) SONGS 1 simulation facility is a real-time, plant-referenced simulator that duplicates the SONGS 1 Control Room. The simulator was built in 1990 and 1991 by CAE Electronics Limited (CAE), and is presently located in Montreal, Quebec, Canada at the vendor's assembly facility.

The simulation models, programs, instructor station, control boards and computer hardware have all undergone thorough vendor testing and SCE factory acceptance testing. These tests verified routine and transient operation capabilities and were used to enhance the simulator dynamic response to match available plant data and ensure simulator realism. Deficiencies were identified and prioritized and either have been corrected or will be corrected in accordance with the Configuration Management System which is discussed below and is described in more detail in Chapter 5.

A computer-based Configuration Management System (CMS) and implementing procedures were created in 1984 to manage the San Onofre Nuclear Generating Station Unit 2 (SONGS 2) simulator and are effective.

The same CMS and procedures have been implemented on the SONGS 1 simulator to ensure fidelity to the reference plant.

Regulatory Guide 1.149 and ANSI/ANS 3.5-1985 formed the bases for the Purchase Specification to the simulator vendor, CAE, and the Certification Program. The specification and the program were developed to satisfy the initial certification requirements as well as the continued certification as required by 10 CFR 55.45(b)(5)(ii).

This report presents the results of the reviews and performance tests conducted for initial certification.

The schedule for the four-year performance testing program for continued certification can be found in Simulator Verification Procedure S0123-XXI-3.1.2, "Performance Test Scheduling,"

(Appendix B).

1.1 REGULATORY REQUIREMENTS Paragraph 55.45(a) of Title 10, Code of Federal Regulations, Part 55 (10 CFR 55), "Operator's Licenses," requires that an applicant for either an operator or senior operator license demonstrate both an understanding of and the ability to perform certain essential job tasks.

Paragraph 55.45(b) specifies that these operating tests will be administered, in part, either in a simulation facility consisting solely of a plant-referenced simulator that has been certified to the Commission by the facility licensee, or in a simulation facility approved by the Commission after application has been made by the facility licensee.

SCE has elected to meet this requirement by purchasing a plant-referenced simulator.

Regulatory Guide (RG) 1.149, "Nuclear Power Plant Simulation Facilities for Use in Operator License Examinations," specifies that the guidance set forth in ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training," constitutes the minimum performance and configuration criteria for a simulator.

RG 1.149 also indicates that ANSI/ANS 3.5-1985 should be used for comparing a simulator to its reference plant, and for upgrading the simulator to reflect changes to reference plant response or control room configuration.

SCE has used both RG 1.149 and ANSI/ANS 3.5-1985 to provide the bases and guidelines for certifying the SONGS 1 Control Room Training Simulator for the conduct of operator testing required by 10 CFR 55.45.

1.2 CERTIFICATION PROGRAM The Certification Program was designed to ensure that the regulatory requirements for a plant-referenced simulator are met.

The program addresses both the initial certification and the four-year performance test requirements. The program provides for:

• Comparing the simulator to the SONGS 1 Control Room,

• Upgrading the simulator due to changes in SONGS 1, and

• Ensuring the facility remains acceptable for both operator training and NRC operator license exams.

The simulator verification procedures are part of the Certification Program. These procedures provide the documentation and step-by-step methodology used for testing and maintaining the certification of the plant-referenced simulator.

Table I (Page 6) lists the simulator verification procedures.

The procedures used for certification are included in this report in Appendices A through Q.

1.3

SUMMARY

OF EXCEPTIONS TO ANSI/ANS 3.5-1985 The SONGS 1 Control Room Training Simulator meets the ANSI/ANS 3.5-1985 standard with the following exceptions:

(1) Appendix B of ANSI/ANS 3.5-1985 requires that simulator data be recorded at a resolution of one-half second or less.

The data resolution was changed to one second to permit the recorded simulator data to more readily match the data resolution of the actual plant.

Plant data resolution varies from one to sixty seconds. The one-second data resolution is deemed sufficient for the required transient comparisons.

(2) ANSI/ANS 3.5-1985, Section 3.1.1, "Normal Plant Evolutions,"

identifies the minimum number of evolutions that a simulator shall be capable of performing.

Item (7),

"Startup, Shutdown, and Power Operations With Less Than Full Reactor Coolant Flow," is limited by the SONGS 1 Technical Specifications, Section 3.1.2.C, to less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with power below 10% or to low power physics testing. Also, the SONGS 1 Reactor Protection System (RPS) does not allow operation above 50% power with less than full flow in 3 reactor coolant loops and above 10% power with less than full flow in 2 reactor coolant loops.

In practice, continued power operation with less than full RCS flow at SONGS 1 is limited to situations when RCP malfunctions occur below the respective RPS trip setpoints and the plant is taken off-line for repair. This abnormal evolution is included as a transient operation test in Section 3.1.2 item (4), "Loss of forced core coolant flow due to single or multiple pump failure".

Therefore, this evolution is not within the scope of normal plant evolutions for the SONGS 1 Control Room Training Simulator.

(3) ANSI/ANS 3.5-1985, Appendix B2.2. "Transient Performance" identifies the transient evolutions that a simulator shall be capable of performing. Item (3),

"Simultaneous Closure of all Main Steam Isolation Valves" cannot be performed on SONGS 1, <which does not have Main Steam Isolation valves.

(4) ANSI/ANS-3.5-1985 Section 3.2.3, "Control Room Environment,"

requires that the Control Room environment be duplicated "as much...as is reasonable and practical."

The simulator is being certified at a temporary location. The environment at this temporary location is different from the Control Room environment. Specific exceptions are documented below and in Section 2.1 "Simulator Control Room."

These exceptions have been determined to have an insignificant effect on operator performance or response.

Exceptions. The desks and consoles, while in the same relative location as the plant, are standard office furniture and, therefore, different from that in the plant.

The control room lighting, location, intensity, and response to loss-of-power do not exactly replicate the plant control room. The lighting fixtures are, however, similar in nature and of the same type (i.e. fluorescent) as used in the plant.

In addition, a similar but different loss of power cue is provided compared to that used in the plant. These differences have been evaluated and it has been determined that the operator response and interface will not be significantly affected. The carpeting installed at the simulator is of the same basic color but a different shade than that in the control room.

The communications system presently used for the simulator comprises two systems.

One is a closed-circuit telephone system confined to the simulator area, and the other is the CAE Electronics internal telephone system. This communication system is different in function and appearance from the one used by the SONGS 1 control room. An intercom between the simulator control room and the instructor station substitutes for the control room radio. A single telephone, located on the control operator's desk can be used to call one of two phones at the instructor facility.

This phone system is part of the CAE Electronics Ltd. phone system and uses specific numbers that are different from those used at SONGS.

1.4 SIMULATOR REVIEW BOARD The Simulator Review Board (SRB) is the management oversight group responsible for ensuring that the simulator conforms with ANSI/ANS 3.5-1985.

In support of the initial certification, the SRB has:

• Reviewed and approved the Certification Program. This review verified that all elements of ANSI/ANS 3.5-1985 are included in the program.

• Reviewed and approved the certification program procedures. This review verified that the ANSI/ANS 3.5-1985 guidance was met by using the procedures.

• Reviewed and approved performance test results, exceptions, and the justification for each exception.

These reviews verified that ANSI/ANS 3.5-1985 test criteria were met or that exceptions were acceptable.

• Reviewed and approved the simulator transient benchmark tests. This review verified that the baseline comparison was appropriate, the justification for any response differences was correct, and these differences would not impact training.

• Reviewed and approved simulator hardware deviations and their justifications in all cases where operator response could be affected.

• Reviewed all open CAE Deficiency Reports (DR) and Simulator Work Orders (SWO) (described below) to determine the impact on training and testing of operators.

This assessment ensured that the board was aware of the current status of all identified performance issues. The review evaluated and approved DR and SWO priorities based on the potential impact to operator training and testing.

Reviewed and approved the initial certification report.

The SRB consists of a panel of experts from organizations that have the most interest in the performance and fidelity of the simulator.

These organizations are: Operations, Station Technical, NES&L Project, and Nuclear Training. The Manager of Nuclear Training, or his designee, chairs the SRB. Simulator Verification Procedure S0123-XXI-3.1.1 "Simulator Review Board",

(Appendix A) details the make-up and responsibilities of the SRB.

When the board lacks sufficient expertise or information for an adequate review of an item, additional personnel are added to the SRB as non-voting guests to complete the evaluation.

The SRB will continue to review simulator performance and provide management oversight after initial certification. The board will continue to assess the impact of changes to the certification procedures. The board will also review the results of annual performance tests, annual simulator reports, SWOs and the results of annual audits.

In support of the four-year recertification requirements in 10 CFR 55.45, the SRB will:

• Review and approve the recertification submittal;

• Review and approve all changes to the Certification Program and its implementing procedures;

• Perform sample audits to ensure the certification is maintained;

• Review and approve all performance results, deviations, and their justifications; and

• Review and approve the annual performance-based audit (Vertical Audit).

These activities will ensure that the SONGS 1 Control Room Training Simulator will remain certified for operator training and NRC operator examinations.

TABLE I LIST OF SIMULATOR VERIFICATION PROCEDURES SIMULATOR ADMINISTRATIVE PROCEDURES REVISION S0123-XXI-3.1.1 SIMULATOR REVIEW BOARD 2

S0123-XXI-3.1.2 PERFORMANCE TEST SCHEDULING 2

S0123-XXI-3.1.3 ANNUAL REPORT GENERATION 2

S0123-XXI-3.1.4 MAINTENANCE OF CERTIFICATION RECORDS 3

S0123-XXI-3.1.5 MAINTENANCE OF OPERATOR VISUAL AIDS S0123-XXI-3.1.6 SIMULATOR AVAILABILITY REPORTS S0123-XXI-3.1.7 REQUEST FOR SIMULATOR ENHANCEMENTS 1

S0123-XXI-3.1.8 REPORTING OF MINOR HARDWARE DEFICIENCIES S0123-XXI-3.1.9 PROTECTED INITIAL CONDITIONS 1

SO123-XXI-3.1.10 SPECIAL TEST PROCEDURE GUIDE LINES CONFIGURATION MANAGEMENT S0123-XXI-3.2.1 TROUBLE REPORTS 2

S0123-XXI-3.2.2 SIMULATOR WORK ORDERS 1

S0123-XXI-3.2.3 REVIEW OF PLANT MODIFICATIONS 3

S0123-XXI-3.2.4 SIMULATOR DESIGN CHANGE PACKAGES 4

S0123-XXI-3.2.5 MAINTENANCE OF SIMULATOR DRAWINGS S0123-XXI-3.2.6 MAINTENANCE OF MALFUNCTIONS 2

S0123-XXI-3.2.7 SCOPING OF SIMULATOR MODIFICATIONS S0123-XXI-3.2.8 MAINTENANCE OF REMOTE FUNCTIONS S0123-XXI-3.2.9 CONFIGURATION MANAGEMENT SYSTEM 1

MODIFICATIONS SIMULATOR PERFORMANCE TESTING S0123-XXI-3.3.1 REAL TIME SIMULATION TEST 1

S0123-XXI-3.3.2 STEADY STATE TESTING 4

S0123-XXI-3.3.3 CORE PHYSICS TESTING

.1 S0123-XXI-3.3.4 SURVEILLANCE TESTING 2

S0123-XXI-3.3.5 NORMAL OPERATIONS TESTING 2

S0123-XXI-3.3.6 TRANSIENT TESTING 2

S0123-XXI-3.3.7 MALFUNCTION TESTING 1

S0123-XXI-3.3.8 RECORDING SIMULATOR DATA HARDWARE VERIFICATION S0123-XXI-3.4.1 HARDWARE VERIFICATION 3

Used for certification.

2.0 FACILITY DESCRIPTION SONGS 1 is a three-loop Westinghouse pressurized water reactor rated at 1347 Megawatts thermal and approximately 450 Megawatts electric and has been in commercial operation since January 1968.

CAE Electronics Limited of Montreal, Canada, manufactured the SONGS 1 Control Room Training Simulator. The factory acceptance testing was completed in January 1992.

The simulator will be released by CAE for operator training in February 1992.

The training simulator duplicates the SONGS 1 control room. The simulation facility consists of simulated vertical control boards, the central operating console, and two panels located behind the boards: nuclear instrumentation rack, and a miscellaneous panel (see Figure No. 1 and Table II).

The control room environment is simulated with temporary lighting, and the following CAE-provided furnishings and communication equipment: Control operator's desk with a phone, an intercom to the Instructor Facility, and a Control Room Supervisor's desk with a PC computer.

The simulation computer system complex consists of three personal computer-based Instructor Facilities (I/F) and two Silicon Graphics (SGI) Model 240 UNIX-based computers, each with 4 Central Processing Units (CPU).

One SGI and one I/F are sufficient to run the simulator. The other computers are provided for backup and development.

All systems that have parameter displays and controls in the SONGS 1 control board are functionally simulated.

This is achieved by providing adequate controls, instrumentation, alarms, and man-machine interfaces to conduct the normal plant evolutions and malfunctions required by ANSI/ANS 3.5-1985, Sections 3.1.1 and 3.1.2.

2.1 SIMULATOR CONTROL ROOM SONGS 1 Control Room. The SONGS 1 Control Room area houses the control panels for SONGS 1. The main vertical control boards form three sides of a rectangular-shaped area, with a central operating console (J-console) in the middle which contains the most frequently used controls and indications.

Behind the control boards are instrumentation racks, relay panels, and power supplies.

(See Figure No. 2 and Table II).

A Control Operator's desk and Control Room supervisor's desk are located within the U-shaped portion of the unit. Across from the open portion of the U-shaped panels are the Technical Support Center HVAC panel, file cabinets, Shift Superintendent's office and windows looking into the Technical Support Center, (refer to Figure 2).

Behind the main control panels are additional panels which are referred to as "back panels."

Table II lists the SONGS 1 panels and cabinets within the main control room area.

Simulator Control Room. The simulator Control Room includes the main control boards, operator's desks, nuclear instrumentation panel and miscellaneous panel as illustrated in Figure No. 1.

Table II also provides a cross-reference between the simulator and the SONGS 1 control board panels. Most hardware provided on the simulator main control boards is functionally simulated.

Some instrumentation is pictorially simulated. These instruments and justification for their use are provided in Section 2.5, "Hardware Verification and Overall Fidelity".

The simulator's control boards physical layout duplicates the SONGS 1 control boards layout. All displays, indications, and controls that are required for normal, abnormal, and emergency evolutions are functionally simulated.

The arrangement of the simulator back panels duplicates the plant's location only to the extent of Cabinet C115, "Nuclear Instrumentation System", (simulator panel A15).

Simulator Panel A16, Miscellaneous Panel, consolidates several back panel controls and indications that are located in various back panels in the plant. This consolidation results in these instruments being in different locations than they are in the plant. During operator training, no problems were noted with this configuration. This assessment is based on feedback received from operators. Therefore, the arrangement of the back panels is not considered an exception.

Communications System and Audible Cues. The communications system presently used for the simulator is made up of a closed circuit telephone system that is confined to the simulator area and of the CAE Electronics internal telephone system.

Exceptions. At the temporary location of the simulator in Montreal, the communications system is different in function and appearance from that in the SONGS 1 control room. An intercom between the simulator control room and the instructor station substitutes for the control room radio.

A single telephone, located on the control operator's desk can be used to call one of two phones at the instructor station. This phone system is part of the CAE Electronics Ltd. phone system and uses specific numbers that are different from those used at SONGS.

Alarms and Audible Cues. The simulator annunciator audible alarms have been tuned to provide the same audio cues as in the SONGS 1 Control Room. Other audible cues originating from within the control room are provided by utilizing the same equipment as the plant whenever possible. Pneumatic instruments were replaced with electrical equivalents in the simulator and no pneumatic sounds are reproduced. It has been determined that there are no pneumatic audible cues used by the operators in the operation of the plant. While the SONGS 1 control room is exposed to external sounds, it was determined during specification and purchasing of the simulator that there were sufficient other cues to the operators that external sound cues were not necessary. This determination has been validated during the testing of the simulator.

Visual Cues and Information. The control boards use the same color scheme, demarcations, and labeling as the SONGS 1 control room. The annunciator's color coding, flash rates, and layout are simulated. The operator visual aids provided in the control room are also provided in the simulator. The visual aids are reviewed on a routine basis, using procedure S0123-XXI-3.4.1, "Hardware Verification,"

to keep the simulator consistent with the.plant.

Exceptions. The desks and consoles, while in the same relative location as the plant, are standard office furniture and, therefore, different from that in the plant.

The control room lighting, location, intensity, and response to loss-of-power do not exactly replicate the plant control room. The lighting fixtures are, however, similar in nature and of the same type (i.e. fluorescent) as used in the plant.

In addition, a similar but different loss of power cue is provided. Carpeting is of the same basic color as in the control room but of a different shade. The simulator review board has reviewed these exceptions and determined that the operator response and interface with the simulator will not be affected.

2.2 INSTRUCTOR STATION Instructor Controls.

The instructor console is located on the operating floor of the simulator at the open end of the "U" formed by the panels. This location provides a view of the control room and the main control board panels. The instructor consoles include controls and indications needed to conduct training/testing exercises from cold shutdown conditions, with the reactor coolant loops filled, to 100% power operation. The instructor console provides three Instructor Facilities (I/F) whose monitors display tables containing the simulator's initial conditions, malfunctions, remote functions, monitored parameters and their summaries.

In addition the Instructor Facility includes system schematics, graphic recording, trainee action monitoring, and lesson plan capabilities.

North Vertical Boards-----

West A 2 East A13 A57 North[

57 J Console Mid A3 A40 INSTRUCTOR STATION AREA A10 A9 UNIT 1 CONTROL ROOM West Mid East A6 A5 A4 A7 South Vertical Boards Niote:

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TABLE II PANELS AND CABINETS WITHIN MAIN CONTROL ROOM CAE SONGS1 Sub-Section No.

Panel No. Panel No.

Description J-CONSOLE Al C04 Turbine, Generator, and Secondary systems A2 C03 Feedwater, Reheaters, and Boric Acid Controls A3 C03 Pressurizer, CVCS, Steam Dump, and Rod Control Systems SOUTH VERTICAL BOARDS A4 C100 Control Room Fire Panels F77, F78', F79 A5 C41 Diesel Generator No.1 Control Panel A6 C42 Diesel Generator No.2 Control Panel A7 C12 East Switchyard Controls and Mimic Bus A7 C12 Mid Electrical Controls and Mimic Bus A7 C12 West Reheater Dump Controls, Electrical and Turbine Recorders WEST VERTICAL BOARDS A8 C1l KO3A Annunciator Reset, OES Alarms, and SV-99 Controls A9 Cl Radiation Monitoring Panel A10 C10 NIS Misc. Panel; Source Range Audio Count Rate, Axial Offset Indication and IR Recorder All C05 South Post Accident Radiation Monitoring Panel All C05 Mid Recorder Board All C09 North Nuclear Control Auxiliary Panel; RHR, CCW, CSAS NORTH VERTICAL BOARDS A12 C09 West Nuclear Control Auxiliary Panel; SI, PZR, VCC, LDS, BAS A12 C09 Mid Nuclear Control Auxiliary Panel; PZR, VCC, LDS, BAS, RCS, RCP, A12 C13 East Auxiliary Equipment Control Panel; NIS, CND, SWC, CWS, ISA and CIS subpanels C13A and C13B A13 C71 Auxiliary Feedwater Control Panel A14 C89 Temperature Averaging Cabinet TABLE II (continued)

REAR PANELS AND RACKS Al5 C115 Nuclear Instrumentation System Rack A16 C116, C117 Miscellaneous Panel; Recorders, Defeat Switches, Trip Resets and Rod Disconnects, NIS Coincidentor A Cabinet, NIS Coincidentor B Cabinet CONTROL ROOM AREA PANELS AND RACKS THAT ARE NOT MODELED C-79 TSC HVAC Panel C-80 Meteorological Data Recorder Panel C-72 Universal Field MultiPlexer Panel (Fox3 computer interface)

C-61 SEP Relay Panel"B" C-16 Vital 120VAC Power Panel C-60 SEP Relay Panel"A" Y-33 Vital Bus 3A 120VAC Power Panel C-08 A,B,C Incore Flux Mapping Panels C-32 CWTMS & Reactivity Computer Rack C-28A Radwaste Slave Annunciator Panel C-27 Reactor Coolant System, Rack R5 C-29 Pressurizer Control Racks R3,R4 C-26 Reactor Protection & Control Racks R1, R2 C-23 Auxiliary Coolant & SI Rack R7 C-25 CVCS Auxiliary Rack R6 C-36 Rod Deviation Rack R15 C-37 Axial Offset Rack R16 C-69 AFW Logic Cabinet, Rack "Al" C-24 Steam Generator Auxiliary Racks R10, R11 C-15 Rod Position Relay Rack R8, R9 C-22 Miscellaneous Relay Rack R12, R13 C-14 Voltage Regulator Rack R14 K-02 Supplemental Annunciator Cabinet C-39 Boron Concentration Measurement System Control C-70 AFW Logic Cabinet, Rack "Bl" C-62 CSAS Logic Cabinet C-121 Undervoltage Relay Panel C-122 Thermal Shield Monitoring Rack Y-29 Vital Bus 5 120VAC Power Supply Panel Y-02 CSAS Invertor Cabinet C-19 Auxiliary Relay Panel -

South C-74 Hydrogen Recombiner Control Panel Train "B" C-73 Hydrogen Recombiner Control Panel Train "A" S02 Sequencer No.1 S03 Sequencer No.2 Initial Conditions. The simulator has 70 initial conditions (IC) available to the instructor. Some ICs are password protected.

The balance of the ICs are normally used for either special training or problem identification and troubleshooting. One IC is a special IC used for snapshot capability. A set of 24 IC sets will be established for training use and will be available prior to commencement of training. These will be maintained for operator training and testing and a list will be kept at the Instructor Station. These 24 ICs will always be available for the instructor to use for training and testing, reflecting a variety of plant conditions, fission product poison con centrations, and reactor core life. The ICs are being established and will be updated and modified as necessary using Simulator Certification procedure S0123-XXI-3.1.9, "Protected Initial Conditions."

Remote Functions. Instructor Station remote functions are used to simulate equipment and controls located outside the control room. The simulator has the necessary remote functions to allow a plant start-up from cold shutdown to 100% power, reactor trip recovery, and shutdown to cold conditions. Remote functions are also used for transferring control systems, resetting safety system actuations, racking out breakers on major equipment, and other required functions that cannot be done from the control boards.

Malfunctions. The simulator has 187 specific malfunction features to simulate abnormal and emergency conditions including the 95 malfunctions that meet ANSI/ANS 3.5-1985 requirements.

Additional malfunctions (beyond those required by the ANSI/ANS standard) are provided that have been identified from operational experience and training needs. Table III provides a cross reference list of the ANSI/ANS required malfunctions to the simulator malfunction identification codes identified in Table V, "Simulator Malfunctions."

The malfunctions can be inserted either sequentially or simultaneously from the Instructor Facility and can be pre-programmed and initiated with timers, or inserted with a remote hand-held unit. The degree of severity for some malfunctions is variable. A list of all malfunctions that may be used in operator testing is kept at the Instructor Station.

Other Special Features.

The simulator can "freeze" the simula tion, operate in slow time, and backtrack. Certain functions can operate in fast time, i.e., Xenon transient, heat-up, and establishing condenser vacuum. The simulator has a switch check system to assure that all switches and controllers are properly aligned for the initial condition set. A Panel Test is performed daily to test light bulbs and calibrate the analog outputs to the recorders and meters.

2.3 OPERATING PROCEDURES AND OPERATOR AIDS Operating Procedures. The procedures used while training on the simulator are the same procedures used in the SONGS 1 Control Room. A controlled set of all SONGS 1 operating procedures is kept in the simulator. However, when a change is made on the simulator either prior to or after it is incorporated in the plant, temporary procedures and/or operating aids will be used during the transition period.

Operator Aids Used. A set of controlled electrical diagrams and piping & instrument drawings (P&IDs) is kept in the simulator.

Other visual aids used in the control room to assist operators in the performance of their duties have also been incorporated into the simulator facility.

Plant Physics Data Book. The SONGS 1 Plant Physics Data Book is updated after each refueling. However, the simulator core model's database will not be updated with each refueling of SONGS 1, since the core response to transients and normal operating maneuvers does not change significantly from cycle to cycle. Thus, updating the core model after each refueling would not significantly improve training on the simulator. The core physics data in the-simulator Plant Physics Data Book will match the core model's database.

2.4 SIMULATOR BASELINE DATA Table IV lists the various documents and reports that form the baseline for the simulator. This listing follows the guidelines of ANSI/ANS 3.5-1985. The list will be revised as needed to ensure the documented record of the present simulator configura tion is up to date.

Differences between'the simulator and the actual plant are documented in Simulator Work Orders (SWO) and CAE Electronics' Deficiency Report Management System.

2.5 HARDWARE VERIFICATION AND OVERALL FIDELITY A hardware verification was conducted using Simulator Certification procedure S0123-XXI-3.4.1 "Hardware Verification,"

(Appendix Q) to verify that the instrumentation, controls, markings, and operator aids on the panels and consoles replicate the size, shape, color, and configuration of the SONGS 1 Control Room. The process compared the physical front panel features with the SONGS 1 control room panels.

All deviations identified during the comparison were documented in the SONGS 1 Simulator Hardware Database.

Then, an assessment was made to determine if the deviation could either impact the operator actions or detract from training.

Minor differences between the simulator and the SONGS 1 Control Room exist. The most significant involve using photographs in place of actual instruments. Twenty one photographs are used.

Fifteen of these involve electric metering not used by the operators. Three are fire protection monitoring panels. These are back-ups to the station fire protection and are not used by the operators for control room response or operation. The remaining three involve the remote plant area post accident sampling system. These are used for long term post accident monitoring in areas not used by operators and have no operator action associated with them.

Other minor differences exist such as meter-scale divisions and name tags.

Experience has shown that these differences impact neither operator training nor any of the operator actions taken during training evolutions.

The overall fidelity of the simulator is expected to be maintained and improved with the assistance of periodic student surveys.

2.6 ACCEPTABILITY FOR OPERATOR TESTING The SONGS 1 Control Room Training Simulator has completed factory acceptance testing and is currently being used to validate training scenarios and NRC examination bank questions. During the period of Factory Acceptance testing (a period of four months),

licensed operators and training instructors were used to test the simulator. The tests were conducted using approved plant procedures and drawings.

Deficiencies were noted and corrected or given a priority for disposition in accordance with Simulator Verification procedure S0123-XXI-3.2.2, "Simulator Work Orders".

Based on the test results and the assessment of these operators and instructors, the simulator has demonstrated that it is acceptable for conducting operator tests.

Following completion of factory acceptance testing the simulator has been upgraded to incorporate plant modifications that could not be incorporated during its manufacture. This insures that the simulator is matched as closely as practical to SONGS 1.

TABLE III is ANSI 3.5 REQUIREMENT vs MALFUNCTION NUMBER UNIT 1 SIMULATOR ANSI 3.5 REQUIREMENT SIMULATOR MALFUNCTION t Ola Loss of Coolant 1.6 Significant S/G leaks Olb Loss of Coolant 1.1 1.2 Inside Containment Olc Loss of Coolant 1.4 Outside Containment Old Loss of Coolant 1.1 1.2 Large RCS breaks Ole Loss of Coolant 1.1 1.2 3.10 Small RCS breaks Olf Loss of Coolant 3.2 Demonstration of saturated conditions Olg Loss of Coolant 3.2 3.4 Failure of safety and relief valves 02 Loss of instrument air 22.1 22.2 03a Loss of off site power 16.1 03b Loss of emergency power 16.4 16.6 16.9 03c Loss of emergency generators 16.7 16.8 16.9 03d Loss of power to plant's electrical 16.4 16.6 16.11 distribution buses.

03e Loss of power to individual instrumentation 16.12 16.13 buses. (AC as well as DC) 04 Loss of forced core cooling flow 2.1 2.2 2.4 (single or multiple pump failures) t See Table V for description of each malfunction listed in this column

-17

TABLE III (continued) 0 ANSI 3.5 REQUIREMENT SIMULATOR MALFUNCTION 05 Loss of condenser vacuum 19.1 05a Loss of condenser level control 19.2 20.1 06 Loss of service water 10.1 (Saltwater Cooling) 07 Loss of shutdown cooling 10.2 8.1 8.2 (Loss of RHR) 08 Loss of component cooling system 10.2 10.3 09 Loss of normal feedwater or normal 21.3 21.7 21.10 feedwater system failure 21.11 21.12 21.13 10 Loss of all feedwater (normal and 21.3 11.1 11.5 emergency) 11 Loss of protective system channel 12.7 12a Control rod failures 14.1 14.6 Stuck rod 12b Control rod failures 5.1 Uncoupled rod 12c Control rod failures 14.2 14.3 Drifting rods 12d Control rod failures 14.5 Rod drops 12e Control rod failures 5.1 14.2 14.3 14.6 Misaligned rods S-18 -

TABLE III (continued)

ANSI 3.5 REQUIREMENT SIMULATOR MALFUNCTION 13 Inability to drive rods 14.1 14.7 14 Fuel cladding failure 1.7 15 Turbine trip 12.5 16 Generator trip 15.1 17 Failure in automatic control systems that 14.1 14.2 14.3 17.6 affect reactivity and core heat removal.

1.8 1.9 18 Failure of reactor coolant pressure and 3.5 3.7 volume control systems 19 Reactor Trip 12.1 20a Steam and feed line breaks 17.1 17.2 Inside containment main steam line break 20b Steam and feed line breaks 21.10 21.11 Inside containment feed line break 20c Steam and feed line breaks 17.3 Outside containment steam line break 20d Steam and feed line breaks 21.12 21.13 Outside containment feed line break 21-Nuclear instrument failures 13.2 -

13.8, inclusive 22 Process instrument, alarms, and control 1.11 1.12 3.5 3.6 system failures 3.7 3.8 4.1 4.2 4.4 4.5 6.14 6.16 6.17 17.6 17.7 18.14 23 Passive malf. in sys. such as engineered 1.4 7.3 7.5 7.7 safety features, emergency feedwater sys 11.3 11.4 24 Failure of auto reactor trip 12.1 12.6 25 Others -

Accidents analyzed in FSAR 21.8 18.2 17.4 17.5 15.8 2.2 2.4 1.9 6.17 1.10 TABLE IV BASELINE DATA Document/Report Description Plant P&ID's Controlled Documents maintained by SCE's Corporate Documentation Management (CDM)

Plant Electrical Dwgs.

Controlled Documents maintained by CDM Plant Procedures Controlled Documents maintained by CDM Core Physics Book Cycle 10 Data Plant DCP File Complete listing of Design Change Packages (DCP) reviewed by the Simulator Support Group.

VARIOUS WCAP's Westinghouse transient and accident analysis documents SONGS 1 UFSAR Updated Final Safety Analysis Report (UFSAR) for SONGS 1 Station Design Manual Plant design information (i.e. pump curves, system characteristics etc.)

Plant performance data Collection of data obtained from various plant evolutions and tests.

Simulator drawings Drawings used to reflect the software representation of the various plant systems Instrument loop drawings Controlled Documents maintained by CDM Simulator test program Data collected during various simulator tests data Station Incidents LERs, Station Incident reports, and various other reports of plant incidents Instrument Calibration Calibration results for plant instruments Data Sheets Technical/Vendor Manuals Manuals containing technical/operating information for plant equipment System Descriptions Controlled Documents maintained by CDM 3.0 SCOPE AND LIMITS OF SIMULATION The simulator models provide sufficient detail and scope to allow the conduct of normal plant evolutions and malfunctions as required by ANSI/ANS 3.5-1985, Sections 3.1.1 and 3.1.2, except as noted in Section 1.3 of this report. The simulator provides the control room operators with a number of sufficient visual, audio, control, and indication inputs to maneuver the plant using normal and emergency operating procedures.

Only those functions which are either not performed in the control room, or are outside the operator's normal job scope, or functions that do not enhance the training process, are considered outside the scope of the simulation.

The instructor is alerted that the "limits of the simulation may be exceeded," when certain key parameters exceed their preset alarm setpoints.

A light on the instructor's console indicates when this condition exists. The key parameters that are used to alert the instructor whether the simulation has gone beyond its limits are:

1.

Any component or system simulation model or other software being driven beyond its intended range of applicability.

2.

Primary containment pressure greater than design limit.

3.

Primary system (reactor) pressure greater than design limit.

4.

Fuel temperature/time history indicative of gross fuel failure.

After being alerted, the instructor may stop the training scenario or take other appropriate actions.

4.0 PERFORMANCE TEST ABSTRACTS This section presents performance test abstracts for those tests required by ANSI/ANS 3.5-1985. Additional tests were also performed but are not in this report. Deficiencies discovered during performance of tests required to meet the ANSI requirements are noted in the abstracts. A summary of all deficiencies is included in Section 4.8, "Deficiency Status."

Initial Certification Tests. The following tests were performed to support the initial simulator certification:

• Real-Time Test This test verifies that sufficient computer spare time is available, and that the models execute within their allotted time frames.

An additional verification of real-time simulation is the comparison of the simulator transients against plant data. The real-time simulation test uses Simulator Verification Procedure SO123-XXI-3.3.1 "Real Time Simulation Test" (Appendix J).

• Steady-State Test This test demonstrates that the steady-state values of key parameters meet the stabilitycriteria of ANSI/ANS 3.5-1985.

Any parameter that did not meet the test criteria has been either corrected, or justified and approved by the Simulator Review Board (SRB).

The steady-state test uses Simulator Verification Procedure S0123-XXI-3.3.2 "Steady-State Testing" (Appendix K).

• Normal Operations Tests These tests demonstrate the ability of the simulator to perform the ANSI/ANS 3.5-1985 required normal evolutions.

The tests are described in Simulator Verification Procedures S0123-XXI-3.3.4 "Surveillance Testing," and S0123-XXI-3.3.5 "Normal Operations" (Appendix N).

• Malfunction Tests These tests verify the response of the simulator's 187 malfunction features. Those malfunctions with unacceptable responses have Deficiency Reports written and prioritized and are restricted from use until the response is corrected.

The tests use Simulator Verification procedure S0123-XXI 3.3.7 "Malfunction Testing" (Appendix P).

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• Transient Tests The transient tests verify that the simulator response to all of the ANSI/ANS 3.5-1985 required transients is acceptable. Two types of transient tests are performed:

benchmark tests and annual operability tests. The benchmark transient test results are compared against either actual plant data, engineering code analyses, or best estimates.

The simulator is compared to plant data whenever possible.

The annual transient operability tests compare the yearly generated data against the baseline data collected following benchmark testing. The tests are conducted and reviewed as described in Simulator Verification Procedure S0123-XXI 3.3.6 "Transient Testing" (Appendix 0).

• Surveillance Tests The surveillance tests are performed to demonstrate that plant surveillance procedures for safety related systems can be conducted on the simulator. The surveillance procedures were selected on the basis of their importance to safety, operator training and testing, and the ability to complete the procedure from the control room. Surveillances on plant equipment outside of the scope of the simulator are not performed. The tests use Simulator Verification Procedure S0123-XXI-3.3.4 "Surveillance Testing" (Appendix M).

• Core Physics Tests The tests demonstrate that the core physics parameters closely match the Plant Physics Data Book, and ensure that the simulator response is similar to the SONGS 1 reactor core. The simulator core model is updated whenever there are any major differences between the SONGS 1 current core physics data and the simulator core physics test results.

The core physics tests use Simulator Verification Procedure S0123-XXI-3.3.3 "Core Physics Testing" (Appendix L).

Re-Certification Performance Testing Program. The four-year re certification program will follow the testing schedule given in Simulator Verification procedure S0123-XXI-3.1.2 "Performance Test Scheduling" (Appendix B).

4.1 REAL-TIME TEST ABSTRACT Verification of real-time simulation was performed according to procedure S0123-XXI-3.3.1, "Real Time Simulation Test."

The following tests were conducted as part of the real-time simulator execution test procedure:

Performance Test Test Date

1.

SGI 4D/240 CPU 0 9/12/91

2.

SGI 4D/240 CPU 1 9/12/91

3.

SGI 4D/240 CPU 2 9/12/91

4.

SGI 4D/240 CPU 3 9/12/91 The 9/12/91 processor timings were done for a Hot Full Power Steady State condition, for a Steam Line Break and for a Hot Leg LOCA transient condition.

Testing Requirements. According to the ANSI/ANS-3.5-1985 standard, real-time simulation is defined as the capability to simulate the power plant dynamics according to the same time base relationships, sequences, durations, rates and accelerations which actually hold for the reference plant. Real time program models which simulate the different power plant systems execute at their designed cycle rate and each model must complete its execution within its allotted time interval.

For the SONGS 1 simulator, the models are scheduled for execution within 50 millisecond interval frames at a rate of 5, 10 or 20 executions per second. The timings performed indicate the following computer processing load requirements due to the real-time power plant system models:

STEADY STATE TRANSIENT SGI 40D/240 CPU 0 16.7%

16.4%

SGI 40D/240 CPU 1 47.0%

46.8%

SGI 40D/240 CPU 2 53.7%

58.2%

SGI 40D/240 CPU 3 55.5%

55.4%

The above results indicate that sufficient computer processing power is available to guarantee execution of every plant system computer model in real-time.

In addition, no overrun was observed on any of the simulator computer processors, since simulation execution would have been automatically alarmed on an overrun condition.

Noted Deficiencies.

No deficiencies to the ANSI/ANS-3.5-1985 standard were noted.

Section 2.4 of Attachment 1 to procedure S0123-XXI-3.3.1 requires that:

(a) no model exceeds real time and (b) the spare time percentage for each processor is 10% or more (100 milliseconds/second).

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As shown above the real time tests performed indicate that no real-time model exceeds the overtime criteria in any of the simulator computers.

The measured Spare Time available in each of the simulator computer processors is the following.

STEADY STATE TRANSIENT SGI 40D/240 CPU 0 83.3%

83.6%

SGI 40D/240 CPU 1 53.0%

53.2%

SGI 40D/240 CPU 2 46.3%

41.8%

SGI 40D/240 CPU 3 44.5%

44.6%

Conclusion.

A computer real-time test as recommended by Appendix A of the ANSI/ANS-3.5-1985 standard was satisfactorily performed according to the REAL TIME SIMULATION TEST procedure S0123-XXI 3.3.1.

This procedure verifies that spare time is available and the processors execute the code in real-time.

An additional verification of real-time simulation was provided by benchmarking the simulator transients against plant data or accident codes.

The parameter time responses were compared with those of the simulator and verified to be consistent.

The test results were presented and approved by the SONGS 1 Simulator Review Board.

4.2 STEADY-STATE TESTS ABSTRACT Introduction.

Steady state testing of the SONGS 1 simulator was conducted in accordance with Simulator Verification Procedure S0123-XXI-3.3.2 and Acceptance Test Procedure SO1-XXI-3.1.N2A.

Testing demonstrated that the simulator steady state values for specified parameters during full power conditions are stable according to the criterion that they do not vary by more than 2%

above or below their initial values over a 60 minute period.

Further testing demonstrated that the specified parameter values computed by the simulator meet the acceptance criteria of ANSI/ANS 3.5-1985 at three different power levels.

The following is a list of the steady state tests performed under S0123-XXI-3.3.2:

SS01 One Hour Stability Test Tested 11/20/91 SSO2 Full Power Steady State Test Tested 12/17/91 SSO3 70% Power Steady State Test Tested 12/17/91 SSO4 30% Power Steady State Test Tested 12/17/91 Testing Requirements. These tests were performed to meet the requirements of ANS 3.5 section 4.1, which specifies in part that the simulator computed values for steady state full power operation shall be stable and not vary more than + 2% of the initial values over a 60 minute period.

ANS 3.5 section 4.1 further specifies the following. The accuracy of computed values shall be determined for a minimum of three points over the power range. The simulator computed values of critical parameters shall agree within + 2% of the reference plant parameters and shall not detract from training. The calculated values of noncritical parameters shall agree within +

10% of the reference plant parameters and shall not detract from training. Examples of critical parameters are listed in ANS 3.5, Section 4.1.

ANS.3.5 Appendix B, Simulator Operability Tests, provides specific information for the conduct of these tests in section B2.1, Steady State Performance, including the specific parameters to monitor.

Deficiencies.

Deficiencies are itemized, discussed and justified below.

Conclusion. All four of the steady state performance tests identified in SO123-XXI-3.3.2 were performed satisfactorily.

Deficiencies are evaluated and justified below.

All deficiencies are acceptable. No deficiency detracts from training.

Deficiency Evaluation.

SS01, One Hour Stability Test There were no deficiencies during the One Hour Stability Test.

SS02, Full Power Steady State Test Plant Description Plant Simulator Allowed Tag Value Value Deviation NIS-1205 Nuclear Flux 91.5%

88.0%

2.4%

NIS-1206 Nuclear Flux 92.0%

87.0%

2.4%

NIS-1207 Nuclear Flux 91.5%

87.5%

2.4%

NIS-1208 Nuclear Flux 92.5%,

87.5%

2.4%

TR402 T-cold Loop C 528*F 588*F 100 F R8(PT2)

Steam Pressure 555 psig 510 psig 30 psig LT-455 S/G C N/R Level 32%

29.5%

2%

FT-457 S/G B Feed Flow 1640 klb/hr 1700 klb/hr 50 klb/hr FT-460 S/G A Steam Flow 1710 klb/hr 1640 klb/hr 50 klb/hr FT-462 S/G C Steam Flow 1580 klb/hr 1680 klb/hr 50 klb/hr JUSTIFICATION Nuclear Flux The simulator values for nuclear flux are the maximum obtainable with turbine governor valves wide open and T-ave equal to T-ref.

Corrective action has been requested of the simulator vendor, CAE Electronics Ltd. under Deficiency Report Number (DR) 1286.

T-cold Loop C This recorder pen was faulty at the time of the test and read high during all three steady state tests.

DR 1290 documents this deficiency.

Steam Pressure This recorder pen was faulty at the time of the test and read low during all three steady state tests.

DR 1288 documents this deficiency.

S/G C Narrow Range Level Plant data show a 1% level spread among loops.

Therefore plant instrument error is at least + 0.5% level.

Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

S/G B Feed Flow The S/G B feed flow recorder pen was faulty at the time of the test and read high during all three steady state tests.

DR 1289 documents this deficiency.

S/G A Steam Flow Plant data show a 130 kilopound per hour (klb/hr) steam flow spread among loops. Therefore plant instrument error is at least

+ 65 klb/hr. Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

S/G C Steam Flow The S/G C steam flow recorder pen was faulty at the time of the test and read high during all three steady state tests.

DR 1289 documents this deficiency'.

SSO3, INTERMEDIATE POWER STEADY STATE TEST Plant Description Plant Simulator Allowed Tag Value Value Deviation none Core Thermal Power 70.3%

74.5%

2.4%

TR402 T-cold Loop B 520.F 530.2.F 10.F TR402 T-cold Loop C 520'F 596*F 10*F LI-431 Pressurizer Level 31.5%

34%

2%

R8(PT2)

Steam Pressure 620 psig 578 psig 30 psig

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Plant Description Plant Simulator Allowed Tag Value Value Deviation FT-456 S/G A Feed Flow 1250 klb/hr 1330 klb/hr 50 klb/hr FT-457 S/G B Feed Flow 1200 klb/hr 1380 klb/hr 50 klb/hr FT-458 S/G C Feed Flow 1200 klb/hr 1330 klb/hr 50 klb/hr FT-460 S/G A Steam Flow 1250 klb/hr 1350 klb/hr 50 klb/hr FT-461 S/G B Steam Flow 1200 klb/hr 1300 klb/hr 50 klb/hr FT-462 S/G C Steam Flow 1150 klb/hr 1360 klb/hr 50 klb/hr JUSTIFICATION Core Thermal Power Core thermal power is calculated from plant and simulator data using the Reactor Thermal Power Calibration surveillance test, where displayed values of feedwater flow are used. Simulator core thermal power is greater than plant core thermal power because simulator feedwater flow is greater than plant feedwater flow. This deficiency will not detract from training because the operator has no display of core thermal power and would use calculated feed flow differential pressure rather than displayed feed flow rates to perform a Reactor Thermal Power Calibration.

T-cold Loop B The accuracy of recorder TR402 is at least + 0.5 *F. Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

T-cold Loop C As per test SSO2.

DR 1290 was written to correct this deficiency.

Pressurizer Level Plant data show a spread of 2.0 % level among the three level channels. Therefore the instrument error is at least + 1.0 %

level.

Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

Steam Pressure As per test SSO2.

DR 1288 was written to correct this deficiency.

S/G A Feed Flow Plant data show a 100 kilopound per hour (klb/hr) feed flow spread among loops at full power. Therefore instrument error is at least + 50 klb/hr. Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

S/G B Feed Flow As per test SSO2.

DR 1289 was written to correct this deficiency.

S/G C Feed Flow The S/G C feed flow recorder pen was faulty at the time the test was performed and read high for both tests SS03 and SS04.

DR 1289 was written to correct this deficiency.

S/G A Steam Flow As explained in the justification for test SSO2, instrument error is at least + 65 klb/hr. Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

S/G B Steam Flow The S/G B steam flow recorder pen was faulty at the time the test was performed and read high for both tests SS03 and SS04.

DR 1289 was written to correct this deficiency.

S/G C Steam Flow As per test SS02.

DR 1289 was written to correct this deficiency.

SS04, LOW POWER STEADY STATE TEST Plant Description Plant Simulator Allowed Tag Value Value Deviation none Core Thermal Power 31.1%

35.0%

2.4%

TR402 T-cold Loop C 530*F 590'F 10*F LI-430 Pressurizer Level 23.0%

26%

2%

LI-431 Pressurizer Level 20.5%

27%

2%

LI-432 Pressurizer Level 23.0%

27%

2%

YR401-23 Control Bank II 290 steps 203 steps 35 steps R8(PT2)

Steam Pressure 750 psig 700 psig 30 psig LT-453 S/G A N/R Level 40.0%

30%

2%

LT-454 S/G B N/R Level 39.0%

30.5%

2%

LT-455 S/G C N/R Level 40.0%

29.5%

2%

FT-457 S/G B Feed Flow 510 klb/hr 650 klb/hr 50 klb/hr FT-458 S/G C Feed Flow 480 klb/hr 550 klb/hr 50 klb/hr FT-461 S/G B Steam Flow 470 klb/hr 540 klb/hr 50 klb/hr FT-462 S/G C Steam Flow 470 klb/hr 600 klb/hr 50 klb/hr JUSTIFICATION Core Thermal Power Core thermal power is calculated from plant and simulator data using the Reactor Thermal Power Calibration surveillance test, where displayed values of feedwater flow are used. Simulator core thermal power is greater than plant core thermal power because simulator feedwater flow is greater than plant feedwater flow. This deficiency will not detract from training because the operator has no display of core thermal power and would use calculated feed flow differential pressure rather than displayed feed flow rates to perform a Reactor Thermal Power Calibration.

T-cold Loop C As per test SSO2.

DR 1290 was written to correct this deficiency.

Pressurizer Level Programmed pressurizer reference level for a T-ave of 537 'F is 26.5%, with a minimum value of 25.0%. Therefore a pressurizer level transient was underway when the plant data were taken. The simulator values of 26%, 27% and 27% are good values, and this deficiency will not detract from training.

Control Bank II Control bank position varies with dissolved boron concentration, and the 203 step simulator value is a reasonable rod height for 30% power.

It is above the insertion limit of 120 steps.

This deficiency will not detract from training.

Steam Pressure As per test SS02.

DR 1288 was written to correct this deficiency.

S/G Narrow Range Levels Steam generator levels are controlled by the operators in accordance with the general operating instructions, which call for levels to be ramped up to about 40% as power is ramped below about 33%.

Ramping was in progress when the plant data were taken, but not when the simulator data were taken. This deficiency will not detract from training.

S/G B Feed Flow As per test SSO2.

DR 1289 was written to correct this deficiency.

8IG C.Feed Flow As explained in the justification for test SSO3, instrument error is at least + 50 klb/hr. Adding instrument error to the value displayed on the simulator as permitted by ANS 3.5 section 4.1 brings the simulator into the allowed deviation range.

S/G B Steam Flow As per test SSO3.

DR 1289 was written to correct this deficiency.

S/G C Steam Flow As per test SSO2.

DR 1289 was written to correct this deficiency.

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4.3 CORE PHYSICS TESTS ABSTRACT Introduction.

Core physics testing of the SONGS 1 Simulator was conducted in accordance with Simulator Verification Procedure S0123-XXI-3.3.3 and Acceptance Test Procedure SO1-XXI-3.1.N3 to verify that key parameters associated with reactivity in a zero power condition closely correspond to the values measured at the plant and predicted in the Nuclear Design Report. The simulator models the Cycle 10 core of SONGS 1. Plant values measured during Cycle 10 startup in May, 1989, are referenced for the beginning of life (BOL) condition. Reference values for middle of life (MOL) and end of life (EOL) are taken from the Cycle 10 Nuclear Design Report.

Tests were conducted to measure each of the following at BOL, MOL and EOL (tests 5 and 6 were not done at EOL):

1. Boron endpoint, all rods out

2. Isothermal temperature coefficient, all rods out

3. Control rod worths

4. Differential boron worth

5. Boron endpoint, control banks in

6. Isothermal temperature coefficient, control banks in

7. Xenon equilibrium and peak worths In addition, surveillance tests of reactivity calculations demonstrated that reactivity changes from full power to zero power are acceptable for the BOL and MOL conditions.

Testing Requirements.

These tests and measurements were performed to meet the requirements of ANSI/ANS 3.5-1985, section 3.1.1 (9). This section requires that a simulator be capable of conducting core performance testing such as the measurement of reactivity coefficients and control rod worths. Section 3.4.1 requires that a simulator have initialization conditions for various times in core life.

SONGS 1 simulator core physics testing helped to verify that the core models for BOL, MOL and EOL are acceptable for licensed operator training and testing.

Dates of Testing.

BOL testing except for xenon worth was done on October 9, 1991.

BOL xenon worths were measured on November 25, 1991.

MOL testing was done on December 13, 1991.

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EOL testing except for xenon worth was done on October 10, 1991.

EOL xenon worths were measured on December 2, 1991.

Reactivity calculations surveillance testing at BOL was done on December 4, 1991.

Reactivity calculations surveillance testing at MOL was done on December 19, 1991.

Individual Test Abstracts.

1.

Boron endpoint, all rods out The boron concentration in the reactor coolant system (RCS) was noted for an exactly-critical condition at normal operating temperature and pressure with no Xenon, all control rods nearly fully withdrawn, below the point of adding heat. The control rods were then fully withdrawn and the resulting positive reactivity was measured by means of the common data base (CDB) dynamic variable RARHO, total core wide reactivity. This reactivity was converted to a delta-boron concentration by dividing it by the differential boron worth. The delta-boron was added to the actual boron concentration to give the boron endpoint with all rods out.

The acceptance criterion of + 500 pcm from plant Engineering Procedure SO1-V-3.3 was applied. This reactivity value is divided by the plant reference differential boron worth to give a range of acceptable boron endpoints.

Results are as follows, with values in parts per million (ppm):

Simulator Plant Acceptance Reference criterion BOL 1884 1840

+ 78 MOL 1264 1170

+ 73 EOL 567 523

+ 71 The MOL value does not meet the acceptance criterion. CAE Electronics, Ltd. Deficiency Report (DR) No. 1250 was written to correct this deficiency.

2.

Isothermal temperature coefficient, all rods out With the reactor critical as above, a series of temperature swings were made using steam dumps to the condenser to lower the RCS temperature and allowing reactor coolant pump heat input to raise RCS temperature. Reactivity as measured by RARHO was noted, along with the magnitude of the corresponding RCS temperature change as measured by average T-average. Reactivity divided by temperature change yields

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the isothermal temperature coefficient in units of pcm per degree Fahrenheit.

The acceptance criterion of + 3 pcm/*F from plant Engineering Procedure SO1-V-3.4 was applied.

Results are as follows, with values in pcm/0 F:

Simulator Plant Acceptance Reference Criterion BOL

-2.59

-4.88

+ 3 MOL

-7.70

-10.8

+ 3 EOL

-15.8

-16.4

+ 3 The MOL value does not meet the acceptance criterion. DR No. 1250, was written to correct this deficiency.

3. Control rod worths With the reactor critical as above, Control Bank 2 was manually inserted a few steps at a time, with the reactor kept near critical by boron dilution. The change in RARHO with each rod movement was recorded as the reactivity associated with the rod movement. This was continued until Control Bank 2 was fully inserted. The cumulative sum of the reactivities is the integral bank worth. The reactivity associated with the delta-boron in test 1 is included so that an integral bank worth from fully withdrawn can be obtained.

The procedure was repeated for Control Bank 1 with Control Bank 2 fully inserted.

The acceptance criteria of plant Engineering Procedure Sol V-3.5 were applied.

Results were as follows, with values in pcm except where noted in %:

BEGINNING OF LIFE Simulator Plant Difference Acceptance Reference Criterion Control Bank 2 2453 2712

-9.6%

+15%

Control Bank 1 1376 1444

-4.7%

+15%

TOTAL 3829 4156

-7.9%

+10%

MIDDLE OF LIFE Simulator Plant Difference Acceptance Reference Criterion Control Bank 2 2373 2829

-16.1%

+15%

Control Bank 1 1436 1612

-10.9%

+15%

TOTAL 3809 4441

-14.2%

+10%

END OF LIFE Simulator Plant Difference Acceptance Reference criterion Control Bank 2 2482 2850

-12.9%

+15%

Control Bank 1 NOT MEASURED

+15%

TOTAL

+10%

The MOL values for Control Bank 2 and TOTAL do not meet the acceptance criteria.

DR No. 1252 was written to correct this deficiency.

Control Bank 1 was not measured at EOL. This is acceptable based on the satisfactory results for bank 1 at BOL and pending completion of DR No. 1252.

4.

Differential boron worth Differential boron worth was obtained by dividing the total control bank worth obtained in test 3 by the difference in boron endpoints obtained in tests 1 and 5, with the reactor just critical in all three tests.

Results are as follows, with values in pcm/ppm:

Simulator Plant Reference BOL 5.38 6.44 MOL 5.61 6.83 EOL 6.16 7.02 By qualitative engineering judgement, the simulator results were deemed unacceptable, and DR No. 1250 was written to correct this deficiency.

5.

Boron endpoint, control banks in For the exactly critical condition at the end of test 3, the boron concentration in the RCS was noted. The control rods were inserted to zero steps and the resulting negative reactivity measured. This reactivity was converted to delta-boron by dividing by the differential boron worth, and the resulting delta-boron was subtracted from the actual 34 -

boron concentration to yield a boron endpoint with the control banks fully inserted.

The acceptance criterion of + 500 pcm from plant Engineering Procedure SO1-V-3.3 was applied. This reactivity value is divided by the measured differential boron worth to give a range of acceptable boron endpoints.

Results are as follows, with values in parts per million (ppm):

Simulator Plant Acceptance Reference Criterion BOL 1174 1195

+ 78 MOL 587 526

+ 73 EOL

-NOT MEASURED-

+ 71 The rods in boron endpoint was not measured at EOL because Control Bank 1 worth was not measured.

6.

Isothermal temperature coefficient, control banks in From the critical condition at the end of test 5, test 2 was repeated to get the isothermal temperature coefficient with control banks inserted.

The acceptance criterion of + 3 pcm/'F from plant Engineering Procedure SO1-V-3.4 was applied.

Results are as follows, with values in pcm/oF:

Simulator Plant Acceptance Reference criterion BOL

-11.35

-13.21

+ 3 MOL

-15.05

-16.5

+ 3 EOL NOT MEASURED -

+

3 The coefficient was not measured at EOL because Control Bank 1 worth was not measured.

7.

Xenon reactivity worths The Xenon reactivity worth at full power steady state was monitored using the CDB variable RRRXT, total core xenon reactivity. The reactor was tripped from full power steady state and the reactor core Xenon transient speedup factor was increased from its normal value of 1.0 to 20 or 24.

The value of RRRXT when it stopped increasing in absolute value was noted. This is the peak Xenon worth.

Results are as follows, with values in pcm. Plant reference values are from the Cycle 10 Nuclear Design Report for 90%

power, with the MOL values interpolated.

BEGINNING OF LIFE Simulator Plant Reference Equilibrium Xenon

-2095

-2097 Peak Xenon

-2462

-2500 MIDDLE OF LIFE Simulator Plant Reference Equilibrium Xenon

-2088

-2149 Peak Xenon

-2457

-2577 END OF LIFE Simulator Plant Reference Equilibrium Xenon

-2196

-2214 Peak Xenon

-2591

-2672 By qualitative engineering judgement, the result of 395 pcm equilibrium-to-peak for the simulator compared to 458 pcm for the plant reference for EOL was deemed unacceptable, and CAE Deficiency Report 1111 was written to correct this deficiency.

Reactivity Calculations Surveillance Tests. Attachment 2, ECP/ECB CALCULATIONS, of S01-12.9-2, Reactivity Calculations, was performed for the BOL and MOL conditions.

The simulator was initialized at full power equilibrium Xenon with control bank 2 at 300 steps and RCS Boron concentrations of 1293 ppm (BOL) and 682.5 ppm (MOL).

These conditions were used for the "previous critical conditions" of Attachment 2.

The reactor was manually tripped and the reactor core speedup factor increased until Xenon was completely decayed. Attachment 2 was completed with "desired critical conditions" of control bank 2 at 120 steps.

The estimated critical boron concentrations and actual critical conditions were as follows:

Estimated Actual Critical Critical Critical Rod Boron Boron Height (ppm)

(ppm)

(steps)

BOL 1587 1584 121 MOL 983 992 120

• ~

These results confirm that the combination of reactivity coefficients in the simulator, including the power coefficient, is such that operators are able to accurately predict reactivity effects during training or examination scenarios on the simulator and that simulator performance is acceptable despite outstanding deficiencies.

Deficiencies.

The following deficiencies were noted during core physics testing of the SONGS 1 simulator.

1. The differential boron worth measured on the simulator is too small in magnitude compared to the plant reference values.

DR No.1250 was written to correct this deficiency.

2. The Control Bank worths measured on the simulator at MOL are too small. DR No.1252 was written to correct this deficiency.

3. The magnitude of the peak Xenon worth on the simulator following a trip from full power equilibrium is too small.

Deficiency Report No.1111 was written to correct this deficiency.

Appropriate retesting will be conducted when the noted Deficiency Reports are accepted. Results will be documented and retained in accordance with Simulator Verification Procedure S0123-XXI-3.1.4.

Conclusions. The core physics tests conducted on the simulator have produced results that are in close agreement with the plant reference data.

Differences in values, except for the noted deficiencies, will not adversely affect the dynamic response of the simulator nor the actions of operators during training and testing scenarios. Subject to the exceptions noted, the simulator satisfies the requirements in ANSI/ANS 3.5-1985 for core performance testing and for initialization conditions at various times in core life.

4.4 SURVEILLANCE TESTS ABSTRACT The surveillance tests were conducted using procedure S0123-XXI 3.3.4 SURVEILLANCE TESTING (Appendix M).

Fourteen safety related plant surveillance tests were performed. These specific surveillance tests were chosen from the entire set of plant surveillances because of their safety related importance, their ability to be performed from the control room, and their relevance to the training and testing of operators.

The surveillance test procedures selected were:

SURVEILLANCE TESTING No.

PLANT PROCEDURE SUBJECT

1.

SSTO1 REACTOR THERMAL POWER CALIBRATION

2.

SSTO2 CONTROL ROOM SHIFT AND DAILY LOG READINGS

3.

SST03 CONTROL ROD POSITION VERIFICATION

4.

SST04 BORIC ACID FLOW PATH VERIFICATION

5.

SST05 MISC. TECH. SPEC. LEVEL SURVEILLANCE

6.

SST06 HOT OPERATIONAL TEST OF THE SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS

7.

SST07 DIESEL GENERATOR LOAD TEST

8.

SST08 TURBINE STOP VALVE TEST

9.

SST09 MONTHLY CONTROL ROD EXERCISE

10.

SST10 MONTHLY-SPHERE ISOLATION CHANNEL TEST

11.

SST11 ACCIDENT MONITORING INSTRUMENTATION CHANNEL CHECK

12.

SST12 REACTIVITY CALCULATIONS

13.

SST13 AUX. FEEDWATER SYSTEM FLOW TEST

14.

SST14 REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE The tests were performed using the latest copy of the plant operating procedures. The record sheets from each surveillance test performed can be found in Appendix R.

Testing Requirements. The tests were performed to meet the requirements of ANSI/ANS 3.5-1985, Section 3.3.1, "Normal Plant Evolutions," Item (10).

The item specifies that the simulator shall be capable of performing operator conducted surveillance testing on safety related equipment or systems.

In addition, the performance testing requirements for principal mass and energy balances in ANSI/ANS 3.5-1985, Section 4.1, "Steady State Operation," were satisfied. Each test is described below.

S01-12.1-2, Reactor Thermal Power Calibration. The four safety channels of the excore nuclear instrumentation system were calibrated to the Turbine Plant calorimetric data.

The surveillance was completed satisfactorily.

Step 6.5.1 of the procedure requires that the "Auto Rod Withdrawal Prohibit Reset" pushbutton lamp be lit if calculated power deviates from indicated power. DR No.919 was written to correct this deficiency (with a priority rating of three assigned).

S01-12.1-4, Control Room Shift and Daily Log Readings.

The routine daily readings were taken per this procedure to compare channel indications with independent instrument channels measuring the same parameter to verify normal channel operation.

The surveillance was completed satisfactorily.

S01-12.1-5, Control Rod Position Verification. This test verifies that the control rod position analog indication is within rod insertion limits and that the analog signals compare within limits with the digital indications.

The surveillance was completed satisfactorily.

Two control rods and both control group bank position indications including shutdown margin indication were not indicated on YR 404. These deficiencies are.judged not to significantly affect operator training. DRs No. 1098, 1136 and 1137 were written to correct these deficiencies (They were all assigned a priority level three).

S01-12.2-10, Boric Acid Flow Path Verification. This surveillance verified the operability of the boric acid flow paths and pumps by operation of the valves, pumps and boric acid controls on the J console.

The surveillance test was completed satisfactorily.

S01-12.2-13, Misc. Tech. Spec. Level Surveillance. This instruction provides for routine surveillance requirements for the Spent Fuel Pit, the Boric Acid Storage Tank, the Refueling Water Storage Tank and the Hydrazine Tank.

The surveillance was completed satisfactorily with the exception of the Spent Fuel Pit which was not modeled.

801-12.3-2, Hot SI and Containment Spray Tests.

This surveillance tested the operability of the Safety Injection System and Containment Spray System components by operation of the respective pumps and valves. It also verified adequate boron concentration in the Safety Injection piping up to MOV-850A, B and C.

The surveillance was completed satisfactorily.

801-12.3-10, Diesel Generator Load Test. Both of the Diesel Generators were started from the control room and operated under the specified loading conditions. They were subsequently unloaded and removed from service.

The surveillance was completed satisfactorily.

When the Reactor Bypass Breakers were operated large load swings were observed on the respective 4KV Bus. This has been recorded on DR No. 1120 and given a priority 2 rating as this condition may have a negative impact on operator training.

801-12.3-15, Turbine Stop Valve Test.

This surveillance tested the operability of the main turbine stop valves by operation on the turbine controls on the J console.

The surveillance was completed satisfactorily.

S01-12.3-24, Monthly Control Rod Exercise Test. The Control Rods were inserted greater than ten steps and then withdrawn to the full out position. The surveillance repeats the same steps for both shutdown groups, control bank 1 and control bank 2.

The surveillance was completed satisfactorily.

Deficiencies noted for YR-404 in test SO1-12.1-5 apply.

S01-12.3-27, Monthly Sphere Isolation Channel Test. This surveillance demonstrates that each containment sphere isolation channel correctly operates and indicates the position of the isolation valves on the containment isolation panel in the control room.

The surveillance was completed satisfactorily.

801-12-3-40, Accident Monitoring Instrumentation Channel Check.

This surveillance assessed the operability of the accident monitoring instrumentation in the control room by quantitative and qualitative comparison, where possible, of channel indication and status with other similar indications.

In step 2.2.2 of this procedure, Steam Generator Wide Range Level indication was too low for plant status. DR 1096 was written.

This may have a negative impact on operator training so a priority 2 was assigned.

The surveillance was completed satisfactorily.

801-12.9-2, Reactivity Calculations. This surveillance was performed to determine Shutdown Boron Calculations, ECP/ECB Calculations and SDM Calculations manually using Cycle 10 data.

The surveillance was completed satisfactorily.

S01-12.9-12, Auxiliary Feedwater System Flow Test.

This surveillance tested the emergency flow path for the auxiliary feedwater pumps.

The surveillance test was not entirely completed on the turbine driven auxiliary feedwater system.

To correct this deficiency, DR 1304 was written and a priority 2 was assigned to this DR.

The surveillance procedure was completed satisfactorily.

801-12.9-21, Reactor Coolant System Water Inventory Balance.

This surveillance tested the RCS leak rate using the control room indications. A known leak rate was induced from the RCS at the instructor station. The leak rate calculation from the procedure matched the pre-programmed leak rate.

The surveillance procedure was completed satisfactorily.

Exceptions.

There were no exceptions taken to the ANSI/ANS 3.5 1985 standard as a result of the surveillance testing.

Conclusion. All of the surveillance tests identified in SO123 XXI-3.3.4 were performed satisfactorily and in accordance with plant operating procedures and instructions.

All noted deficiencies were presented to and approved by the Simulator Review Board.

Each completed surveillance procedure along with its test record sheet will be maintained per procedure S0123-XXI-3.1.4, Maintenance of Certification Records.

4.5 NORMAL OPERATIONS TESTS ABSTRACT The Normal Operations performance tests are documented using procedure S0123-XXI-3.3.5 (Appendix S).

The Normal Operations Performance Tests were performed satisfactorily and in accordance with plant operating procedures and instructions. The test data sheets from each Normal Operations Performance test can be found in Appendix S.

Certification Testing of the SONGS 1 simulator was accomplished in conjunction with Factory Acceptance Testing. The certification testing used the approved factory acceptance test procedures to meet the certification procedural requirements. In order to meet vendor acceptance requirements these acceptance test procedures (ATP) directed that approved SONGS procedures be used in all cases to test the simulator. However, some deviations from certification procedural requirements were needed. In every case these deviations were of a minor nature, primarily in the level of documentation.

The following evolutions were performed:

NORMAL OPERATIONS No.

TITLE

1.

NOP1 COLD SHUTDOWN TO HOT STANDBY (ATP SO1-XXI-3.1.N1B Plant Startup from Cold Shutdown to Hot Standby)

Tested: 9/11/91

2.

NOP2 HOT STANDBY TO RATED POWER (ATP SOl-XXI-3.1.NIC Plant Startup from Hot Standby to Minimum Load)

(ATP SO1-XXI-3.1.NlE Plant Startup from Minimum Load to Full Power)

Tested: 10/11/91

3.

NOP3 RATED POWER TO HOT STANDBY (ATP SO1-XXI-3.1-N2C Plant Shutdown from Full Power to Hot Standby)

Tested: 9/26/91

4.

NOP4 HOT STANDBY TO COLD SHUTDOWN (ATP SOl-XXI-3.1-N2D Plant Shutdown from Hot Standby to Cold Shutdown)

Tested: 9/27/91

5.

NOP5 REACTOR TRIP FOLLOWED BY RECOVERY TO RATED POWER (ATP SO1-3.1.Tl Manual Reactor Trip and Recovery)

Tested: 10/18/91 The tests were performed using the latest copy of the plant operating procedure and associated operating instructions.

Testing Requirements. The tests were performed to meet the requirements of ANSI/ANS 3.5-1985, Section 3.1.1, "Normal Plant Evolutions," Items (1) through (6), and Item (8) as follows:

1. Plant Startup -

cold to hot standby. The starting conditions shall be cold shutdown conditions of temperature and pressure. Removal of the reactor vessel head is not a required condition for simulation.

2. Nuclear startup from hot standby to rated power.

3. Turbine startup and generator synchronization.

4. Reactor trip followed by recovery to rated power.

5. Operations at hot standby.

6. Load changes.

8. Plant shutdown from rated power to hot standby and cooldown to cold shutdown conditions.

42 -

Noted Deficiencies. The amount and type of deficiencies are, to a large extent, the result of factory acceptance testing and are included as part of CAE's commitments to be resolved prior to completion of warranty. The following is a list of DRs (Deficiency Reports) to be resolved as a result of Normal Operations performance testing:

DR#

PROC.

DESCRIPTION PRIORITY 860 NOP1 Heat Up Rate Not Duplicating 2

Plant Data 859 NOP1 Pressurizer Level Spiking 2

During Heatup 851 NOP1 RCP Starting Current Not in 3

Long Enough 1145 NOP1 TCV-601B Controller Dev. Sig.

3 1146 NOP1 LC430F Response to FCV-1112 3

1104 NOP1 SR High Voltage High 3

835 NOP1 RCP #1 Seal Low Delta P Alarm 3

(Rx Plant #)

843 NOP1 Main FW Pp Low Suction Pressure 3

Alarms 1173 NOP2 Audio Count Rate Problems 2

1096 NOP2 S/G Level Correlation is Wrong 2

866 NOP2 Turbine Lube -Oil Reservoir Hi 3

868 NOP2 NIS Source Range Indicators 3

Rack vrs CR 888 NOP2 Reheater Tube Side Level Control 3

886 NOP2 Boric Acid Flow Indication On 3

Auto Make-Up 885 NOP2 Unit Net MWe Recorder Indication 4

906 NOP2 Control Rod, Rod Bottom Lights 4

1163 NOP2 Shutdown Group II Step Counter 4

812 NOP3 Power Ossc @ 100Mw 2

439 NOP3 RCS Delta T on RHR 2

528 NOP3 PSS Should Trip When Voltage 4

Regulator Trips 544 NOP3 Alert SW NIS Mode to HI Range 4

Alarm in at PWR 102 NOP4 PZR Spray Valve Response 2

816 NOP4 B RCP Bearing Temp < Cooling 3

Supply Temp.

805 NOP4 Vacuum Pump AMPS When Aligned 3

to the Waterbox 804 NOP4 TIC-433A/TIC-433B TIC-1105 Alarm 4

Set Point 1147 NOP4 Load Limit Pushbutton Response 4

63 NOP5 Auto Rod Withdrawal Block Reset 2

Light Blinks After Trip 1111 NOP5 Peak Xenon Reactivity Low 2

43 -

Exceptions. An exception to ANSI/ANS 3.5-1985, Section 3.1.1, "Normal Plant Evolutions," Item (7), "Startup, Shutdown, and Power Operations with Less Than Full Reactor Coolant Flow,"

is discussed in Section 1.3 of this report.

Conclusion. All Normal Operations Performance testing was performed satisfactorily and in accordance with plant operating procedures and instructions.

All noted deficiencies were presented to and approved by the simulator review board.

Test data collected during the Normal Operation Performance testing will be maintained per procedure S0123-XXI-3.1.4, Maintenance of Certification Records.

4.6 TRANSIENT TESTS ABSTRACT The transient performance tests were documented using procedure S0123-XXI-3.3.6 "Transient Testing (Appendix 0). Nine transients were selected as benchmark tests. The test data sheets can be found in Appendix T.

Certification testing of the SONGS 1 simulator was accomplished in conjunction with the factory acceptance testing. The certification testing used the approved factory acceptance test procedures to meet the certification procedural requirements. In order to meet vendor acceptance requirements these acceptance test procedures (ATP) directed that approved SONGS procedures be used in all cases to test the simulator.

The following evolutions were performed:

TRANSIENT TESTS No.

TITLE TTO1 Reactor Trip From Full Power (ATP SO1-XXI-3.1.Tl Manual Trip and Recovery)

Tested: 11/23/91 TT02 Loss of All Main Feedwater Pumps (ATP SO1-XXI-3.1.T11 Loss of Main Feedwater)

Tested: 10/30/91 TT03 Simultaneous Trip of All RCPs (ATP SO1.XXI-3.1.T5 Loss of All Reactor Coolant Pumps)

Tested: 10/22/91 TT04 Trip of a Single RCP (ATP SO1-XXI-3.1.T4 Loss of One Reactor Coolant Pump)

Tested: 10/12/91 TTO5 Turbine Trip Without Reactor Trip (ATP SO1-XXI-3.1.T2 Turbine Trip from Below 10% Reactor Power)

Tested: 10/18/91 TTO6 Small Break LOCA Inside Containment (ATP SO1-XXI-3.1.T6 Small Break Loss of Coolant Accident inside Containment)

Tested: 12/5/91 TTO7 Main Steam Line Break Inside Containment (ATP SO1-XXI-3.1.T17 Main Steam Line Break)

Tested: 11/23/91 TTO8 Failure of Pressurizer Power Operated Relief Valve (ATP SO1-XXI-3.1.T9 Failure of Pressurizer Power Operated Relief Valve)

Tested: 12/5/91 TTO9 100%-70%-100% Maximum Rate Ramp (ATP SO1-XXI-3.1.T3 Max Rate Ramp 100%-70%-100%)

Tested: 10/2/91 Simulation Verification Procedure S0123-XXI-3.3.6, "Transient Testing" allows comparison of simulator data with (1) actual plant data, or (2) analytical or plant design data, or (3) results from similar plants, or (4) best estimate. When the transient tests were performed for comparison against plant data, manual operations were performed as needed to match the actions taken in the plant. When the transient tests were performed for comparison against either analytical or plant design data, i.e.

Westinghouse Transient analysis, the Updated Final Safety, Analysis Report (UFSAR), or the Combustion Engineering Plant Analysis Code (CEPAC), manual operations on the simulator were performed as needed to match the initial conditions to that of the analysis code.

Each transient was performed and then compared against the appropriate data.

The data used in the transient operability tests was collected from transients run at about full power conditions with reduced Tavg (about 90% Full Power), equilibrium xenon with no operator follow up action.

Only two transients were performed at less than 90% Full Power.

TTO4 was performed at less than 50% Full Power. This was the maximum power level which did not result in an immediate reactor trip with loss of a Reactor Coolant Pump.

TTO5 was performed at less than 10% Full Power. This was the maximum power level which did not result in an immediate reactor trip.

Testing Requirements. The tests were performed to meet the requirements of ANSI/ANS 3.5-1985, Section 4.2 "Transient Operation" and were-selected from Appendix B 2.2 "Transient Performance". Section 4.2 requires that tests shall be conducted to prove the capability of the simulator to perform correctly during the limiting cases of those evolutions identified in 3.3.1. "Normal Plant Evolutions" and 3.1.2 "Plant Malfunctions" of this standard.

Noted Deficiencies. Each transient is listed below and followed by a comment concerning any deficiencies noted during the test.

TTO1 Reactor Trip From Full Power. Following a reactor trip the main steam pressure response was greater than the plant.

Deficiency report DR-1249 with a priority 2 identifies this concern. The simulator pressurizer pressure response follows the plant response but the response rate of recovery on the simulator is greater than the plant response. This deficiency has been identified in DR-1248 with a priority 2 assigned.

TTO2 Loss of All Main Feedwater Pumps.

No deficiencies noted.

TTO3 Simultaneous Trip of All RCPs. During natural circulation an unrealistic RCS pressure drop was experienced with no change in pressurizer level. This deficiency is identified in DR-928 with a priority 2 assigned.

TTO4 Trip of a Single RCP. No outstanding deficiencies remain.

TTO5 Turbine Trip Without Reactor Trip.

During middle of coastdown the voltage regulator went to the "test" position. This deficiency is identified by DR-1092 with a priority 2 assigned.

TTO6 Small Break LOCA Inside Containment. The ANSI/ANS 3.5-1985 transient list includes a Design Basis Accident LOCA with loss of offsite power accident.Section V of ANSI/ANS 3.5 however states that the "simulator user should substitute other applicable transient tests if such tests provide a more representative comparison to actual or predicted reference plant performance". A Small Break LOCA UFSAR analysis performed by the Westinghouse NOTRUMP analysis code was instead chosen to benchmark the simulator, since the analysis provides a greater amount and higher accuracy data for comparison with the simulator.

The simulator break size was adjusted to closely match the initial 2400 lbm/sec flow of the W SBLOCA Analysis.

No outstanding deficiencies remain.

TTO7 Main Steam Line Break Inside Containment. No deficiencies noted.

TTO8 Failure of Pressurizer Power Operated Relief Valves. No outstanding deficiencies remain.

TTO9 100%-70%-100% Maximum Rate Ramp. No outstanding deficiencies remain.

Exceptions. An exception to ANSI/ANS 3.5-1985, Appendix B2.2 "Transient Performance" is discussed in Section 1.3 of this report.

Conclusion. All of the transient tests identified in S0123-XXI 3.3.6 were performed satisfactorily and in accordance with plant operating procedures and instructions.

All noted deficiencies were presented to and approved by the simulator review board.

Each completed transient test procedure along with its test record sheet is being maintained per procedure S0123-XXI-3.1.4, Maintenance of Certification Records.

4.7 MALFUNCTION TESTS ABSTRACT The simulator malfunctions were tested during the Acceptance Test Program using procedure SO123-XXI-3.3.7 MALFUNCTION TESTING (Appendix P).

Table V is a complete list of the simulator malfunctions.

Table III lists the ANSI/ANS 3.5-1985 required malfunctions and indicates the 95 simulator malfunctions that meet the requirement.

All of the required 95 malfunctions have been tested. All of these will be available at commencement of training. Three malfunctions have Priority 2 Deficiency Reports written against them and will be available for training within a month of simulator certification submittal.

Of the remaining 91 malfunctions, all are available for training.

While every malfunction has not yet been tested per S0123-XXI-3.3.7, every malfunction will be validated prior to its use during training.

Deficiencies. There-are a total of 84 Deficiency Reports (DRs) outstanding against the simulator malfunctions. 13 of these are Priority 2, 64 are priority 3 and the remaining 7 are Priority 4.

O The large number of DRs is due to initial testing having been conducted in parallel with factory acceptance testing.

Estimated Completion Schedule. Malfunctions that have been tested but are not available for training will be made available within one month of certification submittal.

Testing Requirements. Initial certification malfunction tests were performed to meet the requirements of ANS/ANSI 3.5-1985, Section 4.2, "Transient Operation," and Appendix A, "Guide for Documenting Simulator Performance," Section A3.4.

Exceptions.

No exceptions to ANSI/ANS 3.5-1985 were taken as a result of malfunctions testing.

Conclusions.

All ANSI/ANS 3.5-1985 required malfunctions have been tested and are available for training. All remaining malfunctions are available for training.

All noted deficiencies were presented to and approved by the Simulator Review Board.

Test data collected during the Malfunction testing will be maintained per procedure S0123-XXI-3.1.4, "Maintenance of Certification Records.

4.8 DEFICIENCY STATUS CAE Electronics' Deficiency Report Management System (described in Section 5.1) is being used to track and prioritize the simulator deficiencies. The deficiencies noted during performance of tests required by ANSI/ANS 3.5-1985 are noted in the abstracts. Additional deficiencies were noted in the simulator hardware and instructor station, and during additional tests not specifically required by the ANSI standard. The following table is a summary of all outstanding Deficiency Reports (DR) and their priorities.

DR CATEGORY PRIORITY 2

3 4

TRANSIENT TEST 14 12 8

MALFUNCTIONS 13 64 7

NORMAL OPS 21 54 17 SURVEILLANCES 0

4 0

HARDWARE 6

11 13 INSTRUCTOR STATION 8

37 38 The Simulator Review Board has reviewed and approved the priorities of the simulator deficiencies.

In accordance with S0123-XXI-3.2.3, outstanding DRs will be resolved as follows:

Priority 2 DRs will be resolved within one month of certification submittal.

Priority 3 DRs will be resolved within six months of certification submittal.

Priority 4 DRs will be resolved within one year of certification submittal.

TABLE V SIMULATOR MALFUNCTIONS 1 REACTOR COOLANT SYSTEM (RCS) 1.1 Break In Hot Leg of Any Loop 1.2 Break In Cold Leg of Any Loop 1.3 Reactor Vessel Head Vent Leak 1.5 Reactor Vessel Flange Leak-Off (Inner/Outer Seal) 1.6 Steam Generator Tube Rupture (ANY S/G) 1.7 Fuel Cladding Failure 1.8 Uncontrolled Boration 1.9 Uncontrolled Dilution 1.10 Control Rod Ejection 1.11 RCS Flow Transmitter Failure 1.12 RCS Temperature Transmitter Failure 2 REACTOR COOLANT PUMP (RCP) 2.1 RCP Trip 2.2 RCP Locked Rotor 2.3 Upper Oil Reservoir HI/LOW Level 2.4 CP Sheared Shaft 2.6 Loss of No. 1 Seal 2.7 Loss of No. 2 Seal 2.8 Loss of No. 3 Seal 2.9 RCP Thermal Barrier Heat Exchanger Leakage 2.10 Loss of CCW to RCP Upper Oil Cooler 2.11 Loss of CCW to RCP Thermal Barrier Heat Exchanger 3 PRESSURIZER 3.1 Pressurizer Spray Valve Failure 3.2 Pressurizer PORV Failure 3.3 Pressurizer Heater Failure 3.4 Pressurizer Safety Valve Failure 3.5 Pressurizer Pressure Controller Failure 3.6 Pressurizer Pressure Transmitter 3.7 Pressurizer Level Controller Failure 3.8 Pressurizer Level Transmitter Failure 3.9 overpressure Mitigation System Failure 3.10 Pressurizer Steam Space Leak TABLE V (continued) 4 STEAM GENERATORS (S/G) 4.1 S/G Level Controller Failure 4.2 S/G Narrow Range/Wide Range Level Transmitter Failure 4.4 S/G Feed Flow Transm. Failure 4.5 S/G Steam Flow Transm. Failure 5 REACTOR CORE 5.1 Radial Flux Tilt 6 CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS) 6.1 Letdown Line Leakage Inside Containment 6.2 Letdown Line Leakage Outside Containment 6.3 RHR Heat Exchanger Tube Leakage 6.4 Charging Line Leakage Inside Containment 6.5 Charging Line Leakage Outside Containment 6.6 Charging Flow Controller Failure 6.7 Charging Pump Breaker Trip/Failure 6.8 Any Charging Pump Loss of Cooling (CCW) 6.9 Loss of a Seal Injection FCV (ANY RCP) 6.10 Failure of Seal Injection Controller 6.11 Loss of RHR Heat Exchanger Cooling 6.12 Letdown Pressure Controller Output 6.14 VCT Level Transmitter Failure 6.15 Plugged Filter (Reactor Coolant, Either Seal Injection, Seal Water Return).

6.16 Reactor Makeup Control Failure (Any Mode) 6.17 Boric Acid Controller Failure 6.18 Boric Acid Pump Failure (Any) 6.19 Boric Acid Storage Tank Heaters Failure (ON/OFF) 7 SAFETY INJECTION SYSTEM 7L.e1 SI Recirc Pump Breaker Trip/Failure 7.2 SI Pump Breaker Trip/Failure 7.3 SI Line Break Inside Containment Upstream of MOVs 7.4 Hydraulic Valve Failure in Feedwater Mode 7.4B Hydraulic Valve Failure in Safety Injection Mode 7.5 RWST Tank Level (High/Low) 7.6 RWST Level Transmitter Failure (Any; HIGH/LOW) 7.7 Feed Pump in SI Mode Discharge Leak Outside Containment at Header TABLE V (continued) 8 RESIDUAL HEAT REMOVAL 8.1 RHR Pump Breaker Trip/Failure 8.2 RHR Pump Air Entrainment 9 CONTAINMENT 9.1 Refueling Water Pump Breaker Trip/Failure 9.2 Containment Pressure Transmitter Failure 10 SALT WATER COOLING (SWC) AND COMPONENT COOLING WATER AND (CCW) 10.1 SW Cooling Pump Breaker Trip/Failure 10.2 CCW Pump Breaker Trip/Failure 10.3 CCW System Leakage 11 AUXILIARY FEEDWATER (AFW) ll.lA Electric Driven AFW Pump Breaker Trip/Failure ll.lB Turbine Driven AFW Trip/Failure 11.2 AFW Flow Control Valve Failure 11.3 AFW Pump Discharge Piping Leakage 11.4 AFW Tank Level 11.5 AFW Pump Steam Binding 12 REACTOR PROTECTION SYSTEM 12.1 Reactor Trip (Actuate/Fail to Actuate) 12.2 Safety Injection (Actuate/Fail to Actuate) 12.3 Containment Isolation (Actuate/Fail to Actuate) 12.4 Containment Spray (Actuate/Fail to Actuate) 12.5 Turbine Trip (Either Train) (Actuate/Fail to Actuate) 12.6 ATWS (Anticipated Transient Without Scram) 12.7 RPS Channel Failure TABLE V (continued) 13 NUCLEAR INSTRUMENTATION SYSTEM (NIS) 13.1 Any Core Exit Thermocouple Fails (High/Low) 13.2 Failure of Source Range Detector 13.3 Failure of Source Range to De-Energize 13.4 Noisy Source Range Channel 13.5 Failure of Source Range Startup Rate Circuit 13.6 Failure of the Intermediate Range Detector 13.7 Failure of Intermediate Range Startup Rate Circuit 13.8 Power Range Detector Failure (Any; Upper/Lower) 14 ROD CONTROL AND DIGITAL ROD POSITION INDICATION (DRPI) SYSTEMS 14.1 Control Rod Group Fails to Move on Demand (Any; Auto/Manual) 14.2 Incorrect Rod Group Direction (Auto/Manual) 14.3 Continuous Group Movement (Any In/Out) 14.4 Control-Rod Interlock Failure (Any; On/Off) 14.5 Rod Drop 14.6 Stuck Rod 14.7 Slave Cycler/Master Cycler Failure 14.8 Failure of a Group Demand Step Counter 14.9 Rod Position Indicators Fail (Any) 15 MAIN GENERATOR 15.1 Generator Trip or Failure to Trip 15.2 Generator Fault 15.3 Voltage Regulator Failure 15.4 Main Generator Synchroscope Failure 15.5 Loss of Excitation 15.6 Field Breaker Trips or Fails to Trip 15.7 Loss of Cooling to Hydrogen System 15.8 Loss of Load 15.9 Out of Step Failure 15.10 Generator Negative Sequence 15.11 Generator Stator Ground (i

TABLE V (continued) 16 ELECTRICAL SYSTEMS 16.1 Loss of All Offsite Power 16.2 Degraded Grid Voltage (High/Low) 16.3 Degraded Grid Frequency (High/Low) 16.4 Transformer Fault (Any) 16.5 Loss of Invertor (Any) 16.6 4.16-KV Bus Fault (Any) 16.7 DG Trip (Either:

Engine/Generator) 16.8 DG Fail-To-Start/Fail-To-Run/Fail-To-Trip (Either) 16.9 DG Output Breaker Trips/Fails to Close/Fails to Open (Either) 16.10 Sequencer Failure (Either) 16.11 Loss of Any 480-V AC Bus 16.12 Loss of Any 120-V AC BUS 16.13 Loss of Any 125-V DC Bus 16.14 4KV, 480V Ground (Any Location) 16.15 120V DC Ground (Any Location) 17 MAIN STEAM SYSTEM 17.1 Main Steam Line Break Upstream of Flow Restrictor 17.2 Main Steam Line Break Downstream of Flow Restrictor 17.3 Main Steam Line Break Outside Containment, Upstream of Stop Valves 17.4 Main Steam Safety Valve Failure 17.5 Steam Dump Valve Failure 17.6 Steam Dump Steam Pressure Controller Failure 17.7 Main Steam Header Pressure Transmitter Failure 17.8 Reheater Controller (RMC3) Failure 18 MAIN TURBINE 18.1 Stop Valve Failure 18.2 Main Turbine Control Valve Failure 18.3 Loss of Turbine Shaft Driven Lube Oil Pump 18.4 Electric Driven Turbine Oil Pumps Breaker Trip/Failure 18.5 Turbine Generator Bearing Vibration 18.6 Main Turbine Eccentricity 18.7 Main Turbine Thrust Bearing Wear 18.8 Spurious Runback or Failure to Runback 18.9A Failure of Turbine Governor 18.9B Failure of Turbine Load Limit 18.10 Turbine Bearing Oil Low Pressure 54-

TABLE V (continued) 18.11 Loss of Cooling to Lube Oil 18.12A Turning Gear Motor Trip 18.12B Turning Gear Fails to Engage 18.13 Gland Seal Regulator Failure 18.14 Failure of Turbine First Stage Pressure Transmitter PT-415 and/or PT-417 18.15 Exhaust Hood Spray Failure 18.16 MSR Steam Supply MOV Fails (Either MSR, Open/Close) 19 CONDENSER AND CIRCULATING WATER (CW) 19.1 Loss of Condenser Vacuum 19.2 Condenser Tube Leakage. Either One of Four, North Half of A or B. South Half of A or B.

19.3 Circulating Water Pump Bearing Failure 19.4 Circulating Water Pump Breaker Trip/Failure 19.5 Turbine Plant Cooling Water Pump Breaker Trip/Failure 19.6 Condenser Vacuum Pump Breaker Trip/Failure 19.7 Plugged Condenser Tube Sheet 20 CONDENSATE 20.1 Hotwell Level Controller Fails 20.2 Condensate Pump Breaker Trip/Failure 20.3 Low-Pressure Heater Tube Leakage 21 MAIN FEEDWATER (MFW) AND HEATER DRAINS 21.1 Heater Drain Pump Breaker Trip/Failure 21.2A Feedwater Heater Normal Level Control Failure 21.2B. Feedwater Heater Level Dump Control Failure 21.3 Main Feedwater Pump Breaker Trip/Failure 21.4 MFW Oil Pressure Failure 21.5 MFW Auxiliary Oil Pump Breaker Failure 21.6 MFW Pump Recirculation Failure 21.7 FW NOV Fails 21.8A MFW Regulator Failure 21.8B MFW Bypass Regulator Failure 21.9 High Pressure FW Heater Tube Leakage 21.10 Feed Break Inside Containment, Upstream of Check Valves 21.11 Feed Break Inside Containment, Downstream of Check Valves TABLE V (continued) 21.12 Feed Break Outside Containment Downstream of Check Valve 21.13 Feed Break Outside Containment on Common Header 22 MISCELLANEOUS 22.1 Instrument Air Compressor Trip 22.2 Loss of Instrument Air 22.3 ARMS Channel Fails 22.4 ORMS Channel Fails 22.5 High Radiation in Auxiliary Building 22.6 Indication of Seismic Event 22.7A High Conductivity in Condensate System 22.7B High Conductivity in Feedwater System 22.8 Post Accident Radiation Monitoring System Failure 22.9 instrument Air Dryer Failure 5.0 CONFIGURATION MANAGEMENT SYSTEM To ensure the simulator is a good representation of SONGS 1, there are multiple cross checks incorporated into the verification program used to identify differences between the plant and the simulator. Some of these checks are incorporated into the Configuration Management System. The following list provides some of the checks that are used to identify differences:

Review of plant DCPs and FCNs for incorporation into the simulator. (SO123-XXI-3.2.3 "Review of Plant Modifications").

Testing of plant modifications incorporated into the simulator. (S0123-XXI-3.2.2, "Simulator Work Orders").

Testing of malfunctions and transients on a routine basis. (S0123-XXI-3.3.7, "Malfunction Testing," and S0123-XXI-3.3.6, "Transient Testing").

Feedback from students and instructors via the simulator trouble report form and simulator fidelity student reaction form, (to be implemented with the commencement of training). (S0123-XXI-3.2.1, "Trouble Reports," and S0123-XXI-3.4.1, "Hardware Verification").

CAE Electronics' Deficiency Report Management system tracks deficiencies (called "discrepancies" in SCE's Configuration Management System) identified by instructors and operators during the factory acceptance testing of the simulator. After the simulator warranty period expires, tracking of deficiencies will be accomplished by implementing the Simulator Verification Procedure S0123-XXI-3.2.1, "Trouble Reports".

The use of actual plant procedures verifies that the response of the simulator is as expected.

Testing of the simulator response against plant transients, Licensee Event Reports, and Station Incident Reports.

Testing of the simulator steady-state conditions against the plant's steady-state values.

(SO123-XXI-3.3.2, "Steady-State Testing").

Hardware verification program to compare the simulator control boards with the plant control boards using either video tapes or photographs.

(SO123-XXI-3.4.1, "Hardware Verification").

The Configuration Management System is used to track the simulator configuration with respect to SONGS 1. The objective is to maintain the simulator as close to SONGS 1 as possible, both in appearance and dynamic response. The Configuration Management System is organized as follows:

Simulator Work Order System

• DCP System

• Discrepancy Report System

• Delta System (Differences between simulator and plant)

• Malfunction Testing System

• Spare Parts System

• Preventive Maintenance System

• Report Generating System

• Software Change Documentation System Simulator Drawing and Historical File System Simulator Work Order (SWO) System. Simulator Work Orders are written using procedure S0123-XXI-3.2.2. "Simulator Work Orders."

SWOs come from several different sources and may have several priority levels. The SWOs can be generated automatically from DCP/FCNs, manually from trouble reports, or other sources.

Any work performed on the simulator, other than routine preventa tive maintenance, is performed with a SWO. The SWO describes the work and the system affected. If applicable, the source requiring the work is identified.

If a source item requires both software and hardware work to be accomplished for the same task then two separate work orders are written: one for software and one for hardware.

When a SWO is completed, testing is performed to verify that the simulator configuration is correct. The comment section of the SWO is used to document this verification if only minor testing is required. If more detailed testing is done, supporting documentation is attached to the SWO and filed.

DCP System. When plant modifications are planned a Field Change Notice (FCN) or Design Change Package (DCP) is generated. A copy of the FCN/DCP is sent to the Simulator Support Group for review and evaluation.

(This group is a part of the SONGS Nuclear Training Division).

A record is kept of all DCP/FCNs reviewed.

If the change has simulator impact, then a Simulator Work Order (SWO) is generated against this change per procedure S0123-XXI 3.2.3, "Review of Plant Modifications."

This information is also entered into the simulator Delta tracking system.

A special type of DCP is sometimes generated. This is a simulator DCP.

Simulator DCPs generally incorporate changes made to one control board or one specific type of change (e.g.,

annunciators).

A simulator DCP is tracked in the same manner as a plant DCP, however, closeout of a simulator DCP usually results in the closeout of several related plant DCP work orders.

Simulator DCPs are developed, reviewed and closed out as described in procedure S0123-XXI-3.2.4, "Simulator Design Change Packages."

Discrepancy Reporting System. The simulator discrepancy reporting system, procedure S0123-XXI-3.2.1, "Trouble Reports,"

is used as an aid in maintaining simulator fidelity. This provides a method for the simulator instructors to report any abnormalities noted during a training exercise.

It also provides a process for feedback from students on items they note different between the simulator and the plant.

Delta System. The delta system is a semi-automatic tracking system that is used to document any differences that exist between the simulator and the plant.

It automatically receives input from DCP/FCNs that have impact on the simulator when they are entered in the Configuration Management System. When a modification is finally installed in the plant, a DCP/FCN turnover/closeout sheet is issued. A copy of this closeout is sent to the Simulator Support Group for review. If the associated DCP/FCN has simulator impact, it is entered into the Delta tracking system using procedure S0123-XXI-3.2.3, "Review of Plant Modifications."

The system can then be used to generate a Delta Report to show which changes have been implemented in the plant but not yet in the simulator. When the associated SWO is closed-out, the associated delta is automatically closed. This Delta System report is reviewed at least once a year to determine the modifications that were incorporated into the plant. The modifications are scheduled for incorporation into the simulator within one year per ANSI/ANS 3.5-1985.

Simulator Test Tracking System. This system keeps track of the operability testing and performance testing status over a four year period for re-certification purposes.

Spare Parts System. To increase the availability of the simulator for training, a partial spare parts inventory is kept in the simulator area.

The spare parts system helps track the inventory. The system generates reports of parts to be ordered and those already on order.

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Preventative Maintenance System. A preventative maintenance scheduling system is used to aid in the continued reliable operation of the simulator. This system automatically schedules periodic preventative maintenance.

The system keeps track of completed as well as incomplete work.

Report Generating System. There are many reports available from the Configuration Management System (CMS) that aid in tracking the required data associated with the simulator, and to ensure that required items are being completed in a timely and efficient manner. Some of the reports are generated on a periodic basis while others are created on an as needed basis.

Software Change Documentation System. Whenever a change is made to any software program, the software engineer fills a simulator software supplement form. The form is filed with the SWO and into the CMS using procedure S0123-XXI-3.2.2, "Simulator Work Orders."

This process provides a record of all the changes that were made to the simulator software.

Simulator Drawing and Historical File System. Drawings are maintained to show how various plant systems are modeled in the simulator software.

These drawings are updated whenever changes are made that affect the model representation shown in the drawing. The SWO is used whenever simulator drawings are affected. This information is then entered into the associated drawing file. When sufficient changes are made to warrant a change, the drawings are updated. The simulator drawings are being incorporated into the new simulator instructor station displays. These new interactive drawings will be maintained using the same drawing control system. The historical file keeps track of the revisions made, the date they were made, and the work order requiring the change.

5.1 CONFIGURATION CONTROL IN MONTREAL Configuration Control of the simulator while in Montreal will be in accordance with the Configuration Management System described in Section 5.0. The remote location and the fact that the simulator is under warranty result in additional configuration control elements being used.

CAE Electronic's Deficiency Report Management System (DRMS). The system functions in.parallel with the Simulator Verification Procedure S0123-XXI-3.2.2, "Simulator Work Orders". The same items that are identified and tracked by the SWO system (deficiencies, Software and hardware upgrades, etc) are placed into the DRMS when it has been identified that the item is the responsibility of the simulator vendor. In this instance the DRMS will be used in place of SWO's. When an item is identified as being the responsibility of SCE, the SWO system is utilized.

The DRMS provides all the documentation that is necessary to meet the requirements of the SCE configuration Management System.

Upgrade Change Orders. When the Configuration Management System identifies changes that are required on the simulator due to changes in the plant, an SWO is generated. A determination is then made as to when the change is to be implemented in the simulator. If it is determined that the change is required while the simulator is in Montreal, a Change Order to the simulator Purchase Order is issued and given to the simulator vendor to install the change on the simulator.

5.2 REPORTS AND RECORDS Annual Report.

Annually, following the anniversary of certification, a report will be written to document the activities of the test cycle and identify the simulator performance tests scheduled. The Simulator Review Board will review and approve the annual report.

The annual reports will serve as the bases for the 4-year certification report to the NRC.

Each annual report will contain the following information and data:

Performance testing conducted in the previous test cycle including: tests scheduled to be performed, test failures or failure to perform tests, status of test failures and justification for non-performance, and the schedule to resolve the failures if required. Performance tests scheduled to be performed during the current test cycle will also be listed.

Plant changes and modifications that impact the simulator and the modifications incorporated into the simulator during the last test cycle.

Special testing or verification of the simulator against plant data to verify simulator response.

Additional malfunctions added because of plant events, operational experiences, or probabilistic risk assessment studies.

Certification Report.

Every fourth year, on the anniversary of the certification, a report shall be made to the Nuclear Regulatory Commission.

The report following the initial cer tification shall contain a description of the test program and a schedule for the conduct of a minimum of 25 percent of all performance tests per year for the subsequent four years.

The report shall also identify any uncorrected test failures with a schedule for correction and any deviations from the previously submitted test schedule.

Records and Documentation. Records and documentation will be maintained for all certification records.

The affected records are identified in procedure S0123-XXI-3.1.4, "Maintenance of Certification Records."

The Simulator Support Group will maintain the records until the certification report supported by the documents is sent to the NRC. At that time, all records supporting the certification report will be transmitted to CDM for storage. All certification records will be maintained for four years after the submittal date.

6.0 CONCLUSION

S SCE has conducted performance tests and reviews to satisfy the requirements for initial certification under 10 CFR 55.45(b) for the SONGS 1 Control Room Training Simulator. The tests were completed satisfactorily, and outstanding deficiencies have been scheduled to be corrected in accordance with established procedures.

The program being implemented to assure simulator fidelity is effective, and will continue to maintain the facility within the certification requirements.

The guidance of ANSI/ANS 3.5-1985 and Regulatory Guide 1.149 has been met. Only four exceptions to ANSI/ANS 3.5-1985 were taken:

Increase the data resolution to one second during transient testing Eliminate an operational evolution test that involves less than full reactor flow, since it violates the SONGS 1 Technical Specifications, Eliminate a transient involving closure of main steam isolation valves, since SONGS 1 does not have these valves, Minor control room environmental differences while the simulator is at the temporary location in Montreal, Quebec, Canada.

These exceptions do not detract from operator training or operator examinations.

APPENDIX A

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 1 OF 11 EFFECTIVE DATE NOV 1. 1991 SIMULATOR VERIFICATION PROCEDURE SIMULATOR REVIEW BOARD TABLE OF CONTENTS SECTION PAGE 1.0 OBJECTIVES 2

2.0 REFERENCES

2 3.0 PREREQUISITES 2

4.0 PRECAUTIONS 2

5.0 CHECKOFF LIST 3

6.0 PROCEDURE 3

6.1 General 3

6.2 Simulator Review Board Membership 3

6.3 Responsibilities 4

6.4 Review Process and Required Actions 5

7.0 RECORDS 7

ATTACHMENT 1

ANSI/ANS 3.5-1985 Performance Criteria 8

2 Simulator Review Board Meeting Minutes 10 3

Simulator Review Board Quarterly Reviews 11 NOT QA PROGRAM AFFECTING

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 2 OF 11 SIMULATOR VERIFICATION PROCEDURE SIMULATOR REVIEW BOARD 1.0 OBJECTIVE 1.1. To describe the makeup and responsibilities of the Simulator Review Board.

1.2 To provide the necessary documentation for the Simulator Review Board meetings and findings.

1.3 To identify the simulator activities that will be reviewed by the Simulator Review Board.

2.0 REFERENCES

2.1 Simulator Verification Program 2.2 Title 10, Chapter 1, Code of Federal Regulations, Part

55.

2.3 NUREG-1258, Evaluation Procedure for Simulation Facilities Certified under 10CFR 55.

2.4 ANSI/ANS-3.5-1985, American National Standard, Nuclear Power Plant Simulators for Use in Operator Training.

2.5 Regulatory Guide 1.149, Nuclear Power Plant Simulation Facilities for use in Operator License Examinations.

2.6 S0123-XXI-3.1.3, Annual Report Generation 2.7 S0123-XXI-3.1.4, Maintenance of Certification Records 2.8 S0123-XXI-3.2.2, Simulator Work Orders 2.9 S0123-XXI-3.2.6, Maintenance of Malfunctions 2.10 S0123-XXI-3.3.3, Steady State Testing 2.11 S0123-XXI-3.3.6, Transient Testing 2.12 S0123-XXI-3.4.1, Hardware Verification 3.0 PREREQUISITES 3.1 None 4.0 PRECAUTION(S) 4.1 None 5.0 CHECKOFF LIST(S) 5.1 None

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 3 OF 11 6.0 PROCEDURE 6.1 General 6.1.1 The Simulator Support Administrator shall be responsible for coordination and scheduling of all Simulator Review Board (SRB) meetings.

The Simulator Support Administrator is responsible for the following activities:

.1 Prepare and distribute the agenda;

.2 Prepare and have available the appropriate review forms and simulator data to be reviewed;

.3 Maintain minutes of the meetings including the attendance and findings of the board.

6.2 Simulator Review Board Membership 6.2.1 The Simulator Review Board will be made up of individuals that have a vested interest in the fidelity and response of the simulator.

6.2.2 The following people or their representatives shall have representation on the board.

Station Operations, Supervisor of Unit Operations Manager, Station Technical Nuclear Training Division (NTD),

Supervisor, Operations Training NTD, Administrator, Simulator Training NTD, Administrator, Simulator Support Group NTD, Manager, Nuclear Training 6.2.3 A quorum necessary for performance of SRB responsibilities shall consist of the chairman and at least three (3) members.

6.2.4 The SRB will be chaired by the Manager of Nuclear Training or his representative.

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 4 OF 11 6.0 PROCEDURE (Continued) 6.2.5 Should the SRB determine that the meeting participants lack sufficient expertise or information for an adequate evaluation of the item under

review, the item will be rescheduled and appropriate expertise added as guests, non-voting members.

6.3 Responsibilities 6.3.1 The SRB shall be responsible for:

.1 Review and approval of the certification program description.

.2 Review and approval of simulator certification required procedures.

.3 Ensure certification process is established and maintained by performing audit samplings and review of all performance deviations.

.4 Provide recommendation for certification submittal.

.5 Perform annual performance based audits.

.6 Review and approval of the simulator benchmark or factory acceptance tests results required for ANS 3.5 certification.

.7 Annually, review and approval of the simulator transient evaluation justification.

.8 Approve any deletion or revision of an ANS 3.5 required malfunction.

.9 Annually, review the priority 4 & 6 Simulator Work Orders (SWOs) generated during the year.

.10 Annually, review the hardware discrepancies and approve the justifications for these discrepancies.

.11 Annually, review and approval of the simulator annual report.

.12 Approval of other activities that may impact training of operators or certification of the simulator.

NUCLEAR TRAINING DIVISION SO123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 5 OF 11 6.0 PROCEDURE (Continued) 6.4 Review Process and Reauired Actions 6.4.1 Simulator transient benchmark test reviews

.1 The simulator transient benchmark tests will be conducted and evaluated.

The results of the testing will be presented to the SRB to verify that the simulator meets the training and certification requirements.

.2 A presentation and written report will be made to the review board that provides the critical parameter response for each of the transients.

The differences between reference data and the simulator will be identified.

.3 Justification of any significant differences will be identified. The simulator benchmark data shall as a minimum meet the ANS 3.5 criteria for simulator response.

See attachment 1,

ANSI/ANS 3.5 Performance Criteria.

.4 The SRB shall approve without comment, approve with recommendations or disapprove the simulator response.

.5 of S0123-XXI-3.3.6, Simulator Transient Testing, will be signed by the Chairman of the SRB to indicate acceptance of the simulator response.

6.4.2 Annual Transient Testing Review

.1 The results of the annual transient testing will be reviewed and the variations in the results from the previous year will be justified to the SRB.

.2 of S0123-XXI-3.3.6, Simulator Transient Testing, will be signed by the Chairman of the SRB to indicate acceptance of the simulator response justification.

6.4.3 ANS 3.5 Malfunction Deletion or Modification

.1 The SRB will review and approve deletions of or modification to malfunctions required by the ANS 3.5 standard.

NUCLEAR TRAINING DIVISION SO123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 6 OF 11 6.0 PROCEDURE (Continued) 6.4.3.2 The listings of malfunctions required by the ANS 3.5 standard are provided in S0123-XXI 3.2.6, Maintenance of Malfunctions.

.3 This review and approval is to ensure that an adequate number of malfunctions is available for the training and testing of operators.

6.4.4 Review of Simulator Work Orders

.1 Annually the SRB will review and. approve priority 4 & 6 SWOs generated that year. This review is to insure that discrepancies do not impact the training or testing of operators.

.2 The SWO priority will be upgraded if training or testing of operators would be impacted.

6.4.5 Review of Hardware Discrepancy Justification

.1 Simulator hardware discrepancies and the justification for the discrepancies will be reviewed and approved by the simulator review board.

.2 If the SRB does not approve the justification for the discrepancy, the simulator will be modified to conform to the SRB requirements.

6.4.6 Review of the Simulator Annual Report

.1 The simulator annual report will be reviewed by the SRB. This report will provide the SRB with an overview of simulator activities during the preceding year. The report will also contain a summary of the SRB meeting activities.

.2 The SRB will approve the publication of the annual report.

6.4.7 other Activities

.1 The SRB may review and make recommendations about other activities involving the simulator, such as simulator enhancements for additional and/or improved training.

S1

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 7 OF 11 7.0 RECORDS 7.1 The SRB meeting minutes and other SRB documentation will be maintained in accordance with S0123-XXI-3.1.4, Maintenance of Certification Records.

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 8 OF 11 ATTACHMENT 1 EXCERPT FROM ANSI/ANS 3.5-1985 SECTION 4 PERFORMANCE CRITERIA 4.1 Steady State Operation The simulator accuracies shall be related to full power values and interim power levels for which valid reference plant information is available. The parameters displayed on control panels may have the instrument error added to the computed values. During testing, the accuracy of computed values shall be determined for a minimum of three points over the power range:

(1) The simulator instrument error shall be no greater than that of the comparable meter, transducer and related instrument system of the reference plant; (2) Principal mass and energy balances shall be satisfied.

Examples are:

(a) Net NSSS thermal power to generated electrical power; (b) Reactor coolant system temperature to steam generator pressure; (c) Feedwater flow to reactor thermal power; (d) Mass balance of pressurizer; (e) Mass balance of steam generator.

The simulator computed values for steady state, full power operation with the reference plant control system configuration shall be stable and not vary more than +/-2% of the initial values over a 60-minute period; (3)

The simulator computed values of critical parameters shall agree within +/-2% of the reference plant parameters and shall not detract from training. Some examples of critical parameters are:

(a) Reactor thermal power; (b) Reactor hot and cold leg temperatures; (c) Feedwater flow; (d) Steam pressure; (e) Generated electrical power; (f) Recirculation flow; (g) Reactor coolant system pressure.

(4) The calculated values of noncritical parameters pertinent to plant operation, that are included on the simulator control room panels, shall agree within +/-10% of the reference plant parameters and shall not detract from training.

ATTACHMNET 1 PAGE 1 OF 2

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 9 OF 11 ATTACHMENT 1 4.2 Transient Operation 4.2.1 Tests shall be conducted to prove the capability of the simulator to perform correctly during the limiting cases of those evolutions identified in 3.1.1 (Normal Plant Evolutions) and 3.1.2 (Plant Malfunctions) of this standard. Acceptance criteria for these tests shall:

(a) where applicable, be the same as plant startup test procedure acceptance criteria; (b) require that the observable change in the parameters correspond in direction to those expected from a best estimate for the simulated transient and do not violate the physical laws of nature; (c) require that the simulator shall not fail to cause an alarm or automatic action if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause an alarm or automatic action.

4.2.2 Malfunctions and transients not tested in accordance with 4.2.1 shall be tested and compared to best estimate or other available information and shall meet the acceptance criteria of 4.2.1(b).

ATTACHMENT 1 PAGE 2 OF 2

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 10 OF 11 ATTACHMENT 2 SIMULATOR REVIEW BOARD MEETING MINUTES DATE UNIT SIMULATOR ATTENDEES:

STATION OPERATIONS _

STATION TECHNICAL NTD, OPS TRG SUP NTD, SIM TRG ADM

NTD, SIM SUP ADM

NTD, MANAGER GUESTS Verification of Quorum:

Items Reviewed:

Meeting Minutes Accepted Chairman Simulator Review Board Date FACSIMILE FOR INFORMATION ONLY ATTACHMENT 2 PAGE 1 OF 1

f NUCLEAR TRAINING DIVISION S0123-XXI-3.1.1 SIMULATOR REVISION 2 PAGE 11 OF 11 ATTACHMENT 3 SIMULATOR REVIEW BOARD CERTIFICATION EVALUATION CERTIFICATION TESTING REPORTS

1.

VERIFICATION TESTS DUE THIS TEST PERIOD.

2.

VERIFICATION ITEMS REQUIRING RETEST.

3.

OVERDUE TEST ITEMS.

4.

COMPLETED vs REQUIRED TESTING COUNT.

5.

OPEN SWOs FROM ANSI TESTS.

SIMULATOR WORK STATUS

1.

SWO STATUS REPORT

2.

SWOs COMPLETED (TRNG REPORT)

3.

SWOs COMPLETED (HARDWARE)

4.

SWOs WRITTEN

5.

DCP REVIEW STATUS

6.

SWO/DCP MONTHLY REPORT REVIEW AND APPROVAL ITEMS

1.

REVIEW AND APPROVAL OF ANY BENCHMARK TESTS PERFORMED.

2.

REVIEW AND APPROVAL OF SIMULATOR TRANSIENT EVALUATION JUSTIFI CATIONS COMPLETED.

3.

REVIEW AND APPROVAL OF ANY DELETION(S) AND/OR REVISION(S) OF ANS 3.5 REQUIRED MALFUNCTIONS.

4.

REVIEW OF THE PRIORITY 4 & 6 SIMULATOR WORK ORDERS (SWOs)

GENERATED DURING THE TEST PERIOD.

5.

REVIEW HARDWARE DISCREPANCIES AND APPROVE JUSTIFICATIONS FOR THESE DISCREPANCIES.

6.

PERFORM ANNUAL REVIEW OF SIMULATOR CERTIFICATION PROGRAM DESCRIPTION.

ATTACHMNET 3 PAGE 1 OF 1

APPENDIX B

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 1 OF 17 EFFECTIVE DATE JAN 10, 1992 SIMULATOR VERIFICATION PROCEDURE PERFORMANCE TEST SCHEDULING TABLE OF CONTENTS SECTION PAGE 1.0 OBJECTIVES 2

2.0 REFERENCE 2

3.0 PREREQUISITES 2

4.0 PRECAUTIONS 2

5.0 CHECKOFF LIST 2

6.0 PROCEDURE 3

6.1 Simulator Operability Testing 3

6.2 Simulator Performance Testing 3

7.0 RECORDS 4

ATTACHMENTS 1

SIMULATOR OPERABILITY TESTING SCHEDULE -

UNIT 1 5

2 SIMULATOR PERFORMANCE TESTING SCHEDULE -

UNIT 1 6

3 SIMULATOR OPERABILITY TESTING SCHEDULE -

UNITS 2/3 12 4

SIMULATOR PERFORMANCE TESTING SCHEDULE -

UNITS 2/3 13 NOT QA PROGRAM AFFECTING

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 2 OF 17 SIMULATOR VERIFICATION PROCECEDURE PERFORMANCE TEST SCHEDULING 1.0 OBJECTIVE 1.1 To provide a four year simulator testing schedule.

2.0 REFERENCES

2.1 Simulator Verification Program.

2.2 Title 10, Chapter 1, Code of Federal Regulations, Part

55.

2.3 NUREG-1258, Evaluation Procedure for Simulation Facili ties Certified under 10CFR 55.

2.4 ANSI/ANS-3.5-1985, American National Standard, Nuclear Power Plant Simulators for Use in Operator Training.

2.5 Regulatory Guide 1.149, Nuclear Power Plant Simulation Facilities for use in Operator License Examinations.

2.6 S0123-XXI-3.3.1, Real Time Simulation Test.

2.7 S0123-XXI-3.3.2, Steady State Testing 2.8 S0123-XXI-3.3.4, Surveillance Testing 2.9 S0123-XXI-3.3.5, Normal Operations Testing 2.10 S0123-XXI-3.3.6, Transient Testing 2.11 S0123-XXI-3.3.7, Malfunction Testing 3.0 PREREQUISITES 3.1 Simulator performance test scheduling is required by reference 2.2, 10CFR55.

4.0 PRECAUTIONS 4.1 None 5.0 CHECKLIST(S) 5.1 None

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 3 OF 17 6.0 PROCEDURE 6.1 Simulator Operability Testing 6.1.1 Simulator operability testing shall be con ducted annually. The intent of this testing is to:

.1 Verify overall simulator model complete ness and integration;

.2 Verify simulator performance against steady state criteria;

.3 Verify simulator performance against transient criteria.

6.1.2 Operability testing shall be performed in accordance with Simulator Verification Proce dures, references 2.6 through 2.11.

6.1.3 Attachments 1 and 3 of this procedure list the testing required to be done annually.

6.2 Simulator Performance Testing 6.2.1 Simulator performance testing shall be per formed on a four year schedule.

Testing shall be performed in the following areas:

.1 Real Time Simulation test;

.2 Normal Operations testing;

.3 Surveillance testing;

.4 Malfunction testing.

6.2.2 Simulator performance testing shall be per formed in accordance with Simulator Verifica tion Procedures, references 2.6 through 2.11.

6.2.3 Approximately 25% of all required performance testing shall be performed each year.

6.2.4 Attachments 2 and 4 of this procedure list the required testing.

The number in the year tested column indicates which year in the four year cycle a specific test will be performed.

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 4 OF 17 7.0 RECORDS 7.1 Records shall be maintained in accordance with Simulator Verification Procedure S0123-XXI-3.1.4, Maintenance of Certification Records.

NUCLEAR TRAINING DIVISION SO123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 5 OF 17 ATTACHMENT 1 SIMULATOR OPERABILITY TESTING SCHEDULE UNIT ONE The following tests are performed annually on the Unit 1 simulator.

STEADY STATE PERFORMANCE TITLE DESCRIPTION

1.

SS01 SIMULATOR STABILITY 60 MINUTE STEADY STATE RUN

2.

SSO2 HEAT BALANCE @ 100%

RECORD STEADY STATE DATA

3.

SSO3 HEAT BALANCE @ 70%

RECORD STEADY STATE DATA

4.

SS04 HEAT BALANCE @ 30%

RECORD STEADY STATE DATA TRANSIENT PERFORMANCE TITLE DESCRIPTION

1.

TTO1 REACTOR TRIP REACTOR TRIP FROM 100%

2.

TTO2 LOSS OF MFW PUMPs SIMULTANEOUS TRIP OF ALL FEEDWATER PUMPS.

3.

TTO3 TRIP OF ALL RCPs SIMULTANEOUS TRIP OF ALL REACTOR COOLANT PUMPS.

4.

TTO4 TRIP SINGLE RCP TRIP OF A

SINGLE REACTOR COOLANT PUMP.

5.

TTO5 MAIN TURBINE TRIP MAIN TURBINE TRIP AT MAX.

POWER THAT DOES NOT CAUSE RX TRIP.

6.

TTO6 LOCA WITH LOSS OF PWR MAX SIZE REACTOR COOLANT SYSTEM RUPTURE COMBINED WITH LOSS OF ALL OFFSITE POWER.

7.

TTO7 MAIN STM LINE BREAK MAX SIZE UNISOLABLE STEAM LINE RUPTURE IN CONTAINMENT.

8.

TTO8 STUCK OPEN PORV SLOW PRIMARY SYSTEM DEPRESSURI ZATION TO SATURATED CONDITIONS USING STUCK OPEN PORV, HPSI INHIBITED.

9.

TTO9 MAX RATE PWR CHANGE MAX RATE OF POWER CHANGE FROM 100% TO 70% AND BACK TO 100%.

ATTACHMENT 1 PAGE 1 OF 1

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 6 OF 17 ATTACHMENT 2 SIMULATOR PERFORMANCE TESTING FOUR YEAR SCHEDULE UNIT ONE REAL TIME SIMULATION TEST TITLE/DESCRIPTION YEAR OF TEST

1.

RTTO1 COMPUTER REAL TIME TEST 4

NORMAL OPERATIONS No.

TITLE YEAR OF TEST

1.

NOP1 COLD SHUTDOWN TO HOT STANDBY 4

2.

NOP2 HOT STANDBY TO RATED POWER 1

3.

NOP3 RATED POWER TO HOT STANDBY 1

4.

NOP4 HOT STANDBY TO COLD SHUTDOWN 2

5.

NOP5 REACTOR TRIP FOLLOWED BY RECOVERY TO RATED POWER 3 SURVEILLANCE TESTING No.

TITLE YEAR OF TEST

1.

SSTO1 REACTOR THERMAL POWER CALIBRATION 4

2.

SST02 C/R SHIFT AND DAILY LOG READINGS 1

3.

SST03 CONTROL ROD POSITION VERIFICATION 4

4.

SST04 BORIC ACID FLOWPATH VERIFICATION 1

5.

SST05 MISC. TECH. SPEC. LEVEL SURVEILLANCE 4

6.

SST06 HOT OPERATIONAL TEST OF THE SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS 3

7.

SST07 DIESEL GENERATOR LOAD TEST 1

8.

SSTO8 TURBINE STOP VALVE TEST 2

9.

SST09 MONTHLY CONTROL ROD EXERCISE 4

ATTACHMENT 2 PAGE 1 OF 6

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 7 OF 17 ATTACHMENT 2 SURVEILLANCE TESTING (Continued)

No.

TITLE YEAR OF TEST

10.

SST10 MONTHLY SPHERE ISOLATION CH. TEST 3

11.

SST11 ACCIDENT MON. INSTR. CHANNEL CHECK 2

12.

SST12 REACTIVITY CALCULATIONS 1

13.

SST13 AUX. FEEDWATER SYSTEM FLOW TEST 3

14.

SST14 REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE 3

MALFUNCTION TESTING No.

TITLE YEAR OF TEST

1.

1.1 HOT LEG BREAK 1

2.

1.2 COLD LEG BREAK 2

3.

1.6 STEAM GENERATOR TUBE RUPTURE 3

4.

1.7 FUEL CLADDING FAILURE 2

5.

1.8 UNCONTROLLED BORATION 3

6.

1.9 UNCONTROLLED DILUTION 4

7.

1.10 CONTROL ROD EJECTION 4

8.

1.11 RCS FLOW TRANSMITTER FAILURE 2

9.

1.12 RCS TEMPERATURE ELEMENT FAILURE 3

10.

2.1 REACTOR COOLANT PUMP TRIP/FAIL TO TRIP 1

11.

2.2 REACTOR COOLANT PUMP LOCKED ROTOR 2

12.

2.4 REACTOR COOLANT PUMP SHEARED SHAFT 4

13.

3.2 PRESSURIZER PORV FAILURE 4

ATTACHMENT 2 PAGE 2 OF 6

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 8 OF 17 ATTACHMENT 2 MALFUNCTION TESTING (Continued)

No. TITLE YEAR OF TEST

14.

3.4 PRESSURIZER SAFETY VALVE FAILURE 2

15.

3.5 PRESSURIZER PRESSURE CONTROLLER FAILURE 3

16.

3.6 PRESSURIZER PRESSURE TRANSMITTER FAILURE 3

17.

3.7 PRESSURIZER LEVEL CONTROLLER FAILURE 4

18.

3.8 PRESSURIZER LEVEL TRANSMITTER FAILURE 4

19.

3.10 PRESSURIZER STEAM SPACE LEAK 1

20.

4.1 STEAM GENERATOR LEVEL CONTROLLER FAILURE 2

21.

4.2 STEAM GENERATOR LEVEL TRANSMITTER FAILURE 2

22.

4.4 FEED FLOW TRANSMITTER FAILURE 3

23.

4.5 STEAM FLOW TRANSMITTER FAILURE 3

24.

5.1 UNCOUPLED CONTROL ROD 4

25.

6.2 LETDOWN LEAK OUTSIDE CONTAINMENT 1

26.

6.5 CHARGING LEAK OUTSIDE CONTAINMENT 4

27.

6.14 VCT LEVEL TRANSMITTER FAILURE 4

28.

6.16 REACTOR MAKEUP CONTROL FAILURE 2

29.

6.17 BORIC ACID CONTROLLER FAILURE 3

30.

7.3 SAFETY INJECTION LEAKAGE INSIDE CONTAINMENT 3

31.

7.5 RWST LEVEL FAILURE 2

32.

7.7 SAFETY INJECTION LEAKAGE OUTSIDE CONTAINMENT 4

33.

8.1 RHR PUMP TRIPS/FAILS TO TRIP 1

34.

8.2 AIR ENTRAINMENT IN RHR PUMP SUCTION 1

35. 10.1 SALTWATER COOLING PUMP TRIPS/FAILS TO TRIP 1

ATTACHMENT 2 PAGE 3 OF 6

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 9 OF 17 ATTACHMENT 2 MALFUNCTION TESTING (Continued)

No. TITLE YEAR OF TEST

36. 10.2 COMPONENT COOLING WATER PUMP TRIPS/FAILS TO TRIP 2

37. 10.3 COMPONENT COOLING WATER SYSTEM LEAKAGE 3

38. 11.1 AUX FEED PUMP TRIPS/FAILS TO TRIP 4

39. 11.3 AUX FEED PUMP DISCHARGE PIPING LEAKAGE 3

40. 11.4 AUX FEED TANK LEVEL FAILURE 3

41. 11.5 AUX FEED PUMP STEAM BINDING 4

42. 12.1 REACTOR TRIP ACTUATES/FAILS TO ACTUATE 1

43. 12.5 TURBINE TRIP ACTUATES/FAILS TO ACTUATE 1

44. 12.6 ATWS -

REACTOR TRIP BREAKERS FAIL TO OPEN 2

45. 12.7 REACTOR PROTECTION SYSTEM CHANNEL FAILURE 3

46. 13.2 SOURCE RANGE DETECTOR FAILURE 3

47. 13.3 SOURCE RANGE FAILS TO DEENERGIZE 4

48. 13.4 NOISY SOURCE RANGE CHANNEL 1

49. 13.5 SOURCE RANGE STARTUP RATE FAILURE 1

50. 13.6 INTERMEDIATE RANGE DETECTOR FAILURE 3

51. 13.7 INTERMEDIATE RANGE STARTUP RATE FAILURE 4

52. 13.8 POWER RANGE DETECTOR FAILURE 1

53. 14.1 CONTROL ROD GROUP FAILS TO MOVE ON DEMAND 2

54. 14.2 CONTROL ROD GROUP MOVES IN WRONG DIRECTION 3

55. 14.3 CONTROL ROD GROUP MOVES CONTINUOUSLY 4

56.

14.5 CONTROL ROD DROPS 1

ATTACHMENT 2 PAGE 4 OF 6

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 10 OF 17 ATTACHMENT 2 MALFUNCTION TESTING (Continued)

No. TITLE YEAR OF TEST

57.

14.6 CONTROL ROD STUCK 1

58.

14.7 CONTROL ROD MASTER OR SLAVE CYCLER FAILURE 2

59.

15.1 GENERATOR TRIPS/FAILS TO TRIP 4

60.

15.8 MAIN GENERATOR LOSS OF LOAD 3

61.

16.1 LOSS OF OFFSITE POWER 3

62.

16.4 TRANSFORMER FAULT 1

63.

16.6 4160 VOLT BUS FAULT 3

64.

16.7 DIESEL GENERATOR TRIP 4

65.

16.8 DIESEL GENERATOR FAILS TO START/RUN/TRIP 1

66.

16.9 DIESEL GENERATOR OUTPUT BREAKER FAILURE 2

67.

16.11 480 VOLT BUS FAULT 4

68.

16.12 120 VOLT BUS FAULT 1

69.

16.13 125 VOLT DC BUS FAULT 2

70.

17.1 MAIN STEAM LINE BREAK UPSTREAM OF FLOW RESTRICTOR 2

71.

17.2 MAIN STEAM LINE BREAK DOWNSTREAM OF FLOW 3

RESTRICTOR INSIDE CONTAINMENT

72.

17.3 MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT 4

73.

17.4 STEAM GENERATOR SAFETY VALVE FAILURE 1

74.

17.5 STEAM DUMP VALVE FAILURE 2

75.

17.6 STEAM DUMP STEAM PRESSURE CONTROLLER FAILURE 3

76.

17.7 MAIN STEAM HEADER PRESSURE TRANSMITTER FAILURE 4

ATTACHMENT 2 PAGE 5 OF 6

NUCLEAR TRAINING DIVISION S0123-XXI-3.1.2 SIMULATOR REVISION 2 PAGE 11 OF 17 ATTACHMENT 2 MALFUNCTION TESTING (Continued)

No.

TITLE YEAR OF TEST

77.

18.2 MAIN TURBINE CONTROL VALVE FAILURE 2

78.

18.14 TURBINE 1ST STAGE PRESSURE TRANSMITTER FAILURE 2

79.

19.1 LOSS OF CONDENSER VACUUM 1

80.

19.2 CONDENSER TUBE LEAKAGE 2

81.

20.1 HOTWELL LEVEL CONTROLLER FAILURE 4

82.

21.3 MAIN FEEDWATER PUMP TRIPS/FAILS TO TRIP 1

83.

21.7 MAIN FEED MOTOR OPERATED ISOLATION VALVE FAILURE 1

84.

21.8 MAIN FEED REG VALVE/BYPASS VALVE FAILURE 2

85.

21.10 FEED BREAK INSIDE CNMNT UPSTREAM OF CHECK VALVE 4

86.

21.11 FEED BREAK INSIDE CNMNT DOWNSTREAM OF CHK VLV 1

87.

21.12 FEED BREAK OUTSIDE CNMNT DOWNSTREAM OF CHK VLV 2

88.

21.13 FEED BREAK OUTSIDE CNMNT UPSTREAM OF CHECK VALVE 3

89.

22.1 INSTRUMENT AIR COMPRESSOR TRIP 4

90.

22.2 LOSS OF INSTRUMENT AIR 1

ATTACHMENT 2 PAGE 6 OF 6