ML18016A866
| ML18016A866 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/31/1999 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18016A865 | List: |
| References | |
| NUDOCS 9903260290 | |
| Download: ML18016A866 (55) | |
Text
HARRIS NUCLEARPOWER PLANT OPERATOR TRAININGSIMJLATOR SIC( JLATOR CERTIFICATION QUADRENNIALREPORT MARCH 1999 CAROLINAPOWER 4 LIGHTCOMPANY NEW HILL,NORTH CAROLINA eeosiioieo iiosi9 PDR ADGCK 05000400' VDa L~
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SHNPP CERTIFICATIONREPORT PACKAGE TABLEOF CONTENTS FORM 474 INTRODUCTION General Information Simulator Configuration Control Exceptions to ANSUANS-3.5-1985 Standard 1.0 SIMULATORINFORMATION 1.1 Simulator General 1.1.1 Owner 1.1.2 Reference Plant/Unit 1.1.3 Simulator Supplier 1.1.4 Ready for Training Date 1.1.5 Type of Report 1.2 Simulator Control Room 1.2.1 Physical Arrangement 1.2.2 Panels and Equipment 1.2.3 Systems 1.2.4 Environment 1.3 Simulator Instructor Interface 1.3.1 General Instructor System 1.3.2 Initial Conditions 1.3.3 Malfunction Selection 1.3.4 Overrides 1.3.5 Local Operator Actions 1.3.6 Parameter and Equipment Monitoring 1.3.7 Simulator Special Features 1.4 Operating Procedures for Reference Plant 1.5 Changes Since Last Report 1.5.1 Plant Modifications 1.5.2 Simulator Upgrades Page 2 of 33
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SIMULATORCERTIFICATIONREPORT PACKAGE TABLEOF CONTENTS 2.0 SIMULATORDESIGN DATABASE 3.0 SIMULATORDISCREPANCY AND UPGRADE PROGRAM 3.1 Simulator Service Request Program 3.2 Engineering Service Request Implementation 3.3 Simulator Configuration Management System 4.0 SIMULATORTESTS 4.1 Certification Test Schedule 4.1.1 Annual Operability Tests 4.1.2 Malfunction Tests 4.2 Simulator Test Procedure to ANSI/ANS-3.5-1985 Cross-Reference 4.3 Summary of Certification Deficiencies 4.4 Certification Test Abstracts APPENDIX A:
APPENDIX B:
APPENDIX C:
APPENDIX D:
APPENDIX E:
SCHEDULE OF ANNUALOPERABILITYTESTS SCHEDULE OF MALFUNCTIONTESTS
SUMMARY
OF CERTIFICATIONDEFICIENCIES SIMULATORCERTIFICATIONTEST ABSTRACTS SCHEDULED MALFUNCTIONTEST TO ANSI 3.5 1985 CROSS REFERENCE Page 3 of 33
NRC FORM 474 (8.1998)
U.S. NUCLEAR REGULATORYCOMMISSION SIMULATIONFACILITYCERTIFICATION APPROVED BYOMBt NO. 31500138 EXPIRESt 08is li2001 Estimated burdon por rosponso to comply with thiS mandatory information collection roqvest:
120 hour* This information ls used to certify a simviatkxt facility. Forward comments regarding burdon estimate to tho Records Management Branch (TAF33), U.S.
Nuclear Regulatory CommMiion. Washingtcn, DC 205554001
~ and to tho Paperwork Reduction Project (31504138), Office of Managomont and Budgot. Washkigton, DC 205CL lfan Accmai'on coaoctke does not rgsptay a currently veld OMBccntrol number, tho NRC mayynot conduct or sponsor. and a porson ts not roqvirod to respond to, the Information colloction.
INBTRUOTIDNs: This form Is to bo Ekxf tor Initialceniyication. recortificadon gl required), and for any chango to a simulation fadlky performance testing plan made aker krklat submktal of such a lan. Provide the foo krformation and chock the a ate box to Indicate reason tor submittal.
FACIUTY Shearon Harris Nuclear Power Plant UCENSEE Carolina Power and Light Company DOCKET NUMBER so400 DATE 3/1 5/99 This is to certify thaL
- 1. lho above named facilityliconsoo Is using a simulation facilityconatstkig sotely of a plantuoforoncod sknvtator that moots tho roqukomonts of 10 CFA 55.45.
2.
Documentation Is ava8abio tor NRO review in accordanco with 10 cFA 55.45(b).
3.
This sknutatkxt facilitymoots tho guidance contained in ANSIJANS 3 5.1985 or ANSI/ANS 3 5.1993. as ondorsod by NRC Regulatory Guide 1.149.
ltthere aro any EXCEPTIONS to Iho certification ot this item CHECK HERE [ X ] and desc6be tullyon additional pages as necessary.
NAME (or other krentryrcatbn) AND LOCATIONOF SIMULATIONFACIUlY.
Harris Simulator - Harris Energy and Environmental Center 3932 New Hill-Holleman Road New Hill, North Carolina 27562-0327 SIMULATIONFACIUlYPERFORMANCE TEST ABSTRACTS ATTACHED. (For performance tests conducted h tire pen'od orxrng wirrrtire date offirJs cenifcatbra)
DESCRIPTION OF PERFORMANCE TESTING COMPLETED. (Attach addkbnaI pages as necessary and kfentlfythe lorn descrr)rtbn bang conrnued)
Abstracts for tests added since the 1995 Certification Quadrennial Report are attached. See Section 4.0, "Simulator Tests," and Appendix D, "Simulator Certification Test Abstracts."
slMULATIQNFAGIUTYpERFDRMANc6 TEsllNG scHEDULE ATTADHED. (For tire conduct olapproxrmereiy25 percent orperrormanco tests per year iortoo lour year perbd commoncng wkh fire dare oftrris ceciTeatJon.)
DEscRI pTIDNoF pERFoRMANGE TEsTING To BE coNDUOTED. (Attach addabnal pages as necessary and kbnliryfire flem descn'prbn behg contnued)
See Section 4.0, "Simulator Tests"; Appendix A, "Schedule of Annual Operability Tests"; and Appendix B, "Schedule of Malfunction Tests."
pERFQRMANGE TEsTING pLANcHANGE (For any modilcarbn to a performanr>> testhg pron suhmlted on a previous cert yicatbra)
DESCRIPTION OF PERFORMANCE TESTING PLAN CHANGE (Artarh addkknar pages as necessly and krentifythe item descrptbn behg conthuaf)
A complete, revised test plan is attached as Appendix A, "Schedule for Annual Operability Tests";
and Appendix B, "Schedule of Malfunction Tests". See Section 4.1, "Certification Test Schedule" for an explanation of the changes.
RECERTIFICATION (Descnbe corrective acrJons taken. anach resorts olcompiefed periormanco testhg h acc>>nfanco wrrrr 10 CFR 5545(b)(5)(v).
(Attach addsbnaf pages as necossary and identyythe kern descrrprbn behg conthued)
Any false statement or omission In this document, indvdng attachments, may bo subject to civiland criminal sanctions.
I cortifyundor penalty of porjury that the tnformadon In this document and attachmonts Is truo and corrocL SIGNATURE UTHORIZEDA P SENTATIVE Vice President - Harris Nuclear Plant DATF.
3 i'q accot with 10 CFA 55.5. Ccmmunlcations, this torm shall be submitted lo the NRC as follows:
YMAILADDRESSED TO:
DIRECTOR, OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORYCOMMISSION WASHINGTON DC 205554001 NRC FORM 474 (8 1998)
BY DEUVEAYIN PERSON TO THE NRC OFFICE AT:
ONE WHITEFUNT NORTH 11555 ROCKVILLEPIKE ROCKVII.LE,MO PAINTEDON RECYCLED PAPER
INTRODUCTION General Information The Shearon Harris Nuclear Power Plant Simulator Certification Quadrennial Report is provided to demonstrate compliance with the requirements of 10CFR55.45(b) including compliance with ANSI/ANS-3.5-1985 as implemented by NRC Regulatory Guide 1.149 Rev 1. The subject simulation facility consists solely of a plant reference full-scope simulator, which is the primary vehicle for providing positive, practical license training and examination.
An upgrade to the simulation computer system was completed approximately two months before this submittal to make the system Y2K compliant. The documentation contained herein is intended to constitute sufficient basis for retention of the certification of the Harris Simulator.
Simulator Confi uration Control A Simulator Review Group (SRG) is tasked with oversight of changes, potential enhancements, identified discrepancies, and proposed upgrades for implementation or resolution on the Harris Simulator.
The SRG is comprised of the Manager of Operations, Supervisor of License Operator Training, Lead Instructor for Operator Initial Training (OIT) and Licensed Operator Continuing Training (LOCT), and a Program Lead from Simulator Support (functioning as facilitator) or their designees.
Other training and plant operations personnel may also participate in SRG meetings as a function of the topics to be addressed.
Plant modifications are reviewed by a member of the Operator Training program.
Those with clear impact to the scope of simulation require no further review and are implemented.
Those changes with questionable impact are presented to the SRG for a training value assessment.
This SRG review ensures that differences between the plant and the simulator do not detract from training.
The SRG also reviews outstanding deficiencies for impact on training to ensure high priority items are properly scheduled for resolution.
The SRG provides guidance for scheduling discrepancy resolutions and modification implementations.
Exce tions to ANSI/ANS-3.5-1985 Standard Exceptions listed below, except for Exceptions ¹3 and ¹7, were identified at the time of the initial certification of the Harris Simulator's compliance with 10CFR55.45(b) stipulations. Exceptions ¹3 and ¹7 were identified in the 1995 quadrennial report. At those times the SRG reviewed the list of exceptions to ensure that the exception did not detrimentally impact the license operator training program and did not prevent 10CFR55 compliant simulator examinations (operating tests) from being conducted.
The exceptions identified in this section are listed by ANSI-3.5 reference and subject.
The justification for Page 4 of 33
each exception is included.
1.
ANS Section 3.1.1(7) Operations at Less than Full Reactor Coolant System (RCS) Flow This section is not applicable.
Power operations with less than three operating reactor coolant pumps is prohibited by Technical Specifications.
- However, the simulator is capable of such operations.
2.
ANS Section 3.1.1(9) Core Performance Testing Rod worth and reactivity coefficient measurement procedures were not performed as a part of the certification test program.
These tests are performed by Reactor Engineering, not Operations.
Tests which were conducted applicable to this section were Estimated Critical Conditions, Shutdown Margin, and Heat Balance.
3.
ANS Section 3.1.2(11) Protective System Channel Failures Protective system channel failures have been replaced by component overrides consisting of process instrumentation transmitter, relay, and bistable failures. This enhancement provides more credible failures for the student to diagnose or respond to.
The instructor has more explicit control over these devices than had been available through the deleted malfunctions.
4.
ANS Section 3.1.2(12) Control Rod Failures Drifting rods are not simulated as this type of failure is not relevant to the rod mechanisms used at the Harris Nuclear Plant.
5.
ANS Section 3.1.2(25) Reactor Pressure Control System Failure including Turbine Bypass Failure (BWR)
This item is specifically related to Boiling Water Reactors.
6.
ANS Section 3.2.1 Degree of Panel Simulation The Seismic Monitoring, Condensate Booster Pump, and Digital Metal Impact Monitoring Panels were not included in the simulation based on an assessment of the training value of having these panels.
Training in this area can be sufficiently accomplished utilizing the actual panels in the Harris Plant control room.
7.
ANS Section 3.2.3 Control Room Environment Page 5 of 33
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(Communications Systems) A telephone page system used at the plant to page outside operators was evaluated by the SRG and determined to be unnecessary in the simulation.
A normal telephone system that emulates the real control room exist for normal operator training and is replaced by a similar system of different colored phones when the simulator is used for Emergency Preparedness drills. A radio simulation is available as well as sound-powered phones. The SRG deemed the provided communications systems to be appropriate.
(Ceiling and Lighting)
The current ceiling is approximately twenty feet above the simulator panels rather than three feet as in the plant to facilitate visitor viewing of the simulator from above.
The lighting provides failure capability and emergency lighting to simulate electrical bus failures. The lighting configuration was altered due to ceiling height differences to provide light intensity level which approximates lighting levels in the plant control room.
c.
(Noise Levels)
Background noise levels in the simulator room is approximately that of the plant control room. A replacement of the simulator room HVACunits occurred since the 1995 Certification Report that established this match in background sound.
8.
ANS Section 4.1(3) Steady State Accuracy Tests (Critical Parameters)
ANS Section 4.1(4) Steady State Accuracy Tests (Non-Critical Parameters)
The criteria used for comparison between the simulator and plant parameters was 2 percent (10 percent for non-critical parameters) of the associated instrument loop range.
In addition, the parameter variation must not detract from training.
The standard states to use 2 percent (10 percent for noncritical parameters) of the reference plant parameter.
Using the percentage of instrument loop range is more limiting and more realistically represents the difference which can be noted by the operators.
This method was reviewed and approved by the SRG at the time of the original certification submittal.
9.
ANS Section Appendix B.1 BWR Simulator Operability Test This item is specifically related to Boiling Water Reactors.
10.
ANS Section Appendix B.2.1(2) Steady State Performance Steam generator temperature was not measured as this parameter is only applicable to once-through type steam generators.
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1.0 SIMULATORINFORMATION 1.1 Simulator General
1.1.1 Owner
1.1.2 Reference Plant/Unit:
1.1.3 Simulator Supplier:
1.1.4 Ready-for-Training Date:
1.1.5 Type of Report:
Carolina Power &Light Company Shearon Harris Nuclear Power Plant, Unit ¹I, Westinghouse 3-Loop PWR Westinghouse Electric Corporation with major upgrades by S3 Technologies (currently GSE Systems)
Initial December 20, 1985 Upgrade December 27, 1994 Year 2000 Upgrade January 11, 1999 Quadrennial (4-Year) Report 1.2 Simulator Control Room 1.2.1 Physical Arrangement The simulator control room is approximately 80 percent as large as the Harris Plant control room.
The simulated control room panels are the same size and color as found in the Harris Plant control room.
Some of the panels have been moved or angled slightly to accommodate space restrictions and the protrusion of the instructor station area into the simulator control room.
The simulated panels are in the same relative location as in the Harris Plant control room and provide the same visual perspective as in the plant.
The raised platform in the middle of the "at the controls" area is approximately 80 percent the size of the platform in the plant due to room size restrictions.
There are other minor differences with carpet color, location/style of handrails, type of furniture, and shape/size of status boards.
The differences have been reviewed and accepted by the Simulator Review Group.
1.2.2 Panels and Equipment Control room panels are included in the simulation except the Condensate Booster Pump Panel, Seismic Monitoring Panel, and the Digital Metal Impact Monitoring Panel.
The Reactivity Computer, which was only used by the reactor engineers at the time of refueling, has also been omitted. Connections for the portable device that they actually use during physics testing are available. These panels and equipment were omitted based on training value assessment.
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Classroom and on-the-job training are the means to provide training on these systems.
With the exception of the Emergency
Response
Facility Information System (ERFIS) peripherals, no panels outside the control room are included in the simulation facility.
Communications equipment capabilities essential to operator training and examination are provided in the simulation facility.
Telephone and radio communications terminate in the instructor station rather than various locations in the plant.
The instructor plays the role of appropriate plant personnel, interacts with the operating crew, and performs the local operator actions requested.
Dialed or automatic ring-down telephone calls made by the operating crew give a lighted indication in the instructor station as to who was the intended recipient of the call.
1.2.3 Systems
,Operative plant systems assessable from the control room are simulated except for Seismic Monitoring, Digital Metal Impact Monitoring, and Waste Processing.
These systems are omitted based on training value assessment.
1.2.4 Environment Some differences exist in the ceiling, and lighting between the simulator and the Harris Plant control rooms (see Exception
¹7). The simulator control room is designed to include a viewing platform for visitors to the Harris Energy and Environmental Center and an instructor station viewing area. This results in a difference between the simulator and main control room ceiling and lighting 1.3 Simulator Instructor Interface 1.3.1 General Instructor System The Harris Simulator has an instructor booth (or station) that is separated from the simulator control room and out of sight (one way mirrored glass) from the operator's view. The instructor is able to observe the actions of the operators in the simulator control room from the booth.
A multiple camera audio/video system is provided in the simulator facility to allow better analysis of operator activity. The audio/video system has been reviewed and accepted by the SRG as a no-training impact difference.
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The instructor has the capability of operating the simulator from the instructor's booth or from a terminal in the simulated control room.
Hand held remote operating controls are also available for inserting pre-planned simulation functions.
1.3.2 Initial Conditions (IC)
The simulation has storage for up to 200 Initial Condition sets. A controlled set of ICs are stabilized and re-snapped after each major simulator modification/upgrade period but prior to training restart.
These ICs contain a minimum of 3 power levels at 3 times in core life (BOL, MOL, and EOL), hot
- standby, and other primary training starting points selected to satisfy training objectives. Training Administrative Procedures provides a method of controlling simulator initial conditions.
1.3.3 Malfunction Selection The simulation contains capability to insert any number of discrete malfunctions individually or in combination.
The selection of malfunctions may be accomplished through command line
- entry, menu selection or available simulation dynamic PAIDs.
Malfunction severity, time of activation, and time to reach selected severity may be entered through the instructor system and modified as training objectives dictate.
Any number of malfunctions may be active at the same time.
Malfunctions may also be initiated based on specific plant conditions.
Deactivation and time delayed deactivation of malfunctions are also facilitated.
The current status of selected malfunctions is readily available to the instructor.
1.3.4 Simulator Overrides 1.3.4.1 Panel Overrides The instructor has the ability to override any simulated device on the control room panels.
For example, a meter may be driven to any
- value, a light may be turned off or on, or a switch may be failed closed.
The override may be inserted with a time delay, and analog values may be ramped in over a specified time band.
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k 1.3.4.2 Transmitter Overrides Most transmitters that have meters on the MCB or others may be overridden or failed to any value in it's range so that corresponding bistable trips and automatic actions willoccur.
The bistables may also be overridden directly.
As with malfunctions, the override may be ramped in over a specified time period.
This capability was expanded since the original certification submittal resulting in several of the previously certified malfunctions being no longer necessary.
1.3.4.3 Relay Overrides Selected relays may be overridden or failed to a specified state.
This capability was added since simulator certification and eliminated the need for related system malfunctions, some of which had been certified as a part of the original submittal.
1.3.4.4 Selection of Overrides The selection of overrides may be accomplished through command line
- entry, through a menu of available overrides or from dynamic system P&IDs.
1.3.5 Local Operator Actions (LOAs)
Local operator actions needed to provide training are available through the same selection methods as malfunctions and overrides. Plant procedures are reviewed to identify needed changes to these LOAs. Additional LOAs identified by training within the scope of simulation are added as needed.
1.3.6 Parameter and Equipment Monitoring The graphical capabilities of the instructor system facilitate visual monitoring of the simulation through dynamic P&IDs and panel mimic displays.
Plot capabilities for up to 400 parameters simultaneously is available through the instructor system.
The standard parameter versus time and X-Y plots are available along with the capability to trend against previously recorded trends, as is necessary to compare a previous test of simulator performance against the current simulator performance.
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1.3.7 Simulator Special Features Industry standard capabilities are available in the areas of switch check status/override, run, freeze, backtrack, replay, snapshot, fast time for certain parameters, slow time, Computer Aided Exercises, and simulation limitexceeded warnings.
Backtrack capabilities allow for four hours of storage at 2 minute intervals. The time between snaps of backtracks can be changed to lengthen or shorten this time. The capability for "nested" batch files allow multiple computer aided exercises to run concurrently, which facilitates simulation of a test (such as a maintenance surveillance test) being run on a system in the plant while other normal plant operations continue without required instructor interaction.
In compliance with ANSI/ANS-3.5 section 4.3, the simulator operating limits exceeded warning to the instructor exist and includes the following:
- Containinent Temperature > 400 degrees
- Containment Pressure > 60 psia
- RCS Pressure > 2700 psia
- Thermocouple Temperature > 2500 degrees
- Steam Generator Pressure > 1400 psia
- Steam Generator Steam Flow > 12.6 MPPH
-Core Power > 120%
- Condenser Pressure >.20 psia 1.4 Operating Procedures for the Reference Plant The Simulator Control Room utilizes a selected set of controlled procedures identical to those used in the Harris Plant control room.
1.5 Changes Since Last Report 1.5.1 Plant Modifications Numerous modifications to the plant have occurred since the last submittal which impact the simulator. The scope of the modifications was significantly less than in the previous certification cycle. Plant modifications continue to be reviewed for simulator and training impact. The more significant modifications are listed below:
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- Reactor core fuel cycles 7, 8 and 9
- ERFIS display system replacement (RTIN) 1.5.2 Simulator Upgrades One operating system upgrade has occurred since the last submittal. That upgrade was completed and the system declared "Ready For Training" on 1/11/99. This operating system upgrade to SGI IRIX 6.5 made the simulator computers Y2K compliant.
The operating system upgrade was followed by an extensive Performance Test performed in conjunction with the Annual Operability Test.
The peripheral computer systems such as ERFIS and the Radiation Monitoring System (RMS) are being addressed by the plant's Information Technology (IT) group's Y2Kcompliance plan.
With the 1994 GSE Systems upgrade graphic based code generator modeling tools were purchased.
Those tools have been used to build improved themo-hydraulic models of the following systems:
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AuxiliaryFeed-water (AFW)
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Pressurizer Spray
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Charging and Safety Injection System (SIS)
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Letdown Each of these system upgrades was followed by a specific series of performance test prior to being turned over to training.
1.5.2.1 Other Upgrades An ongoing effort to replace the generic controllers with system specific controllers continues. This effort allows for the use ofplant specific settings since the mathematics of the simulated controller matches that of the plant.
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2.0 SIMULATORDESIGN DATABASE The original simulator design data base consists of plant reference drawings (logics, CWDs, PEcIDs), FSAR, Plant Operating Manuals (POMs) including system descriptions, and system test results.
A complete set of these reference documents is available for use in simulator modification, troubleshooting, and updating.
The design data base was pre-start-up data.
Updated Harris Plant design data subsequently obtained is being used to perform simulator modifications. This design data is maintained as part of the Simulator Update Design Data.
Plant modification/change data have continued to be collected and analyzed for simulator applicability through formally controlled distribution of Engineering Service Requests (ESR's),
documentation updates, and plant procedure changes.
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3.0 SIMULATORDISCREPANCY AND UPGRADE PROGRAMS 3.1 Simulator Service Request Program Discrepancies noted in the simulator during testing or training sessions are documented by a Simulator Service Request (SSR). The SSR is used by the simulator staff to evaluate the problem and to identify corrective actions.
Documentation used to research the problems is attached to the SSR for inclusion as part of the Simulator Update Design Data Base.
3.2 Engineering Service Request Implementation Engineering Service Requests (ESRs) which are approved for work and which have the potential to impact the simulator, are reviewed by the staff concurrent with the plant review for applicability to the simulator. ESR's which are applicable to the scope of simulation are used to generate a Simulator Service Request.
Plant modification SSRs are scheduled to be completed in the simulator within twelve months of their operability in the plant. If requested by the plant operations staff, the modification may be performed in the simulator prior to its completion in the plant in order that the operators may be trained prior to plant modification completion.
This is particularly true for many modifications performed during a scheduled plant outage so as to be available for training operators prior to plant start-up.
The package is maintained as part of the simulator Update Design Data Base.
3.3 Simulator Configuration Management System The simulator Configuration Management System (CMS) is a PC-based management and design control system which is used to track the simulator's consistency with Harris Plant, performance or certification testing, modifications, and maintenance.
This system is used for recording and tracking plant changes and Simulator Service Requests.
Based on the relative importance of the modification or severity of the problem, a four-level schedule system is applied to the SSR or SMR.
This schedule is used to determine the order in which items are worked.
When SSRs are completed, their status is updated in the CMS computer.
The CMS computer is used to provide necessary reports as to the status of outstanding plant modifications and service requests.
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4.0 SIMULATORTESTS The simulator certification testing is carried out in accordance with the Simulator Certification test schedule.
The testing is typically accomplished by SRO licensed or certified individuals using test procedures developed by currently or previously licensed or certified personnel, engineers or others as appropriate. The tests were based on Harris Plant data, similar plant performance data, best estimate analysis, or a panel of experts. The selection of simulator performance test topics was determined based on ANSVANS-3.5-1985 requirements and a comprehensive review of the licensed operator training program.
Listed in Appendix C are those certification test deficiencies identified during testing that remain unresolved at the time of this report submittal.
4.1 Certification Test Schedule The test programs in place at Harris Nuclear Plant (HNP) for the past two certification cycles have exceeded the requirements of ANSI 3.5 1985 and Regulatory Guide 1.149 (1987). Because of improvements in the simulator model fidelity and simulator reliability this additional testing is no longer considered necessary.
Accordingly, the annual test program was reduced in scope but continues to meet the requirements of the applicable standards and guides.
4.1.1 Annual Operability Tests With this submittal, the requirements of ANSI 3.5 1985 were reviewed and the total number of annual operability test was reduced to those test specified in ANSI 3.5 1985 Appendix B. The annual operability tests program now includes the simulator Real Time Test, the Steady State Stability and Accuracy Tests, and the Transient Tests.
These tests are listed in Appendix A.
4.1.2 Malfunction Tests Malfunctions and component overrides available on the simulator and incorporated in the operator training program have been formally tested via an individual performance
- test, typically at the time of inclusion. In addition, scenario validation performed at the time that a scenario is added verifies that the malfunctions and component failures function as expected. With this submittal, the requirements of ANSI 3.5 1985 were reviewed and the total number of malfunction test to be performed annually was reduced.
Only the malfunctions included in the simulation to meet the requirements of ANS/ANSI 3.5 -1985 Section 3.1.2 are tested on a periodic basis. These tests are scheduled such that approximately 25 percent of these required malfunctions are tested each year.
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The number of performance tests willbe adjusted as malfunctions are added to or deleted from the certification test program as dictated by operator training program requirements.
These additions and/or deletions will be noted in subsequent quadrennial reports, however, the test program will be maintained in compliance with ANSI/ANS-3.5-1985.
Appendix B lists the malfunctions which meet this requirement and schedule for performing them over the next four years.
4.2 Simulator Test Procedure to ANSI/ANS-3.5-1985 Cross-Reference Appendix E reflects those malfunctions and malfunction test used to show compliance to the ANSI 3.5 1985 section 3.1.2 list of required malfunctions.
4.3 Summary of Certification Deficiencies Certification deficiencies are listed in Appendix C to this report. To be listed in this
- appendix, test results must be identified as either "Satisfactory with Deficiencies" or "Unsatisfactory".
Deficiencies against these tests will be resolved based on training impact in accordance with the four-level scheduling system outlined in Harris Training Administrative Procedures.
There are no tests identified as "Unsatisfactory" at this time.
4.4 Certification Test Abstracts Abstracts of the certification tests were included in the original certification submittal or subsequent quadrennial reports.
Abstracts for those new tests identified in the above appendix are attached to this report.
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APPENDIXA SCHEDULE OF ANNUALOPERABILITYTESTS The following tests are performed on an annual basis.
Real Time Test RTI'-001 Computer Real Time Test R~
SST-001 100 Percent Power and Accuracy Test SST-002 75 Percent Power Accuracy Test SST-003 30 Percent Power Accuracy Test Transient Tests TT-001 Manual Reactor Trip Tl'-002 Simultaneous Trip of all Feedwater Pumps TT-003 Simultaneous Closure ofAllMain Steam Isolation Valves Ti'-004 Simultaneous Trip ofAllReactor Coolant Pumps TT-005 One Reactor Coolant Pump Trip TT-006 Turbine Trip Below P-10 TT-007 Maximum Rate Power Ramp TT-008 Maximum Size RCS Leak Inside Containment With Loss of Off-site Power TT-009 Maximum Size Steam Leak Inside Containment TT-010 Slow RCS Depressurization to Saturation Using PORV's and No SI Page 17 of 33
APPENDIX B SCHEDULE OF MALFUNCTIONTESTS MALFUNCTIONTESTS FIRST YEAR TEST NUMBER TEST TITLE ANSI 3.5 1985 Reference MT-1042 MT-111A MT-12 MT-1222 MT-1231 MT-135 MT-42 MT-44 MT-51 MT-61 MT-623 MT-710 MT-724 MT-86 MT-97 MT-MSC3 RCS-18 Reactor Trip Breakers Fail (B fails to open)
Pressurizer Steam Space Leak NSW Pump Trip and Loss of NSW Steam Generator Tube Rupture (S/G A)
RCP Trip From 100 Percent Power (RCP-C)
RHR Bypass Line Leak (Train A)
Logic Cabinet Urgent Failure Stuck Rod Letdown Isolation Valve Failure (1CS-11)
Station Blackout Loss of 120-VAC Uninterruptible Power (Power Supply SIII)
Condensate Pump Trip (Pump A)
Feedline Break Outside Containment Steam Generator Relief Valve Failure (Open)
Power Range Channel Detector Failure (Low)
Annunciator System Failure Small Break LOCA 3.1.2(24) 3.1.2(lc) 3.1.2(6) 3.1.2(la) 3.1.2(4) 3.1.2(7) 3.1.2(13) 3.1.2(12) 3.1.2(23) 3.1.2(3) 3.1.2(3) 3.1.2(9) 3.1.2(20) 3.1.2(20) 3.1.2(21) 3.1.2(22) 3.1.2(lc)
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MALFUNCTIONTESTS SECOND YEAR TEST NUMBER TEST TITLE ANSI 3.5 1985 Reference MT-1032 MT-1041 MT-112 MT-1132 MT-114 MT-1211 MT-1212 MT-1214A MT-1232 MT-151 MT45 MT-571 MT-651 MT-67 MT-712A MT-719 MT-723 MT-82 Safety Injection Failure (Train A, Fail to Initiate)
Reactor Trip Breakers Fail (Both Inadvertent Open)
Loss of Instrument Air to the Reactor (Reactor Auxiliary Building)
Pressurizer Relief Valve Failure (444B Without P-11 Interlock)
Pressurizer Safety Valve Failure (8010C Open)
RCS Leak Within Capacity of Charging Pumps LOCA Within Capacity of the SI Pumps LOCA on RHR Reactor Coolant Pump Trips (RCP-C)
Inadvertent Turbine Trip Ejected Rod Letdown Pressure Control Valve Failure (PK-145 Open)
Loss of 6.9-KV Emergency Bus (1A-SA)
Diesel Generator Failure Turbine Driven AuxiliaryFeedwater Pump Trip Main Feedwater Pump Trip (Pump B)
Feedline Break Inside Containment Steam Break Outside Containment 3.1.2(23) 3.1.2(19) 3.1.2(2) 3.1.2(18) 3.1.2(1d) 3.1.2(lc) 3.1.2(1c) 3.1.2(17) 3.1.2(4) 3.1.2(15) 3.1.2(12) 3.1.2(18) 3.1.2(3) 3.1.2(3) 3.1.2(23) 3.1.2(10) 3.1.2(20) 3.1.2(20)
Page 19 of 33
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MALFUNCTIONTESTS THIRDYEAR TEST NUMBER TEST TITLE ANSI 3.5 1985 Reference MT-1072 MT-1110 MT-113
'T-1211B MT-1213A MT-1221 MT-136 MT-333 MT-34 MT-612 MT-68 MT-711A MT-76 MT-81 MT-815 MT-91 RCS-6 Turbine Runback Failure (Failure to Runback)
Pressurizer Level Control Band Shift Down Loss of Instrument Airto the Containment Building Uncoupled Control Rod RCS Leak (Large Break)
Steam Generator Tube Leak (S/G B)
RHR Sump Valves Fail to Open Hotwell Level Controller Failure (LC-1901 Low)
Loss of Condenser Vacuum Generator Output Breakers Fail to Trip Automatic Voltage Regulator Failure (High)
Motor Driven AuxiliaryFeedwater Pump Trip AuxiliaryFeedwater Flow Control Valve Failure (Open)
Steamline Break Inside Containment Main Steam Header Break Source Range Instrument Failure (N31 High)
Median Select Circuit Failure 3.1.2(22) 3.1.2(22) 3.1.2(2) 3.1.2(12) 3.1.2(1b) 3.1.2(1a) 3.1.2(23) 3.1.2(5) 3.1.2(5) 3.1.2(16) 3.1.2(16) 3.1.2(10) 3.1.2(23) 3.1.2(20) 3.1.2(20) 3.1.2(21) 3.1.2(22)
Page 20 of 33
MALFUNCTIONTESTS FOURTH YEAR TEST NUMBER TEST TITLE ANSI 3.5 1985 Reference MT-1015 MT-111 MT-112A MT-1211A MT-1212A MT-21 MT-22A MT-28 MT-331 MT-431 MT-512 MT-616 MT-912 CRF-16 CVC-30 MSC-4 SWS-7 Diesel Generator Sequencer Fails to Complete Block 1 Loss of Instrument Air(Turbine Building)
Pressurizer Spray Valve Failure RCS Fuel Rod Breach RCS Leakage into an Accumulator Component Cooling Water Pump Trip Loss of CCW to RHR Heat Exchanger Loss of CCW to the Reactor Coolant Pumps Hotwell Level Controller Failure (LC-1900 High)
Dropped Rod (One Rod)
RCP Number 1 Seal Failure (RCP B)
Diesel Generator Breaker Inadvertent Trip Intermediate Range Control Power Fuse Blown Control Rod Stuck on Trip (NEW)
Charging/Safety Injection Pump Speed Changer Failure (NEW)
Inadvertent Containment Isolation Phase A Normal Service Water Pump Shaft Shear (NEW) 3.1.2(23) 3.1.2(2) 3.1.2(18) 3.1.2(14) 3.1.2(23) 3.1.2(8) 3.1.2(8) 3.1.2(8) 3.1.2(5) 3.1.2(12) 3.1.2(8) 3.1.2(3) 3.1.2(21) 3.1.2(12) 3.1.2(23) 3.1.2(22) 3.1.2(6)
Page 21 of 33
APPENDIX C
SUMMARY
OF CE<RTIFICATIONDEFICIENCIE<S Performance tests were run as a part of the current (March 1995 - March 1999) certification testing program.
The resulting deficiencies that remain unresolved at this time are shown below.
TEST/RE<SULTS CMS/DR ¹ TITLE</DE<SCRIPTION MT-1211A 98-261 RCS Fuel Rod Breach The acceptance criteria is 1000X normal and 800X normal was achieved.
Efforts are ongoing to determine if the acceptance criteria is correct for a best estimate situation vice a worst case situation of the normal EP environment.
Page 22 of 33
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APPENDIX D SIMULATORCERTIFICATIONTEST ABSTRACTS This appendix contains a complete list (index) of the performance test performed per schedule, in the previous certification period. Abstracts of these test were included in the previous certification submittals and are not being included in this submittal unless it is one of the tests added in the past four years.
INDEXOF ABSTRACTS Simulator Ph sical Fidelit Test (2)
FT-001 FT-002 Simulator Physical Fidelity Test Simulator Model Limits Exceeded Test Malfunction Tests (180)
MT-1013 MT-1014 MT-1015 MT-10161 MT-10162 MT-10165 MT-1017 MT-1031 MT-1032 MT-1041 MT-1042 MT-106 MT-1071 MT-1072 MT-111 MT-1110 MT-111A MT-112 MT-112A MT-113 MT-1131 MT-1132 MT-114 MT-1151 MT-1152 MT-1162 Inadvertent Feedwater Isolation Inadvertent Main Steam Isolation Diesel Generator Sequencer Fails to Complete Block 1 Failure ofRod Blocks to Block (C-1)
Failure ofRod Blocks to Block (C-2, C-3, C-4)
Failure of Rod Block to Block (C-5)
Failure of Permissive Interlock P-14 Safety Injection Failure (Train B, Inadvertent)
Safety Injection Failure (Train A, Fail to Initiate)
Reactor Trip Breakers Fail (Both Inadvertent Open)
Reactor Trip Breakers Fail (B fails to open)
False Containment Spray Actuation Turbine Runback Failure (Erroneous Runback)
Turbine Runback Failure (Failure to Runback)
Loss of Instrument Air(Turbine Building)
Pressurizer Level Control Band Shift Down Pressurizer Steam Space Leak Loss of Instrument Airto the Reactor (Reactor AuxiliaryBuilding)
Pressurizer Spray Valve Failure Loss of Instrument Airto the Containment Building (AIR-1,1)
Pressurizer Relief Valve Failure (445A With P-11 Interlock)
Pressurizer Relief Valve Failure (444B Without P-11 Interlock)
Pressurizer Safety Valve Failure (8010C Open)
Pressurizer Pressure Channel Failure (PT~ High)
Pressurizer Pressure Channel Failure (PTM5 Low)
Pressurizer Pressure Channel Failure (PT-457 Low)
Page 23 of 33
MT-117'1 MT-118 MT-12 MT-121 MT-1210 MT-1211 MT-1211A MT-1211B MT-1212 MT-1212A MT-1213 MT-1213A MT-1214 MT-1214A MT-1215 MT-1216 MT-1221 MT-1222 MT-1231 MT-1232 MT-124 MT-125 MT-126 MT-1281 MT-129 MT-131 MT-131A MT-1321 MT-1322 MT-1331 MT-1332 MT-134 MT-135 MT-136 MT-137 MT-138 MT-141 MT-151 MT-152 MT-154 MT-155 Pressurizer Level Channel Failure (LTC59 Low)
Pressurizer Backup Heaters Groups A and B Failure NSW Pump Trip and Loss of NSW Emergency Service Water Pump Trip RCP A, B, C High Vibration RCS Leak Within Capacity of Charging Pumps RCS Fuel Rod Breach Uncoupled Control Rod LOCA Within Capacity of the SI Pumps RCS Leakage into an Accumulator RCS Vessel Flange Leak RCS Leak (LOCA)
RCP Bearing Oil Reservoir Leak LOCA on RHR RCS Thermal Barrier Leak into CCW System RCS Flow Transmitter Failure (FT-436 w)
Steam Generator Tube Leak (S/G B)
Steam Generator Tube Rupture (S/G A)
RCP Trip From 100 Percent Power (RCP-C)
Reactor Coolant Pump Trips (RCP-C)
Reactor Coolant Pump Trip (Locked Rotor)
RCP Shaft Break Accident (RCP B)
RCS Boron Dilution RCS Protection RTD Failure (TE-412B Low)
RCS WR Pressure Transmitter Failure (PT-403 High)
RHR Pump Trip (Pump A)
RHR Pump Trip (Pump A)
RHR HX Flow Control Valve Failure (FCV-603A Closed)
RHR HX Flow Control Valve Failure (FCV-603B Open)
RHR HX Bypass FCV Failure (FK-605A1 Open)
RHR HX Bypass FCV Failure (FK-605B1 Closed)
RHR to Letdown Valve Failure (HCV-142.1 Open)
RHR Bypass Line Leak (Train A)
RHR Sump Valves Fail to Open Containment Spray Pump Failure Containment Spray Pump Discharge Valve Failure Containment Fan Cooler Unit Trip Inadvertent Turbine Trip Turbine Protection Trip Failure Turbine Vibration Governor Valve Failure (GV-3 Closed)
Page 24 of 33
I MT-157 MT-17 MT-21 MT-210 MT-22 MT-22A MT-23 MT-24 MT-25 MT-26 MT-271 MT-272 MT-28 MT-31 MT-32 MT-331 MT-333 MT-34 MT-35 MT-41 MT-410 MT-411 MT-412 MT-413 MT-42 MT431 MT'T'T-461 MT462 MT-47 MT48 MT'T-51 MT-5111 MT-512 MT-513 MT-514 MT-5151 MT-5152 MT-516 Turbine DEH Computer Failure Refueling Water Storage Tank Leak Component Cooling Water Pump Trip Seal Water Heat Exchanger Tube Leak Loss of CCW to RHR Heat Exchanger Loss of CCW to RHR Heat Exchanger CCW Leak into the Service Water System Component Cooling Water Header Supply Valve Failure (Closed)
Letdown Heat Exchanger Tube Leak Loss of CCW to RCP Thermal Barrier Letdown Temperature Controller Failure (TK-144 Low)
Letdown Temperature Controller Failure (TK-144 High)
Loss of CCW to the Reactor Coolant Pumps Circulating Water Pump Trip Main Condenser Tube Leak Hotwell Level Controller Failure (LC-1900 High)
Hotwell Level Controller Failure (LC-1901 Low)
Loss of Condenser Vacuum Loss of Condenser Vacuum Pump Power Cabinet Urgent Failure DRPI-Open or Shorted Coil Improper Bank Overlap Control Bank Rod Step Counter Failure Rod Speed Deadband Control Failure Logic Cabinet Urgent Failure Dropped Rod (One Rod)
Stuck Rod Ejected Rod Uncontrolled Automatic Rod Motion Uncontrolled Manual Rod Motion Failure of Auto Rod Blocks to Block (C-11)
TREF Failure DRPILoss ofVoltage Letdown Isolation Valve Failure (1CS-11)
VCT Level Transmitter Failure (LT-112 High)
RCP Number 1 Seal Failure (RCP B)
RCP Number 2 Seal Failure (RCP A)
RCP Number 3 Seal Failure (RCP C)
Boric Acid Flow Xmtr. Failure (FT-113 to 20 gpm)
Boric Acid Flow Xmtr. Failure (FT-113 to 0 gpm)
Boric Acid Filter Plugged Page 25 of 33
MT-'518'1 MT-5182 MT-52 MT-5201 MT-5202 MT-523 MT-524 MT-525 MT-526 MT-527 MT-5281 MT-5282 MT-5283 MT-5284 MT-5285 MT-53 MT-54 MT-55 MT-56 MT-571 MT-572 MT-58 MT-59 MT-61 MT-6101 MT-6102 MT-612 MT-615 MT-616 MT-623 MT-632 MT-64 MT-645 MT-651 MT-661 MT-662 MT-67 MT-68 MT-69 MT-692 MT-710 Seal Injection Flow Control Valve Failure (HC-186 Open)
Seal Injection Flow Control Valve Failure (HC-186 Closed)
VCT Outlet Isolation Valve Failure (LCV-115E Closed)
Failure of Charging Flow Control Valve Failure of Charging Flow Control Valve (Closed)
High Temperature Divert Valve (TCV-143) Failure Charging Pump Suction From RWST Failure (115D Open)
Charging Pump MiniFlow Valve Failure (1CS-182 Closed)
Boric Acid Pump Trip Charging Line Containment Isolation Valve Failure Charging Line Leak on Charging Pump Suction Charging Pump Discharge Line Leak Before FT-122 Charging Line Leak Between FT-122 and 1CS-235 Charging Line Leak in Containment Before Regen HX Charging Line Leak Between Regen HX and 1CS-492 Letdown Line Leak Inside Containment Letdown Line Leak Outside Containment Charging Pump Trip Reactor Makeup Water Pump Trip Letdown Pressure Control Valve Failure (PK-145 Open)
Letdown Pressure Control Valve Failure (PK-145 Closed)
Loss of Normal Letdown VCT Divert Valve Control Failure (HUT)
Station Blackout Loss of Unit AuxiliaryTransformer A phase Loss of Unit AuxiliaryTransformer B phase Generator Output Breakers Fail to Trip Diesel Generator Governor Failure Diesel Generator Breaker Inadvertent Trip Loss of 120-VAC Uninterruptible Power (Power Supply SIII)
Loss of 1&-VDCEmergency Bus (DP 1B-SB)
Loss of 6.9 KVAuxiliaryBus (1B)
Loss of 6.9 Aux Bus 1E Loss of 6.9-KV Emergency Bus (1A-SA)
Loss of a 250-VDC Nonvital Bus (DP-1-250)
Loss of a 125-VDC Nonvital Bus (DP 1A)
Diesel Generator Failure Automatic Voltage Regulator Failure (High)
Loss ofStart-up Transformer 1A Loss ofStart-up Transformer 1B Condensate Pump Trip (Pump A)
Page 26 of 33
MT-'711A MT-712 MT-712A MT-714 MT-715 MT-719
. MT-72 MT-720 MT-721 MT-722 MT-723 MT-724 MT-725 MT-73 MT-74 MT-76 MT-771 MT-772 MT-78 MT-81 MT-810 MT-811 MT-812 MT-814 MT-815 MT-82 MT-83 MT-84 MT-85 MT-86 MT-87 MT-88 MT-89 MT-91 MT-911 MT-912 MT-913 MT-92 MT-93 MT-94 MT-95 Motor Driven AuxiliaryFeedwater Pump Trip Failure of Excess Condensate Dump Valve (Closed)
Turbine Driven AuxiliaryFeedwater Pump Trip Condensate Storage Tank Leak Heater Drain Pump Trip (Pump B)
Main Feedwater Pump Trip (Pump B)
Condensate Booster Pump Trip (Pump B)
Main Feedwater Pump Recirc Valve Failure (Pump 1B)
Feedwater Flow Transmitter Failure (FT466 Low)
Feedwater Control Valve Position Failure (LCV-488 Open)
Feedline Break Inside Containment Feedline Break Outside Containment Steam Generator Level Chan. Failure (LT-496 Low)
Turbine Driven AuxiliaryFeedwater Pump Speed Control Oscillates Steam Generator Backleakage AuxiliaryFeedwater Flow Control Valve Failure (Open)
Feedwater Bypass Valve Failure (Closed)
Feedwater Bypass Valve Failure (Open)
Turbine Driven AuxiliaryFeedwater Flow Control Valve Failure (Closed)
Steamline Break Inside Containment Steam Dump Control Failure (Closed)
Mechanically Stuck Condenser Dump Valve (PCVA08 Open)
Steam Dump Permissive (P-12) Failure Steam Failure to TDAFW Pump (1MS-72 Closed)
Main Steam Header Break Steam Break Outside Containment Steam Header Press. Detector Failure (PT-464 High)
Steam-Line Flow Transmitter FT-494)
Steam Generator Press. Xmtr. Failure (PT-485 High)
Steam Generator Relief Valve Failure (Open)
MSIV Failure (S/G B Shut)
Steam Generator Safety Valve Failure (Open)
Atmospheric Steam Dump Valve Failure (PCV408J Open)
Source Range Instrument Failure (N31 High)
Source Range Instrument Power Fuse Blown Intermediate Range Control Power Fuse Blown Power Range Control Power Fuse Blown Source Range Pulse Height Discriminator Failure Failure of Source Range High Voltage to Disconnect Source Range Channel High Voltage Failure Intermediate Range Channel Failure Page 27 of 33
I MT-'96 MT-97 MT-98 MT-MSC3 MT-RCS-18 MT-RCS-6 MT-RPS4 MT-CRF16 MT-CVC29 MT-CVC30 MT-DSG5 MT-HVA4 MT-SWS4 MT-SWSS MT-SWS6 MT-SWS7 Intermediate Range Channel Gamma Compensation Failure Power Range Channel Detector Failure (Low)
Power Range Channel Failure (Low)
Annunciator System Failure Small Break LOCA Median Select Circuit Failure Inadvertent Containment Isolation Phase B Control Rod Stuck on Trip (NEW)
CSIP Shaft Shear (NEW)
CSIP Speed Changer Failure (NEW)
Diesel Generator Emergency Trip (NEW)
Essential Services Chiller Trip Service Water Discharge Valve Fails to Open (NEW)
B NSW Pump Fails to Auto Start (NEW)
SW From Containment Fan Coolers Back Pressure Valve Failure (NEW)
NSW Pump Shaft Shear (NEW)
Page 28 of 33
Normal 0 erator Surveillance Tests (25)
NOST-1004 NOST-1005 NOST-1007 NOST-1008 NOST-1009 NOST-1013 NOST-1014 NOST-1018 NOST-1021 NOST-1022 NOST-1026 NOST-1036 NOST-1039 NOST-1046 NOST-1054 NOST-1073 NOST-1075 NOST-1076 NOST-1080 NOST-1087 NOST-1092 NOST-1126 NOST-1211 NOST-1316 NOST-1411 OST-1004, Power Range Heat Balance OST-1005, Control Rod and Rod Position Exercise OST-1007, CVCS/SI System Operability OST-1008, RHR Pump Operability OST-1009, Containment Spray Operability OST-1013, 1A-SA Emergency Diesel Generator Operability OST-1014, Turbine Valve Test OST-1018, Main Steam Isolation Valve and Main Feedwater Isolation Valve Operability Test OST-1021, Daily Surveillance Requirements Modes 1 and 2 OST-1022, Daily Surveillance Requirements Modes 3 and 4 OST-1026, Reactor Coolant System Leakage Evaluation OST-1036, Shutdown Margin Calculation OST-1039, Calculation of Quadrant Power TiltRatio OST-1046, Main Steam Isolation Valve Operability Test OST-1054, Permissives P-6 and P-10 Verification OST-1073, 1B-SB Emergency Diesel Generator Operability OST-1075, Turbine Mechanical Overspeed Trip Test OST-1076, AFW Pump 1BCB Operability Test - Quarterly OST-1080, Turbine Driven AFW Pump Full Flow Test OST-1087, Motor Driven AFW Pumps Flow Test OST-1092, RHR Pump 1B-SB Operability OST-1126, Reactor Coolant Pump Seals Controlled Leakage Evaluation OST-1211, AFW Pump 1A-SA Operability Test - Quarterly OST-1316, CCW System Operability - Quarterly OST-1411, AFW Pump 1X-SAB Operability Normal 0 erations Tests (9)
NOT-001 NOT-002 NOT-003 NOT-004 NOT-005 NOT-006 NOT-007 NOT-008 NOT-009 GP-001, Plant Fill and Vent GP-002, Plant Heatup Recovery to Rated Power Following Reactor Trip GP-004, Reactor Startup GP-005, Plant Startup GP-006, Plant Shutdown GP-007, Plant Cooldown GP-008, Plant Drain to Mid-Loop GP-009, Refueling with Cavity Fill and Drain page 29 of 33
Real Time Test (1)
RTT-001 Computer Real Time Test SST-001 SET-002 SST-003 100 Percent Power Accuracy Test 75 Percent Power Accuracy Test 30 Percent Power Accuracy Test Transient Tests (10)
TT-001 TT-002 TT-003 TT-004 TI'-005 TT-006 TT-007 Tl'-008 TT-009 TT-010 Manual Reactor Trip Simultaneous Trip of all Feedwater Pumps Simultaneous Closure ofAllMain Steam Isolation Valves Simultaneous Trip ofAllReactor Coolant Pumps One Reactor Coolant Pump Trip Turbine Trip Below P-10 Maximum Rate Power Ramp Maximum Size RCS Leak Inside Containment With Loss of Off-site Power Maximum Size Steam Leak Inside Containment Slow RCS Depressurization to Saturation Using PORV's and No SI Page 30 of 33
APPE<NDIX E Scheduled Malfunction Test to ANSI 3.5 1985 Cross Reference This appendix contains a list (index) of the malfunction test to be performed in the upcoming certification cycle with a cross reference to the specific ANSI 3.5 1985 requirement.
Abstracts of these test were included in the previous certification submittals and are not being included in this submittal unless it is one of the tests added in the past four years.
ANSI 3.5 1985 Reference TEST TITLE Test Number 3.1.2(la) 3.1.2(la) 3.1.2(lb) 3.1.2(lc) 3.1.2(lc) 3.1.2(lc) 3.1.2(lc) 3.1.2(ld) 3.1.2(2) 3.1.2(2) 3.1.2(2) 3.1.2(3) 3.1.2(3) 3.1.2(3) 3.1.2(3) 3.1.2(3) 3.1.2(4) 3.1.2(4) 3.1.2(5) 3.1.2(5) 3.1.2(5) 3.1.2(6) 3.1.2(6) 3.1.2(7) 3.1.2(8)
Steam Generator Tube Rupture (S/G A)
Steam Generator Tube Leak (S/G B)
RCS Leak (Large Break)
Small Break LOCA RCS Leak Within Capacity of Charging Pumps LOCA Within Capacity of the SI Pumps Pressurizer Steam Space Leak Pressurizer Safety Valve Failure (8010C Open)
Loss of Instrument Airto the Containment Building Loss of Instrument Air to the Reactor (Reactor Auxiliary Building)
Loss of Instrument Air(Turbine Building)
Station Blackout Loss of 120-VAC Uninterruptible Power (Power Supply SIII)
Loss of 6.9-KV Emergency Bus (1A-SA)
Diesel Generator Failure Diesel Generator Breaker Inadvertent Trip RCP Trip From 100 Percent Power (RCP-C)
Reactor Coolant Pump Trips (RCP-C)
Hotwell Level Controller Failure (LC-1901 Low)
Loss of Condenser Vacuum Hotwell Level Controller Failure (LC-1900 High)
NSW Pump Trip and Loss of NSW Normal Service Water Pump Shaft Shear (NE<W)
RHR Bypass Line Leak (Train A)
Component Cooling Water Pump Trip MT-1222 MT-1221 MT-1213A RCS-18 MT-1211 MT-1212 MT-111A MT-114 MT-113 MT-112 MT-111 MT-61 MT-623 MT-651 MT-67 MT-616 MT-1231 MT-1232 MT-333 MT-34 MT-331 MT-12 SWS-7 MT-135 MT-21 Page 31 of 33
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3.1.2(8) 3.1.2(8) 3.1.2(8) 3.1.2(9) 3.1.2(10) 3.1.2(10) 3.1.2(1 1) 3.1.2(12) 3.1.2(12) 3.1.2(12) 3.1.2(12) 3.1.2(12) 3.1.2(13) 3.1.2(14) 3.1.2(15) 3.1.2(16) 3.1.2(16) 3.1.2(17) 3.1.2(18) 3.1.2(18) 3.1.2(18) 3.1.2(19) 3.1.2(20) 3.1.2(20) 3.1.2(20) 3.1.2(20) 3.1.2(20) 3.1.2(20) 3.1.2(21) 3.1.2(21) 3.1.2(21) 3.1.2(22) 3.1.2(22) 3.1.2(22) 3.1.2(22) 3.1.2(22) 3.1.2(23) 3.1.2(23) 3.1.2(23)
Loss of CCW to RHR Heat Exchanger Loss of CCW to the Reactor Coolant Pumps RCP Number 1 Seal Failure (RCP B)
Condensate Pump Trip (Pump A)
Main Feedwater Pump Trip (Pump B)
Motor Driven AuxiliaryFeedwater Pump Trip N/A to HNP. See exceptions to ANSI 3.5 item 3 on page 5 Stuck Rod Uncoupled Control Rod Dropped Rod (One Rod)
Ejected Rod Control Rod Stuck on Trip (NEW)
Logic Cabinet Urgent Failure RCS Fuel Rod Breach Inadvertent Turbine Trip Generator Output Breakers Fail to Trip Automatic Voltage Regulator Failure (High)
LOCA on RHR Pressurizer Relief Valve Failure (444B Without P-11 Interlock)
Letdown Pressure Control Valve Failure (PK-145 Open)
Pressurizer Spray Valve Failure Reactor Trip Breakers Fail (Both Inadvertent Open)
Feedline Break Outside Containment Steam Generator Relief Valve Failure (Open)
Feedline Break Inside Containment Steam Break Outside Containment Steamline Break Inside Containment Main Steam Header Break Power Range Channel Detector Failure (Low)
Source Range Instrument Failure (N31 High)
Intermediate Range Control Power Fuse Blown Annunciator System Failure Turbine Runback Failure (Failure to Runback)
Pressurizer Level Control Band Shift Down Median Select Circuit Failure Inadvertent Containment Isolation Phase A Letdown Isolation Valve Failure (1CS-11)
Safety Injection Failure (Train A, Fail to Initiate)
Turbine Driven AuxiliaryFeedwater Pump Trip MT-22A MT-28 MT-512 MT-710 MT-719 MT-711A MT-44 MT-1211B MT-431 MT'RF-16 MT'T-1211A MT-151 MT-612 MT-68 MT-1214A MT-1132 MT-571 MT-112A MT-1041 MT-724 MT-86 MT-723 MT-82 MT-81 MT-815 MT-97 MT-91 MT-912 MT-MSC3 MT-1072 MT-1110 RCS-6 MS'T-51 MT-1032 MT-712A Page 32 of 33
3.1:2(23) 3.1.2(23) 3.1.2(23) 3.1.2(23) 3.1.2(23) 3.1.2(24) 3..1,2(25)
RHR Sump Valves Fail to Open AuxiliaryFeedwater Flow Control Valve Failure (Open)
Diesel Generator Sequencer Fails to Complete Block 1 RCS Leakage into an Accumulator Charging/Safety Injection Pump Speed Changer Failure (NEW)
Reactor Trip Bnreakers Fail (B fails to open)
N/A to HNP. See Exceptions to ANSI 3.5 item 5 on page 5.
MT-136 MT-76 MT-1015 MT-1212A CVC-30 MT-1042 Abstracts for those new tests identified in the above appendix are attached to this report following this page.
Page 33 of 33
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4 Harris Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-SWSS or:
MT-SWSS Malfunction Number:
CO SWS Z21 Year:
3
Title:
B NSW Pump Fails to Auto Start on a A Pump Faiure ANSUANS-3.5-1985 Section 3.2.3(6)
Available Options:
Tested Options:
Test the proper response of the B NSW Pump to a Malfunction ofthe A NSW Pump simultaneous with a failure of the B NSW Pump Auto Start Relay 2-2181. It is not the intent of the test to address the hydraulic response ofthe NSW system. Either A or B pump malfunctions are available.
This test address the B NSW Pump.
Initial Conditions:
Any power with A NSW Pump in service Test
Description:
This test verifies that with a relay failure inserted on the auto start relay of B NSW Pump and a malfunction occurring on the A NSW Pump via the simulated malfunction SWS1A that the B NSW pump willnot automatically start. The result should prompt the operator to manually start the B NSW Pump and it willstart from a switch input. The test is complete when the B NSW pump has been started and the discharge valve has been opened.
Baseline Data:
Current Status:
(1) Panel ofexperts.
Date Achieved Status:
2/18/97 Year 1:
Results 1:
Year 3:
2/18/97 Results 3: S Year 2:
Year 4:
Results 2:
Results 4:
Open SSRs/DRs:
Comments on Current Status and Training Impact ofAny Deficiencies:
Abslract Page
C 4
A I
1
Harris Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-SWS6 or:
MT-SWS6 Malfunction Number:
SWS5a Year:
3
Title:
SW From CNM Fan Cooler Back Pressure Valve fail open ANSVANS-3.5-1985 Section 3.2.3(6)
Available Options:
This malfunction allows the operator to fail the SWS back pressure control valve for the Containment Fan Coolers to any selected valve. Options are available for the A train as well as the B train valve.
Tested Options:
This test address the A train valve SW-116.
Initial Conditions:
Mode 1, approximately 100% power.
Test
Description:
This test verifies that with the malfunction inserted at 100% that SW 116 does not shut in response to a ESW Booster Pump start when SI is activated. The test willinsert the malfunction, verify that the valve is open, insert a large LOCA to cause SI activation and thc verify the valve position based on MCB indications and SW pressures.
Baseline Data:
Current Status:
(1) Panel ofexperts.
Date Achieved Status:
10/9/97 Year 1:
Results 1:
Year 3:
10/9/97 Results 3:
S Year 2:
Year 4; Results 2:
Results 4:
Open SSRs/DRs:
Comments on Current Status and Training Impact of Any Deficiencies:
hbslraer Page
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Harris Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-CRF16 or:
MT-CRF16 Malfunction Number:
CRF-16
Title:
Control Rod Stuck on Trip Year:
4 ANSUANS-3.5-1985 Section 2.1.2(12)
Available Options:
The malfunction is one of 8 (CRF16a-h) that can select any ofthe 52 control rods.
Tested Options:
CRF16A for Rod G-3 at 24 steps. CRF16C for Rod E-11 at 48 steps. CRF16E for Rod D-4 at 72 steps. CRF16h for Rod F-6 at a variable position above the current value.
InitialConditions:
Mode 1, approximately 50% power.
Test
Description:
Eight identical malfunctions are available that willaffect any ofthe Control or Shutdown Rods. This test uses four of these malfunctions simultaneously to verify the ability to stick multiple rods on a Reactor Trip. In at least one case the selected stuck position is above the current rod position ofthe rod. In that case thc rod should stick at the actual rod height at the time of the trip.
Baseline Data:
Current Status:
(1) Panel ofexperts Date Achicvcd Status:
9/23/98 Year I:
Year 3:
Results 1:
Results 3:
Year 2:
Results 2:.
Year 4:
9/23/98 Results 4:
S Open SSRs/DRs:
Comments on Current Status and Training Impact of Any Deficiencies:
Absrrael Page 261
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Harris Simulator Performance Test Abstracts
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09-Mar-99 Test Number:
MT-CVC29 or:
MT-CVC29 Malfunction Number:
CVC-29
Title:
Charging Pump Shaft Shear Year:
4 ANSVANS-3.5-1985 Section 3.1.2 (18)
Available Options:
Three malfunctions are available as CVC29A-C. CVC29C willimpact the "C" pump reguardless ofwhich power bus it is attached Tested Options:
Charging Pump A shaft shear.
Initial Conditions:
Any at power condition with CSIP "A"in service Test
Description:
Baseline Data:
The shaft on the "A"CSIP shears between the speed changer and the pump. The charging header willdepressurize to less than RCS pressure causing a loss ofcharging flowand seal injection. Ifcharging flow is not restored letdown system willheat up and activate alarms accordingly. The CSIP amps should decrease to a value representing the motor with only the speed changer load.
(1) Panel ofexperts.
(2) OP-107, Chemical and Volume Control System. (3) Hot functional test CVCS Current Status:
Date Achieved Status:
10/19/98 Year 1:
Year 3:
Results 1:
Results 3:
Year 2:
Results 2:
Year 4:
10/19/98 Results 4: S Open SSRs/DRs:
Comments on Current Status and Training Impact ofAny Deficiencies:
Abstract Page 262
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Harris Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-CVC30 or:
MT-CVC30 Malfunction Number:
CVC-30
Title:
Charging/Safety Injection Pump Speed Changer Failure ANSVANS-3.5-1985 Section 3.1.2 (18)
Year:
4 Available Options:
Three malfunctions are available as CVC29A-C. CVC29C willimpact the "C" pump reguardless of which power bus it is attached Tested Options:
Charging Pump B speed changer failure Initial Conditions:
Any at power condition with CSIP "B" in service Test
Description:
Baseline Data:
Current Status:
The shaft on the "B" CSIP shears in the speed changer causing a reduction ofabout 10% in the speed of the pump. The charging header willdepressurize to less than RCS pressure causing a loss ofcharging flowand seal injection. Ifcharging flow is not restored the letdown system willheat up and activate alarms accordingly. The CSIP amps should increase about 15 amps greater than thc previous value representing the motor plus a binding in the speed changer. This malfunction is based on event that occurred at the HNP.
Plant data from ACR 93-0111 Date Achieved Status:
10/19/98 Year 1:
Year 3:
Results I:
Results 3:
Year 2:
Results 2:
Year 4:
10/19/98 Results 4:
S Open SSRs/DRs:
Comments on Current Status and Training Impact ofAny Dcficicncics:
Abstract Page 263
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Harris Simulator Performance Test Abstracts a
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/0-Mar-99 Test Number:
MT-DSG5 or:
MT-DSG5 Malfunction Number:
DSG-5 Year:
2
Title:
Diesel Generator Emergency Trip ANSVANS-3.5-1985 Section 3.1.2(23)
Available Options:
The malfunction is available for either Emergency Diesel Generator.
Tested Options:
Diesel Generator A Emergency Failure Initial Conditions:
Mode l, approximately 100% power Test
Description:
Baseline Data:
This test inserts a diesel generator emergency trip following a normal start. After the trip is verified, an emergency start signal is initiated. Indications and alarms consistent with starting air being depleted are verified as well as the removal ofthe start signal by the low air pressure condition. To complete the test the malfunction is cleared, the trip is reset, and diesel generator is started.
(I) Panel ofExperts Current Status:
S Date Achieved Status:
9/25/96 Year 1:
Year 3:
Results 1:
Results 3:
Year 4:
Results 4:
Year 2:
9/25/96, Results 2: S Open SSRs/DRs:
Comments on Current Status and Training Impact ofAny Deficiencies:
Abslract Page 26'4
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Harris Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-SWS4 or:
MT-SWS4
Title:
Service Water Discharge Valve Fails to Open ANSVANS-3.5-1985 Section 3.2.3(6)
Malfunction Number:
SWS-4 Year:
2 Available Options:
This procedure tests the proper response to a failure of a Normal Service Water (NSW) pump discharge valve to stroke to 10% within 10 seconds. This results in a trip ofthe NSW pump. The malfunction is available for either A or B pump.
Tested Options:
This test addresses the A NSW pump.
Initial Conditions:
Mode 3 with A NSW Pump in service.
Test
Description:
This test verifies a failure of the NSW pump discharge valve to stroke to 10% in less than 10 seconds. This results in a trip of the NSW pump.
Basclinc Data:
Current Status:
(I) Panel ofexperts.
Date Achieved Status:
10/17/96 Year 1:
Year 3:
Results 1:
Results 3:
Year 4:
Results 4:
Year 2:
10/17/96 Results 2: S Open SSRs/DRs:
Comments on Current Status and Training Impact of Any Deficiencies:
Abslract Page 268
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HyrrIIs Simulator Performance Test Abstracts 09-Mar-99 Test Number:
MT-SWS7 or:
MT-SWS7
Title:
Nortnal Service Water Pump Shaft Shear ANSVANS-3.5-1985 Section Malfunction Number:
SWS-7 Year:
4 Available Options:
This malfunction allows the simulator operator to break the Normal Service Water (NSW)
Pump shaft at the pump to motor coupling. Options are available for either A or B NSW pumps.
Tested Options:
This test address the A NSW pump.
Initial Conditions:
Any power with A NSW pump in service.
Test
Description:
The simulator is initialized to an at power condition with "A"NSW pump in service. The malfunction to cause the shaft shear is inserted and plant response noted. The ESW headers willdepressurize causing an auto start ofboth ESW pumps and alignment ofthe ESW to return water to the Aux Reservoir willoccur. The NSW system willalso depressurize since an auto start ofthe "B"NSW pump willnot occur.
Baseline Data:
Current Status:
(l) Panel of experts.
Date Achieved Status:
8/20/98 Year 1:
Year 3:
Results 1:
Results 3:
Year 2:
Results 2:
Year 4:
8/20/98 Results 4: S Open SSRs/DRs:
Comments on Current Status and Training Impact ofAny Deficicncics:
Abslract Page 269
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