ML18011A904
| ML18011A904 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/03/1995 |
| From: | Robinson W CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18011A903 | List: |
| References | |
| NUDOCS 9504240316 | |
| Download: ML18011A904 (97) | |
Text
SHEARON HARRIS NUCLEAR POWER PLANT OPERATOR TRAININGSIMIULATOR SIMULATORCERTIFICATION QUADRENNIALREPORT MARCH 1995 CAROLINAPOWER R LIGHT COMPANY NEW HILL,NORTH CAROLINA 950424031b 950315 PDR ADOCK.05000400 PDR
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SHNPP CERTIFICATIONREPORT PACKAGE TABLEOF CONTENTS FORM 474 INTRODUCTION General Information Simulator Configuration Control Exceptions to ANSI/ANS-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 certifyzpt
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:
SCHEDULE OF ANNUALOPERABILITYTESTS APPENDIX B:
SCHEDULE OF MALFUNCTIONTESTS APPENDIX C:
SUMMARY
OF CERTIFICATION DEFICIENCIES APPENDIX D:
SIMULATORCERTIFICATIONTEST ABSTRACTS cenify.re Msscb 3, l99$
NRC FORM 474 (1042j U.S. NUCLEAR REGUIATORYCO)V!M)SS)ON APPRI7VED BYOIEL HCL8150018$
EXFfKR 10/St/IS5 SIMULATIONFACILITYCERTIFICATION ESR MATED BURDEN PER
RESPONSE
TD COMPLY WIIH THIS UIFORMATION COLIECRON REQUEST:
120 HOURS.
FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AHD RECORDS MANAGEMENT BRANCH (MMBB7714),
US.
NUCLEAR REGULATORY COMMISSON, WASHINGTON, DC 205550&I ~ AMDTO THE PAPERWORK REDUCTIONPROJECT (81 500188)> OFRCE OF MANAGEMENT AMDBUDGET, WASHINGTON,DC 206CL UISTTUJCRDN>k Ihb form b to be Sled for In5slce~
recerUUca5on (5 required), and tor any change to a stmutsUon facUtrypedormsnce testing pbn made atter inlUal submtttal of such a pbn. Provkfe the folkwdng Informatkm and check the appoprbte box to In>y>ceto reason for submIUaL FAQLIIY Shearon Harris Nuclear Power Plant Unit l Carolina Power and Light Company DOCKET NUMBER 50-4 pp 3/l5/95 lhb b to certifythat
- 1. The above named facUUyUcensee b using a slmutaUon factly consbUng solely ot a plant>>oferenced slmu!ator Uud meob the requlremerrts of 10 CFR 55A5.
2.
Documenu>Uon b~ for NRC review In accordance wkh 10 CFR 55A5(b).
- 0. Thb stmvfaUon fadUty meets the guktance contlned In AHSt/ANSSE, 1085, as endoaed by NRC Regubtay Gukb 1.1CS.
5 there are any EXCEPIIONS to the cerUUcathn ot ihb Item, Cf%CKl%fK IX] and describe Mlyon addMonal pages as necesauy.
NAME(or oN>>r fdonrrffcadcn) AND LOCATIONOF SIMULAllONFACILIIY.
Harris Simulator Harris Energy 6 Environmental Center 3932 New Hill Holleman Road New Hill, North Carolina 27562-0327 SIMULATIOHFACIL(IYPERFORMANCE TEST AIISTIIACTSATTACHED. (Forp<<fcmrsnce tests cnductod ln N>> period ending wfNthe date ofNb c<<rfdcorfas)
DESCRIPRON OF PERFOIUONCE TESTUIG COMPLETED. (Attach adrgrfcnaf pages as necessary and fdonbfyN>> item desafprfcn heing ccnSued)
Abstracts for tests added since Initial Certification are attached.
. See Section 4.0, "Simulator Tests,"
and Appendix D, "Simulator Certification Test Abstracts."
stMULATICNFActtflYpERFoRMANGE TEsRHG scHEDULE ATTAGHED. (For N>> conduct ofapprordmstoly25'f p<<fonnsnce tests per year fbrN>> fcunyesr period convnoncrng wrrh X
the dere ofttA c<<rrrfcadcn)
DESCBPIION OF PERFORMANCE TESTING TO BE CONDUCTED. (Alrach eddrrfcnsr psgar as necess<<y and IdonrfryNe from descrrptrar heing ccnrrnuert)
See Section 4.0, "Simulator Tests;" Appendix A, "Schedule of Annual Operability Tests;"
and Appendix B, "Schedule of Malfunction Tests."
PERFoRMANGE TEBRNG PLAN cHANGE. (For any mod>Ucsrfrn to a perfcmrence terrfng plan t>mmlrrod on a prevfcus o<<tfifcstfas)
DESCIUPTON OF PERFORMAHCE TESllHG PLAN CHANGE (Attach <<fdldcnsl pages ar necesr<<y and IdentifyN>> rrem dosarprfcn heing ccnrrnuert)
A complete, revised test plan is attached.
See Section 4.0, "Simulator Tests;" Appendix,A, "Schedule of Annual Operability Tests;"
and Appendix B, "Schedule of Malfunction Tests."
REOEfmptcATIDN (Dos<<It>> caroctfre ectrons tafr<<L srrsch results ofcompleted perfonnance rosdng ht ace<<donee wrN 10 cFR M 45(b)(5)(v).
(Alrsch addrrrcnar pages as necessary and fdenrrfyN>> from dosafptfon heing conrfnuert)
Anytahe statement or ombslon In Utb document, Indudlng attachments, may be subject to clvg and almtnal sancUons.
I corufy under penalty of perjury that the Informaucn In thb document and attachments b true and correcL SIG RE A
O SENTAllVE Vice President Harris Nuclear Plant In accordance with 10 CFR 655, DommunbaUons, thb tonn shall be submitted to the NRC as foUovn:
BY MAILADDRESSED Ttk LLS. %ICLEARfKGULAlTIRYCOMMISSION WASt%IGTDM,DC~1 NRC FORM 474 (IM2)
BY DEUVERYUt PERSON ONE WNIEANTHORIH TD THE NRC OFFICE ATl 11555 fKCKVULEPULE
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.
The subject simulation facilityconsists solely of a plant reference full-scope simulator, which is the primary vehicle for providing positive, practical license training and examination.
A major simulator upgrade was carried out and training resumed on the upgraded simulator approximately two months prior to submittal of this report.
The documentation contained herein is intended to constitute sufficient basis for retention of the certification of the Harris Simulator.
Simulator Configuration Control A Simulator Review Group (SRG) is tasked with the responsibility of reviewing changes, potential enhancements, identified discrepancies, and proposed upgrades for implementation or resolution on the Harris Simulator.
The SRG provides recommendations to training management.
The SRG is comprised of the Managers of Operations, License Operator Requalification (LOR)
Training, Operator Initial Training (OIT), and Simulator (functioning as Chairman) or their designees.
The simulator operations specialist (functioning as the facilitator) and other training and plant operations personnel also participate in SRG meetings, which are scheduled each month.
There must be at least one representative each from plant operations, license training (LOR or OIT), and simulator support to conduct a meeting.
The SRG includes at least one Senior Reactor Operator (SRO) licensed or certified individual and one degreed engineer.
The SRG reviews the impact of plant changes on the simulator physical or functional design, other proposed simulator design changes, and simulator discrepancies.
The SRG conducts a
training value assessment of proposed changes to the simulation facility.
The SRG reviews ensure 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.
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Exceptions to ANSI/ANS-3.5-1985 Standard Allexceptions listed herein, 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.
At that time, 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 SRG again reviewed these exceptions within 30 days prior to submittal of this report.
The exceptions identified in this section are listed by ANSI-3.5 reference and subject.
The justification for 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 Technical Support, 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, protective relay, and bistable failures.
This enhancement provides more credible failures for the student to diagnose.
The instructor has more explicit control over these devices than had been available through the deleted malfunctions.
A total of 23 of this type malfunction was replaced by 562 transmitter overrides, 257 relay overrides, and 877 bistable overrides.
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.
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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 (Communications Systems)
Sound-powered phone circuits have been installed for one of the two locations in the control room since simulator certification.
This system is operative for communications with the instructor station.
The installation in the seismic panel was not performed due to this panel representing a non-simulated system.
The radio beeper system was evaluated and deemed unnecessary for training; therefore, a facade only was provided on the radio console.
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 was modified since simulator certification to provide failure capability and emergency lighting to simulate electrical bus failures; however, the lighting configuration was altered to provide light intensity level which approximates lighting levels in the plant control room.
(Noise Levels)
Background noise levels in the simulator room are higher than those found in the plant control room for certain frequencies.
The cause of the high noise level is the HVACunits supporting the simulation facility. These units are projected to be replaced this year (1995) to correct a system reliability concern and reduce system noise.
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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'riteria 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 non-critical 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
¹1, Westinghouse 3-Loop PWR Westinghouse Electric Corporation with major upgrades by S3 Technologies Initial December 20, 1985 Upgrade December 27, 1994 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 deemed acceptable by the Simulator Review Group.
1.2.2 Panels and Equipment All 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 is only used by the reactor engineers at the time of refueling, has also been omitted. These panels and equipment were omitted based on training value assessment.
Classroom and on-the-job training are the means to cevil.re Page 5 of 40 bsarch 3, l99$
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.
Alldialed 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 Alloperative 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, lighting, and sound environment 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 simulator.
This results in a difference between the simulator and main control room ceiling and lighting.
None of the differences have severe training impact.
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 certify,qtt Page 60f 40 h$rrch 3, l99$
camera audio/video system is provided in the simulator facility to allow better analysis of operator activity.
The audio/video system has been reviewed by the SRG and deemed acceptable as a no-training impact difference.
The major simulator upgrade, which reached "Ready-for-Training" status in December
- 1994, included the replacement of the instructor system.
The instructor console was slightly modified to accept installation of the new instructor system hardware.
The upgrade added the capability for the instructor to interact with the simulation through a variety of graphical displays (P8cIDs, panel mimics, and so forth). The instructor station has been modified to accommodate observers without interfering with instructor activities.
The instructor has the capability of operating the simulator from the instructor's booth or from the simulator control room using either of two types of hand held remote operating devices.
1.3.2 Initial Conditions The simulator upgrade project increased the number of available Initial Conditions (ICs) from 50 to 200.
The first 30 ICs axe stabilized and resnapped after each major simulator modification/upgrade period but prior to training restart.
These first 30 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 Procedure 706 provides the method of controlling maintenance and update of simulator initial conditions.
1.3.3 Malfunction Selection The simulation contains capability to insert any number of the over 150 discrete malfunctions individually or in combination.
The selection of malfunctions may still be accomplished through command line entry or through a menu of available malfunctions.
The improvement through the simulator upgrade willalso add the capability for the instructor to select many of the malfunctions through the more than 250 system diagrams ccrrify.rpt Page 7 of 40 March 3. 199$
displayed on the instructor system monitor.
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 device on the simulated control room panels.
For example, a meter may be driven to any value, a light may be turned offor 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.
1.3.4.2 Transmitter Overrides Transmitters and controllers may be overridden or failed to any value in it's range so that corresponding bistable trips and automatic actions will occur.
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 Reactor Protection System 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, two of which had been certified as a part of the original submittal.
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1.3.4.4 Selection of Overrides The selection of overrides may still be accomplished through command line entry or through a menu of available overrides.
The improvement through the upgrade will also add the capability for the instructor to select overrides through the more than 250 system diagrams or through a complete control room set of panel mimics available on the instructor system monitor..
1.3.5 Local Operator Actions (LOAs)
The simulator upgrade project retained all LOAs originally available to the instructor plus expanded capabilities further.
The instructor may still select LOAs from a menu or by direct command line input.
The improvement through the upgrade will also add the capability for the instructor to select LOAs through the more than 250 system diagrams.
1.3.6 Parameter and Equipment Monitoring Allparameter and equipment monitoring capabilities originally available are still retained after the upgrade.
The graphical capabilities of the new instructor system facilitate additional visual monitoring through PAID and panel mimic displays.
Trending capabilities have also been enhanced.
Plot capabilities for up to 400 parameters simultaneously is now available through the instructor system.
The parameter versus time and X-Y plots are still available.
Added is the capability to trend against previously recorded trends, such as necessary to compare a previous test of simulator performance against the current simulator performance.
1.3.7 Simulator Special Features Previously available capabilities are retained in the areas of switch check status/override, run, freeze, backtrack, replay, snapshot, fast time for certain parameters, slow time, computer aided (automatic) exercises, and simulation limitexceeded warnings.
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Backtrack capabilities have been improved by increasing number of backtrack ICs from 10 (at 5 minute intervals) to 120 at an instructor selectable rate (defaults to 2 minute intervals). 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.
The original instructor system only allowed a single computer aided exercise to be active at a time.
In compliance with ANSI/ANS-3.5 section 4.3, the simulator operating limits exceeded warning to the instructor still exist with the upgraded system and still include the following:
- Containment 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 Previously, the loss of the instructor system meant loss of the simulator.
The upgraded instructor system is in many respects independent of the simulation in a manner which allows the simulation to continue with loss of the instructor system.
A separate system is available to interface with, monitor, and operate the simulator temporarily until the instructor system can be returned to service.
1.4 Operating Procedures for the Reference Plant The Simulator 'Control Room continues to utilize a selected set of controlled procedures identical to those used in the Harris Plant control room.
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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.
All such changes have been implemented successfully on the simulator.
The following is a list of the more significant changes:
- Reconfiguration of 4 shutdown bank rods into control bank rods
- Reactor core fuel cycles 5 and 6
- T-hot reduction
- RTD-Bypass removal
- DEH operator panel replacement
- Turbine supervisory recorder replacement
- Digital NI meters
- ERFIS computer system upgrade
- ERFIS display system replacement 1.5.2 Simulator Upgrades Two major simulator upgrades and several limited upgrades have occurred since the last submittal.
The changes made are discussed below.
The latest upgrade was implemented and declared "Ready-for-Training" on December 27, 1994.
1.5.2.1 1991 Upgrade Pxoject As committed in the initial submittal, the decision on the ceiling and lighting modifications was made in June 1991.
The ceiling height was deemed as not practical to modify. A lighting modification was, however, carried out to provide electrical bus failure capability to facilitate effective training in that arena.
This project was completed in December 1991.
1.5.2.2 1991/1992 Major Upgrade Projects These upgrades focused on correcting real time performance deficiencies and RCS modeling deficiencies.
One project carried out Page 11 of 40 March 3. l99$
by GP International interfaced UNIX-based SGI 4D340S computers to the existing simulator computers.
This involved moving the primary NSSS modeling (Reactor Coolant System, Reactor Core, Steam Generator, and Pressurizer Relief Tank models) to the new UNIX-based platform and creating executive software to communicate and interface with the old Encore computers.
Post-upgrade testing verified that Encore-SGI combination performance was well within real time criteria.
This project was concluded in August 1992.
In a parallel project, Westinghouse replaced existing RCS modeling with the latest compatible models.
This upgrade was to correct LOCA response problems in the RCS. Full RCS-related malfunction testing was carried out to validate the results of the upgrade.
The upgrade had successfully corrected the identified deficiencies.
This project was concluded in December 1992.
1.5.2.3 1993 - 1995 Major Upgrade Project The purpose of this project was to correct inconsistent and repeatability problems, to ensure continued real time performance, correct specific problems identified with simulation modeling (RCS, Reactor Core, Containment, and Main Steam models), enhance user friendliness of the instructor
- system, and reduce long-term maintenance costs.
Anticipated side-benefits of such an upgrade were the capability to perform nearly all
- testing, including certification testing, and the capability to carry out instructional and software development without requiring the use of the simulator.
In September 1993, a contract was awarded to S3 Technologies to upgrade the simulator computer and instructor systems and replace the RCS, Reactor Core, Containment, and Main Steam simulation models.
This upgrade was completed and reached the "Ready-for-Training" status on December 27, 1994.
The upgrade warranty period extends through December 27, 1995. Training resumed using upgraded simulator system on January 30, 1995.
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1.5.2.F 1 Scope of Upgrade The upgrade replaced the Encore 32/8780 - Silicon Graphics (SGI) 4D/340S computer system combination. with the SGI Challenge L computer system.
The new system is a UNIX-based computer platform.
The display system used to support the new instructor system is based on an SGI Indigo2 workstation connected by Ethernet to the SGI Challenge L computers.
The instructor system was replaced by a newly developed system which continued to provide the same capabilities as before plus an advanced graphical interface for controlling the simulator.
This graphical interface provides P&ID displays based on Harris Plant designs for most modeled systems and mimics for all simulated panels.
The mimics provide a
user on the development system the capability to operate the plant control board devices as ifusing the actual panel controls.
The mimics also provide one method for insertion of overrides for any panel instrument or control.
The P&ID displays facilitate activation of malfunctions, overrides, local operator actions (LOAs), and normal control functions for the plant component.
The primary thermohydraulic modeling was replaced by S3's RETACT model a design code based generic modeling package which can support a variety of PWR and BWR designs.
This model includes the RCS, pressurizer,
- core, and steam generator thermohydraulics.
The 26 (1 x 26) node reactor core neutronics model was replaced by S3's 300 (12 x 25) node STK model.
The containment, main steam, and pressurizer relief tank modeling was replaced using S3's TOPMERET model building tool. Allof the replaced models use near engineering grade modeling techniques which obey the physical laws.
The executive software structure of the simulation was replaced by S3'sstandard design which ensures more faithful consistency and repeatability in simulator performance.
Interfaces were made significantly more reliable to stimulated systems such as the Emergency Response Facility Information System (ERFIS),
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Radiation Monitoring System (RMS), and Reactor Vessel Level Indication System (RVLIS).
1.5.2.3.2 Upgrade Testing Conducted Four phases ofupgrade testing were carried out. The first phase ensured new hardware performance in both a
stand-alone environment and connected to the simulation panels through he existing Computer Products Input/Output (I/O) system.
The second phase ensured the new modeling was providing accurate simulation of their respective systems.
The third phase ensured the converted models still performed as they had prior to conversion.
The fourth and final phase of testing, prior to turning over the simulator to training for scenario revalidation, was made up of the fourth year certification test program plus additional testing, continued validation ofthe new modeling, and execution ofthose Harris Plant abnormal and emergency operating procedures routinely utilized as a part of the operator training program.
Future testing efficiency has been enhanced by facilitating the conduct of certification testing utilizing the development computer system in lieu of the simulator control room.
This also makes the simulator more available for other training activities.
The panel mimics included as a part of the upgrade allow control manipulation through the instructor workstation in effectively the same manner as ifusing the control room panels.
Running tests on both the simulator and the development system and comparing the results demonstrated that the outcome of either testing method was the same.
Therefore, it is planned that future certification testing willmaximize use of the development system in lieu of using the simulator control room.
1.5.2.3.3 Training Program Validation At the conclusion of the upgrade testing activities, the simulator training and examination scenarios required revalidation to ensure preparedness for resumption of training. This effort was certify.ritt Page 14 of 40 hferch 3, les
conducted during December 1994 and January 1995.
1.5.2.4 Other Upgrades Two significant enhancements to the simulation in the past four years have allowed more credible failures for the students to diagnose.
These enhancements were the addition of Process Instrument Cabinet simulation and selected component override capability.
Detailed testing of these changes against control wiring diagrams (CWDs) ensured faithful replication of the simulated plant system.
The component overrides include the reactor protection system relays, bistables, and process instrumentation transmitters. It deletes the need for a number of specific malfunctions. It also provides the instructor explicit control over more bistables,
- relays, and transmitters than had been available through the deleted malfunctions.
Inclusion of selected relay failures in the Load Sequencer model expanded instructor options in this arena as well.
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2.0 SIMULATORDESIGN DATABASE The original simulator design data base consists of plant reference drawings (logics, CWDs, P8cIDs, and so forth), 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.
Potential simulator modifications are presented to the Simulator Review Group (SRG) for review and recommendation to management for implementation or rejection on the basis ofa training value assessment, which would include a cost benefit analysis.
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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 will be documented by a simulator Trouble Report (TR).
The TR is one of two categories identified on the Simulator Service Request (SSR) form.
The other category is potential change, which is used to initiate consideration of desired simulator enhancements.
Persons noting a problem in the simulator may submit an 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.
The Simulator Review Group (SRG) reviews SSRs submitted/generated since the last meeting for impact on training.
3.2 Engineering Service Request Implementation AllEngineering Service Requests (ESRs) which are approved for work and which have the potential to impact the simulator, are reviewed by the simulator staff for applicability to the simulator.
ESRs are reviewed by the simulator staff in the same time frame during which they are presented to the Harris Plant staff. ESR's which are applicable to the scope of simulation are used to generate a Simulator Modification Request (SMR).
ESRs were identified in the original certification submittal as Plant Change Requests or PCRs.
ESR's which may have an impact on simulator training but are outside the scope of simulation are reviewed by the SRG to determine training value. Ifthe SRG decides that an ESR should be implemented in the simulator, an SMR is generated to increase the scope of simulation and have the modification included.
SMRs are scheduled to be completed in the simulator within twelve months of their completion 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 to be available for training operators. prior to plant start-up. The SMR package is maintained as part of the simulator Update Design Data Base.
ccrr ify.rpr Page 17 of 40 March 3, l993
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, Simulator Service Requests, and Simulator Modification 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 SMRs or 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.
cert i'.rpr Page 18 of 40 March 3, l993
4.0 SIMULATORTESTS The first four years of simulator certification testing was carried out in accordance with the Initial Simulator Certification submittal test schedule.
The testing was accomplished by SRO licensed or certified individuals using test procedures developed by currently or previously licensed or certified personnel.
The tests were based on Harris Plant or similar plant performance data, best estimate analysis, or a panel of experts.
The results of the testing was reviewed by a qualified engineer and by an SRO licensed or certified individual. The selection of simulator performance test topics was determined based on ANSI/ANS-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 report submittal.
These deficiencies are reviewed for their impact on training by the Simulator Review Group.
4.1 Certification Test Schedule 4.1.1 Annual Operability Tests The annual operability tests include the simulator Real Time Test, the Physical Fidelity Tests, the Steady State Stability and Accuracy Tests, selected Normal Operating Tests, selected Normal Operator Surveillance Tests, and the Transient tests.
These tests are listed in Appendix A. All of the tests included in the list are performed on an annual basis except those tests for which a refueling cycle schedule is more appropriate.
The refueling cycle
- tests, including all Normal Operations and Operator Surveillance Tests, are performed after each reactor core fuel cycle update in conjunction with establishing a new set of initial conditions.
4.1.2 Malfunction Tests Malfunctions available on the simulator and incorporated in the operator training program are included in the certification test program.
These tests willbe scheduled for continuing testing such that approximately 25 percent are tested each year and all are tested during the four year period.
The number of certified tests will be 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 willbe noted in subsequent quadrennial reports.
However, the test program will be maintained in compliance with ANSI/ANS-3.5-1985. Appendix B lists certify.rpt Page19of 40 March 3, 1993
the malfunctions which are currently certified and the schedule for testing them over the next four years.
The malfunction tests are divided in such a manner that, to the extent possible, each plant system is tested each year.
4.2 Simulator Test Procedure to ANSI/ANS-3.5-1985 Cross-Reference The initial certification submittal contained a cross-reference of the test program to ANSI/ANS-3.5-1985 to demonstrate satisfaction of regulatory requirements.
contained therein.
Also, the test abstracts contain a reference to the ANSI standard section being satisfied by the test.
Therefore, the cross-reference is not being included with this report.
The current schedule of certification testing was compared against the previously submitted cross-reference to ensure the program still satisfies the ANSI standard.
The conversion of malfunctions to component overrides, such as relay overrides and instrument failures, still allows the instructor to evaluate student performance in response to functionally the same set of equipment failures.
4.3 Summary of Certification Deficiencies Certification deficiencies listed are in Appendix C to this report. Alldeficiencies listed were identified during or following upgrade acceptance testing.
To be listed in this appendix, test results must be identified as either "Satisfactory with Deficiencies" or "Unsatisfactory".
Deficiencies against "Satisfactory" tests will be resolved based on training impact in accordance with the four-level scheduling system outlined in Training Administrative Procedure 702 but not later than March 29, 1996.
There are no tests identified as "Unsatisfactory" in this report.
4.4 Certification Test Abstracts Abstracts of all certification tests were included in the original certification submittal.
Any new certification test willhave an abstract included in Appendix D to this report.
Any deleted certification test willbe annotated in Appendix D as to the justification for the deletion.
ccrri(y.rirr Page 20 of 40 March 3, 1993
APPENDIX A SCHEDULE OF ANNUALOPERABILITYTESTS r
All of the following tests are performed on an annual basis except those tests for which a refueling cycle schedule, normally eighteen to twenty-four months, is more appropriate.
Those refueling cycle tests include all *Normal 0 erations and 0 erator Surveillance Tests.
The Harris Plant refueling outages scheduled for the next four years are currently planned for the Fall of 1995, the Spring of 1997, and the Fall of 1998.
Real Time Test RTT-001 Computer Real Time Test Simulator Ph sical Fidelit Test FT-001 Simulator Physical Fidelity Test FT-002 Simulator Model Limits Exceeded Test (NEW)
Stead State Tests SST-001 100 Percent Power Accuracy Test SST-002 75 Percent Power Accuracy Test SST-003 30 Percent Power Accuracy Test Normal 0 NOT-001 NOT-002 NOT-003 NOT-004 NOT-005 NOT-006 NOT-007 NOT-008 NOT-009 erations Tests
("See Note Above)
GP-006, Plant Shutdown GP-007, Plant Cooldown GP-002, Plant Heatup GP-004, Reactor Startup GP-005, Plant Startup Recovery to Rated Power Following Reactor Trip GP-008, Plant Drain to Mid-Loop (NEW)
GP-001, Plant Fill and Vent (NEW)
GP-009, Refueling with Cavity Fill and Drain (NEW)
Normal 0 NOST-001 NOST-002 NOST-003 NOST-004 NOST-005 NOST-006 erator Surveillance Tests
(*See Note Above)
Power Range Heat Balance CVCS/SI System Operability RHR Pump Operability Containment Spray Operability 1A-SA Emergency Diesel Generator Operability Turbine Valve Test (NEW) certify.qx Page 21 of 40 March 3, l99$
NOST-007 NOST-008 NOST-009 NOST-010 NOST-011 NOST-012 NOST-013 NOST-014 NOST-015 NOST-016 NOST-017 NOST-018 NOST-019 NOST-020 NOST-021 NOST-022 NOST-023 NOST-024 NOST-025 Main Steam Isolation Valve and Main Feedwater Isolation Valve Operability Test Daily Surveillance Requirements Modes 1 and 2 Daily Surveillance Requirements Modes 3 and 4 Reactor Coolant System Leakage Evaluation Shutdown Margin Calculation Calculation of Quadrant Power Tilt Ratio Main Steam Isolation Valve Operability Test Permissives P-6 and P-10 Verification 1B-SB Emergency Diesel Generator Operability Turbine Mechanical Overspeed Trip Test AFW Pump 1B-SB Operability Test - Quarterly RHR Pump 1B-SB Operability Reactor Coolant Pump Seals Controlled Leakage Evaluation AFW Pump 1A-SA Operability Test - Quarterly CCW System Operability - Quarterly AFW Pump 1X-SAB Operability (NEW)
Control Rod and Rod Position Indication Exercise Motor Driven AFW Pump Flow Test (NEW)
Turbine Driven AFW Pump Full Flow Test (NEW)
Transient Tests TT-001 TT-002 TT-003 TT-004 TT-005 TT-006 TT-007 TT-008 TT-009 TT-010 Manual Reactor Trip Simultaneous Trip of all Feedwater Pumps Simultaneous Closure of AllMain Steam Isolation Valves Simultaneous Trip of AllReactor 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 certify.re Page22of 40 Much 3, l99$
APPENDIX B SCHEDULE OF MALFUNCTIONTESTS MALFUNCTIONTESTS FIRST YEAR MT-10162 MT-10165 MT-1031 MT-1042 MT-111A MT-113 MT-1131 MT-12 MT-1210 MT-1221 MT-1222 MT-1231 MT-126 MT-135 MT-136 MT-152 MT-22 MT-25 MT-42 MT-431 MT-44 MT-51 MT-5182 MT-513 MT-523 MT-526 MT-53 MT-61 MT-6102 MT-612 MT-623 MT-632 Failure of Rod Blocks to Block (C-2, C-3, C-4)
Failure of Rod Block to Block (C-5)
Safety Injection Failure (Train B, Inadvertent)
Reactor Trip Breakers Fail (B fails to open)
Pressurizer Steam Space Leak Loss of Instrument Airto the Containment Building (AIR-1,1)
Pressurizer Relief Valve Failure (445A With P-11 Interlock)
NSW Pump Trip'and Loss of NSW RCP A, B, C High Vibration Steam Generator Tube Leak (S/G B)
Steam Generator Tube Rupture (S/G A)
RCP Trip From 100 Percent Power (RCP-C)
RCS Boron Dilution RHR Bypass Line Leak (Train A)
RHR Sump Valves Fail to Open Turbine Protection Trip.Failure Loss of CCW to RHR Heat Exchanger Letdown Heat Exchanger Tube Leak Logic Cabinet Urgent Failure Dropped Rod (One Rod)
Stuck Rod Letdown Isolation Valve Failure (1CS-11)
Seal Injection Flow Control Valve Failure (HC-186 Closed)
RCP Number 2 Seal Failure (RCP A)
High Temperature Divert Valve (TCV-143) Failure Boric Acid Pump Trip Letdown Line Leak Inside Containment Station Blackout Loss of Unit Aux Transformer 1B (NEW)
Generator Output Breakers Fail to Trip Loss of 120-VAC Uninterruptible Power (Power Supply SIII)
Loss of 125-VDC Emergency Bus (DP 1B-SB) aaify.qa Page23 of 40 hitch 3. l99$
MT-710 MT-712 MT-724 MT-772 MT-78 MT-814 MT-86 MT-92 MT-93 MT-96 MT-97 MT-MSC3 RCS-18 Condensate Pump Trip (Pump A)
Failure of Excess Condensate Dump Valve (Closed)
Feedline Break Outside Containment Feedwater Bypass Valve Failure (Open)
Turbine Driven Auxiliary Feedwater Flow Control Valve Failure (Closed)
Steam Failure to TDAFW Pump (1MS-72 Closed)
Steam Generator Relief Valve Failure (Open)
Source Range Pulse Height Discriminator Failure Failure of Source Range High Voltage to Disconnect Intermediate Range Channel Gamma Compensation Failure Power Range Channel Detector Failure (Low)
Annunciator System Failure (NEW)
Small Break LOCA (NEW) ccrrify.rpc Page24 of 40 hlrreh 3, 199S
MALFUNCTIONTESTS SECOND YEAR MT-1032 MT-1041 MT-106 MT-112 MT-1132 MT-114 MT-1211 MT-1212 MT-1214 MT-1215 MT-1232 MT-131 MT-1321 MT-151 MT-17 MT-210 MT-24 MT-26 MT-31 MT-35 MT-41 MT-45 MT-461 MT-47 MT-514 MT-5284 MT-52 MT-5285 MT-54 MT-55 MT-56 MT-571 MT-651 MT-662 Safety Injection Failure (Train A, Fail to Initiate)
Reactor Trip Breakers Fail (Both Inadvertent Open)
False Containment Spray Actuation 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 RCP Bearing Oil Reservoir Leak RCS Thermal Barrier Leak into CCW System Reactor Coolant Pump Trips (RCP-C)
RHR Pump Trip (Pump A)
RHR HX Flow Control Valve Failure (FCV-603A Closed)
Inadvertent Turbine Trip Refueling Water Storage Tank Leak Seal Water Heat Exchanger Tube Leak Component Cooling Water Header Supply Valve Failure (Closed)
Loss of CCW to RCP Thermal Barrier Circulating Water Pump Trip Loss of Condenser Vacuum Pump Power Cabinet Urgent Failure Ejected Rod Uncontrolled Automatic Rod Motion Failure of Auto Rod Blocks to Block (C-11)
RCP Number 3 Seal Failure (RCP C)
Charging Line Leak in Containment Before Regen HX VCT Outlet Isolation Valve Failure (LCV-115E Closed)
(NEW)
Charging Line Leak Between Regen HX and 1CS-492 Letdown Line Leak Outside Containment Charging Pump Trip Reactor Makeup Water Pump Trip Letdown Pressure Control Valve Failure (PK-145 Open)
Loss of 6.9-KV Emergency Bus (1A-SA)
Loss of a 125-VDC Nonvital Bus (DP 1A) arrify.rpr Page 25 of 40
MT-67 MT-692 MT-712A MT-719 MT-72 MT-720 MT-723 MT-73 MT-82 MT-94 MT-95 Diesel Generator Failure Loss of Start-up Transformer 1B (NEW)
Turbine Driven Auxiliary Feedwater Pump Trip Main Feedwater Pump Trip (Pump B)
Condensate Booster Pump Trip (Pump B)
Main Feedwater Pump Recirc Valve Failure (Pump 1B)
Feedline Break Inside Containment Turbine Driven Auxiliary Feedwater Pump Speed Control Oscillates Steam Break Outside Containment Source Range Channel High Voltage Failure Intermediate Range Channel Failure ccN fy.rpt Page26of 40
MALFUNCTIONTESTS THIRD YEAR MT-1013 MT-1072 MT-1110 MT-121 MT-1211B MT-1213A MT-124 MT-131A MT-1322 MT-1331 MT-137 MT-138 MT-141 MT-154 MT-271 MT-32 MT-333 MT-34 MT-410 MT-411 MT-412 MT-462 MT-5201 MT-5282 MT-5283 MT-572 MT-58 MT-59 MT-6101 MT-64 MT-661 MT-68 MT-69 MT-711A Inadvertent Feedwater Isolation Turbine Runback Failure (Failure to Runback)
Pressurizer Level Control Band Shift Down Emergency Service Water Pump Trip (NEW)
Uncoupled Control Rod (NEW)
RCS Leak (Large Break)
(NEW)
Reactor Coolant Pump Trip (Locked Rotor)
RHR Pump Trip (NEW)
RHR HX Flow Control Valve Failure (FCV-603B Open)
RHR HX Bypass FCV Failure (FK-605A1 Open)
Containment Spray Pump Failure Containment Spray Pump Discharge Valve Failure Containment Fan Cooler Unit Trip Turbine Vibration Letdown Temperature Controller Failure (TK-144 Low)
Main Condenser Tube Leak Hotwell Level Controller Failure (LC-1901 Low)
Loss of Condenser Vacuum DRPI-Open or Shorted Coil Improper Bank Overlap Control Bank Rod Step Counter Failure Uncontrolled Manual Rod Motion Failure of Charging Flow Control Valve Charging Pump Discharge Line Leak Before FT-122 Charging Line Leak Between FT-122 and 1CS-235 Letdown Pressure Control Valve Failure (PK-145 Closed)
Loss of Normal Letdown VCT Divert Valve Control Failure (HUT)
Loss of Unit Auxiliary Transformer Loss of 6.9 KVAuxiliary Bus (1B)
Loss of a 250-VDC Nonvital Bus (DP-1-250)
(NEW)
Automatic Voltage Regulator Failure (High)
Loss of Start-up Transformer Motor Driven Auxiliary Feedwater Pump Trip ccrrlfy.re Page 27 of 40 hfrrch 3. l99S
MT-722 MT-74 MT-75 MT-76 MT-81 MT-815 MT-87 MT-89 MT-91 MT-98 RCS-6 Feedwater Control Valve Position Failure (LCV-488 Open)
Steam Generator Backleakage (NEW)
Auxiliary Feedwater Line Ruptures (NEW)
Auxiliary Feedwater Flow Control Valve Failure (Open)
Steamline Break Inside Containment Main Steam Header Break MSIV Failure (S/G B Shut)
Atmospheric Steam Dump Valve Failure (PCV-408J Open)
Source Range Instrument Failure (N31 High)
Power Range Channel Failure (Low)
Median Select Circuit Failure (NEW) certify.re Page 28 of 40 hlarch 3, l995
MALFUNCTIONTESTS FOURTH YEAR MT-1014 MT-1015 MT-10161 MT-1017 MT-1071 MT-111 MT-112A MT-118 MT-1211A MT-1212A MT-1213 MT-125 MT-1332 MT-134 MT-155 MT-157 MT-21 MT-22A MT-23 MT-272 MT-28 MT-331 MT-413 MT-48 MT-49 MT-512 MT-516 MT-5181 MT-5202 MT-524 MT-525 MT-527 MT-5281 MT-615 Inadvertent Main Steam Isolation Diesel Generator Sequencer Fails to Complete Block 1 Failure of Rod Blocks to Block (C-1)
Failure of Permissive Interlock P-14 Turbine Runback Failure (Erroneous Runback)
Loss of Instrument Air (Turbine Building)
Pressurizer Spray Valve Failure Pressurizer Backup Heaters Groups A and B Failure RCS Fuel Rod Breach RCS Leakage into an Accumulator RCS Vessel Flange Leak RCP Shaft Break Accident (RCP B)
RHR HX Bypass FCV Failure (FK-605B1 Closed)
RHR to Letdown Valve Failure (HCV-142.1 Open)
Governor Valve Failure (GV-3 Closed)
Turbine DEH Computer Failure Component Cooling Water Pump Trip Loss of CCW to RHR Heat Exchanger (NEW)
CCW Leak into the Service Water System Letdown Temperature Controller Failure (TK-144 High)
Loss of CCW to the Reactor Coolant Pumps Hotwell Level Controller Failure (LC-1900 High)
Rod Speed Deadband Control Failure TREF Failure DRPILoss of Voltage RCP Number 1 Seal Failure (RCP B)
Boric Acid Filter Plugged Seal Injection Flow Control Valve Failure (HC-186 Open)
Failure of Charging Flow Control Valve (Closed)
Charging Pump Suction From RWST Failure (115D Open)
Charging Pump Mini Flow Valve Failure (1CS-182 Closed)
Charging Line Containment Isolation Valve Failure Charging Line Leak on Charging Pump Suction Diesel Generator Governor Failure cenify.qx Page 29 of 40 bsarch 3, l99S
MT-616 MT-714 MT-715 MT-771 MT-810 MT-811 MT-812 MT-88 MT-911 MT-912 MT-913 Diesel Generator Breaker Inadvertent Trip Condensate Storage Tank Leak Heater Drain Pump Trip (Pump B)
Feedwater Bypass Valve Failure (Closed)
Steam Dump Control Failure (Closed)
Mechanically Stuck Condenser Dump Valve (PCV-408 Open)
Steam Dump Permissive (P-12) Failure Steam Generator Safety Valve Failure (Open)
Source Range Instrument Power Fuse Blown Intermediate Range Control Power Fuse Blown Power Range Control Power Fuse Blown ceniky.re Page 30 of 40 March 3, l99$
APPENDIX C
SUMMARY
OF CERTIFICATIONDEFICIENCIES Deficiencies identified during certification testing are listed in this appendix.
They are split between deficiencies against the testing conducted over the last four years and deficiencies against new tests being added to the plan for the next four years.
There are currently 22 deficiencies outstanding against 19 of 230 certification tests.
These deficiencies are made up of either Simulator Service Requests (SSRs) or upgrade warranty Deficiency Reports (DRs).
DRs are identified as W-xxxx. Test results are shown as "Satisfactory with Discrepancies" (SD) or "Unsatisfactory" (U). There are no tests identified as "Unsatisfactory" in this report.
TEST/RESULTS CMS/DR ¹ TITLE/DESCRIPTION The following tests were run as a part of the initial (March 1991 - March 1995) certification testing program.
The resulting deficiencies that remain unresolved at this time are shown below.
Except as noted, these tests willcontinue to be included in the next four year testing cycle.
FT-001/SD SST-001/SD SST-002/SD SST-003/SD NOT-005/SD TT-003/SD TT-004/SD SSR 95-0152 SSR 95-0148 SSR 95-0147 SSR 94-0693 W-0012 W-0173/
SSR 95-0153 SSR 95-0153 Physical Fidelity Test.
11 differences.
4 of these awaiting PCR completion.
7 minor label/engraving differences.
Steady State Test 100%.
Stability portion satisfactory.
Discrepancies were identiTied between actual plant board data and simulator board
- data, potentially caused by differences in equipment line-ups.
Steady State Test 75%.
Stability portion satisfactory.
Same type discrepancies as with SST-001 were identified.
Steady State Test 30%.
Identified 7 specific differences between plant and simulator.
Plant Start up test.
Impulse pressure changes of 20 psi on turbine roll up.
Closure of all MSIVs Transient Test.
AFW flow spiked on pump start.
Cause of the Rx trip is unknown.
Simultaneous Trip of all Reactor Coolant Pumps.
AFW flow spiked on pump start.
ceniyy.qe Page 31 of 40 Marcb 3, l99S
TT-006/SD TT-008/SD TT-009/SD TT-010/SD MT-1031/SD MT-1211A/SD MT-5152/SD SSR 95-0151 W-0168/W-170 W-0171 SSR 95-0149 W-0172 SSR 94-0610 SSR 95-0223 SSR 94-0607 Turbine Trip below P10 transient test.
Feed Reg Bypass valve did not maintain level at 66% during transient.
0.5 degree step change in cold leg temperature being evaluated.
LBLOCATransient Test.
Containment pressure greater than FSAR peak.
Evaluate differences in AFW flow to individual SGs and containment pressure values.
Slow RCS depressurization transient.
Flow spike on AFW and SI systems noted.
Relay problem on inadvertent SI - no red first out annunciator.
Area Monitors did not increase ) 1000x normal Boric acid flow transmitter failure.
Flow spiked to 84gpm.
Max value at plant ( 50gpm.
(This test has been converted to component override.)
The following are new tests being included in the next four year cycle ofcertification testing and having outstanding deficiencies.
These tests are among 22 new tests included as a part of simulator upgrade project acceptance testing and being added to the certification test program.
NOT-007/SD NOT-008/SD MT-1213A/SD MT-692/SD MT-MSC3/SD W-0069/
W-0048 W-0004 SSR 95-0180 SSR 94-0536 SSR 95-0225 Vessel vent is sized too small Audio count rate changes without changes in counts Improper pressure response on RCP start RCDT Level/Mass increases at point of break/SI Differential target relay doesn't drop Chime/Horn 2 improper operation with failed power supply certify.rpt Page 32 of 40 bsarch 3, l99S
APPENDIX D SIMULATORCERTIFICATIONTEST ABSTRACTS There remains a total of 230 tests, after additions and deletions, included in the certification test package for the next four years.
This appendix contains a complete list (index) of test abstracts.
These abstracts were included in the original certification submittal and are not being included in this appendix unless it is one of the 22 NEW tests being added to the certification package.
Deleted tests will be so annotated (in the index of abstracts) with justification for deletion indicated, generally due to conversion from malfunctions to component (relay or instrument) overrides.
INDEX OF ABSTRACTS Simulator Ph sical Fidelit Test (2)
FT-001 FT-002 Simulator Physical Fidelity Test Simulator Model Limits Exceeded Test (NEW)
Malfunction Tests (180)
MT-101 MT-1013 MT-1014 MT-1015 MT-10161 MT-10162 MT-10165 MT-1017 MT-102 MT-1031 MT-1032 MT-1041 MT-1042 MT-106 MT-1071 MT-1072 MT-111 MT-1110 MT-111A Inadvertent Containment Isolation Phase A (Deleted - Relay Override)
Inadvertent Feedwater Isolation Inadvertent Main Steam Isolation Diesel Generator Sequencer Fails to Complete Block 1 Failure of Rod Blocks to Block (C-1)
Failure of Rod Blocks to Block (C-2, C-3, C-4)
Failure of Rod Block to Block (C-5)
Failure of Permissive Interlock P-14 Inadvertent Containment Isolation Phase B (Deleted - Relay Override)
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 certify.re Page 33 of 40 bsarch 5, l99S
MT-112 MT-112A MT-113 MT-1131 MT-1132 MT-114 MT-1151 MT-1152 MT-1161 MT-1162 MT-1171 MT-1172 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-1215 MT-1216 MT-1221 MT-1222 MT-1231 MT-1232 MT-124 MT-125 MT-126 MT-1271 MT-1281 MT-1282 MT-129 MT-13 MT-131 Loss of Instrument Airto the Reactor (Reactor Auxiliary Building)
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-444 High) (Deleted - Inst. Override)
Pressurizer Pressure Channel Failure (PT-445 Low) (Deleted - Inst. Override)
Pressurizer Pressure Channel Failure (PT-456 High) (Deleted - Inst. Override)
Pressurizer Pressure Channel Failure (PT-457 Low) (Deleted - Inst. Override)
Pressurizer Level Channel Failure (LT-459 Low) (Deleted - Inst. Override)
Pressurizer Level Channel Failure (LT-459 High) (Deleted - Inst. Override)
Pressurizer Backup Heaters Groups A and B Failure NSW Pump Trip and Loss of NSW Emergency Service Water Pump Trip (NEW)
RCP A, B, C High Vibration RCS Leak Within Capacity of Charging Pumps RCS Fuel Rod Breach Uncoupled Control Rod (NEW)
LOCA Within Capacity of the SI Pumps RCS Leakage into an Accumulator RCS Vessel Flange Leak RCS Leak (LOCA) (NEW)
RCP Bearing Oil Reservoir Leak RCS Thermal Barrier Leak into CCW System RCS Flow Transmitter Failure (FT-436 w) (Deleted - Inst. Override)
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 Control RTD Failure (TE-411B High) (Deleted - Inst. Override)
RCS Protection RTD Failure (TE-412B Low) (Deleted - Inst. Override)
RCS Protection RTD Failure (TE-422B High) (Deleted - Inst. Override)
RCS WR Pressure Transmitter Failure (PT-403 High) (Deleted - Inst. Override)
Containment Fan Cooler Unit Trip (Deleted - Renamed MT-141)
RHR Pump Trip (Pump A) ccNly.qa Page34of 40 Marcb 3, 199S
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-1519 MT-152 MT-154 MT-155 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 RHR Pump Trip (Pump A) (NEW)
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 (Was MT-13)
Inadvertent Turbine Trip Turbine 1st Stage Press. Xmtr Failure (PT-446 Low) (Deleted - Inst. Override)
Turbine Protection Trip Failure Turbine Vibration Governor Valve Failure (GV-3 Closed)
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 (NEW)
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 DRPIOpen or Shorted Coil Improper Bank Overlap Control Bank Rod Step Counter Failure certify.rpt Page 35 of 40 hiarch 3, l99$
MT-413 MT-42 MT-431 MT-44 MT-45 MT-46 MT-461 MT-462 MT-47 MT-48 MT-49 MT-51 MT-5111 MT-5112 MT-512 MT-513 MT-514 MT-5151 MT-5152 MT-516 MT-5181 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 Rod Speed Deadband Control Failure Logic Cabinet Urgent Failure Dropped Rod (One Rod)
Stuck Rod Ejected Rod Uncontrolled Rod Motion (Deleted - Split into MT-461 & MT-462)
Uncontrolled Automatic Rod Motion (Was part of MT-46)
Uncontrolled Manual Rod Motion (Was part of MT-46)
Failure of Auto Rod Blocks to Block (C-11)
TREF Failure DRPILoss of Voltage Letdown Isolation Valve Failure (1CS-11)
VCT Level Transmitter Failure (LT-112 High) (Deleted - Inst. Override)
VCT Level Transmitter Failure (LT-115 Low) (Deleted - Inst. Override)
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) (Deleted - Inst. Override)
Boric Acid Flow Xmtr. Failure (FT-113 to 0 gpm) (Deleted - Inst. Override)
Boric Acid Filter Plugged 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)
(NEW)
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 Mini Flow 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 Page 36 of 40 March 3, 199S
MT-571 MT-572 MT-58 MT-59 MT-61 MT-610 MT-6101 MT-6102 MT-612 MT-615 MT-616 MT-62 MT-623 MT-63 MT-632 MT-64 MT-65 MT-651 MT-66 MT-661 MT-662 MT-67 MT-68 MT-69 MT-692 MT-710 MT-711A MT-712 MT-712A MT-714 MT-715 MT-719 MT-72 MT-720 MT-721 MT-722 MT-723 MT-724 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 (Deleted - Renamed MT-6101)
Loss of Unit Auxiliary Transformer A phase (Was MT-610)
Loss of Unit AuxiliaryTransformer B phase (NEW)
Generator Output Breakers Fail to Trip Diesel Generator Governor Failure Diesel Generator Breaker Inadvertent Trip Loss of 120-VAC Uninterruptible Power (Power Supply SIII) (Deleted-Renamed MT-623)
Loss of 120-VAC Uninterruptible Power (Power Supply SIII) (Was MT-62)
Loss of 125-VDC Emergency Bus (DP 1B-SB) (Deleted - Renamed MT-632)
Loss of 125-VDC Emergency Bus (DP 1B-SB) (Was MT-63)
Loss of 6.9 KVAuxiliary Bus (1B)
Loss of 6.9-KV Emergency Bus (1A-SA) (Deleted - Renamed MT-651)
Loss of 6.9-KV Emergency Bus (1A-SA) (Was MT-65)
Loss of a 125-VDC Nonvital Bus (DP 1A) (Deleted - Renamed MT-662)
Loss of a 250-VDC Nonvital Bus (DP-1-250)
(NEW)
Loss of a 125-VDC Nonvital Bus (DP 1A) (Was MT-66)
Diesel Generator Failure Automatic Voltage Regulator Failure (High)
Loss of Start-up Transformer 1A Loss of Start-up Transformer 1B (NEW)
Condensate Pump Trip (Pump A)
Motor Driven Auxiliary Feedwater Pump Trip Failure of Excess Condensate Dump Valve (Closed)
Turbine Driven Auxiliary Feedwater 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 (FT-466 Low) (Deleted - Inst. Override)
Feedwater Control Valve Position Failure (LCV-488 Open)
Feedline Break Inside Containment Feedline Break Outside Containment ccrrify.rpr Page 37 of 40 March 3, l993
0
MT-725 MT-73 MT-74 MT-75 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 MT-96 MT-97 MT-98 MT-MSC3 RCS-18 RCS-6 Override) osed)
Override)
Override)
Steam Generator Level Chan. Failure (LT-496 Low) (Deleted - Inst.
Turbine Driven Auxiliary Feedwater Pump Speed Control Oscillates Steam Generator Backleakage (NEW)
Auxiliary Feedwater Line Ruptures (NEW)
Auxiliary Feedwater Flow Control Valve Failure (Open)
Feedwater Bypass Valve Failure (Closed)
Feedwater Bypass Valve Failure (Open)
Turbine Driven Auxiliary Feedwater Flow Control Valve Failure (Cl Steamline Break Inside Containment Steam Dump Control Failure (Closed)
Mechanically Stuck Condenser Dump Valve (PCV-408 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) (Deleted - Inst.
Steam-Line Flow Transmitter FT-494 (Deleted - Inst. Override)
Steam Generator Press. Xmtr. Failure (PT-485 High) (Deleted - Inst.
Steam Generator Relief Valve Failure (Open)
MSIV Failure (S/G B Shut)
Steam Generator Safety Valve Failure (Open)
Atmospheric Steam Dump Valve Failure (PCV-408J 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 Intermediate Range Channel Gamma Compensation Failure Power Range Channel Detector Failure (Low)
Power Range Channel Failure (Low)
Annunciator System Failure (NEW)
Small Break LOCA (NEW)
Median Select Circuit Failure (NEW) ccnify.rior Page38of 40 March 3, 199S
Normal 0 erator Surveillance Tests (25)
NOST-001 NOST-002 NOST-003 NOST-004 NOST-005 NOST-006 NOST-007 NOST-008 NOST-009 NOST-010 NOST-011 NOST-012 NOST-013 NOST-014 NOST-015 NOST-016 NOST-017 NOST-018 NOST-019 NOST-020 NOST-021 NOST-022 NOST-023 NOST-024 NOST-025 OST-1004, OST-1007, OST-1008, OST-1009, OST-1013, OST-1014, OST-1018, Operability OST-1021, OST-1022, OST-1026, OST-1036, OST-1039, OST-1046, OST-1054, OST-1073, OST-1075, OST-1076, OST-1092, OST-1126, OST-1211, OST-1316, OST-1411, OST-1005, OST-1087, OST-1088, Power Range Heat Balance CVCS/SI System Operability RHR Pump Operability Containment Spray Operability 1A-SA Emergency Diesel Generator Operability Turbine Valve Test (NEW)
Main Steam Isolation Valve and Main Feedwater Isolation Valve Test Daily Surveillance Requirements Modes 1 and 2 Daily Surveillance Requirements Modes 3 and 4 Reactor Coolant System Leakage Evaluation Shutdown Margin Calculation Calculation of Quadrant Power TiltRatio Main Steam Isolation Valve Operability Test Permissives P-6 and P-10 Verification 1B-SB Emergency Diesel Generator Operability Turbine Mechanical Overspeed Trip Test AFW Pump 1B-SB Operability Test - Quarterly RHR Pump 1B-SB Operability Reactor Coolant Pump Seals Controlled Leakage Evaluation AFW Pump 1A-SA Operability Test - Quarterly CCW System Operability - Quarterly AFW Pump 1X-SAB Operability (NEW)
Control Rod and Rod Position Exercise Motor Driven AFW Pumps Flow Test (NEW)
Turbine Driven AFW Pump Full Flow Test (NEW)
Normal 0 erations Tests (9)
NOT-001 NOT-002 NOT-003 NOT-004 NOT-005 NOT-006 NOT-007 GP-006, Plant Shutdown GP-007, Plant Cooldown GP-002, Plant Heatup GP-004, Reactor Startup GP-005, Plant Startup Recovery to Rated Power Following Reactor Trip GP-008, Plant Drain to Mid-Loop (NEW) ccrtiCy.rpc Page39of 40 hluch 3, l99$
NOT-008 GP-001, Plant Fill and Vent (NEW)
NOT-009 GP-009, Refueling with Cavity Fill and Drain (NEW)
Real Time Test (1)
RTT-001 Computer Real Time Test Stead State Tests (3)
SST-001 SST-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 TT-005 TT-006 TT-007 TT-008 TT-009 TT-010 Manual Reactor Trip Simultaneous Trip of all Feedwater Pumps Simultaneous Closure of AllMain Steam Isolation Valves Simultaneous Trip of AllReactor 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 Abstracts for those new tests identified in the above index are attached to this report following this page.
ccrri(y.rpr Page 40 of 40 h(rrch 3, l99S
PERFORMANCE TEST ABSTRACT
'T-002 1.0 PROCEDURE TITLE/ANSI 3.5 REFERENCE 1.1 Simulator Model Limits Exceeded Test 1.2 ANSUANS 3.5, 1985, 4.3 1.3 Regulatory Guide 1.149, Rev 1.
2.0 AVAILABLEOPTIONS Not applicable 3.0 TESTED OPTIONS Not applicable 4.0 INITIALCONDITIONS Normal operating temperature and pressure.
5.0 TEST DESCRIPTIONS Nine variables are used to alert the instructor in event they exceed a preset value. To test this, the simulator models are &ozen and the variable is set above the assigned alert value. The models are lrozen because the variables are calculated and willnot stay at the value they are set to. The simulator message alerting the instructor of the limit exceeded is displayed. This is repeated for each variable.
6.0 BASELINE DATA/REFERENCES 6.1 Plant design criteria 6.2 Model limits as defined by the software group FT-002ABS Page.,l of 2
7 0 DATE PERFORMED/TEST RESULTS c
03-01-95/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAINING IMPACT None c
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PERFORMANCE TEST ABSTRACT MT-121 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 SWS3, Emergency Service Water Pump Trip 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (6) 2.0 AVAILABLEOPTIONS ESW Pump A trip or ESW Pump B trip on overcurrent 3.0 TESTED OPTIONS Both ESW pumps are tripped.
One at startup and the other while it is running.
4.0 INITIALCONDITIONS Mode 1, with ESW Pumps secured.
5.0 TEST DESCRIPTION The test is started with a Safety Injection Actuation signal to start the Emergency Safeguards Sequencer.
ESW Pump A willstart and trip. After the indications of ESW Pump A trip are verified, a trip is entered on ESW Pump B that is running.
The indications of a trip of a running ESW Pump are verified to be correct and the test is complete.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 AOP-022, Loss of Service Water hfI12IABS
7.0 DATE PERFORMED/TEST RESULTS 11-08-93/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT MT-1211B 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 RTC-2, Uncoupled Control Rod 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (12) 2.0 AVAILABLEOPTIONS Any one of 52 control rods can be selected to appear uncoupled.
In addition any percentage of the selected rod can be chosen which allows for a lesser reactivity effect ifless than 100% of the rod is chosen as uncoupled.
3.0 TESTED OPTIONS Any rod is selected at 100% failure and a reactor startup is performed.
4.0 INITIALCONDITIONS Mode 3, xeady for a Reactor Startup.
5.0 TEST DESCRIPTION This malfunction can fail any one of 52 control xods to appear uncoupled.
After selecting the rod, up to 100% of the rod can be selected to give the appearance of individual xodlets broken off of the control rod.
This test will perform two Rx startups to 10'mps.
The first with an uncoupled rod malfunction active, the second with no malfunction to show the difference in control 'bank rod heights needed to achieve criticality. The affected rod is chosen by the person running the test since the results will be the same for any rod.
Aftex the first startup, the uncoupled rod malfunction willbe inserted on a control rod to allow it to drop into the coxe and the reactivity effect willbe observed and verified.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts CZWl721IRABS
7.0 DATE PERFORMED/TEST RESULTS 10-19-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING None C791TI21IBA8$
PERFORMANCE TEST ABSTRACT MT-1213A 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 RCS-1, RCS Leak (LOCA) 1.2 ANSI/ANS 3.5, 1985, 3.1.2(l) 2.0 AVAILABLEOPTIONS This malfunction tests the proper response to a large break loss of coolant accident.
More than one LOCA may be active from any of the loop's hot or cold legs.
The leak rate is based on normal operating pressure and willvary as pressure varies, and the break size is selectable in a range of 0 to 100% which is a break size from 0 to 4.3 sq. ft.
3.0 TESTED OPTIONS RCS loop 3 cold leg at 100%
4.0 INITIALCONDITIONS Mode 1, at approximately 100% power.
5.0 TEST DESCRIPTION This malfunction causes rapid and complete depressurization of the RCS with a reactor trip and safety injection.
Containment pressure, temperature and dew point increase rapidly.
Containment Phase A and B isolation spray actuation occurs.
The test is complete when all alarms, bistables and indications are verified correct.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 OP-100, Reactor Coolant System MT-1213AJESS
7.0 DATE PERFORMED/TEST RESULTS 02-01-95/Sat w/Deficiencies 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT SSR 95-0180, RCDT Level/Mass increases at point of break/SI MT-1213AABS
PERFORMANCE TE<ST ABSTRACT MT-131A 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 RHR-I, RHR Pump Jrip (Pump A) 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (7) 2.0 AVAILABLEOPTIONS This malfunction tests the proper response to the trip of an RHR pump while on RHR in a "mid-loop" condition due to an overcurrent fault. Either train's pump (A-SA or B-SB) may be selected for the failure.
3.0 TESTED OPTIONS RHR Pump A trip while in mid-loop lineup.
~.0 INITIALCONDITIONS Mode 5, on RHR.
5.0 TEST DESCRIPTION The affected train's flow decreases to zero. Core exit temperatures and Reactor Vessel level increases.
The test is complete when definite indication of heat-up is observed.
6.0 BASELINE DATA/RE<FERENCE<S 6.1 Panel of experts 6.2 AOP-020, Loss of RCS Inventory, or RHR while Shutdown.
UMSV59ASTP$1TI3hL4BS
7.0 DATE PERFORMED/TEST RESULTS 10-14-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None V:lOSU59AS1PSfE131riABS
PERFORMANCE TEST ABSTRACT MT-22A 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 CCW-2, Loss of CCW to RHR Heat Exchanger in Partial Drain 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (7) and 3.1.2 (8) 2.0 AVAILABLEOPTIONS The malfunction tests proper response to a loss of CCW to an RHR heat exchanger due to the outlet valve failing shut.
Any combination of the heat exchanger valves may be selected for the failure in the position of 0 to 100 percent, closed to open.
3.0 TESTED OPTIONS RHR heat exchanger B component cooling water outlet valve fails closed.
4.0 INITIALCONDITIONS Mode 5 at mid-loop with B RHR heat exchanger in service.
5.0 TEST DESCRIPTION The test willbe complete when 1CC-167 fails closed and the listed RCS parameters are verified to change in the correct direction, which willverify the loss of CCW.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 OP-145, Component Cooling Water 6.3 AOP-014, Loss of Component Cooling Water
7.0 DATE PERFORMED/TEST RESULTS 11-16-94/SAT S.O DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT MT-52 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 CVC-2, VCT Outlet Isolation Valve Failure (LCV-115E Closed) 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (18) 2.0 AVAILABLEOPTIONS The malfunction tests the response to a VCT outlet isolation valve failure. Either of the outlet valves may be selected for the failure, with a range 0 to 100 percent, open or closed.
3.0 TE<STED OPTIONS VCT Outlet Isolation Valve (LCV-115E) failure, fully closed.
4.0 INITIALCONDITIONS Mode 1, 2, or 3, at normal pressure and temperature.
5.0 TEST DESCRIPTION The VCT outlet valve, providing suction to the pumps, shuts.
Charging flow decreases, letdown temperature increases and pressurizer level decreases.
The test is complete when all indications and alarms are verified to be correct.
6.0 BASELINE DATA/REFERENCE<S 6.1 Panel of experts 6.2 OP-107, Chemical and Volume Control System
E 7.0 DATE PERFORMED/TEST RESULTS 1-18-93/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None IP
PERFORMANCE TEST ABSTRACT MT-6102 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 EPS-8, Loss fo a Unit Auxiliary Transformer (UAT) 1.2
- ANSI/ANS 3.5, 1985, 3.1.2 (3) 2.0 AVAILABLEOPTIONS The malfunction tests proper response to the'loss of a unit auxiliary transformer due to a phase differential relay actuation.
The lockout relay is actuated by the differential relay, tripping various breakers.
Either A or B UAT is selectable for the malfunction.
3.0 TESTED OPTIONS Unit Auxiliary Transformer B phase differential relay trip.
4.0 INITIALCONDITIONS Mode 1 5.0 TEST DESCRIPTION The unit output breakers trip immediately, the exciter field breaker trips and a fast bus transfer is initiated to shift the in-house loads to the startup transformers. Ifthe plant is above the P-7 Permissive setpoint, a reactor trip will occur follwoing the turbine trip.
No busses lose power with this malfunction.
The test is complete when all indications and alarms are verified to be correct.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 OP-156.02, AC Electrical Distribution
I 7.0 DATE PERFORMED/TEST RESULTS 11-19-94/SAT S.O DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT MT-661 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 EPS-6, Loss of a 250 VDC Nonvital Bus (DP-1-250) 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (3) 2.0 AVAILABLEOPTIONS The breaker supplying the DC bus trips.
The options avaialble are a loss of bus DP 250, 1A, 1A-1, and bus 1A-2.
3.0 TESTED OPTIONS Loss of the 250 VDC bus DP-1-250 during power operation.
4.0 INITIALCONDITIONS Mode 1, at approximately 100% power.
5.0 TEST DESCRIPTION The 250-V DC bus loads deenergize on the loss of power.
The test is complete when all expected actions occur and all indications and alarms are verified to be correct.
6.0 BASELINE DATA/REFERENCES 6.1 APP-ALB-015, Annunciator Procedure 6.2 APP-ALB-20, Annunciator Procedure
7.0 DATE PERFORMED/TEST RESULTS 10-10-92/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT MT-692 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 EPS-7, Loss of Start-up Transformer 1B 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (3) 2.0 AVAILABLEOPTIONS The malfunction tests the proper response to the loss of a start-up transformer due to load side ground overcurrent.
Either of the two start-up transformers may be selected for the failure.
3.0 TESTED OPTIONS Start-up Transformer B lockout relay actuation on ground overcurrent.
4.0 INITIALCONDITIONS Mode 3.
5.0 TEST DESCRIPTION The start-up transformer trips and locks out.
Busses 1B and 1E deenergize and the EDG picks Bus 1E.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 OP-156.02, AC Electrical Distribution 6.3 APP-ALB-022 MY692ABS
7.0 DATE PERFORMED/TEST RESULTS 03-03-95/SAT w/Deficiencies 8.0 DEFICIENCIES FOUND DURING TESTING SSR 94-0536, Differential Phase B relay fiag doesn't drop.
MT692.ABS
PERFORMANCE TE<ST ABSTRACT MT-74 1.0 PROCEDURE TITLE/ANSI 3.5 REFERENCE 1.1 SGN3, Steam Generator Backleakage 1.2 ANSVANS 3.5, 1985, 3.1.2 (23) 2.0 AVAILABLEOPTIONS Backleakage to MDAFW Pump A or MDAFW Pump B 3.0 TESTED OPTIONS Backleakage to each of the MDAFW Pumps under different conditions.
4.0 INITIALCONDITIONS Mode 1 at approximately 60 percent power.
5.0 TEST DESCRIPTION AFW backleakage is first initiated with the pre-heater bypass lines open.
The backleakage source is from main feed. After the thermocouple response is verified, the affected line is vented and cooled and the malfunction is removed.
SG A pre-heater bypass line is then isolated.
The backleakage source in this condition is from the Steam Generator.
This affects another set of thermocouples and the response is verified correct.
The AFW pumps are run with the malfunction active to verify the malfunction does not interfere with normal system operation.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 AOP-010, Feedwater Malfunctions MT-74.ABS
7.0 DATE PERFORMED/TEST RESULTS 03-10-95/SAT S.O DEFICIENCIES FOUND DURING TESTING, AND TRAINING IMPACT None MT-74.ABS
PERFORMANCE TEST ABSTRACT 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 AUXILIARYFEED7fATER LINERUPTURES 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (6) 2.0 AVAILABLEOPTIONS This malfunction tests the proper response to an AFW line rupture to any one of the Steam Generators (A,B or C). The rupture can be as large as a 100% break which is a guillotine rupture of the line.
3.0 TESTED OPTIONS CFWSA, a break in the line going to SG A at a 50% break size.
4.0 INITIALCONDITIONS Mode 1 or 2 with AFW pump 1A-SA maintaining SG levels.
5.0 TEST DESCRIPTION With SG levels being maintained by AFWpump 1A-SA, an AFW feedline rupture occurs on the line feeding SG A. Indicated flow increases to the A SG but actual feed flow decreases.
The pump discharge flow control valve throttles closed to prevent pump runout.
6.0 BASELINE DATA/REFERENCES 6.1 System Description SD-137 u lOSV59551P511-'7SABS
7.0 DATE PERFORMED/TEST RESULTS 1-6-93/SAT 8.0 DEFICIE<NCIE~S FOUND DURING TESTING AND TRAININGIMPACT None MU59$PSQ-'75AM
PERFORMANCE TEST ABSTRACT MT-MSC3 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 ANNUNCIATORSYSTEM FAILURE 1.2 ANSI/ANS 3.5, 1985, 3.1.2 (3) 2.0 AVAILABLEOPTIONS Loss of power from individual 24v power supplies, inverters supplying multiple 24v power supplies, or the power distribution panel supplying the system inverters. Both systems (1 and 2) are available for malfunction simultaneously.
3.0 TESTED OPTIONS MSC3A for System 1 AND MSC3B for System 2.
4.0 INITIALCONDITIONS Mode 1 at approximately 100% power.
5.0 TEST DESCRIPTION The annunciator response to a loss of individual 24v power supplies, loss ofinverter, and loss of the power distribution panel is verified. Tests are performed on System 1 and System 2. Control features and audible alarms are verified relative to the effect of a loss of power.
6.0 BASELINE DATA/REFERENCES 6.1 The following plant drawings were researched:
EMDRAC 1364-47390,47389 Vendor manual PCP MT-MSC3.ASS
7.0 DATE PERFORMED/TEST RESULTS 03-02-95/SAT w/Deficiencies S.O DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT l
SSR 95-0225, Failed power supply on horn 2 F
MT-MSC3.ASS
PERFORMANCE TEST ABSTRACT RCS-18 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 Small Break LOCA (0.5")
1.3 ANSI/ANS 3.5, 1985, 3.1.2 (18) 2.0 AVAILABLEOPTIONS Any of the RCS hot or cold legs can be selected.
The malfunction is selected in a percentage of 0 to 100% where 100% corresponds to a 4.5 in'reak.
3.0 TESTED OPTIONS Loop B cold leg to 3% failure over 5 seconds.
4.0 INITIALCONDITIONS Mode 1, at approximately 100% power.
5.0 TEST DESCRIPTION This test is designed to test the small break LOCA malfunction.
The range of the malfunction is 0 to 100% where 100% equals a 4.5" pipe break.
The.5" break is tested here.
The results of this test will be compared to data supplied by the nuclear fuels section for this size break. At the end of the test the malfunction willbe cleared to verify leak flow decreases to zero.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 IPE-90.005 SHNPP SMALLBREAK LOCA 6.3 RAT-018 SMALLBREAKLOCA RESPONSE
7.0 DATE PERFORMED/TEST RESULTS 11-16-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None I'
~
RCS1RABS
PERFORMANCE TEST ABSTRACT RCS-6 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 Median Select Circuit Failure 1.3 ANSI/ANS 3.5, 1985, 3.1.2 (22) 2.0 AVAILABLEOPTIONS This malfunction tests the proper response to a failure of the Median select circuits. Any or all (of three) circuits can be selected for failure.
Circuit may be failed high or low.
Circuits available for failure are; Median Tave, Median Delta-T, and/or Median Tave to Load Follow (C-16).
3.0 TESTED OPTIONS RCS6A,B,C which tests all three circuits.
4.0 INITIALCONDITIONS Mode 1, at approximately 100% power.
5.0 TEST DESCRIPTION With the plant at 100% power, failures of the median select circuits are activated.
The circuit failures willbe performed independently to verify proper plant response.
Failures will be performed in the high and low directions.
The test is complete when the simulator responses of indication, annunciation, and control have been verified.
6.0 BASELINE DATA/REFERENCES EMDRAC 1364-46580, SH. 21,27,28,29,40,41,42,43,44,45,46,47,48,49,50
7.0 DATE PERFORMED/TEST RESULTS 02-03-95/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT OST-1014 NOST-006 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 OST-1014, Turbine Valve Test, Mode 1 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (10) 2.0 AVAILABLEOPTIONS The procedure tests the ability to perform the turbine valve test during normal plant operations.
3.0 TESTED OPTIONS The Turbine Overspeed Protection System test will be performed in accordance with approved plant procedures.
4.0 INITIALCONDITIONS Mode 1.
5.0 TEST DESCRIPTION The test is complete when all data collection and procedural steps are performed in accordance with the surveillance test procedure and are verified to be correct.
6.0 BASELINE DATA/REFERENCES OST-1014, Turbine Valve Test, Mode 1 1
7.0 DATE PERFORMED/TEST RESULTS 10/07/94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT.
OST-1411 NOST-022 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 OST-1411, Auxiliary Feedwater Pump 1X-SAB Operability Surveillance Test 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (10) 2.0 AVAILABLEOPTIONS This procedure tests the ability to perform the Auxiliary Feedwater Pump 1X-SAB operability surveillance test during normal plant operations, Modes 1, 2 or 3 ~
3.0 TESTED OPTIONS AuxiliaryFeedwater Pump 1X-SAB operability surveillance test performed in accordance with approved plant procedures.
4.0 INITIALCONDITIONS Mode 1.
5.0 TEST DESCRIPTION The test is complete when all data collection and procedural steps are performed in accordance with the surveillance test procedure and are verified to be correct.
6.0 BASELINE DATA/REFERENCES OST-1411, Auxiliary Feedwater Pump 1X-SAB Operability Surveillance Test UhosiMIh6STPWOSTO22.ABS
7.0 DATE PERFORMED/TEST RESULTS 11-18-94/SAT S.O DEFICIENCIES FOUND DURING TESTING ANDTINNING IMPACT None UAOSUSIhGSTPWOST022.ABS
PERFORMANCE TEST ABSTRACT OST-1087 NOST-024 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 OST-1087, Motor Driven AFW Pumps Flow Test, Mode 3 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (10) 2.0 AVAILABLEOPTIONS The surveillance procedure tests the operation of the Motor Driven AuxiliaryFeedwater Pumps under full flow conditions.
3.0 TESTED OPTIONS The test is performed in accordance with the surveillance procedure.
4.0 INITIALCONDITIONS Mode 3 5.0 TEST DESCRIPTION The test verifies AFW pump performance by recording pump differential pressure and flow which is compared to an acceptance criteria. It also verifies proper performance of the pumps pressure control valves.
6.0 BASELINE DATA/REFERENCES 6.1 OST-1087
7.0 DATE PERFORMED/TEST RESULTS 10-03-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None U:NS059AS1FWOS7VÃABS
PERFORMANCE TEST ABSTRACT OST-1080 NOST-025 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 OST-1080, Auxiliary Feedwater Pump 1X-SAB Full Flow Test 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (10) 2.0 AVAILABLEOPTIONS The surveillance procedure tests the operation ofthe Turbine Driven AuxiliaryFeedwater Pumps under full fiow conditions.
3.0 TESTED OPTIONS The test is performed in accordance with the surveillance procedure.
4.0 'NITIALCONDITIONS As described by the surveillance test procedure.
5.0 TEST DESCRIPTION The test verifies TDAFWpump performance by recording pump differential pressure and flow which is compared to an acceptance criteria. It also verifies fail safe and fullstroke times of various valves.
6.0 BASELINE DATA/REFERENCES 6.1 OST-1080
7.0 DATE PERFORMED/TEST RESULTS 11-19-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT None
PERFORMANCE TEST ABSTRACT GP-008 NOT-007 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 Plant Drain to Mid-Loop 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (8) 2.0 AVAILABLEOPTIONS The procedure tests the ability of the simulator to perform RCS drain to mid-loop from cold shutdown.
3.0 TESTED OPTIONS Reactor coolant system drain to mid-loop conditions in accordance with approved plant procedures.
4.0 INITIALCONDITIONS The plant is in cold shutdown in accordance with GP-007.
5.0 TEST DESCRIPTION The Reactor Coolant System is drained from solid plant conditions to between 70 and 75 inches below the reactor vessel flange (mid-loop).
The RCS is then filled to between 4 and 36 inches below the reactor vessel flange in accordance with GP-008.
6.0 BASELXNE DATA/REFERENCE<S 6.1 Panel of experts 6.2 GP-008, Draining the Reactor Coolant System.
7.0 DATE PERFORMED/TEST RESULTS 10-04-94/SAT w/Deficiencies 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT W0069 Vessel vent is sized too small
%0048 Audio Count Rate changes without changes in counts
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PERFORMANCE TEST ABSTRACT GP-001 NOT-008 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 Plant filland vent 1.2 ANSUANS 3.5, 1985, 3.1.1 (8) 2.0 AVAILABLEOPTIONS The procedure tests the ability of the simulator to perform RCS filland vent procedure.
3.0 TESTED OPTIONS Reactor coolant system filland vent in accordance with approved plant procedures.
4.0 INITIALCONDITIONS The plant is in cold shutdown in accordance with GP-008 5.0 TEST DESCRIPTION The Reactor Coolant System is filled from a level below the vessel flange to solid plant conditions. The RCS is then pressurized and the Reactor Coolant Pumps are run one at a time to remove the voids in the system. Between pump runs the system is depressurized
'nd vented.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 GP-001 Plant Fill and Vent NOTOOS.ASS
7.0 DATE PERFORMED/TEST RESULTS 10-06-94/SAT w/Deficiencies 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAININGIMPACT W0004 Improper pressure response on RCP start.
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PERFORMANCE TEST ABSTRACT GP-009 NOT-009 1.0 PROCEDURE TITLE/ANSI3.5 REFERENCE 1.1 Refueling 1.2 ANSI/ANS 3.5, 1985, 3.1.1 (8) 2.0 AVAILABLEOPTIONS The procedure tests the ability of the simulator to perform plant refueling procedure.
3.0 TESTED OPTIONS Plant refueling procedure.
4.0 INITIALCONDITIONS The plant is in cold shutdown, ready for refueling in accordance with GP-008.
5.0 TEST DESCRIPTION The test is complete when the plant has been refueled per GP-009 and is ready for GP-001, RCS Fill and Vent.
6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 GP-009 Refueling
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7.0 DATE PERFORME<D/TEST RESULTS 10-5-94/SAT S.O DEFICIENCIES FOUND DURING TESTING None