ML18011A904

Shnpp Operator Training Simulator Certification Quadrennial Rept.

ML18011A904
Person / Time
Site:	Harris
Issue date:	03/03/1995
From:	Robinson W CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18011A903	List: ML18011A902 ML18011A904
References
NUDOCS 9504240316
Download: ML18011A904 (97)

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SHEARON HARRIS NUCLEAR POWER PLANT OPERATOR TRAINING SIMIULATOR SIMULATOR CERTIFICATION QUADRENNIALREPORT MARCH 1995 CAROLINA POWER R LIGHT COMPANY NEW HILL, NORTH CAROLINA 950424031b 950315 PDR ADOCK .05000400 PDR

h SHNPP CERTIFICATION REPORT PACKAGE TABLE OF CONTENTS FORM 474 INTRODUCTION General Information Simulator Configuration Control Exceptions to ANSI/ANS-3.5-1985 Standard 1.0 SIMULATOR INFORMATION 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

SIMULATOR CERTIFICATION REPORT PACKAGE TABLE OF CONTENTS 2.0 SIMULATOR DESIGN DATABASE 3.0 SIMULATOR DISCREPANCY AND UPGRADE PROGRAM 3.1 Simulator Service Request Program 3.2 Engineering Service Request Implementation 3.3 Simulator Configuration Management System 4.0 SIMULATOR TESTS 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 ANNUALOPERABILITY TESTS APPENDIX B: SCHEDULE OF MALFUNCTIONTESTS APPENDIX C:

SUMMARY

OF CERTIFICATION DEFICIENCIES APPENDIX D: SIMULATOR CERTIFICATION TEST ABSTRACTS cenify.re Msscb 3, l99$

APPRI7VED BY OIEL HCL 8150018$

NRC FORM 474 U.S. NUCLEAR REGUIATORY CO)V!M)SS)ON EXFfKR 10/St/IS5 (1 042j ESR MATED BURDEN PER RESPONSE TD COMPLY WIIH THIS UIFORMATION COLIEC RON REQUEST: 120 HOURS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AHD SIMULATION FACILITYCERTIFICATION RECORDS MANAGEMENT BRANCH (MMBB7714), US. NUCLEAR REGULATORY COMMISSON, WASHINGTON, DC 205550&I AMD TO THE ~

PAPERWORK REDUCTION PROJECT (81 500188)> OFRCE OF MANAGEMENT AMD BUDGET, WASHINGTON, DC 206CL UISTTUJCRDN>k Ihb form b to be Sled for In5sl ce~ recerUUca5on (5 required), and tor any change to a stmutsUon facUtry pedormsnce testing pbn made atter inlUal submtttal such a pbn. Provkfe the folkwdng Informatkm and check the appoprbte box to In>y>ceto reason for submIUaL of DOCKET NUMBER FAQLIIY Shearon Harris Nuclear Power Plant Unit l 50-4 pp Carolina Power and Light Company 3/l5/95 lhb b to certify that

2. Documenu>Uon b ~

1. The above named facUUy Ucensee b using a slmutaUon factly consbUng solely ot a for NRC review In accordance wkh 10 CFR 55A5(b).

plant>>oferenced slmu!ator Uud meob the requlremerrts of 10 CFR 55A5.

0. Thb stmvfaUon fadUty meets the guktance contlned In AHSt/ANS SE, 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 LOCATION OF SIMULAllONFACILIIY.

Harris Simulator Harris Energy 6 Environmental Center 3932 New Hill Holleman Road New Hill, North Carolina 27562-0327 SIMULATIOHFACIL(IYPERFORMANCE TEST AIISTIIACTSATTACHED. (For p<<fcmrsnce tests cnductod ln N>> period ending wfN the date of Nb c<<rfdcorfas)

DESCRIPRON OF PERFOIUONCE TESTUIG COMPLETED. (Attach adrgrfcnaf pages as necessary and fdonbfy N>> 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 of apprordmstoly25'f p<<fonnsnce tests per year fbr N>> fcunyesr period convnoncrng wrrh X the dere of ttA c<<rrrfcadcn)

DESCBPIION OF PERFORMANCE TESTING TO BE CONDUCTED. (Alrach eddrrfcnsr psgar as necess<<y and Idonrfry Ne 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 Identify N>> 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 of completed perfonnance rosdng ht ace<<donee wrN 10 cFR M 45(b)(5)(v).

(Alrsch addrrrcnar pages as necessary and fdenrrfy N>> from dosafptfon heing conrfnuert)

Any tahe 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 BY DEUVERY Ut PERSON ONE WNIE ANT HORIH NRC FORM 474 (IM2)

~1 LLS. %ICLEAR fKGULAlTIRYCOMMISSION WASt%IGTDM, DC 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 facility consists 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 All exceptions 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 HVAC units 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 SIMULATOR INFORMATION 1.1 Simulator General

1.1.1 Owner

Carolina Power & Light Company 1.1.2 Reference Plant/Unit: Shearon Harris Nuclear Power Plant, Unit

¹1, Westinghouse 3-Loop PWR 1.1.3 Simulator Supplier: Westinghouse Electric Corporation with major upgrades by S3 Technologies 1.1.4 Ready-for-Training Date: Initial December 20, 1985 Upgrade December 27, 1994 1.1.5 Type of Report: 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. All dialed or automatic ring-down telephone calls made by the operating crew give a lighted indication in the instructor station as to who was the intended recipient of the call.

1.2.3 Systems 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 will also 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 off or on, or a switch may be failed closed. The override.may be inserted with a time delay, and analog values may be ramped in over a specified time band.

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 All parameter 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 limit exceeded 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

- RCS Boron ( 0 ppm

- 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 if using 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. All of 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 of upgrade 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 of the new modeling, and execution of those 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 SIMULATOR DESIGN 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 of a training value assessment, which would include a cost benefit analysis.

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3.0 SIMULATOR DISCREPANCY 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 All Engineering 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. If the 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 i fy. 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, l 993

4.0 SIMULATOR TESTS 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 will be 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 will be noted in subsequent quadrennial reports. However, the test program will be maintained in compliance with ANSI/ANS-3.5-1985. Appendix B lists 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. All deficiencies 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 will have an abstract included in Appendix D to this report. Any deleted certification test will be 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 ANNUALOPERABILITY TESTS 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 erations Tests ("See Note Above)

NOT-001 GP-006, Plant Shutdown NOT-002 GP-007, Plant Cooldown NOT-003 GP-002, Plant Heatup NOT-004 GP-004, Reactor Startup NOT-005 GP-005, Plant Startup NOT-006 Recovery to Rated Power Following Reactor Trip NOT-007 GP-008, Plant Drain to Mid-Loop (NEW)

NOT-008 GP-001, Plant Fill and Vent (NEW)

NOT-009 GP-009, Refueling with Cavity Fill and Drain (NEW)

Normal 0 erator Surveillance Tests (*See Note Above)

NOST-001 Power Range Heat Balance NOST-002 CVCS/SI System Operability NOST-003 RHR Pump Operability NOST-004 Containment Spray Operability NOST-005 1A-SA Emergency Diesel Generator Operability NOST-006 Turbine Valve Test (NEW) certify.qx Page 21 of 40 March 3, l99$

NOST-007 Main Steam Isolation Valve and Main Feedwater Isolation Valve Operability Test NOST-008 Daily Surveillance Requirements Modes 1 and 2 NOST-009 Daily Surveillance Requirements Modes 3 and 4 NOST-010 Reactor Coolant System Leakage Evaluation NOST-011 Shutdown Margin Calculation NOST-012 Calculation of Quadrant Power Tilt Ratio NOST-013 Main Steam Isolation Valve Operability Test NOST-014 Permissives P-6 and P-10 Verification NOST-015 1B-SB Emergency Diesel Generator Operability NOST-016 Turbine Mechanical Overspeed Trip Test NOST-017 AFW Pump 1B-SB Operability Test - Quarterly NOST-018 RHR Pump 1B-SB Operability NOST-019 Reactor Coolant Pump Seals Controlled Leakage Evaluation NOST-020 AFW Pump 1A-SA Operability Test - Quarterly NOST-021 CCW System Operability - Quarterly NOST-022 AFW Pump 1X-SAB Operability (NEW)

NOST-023 Control Rod and Rod Position Indication Exercise NOST-024 Motor Driven AFW Pump Flow Test (NEW)

NOST-025 Turbine Driven AFW Pump Full Flow Test (NEW)

Transient Tests TT-001 Manual Reactor Trip TT-002 Simultaneous Trip of all Feedwater Pumps TT-003 Simultaneous Closure of All Main Steam Isolation Valves TT-004 Simultaneous Trip of All Reactor Coolant Pumps TT-005 One Reactor Coolant Pump Trip TT-006 Turbine Trip Below P-10 TT-007 Maximum Rate Power Ramp TT-008 Maximum Size RCS Leak Inside Containment With Loss of Off-site Power TT-009 Maximum Size Steam Leak Inside Containment TT-010 Slow RCS Depressurization to Saturation Using PORV's and No SI certify.re Page22of 40 Much 3, l99$

APPENDIX B SCHEDULE OF MALFUNCTIONTESTS MALFUNCTIONTESTS FIRST YEAR MT-10162 Failure of Rod Blocks to Block (C-2, C-3, C-4)

MT-10165 Failure of Rod Block to Block (C-5)

MT-1031 Safety Injection Failure (Train B, Inadvertent)

MT-1042 Reactor Trip Breakers Fail (B fails to open)

MT-111A Pressurizer Steam Space Leak MT-113 Loss of Instrument Air to the Containment Building (AIR-1,1)

MT-1131 Pressurizer Relief Valve Failure (445A With P-11 Interlock)

MT-12 NSW Pump Trip'and Loss of NSW MT-1210 RCP A, B, C High Vibration MT-1221 Steam Generator Tube Leak (S/G B)

MT-1222 Steam Generator Tube Rupture (S/G A)

MT-1231 RCP Trip From 100 Percent Power (RCP-C)

MT-126 RCS Boron Dilution MT-135 RHR Bypass Line Leak (Train A)

MT-136 RHR Sump Valves Fail to Open MT-152 Turbine Protection Trip. Failure MT-22 Loss of CCW to RHR Heat Exchanger MT-25 Letdown Heat Exchanger Tube Leak MT-42 Logic Cabinet Urgent Failure MT-431 Dropped Rod (One Rod)

MT-44 Stuck Rod MT-51 Letdown Isolation Valve Failure (1CS-11)

MT-5182 Seal Injection Flow Control Valve Failure (HC-186 Closed)

MT-513 RCP Number 2 Seal Failure (RCP A)

MT-523 High Temperature Divert Valve (TCV-143) Failure MT-526 Boric Acid Pump Trip MT-53 Letdown Line Leak Inside Containment MT-61 Station Blackout MT-6102 Loss of Unit Aux Transformer 1B (NEW)

MT-612 Generator Output Breakers Fail to Trip MT-623 Loss of 120-VAC Uninterruptible Power (Power Supply SIII)

MT-632 Loss of 125-VDC Emergency Bus (DP 1B-SB) aaify.qa Page23 of 40 hitch 3. l99$

MT-710 Condensate Pump Trip (Pump A)

MT-712 Failure of Excess Condensate Dump Valve (Closed)

MT-724 Feedline Break Outside Containment MT-772 Feedwater Bypass Valve Failure (Open)

MT-78 Turbine Driven Auxiliary Feedwater Flow Control Valve Failure (Closed)

MT-814 Steam Failure to TDAFW Pump (1MS-72 Closed)

MT-86 Steam Generator Relief Valve Failure (Open)

MT-92 Source Range Pulse Height Discriminator Failure MT-93 Failure of Source Range High Voltage to Disconnect MT-96 Intermediate Range Channel Gamma Compensation Failure MT-97 Power Range Channel Detector Failure (Low)

MT-MSC3 Annunciator System Failure (NEW)

RCS-18 Small Break LOCA (NEW) ccrrify.rpc Page24 of 40 hlrreh 3, 199S

MALFUNCTIONTESTS SECOND YEAR MT-1032 Safety Injection Failure (Train A, Fail to Initiate)

MT-1041 Reactor Trip Breakers Fail (Both Inadvertent Open)

MT-106 False Containment Spray Actuation MT-112 Loss of Instrument Air to the Reactor (Reactor Auxiliary Building)

MT-1132 Pressurizer Relief Valve Failure (444B Without P-11 Interlock)

MT-114 Pressurizer Safety Valve Failure (8010C Open)

MT-1211 RCS Leak Within Capacity of Charging Pumps MT-1212 LOCA Within Capacity of the SI Pumps MT-1214 RCP Bearing Oil Reservoir Leak MT-1215 RCS Thermal Barrier Leak into CCW System MT-1232 Reactor Coolant Pump Trips (RCP-C)

MT-131 RHR Pump Trip (Pump A)

MT-1321 RHR HX Flow Control Valve Failure (FCV-603A Closed)

MT-151 Inadvertent Turbine Trip MT-17 Refueling Water Storage Tank Leak MT-210 Seal Water Heat Exchanger Tube Leak MT-24 Component Cooling Water Header Supply Valve Failure (Closed)

MT-26 Loss of CCW to RCP Thermal Barrier MT-31 Circulating Water Pump Trip MT-35 Loss of Condenser Vacuum Pump MT-41 Power Cabinet Urgent Failure MT-45 Ejected Rod MT-461 Uncontrolled Automatic Rod Motion MT-47 Failure of Auto Rod Blocks to Block (C-11)

MT-514 RCP Number 3 Seal Failure (RCP C)

MT-5284 Charging Line Leak in Containment Before Regen HX MT-52 VCT Outlet Isolation Valve Failure (LCV-115E Closed) (NEW)

MT-5285 Charging Line Leak Between Regen HX and 1CS-492 MT-54 Letdown Line Leak Outside Containment MT-55 Charging Pump Trip MT-56 Reactor Makeup Water Pump Trip MT-571 Letdown Pressure Control Valve Failure (PK-145 Open)

MT-651 Loss of 6.9-KV Emergency Bus (1A-SA)

MT-662 Loss of a 125-VDC Nonvital Bus (DP 1A) arri fy.rpr Page 25 of 40

MT-67 Diesel Generator Failure MT-692 Loss of Start-up Transformer 1B (NEW)

MT-712A Turbine Driven Auxiliary Feedwater Pump Trip MT-719 Main Feedwater Pump Trip (Pump B)

MT-72 Condensate Booster Pump Trip (Pump B)

MT-720 Main Feedwater Pump Recirc Valve Failure (Pump 1B)

MT-723 Feedline Break Inside Containment MT-73 Turbine Driven Auxiliary Feedwater Pump Speed Control Oscillates MT-82 Steam Break Outside Containment MT-94 Source Range Channel High Voltage Failure MT-95 Intermediate Range Channel Failure ccN fy.rpt Page26of 40

MALFUNCTIONTESTS THIRD YEAR MT-1013 Inadvertent Feedwater Isolation MT-1072 Turbine Runback Failure (Failure to Runback)

MT-1110 Pressurizer Level Control Band Shift Down MT-121 Emergency Service Water Pump Trip (NEW)

MT-1211B Uncoupled Control Rod (NEW)

MT-1213 A RCS Leak (Large Break) (NEW)

MT-124 Reactor Coolant Pump Trip (Locked Rotor)

MT-131A RHR Pump Trip (NEW)

MT-1322 RHR HX Flow Control Valve Failure (FCV-603B Open)

MT-1331 RHR HX Bypass FCV Failure (FK-605A1 Open)

MT-137 Containment Spray Pump Failure MT-138 Containment Spray Pump Discharge Valve Failure MT-141 Containment Fan Cooler Unit Trip MT-154 Turbine Vibration MT-271 Letdown Temperature Controller Failure (TK-144 Low)

MT-32 Main Condenser Tube Leak MT-333 Hotwell Level Controller Failure (LC-1901 Low)

MT-34 Loss of Condenser Vacuum MT-410 DRPI-Open or Shorted Coil MT-411 Improper Bank Overlap MT-412 Control Bank Rod Step Counter Failure MT-462 Uncontrolled Manual Rod Motion MT-5201 Failure of Charging Flow Control Valve MT-5282 Charging Pump Discharge Line Leak Before FT-122 MT-5283 Charging Line Leak Between FT-122 and 1CS-235 MT-572 Letdown Pressure Control Valve Failure (PK-145 Closed)

MT-58 Loss of Normal Letdown MT-59 VCT Divert Valve Control Failure (HUT)

MT-6101 Loss of Unit Auxiliary Transformer MT-64 Loss of 6.9 KV Auxiliary Bus (1B)

MT-661 Loss of a 250-VDC Nonvital Bus (DP-1-250) (NEW)

MT-68 Automatic Voltage Regulator Failure (High)

MT-69 Loss of Start-up Transformer MT-711A Motor Driven Auxiliary Feedwater Pump Trip ccrrlfy.re Page 27 of 40 hfrrch 3. l99S

MT-722 Feedwater Control Valve Position Failure (LCV-488 Open)

MT-74 Steam Generator Backleakage (NEW)

MT-75 Auxiliary Feedwater Line Ruptures (NEW)

MT-76 Auxiliary Feedwater Flow Control Valve Failure (Open)

MT-81 Steamline Break Inside Containment MT-815 Main Steam Header Break MT-87 MSIV Failure (S/G B Shut)

MT-89 Atmospheric Steam Dump Valve Failure (PCV-408J Open)

MT-91 Source Range Instrument Failure (N31 High)

MT-98 Power Range Channel Failure (Low)

RCS-6 Median Select Circuit Failure (NEW) certify.re Page 28 of 40 hlarch 3, l 995

MALFUNCTIONTESTS FOURTH YEAR MT-1014 Inadvertent Main Steam Isolation MT-1015 Diesel Generator Sequencer Fails to Complete Block 1 MT-10161 Failure of Rod Blocks to Block (C-1)

MT-1017 Failure of Permissive Interlock P-14 MT-1071 Turbine Runback Failure (Erroneous Runback)

MT-111 Loss of Instrument Air (Turbine Building)

MT-112A Pressurizer Spray Valve Failure MT-118 Pressurizer Backup Heaters Groups A and B Failure MT-1211 A RCS Fuel Rod Breach MT-1212A RCS Leakage into an Accumulator MT-1213 RCS Vessel Flange Leak MT-125 RCP Shaft Break Accident (RCP B)

MT-1332 RHR HX Bypass FCV Failure (FK-605B1 Closed)

MT-134 RHR to Letdown Valve Failure (HCV-142.1 Open)

MT-155 Governor Valve Failure (GV-3 Closed)

MT-157 Turbine DEH Computer Failure MT-21 Component Cooling Water Pump Trip MT-22A Loss of CCW to RHR Heat Exchanger (NEW)

MT-23 CCW Leak into the Service Water System MT-272 Letdown Temperature Controller Failure (TK-144 High)

MT-28 Loss of CCW to the Reactor Coolant Pumps MT-331 Hotwell Level Controller Failure (LC-1900 High)

MT-413 Rod Speed Deadband Control Failure MT-48 TREF Failure MT-49 DRPI Loss of Voltage MT-512 RCP Number 1 Seal Failure (RCP B)

MT-516 Boric Acid Filter Plugged MT-5181 Seal Injection Flow Control Valve Failure (HC-186 Open)

MT-5202 Failure of Charging Flow Control Valve (Closed)

MT-524 Charging Pump Suction From RWST Failure (115D Open)

MT-525 Charging Pump Mini Flow Valve Failure (1CS-182 Closed)

MT-527 Charging Line Containment Isolation Valve Failure MT-5281 Charging Line Leak on Charging Pump Suction MT-615 Diesel Generator Governor Failure cenify.qx Page 29 of 40 bsarch 3, l99S

MT-616 Diesel Generator Breaker Inadvertent Trip MT-714 Condensate Storage Tank Leak MT-715 Heater Drain Pump Trip (Pump B)

MT-771 Feedwater Bypass Valve Failure (Closed)

MT-810 Steam Dump Control Failure (Closed)

MT-811 Mechanically Stuck Condenser Dump Valve (PCV-408 Open)

MT-812 Steam Dump Permissive (P-12) Failure MT-88 Steam Generator Safety Valve Failure (Open)

MT-911 Source Range Instrument Power Fuse Blown MT-912 Intermediate Range Control Power Fuse Blown MT-913 Power Range Control Power Fuse Blown ceniky.re Page 30 of 40 March 3, l99$

APPENDIX C

SUMMARY

OF CERTIFICATION DEFICIENCIES 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 will continue to be included in the next four year testing cycle.

FT-001/SD SSR 95-0152 Physical Fidelity Test. 11 differences. 4 of these awaiting PCR completion. 7 minor label/engraving differences.

SST-001/SD SSR 95-0148 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.

SST-002/SD SSR 95-0147 Steady State Test 75%. Stability portion satisfactory.

Same type discrepancies as with SST-001 were identified.

SST-003/SD SSR 94-0693 Steady State Test 30%. Identified 7 specific differences between plant and simulator.

NOT-005/SD W-0012 Plant Start up test. Impulse pressure changes of 20 psi on turbine roll up.

TT-003/SD W-0173/ Closure of all MSIVs Transient Test. AFW flow spiked SSR 95-0153 on pump start. Cause of the Rx trip is unknown.

TT-004/SD SSR 95-0153 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 SSR 95-0151 Turbine Trip below P10 transient test. Feed Reg Bypass valve did not maintain level at 66% during transient.

W-0168/W-170 0.5 degree step change in cold leg temperature being evaluated.

TT-008/SD W-0171 LBLOCA Transient Test. Containment pressure greater than FSAR peak.

TT-009/SD SSR 95-0149 STM Break in CNM transient. Evaluate differences in AFW flow to individual SGs and containment pressure values.

TT-010/SD W-0172 Slow RCS depressurization transient. Flow spike on AFW and SI systems noted.

MT-1031/SD SSR 94-0610 Relay problem on inadvertent SI - no red first out annunciator.

MT-1211A/SD SSR 95-0223 Area Monitors did not increase ) 1000x normal MT-5152/SD SSR 94-0607 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 of certification 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 W-0069/ Vessel vent is sized too small W-0048 Audio count rate changes without changes in counts NOT-008/SD W-0004 Improper pressure response on RCP start MT-1213 A/SD SSR 95-0180 RCDT Level/Mass increases at point of break/SI MT-692/SD SSR 94-0536 Differential target relay doesn't drop MT-MSC3/SD SSR 95-0225 Chime/Horn 2 improper operation with failed power supply certify.rpt Page 32 of 40 bsarch 3, l99S

APPENDIX D SIMULATOR CERTIFICATION TEST 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 Simulator Physical Fidelity Test FT-002 Simulator Model Limits Exceeded Test (NEW)

Malfunction Tests (180)

MT-101 Inadvertent Containment Isolation Phase A (Deleted - Relay Override)

MT-1013 Inadvertent Feedwater Isolation MT-1014 Inadvertent Main Steam Isolation MT-1015 Diesel Generator Sequencer Fails to Complete Block 1 MT-10161 Failure of Rod Blocks to Block (C-1)

MT-10162 Failure of Rod Blocks to Block (C-2, C-3, C-4)

MT-10165 Failure of Rod Block to Block (C-5)

MT-1017 Failure of Permissive Interlock P-14 MT-102 Inadvertent Containment Isolation Phase B (Deleted - Relay Override)

MT-1031 Safety Injection Failure (Train B, Inadvertent)

MT-1032 Safety Injection Failure (Train A, Fail to Initiate)

MT-1041 Reactor Trip Breakers Fail (Both Inadvertent Open)

MT-1042 Reactor Trip Breakers Fail (B fails to open)

MT-106 False Containment Spray Actuation MT-1071 Turbine Runback Failure (Erroneous Runback)

MT-1072 Turbine Runback Failure (Failure to Runback)

MT-111 " Loss of Instrument Air (Turbine Building)

MT-1110 Pressurizer Level Control Band Shift Down MT-111 A Pressurizer Steam Space Leak certify.re Page 33 of 40 bsarch 5, l99S

MT-112 Loss of Instrument Air to the Reactor (Reactor Auxiliary Building)

MT-112A Pressurizer Spray Valve Failure MT-113 Loss of Instrument Air to the Containment Building (AIR-1,1)

MT-1131 Pressurizer Relief Valve Failure (445A With P-11 Interlock)

MT-1132 Pressurizer Relief Valve Failure (444B Without P-11 Interlock)

MT-114 Pressurizer Safety Valve Failure (8010C Open)

MT-1151 Pressurizer Pressure Channel Failure (PT-444 High) (Deleted - Inst. Override)

MT-1152 Pressurizer Pressure Channel Failure (PT-445 Low) (Deleted - Inst. Override)

MT-1161 Pressurizer Pressure Channel Failure (PT-456 High) (Deleted - Inst. Override)

MT-1162 Pressurizer Pressure Channel Failure (PT-457 Low) (Deleted - Inst. Override)

MT-1171 Pressurizer Level Channel Failure (LT-459 Low) (Deleted - Inst. Override)

MT-1172 Pressurizer Level Channel Failure (LT-459 High) (Deleted - Inst. Override)

MT-118 Pressurizer Backup Heaters Groups A and B Failure MT-12 NSW Pump Trip and Loss of NSW MT-121 Emergency Service Water Pump Trip (NEW)

MT-1210 RCP A, B, C High Vibration MT-1211 RCS Leak Within Capacity of Charging Pumps MT-1211A RCS Fuel Rod Breach MT-1211B Uncoupled Control Rod (NEW)

MT-1212 LOCA Within Capacity of the SI Pumps MT-1212A RCS Leakage into an Accumulator MT-1213 RCS Vessel Flange Leak MT-1213 A RCS Leak (LOCA) (NEW)

MT-1214 RCP Bearing Oil Reservoir Leak MT-1215 RCS Thermal Barrier Leak into CCW System MT-1216 RCS Flow Transmitter Failure (FT-436 w) (Deleted - Inst. Override)

MT-1221 Steam Generator Tube Leak (S/G B)

MT-1222 Steam Generator Tube Rupture (S/G A)

MT-1231 RCP Trip From 100 Percent Power (RCP-C)

MT-1232 Reactor Coolant Pump Trips (RCP-C)

MT-124 Reactor Coolant Pump Trip (Locked Rotor)

MT-125 RCP Shaft Break Accident (RCP B)

MT-126 RCS Boron Dilution MT-1271 RCS Control RTD Failure (TE-411B High) (Deleted - Inst. Override)

MT-1281 RCS Protection RTD Failure (TE-412B Low) (Deleted - Inst. Override)

MT-1282 RCS Protection RTD Failure (TE-422B High) (Deleted - Inst. Override)

MT-129 RCS WR Pressure Transmitter Failure (PT-403 High) (Deleted - Inst. Override)

MT-13 Containment Fan Cooler Unit Trip (Deleted - Renamed MT-141)

MT-131 RHR Pump Trip (Pump A) ccNly.qa Page34of 40 Marcb 3, 199S

MT-131A RHR Pump Trip (Pump A) (NEW)

MT-1321 RHR HX Flow Control Valve Failure (FCV-603A Closed)

MT-1322 RHR HX Flow Control Valve Failure (FCV-603B Open)

MT-1331 RHR HX Bypass FCV Failure (FK-605A1 Open)

MT-1332 RHR HX Bypass FCV Failure (FK-605B1 Closed)

MT-134 RHR to Letdown Valve Failure (HCV-142.1 Open)

MT-135 RHR Bypass Line Leak (Train A)

MT-136 RHR Sump Valves Fail to Open MT-137 Containment Spray Pump Failure MT-138 Containment Spray Pump Discharge Valve Failure MT-141 Containment Fan Cooler Unit Trip (Was MT-13)

MT-151 Inadvertent Turbine Trip MT-1519 Turbine 1st Stage Press. Xmtr Failure (PT-446 Low) (Deleted - Inst. Override)

MT-152 Turbine Protection Trip Failure MT-154 Turbine Vibration MT-155 Governor Valve Failure (GV-3 Closed)

MT-157 Turbine DEH Computer Failure MT-17 Refueling Water Storage Tank Leak MT-21 Component Cooling Water Pump Trip MT-210 Seal Water Heat Exchanger Tube Leak MT-22 Loss of CCW to RHR Heat Exchanger MT-22A Loss of CCW to RHR Heat Exchanger (NEW)

MT-23 CCW Leak into the Service Water System MT-24 Component Cooling Water Header Supply Valve Failure (Closed)

MT-25 Letdown Heat Exchanger Tube Leak MT-26 Loss of CCW to RCP Thermal Barrier MT-271 Letdown Temperature Controller Failure (TK-144 Low)

MT-272 Letdown Temperature Controller Failure (TK-144 High)

MT-28 Loss of CCW to the Reactor Coolant Pumps MT-31 Circulating Water Pump Trip MT-32 Main Condenser Tube Leak MT-331 Hotwell Level Controller Failure (LC-1900 High)

MT-333 Hotwell Level Controller Failure (LC-1901 Low)

MT-34 Loss of Condenser Vacuum MT-35 Loss of Condenser Vacuum Pump MT-41 Power Cabinet Urgent Failure MT-410 DRPI Open or Shorted Coil MT-411 Improper Bank Overlap MT-412 Control Bank Rod Step Counter Failure certify.rpt Page 35 of 40 hi arch 3, l99$

MT-413 Rod Speed Deadband Control Failure MT-42 Logic Cabinet Urgent Failure MT-431 Dropped Rod (One Rod)

MT-44 Stuck Rod MT-45 Ejected Rod MT-46 Uncontrolled Rod Motion (Deleted - Split into MT-461 & MT-462)

MT-461 Uncontrolled Automatic Rod Motion (Was part of MT-46)

MT-462 Uncontrolled Manual Rod Motion (Was part of MT-46)

MT-47 Failure of Auto Rod Blocks to Block (C-11)

MT-48 TREF Failure MT-49 DRPI Loss of Voltage MT-51 Letdown Isolation Valve Failure (1CS-11)

MT-5111 VCT Level Transmitter Failure (LT-112 High) (Deleted - Inst. Override)

MT-5112 VCT Level Transmitter Failure (LT-115 Low) (Deleted - Inst. Override)

MT-512 RCP Number 1 Seal Failure (RCP B)

MT-513 RCP Number 2 Seal Failure (RCP A)

MT-514 RCP Number 3 Seal Failure (RCP C)

MT-5151 Boric Acid Flow Xmtr. Failure (FT-113 to 20 gpm) (Deleted - Inst. Override)

MT-5152 Boric Acid Flow Xmtr. Failure (FT-113 to 0 gpm) (Deleted - Inst. Override)

MT-516 Boric Acid Filter Plugged MT-5181 Seal Injection Flow Control Valve Failure (HC-186 Open)

MT-5182 Seal Injection Flow Control Valve Failure (HC-186 Closed)

MT-52 VCT Outlet Isolation Valve Failure (LCV-115E Closed) (NEW)

MT-5201 Failure of Charging Flow Control Valve MT-5202 Failure of Charging Flow Control Valve (Closed)

MT-523 High Temperature Divert Valve (TCV-143) Failure MT-524 Charging Pump Suction From RWST Failure (115D Open)

MT-525 Charging Pump Mini Flow Valve Failure (1CS-182 Closed)

MT-526 Boric Acid Pump Trip MT-527 Charging Line Containment Isolation Valve Failure MT-5281 Charging Line Leak on Charging Pump Suction MT-5282 Charging Pump Discharge Line Leak Before FT-122 MT-5283 Charging Line Leak Between FT-122 and 1CS-235 MT-5284 Charging Line Leak in Containment Before Regen HX MT-5285 Charging Line Leak Between Regen HX and 1CS-492 MT-53 Letdown Line Leak Inside Containment MT-54 Letdown Line Leak Outside Containment MT-55 Charging Pump Trip MT-56 Reactor Makeup Water Pump Trip Page 36 of 40 March 3, 199S

MT-571 Letdown Pressure Control Valve Failure (PK-145 Open)

MT-572 Letdown Pressure Control Valve Failure (PK-145 Closed)

MT-58 Loss of Normal Letdown MT-59 VCT Divert Valve Control Failure (HUT)

MT-61 Station Blackout MT-610 Loss of Unit Auxiliary Transformer (Deleted - Renamed MT-6101)

MT-6101 Loss of Unit Auxiliary Transformer A phase (Was MT-610)

MT-6102 Loss of Unit Auxiliary Transformer B phase (NEW)

MT-612 Generator Output Breakers Fail to Trip MT-615 Diesel Generator Governor Failure MT-616 Diesel Generator Breaker Inadvertent Trip MT-62 Loss of 120-VAC Uninterruptible Power (Power Supply SIII) (Deleted-Renamed MT-623)

MT-623 Loss of 120-VAC Uninterruptible Power (Power Supply SIII) (Was MT-62)

MT-63 Loss of 125-VDC Emergency Bus (DP 1B-SB) (Deleted - Renamed MT-632)

MT-632 Loss of 125-VDC Emergency Bus (DP 1B-SB) (Was MT-63)

MT-64 Loss of 6.9 KV Auxiliary Bus (1B)

MT-65 Loss of 6.9-KV Emergency Bus (1A-SA) (Deleted - Renamed MT-651)

MT-651 Loss of 6.9-KV Emergency Bus (1A-SA) (Was MT-65)

MT-66 Loss of a 125-VDC Nonvital Bus (DP 1A) (Deleted - Renamed MT-662)

MT-661 Loss of a 250-VDC Nonvital Bus (DP-1-250) (NEW)

MT-662 Loss of a 125-VDC Nonvital Bus (DP 1A) (Was MT-66)

MT-67 Diesel Generator Failure MT-68 Automatic Voltage Regulator Failure (High)

MT-69 Loss of Start-up Transformer 1A MT-692 Loss of Start-up Transformer 1B (NEW)

MT-710 Condensate Pump Trip (Pump A)

MT-711A Motor Driven Auxiliary Feedwater Pump Trip MT-712 Failure of Excess Condensate Dump Valve (Closed)

MT-712A Turbine Driven Auxiliary Feedwater Pump Trip MT-714 Condensate Storage Tank Leak MT-715 Heater Drain Pump Trip (Pump B)

MT-719 Main Feedwater Pump Trip (Pump B)

MT-72 Condensate Booster Pump Trip (Pump B)

MT-720 Main Feedwater Pump Recirc Valve Failure (Pump 1B)

MT-721 Feedwater Flow Transmitter Failure (FT-466 Low) (Deleted - Inst. Override)

MT-722 Feedwater Control Valve Position Failure (LCV-488 Open)

MT-723 Feedline Break Inside Containment MT-724 Feedline Break Outside Containment ccrrify.rpr Page 37 of 40 March 3, l 993

0 MT-725 Steam Generator Level Chan. Failure (LT-496 Low) (Deleted - Inst. Override)

MT-73 Turbine Driven Auxiliary Feedwater Pump Speed Control Oscillates MT-74 Steam Generator Backleakage (NEW)

MT-75 Auxiliary Feedwater Line Ruptures (NEW)

MT-76 Auxiliary Feedwater Flow Control Valve Failure (Open)

MT-771 Feedwater Bypass Valve Failure (Closed)

MT-772 Feedwater Bypass Valve Failure (Open)

MT-78 Turbine Driven Auxiliary Feedwater Flow Control Valve Failure (Closed)

MT-81 Steamline Break Inside Containment MT-810 Steam Dump Control Failure (Closed)

MT-811 Mechanically Stuck Condenser Dump Valve (PCV-408 Open)

MT-812 Steam Dump Permissive (P-12) Failure MT-814 Steam Failure to TDAFW Pump (1MS-72 Closed)

MT-815 Main Steam Header Break MT-82 Steam Break Outside Containment MT-83 Steam Header Press. Detector Failure (PT-464 High) (Deleted - Inst. Override)

MT-84 Steam-Line Flow Transmitter FT-494 (Deleted - Inst. Override)

MT-85 Steam Generator Press. Xmtr. Failure (PT-485 High) (Deleted - Inst. Override)

MT-86 Steam Generator Relief Valve Failure (Open)

MT-87 MSIV Failure (S/G B Shut)

MT-88 Steam Generator Safety Valve Failure (Open)

MT-89 Atmospheric Steam Dump Valve Failure (PCV-408J Open)

MT-91 Source Range Instrument Failure (N31 High)

MT-911 Source Range Instrument Power Fuse Blown MT-912 Intermediate Range Control Power Fuse Blown MT-913 Power Range Control Power Fuse Blown MT-92 Source Range Pulse Height Discriminator Failure MT-93 Failure of Source Range High Voltage to Disconnect MT-94 Source Range Channel High Voltage Failure MT-95 Intermediate Range Channel Failure MT-96 Intermediate Range Channel Gamma Compensation Failure MT-97 Power Range Channel Detector Failure (Low)

MT-98 Power Range Channel Failure (Low)

MT-MSC3 Annunciator System Failure (NEW)

RCS-18 Small Break LOCA (NEW)

RCS-6 Median Select Circuit Failure (NEW) ccnify.rior Page38of 40 March 3, 199S

Normal 0 erator Surveillance Tests (25)

NOST-001 OST-1004, Power Range Heat Balance NOST-002 OST-1007, CVCS/SI System Operability NOST-003 OST-1008, RHR Pump Operability NOST-004 OST-1009, Containment Spray Operability NOST-005 OST-1013, 1A-SA Emergency Diesel Generator Operability NOST-006 OST-1014, Turbine Valve Test (NEW)

NOST-007 OST-1018, Main Steam Isolation Valve and Main Feedwater Isolation Valve Operability Test NOST-008 OST-1021, Daily Surveillance Requirements Modes 1 and 2 NOST-009 OST-1022, Daily Surveillance Requirements Modes 3 and 4 NOST-010 OST-1026, Reactor Coolant System Leakage Evaluation NOST-011 OST-1036, Shutdown Margin Calculation NOST-012 OST-1039, Calculation of Quadrant Power Tilt Ratio NOST-013 OST-1046, Main Steam Isolation Valve Operability Test NOST-014 OST-1054, Permissives P-6 and P-10 Verification NOST-015 OST-1073, 1B-SB Emergency Diesel Generator Operability NOST-016 OST-1075, Turbine Mechanical Overspeed Trip Test NOST-017 OST-1076, AFW Pump 1B-SB Operability Test - Quarterly NOST-018 OST-1092, RHR Pump 1B-SB Operability NOST-019 OST-1126, Reactor Coolant Pump Seals Controlled Leakage Evaluation NOST-020 OST-1211, AFW Pump 1A-SA Operability Test - Quarterly NOST-021 OST-1316, CCW System Operability - Quarterly NOST-022 OST-1411, AFW Pump 1X-SAB Operability (NEW)

NOST-023 OST-1005, Control Rod and Rod Position Exercise NOST-024 OST-1087, Motor Driven AFW Pumps Flow Test (NEW)

NOST-025 OST-1088, Turbine Driven AFW Pump Full Flow Test (NEW)

Normal 0 erations Tests (9)

NOT-001 GP-006, Plant Shutdown NOT-002 GP-007, Plant Cooldown NOT-003 GP-002, Plant Heatup NOT-004 GP-004, Reactor Startup NOT-005 GP-005, Plant Startup NOT-006 Recovery to Rated Power Following Reactor Trip NOT-007 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-001 100 Percent Power Accuracy Test SST-002 75 Percent Power Accuracy Test SST-003 30 Percent Power Accuracy Test Transient Tests (10)

TT-001 Manual Reactor Trip TT-002 Simultaneous Trip of all Feedwater Pumps TT-003 Simultaneous Closure of All Main Steam Isolation Valves TT-004 Simultaneous Trip of All Reactor Coolant Pumps TT-005 One Reactor Coolant Pump Trip TT-006 Turbine Trip Below P-10 TT-007 Maximum Rate Power Ramp TT-008 Maximum Size RCS Leak Inside Containment With Loss of Off-site Power TT-009 Maximum Size Steam Leak Inside Containment TT-010 Slow RCS Depressurization to Saturation Using PORV's and No SI 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 will not 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

70 DATE PERFORMED/TEST RESULTS c 03-01-95/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAINING IMPACT None c

P~ J ~ ~J 1 g'>

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E gk FT402ABS Page 2 of 2

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 will start 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 TRAINING IMPACT 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 if less 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 The first with an uncoupled rod malfunction active, the second with no 10'mps.

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 will be inserted on a control rod to allow it to drop into the coxe and the reactivity effect will be 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/ANSI 3.5 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 TRAINING IMPACT 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 TRAINING IMPACT 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 will be complete when 1CC-167 fails closed and the listed RCS parameters are verified to change in the correct direction, which will verify 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 TRAINING IMPACT 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 TRAINING IMPACT 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. If the 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 TRAINING IMPACT None

PERFORMANCE TEST ABSTRACT MT-661 1.0 PROCEDURE TITLE/ANSI 3.5 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 TRAINING IMPACT None

PERFORMANCE TEST ABSTRACT MT-692 1.0 PROCEDURE TITLE/ANSI 3.5 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 INITIAL CONDITIONS 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/ANSI 3.5 REFERENCE 1.1 AUXILIARYFEED7fATER LINE RUPTURES 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 AFW pump 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 TRAINING IMPACT None MU59$PSQ-'75AM

PERFORMANCE TEST ABSTRACT MT-MSC3 1.0 PROCEDURE TITLE/ANSI 3.5 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 of inverter, 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 TRAINING IMPACT 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 will be cleared to verify leak flow decreases to zero.

6.0 BASELINE DATA/REFERENCES 6.1 Panel of experts 6.2 IPE-90.005 SHNPP SMALL BREAK LOCA 6.3 RAT-018 SMALLBREAK LOCA RESPONSE

7.0 DATE PERFORMED/TEST RESULTS 11-16-94/SAT 8.0 DEFICIENCIES FOUND DURING TESTING AND TRAINING IMPACT 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 TRAINING IMPACT 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 TRAINING IMPACT 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 Auxiliary Feedwater 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 AND TINNING 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 Auxiliary Feedwater 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 TRAINING IMPACT 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 of the Turbine Driven Auxiliary Feedwater 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 TDAFW pump performance by recording pump differential pressure and flow which is compared to an acceptance criteria. It also verifies fail safe and full stroke 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 TRAINING IMPACT 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 TRAINING IMPACT 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/ANSI 3.5 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 TRAINING IMPACT 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