ML18095A409
| ML18095A409 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/09/1990 |
| From: | Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLR-N90163, NUDOCS 9008130282 | |
| Download: ML18095A409 (179) | |
Text
Public Service Electric and Gas Company Stanley LaBruna
- Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609-339-4800 Vice President - Nuclear Operations
'AUG O 9 1990 NLR-N90163 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Gentlemen:
SALEM SIMULATOR CERTIFICATION FOR 10CFR55.45(b) (2)
SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 Public Service Electric and Gas Company (PSE&G) certified that the Salem plant-referenced simulator meets NRC regulations, in a letter dated December 28, 1989.
The NRC staff conducted a review of this submittal for the purpose of establishing the completeness of documentation.
This review uncovered several examples of inadequate information/data.
bn May 12, 1990, the staff notified PSE&G of these deficiencies and requested additional information/data.
We have enclosed responses to the ten staff comments.
Additional information, where required, is supplied in a tabbed binder.
The binder tabs correspond to the comment number.
Please contact us if you have any questions regarding this transmittal.
Sincerely, Attachment
~-~-~ ~~~~~~~~-
r 9oos1302s2 90~309
~
PDR ADOCK 05000272 P
Document Control Desk NLR-N90163 w/o attachment c
Mr. J. c. Stone 2
Licensing Project Manager - Salem Mr. T. Johnson Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 AUG o 9 1990
NLR-N90163 ATTACHMENT comment 1 In Appendix 1 to Attachment 3 on page 2 of 2 it is noted that the "Impact assessment" of hardware differences is not applicable to Unit 1.
However, there are over 300 items in the Unit 1 list of hardware differences.
ANSI/ANS-3.5-1985, Section 3.2, allows deviations in dimensions and arrangement of panels provided they do not detract from training and dimensional deviations in the configuration of components and instrumentation provided they do not impact the actions to be taken by the operator.
With regard to Unit 1 simulation, please provide a description of the method you use to determine whether deviations "detract from training" or "impact the actions to be taken by the operator" or provide justification for exception to making such determinations.
Response
PSE&G has revised the evaluation process (SP-206, Appendix 1, Revision 1) to include an assessment of Unit 1 differences.
A "Simulator Performance Engineer (SRO)" determines whether the difference detracts from training and the impact on operator action.
These assessments are reevaluated during our annual Training Review Committee meeting (Operations Department, Training Department and Simulator Support Group).
Existing differences between the reference plants and_the simulator are addressed during operator initial and continuing training classes.
Refer to <Tab 1> for a copy of the revised procedure SP-206, Appendix 1 and the most recent Unit 1 differences database.
Comment 2 Pages 2 through 6 of Appendix 3 to Attachment 3 are missing.
Please provide these pages.
Response
Page 1 of Appendix 3 to Attachment 3 was incorrectly numbered.
It should have read "Page 1 of 1." Refer to <Tab 2> for the page correction.
Comment 3 There is a table in Appendix 6 to Attachment 3 entitled, "SALEM SIMULATOR PROCEDURE LIST - UNIT 1 & UNIT 2_ SORTED BY PROCEDURE NUMBER."-
On pages 6 and 9 of this bible it is noted that the units have different Radiation Monitoring Systems.
ANSI/ANS-3.5-1985, Section 3.3.1, requires the inclusion of systems and the degree of simulation to be to the extent necessary to perform the normal plant evolutions in Section 3.1.1 and respond to the malfunctions in Section 3.1.2.
Please confirm that simulation of the Unit 1 Radiation Monitoring System is not required to meet the guidance of ANSI/ANS-3.5-1985 or provide justification for exception.
Response
PS.E&G requests exception from this requirement and provides the following justification. Unit 1 and Unit 2 Radiation Monitoring Systems perform identical functions.
The difference lies in the processing and displaying of information.
Unit 2 is a more advanced computerized system.
Operators obtain Unit 2 radiological information via a computer terminal; class lE instrumentation data is also available from displays in the control room.
Operators obtain Unit 1 radiological information from analog meters_ (control room) and strip chart recorders (protection racks).
Since Unit 2 is the more advanced complex system, simulating that system provides a higher confidence level that the operators will respond correctly when required.
The simpler Unit 1 analog system can be adequately addressed during training classes.
Both systems are covered during operator initial and continuing training sessions.
Comment 4 On pagesl5 through 22 of the table mentioned in question 3 of this enclosure many differences are noted in the alarm response procedures between Unit 1 and Unit 2.
ANSI/ANS-3.5-1985 requires that the simulator produce alarms in a manner of the reference plant.
Please confirm that, for Unit 1, the simulator meets this guidance or provide justification for exception.
Response
PSE&G requests exception from this requirement and provides the following justification. Unit 1 overhead annunciator (OHA) differences were evaluated for impact on operator actions and training (refer to question 1 response) and received a disposition of "slight" (may result in minor disorientation, no error likely).
Unit 1 OHAs and associated alarm response procedures are covered during operator initial and continuing training sessions.
Comment 5 On page 27 of the table mentioned in comment 3 of this enclosure it is noted that, for procedures 2-EOP-TRIP-5 and 2-EOP-TRIP-6, the simulator cannot produce "a bubble in the head."
ANSI/ANS-3.5-1985, in Section 3.1.2, under malfunction (1) (c),
requires "demonstration of saturation conditions."
Please confirm the simulator meets this guidance or provide justification for exception.
Response
Salem simulator fidelity does demonstrate "saturation conditions" within the context of ANSI/ANS-3.5-1985.
our performance test procedure PTP-010, "Slow Primary System Depressurization to Saturated Conditions" exercises this feature.
The Reactor Coolant System (RCS) fluid condition at equilibrium pressure
(saturation conditi. may be characterized by-iding in the core, upper plenum and hot legs; and saturated or slightly subcooled liquid in the cold legs.
The saturation condition is homogeneous throughout the RCS.
Drawing a bubble in the "head" is presently OUT-OF-SCOPE.
Drawing a "bubble in the head" is addressed during operator initial and continuing training classroom sessions.
Comment 6 Some pages appear to be missing from Appendix 9 to Attachment 3.
For the "Salem Simulator Steady-State Operations Procedures,"
SSOP-1, SSOP-2 and SSOP-3 the pages between "i" and "page 2 of 10 11 appear to be missing.
For SSOP-3, pages 1 through 3 of the calorimetric calculation at the end of the procedure appear to be missing.
Please provide these missing pages.
Response
Refer to <Tab 6> for the missing pages from "Salem Simulator Steady-State Operations Procedures" SSOP-1, SSOP-2 and SSOP-3.
PSE&G intentionally withheld pages 1 through 3 of the "Calorimetric Calculation."
The Calorimetric calculation data sheets used by the simulator staff are extracted from the reference plant Reactor Engineers Manual, Part 2, Section 2.6.
The first three pages contain the purpose, applicability, initial conditions, precautions and body of the procedure.
These pages were not sent because ~hey contain no historical information.
Refer to <Tab 6> for a copy of pages 1 through 3.
Comment 7 In Appendix 10 to Attachment 3 pages 4 and 5 of 6 are missing.
Please provide these pages.
Response
Refer to <Tab 7> for a coinplete copy of Appendix 10.
Comment 8 With regard to Appendix 11 to Attachment 3 it is noted that you have substituted "Reactor Trip Due To Main Generator Trip" (Procedure S-PTP-001) and "Pressurizer PORV Stuck Open" (Procedure S-PTP-010) for the ANSI/ANS-3.5-1985 Appendix B transients "Manual Reactor Trip" and "Slow primary system depressurization to satu_rated condition using pressurizer rel-ief
-- or safety valve stuck open," respectively [emphasis added].
Please provide test abstracts for the Appendix B transients or provide justification for exception to these tests.
Response
PSE&G intentionally replaced "Manual Reactor Trip" with "Main Generator Trip" because we had actual plant data for the Main Generator Trip.
Simulator response is comparable for a Manual Reactor Trip and a Main Generator Trip, so we made a direct substitution.
We have since added Manual Reactor Trip to our list of Appendix B annual malfunctions.
Refer to <Tab 8> for the test abstract.
Our procedure S-PTP-010 does demonstrate depressurization to saturated condition.
The test procedure title has been changed to use the terminology in Appendix B of ANSI/ANS-3.5-1985.
We expanded the procedure general sequence and termination point description to clearly reflect "saturation" condition.
The test results remain unchanged so we are forwarding only a revised copy of the Performance Test Procedure.
Refer to <Tab 8> for revised copies of S-PTP-010 and S-PTP-034.
Comment 9 The majority of the computer printouts in Appendix 13 and 14 to are not legible.
Please provide darker, more legible, copies.
Response
Refer to <Tab 9> for copies of Appendices 13.and 14 of SP-206.
Comment 10 on page 12 of Appendix 14 to Attachment 3, under "Feedback Number S-FB-89-061," it is noted that the simulator does not have a radio.
ANSI/ANS-3.5-1985, in Section 3.2.3, requires "Communication systems that a control room operator would use to communicate with an auxiliary operator or other support activities shall be operational to the extent that the simulator instructor, when performing these remote activities, shall be able to communicate over the appropriate communication system."
Please confirm the simulator meets this guidance or provide justification for exception.
Response
PSE&G requests exception from this requirement and provides the following justification.
Feedback report number S-FB-89-061 was in error.
The Salem simulator has the identical radio used in the reference plant.
This radio is used only during Emergency Preparedness drills to communicate with the- "actual" plant remote shutdown-panel auxiliary operator.
The plant control room operators can use either the PA system or a radio to communicate with field operators.
The Salem simulator currently possesses a fully simulated PA system.
We are evaluating whether to simulate the radio function or modify the existing radio for everyday training.
-~
(saturation conditi!) may be characterized by.iding in the core, upper plenum and hot legs; and saturated or slightly subcooled liquid in the cold legs.
The saturation condition is homogeneous throughout the RCS.
Drawing a bubble in the "head" is presently OUT-OF-SCOPE.
Drawing a "bubble in the head" is addressed during operator initial and continuing training classroom sessions.
Comment 6 Some pages appear to be missing from Appendix 9 to Attachment 3.
For the "Salem Simulator Steady-State Operations Procedures,"
SSOP-1, SSOP-2 and SSOP-3 the pages between "i" and "page 2 of 10 11 appear to be missing.
For SSOP-3, pages 1 through 3 of the calorimetric calculation at the end of the procedure appear to be missing.
Please provide these missing pages.
Response
Refer to <Tab 6> for the missing pages from "Salem Simulator Steady-State Operations Procedures" SSOP-1, SSOP-2 and SSOP-3.
PSE&G intentionally withheld pages 1 through 3 of the "Calorimetric Calculation."
The Calorimetric calculation data sheets used by the simulator staff are extracted from the reference plant Reactor Engineers Manual, Part 2, Section 2.6.
The first three pages contain the purpose, applicability, initial conditions, precautions and body of the procedure.
These pages were not sent because they contain no historical information.
Refer to <Tab 6> for a copy of pages 1 through 3.
Comment 7 In Appendix 10 to Attachment 3 pages 4 and 5 of 6 are missing.
Please provide these pages.
Response
Refer to <Tab 7> for a complete copy of Appendix 10.
Comment 8 With regard to Appendix 11 to Attachment 3 it is noted that you have substituted "Reactor Trip Due To Main Generator Trip" (Procedure S-PTP-001) and "Pressurizer PORV Stuck Open" (Procedure S-PTP-010) for the ANSI/ANS-3.5-1985 Appendix B transients "Manual Reactor Trip" and "Slow primary system depressurization to saturated condition using pressurizer relief_
or safety valve stuck open," respectively [emphasis added].
Please provide test abstract.s for the Appendix B transients or provide justification for exception to these tests.
Response
PSE&G intentionally replaced "Manual Reactor Trip" with "Main Generator Trip" because we had actual plant data for the Main Generator Trip.
Simulator response is comparable for a Manual Reactor Trip and a Main Generator Trip, so we made a direct substitution.
We have since added Manual Reactor Trip to our list of Appendix B annual malfunctions.
Refer to <Tab 8> for the test abstract.
Our procedure S-PTP-010 does demonstrate depressurization to saturated condition.
The test procedure title has been changed to use the terminology in Appendix B of ANSI/ANS-3.5-1985.
We expanded the procedure general sequence and termination point description to clearly reflect "saturation" condition.
The test results remain unchanged so we are forwarding only a revised copy of the Performance Test Procedure.
Refer to <Tab 8> for revised copies of S-PTP-010 and S-PTP-034.
Comment 9 The majority of the computer printouts in Appendix 13 and 14 to are not legible.
Please provide darker, more legible, copies.
Response
Refer to <Tab 9> for copies of Appendices 13 and 14 of SP-206.
Comment 10 On page 12 of Appendix 14 to Attachment 3, under "Feedback Number S-FB-89-061," it is noted that the simulator does not have a radio.
ANSI/ANS-3.5-1985, in Section 3.2.3, requires "Communication systems that a control room operator would use to communicate with an auxiliary operator or other support activities shall be operational to the extent that the simulator instructor, when performing these remote activities, shall be able to communicate over the appropriate communication system."
Please confirm the simulator meets this guidance or provide justification for exception.
Response
PSE&G requests exception from this requirement and provides the following justification.
Feedback report number S-FB-89-061 was in error.
The Salem simulator has the identical radio used in the reference plant.
This radio is used only during Emergency Preparedness drills to communicate with the "actual" plant auxiliary operators.
The plant control room operators can use either the PA system or a radio to communicate with field operators.
The Salem simulator currently possesses a fully simulated PA system.
We are evaluating whether to simulate the radio function or modify the existing radio for everyday training.
SP-206 HOPE CREEK & SALEM SIMULATORS Station vs Simulator Hardware Differences Database The physical fidelity between the simulator and station control room hardware is reviewed as part of the Annual Review.
The arrangement and control room environment are evaluated utilizing one or more of the following as determined by the Simulator Configuration Engineer:
Static Photography:
All simulated panels of the control room, are photographed in color with sufficient resolution to determine:
all switch, status light, and annunciator engravings all component, meter, recorder, and indicator scale graduations, mimics, and labeling all operator aid lamicoid engravings and placement Video Recording:
The entire control room is recorded on videotape to obtain an overall orientation and placement of movable objects/materials such as tables, bookcases, procedures, etc.
Audio Recording:
The control room audio environment is recorded to obtain a background "noise" level.
All annunciators are individually recorded for volume, tone, and rate characteristics.
The collected data is reviewed and the simulator panel drawings are updated by the Simulator Support Group as required.
All differences are identified in the associated Simulator Differences Database and evaluated for training/operator impact and assigned one of the following status codes by the Simulator Performance Engineer:
N None No confusion or error likely (e.g. change in font)
S Slight May result in minor disorientation, no error likely (e.g. smoked vs. clear bezel, letter spacing, etc)
M Moderate May result in confusion, recoverable operator error possible (e.g. labeling or scale graduation errors)
L Large May inhibit operator attention/action, non-recoverable operator error possible (e.g.
equipment/control relocation, addition, or removal)
All data will be entered into a database from which a monthly Simulator Status Report is generated.
This report will be used to track progress and aid the Principal Training Supervisor - Simulator Support in material acquisition and work scheduling.
The database is comprised of the following fields:
PHOTO NUMBER/VIDEO FRAME Reference documenting the hardware differences.
COMPONENT The affected component/device.
PANEL The affected panel where the difference is located.
Appendix 1 Page 1 of 2 Date: 07/06/90 Rev. : --=1=---
ITEM NUMBER The simulator designator for the affected component/device GROUP NUMBER The bezel group number for the affected component/device SP-206 CATEGORY Component group where the difference exists.
Possible entries include:
Annunciators Controllers Demarcation lines Meters Mimics Operator Aids Recorders Status lights Switches DESCRIPTION Brief description of the identified difference.
DATE IDENTIFIED Date the hardware difference was noted.
IMPACT An "SRO" review of the difference between the simulator and the reference plant, and how the difference will "impact" the actions taken by an operator.
Impact assessment as assigned by Simulator Performance Engineer.
All proposed changes to the simulator hardware are evaluated by the Simulator Configuration Engineer for possible software modification to support the change (i.e.
meter scale change, the addition/deletion of a status light, switch, or annunciator).
If a software change is required to support the hardware modification, a work package is developed in accordance with SP-205.
If a software change is not required (i.e. a change in font on a label plate, or the addition of an operator aid) the change is made by the Simulator Support Group as scheduled.
The simulator database is then updated to reflect the modification.
Appendix 1 Page 2 of 2 Date: 07/06/90 Rev. : --=1=---
- I SP-206 HOPE CREEK/SALEM SIMULATOR SIMULATOR vs STATION DIFFERENCES PHOTO NUMBER/VIDEO FRAME:
PANEL:
ITEM NUMBER:
GROUP NUMBER:
COMPONENT:
CATEGORY:
DESCRIPTION:
IMPACT:
DATE IDENTIFIED:
DATE CLEARED:
CATEGORY Possible entries include:
Annunciators Demarcation lines Mimics Controllers Meters Operator Aids Status lights Recorders Switches IMPACT Possible entries include:
N None No confusion or error likely (e.g. change in font)
S Slight May result in minor disorientation, no error likely (e.g. smoked vs. clear bezel, letter spacing, etc)
M Moderate May result in confusion, recoverable operator L
Large error possible (e.g. labeling or scale graduation errors)
May inhibit operator attention/action, non-recoverable operator error possible (e.g.
equipment/control relocation, addition, or removal)
I Appendix 1 Date: 07/06/90 Rev. : --=1 __
Page No.
07/05/90 PHOTO #/
VIDEO FRAME COMPONENT SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL ITEM GROUP PANEL NUMBER NUMBER CATEGORY DESCRIPTION
=======================================================================
DATE IDENTIFIED IMPACT
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2 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PMIEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=================================================================================== ---------- ------
RC 21-3 lCCl A621 1-19 R2 P/B L PUSHBUTTON IS 12/02/89 M
BLANK RC22-3 lCCl A662 1-19 R2 P/B L PUSHBUTTON rs 12/02/89 M
BLANK lCCl A706 1-16 R2 STATION THIS POSITION 12/04/89 M
COVERED WITH BLANK lCCl A807 1-17 RY STATION THIS POSITION 12/04/89 M
COVERED WITH BLANK lCCl A903 1-18 LABEL BOTTOM LABEL READS: 12/04/89 RC13-6 lCCl A903 1-18 LABEL TOP LABEL READS: 11 12/04/89 1C-4KV lCCl A905 1-18 LABELS TOP LABEL READS:
12 12/04/89 s 1C-4KV BOTTOM LABEL READS:
RC13-5 lCCl A909 1-18 LABEL BOTTOM LABEL READS: 12/04/89 s RCll-6 lCCl A909 1-18 LABELS TOP LABEL READS: 15 12/04/89 s 1A-4KV lCCl A911 1-18 LABELS BOTTOM LABEL READS: 12/04/89 s RCll-5 TOP LABEL READS: 16 1A-4KV lCCl A912 1-18 LABELS BOTTOM LABEL READS: 12/04/89 s RC16-5 lCCl A913 1-18 LABEL BOTTOM LABEL READS: 12/04/89 s RC16-7 lCCl A915 1-18 LABEL BOTTOM LABEL READS: 12/04/89 s RC14-7 I
___ j
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07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO *!
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=================================================================================== ---------- ------
lCCl A916 1-18 LABEL BOTTOM LABEL READS:
12/04/89 s RC14-6 lCCl AA04 1-15 R2 P/B LENS A P/B IS BLANK 12/04/89 M
lCCl AA05 1-15 R2 P/B LENS A P /B IS BLANK 12/04/89 M
lCCl AA20 1-15 SCALE SCALE IS 0 - 75 12/04/89 s MULTIPLIER LABEL IS:
GPM XlOO lCCl AA21 1-15 SCALE BOTTOM LABEL READS: 12/04/89 s RCll-3 1CC1 AD03,07 1-2 R2 LENS LENS IS ENGRAVED (A) 11/19/89 s 11,15, AUTO 19 TA-5489 lCCl AD04,08 1-2 RY SCALES SCALE IS 0 - 20 11/19/89 M
12,16, 20 lCCl AD15 1-2 LABEL LOWER LABEL IS:
RC 11/19/89 s 12 - 7 lCCl AD19 1-2 LABEL LOWER LABEL IS:
RC 11/19/89 s 13 - 6 lCCl AD22 1-2 LABEL TOP LABEL IS:
LK 11/19/89 M
DET LEVEL CONT DIFF PRESS lCCl AD22 1-2 RY SCALE RIGHT SCALE IS: -2 11/19/89 M
-1 0
+1
+2 PA-5511 lCCl AFOl 1-3 RECORDER LEFT SCALE SHOULD BE 11/19/89 M
-2 TO 160 PA-5511 lCCl AFOl 1-3 RECORDER LOWER LABEL SHOULD 11/19/89 s BE PR-948 A AND 1-3 (202071)
PA-5512 lCCl AF02 1-4 RECORDER LEFT SCALE SHOULD BE 11/19/89 M
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4 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
-2 TO 160.
LOWER LABELS SHOULD BE PR-9488 1-4 (202071) lCCl AH13(A) 1-7 R2 LENS LENS IS BLANK 11/19/89 M
AH14(A) lCCl AH14 1-7 LABEL BOTTOM LABEL IS:
RC 11/19/89 s 12-6; RC 12-7 lCCl AI05 1-8 RY SCALE LEFT SCALE IS BLANK 11/19/89 M
lCCl AI06 1-8 RY SCALE LEFT SCALE IS 0 - 80 11/19/89 M
lCCl AI07,08 1-8 R2 STATIONS R2 STATION LOCATIONS 11/19/89 M
ARE COVERED WITH BLANK PLATES lCCl AI09 1-8 LABEL LOWER LABEL IS:
11/19/89 s B-VIB; LI-1921 lCCl AilO 1-8 LABEL LOWER LABEL IS:
11/19/89 D-VIB; LI-920 lCCl Aill,12 1-8 RY STATIONS RY STATIONS 11/19/89 M
LOCATIONS ARE COVERED WITH BLANK PLATES lCCl AJOl,06 2-5 LABELS TOP LABELS MISSING 12/04/89 M
(SIMULATOR) lCCl AK06 2-6 LABEL TOP LABEL READS:
12/04/89 1G-460V; 1E-4KV lCCl AL06 2-7 LABEL TOP LABEL READS:
12/04/89 s 1E-460V; 1F-4KV lCCl AM04,05 2-8 LABELS TOP LABELS MISSING 12/04/89 M
(SIMULATOR) lCCl AN03 2-9 LABEL BOTTOM LABEL READS: 12/04/89 s A-VIB lCCl APOl 2-2 ALL LENSES DO NOT 11/19/89 s ADHLPT HAVE SG INLET AS A PREFIX
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07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP.
DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================
~=============~=
LA-5193 lCCl AQ05 2-3 R2 STATION R2 STATION AT THIS 11/19/89 s LOCATION UNDER THE LABEL: 3 STATION AIR COMPRESSOR AND:
1G-4KV lCCl AQ06 2-3 RY INDICATOR RY INDICATOR AT THIS 11/19/89 s LOCATION UNDER THE LABEL OF: STATION AIR AND: HEADER PRESSURE A B lCCl AQ07 2-3 R2 STATION R2 STATION AT THIS 11/19/89 s LOCATION UNDER THE LABEL OF: EMER AIR COMPR AND:
1C-460V lCCl AQ08 2-3 RY STATION RY 2 CHAN. STATION 11/19/89 s AT THIS LOCATION UNDER THE LABEL OF:
CONTROL AIR HEADER PRESSURE B
A lCCl ASOl LABEL TOP LABEL READS:
11/19/89 s GROUP 2 1CC2 GRAPH PRESSURE -
11/19/89 s TEMPERATURE CURVE GRAPH OVER 3 - 6 AFFIXED 1CC2
- LABEL LOWER LABEL OF BORID 11/19/89 s ACID FLOW COUNTERS READS:
FD-5300 1CC2 LABEL LOWER LABEL OF 11/19/89 s PRIMARY WATER FLOW COUNTERS READS:
FD-5301 1CC2 3-17 PLACKARD D/T FEEDS: PLACKARD 12/04/89 s IN PLACE NEXT TO ITEM B221 1CC2 3-18 PLACKARD TAVE FEEDS: PLACKARD 12/04/89 s
I Page No.
7 07 /05/90 SALEM SIMULATOR I
STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
IN PLACE NEXT TO ITEM B307 1CC2 B144,45 3-20 P/B ASSEMBLIES THESE ITEMS ARE NOT 12/04/89 M
54,55, ENGRAVED THEY ARE 56 BLANK lCC2 B201 3-14 LABELS TOP LABEL READS:
12/04/89 s 1PR6, lPRl 1CC2 B202 3-14 LABEL TOP LABEL READS:
12/04/89 s 1PR7, 1PR2 lCC2 B412 4-11 R2 P/B LENS P PUSHBUTTON LENS 12/04/89 s READS: /\\ FLUS CHANNEL 1 1CC2 B413 4-11 R2 P/B LENS P PUSHBUTTON LENS 12/04/89 s READS: /\\ FLUX CHANNEL III 1CC2 B415 Hl R2 P/B LENS P PUSHBUTTON LENS 12/04/89 s READS: /\\ FLUX CHANNEL II 1CC2 B416 4-11 R2 P/B LENS P PUSHBUTTON LENS 12/04/89 s READS: /\\ FLUX CHANNEL IV 1CC2 B426 4-14 RY SCALE LEFT SCALE IS 0 -
12/04/89 s 30; RIGHT SCALE IS 0
- 20 1CC2 B522 4-15 LABEL TOP LABEL:
12/06/89 s EMERGENCY 250V IS BLACK ON WHITE lCC2 B526 4-15 LABEL BOTTOM LABEL READS: 12/06/89 s D-VIB: SA-2270; F1500K 1CC2 B526 4-15 RY SCALES BOTH RIGHT AND LEFT 12/06/89 M
SCALES ARE 0 - 80 1CC2 B527 4-15 RY SCALE RIGHT SCALE IS 0 -
12/06/89 M
10
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07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=================================================================================== ---------- ------
1CC2 B531 4-20 LABEL TOP LABEL READS:
SG 12/06/89 s INLET PRESS; FW SG D/P DIFF PRESS 1CC2 B531 4-20 LABEL BOTTOM LABEL READS: 12/06/89 s D-VIB; P1508, PA4932 lCC2 B531 4-20 RY SCALES RIGHT INDICATOR & 12/06/89 M
SCALE EXISTS AND IS 0 - 24 1CC2 B532 4-20 LABELS BOTTOM LABEL READS: 12/06/89 D-VIB; P1509, PA5099 1CC2 B541 4-18 RY SCALE LEFT SCALE IS 0 - 10 12/06/89 M
1CC2 B546 4-18 RY SCALE LEFT SCALE IS 0 -32 12/06/89 M
1CC2 B549 4-18 RY SCALE RIGHT SCALE DOES NOT 12/06/89 M
EXIST 1CC2 B607 4-13 RY SCALE LEFT SCALE IS 0 - 60 12/06/89 M
1CC2 B614 4-16 LABEL TOP LABEL: EMERGENCY 12/06/89 s 250 V IS BLACK ON WHITE 1CC2 B618 4-16 LABEL BOTTOM LABEL READS: 12/06/89 s D-VIB; SA5106, F1500M 1CC2 B618 4-16 SCALES LEFT AND RIGHT 12/06/89 M
SCALES ARE 0 - 80 1CC2 B619 4-16 LABEL BOTTOM LABEL READS: 12/06/89 22 MAC; PA5108 I FA598 1CC2 B619 4-16 RY SCALE RIGHT SCALE IS 0 -
12/06/89 M
10 SHUTDOWN COUN-1CC2 BBOl -
3-6 LABELS ALL LABELS ON THE 11/19/89 N
TERS BB06 COUNTERS ARE BLACK ON WHITE.
DOOR
Page No.
10 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
1CC3 6-11 LABEL TOP LABEL OVER ITEMS 12/14/89 s C301 AND C302 READS:
lA AUX PWR XFMR 1CC3 6-16 LABEL TOP MAIN LABEL 12/15/89 READS:
13 KV RING BUS 1CC3 6-4 LABEL TOP LABEL IS:
11/19/89 s STATION POWER lCC3 6-5 LABEL TOP LABEL SAYS:
11/19/89 s GRID SYNCHRONIZATION 1CC3 6-8 MIMIC RED MIMIC HAS 11/19/89 s DIFFERENCES 1CC3 6-1 PLACKARD RESARANT SPEEDS 12/14/89 M
PLACKARD IS MISSING (AT SIMULATOR) 1CC3 PLACKARD LOAD DISPATCHER 11/19/89 s VOLTAGE REQUIREMENTS PLACKARD IS AFFIXED NEXT TO SUPERVISORY LIGHTS 1CC3 6-10 R2 LENSES WORD (DIESEL GEN) IS 12/14/89 REPLACED BY (lC) 1CC3 C103 6-1 LENS LENS IS CLAR BLANK 12/14/89 M
1A5850 1CC3 C204 6-10 INDICATOR SCALE RIGHT SCALE IS: 0 - 12/14/89 M
6 1CC3 C301 6-11 R2 LENS REFERENCE MADE TO lA 12/15/89 ON P/B's H, L, P, T (3RD THRU LAST PUSH BUTTONS) lCC3 C303 6-11 R2 LENSES LAST TWO LENSES 12/15/89 s (P,T) MAKE REFERENCE TO lA:
lA HGD CLOSE 7A HGD OPEN
r Page No.
11 07 /05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
1CC3 C305 6-11 LABEL TOP LABEL READS:
12/15/89 s 1H5D TO 460 & 230 1CC3 C307 6-11 R2 LENS LENS IS WHITE BLANK 12/15/89 M
THIRD FROM TOP (H) 1CC3 C308 6-11 R2 LENSES BOTTOM 2 P/B's MAKE 12/15/89 s REFERENCE TO lA:
lAEGD CLOSED; lAEGD OPEN 1CC3 C310 6-11 R2 LENS 5TH LENS FROM TOP 12/15/89 s (P) READS:
lEGD CLOSE 1CC3 C310 6-11 R2 LENS 6TH LENS FROM TOP 12/15/89 s (T) READS:
lEGD OPEN 1CC3 C312 6-11 R2 LENS 3RD LENS FROM TOP 12/15/89 M
(H) IS WHITE BLANK 1CC3 C501 6-15 LABEL TOP LABEL READS: 2B 12/15/89 s C502 AUX PWER XFMR 1CC3 C503 6-15 R2 LENS 5TH AND 6TH LENS 12/15/89 FROM TOP (P, T)
READ:
lBFGD CLOSE; lBFGD OPEN 1CC3 C505 6-15 LABEL TOP LABEL READS:
12/15/89 s 1F5D TO 460 & 230 1CC3 C508 6-15 R2 LENS LENS H IS WHITE 12/15/89 M
BLANK 1CC3 C509 6-15 R2 LENS LENS P, T READ lBGGD 12/14/89 s CLOSE; lBGGD OPEN 1CC3 C510 6-15 R2 LENS LENSES P & T READ:
12/15/89 s lGGD CLOSE; lGGD OPEN lCC3 C512 6-15 R2 LENS LENS H IS WHITE 12/15/89 M
BLANK
Page No.
13 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
LENS P READS:
2-8 CLOSE.
LENS T READS:
2-8 OPEN 1CC3 C606,07 6-17 LABEL TOP LABELS OVER C606 12/15/89 s
& C607 READ:
500 KV GROUP BUS BREAKER SECTIONS 2-8 21X; 1-8 20X 1CC3 C607 6-17 LABEL BOTTOM LABEL READS 12/15/89 s RC 11-4 1CC3 C607 6-17 R2 LENS LENS A READS:
1-B 12/15/89 s SYNCH POT ON.
LENS D READS:
1-B SYNCH POT OFF.
LENS H READS:
H REMOTE.
LENS L READS: 1-8 LOCAL MIMIC BUS (UNINTELIGIBLE).
LENS P READS:
1-8 CLOSE.
LENS T READS:
1-8 OPEN.
1CC3 C708,09 P/B SWITCHES ITEM C708 READS:
12/14/89 M
LOAD CHAN ITEM C709 READS LOAD REF CHAN 1CC3 C713 -
6-9 LABEL THESE ITEMS ARE 12/14/89 M
C716 LABELED AS FOLLOWS:
llLV I 12LL, 13RL, 14RV 1CC3 C802 6-18 R2 LENS LENS A IS BLACK ON 12/15/89 s YELLOW 1CC3 C806 6-18 R2 LENS LENS A IS BLACK ON 12/15/89 s YELLOW 1CC3 C810 6-18 R2 LENS LENS A IS BLACK ON 12/15/89 s YELLOW 1CC3 C901 6-19 R2 LENS LENS P READS:
12/15/89 s EXCITER FIELD BKR
Page No.
14 D7/D5/9D SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
CLOSE 1CC3 C903 6-19 LABEL TOP LABEL READS:
12/15/89 M
MANUAL ADJUSTER 1CC3 C903 6-19 R2 LENS LENS P READS:
RAISE 12/15/89 M
OUTPUT VOLTS 1CC3 C904 6-19 LABEL BOTTOM LABEL READS: 12/15/89 s 11-MAC; VA-5115, NA-5116 1CC3 C904 6-19 LABEL TOP LABEL READS:
12/15/89 M
MANUAL ADJUSTER; VOLTS POSITION 1CC3 C905 6-19 LABEL TOP LABEL READS:
12/15/89 M
REGULATOR I
- 1CC3 C905 6-19 R2 LENS LENS P READS:
12/15/89 M
REGULATOR AUTO.
LENS T READS:
REGULATOR MANUAL 1CC3 CA02 6-20 LABEL TOP LABEL READS:
12/15/89 s 2-6 llX 1CC3 CA02 6-20 LABEL BOTTOM LABEL READS: 12/15/89 s RC 13-4 1CC3 CA02 6-20 R2 LENS ALL LENSES MAKE 12/15/89 s REFERENCE TO 2-6 1CC3 CA03 6-20 R2 LENS ALL LENSES MAKE 12/15/89 s REFERENCE TO 5-6 1CC3 CA03, 04 6-20 LABELS TOP LABELS READ:
12/15/89 s CA03 5-6 lOX; CA04 1-5 12X 1CC3 CA04 6-20 LABEL BOTTOM LABEL READS: 12/15/89 s RC 11-4 1CC3 CA04 6-20 R2 LENS ALL LENSES MAKE 12/15/89 s REFERENCE TO 1-5
Page No.
15 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
1CC3 CB03 LABEL LABEL ON TOP OF CB03 11/19/89 N
SAYS:
GROUP 6 1CC3 CB04,05 6-2 R2 LENS ALL REFERENCES TO 11/19/89 s DIESEL GENERATOR ON DG IS REPLACED BY lA 1CC3 CB18 6-4 LABEL TOP LABEL IS:
PHASE 11/19/89 s SELECT 1CC3 CDOl,02 6-7 R2 LENS ALL REFERENCE TO 11/19/89 s DIESEL GENERATOR ON DG IS REPLACED BY lA 1CC3 CI05 5-6 LABEL TOP LABEL:
12/14/89 s EMERGENCY IN WHITE ON RED 1CC3 CI06 5-6 LABEL TOP LABEL READS:
12/14/89 s BACKUP LIFT 1H-230V 1CC3 CI07 5-6 LABEL TOP LABEL READS:
12/14/89 s LIFT 1F-230V 1CC3 CKOl 5-1 LABEL TOP LABEL READS:
11/19/89 N
- 11AR25, 1AR19, 11AR14 BOTTOM LABEL READS:
RC 14-4, RC 14-5 1CC3 CKOl -
LABELS & R2 CK02 TOP LABEL: 11 11/19/89 N
CK03 LENS 11AR37 I 1H-460V BOTTOM LABEL RC 14-4 CK03 TOP LABEL:
12 12AR25, 1H-460V BOTTOM LABEL: RC 14-5 CK04 TOP LABEL: 13 13AR25, lF-460 BOTTOM LABEL: RC 16-4, RC 16-5 CK05 TOP LABEL: 14 14AR25, 1F-460V
Page No.
16 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
BOTTOM LABEL: RC 16-5 CK06 BOTTOM LABEL:
RC 17-3, RC 18-6 R2 LENS ENGRAVINGS REFERENCE TOP LABEL NUMBERS lCC3 CK11 5-1 LABELS TOP LABEL:
21A 11/19/89 s
- LEVEL, 218 LEVEL 1CC3 CK12 5-1 LABELS TOP LABEL:
13 LEVEL 11/19/89 s BOTTOM LABEL:
11 MAC; LA-228 1CC3 CK12 5-1 RY IND.
SINGLE CHANNEL RY 11/19/89 M
INDICATOR ONLY RIGHT SIDE IS BLANK 1CC3 CK13,14 5-1 RY IND,R2 P/B INSTRUMENTS AT THESE 11/19/89 M
SWITCH LOCATIONS DO NOT EXIST 1CC3 CNOl P/B SWITCH DOES NOT EXIST AT 11/19/89 M
SIMILAR TO SIMULATOR LOCATION 1CC3 CP08 (D) 6-22 R2 LENS R2 LENS ENGRAVED:
11/19/89 s CONDENSER VACUUM LO lRPl LAYOUT, ARRANGEMENT 11/19/89 L
AND TYPE OF INSTRUMENTS ON lRPl IS DIFFERENTTHAN THAT OF SALEM 2.
REFER TO PLANT DRAWING OF THIS PANEL.
1RP2 KEY SWITCH CONTAINMENT 11/27 /89 s EVACUATION HORN BLOCK OPERATE KEY SWITCH AND LABEL ABOV~ EXIST BELOW
I I
Page No.
17 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=================================================================================== ---------- ------
ITEM 2406 KA-3278 1RP2 POINT PLACKARD TEMPERATURE FOR ALL 11/19I8 9 M 4 POINTS IS 25xC
[250xc ON SIMULATOR)
XA-8493 1RP2 RECORDER THIS RECORDER EXISTS 11/19/89 M
ON 1RP2 KA-3277 1RP2 RECORDERS I SCALE IS 0 - 1.0 11/19/89 s SCALES POINT DESCRIPTION PLACKARD IS DIFFERENT 1RP2 2113,14 LABEL LABEL OVER BOTH 11/19/89 s SWITCHES READS:
PIPE GALLERY CIRCULATING FANS 1RP2 2114 CMC SW WINDOW CMC 2 POSITION 11/19/89 s ENGRAV MAINTAIN:
STOP I START, NO. 12 PIPE GALLERY CIRCULATING FAN TOP LABEL:
12 1RP2 2116,17 CMC SW WINDOW CMC 2 POSITION SRC 11/19/89 s ENGRAV SWITCH:
AUTO I START STOP - GREEN 1RP2 21B CMC SW WINDOW CMC 2 POSITION 11/19/89 s ENGRAV MAINTAIN:
STOP I START, NO. 11 PIPE GALLERY CIRCULATING FAN.
TOP LABELl 77 1RP2 2203,04 CMC P/B SWITCHES ARE 2 11/27 /89 s SWITCHES POSITION MAINTAIN 1RP2 2210 CMC SWITHCES SWITCH IS 2 POSITION 11/19/89 s SRC WITH 4 WINDOWS 1RP2 2211 CMC SWITCHES SWITCH IS 2 POSITION 11/27/89 MAINTAIN WITH 4 WINDOWS
I I
I I
Page No.
18 I
I 07/05/90 I
SALEM SIMULATOR I
STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 I
I
- SORTED BY PANEL I
I I
I PHOTO #/
ITEM GROUP DATE I
VIDEO FRAME COMPONENT I
PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT I
=================================================================================== ---------- ------
I I
I 1RP2 2221 LABEL TOP LABEL READS:
11/27 /89 s I
I 1CH30 I
I 1RP2 2222 LABEL TOP LABEL READS:
11/27 /89 I
1CH151 1RP2 2224 CMC SWITCH SWITCH IS 2 POSITION 11/27/89 SRC WITH 4 WINDOWS 1RP2 2302 LABEL RIGHT HAND LABEL 11/27 /89 s READS:
DISCHARGE LOUVER 1RP2 2317 LABEL RIGHT HAND LABEL 11/27/89 s READS:
OUTSIDE AIR INTAKE 1RP2 2407 -
SWITCHES &
THE ORDER IN WHICH 11/29/89 M
2410 LABELS THESE ITEMS ARE INSTALLED IS FROM LEFT TO RIGHT:
- 2RC40, 2RC43,
- 2RC41, 2RC42 HA-3280,KA-3277 1RP2 2501 -
RECORDERS 11/19/89 s TO KA-3279 2504 KA-3279 1RP2 2504 POINT PLACKARD POINT DESCRIPTION 11/19/89 s PLACKARD IS DIFFERENT 1RP3 KEYBOARD SPDS KEYBOARD IS 11/29/89 N
DRAWER TYPE 1RP3 LABEL LABEL BELOW (REACTOR 11/29/89 s COREi LABEL READS:
270x GAMMAMETRICS CHANN EL C XA6560 1RP3 LABEL LABEL BELOW REACTOR 11/29/89 s CORE READS: 90x GAMMAMETRICS CHANNEL D XA6561 1RP3 3505 -
SWITCHES ORDER OF SWITCHES 11/29/89 M
3507 FROM LEFT TO RIGHT
Page No.
19 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
IS:
11 13 1RP3 3605 CORE MAP 9A IS BLANK 11/29/89 N
INDICATORS 1RP3 3607 CORE MAP llA IS BLANK 11/29/89 N
INDICATORS 1RP3 3609 CORE MAP 4B BLANK 11/29/89 N
INDICATORS 1RP3 3611 CORE MAP 6B READS:
T39; LIFT 11/29/89 N
INDICATORS COIL DISC - 2Dl 1RP3 3613 CORE MAP 8B READS:
T35; LIFT 11/19/89 N
INDICATORS COIL DISC - 1SC2 1RP3 3615 CORE MAP lOB READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 102 1RP3 3617 CORE MAP 12B IS BLANK 11/29/89 N
INDICATORS 1RP3 3620 CORE MAP 3C READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - lCl 1RP3 3622 CORE MAP 5C READS:
T33; LIFT 11/29/89 N
INDICATORS COIL DISC - 2Al 1RP3 3624 CORE MAP 7C READS: T28; LIFT 11729/89 N
INDICATORS COIL DISC - 2SB1 1RP3 3628 CORE MAP llC READS:
LIFT 11/19/89 N
INDICATORS COIL DISC - 1A2 1RP3 3630 CORE MAP 13C READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 1C2 1RP3 3632 CORE MAP 2D BLANK 11/29/89 N
INDICATORS 1RP3 3634 CORE MAP 4D READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - lSAl 1RP3 3636 CORE MAP 6D READS:
LIFT COIL 11/29/89 N
I Page No.
20 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #I ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
INDICATORS DISC 1RP3 3638 CORE MAP 8D READS:
T15; LIFT 11/29/89 N
INDICATORS COIL DISC - lBl 1RP3 3640 CORE MAP lOD READS:
T23; 11/29/89 N
INDICATORS LIFT COIL DISC TO_VID 1RP3 3642 CORE MAP 12D READS:
T31; 11/29/89 N
IND I CATO.RS LIFT COIL DISC -
1SA2 1RP3 3644 CORE MAP 14D READS:
T49 11/29/89 N
INDICATORS 1RP3 3647 CORE MAP 3E READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - lSCl 1RP3 3653 CORE MAP 9E READS:
T12; LIFT 11/29/89 N
INDICATORS COIL DISC - 2SA2 1RP3 3657 CORE MAP 13E READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2A2 1RP3 3665 CORE MAP 6F READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - lSDl 1RP3 3667 CORE MAP 8F READS:
T5; LIFT 11/29/89 N
INDICATORS COIL DISC - 2D2 1RP3 3669 CORE MAP lOF READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - lSD2 1RP3 367 3 CORE MAP 14F READS:
T40; 11/29/89 N
INDICATORS LIFT COIL DISC - 2D2 1RP3 3679 CORE MAP 5G READS: Tll; LIFT 11/29/89 N
INDICATORS COIL DISC - 2SA1 1RP3 3687 CORE MAP 13G READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2SB2 1RP3 3691 CORE MAP 2H READS:
T34; LIFT 11/29/89 N
INDICATORS COIL DISC - lSCl
---~---------------------------------~~---------------
Page No.
21 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
==============~========================================================================
1RP3 3693 CORE MAP 4H READS:
T14; LIFT 11/29/89 N
INDICATORS COIL DISC - 2Bl 1RP3 3695 CORE MAP 6H READS:
T4; LIFT 11/29/89 N
INDICATORS COIL DISC - 2Cl 1RP3 3699 CORE MAP lOH READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2C3 1RP3 3702 CORE MAP 12H READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2B2 1RP3 3704 CORE MAP 14H READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 1SC3 1RP3 3708 CORE MAP 3J READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 2SB4 1RP3 3716 CORE MAP llJ READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2SA3 1RP3 37 22 CORE MAP 2K READS:
T38; LIFT 11/29/89 N
INDICATORS COIL DISC 2D4 1RP3 3726 CORE MAP 6K READS:
TB; LIFT 11/29/89 N
INDICATORS COIL DISC - 1SD4 1RP3 3728 CORE MAP BK READS:
LIFT COIL 11/29/89 N
INDICATORS
- DISC - 2C4 1RP3 37 30 CORE MAP lOK READS:
LIFT 11/29/ 89 N
INDICATORS COIL DISC - 2BC 1RP3 3732 CORE MAP 12K READS:
T22; 11/29/89 N
INDICATORS LIFT COIL DISC 1RP3 3738 CORE MAP 3L READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 2A4 1RP3 3742 CORE MAP 7L READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 2SA4 1RP3 3748 CORE MAP 13L READS:
T32; 11/29/89 N
INDICATORS LIFT COIL DISC - 1A3 1RP3 3751 CORE MAP 2M:
BLANK 11/29/89 N
Page No.
22 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
INDICATORS 1RP3 37 53 CORE MAP 4M READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 1SA4 1RP3 37 57 CORE MAP 8M READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 1B2 1RP3 3761 CORE MAP 12M READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 1SA3 1RP3 3763 CORE MAP 14M:
BLANK 11/29/89 N
INDICATORS 1RP3 3765 CORE MAP 3N READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 1C4 1RP3 3767 CORE MAP 5N READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 1A4 1RP3 3771 CORE MAP 9N READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 2SB3 1RP3 3773 CORE MAP llN READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2A3 1RP3
- 3775 CORE MAP 13N READS:
T45; 11/29/89 N
INDICATORS LIFT COIL DISC - 1C3 1RP3 3778 CORE MAP 4P READS:
T48 11/29/89 N
INDICATORS 1RP3 3782 CORE MAP 8P READS:
LIFT COIL 11/29/89 N
INDICATORS DISC - 1SC4 1RP3 37 84 CORE MAP lOP READS:
LIFT 11/29/89 N
INDICATORS COIL DISC - 2D3 1RP3 3786 CORE MAP 12P:
BLANK 11/29/89 N
INDICATORS llSJ 44 1RP4 CMC SWITCH EXIST ON THE LEFT OF 11/29/89 s ITEN 4D04 12SJ44 1RP4 CMC SWITCH EXISTS ON THE LEFT 11/29/89 s OF ITEM 4D07
Page No.
23 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
LA-3977 1RP4 LABEL LABEL MISSING 11/29/89 M
LA-3977 LA-3978 1RP4 LABEL 11/29/89 1RP4 P/B SWITCH 6 R2 P/B SWITCH 11/29/89 s EXIST DIRECTLY UNDER ITEMS 4C01, 4C02, 4C03, 4C04, 4C05, 4C06 LA-8639 1RP4 RECORDER RECORDER EXISTS 11/29/89 s UNDER THE LABEL OF:
TIDE LEVEL 1RP4 4611 P/B SWITCH LENS READS:
15 11/29/89 N
SERVICE WATER PUMP 1RP4 4612 P/B SWITCH LENS READS:
16 11/29/89 N
SERVICE WATER PUMP 1RP4 4631 P/B SWITCH LENS READS:
25 11/29/89 N
SERVICE WATER PUMP 1RP4 4633 P/B SWITCH LENS READS:
25 11/29/89 N
SERVICE WATER PUMP 1RP4 4706 P/B SWITCH LENS READS:
15 11/29/89 N
SERVICE WATER PUMP 1RP4 47 07 P/B SWITCH LENS READS:
16 11/29/89 N
SERVICE WATER PUMP 1RP4 4728 P/B SWITCH LENS READS:
11 11/29/89 N
SERVICE WATER PUMP 1RP4 4729 P/B SWITCH LENS READS:
12 11/29/89 N
SERVICE WATER PUMP 1RP4 4H01 CMC SWITCH LENS READS:
OFF I 11/29/89 ON - POWER SUPPLY A LOCA SAMPL ISOL VAS 1RP4 4H18 CMC SWITCH LENS READS:
OFF I 11/29/89 s ON - POWER SUPPLY C
=
Z DC n
rzcz De E
Page No.
24 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #I ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
LOCA SAMPL ISOL VAS 1RP5 CMC SWITCH THESE 2 POSITIONS 11/29/89 M
MAINTAIN CMC SWITCHES UNDER LABELS OF: A, B, C, NORMAL I EMERG -
POWER AVAIL AND UNDER LARGE LABEL:
DIESEL GENERATOR EXHAUST FANS EMERG BYPASS OF C02 SHUTDOWN ARE MISSING ON 1RP5 XA-3361 1RP5 5500 RECORDERS XA-3361 CONTAINMENT 11/29/89 s HYDROGEN CONCENTRATION RECORDER IS A TIGRAPH RECORDER XA-3362 1RP5 5600 RECORDER XA-3362 CONTAINMENT 11/29/89 s HYDROGEN CONCENTRATION RECORDER IS A TIGRAPH RECORDER 1RP5 5701, 02 LABEL LABEL ABOVE THESE 11/29/89 s INDICATORS READ:
STEAM GENERATOR METAL TEMPERATURE 1RP5 5705,06 LABEL LABEL ABOVE THESE 11/29/89 s SELECTORS READ:
STEAM GENERATOR TEMPERATURE SELECT 1RP5 5705,06 LABELS POSITION LABELS 11/29/89 s ABOVE THESE SELECTORS ARE SWAPPED AT THE STATION 1RP5 5901 LABEL LABEL ON THE LOWER 11/29/89 s MID SECTION OF RECORDER XA-6846
l Page No.
25 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
1RP5 5902 LABEL LABEL ON THE LOWER 11/29/89 s AID SECTION OF RECODER XA-8550 1RP5 5903 LABEL LABEL ON THE LOWER 11/29/89 s MID SECTION OF RECORDER XA-8551 1RP6 ANALOG 4" ROUND INDICATORS 12/02/89 s INDICATORS HAVE WHITE FRAME 1RP6 LAMP INDICATORS BREAKER FAILURE 12/02/89 MULTI TRIP LAMP INDICATORS ARE SITUATED AS FOLLOWS:
INDICATOR LABELED SECTION 1-2 MT 86/3-4 IS DIRECTLY BELOW ITEM 6219 INDICATOR LABELED SECTION 2-3 MT 86/2-3 IS DIRECTLY BELOW ITEM 6220 INDICATOR LABELED SECTION 4-5 MT 86/4-5 IS DIRECTLY BELOW ITEM 6322 1RP6 MIMIC NO MIMIC LINE 12/02/89 SEPARATING 11 AND 12 POWER TRANSFORMER BLACK AND 13KV BUS BLACK 1RP6 PANEL SECTION (BREAKER 21X/LINE 12/02/89 s 5021) THIS SECTION DOES NOT EXIST AT THE SIMULATOR 1RP6 PANEL SECTION (BREAKER llX/LINE 12/02/89 s 5024) THIS SECTION IS VERY SIMILAR TO:
BREAKER 31X/LINE 5037 AT THE SIMULATOR.
THE
Page No.
26 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
1RP6 1RP6 1RP6 DIFFERENCES ARE:
- 1.
MAIN TOP LABEL
- 2.
LABEL:
NEW FREEDOM LINE
- 3.
MEGAWATTS INDICATOR SCALE IS 4000-0-4000
- 4.
LABEL TO LAMP INDICATOR NEXT TO VOLTMETER SELECTOR SWITCH NEEDS CARRIER RELAYS
- 5.
VOLTMETER SELECTOR SWITCH HAS LABEL:
5024 VOLTMETER
- 6.
TOGGLE SWITCH UNDER THE LABEL:
500KV SECT 2-6 BKR REC LOSING PANEL SECTION 10 - ITEM 6430 HAS A 12/02/89 S
LABEL REMOTE TRIP PANEL SECTION 3 UNIT GAS TURBINE 12/02/89 M
GENERATOR DOES EXIST AT SALEM 1 BUT WAS REMOVED FROM SIMULAOR I 1988 PANEL SWITCH ALSO LARGER LABEL ON 12/02/~9 S THE RIGHT OF SWITCH:
BS 2-6 BREAKER REC LOSING
- 7.
UNDER BUS SECTION AMPERES LABEL, THE LEFT SMALL LABEL READS:
BS 2-6; CENTER SMALL LABEL READS:
BS 5-6; RIGHT SMALL LABEL READS BS 1-5.
LABELS PHASE 1, PHASE 2, PHASE 3 ARE ON THE RIGHT SIDE OF THE
I Page No.
27 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
INDICATORS
- 8.
ITEMS 6427, 6429 HAVE A LABEL:
BS 2-10 BREAKER
- 9.
ITEM 6428 HAS A LABEL:
BACKUP RELAYS 1RP7 METERS TWO KW INPUT DEMAND 11/19/89 s METERS INSTALLED BETWEEN XA-0883 AND GENERATOR STATION WATER CONDUCTIVITY 1RP7 7101 LABEL TOP LABEL IS:
11/19/89 s KILOVOLTS BOTTOM LABEL IS:
XA-3975 1RP7 7101,02 LABELS LARGE LABEL COVERING 11/19/89 BOTH RECORDERS IS:
GENERATOR 1RP7 7102 LABEL TOP LABEL IS:
11/19/89 KILOAMPERES BOTTOM LABEL IS:
XA-3976 XA-0884,SA-0886 1RP7 7103,04 RECORDERS RECORDERS ARE 11/19/89 N
DIFFERENT MAKE
[ESTERLINE ANGUS)
XA-0883 1RP7 7105 LABEL TOP LABEL IS: LARGE 11/19/89 s AND BOLD.
BOTTOM LABEL IS:
XA-0883 W552 TO W560 1RP7 7200 -
THESE INSTRUMENTS 11/19/89 s 7900 HAVE BEEN CROSSED 7AOO OUT BY RED TAPE TA-0885 1RP7 7B01 LABEL BOTTOM LABEL IS:
11/19/89 s TA-0885 1RP8 LABEL LABEL OVER GREEN 11/19/89 s PUSHBUTTON IS:
Page No.
07/05/90 PHOTO #/
28 VIDEO FRAME COMPONENT SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL ITEM GROUP PANEL NUMBER NUMBER CATEGORY DESCRIPTION DATE IDENTIFIED IMPACT
=================================================================================== ---------- ------
OVERPRESSURIZATION U2027 :RCL SYSTEM PESS. 1/2 AVG" 1RP8 P/B ENGRAVING INCORE T/C TREND AND 11/19/89 N
TW TURB TREND ARE RED IN COLOR 1RP8 8177 -
P/B COVERS THESE P/B's ARE 11/19/89 N
8179 BLANK YELLOW 1RP8 8818 -
P/B COVERS THESE P/B's ARE 11/19/89 N
8823 YELLOW 1RP9 9102,03 LABEL LABEL OVER THESE 11/19/89 s P/B's IS:
CONTROL ROOM DOOR ALL PANELS ALL CNTRL RM ALL PLANT EQUIPMENT 11/19/89 s INSTRUMENTS REFERENCE NUMBERS BEGIN WITH 1 (AT SALEM lSTATION)
OHA 10 K
READS:
500 KV BS 11/19/89 s 5-6 BKR lOX TRBL OHA 11 K
READS:
500KV BS 1-8 11/19/89 s BKR 20X TRBL OHA 12 B
READS:
RMS PROCESS 11(19/89 s RAD HI OHA 12 D
READS:
RH 1 NOT 11/19/89 s FULL CLS & RX PRESS HIGH OHA 12 K
READS:
500KV LN 21X 11/19/89 s CARR CHK TRBL OHA 13 B
READS:
FRESHWTR SYS 11/19/89 s PRESS LO OHA 13 K
READS:
500KV LN llX 11/19/89 s REG RELAY PWR FAIL OHA 14 H
READS:
22 STA XFMR 11/19/89 s PROT BU
Page No.
29 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
OHA 14 J
READS:
13 KV BUS 11/19/89 s SECT 1 GRND FAULT OHA 14 K
READS:
500KV LN 21X 11/19/89 s REG RELAY PWR FAIL OHA 15 c
CROSSED OUT BY TAPE 11/19/89 s OHA 15 J
READS:
22 STA PWR 11/19/89 s XFMR BKR FAIL OHA 16 J
CROSSED OUT BY TAPE 11/19/89 s OHA 17 D
READS:
NON RAD LIQ 11/19/89 s WSTE DISP TRBL OHA 18 K
READS:
500KV BS 2-6 11/19/89 s BKR llX GRND / FAIL OHA 19 B
READS:
RMS CH. TEST 11/19/89 s OHA 19 K
READS:
500 KV BS 11/19/89 s 2-8 BKR 21X GRND I FAIL OHA 2
K READS:
500 KV BS 11/19/89 s 5-6 BKR lOX GRND I FAIL OHA 20 B
READS:
RMS PROCESS 11/19 / 8 9 s FLTR RADI OHA 20 D
READS:
RH2 NOT FULL 11/19/89 s CLS & RX PRESS HI OHA 20 K
READS:
500KV BS 11/19/89 s 2-10 BKR 31X GRND I FAIL OHA 21 B
READS:
FRESHWTR 11/19/89 s STOR TK 1/2 LVL LO OHA 21 K
READS:
500KV LN llX 11/19/89 s REG CARR TRBL
Page No.
30 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
OHA 22 A
READS:
FIRE PROT 11/19/89 s WTR PRESS LO OHA 22 H
READS:
BLANK 11/19/89 s OHA 22 K
READS:
500KV LN 21X 11/19/89 REG CARR TRBL OHA 23 J
BLANK 11/19/89 s OHA 24 D
READS:
ROD DRIVE 11/19/89 s FAN VIB HI ALERT OHA 24 E
READS:
1/3 CW BRG 11/19/89 LUBE PRESS LO/PMP TRIP OHA 24 K
READS:
POPS CH. 1 11/19/89 s AUX AIR PRESS LO OHA 25 A
ANN.
CROSSED OUT BY TAPE 11/19/89 N
OHA 26 D
READS:
VTL HT TRACE 11/19/89 TRBL OHA 26 K
READS:
500 KV BS 11/19/89 s 2-6 BKR llX TRBL OHA 27 H
READS:
AUX XFMR OIL 11/19/89 s LVL LO/INERT AIR TRBL OHA 27 K
READS:
50 KV BS 2-8 11/19/89 s BKR 21X TRBL OHA 28 K
READS:
CNTMT VENT 11/19/89 s ISOL RESET W/ACT SIGNL OHA 29 B
READS:
FRESHWTR 11/19/89 s PRE-TREAT TRBL OHA 29 K
READS:
500KV LN llX 11/19/89 s BU CARR TRBL OHA 30 A
READS:
FIRE PMP 1/2 11/19/89 s
Page No.
31 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
RUN OHA 30 G
READS:
C02 STOR 11/19/89 s PRESS HI/LO OHA 30 K
READS:
500KV LN 21X 11/19/89 s BU CARR TRBL OHA 31 H
READS:
ISTA XFMR 11/19/89 s CLG TRBL OHA 32 H
READS:
COND POL 11/19/89 s IMMED BYP OHA 32 J
READS:
13KV BUS 11/19/89 s SECT 4 GRND FAULT OHA 33 H
READS:
MN XFMR 02 11/19/89 s OIL LVL LO/MISC TRBL OHA 34 H
READS:
SMOKE IN GEN 11/19/89 s OHA 34 K
READS:
500 KV BS 11/19/89 s 1-5 BKR 12X GRND I FAIL OHA 35 A
CROSSED OUT BY TAPE 11/19/89 N
OHA 35 D
READS:
RWST LVL 11/19/89 s HI/LO OHA 35 K
READS:
BS 1-5/5-6 11/19/89 s 500KV BKR FLASHOVER OHA 36 A
READS:
ESS CNTRL & 11/19/89 s EMER LTG INVRT FAIL OHA 36 D
READS I HTG WTR STM 11/19/89 s SYS TRBL OHA 36 D
READS:
RWST LVL LO 11/19/89 s BU OHA 36 K
READS:
500KV BS 1-9 11/19/89 s BKR 32X GRND I FAIL
Page No.
32 07/05/90 SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL PHOTO #/
ITEM GROUP DATE VIDEO FRAME COMPONENT PANEL NUMBER NUMBER CATEGORY DESCRIPTION IDENTIFIED IMPACT
=======================================================================================
OHA 37 K
READS:
500KV LN llX 11/19/89 s RECVR REN TRIP OHA 38 A
READS:
FIRE PMP 1/2 11/19/89 s TRBL OHA 38 K
READS:
500 KV LN 11/19/89 s 21X RECVR REN TRIP OHA 39 D
READS:
CH. C 11/19/89 s DECREASING SHUTDN MARGIN OHA 39 F
CROSSED OUT BY TAPE 11/19/89 s OHA 39 H
READS:
ISTA XFMR 11/19/89 s OIL LVL LO/TEMP HI OHA B
READS:
FRESHWTR SYS 11/19/89 s TRBL OHA K
READS:
500KV LN llX 11/19/89 CARR CHK TRBL OHA 40 40 READS:
GAS TURB 11/19/89 s TRBL OHA 40 H
READS:
COND POL 11/19/89 s WSTE RECOV TRBL OHA 41 c
CROSSED OUT BY TAPE 11/19/89 s OHA 41 D
READS:
BIT DISCH 11/19/89 PRESS HI OHA 42 H
READS:
MN XFMR 11/19/89 s HARMONIC OVERCURRENT OHA 42 K
READS:
500KV BS 1-5 11/19/89 s BKR 12X TRBL OHA 43 B
READS:
SEIS RCDR 11/19/89 s SYS ACT OHA 43 D
READS:
RWST/PWST 11/19/89 s OVRFLO
Page No.
07/05/90 PHOTO #/
33 VIDEO FRAME COMPONENT SALEM SIMULATOR STATION vs SIMULATOR HARDWARE DIFFERENCES DATABASE UNIT 1 SORTED BY PANEL ITEM GROUP PANEL NUMBER NUMBER CATEGORY DESCRIPTION DATE IDENTIFIED IMPACT
=================================================================================== ---------- ------
OHA 43 G
READS:
BULK H2 TRBL 11/19/89 s OHA 43 H
READS:
AUX XFMR CLG 11/19/89 s TRBL OHA 43 K
READS:
BS 1-9/9-10 11/19/89 s 500 KV BKR FLASHOVR OHA 44 A
READS:
FIRE PROT 11/19/89 WTR FLO IN 1/2 AUX BLDG OHA 44 READS:
AUX BOIL 11/19/89 s TRBL OHA 44 D
READS:
RWST LVL LO 11/19/89 s LO OHA 44 E
READS:
BLANK 11/19/89 s OHA 44 K
READS:
OSCILLO TRBL 11/19/89 s OHA 45 K
READS:
500KV LN llX 11/19/89 s RT RECVR TRBL OHA 46 D
READS: CH. D 11/19/89 s DECREASING SHUTDN MARGIN OHA 46 K
READS:
500KV/LN21X 11/19/89 s RT RECVR TRBL OHA 47 E
READS:
SOUTH PENET 11/19/89 s AREA AMB TEMP HI OHA B
READS:
FRESHWTR SYS 11/19/89 s TRBL OHA K
READS:
500KV LN llX 11/19/89 s REG/BU POT FAIL OHA K
READS:
500KV LN 21X 11/19/89 s REG/BU POT FAIL
SP-206 HOPE CREEK/SALEM SIMULATOR Initial Condition Validation and Database All simulator modifications or revisions to any procedure used to set the protected ICs must be evaluate~ for impact on all protected initial conditions by the Simulator Performance Engineer.
All changes to a protected initial condition must also be under the direction of the Simulator Performance Engineer.
I Hope Creek Each of the ICs presented in Appendix 3 were developed from IC-01, 02, or 03 representing a cold shutdown plant condition at BOL, MOL, and EOL respectively.
A complete plant startup was performed utilizing all applicable integrated and system operating procedures.
All subsequent ICs are shot recording the various stages of the startup valuable to training such as initial criticality, reactor feed pump startup, main turbine roll, and main generator synchronization.
II Salem Each of the ICs presented in Appendix 3 were developed from the 100% power condition at BOL, MOL, and EOL respectively, and were created utilizing all applicable integrated and system operating procedures.
All subsequent ICs were created at various plant conditions valuable to training such as initial criticality, steam generator feed pump startup, main turbine roll, and main generator synchron-ization etc.
Appendix 3 Page 1 of 1 Date: 08/04/89 Re\\*.:
0
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SALEM SIMULATOR STEADY STATE OPERATION PROCEDURE SSOP-1 28% POWER I.
OVERVIEW - This procedure tests the simulator's steady state performance at 28% power and meets the requirements for a 25% power Operability Test described in ANS-3.5 Appendix B, Section B2.1.
The primary technical basis for this test is a Salem Unit 2 Calorimetric.
II.
REFERENCES A.
ANSI/ANS -
3.5 -
1985, Appendix B, Section B2.1 B.
Salem Unit 2 Calorimetric dated 12/23/86 III.
PERIODICITY -
Annual IV.
INITIAL CONDITIONS A.
28% Power at MOL B.
Equilibrium Xenon
- c.
All Control Systems in AUTO D.
Steady State E.
SG BID at 40,000 lbm/hr F.
Tavg on the Program V.
ACCEPTANCE CRITERIA A.
Instrument error shall not be added to computed values.
B.
Critical parameters shall agree with the expected values to within plus or minus 2%, and shall not detract from training.
C.
Non-Critical parameters shall agree with the expected values to within plus or minus 10%, and shall not detract from training.
Page 1 of 10
SALEM SIMULATOR STEADY STATE OPERATION PROCEDURE SSOP-2 75% POWER I.
OVERVIEW - This procedure tests the simulator's steady state performance at 75% power and meets the requirements for a 75% power Operability Test described in ANS-3.5 Appendix B, Section B2.1.
The primary technical basis for this test are the Salem Unit 2 Control Room Reading Sheets.
II.
REFERENCES A.
ANSI/ANS -
3.5 -
1985, Appendix B, Section B2.1 B.
Salem Unit 2 Control Room Reading Sheets dated 8/24/88 C.
Salem Unit 2 Calorimetric dated 12/29/86 III.
PERIODICITY -
Annual IV.
INITIAL CONDITIONS A.
75% Power at MOL B.
Equilibrium Xenon C.
All Control Systems in AUTO D.
Steady State E.
SG BID at 40,000 lbm/hr F.
Tavg on the Program V.
ACCEPTANCE CRITERIA A.
Instrument error shall not be added to computed values.
B.
Critical parameters shall agree with the expected values to within plus or minus 2%, and shall not detract from training.
C.
Non-Critical parameters shall agree with the expected values to within plus or minus 10%, and shall not detract from training.
Page 1 of 10
SA.LEM SIMULA.TOR STEADY STA.TE OPERATION PROCEDURE SSOP-3 100% POWER I.
OVERVIEW - This procedure tests the simulator's steady state performance at 100% power and meets the requirements for a 100% power Operability Test described in ANS-3.5 Appendix B, Section B2.1.
The primary technical basis for this procedure is Salem Unit 2 Statepoint Data and Control Room Readings.
II.
REFERENCES A.
ANSI/ANS - 3.5 -
1985, Appendix B, Section B2.1 B.
Salem Unit 2 Statepoint Data dated 1/9/87 C.
Salem Unit 2 Strip Charts for 21 SG FF & SF dated 3/12/87 D.
Salem Unit 2 Control Room Readings dated 9/30/87 III.
PERIODICITY -
Annual IV.
INITIAL CONDITIONS A.
100% Power at MOL B.
Equilibrium Xenon C.
All Control Systems in AUTO D.
Steady State E.
SG BID at 40,000 lbm/hr F.
Tavg on the Program V.
ACCEPTANCE CRITERIA.
A.
Instrument error shall not be added to computed values.
B.
Critical parameters shall agree with the expected values to within plus or minus 2%, and shall not detract from training.
C.
Non-Critical parameters shall agree with the expected values to within plus or minus 10%, and shall not detract from training.
Page 1 of 10
REM Part i REACTOR* ENGINEERING PROCBDURB COVER ~HBBT Number
- {
Rev. *_J__
Temporary Y/N A)
Duration of Authorization __
4.J.;..;/.;..;'.4;,;._ _____
.J.
SSI Y/N A)
USQ Y/N DESCRIPTION or CHANGES AND REASONS:
fJ'.f$3:eL-Jl~J e4-/eu/4-/-.r h e-n~.J-bw.
'J 3 ~ c~1 e.J 'XeA4~ k ~~ ~
"'"11',.J~/(/*
_ci.-... p!ef ~ef/~f c.1.u.-.wl- /'-"~CJ!!.J REASON FOR SSI DETERMINATION:
. e'cL,~-1 ~(!Jl ~7*
.
- OTHER. DISCIPLINES REQUIRING REVIEW*:
,Vu-rVe..
TITLE ORIGINATOR.
'\\soR.c MTG. NO.
'... :,_GM -
SO TECH. MGR *
. SIGNATURE OC1.2 9,sss DATE ALL BLANKS ON THIS COVER SHALL BE FILLED BEFORE IMPLEMENTATION Salem Unit *1/S REM Part i
RX. ENG. MAN. PART 2 CALORIMETRIC CALCULATION 2.0 PURPOSE 1~.
This procedure provides a method of calculating the net thermal power output of the reactor core (>15%).
- 2.
This procedure can be used (when required) to evaluate and, if necessary, adjust the gain of the power range excore detector ~hannels.
2.1 APPLICABILITY
- 1.
- This procedure is to be completed daily when the reactor*
power is greater than or equal to 15% (Technical Specification 4.3~1._l.1, Table 4.3-1), or
- 2.
The operator.desires to check the NIS power range instrument.
2.2 INITIAL CONDITIONS
- 1.
- The reactor is being operated in a steady state ~ondition.
- 2.
Tha,the~mal power level: is greater-than 15% of rated thermal power._
- 3.
The reactor coolant average temperature (T
) is within +
o ave
.5 F of the reference temperature {T f).
re 2.3 PRECAUTIONS
- 1.
- 2.
- 3.
With the reactor thermal power less than 15% of rated, inaccuracies in the calorimetric calculation could nullify the results of this procedure.
All nuclear channel r*eadings must be taken off the NIS racks
{not the control console).
Reactor coolant. temperature indication strongly (due to is important that T
= T ave (T
) affects the excore power ave density affects) therefore, it*
0
+. 5 F.
ref
_O_CJ 2. 9 1988
~~GR Df.Et Salem Unit 2 Page 1 of 5 REM Part 2
RX. ENG*. MAN. PAR.'!" 2
- 4.
All data should be taken at appro~imately the same time.
- s.
- Calorimetrics should never be taken immediately after power changes.
An interval should elapse* to allow the plant to come to equilibrium~
- 6.
Once adjusted at the specific power. level the.NIS percent power meter should hold within specification for extended
- time periods.
Some deviation can be expected at lower power levels *du~ to the non-linearity of the channels.
Do not adjust tbe NIS power levels less than full power unless the (NIS~CAL) difference is greater than +/- 2-. If extended operation at reduced power is expected, then it may be advantageous to adjust the NIS to the calorimetric.
2.4 REQUIRED RECORDS
- l.
When this procedure is completed, the operator shall log the
- time and % power in the control room log book. *
- 2.
The original copy of the completed work sheet or computer 1~
printout, dated and signed by a senior reactor operator, shall be retained in the a°perating Department files.*
2.5 PROCEDURE.
- 2. 5.1*
Salem Unit 2 Thermal Power Determination The power delivered by each steam generator will be determined by measurement of -feedwater flow, feedwater temperature, and steam pressure.
Percent reactor power is determined by summing the power being delivered by each steam generator plus radiant heatloss, blowdown and letdow1 losses less the.net input due to pump operation and dividir by the licensed-full power ou~put.
- 1.
- 2.
Stabilize the reactor power level and hold in as near steady state conditions as* possible by minimizing changes in rod position, boron concentration, and steam generator level.
Obtain the required data (feedwater temperature, feedwater flow Delta P, and steam pressure) and record it on the "Data Sheet", Section 2.. 6 of. *this procedure.
Page_ 2 of 5 9 \\'!;~%
-'iP~l '6
,.., i:1 f.M.GA p'£Zi REM Part 2
RX.- ENG. MANG PART 2
- 2.5 PROCEDUR~ (Cont'd.)
2.5.2 Salem Unit 2 3 *.
Complete the _attached "Work Sheet", Section 2. 7 or use the computer code, to obtain reactor thermal lf!b..
power output.
Power Range Nuclear Channel Calibration if the thermal
. power calorimetric was conducted for the purpose of calibrating the power range nuclear channels then complete the following steps:
- 1.
If Reactor Power is <98% and any NIS channel total power indicator is more than 2% below the calorimetric power (i.e. NIS-CAL difference on the data sheet is less ~han -2%) then recheck_
- computation or run another calorimetric. If -2.0%
difference still exists.notify Reactor
'A Engineering.
Adjustment may be required.
CJ).
. 2.
If Reacto_r
- power is > 98% and any NIS channel total power indicator is.more than 1% below the calorimetric power (i.e *. NIS-CAL difference is -.
less than -1,)*or if the average of.the 4NI channels is less than the calorimetric power, then
- adjust each NIS % power meter as follows:
NOTE:
T~e gain pots are very sensitive.
To avoid a "RATE TRIP" slowly adju~t 1 ch~nnel at a time.
On the N41 channel unlock and adjust the gain potentiometer on the "Power Range B" NIS drawer so that the % power indication is corrected for the difference between the* % power indication and the calorimetric power (See Sect. 2.6 "Data Sheet"}.
Adjust the other channels to match N41 in a similar manner.
Lock the "GAIN" pots.
Record each channel power level meter reading, after adjustment, on the Data Sheet (Sect. 2.6).
Page 3 of 5 REM Part 2
!"-~*
~
.i..
PURPOSE HOPE CREEK AND SALEM SIMULATOR Normal Plant Evolutions SP-206 The simulator is required to demonstrate the ability to conduct the minimum evolutions presented in ANS 3.5-1985 Section. 3.1.1, Normal Evolutions.
ACCEPTANCE CRITERIA The simulator shall present continuous real-time simulation of the normal evolutions of the reference plant.
During the conduct of a specific evolution the operator shall not discern a difference between the response of the simulator and reference plant control room instrumentation.
When available the conduct of a specific evolution shall meet the applicable reference plant startup test criteria.
The observable change in the monitored parameters will correspond in direction to those expected from a best estimate analysis and will not violate the physical laws of nature or adversely influence an operators evaluation or decision regarding the evolution.
The simulator shall not cause or prevent an alarm or automatic action if the reference plant would have caused an alarm or automatic action, and conversely, the simulator shall not cause an alarm or automatic action if the reference plant would not cause and alarm or automatic action.
To the extent of reference plant system simulation all applicable reference plant integrated and system operating procedures shall be able to be performed during the conduct of a specific evolution.
DESCRIPTION A complete plant startup is conducted using all applicable plant procedures.
The integrated operating procedures are used as the base procedure, all referenced system operating, surveillance, and other test procedures are performed as directed.
Any procedural differences encountered during the performance is noted and recorded in the Station Procedure Performance Database, presented in Appendix 6.
The startup conducted not only demonstrates the ability to execute each procedural step but is used to set the initial simulator conditions at points selected as valuable to training (Appendix 3)
The following evolutions were performed and verified to meet the stated acceptance criteria using the identified plant operating procedures as the performance standard:
_\\ppendix 10 Pag8 1 r)f 6 Date: 08/04/89 Re'.- - :
0
I I
- I I
- 1.
Plant Startup from Cold to Hot Standby Hope Creek
!I Salem I
OP-IO.ZZ-002 Preparation IOP-2 Cold Shutdown to for Plant Hot Standby Startup OP-IO.ZZ-003 Startup from Cold Shutdown to Rated Power
- 2.
Nuclear Startup from Hot Standby to Rated Power Hope Creek II Salem I
OP-IO.ZZ-002 Preparation IOP-3 Hot Standby to for Plant Minimum Load Startup OP-IO.ZZ-003 Startup from Cold Shutdown to Rated Power
- 3.
Turbine Startup and Generator Synchronization Hope Creek II Salem I
OP-IO.ZZ-003 Startup from IOP-3 Hot Standby to Cold Shutdown Minimum Load to Rated Power
- 4.
Reactor Trip followed by Recovery to Rated Power Hope Creek II Salem I
OP-EO.ZZ-100 Reactor Scram EOP-TRIP-1 Reactor Trip or Safety Injection OP-EO.ZZ-099 Post Scram Restoration EOP-TRIP-2 Reactor Trip
Response
OP-IO.ZZ-002 Preparation for Plant IOP-8 Maintaining Hot Startup Standby OP-IO.ZZ-003 Startup from Cold Shutdown to Rated Power
_i\\ppendix 10 Page 2 of 6 Date: 08/04/89 P. e \\ *. : _ ____oo __ _
- I I
I I
.1 SP-206-
- 5.
Operations at Hot Standby Hope Creek II Salem OP-IO.ZZ-007 Operations IOP-3 Hot Standby to from Hot Minim*um Load Standby
- 6.
Load Changes Hope Creek II Salem OP-IO.ZZ-006 Power Changes I
IOP-4 Power Operation During Operation
- 7.
Startup, Shutdown and Power Operations with less than Full Reactor Coolant Flow Hope Creek II Salem OP-IO.ZZ-006 Power Changes During Operation N/A Technical Recirculation Specification Loops LCO 3.4.1.1
- 8.
Plant Shutdown from Rated Power to Hot Standby and Cooldown to Cold Shutdown conditions Hope Creek II Salem OP-IO.ZZ-004
~
IOP-6 Hope Heat Creek IOP-5
- 9.
Core Performance Testing such as Plant Heat Balance, Determination of Shutdown Margin, and Measurement of Reactivity Coefficients and Control Rod Worth using Permanently Installed Instrumentation ll Salem Balance II Heat Balance I
I I
I Appendi:;: 10 P<1ge 3 of ')
Date: 08/0-J/89 Rev. : _ __co __ _
- I
(.
\\'_
- 10.
Operator Conducted Surveillance Testing ori.Safety~
~. -
Related Equipment or Systems Hope Creek See Procedure Index for a complete listing I Salem See Procedure Ind~x complete listing Appendix 10 Page 4 of 6 for a Date: 08/04/89 Rev.: ----=-0 __
- 1.
Plant Startup from Cold to Hot Standby Hope Creek Initial Date al em Initial Date OP-IO.ZZ-002
~ 2-IOP-2
?IE OP-IO.ZZ-003 SH
'l'l/&~
2.
Nuclear Startup from Hot Standby__ to Rated Power Hope Creek Initial Date Salem Initial Date OP-IO.ZZ-002 814 U2/6'i IOP-3 A-F 1j!S/IJ~
'4iLf16 OP-IO.ZZ-003 fSH
- 3.
Turbine Startup and Generator Synchronization Hope Creek Initial Date al em Initial Date OP-IO.ZZ-003 IOP-3 GF
- 4.
Reactor Trip followed by Recovery to Rated Power Hope Creek OP-EO.ZZ-100 OP-EO.ZZ-099 OP-IO.ZZ-002 OP-IO.ZZ-003
- 5.
Hope Creek OP-IO.ZZ-007
- 6.
Hope Creek*
OP-IO.ZZ-006 Initial Date Salem
~ I rl zrl'i"'
EOP-TRIP-1 l\\zrJ ~()""
EOP-TRIP-2 Ci
- ~(:~
~
Operations at Hot Standby Initial Date Salem
&H IOP-8 Load Chanqes Initial
~H Date Salem IP IOP-4 Appendix 10 Page 5 of 6 Initial Date a f:."
u>l 10/8'!
~E
~tj Initial Date A-
- r Initial Date 1-Date: 08/04/89 Rev. :
0
- 7.
Startup, Shutdown and Power Operations with less than Full Reactor Coolant Flow Hope Creek Initial Date I Salem Initial Date OP-IO.ZZ-006
~
ll\\n\\'lfn I
Technical
~).\\
-4\\ \\7\\<6 N/A Specification LCO 3.4.1.1
- 8.
Plant Shutdown from Rated Power to Hot Standby and Cooldown to Cold Shutdown conditions Hope Creek Initial Date I
Salem Initial Date I
I OP-IO.ZZ-004
~\\.\\
L.l'J I "6 I
IOP-6 It F =
l I
/}_ ;:-
IOP-5 Hope Creek
- 9.
Core Performance Testing such as Plant Heat Balance, Determination of Shutdown Margin, and Measurement of Reactivity Coefficients and Control Rod Worth using Permanently Installed Instrumentation Initial Date Salem Initial Date HE_\\ T BALANCE HEAT BALANCE Hope Creek
- 10. Operator Conducted Surveillance Testing on Safety-Related Equipment or Systems Initial b\\(
Date Salem
~'\\=\\.\\e.,
Appendix 10 Page 6 of 6 Initial Date.
A~
a./4/Aq
/ '/
Date: 08/04/89 Rev. : ---'-0 __
I.
II.
III.
IV.
v.
VI.
VII.
VIII.
IX.
HOPE CREEK/SALEM SIMULATOR PERFORMANCE TEST PROCEDURE S-PTP-034 MANUAL REACTOR TRIP TABLE OF CONTENTS OVERVIEW REFERENCES PERIODICITY INITIAL CONDITIONS ACCEPTANCE CRITERIA DATABASE RECORDING REQUIREMENTS PROCEDURE TEST
SUMMARY
Appendix 11 Page 1 of 10 SP-206 Page 2
2 2
2 2
3 4
5 10 Date: 11/01/89 Rev.:
0
SP-206 I.
OVERVIEW A
- TEST NUMBER: S-PTP-034 B.
TITLE: MANUAL REACTOR TRIP
- c.
ANS-3.5 REQUIREMENTS SECTION NUMBER: 5.4.2, Appendix B.2.2 (1)
D.
CLASSIFICATION:
Best Estimate E.
TERMINATION POINT: Until such time that a stable, controllable and safe condition is attained which can be continued to Cold Shutdown conditions.
II.
REFERENCES A. Salem Unit-2 Licensee Event Report (LER)87-004 B. Salem Unit-2 Strip Chart Recordings:
- 1. Pressurizer Pressure (3/12/87)
- 2. Pressurize Level (3/12/87)
- 3. Stearn-Flow, Feed-Flow & Stearn Generator Level (3/12/87)
- 4. 21 RC Loop Th & Tc (3/12/87)
C. Salem Unit-1 Strip Chart Recordings
- 1. 11 Stearn Generator Level (10/6/85)
- 2. Reactor Power (6/6/86)
D. Salem Unit-2 Startup Report - Figure 2.10.1 III. PERIODICITY: -
Annual IV.
INITIAL CONDITIONS A.
100% Power B.
BOL C.
Equilibrium Xenon V.
ACCEPTANCE CRITERIA A.
During steady state testing for which valid reference plant information is available, critical parameters shall agree within +/-2% of the reference plant parameters.
Non-critical parameters shall agree within +/-10% of the reference plant parameter.
B.
Observed changes in parameters do not violate physical laws.
- c.
All alarms and automatic actions occur as expected.
D.
Unexpected alarms or automatic actions do not occur.
Appendix 11 Date: 11/01/89 Page 2 of 10 Rev.:
0
~
TEST DATE:
VI.
DATABASE A.
Performance/Malfunction
Title:
Malfunction Number:
None
- c.
Ramp Rate:
None D.
Severity:
None E.
General Sequence:
The test operator initiates a manual Reactor Trip from 100%
power BOL. Both the Reactor and Turbine will trip at time zero followed by a Generator trip 30 seconds later.
F.
Probable Causes:
- 1. Failure of the reactor to automatically trip when demanded.
G.
Initial Alarms:
- 1. OHA F-36 MAN RX TRIP INITIATED Appendix 11 Page 3 of 10 Date: 11/01/89 Rev. :
0
SP-206 VII.
RECORDING REQUIREMENTS A.
Parameters
- 1.
Reactor Power
- 2.
RCS Tavg 3.
Pressurizer Pressure 4.
Pressurizer Level
- 5.
Pressurizer Vapor Space Temperature
- 6.
21 Steam Generator Steam Flow
- 7.
21 Steam Generator Feed Flow
- 8.
21 RC Hot Leg Temperature
- 9.
21 RC Cold Leg Temperature
- 10. 21 Steam Generator Steam Pressure
- 11. 21 Steam Generator Level B.
Alarms
- 1.
OHA F-36 MAN RX TRIP INITIATED
- 2.
OHA F-30 TURB TRIP and P-9
- 3.
OHA F-46 RX TRIP
- 4.
OHA G-24 UNIT ISOL TRIP REG
- 5.
OHA G-32 UNIT ISOL TRIP BU
- 6.
OHA E-27 PZR HTR ON PRESS LO
- 7.
OHA E-37 RX TRIP & TAVG LO
- c.
Automatic Actions
- 1.
- 2.
- 3.
Group Busses realign to offsite power
- 4.
Generator Field Breaker opens
- 5.
BS 1-9 and BS 9-10 open
- 6.
Pzr backup heaters energize I deenergize
- 7.
21, 22 and 23 AFW Pumps auto start
- 8.
Feedwater interlock actuates
- 9.
Pzr sprays open Appendix 11 Page 4 of 10 Date: 11/01/89 Rev.:
0
\\~
SP-206 VII
I. PROCEDURE
A.
General
- 1.
Parameters
- a.
Record all Required Parameters on the Data Recording System.
Parameters shall be recorded every 0.5 seconds.
- b.
The failure of a Required Parameter to respond as expected, within Acceptance Criteria of Section V will result in a failure of this test.
Corrective action will be taken and documented using a Discrepancy Report, then this test shall be conducted again.
- c.
A Discrepancy Report may be generated, even though a Parameter response meets the Acceptance Criteria, if the Test Engineer determines that the response could be significantly improved.
- 2.
Alarms and Automatic Actions
- a.
Log the occurrence of the required alarms and automatic actions.
- b.
The time of occurrence of the alarms or actions should be reasonably close to any times specified in the Sequence of Events.
If an alarm or action does not occur reasonably close to times, in the Sequence of Events, then the alarms/action shall be evaluated as UNSAT.
The following factors should be considered when evaluating the time of occurrence:
- 1)
Time in the sequence -
time differences near the end of the sequence can be expected to be greater than time differences at the beginning of the sequence.
- 2)
Related parameter response -
a parameter response that is different from the expected, but satisfactory within the Acceptance Criteria, may cause the times of occurrence to vary considerably.
- c.
The failure of an Alarm or Automatic Action to occur as expected will result in the failure of this test.
The occurrence of an unexpected Alarm or Automatic Action will result in the failure of this test.
In either case, corrective action will be taken and documented using a Discrepancy Report, then this test shall be conducted again.
Appendix 11 Page 5 of 10 Date: 11/01/89 Rev. :
0
- d.
The occurrence of Alarms and Automatic Actions will be evaluated during the Sequence of Events (Section VIII.B).
The Test Operator may use additional personnel as well as the simulator's freeze, backtrack and slow motion capabilities when multiple alarms/
actions are occurring in a short period of time.
- 3.
Other variables -
The following variables may have a significant impact on the results of this test:
- a.
Pressurizer Spray Valve Leakage,
- b.
Exact time 23 AFW flow was reduced,
- c.
Which charging pump was in service.
Appendix 11 Page 6 of 10 Date: 11/01/89 Rev.:
0
B.
Anticipated Response (Sequence of Events)
Test Date:
- 1.
T = 2 sec.
- a.
Manual Reactor Trip OHA F-36
- b.
Reactor Trip OHA F-46
- c.
Turbine Trip and P-9 OHA F-30
- d.
- e.
- 2.
T = 10 sec.
- a.
PZR HTR ON PRESS LO OHA E-27 @
2210 psig
- b.
PZR Backup Heaters energize @
2210 psig
- c.
21, 22 and 23 Auxiliary Feedwater Pump actuated @ 8.5% S/G level
- 3.
T = 30 sec.
- a.
Generator Protection OHA F-47
- b.
Unit Isolation Trip Reg. OHA G-24
- c.
Unit Isolation Trip BU OHA G-32
- d.
Swap of Group Bus Power Supplies
- e.
Generator Field Breaker open
- f.
500 Kv Breakers 9-10 and 1-9 open
- g.
Feedwater Interlock actuation (554°F)
- h.
Reactor Trip and Tavg Low OHA E-37 (554°F)
- 4.
T = 3 min
- a.
Operator secures 23 AF Pump Appendix 11 Page 7 of 10 SAT UN SAT Date: 11/01/89 Rev. :
0
- 4.
T = 8 min.
- a.
Pzr HTR ON PRESS LO OHA E-27 clears at 2218 psig increasing
- b.
PZR Backup Heaters de-energize at 2218 psig increasing
- 5.
T = 10 min.
- a.
Pressurizer spray actuates at 2250 psig
- 6.
T = 20 min.
- a.
End of Test Appendix 11 Page 8 of 10 Date:
Rev.:
SP-206- -
11/01/89 0
- c.
, SP-206
~4i1kv I
I Parameter Evaluation Test Date:
Evaluate the recorded parameters against the Acceptance Criteria of Section V.
- 1.
Parameters A.
Reactor Power B.
RCS Tavg C.
Pressurizer Pressure D.
Pressurizer Level E.
Pressurizer Vapor Space Temperature F.
21 Steam Generator Steam Flow G.
21 Steam Generator Feed Flow H.
21 RC Hot Leg Temperature I.
21 RC Cold Leg Temperature J.
21 Steam Generator Stearn Pressure K.
21 Steam Generator Level Appendix 11 Page 9 of 10 SAT UN SAT Date: 11/01/89 Rev.:
0
IX.
TEST
SUMMARY
Test Date:
SP-206 fl/nf?IJ (For any UNSAT items, include the Discrepancy Report number)
A.
UNSAT Anticipated Response (from section VIII.B)
/Jo JC B.
UNSAT Anticipated Parameter Response (from Section VIII.C)
Simulator Appendix 11 Page 10 of 10 Engineer Date: 11/01/89 Rev.:
0
PSIS DEG F DEG FDEG F 3000 630.
700. 700.
2700 620.
630. 630.
2400 610.
560. 560.
100 600.
490. 490.
1800 590.
420. 420.
1500 580.
350. 350.
1200 570.
280. 280.
- 00. 560.
210. 210.
- 00. 550.
140. 140.
300. 540.
70.0 70
- SALEM SIMULATOR PERFORMANCE TEST 034 MANUAL REACTOR TRIP FROM 100% POWER 4
3
- ooo 530..ooo.000-t-~~~~1--~~~~~~~---;~~~~-t-~~~--t~~~~+-~~~-+~~~~+-~~~-t-~~~--t 4
3 2
1 1
- 21 RC LOOP Th 2
- HOT LEG PRESS PT-405 120.
-SOLID LINES indicate SIMULATOR DATA
---DASHED LINES indicate REFERENCE DATA 240.
380.
480.
BOO.
720.
840.
SECONDS 960.
1080 1200 RUN DATE
- 06/29/90 SIMULATOR FILE: PTP034.SIM REFERENCE FILE: N/ A
1200 100.
108. 108.
1080 90.
6.0 96.0 960. 80.C 84.0 84.0 840. 70.
72.0 72.0 720. 60.0 0.0 60.0 600. 50.
48.0 48.0 480. 40.
36.0 36.0 360. 30.
24.0 24.0 240. 20.0 12.0 12.0 120. 10
- I SALEM SIMULATOR PERFORMANCE TEST 034 MANUAL REACTOR TRIP FROM 100% POWER 2
1
. ooo.ooo.ooo.ooo-lJLI:___:::;::::::::::::;::=:=:=:=i=:=======s=====~==i111111111............... ~;.......................... a 1200 4
3 2
1 120.
1
- 21 STM GEN LEVEL 2
- 21 STM GEN PRESSURE 3
- 21 STM GEN STEAM FLOW 4
- 21 STM GEN FEED FLOW
~SOLID LINES indicate SIMlLATOR DATA
---DASHED LINES indicate REFERENCE DATA 240.
360.
490.
600.
720.
840.
SECONDS 960.
1080 RUN DATE
- 06/29/90 SIMULATOR FILE: PTPOS4.SIM REFERENCE FILE: N/ A
DEG F PCT PSIS 700.
100. 250 560.
80.0 234 490.
70.0 226 420.
60.0 2180 350.
50.0 210 280.
40.0 202 210.
30.0 194 140.
20.0 1860 70.0 10.0 178 SALEM SIMULATOR PERFORMANCE TEST 034 MANUAL REACTOR TRIP FROM 100% POWER
- 3 2
.ooo.ooo 1700-t-~~~-+-~~~~~~~~r-~~~-+-~~~--1-~~~~1--~~~~~~~-+~~~-+~~~--1 3
2 1
1. PRESSURIZER PRESSURE 2. PRESSURIZER LEVEL 120.
3
- PRESSURIZER V APDR TEMPERATURE
-SOLID LINES indicate SIMULATOR DATA
---DASfED LINES indicate REFERENCE DATA 240.
360.
480.
600.
720.
840.
SECONDS 960.
1080 1200 xEO RUN DATE
- 06/29/90 SIMULA TOR FILE: PTP034. SIM REFERENCE FILE: N/ A
AMP 1.0E-3 1.0E-4 1.0E-5 1.0E-6 1.0E-7 1.0E-B 1.0E-9 1.0E-10 SALEM SIMULATOR PERFORMANCE TEST 034 MANUAL REACTOR TRIP FROM 100% POWER *
!.OE-11 "'f-'~~---+-~~----t~-------1--------+-----...__. ________.__ ______,.._ ______ -+--------t--------1 1 1200 1
120.
.,.._ 1
- INTERMEDIATE RANGE AMPS
~SOLID LINES indicate SIMll.ATOR DATA
---DAstED LINES indicate REFERENCE DATA 240.
360.
480.
600.
720.
IMO.
SECONDS 960.
1080 RUN DATE
- 06/29/90 SIMULATOR FILE: PTP034.SIM REFERENCE FILE: N/ A
PCT 120.
108.
- 96.
- 84.
- 72.
- 60.
- 48.
- 36.
- 24.
12.
SALEM SIMULATOR PERFORMANCE TEST 034 MANUAL REACTOR TRIP FROM 100% POWER *
- ooo+_...;;:=----f---------4--------1--------+-------+--------+-------_..-------+--------1--------1 1 1200 1
120.
1. POWER RANGE N-41
~SOLID LINES indicate SIMlLATOR DATA
---DASHED UNES indicate REFERENCE DATA 240.
360.
.oao.
600.
720.
640.
SECONDS 960.
1080 RUN DATE
- 06/29/90 SIMULA TOR FILE: PTP034. SIM REFERENCE FILE: N/ A
I.
II.
III.
IV.
- v.
VI.
VII.
VIII.
IX.
OVERVIEW REFERENCES PERIODICITY HOPE CREEK/SALEM SIMULATOR PERFORMANCE TEST PROCEDURE S-PTP-010 SLOW PRIMARY SYSTEM DEPRESSURIZATION TO SATURATED CONDITIONS TABLE OF CONTENTS INITIAL CONDITIONS ACCEPTANCE CRITERIA DATABASE RECORDING REQUIREMENTS PROCEDURE TEST
SUMMARY
Appendix 11 Page 1 of 10 SP-206 Page 2
2 2
2 2
3 4
5 10 Date: 07/18/90 Rev * : -----=-1 __
~
I.
OVERVIEW A.
TEST NUMBER:
S-PTP-010 B.
TITLE:
SLOW PRIMARY SYSTEM DEPRESSURIZATION TO SATURATED CONDITIONS SP-206 C.
ANS-3.5 REQUIREMENTS SECTION NUMBER: 5.4.2, Appendix B2.2(10),
3.1.2 (le) & (ld)
D.
CLASSIFICATION:
BEST ESTIMATE DATA E.
TERMINATION POINT: The end point of this test will be when saturated conditions are attained in the Reactor Coolant System.
II.
REFERENCES A.
Westinghouse Owner's Group Emergency Response Guideline Background Document (HP-Rev. 1): Section ES-1.2 & Executive volume - Generic Issues, Section 2.5.2 B.
Westinghouse Vapor Space Break Analysis (8/82)
C.
Salem Unit 2 Strip Chart for 21 RC Loop Th and Tc (3/12/87)
D.
AD-16 Report 84-13 E.
Salem Unit 1 Strip Chart for Reactor Power (6/6/86)
F.
PRT System Description G.
Salem Unit 2 PRT Tank Graph H.
Salem Unit 2 Technical Specifications: Table 2.2-1 III. PERIODICITY: -
Annual IV.
INITIAL CONDITIONS A.
100% Power B.
EOL C.
Equilibrium Xenon D.
21 and 22 Charging pumps out of service E.
2PR15 open V.
ACCEPTANCE CRITERIA A.
During steady state testing for which valid reference plant information is available, critical parameters shall agree within +/-2% of the reference plant parameters.
Non-critical parameters shall agree within +/-10% of the reference plant parameter.
B.
Observed changes in parameters do not violate physical laws.
C.
All alarms and automatic actions occur as expected.
D.
Unexpected alarms or automatic actions do not occur.
Appendix 11 Page 2 of 10 Date: 07/18/90 Rev. :
1
SP-206 TEST DATE:
VI.
DATABASE A.
Performance/Malfunction
Title:
SLOW PRIMARY SYSTEM DEPRESSURIZATION TO SATURATED CONDITIONS.
B.
Malfunction Number: 18 C.
Ramp Rate: 0 to 60 Minutes D.
Severity: 0 to 210,000 lb/hr E.
General Sequence:
This malfunction simulates a pressurizer vapor space leak by failing a PORV open. The RCS will depressurize and an automatic reactor trip will be generated. Immediately following reactor trip, the system will rapidly depressurize and a safety injection signal is quickly generated.
Within a few minutes subsequent to the reactor trip/safety injection equilibrium pressure is established. The fluid conditions in the RCS at the time of equilibrium pressure establishment may be characterized by slight voiding in the core and upper plenum and hot legs, and saturated or slightly subcooled liquid in the cold legs.
Equilibrium pressure is determined by means of a mass balance of safety injection pump flow and flow through the PORV.
F.
Probable Causes:
- 1. Valve seat cutting.
- 2. Mechanical failure of the valve.
G.
Initial Alarms:
- 1.
OHA K-7 2PR1 NOT FULL CLS
- 2.
OHA E-27 PZR HTR ON PRESS LO
- 3.
OHA E-11 RC PRESS LO Appendix 11 Page 3 of 10 Date: 07/18/90 Rev.:
1
~~~~-
SP-206 VII.
RECORDING REQUIREMENTS A.
Parameters
- 1.
Pressurizer Pressure
- 2.
Pressurizer Vapor Space Temperature
- 3.
Pressurizer Level
- 4.
21 RC Loop Flow Rate
- 5.
22 RC Loop Flow Rate
- 6.
23 RC Loop Flow Rate
- 7.
24 RC Loop Flow Rate
- 8.
Pressurizer Surge Line Temperature
- 9.
Source Range NI-31 Count Rate
- 10. RVLIS Dynamic Range
- 11. Saturation Margin
- 12. 23 RC Loop Tavg B.
Alarms
- 1.
OHA K-7 2PR1 NOT FULL CLS
- 2.
OHA E-27 PZR HTR ON PRESS LO
- 3.
OHA G-42 EH RUN BACK OPER OHA
- 4.
OHA F-32 OT Delta-T
- 5.
PRT High Pressure console alarm
- 6.
PRT High Temperature console alarm
- 7.
PORV High Temperature computer alarm
- 8.
OHA E-19 PZR PRESS LO
- 9.
- 10. PRT High Level console alarm C.
Automatic Actions
- 1.
Pressurizer Backup Heaters energize
- 2.
Turbine Runback
- 3.
- 4.
- 5.
2PR15 closes
- 6.
Safety Injection Actuation (less Charging Pumps)
- 7.
21, 22 and 23 AF Pumps auto start
- 8.
PRT Rupture Discs break Appendix 11 Page 4 of 10 Date: 07/18/90 Rev.:
1
SP-206 VII
I. PROCEDURE
A.
General
- 1.
Parameters
- a.
Record all Required Parameters on the Data Recording System.
Parameters shall be recorded every 0.5 seconds.
- b.
The failure of a Required Parameter to respond as expected, within Acceptance Criteria of Section V will result in a failure of this test.
Corrective action will be taken and documented using a Discrepancy Report, then this test shall be conducted again.
- c.
A Discrepancy Report may be generated, even though a Parameter response meets the Acceptance Criteria, if the Test Engineer determines that the response could be significantly improved.
- 2.
Alarms and Automatic Actions
- a.
Log the occurrence of the required alarms and automatic actions.
- b.
The time of occurrence of the alarms or actions should be reasonably close to any times specified in the Sequence of Events.
If an alarm or action does not occur reasonably close to times, in the Sequence of Events, then the alarms/action shall be evaluated as UNSAT.
The following factors should be considered when evaluating the time of occurrence:
- 1)
Time in the sequence -
time differences near the end of the sequence can be expected to be greater than time differences at the beginning of the sequence.
- 2)
Related parameter response -
a parameter response that is different from the expected, but satisfactory within the Acceptance Criteria, may cause the times of occurrence to vary considerably.
- c.
The failure of an Alarm or Automatic Action to occur as expected will result in the failure of this test.
The occurrence of an unexpected Alarm or Automatic Action will result in the failure of this test.
In either case, corrective action will be taken and documented using a Discrepancy Report, then this test shall be conducted again.
Appendix 11 Page 5 of 10 Date: 07/18/90 Rev.:
1
- d.
The occurrence of Alarms and Automatic Actions will be evaluated during the Sequence of Events (Section VIII.B).
The Test Operator may use additional personnel as well as the simulator's freeze, backtrack and slow motion capabilities when multiple alarms/
actions are occurring in a short period of time.
- 3.
Other variables -
The following variables may have a significant impact on the results of this test:
None Appendix 11 Page 6 of 10 Date: 07/18/90 Rev.:
1
______ __J
B.
Anticipated Response (Sequence of Events)
Test Date:
- 1.
T = 1 sec.
- 2.
- 3.
- 4.
- a.
Pressurizer Heaters On Pressure Low OHA E-27 @ 2210 psig
- b.
Pressurizer Backup Heaters energize
@ 2210 psig T = 20 sec.
- a.
EH RUNBACK OPER OHA G-42
- b.
OT Delta-T Runback T = 30 sec.
- a.
OT Delta-T OHA F-32
- b.
- c.
- d.
PRT High Pressure alarm @ 10 psig
- e.
2PR15 closes @ 10 psig T = 1 min.
- a.
Pressurizer Pressure Low SI OHA E-19
@ 1765 psig
- b.
Pressurizer Pressure Low SI OHA F-21
@ 1765 psig
- c.
Safety Injection (less Charging Pump Operation)
- d.
21, 22 and 23 AF Pumps actuated (S/G level @ 8.5%)
- 5.
T = 4 min.
- a.
PRT High Level alarm @ 87%
- 6.
T = 6 min.
- a.
PRT Rupture Discs break (PRT Pressure reaches 100 psig then drops suddenly)
Appendix 11 Page 7 of 10 SAT SP-206 UN SAT Date: 07/18/90 Rev. :
1
- 7.
T = 10 min.
RCS Equilibrium Pressure Established, Tsat/Psat Margin Indicates Saturated conditions exist in the RCS.
End of Test Appendix 11 Page 8 of 10 Date: 07/18/90 Rev. : ----'1=----
SP-206 C.
Parameter Evaluation Test Date:
Evaluate the recorded parameters against the Acceptance Criteria of Section V.
- 1.
Parameters
- a.
- b.
- c.
- d.
- e.
- f.
- g.
- h.
- i.
j.
k
- Pressurizer Pressure Pressurizer Vapor Space Temperature Pressurizer Temperature 21 RC Loop Flow Rate 22 RC Loop Flow Rate 23 RC Loop Flow Rate 24 RC Loop Flow Rate Pressurizer Surge Line Temperature Source Range NI-31 Count RVLIS Dynamic Range Saturation Margin Appendix 11 Page 9 of 10 Rate SAT UN SAT Date: 07/18/90 Rev. :
1
SP-206 IX.
TEST
SUMMARY
Test Date:
(For any UNSAT items, include the Discrepancy Report number)
A.
UNSAT Anticipated Response (from section VIII.B)
B.
UNSAT Al1ticipated Parameter Response (from Section VIII.C)
(Signature)
Simulator Performance Engineer Appendix 11 Page 10 of 10 Date: 07/18/90 Rev. : __
1 __ _
Page No.
07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEt1 SH1ULA TOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-84-001 I I y
S-1-84-002 I I N
S-1-84-003 I I y
S-1-84-004 I
y Unit 1 Reactor Trip from 10% Power During Startup Ope rat ions, Unit 1 Reactor Trip from 100% -
High-High Level No.
14 SteamGenerator.
The initiating event was a high level spike in 15Cf eedwater Heater, which caused the heater string inlet valve(13CN27) to close. (1CN47l failed to open automatically.
Thiscaused No. 11 Steam Generator Feed Pump to trip on low suctionpressure.
Unit 1 Reactor Trip from 10% During Unit Startup Operations.
Unit 1 Reactor Trio from 60% - Low Low Level No. 13 Steam Generator.
The initiating event was the closure of the inlet valve to a Low Pressure Feedwater Heater string, due to a high level spike in a feedwater heater.
The bypass N
N N
y I I I I l~ill install mal f. during S.S. upgrade
I le I
I I
I I
I I
I I
I I I I
I I
I I
I I
I
~
Page No.
2 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULA TOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-84-005 y
S-1-84-006 I I y
S-1-84-007 I I N
S-1-84-008 I I S-1-84-009 I I N
valve around the heaters (1CN47l failed to open automatically.
Reactor Trip from 100% Due to Turbine Generator Failure.
Unit 1 Service Water Leak Inside Containment on No.
15 containment Fan Coil Unit service water vent line.
Unit 1 Containment System - TYPe B and C Leak Rate - Out of Specification.
Unit 1 Weld Area Degradation of No.
12 Component Cooling Heat Exchanger Service \\*later Piping.
Unit 1 Technical Discrepancy in Design Evaluation Calculations for Service Water Intake Structure.
N I I N
I I N
I I N
I I N
I I
r--------
Page No.
07/16/90 iYPE NUMBER 3
DAiE ENTERED APPLICABLE (Y /NJ SALEJ1 Sii1ULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
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S-1-84-010 S-1-84-011 S-1-84-012 S-1-84-013 LL _____ _
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y Unit 1 Reactor Coolant system - RTD Bypass Line Valve Failures.
Unit 1 No. 2 Fire Suppression Pump Inoperable for Greater than Seven Days.
Unit 1 Charging/Safety Injection Throttling Valves - Disks Becoming Detached from Stems.
Unit 1 Loss of all 4KV Group and Vital Buses - Units 1 and 2 PO'~er was interrupted between the 500 KV yard and the 13KV bus, resulting in a loss of on-site Power to the Unit 1 and Unit 2 4KV Group and Vital Busses.
The event was the result of a Nuclear Control Operator opening the wrong 5DOKV Circuit Switchgear. This was due to not fully understanding the switchgear controls that were available to him.
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4 07/16/90 TYPE NUMBER
- DATE APPLICABLE SALEl1 SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
====================================================================================================
S-1-84-014 I I N
S-1-84-015 I I N
S-1-84-016 I I N
S-1-84-017 I I N
Unit 1 - Unit 1 Vital Bus Blackout Actuation during a refueling outage, 1A Vital Bus was deenergized when the 1A Vital Bus Infeed Breaker failed to close during breaker testing.
18 Vital Bus was deenergized for inspection at the time, a Blackout Loading signal started 1A and 1C Diesels.
Unit 1 Inconsistency Bwtween Tehcnical Specifications and Safety Analysis.
The issue in question concerns the number of operating Reactor Coolant Pumps when in 11ode 3.
Unit 1 Late Submittal of Procedure Change to SORC for Review.
Salem Unit 1 Foreign Material in Charging Pump Suction Lines.
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r Page No.
07/16/90 TYPE NUMBER 5
DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST S-1-84-018 I I N
S-1-84-019 N
S-1-84-020 N
S-1-84-021 I I N
S-1-84-022 I I N
S-1-84-023 I I y
Unit 1 lnadvertant Safety Injection Signal.
Unit 1 Impingment of Sea Turtle in the Circulating Water Intake.
Unit 1 Containment Air Locks - Design Deficiency Salem Unit 1 Containment Isolation Valve 1CC131 Inoperable (Unit Shutdown) -
Equipment Failure a trip of No. 21 Steam Generator Feed Pumo resulted in a steam flow/feed flow mismatch on No. 23 Steam Generator.
Unit 1 Containment Isolation Valves Eronously Reported as Inoperable.
Salem Unit 1 Reactor Trip form 8% While Performing Turbine Oversoeed Test.
P-7 permissive, which is N
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Page No.
07/16/90 TYPE NUMBER 6
DATE APPLICABLE SALEI1 SII1ULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGfl CAUSE ANS 3.5 srn TEST
====================================================================================================
S-1-84-024 I I N
S-1-84-025 I I y
provided by signals from turbine first stage pressure and from power range flux channels, was not available due to an erroneous output from one of the turbine first stage Pressure transmitters.
Salem Unit 1 Engineered Safety Feature Actuation System Feedwater Isolation Malfunction during Unit startup operations, while process of shifting from Auxiliary Feedwater to Main Feedwater, 11, 12 and 13BF13 Valves closed. It was apparently caused by a spurious feedwater actuation singal.
Unit 1 Reactor Trips from 91% and 93% due to Low-Low Level No.
13 Steam Generator malfunctions of the turbine Electro-Hydraulic Control System (EHCSl resulted in reactor trips, Rapid load rejection caused a sudden reduction of steam N
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Page No.
7 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-84-026 I I N
S-1-85-001 I I N
S-1-85-002 I I y
S-1-85-003 I I N
flo\\.J, and a 'shrink" of the steam generator indicated 1 eve l.
Unit 1 Containment Isolation Valve 11MS18 - Inoperable.
Unit 1 AFW Pump Circuitry dose not Meet Single Failure Criteria.
Unit 1 Containment Pressure Relief Operations Not IAW Tech. Spec.
ReQuirements the 1R41B channel setpoints were not reduced.
The event was attributed to personnel error.
Unit 1 12MS28 Closed Signal to SSPS Train
'B' Inoperable while performing weekly turbine valve testing, it was discovered that No.
12 Turbine Stop Valve ( 12MS28) instrument channel was possibly inoperable. It was subseQuently discovered that one N
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Page No.
8 07/16/90 iYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-85-004 I I N
S-1-85-005 I I N
S-1-85-006 I I y
S-1-85-007 N
of the two closed limit switches on 12MS28 was malfunctioning.
Salem Unit 1 Foreign Matter Contamination of new Terrestic T-68 Lube Oil.
New oil was obtained from another generating station, and all pumps utilizing T-68 oil were drained, flushed and refilled.
Unit 1 Waste Gas Decay Tanks Not Sampled Prior to Releasing contents the pre-release were not representative of the contents of the tanks.
Unit 1 Service Water Leak In.side of Containment.
A containment entry revelaed a service water leak had developed on No. 13 CFCU motor cooler.
Unit 1 No. 14 Waste Gas Decay Tank -
Inadvertant Release N
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9 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y /N1 SALEt1 SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIN MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-85-008 I I y
S-1-85-009 I I N
S-1-85-010 I I N
of Contents while leak checking the Waste Gas Analyzer System, No. 14 Waste Gas Decay Tank depressurized through the Plant vent.
The plant vent monitors were operational, and the inadvertent release was monitored and recorded.
Unit 1 Service Water Leak Inside of Containment. A Containment entry revealed a service water leak from a pipe plug on No.11 Containment Fan Coil Unit ( CFCU) motor cooler.
Unit 1 Fire Watch not Continuously Maintained I.A.W.
Technical Specifications.
The root cuase of this event was the lack of coordination between the maintenance supervisor and the operations shift supervisor.
Unit 1 Waste Gas Holdup System not N
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10 07/16/90 SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM
~10D TYPE DATE APPLICABLE MODIFICATION COMPLETE PLANT ANS 3.5 NUMBER ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE srn TEST
====================================================================================================
S-1-85-011 I I y
S-1-85-012 I I N
Continuously Sampled for Oxygen.
The event was caused by an incorrect valve lineup which apparently occurred on following sampling of No. 13 WGDT.
The event was therefore attributed to personnel error and failure to follow procedures.
Unit 1 Reactor Coolant System Unidentifed Leakage Exceeded Allowable Limit. Technical Specification Action Statement 3.~.6.2.b was not entered (as required by procedures), and that a visual
' estimate" of leakage from the Pressurizer spray valves was used to quantify the leakage.
In order to classify leakage as 'identified', it must be accurately quantified; visual estimation of leakage is not acceptable.
Salem Unit 1 Turbine Trip/Reactor Trip From 99% due to N
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11 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM Sli1ULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-85-013 I I y
S-1-86-001 I I N
False Low Condenser Vacuum Signal a turbine trip
- occurred when the Senior Shift Supervisor, while troubleshooting an apparent false condenser 'Low-Low Vacuum " alarm, opened the vent valve.for the alarm sensor.
3 Unit 1 Inadvertent Loss of two Emergency Core Cooling System Subsystems while attempting to tag No. 22 centrifugal Charging Pump inadvertently deenergized the control power for No. 12 centrifugal Charging Pump charging pump (No.
- 11) was inoperable at the time.
Unit 1 Reactor Trip from 100% on High Negative Flux Rate.
Trip occurred immediately follwing inadvertent de-energization of the lA Vital Bus.
The loss of the bus resulted in the loss of one of the two power supplies for the rod control N
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12 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SHl TEST
====================================================================================================
S-1-86-002 S-1-86-003 S-1-86-004 S-1-86-005 S-1-86-006 I I N
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power and logic cabinets.
- However, the remaining power supply failed due to an internal fault, resulting in the loss of all power to the logic cards.
1 - Failure to Implement Portions of the Inservice Testing Program.
It was discovered that current testing procedures did not reflect all testing specified in the program.
Unit 1 Reactor Trip from 100% caused by partial closure of 11BF19.
11BF19 drifted partially closed and failed to respond to manual control.
1 - Plant Vent Samole not Obtained as Required by the RETS.
1 - Diesel Generator Surveillance Performed Late.
1 - Reactor Trip N
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13 07/16/90 TYPE NUMBER DATE ENTERED APPUCABLE SALEM Sii1ULA iOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG~ CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-86-007 I I N
S-1-86-008 I I N
S-1-86-009 N
S-1-86-010 I I y
from 100% caused by the closure of 14BF19 and failed to respond to the full open demand signal in the automatic or manual mode of operation.
1 - Environmental Qualification Discrepancies Unit 1 Not all Required l/alves Listed in the Valve Position Verification Surveillances following a routine review of Unit 2 Service Water System operating Procedures, it was determined that several valves should be added to Surveillance Procedure SP(O)
- 4. 7. 4. a.
Unit 1 Oxygen Content of Waste Gas Decay Tanks Exceeded Allowable Limits.
Unit 1 Reactor Trip from 95% due to the Loss of Both Steam N
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I I I I BF-65 Trip removed.
Page No.
14 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE
[Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-86-011 I I N
S-1-86-012 I I N
S-1-86-013 I I y
Generator Feedwater pumps.
The simultaneous trip of the SGFP's was caused by the failure of a manual restraining device utilized to maintain the 'closed" limit on 1BF65.
This limit switch, which was defective, provides an interlock to trip the SGFP's if 1BF65 is not fully cloaed.
Fire Watch not Continuously Maintained IAW Tech.
Specs.
1 - Reactor Trip from 100% - Main Generator Protection (APT Differential Realy Actuation) which resulted from shorted windings within the APT.
1 - Reactor Trip from 64% - No. 13 S/G Steam Flow/Feed Flow Mismatch with Low S/G Water Level.
13CN27 closed, resulting in the loss of one feedwater heater train and the N
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15 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 srn TEST
====================================================================================================
S-1-86-014 I I N
S-1-86-015 I I N
S-1-86-016 I I y
subsequent tripping of No. 11 Steam Generator Feedwater Pump (SGFP) on low suction pressure.
1 - Reactor Trip from 15% - Turbine Trip and P-7 while in the process of shifting Main Turbine Lubricating Oil Coolers a momentary drop in pressure which occurred in the Main Turbine.
1 - Environmental Qualification of Raychem Heat Shrinkable Tubing.
1 - No. 11 Steam Generator Feed Pump tripped on the overspeed trip.
The operators p~rf ormed the normal initial actions.
The steam dump opened due to the load rejection signal and remained ooen following the power reduction to 70%.
The steam dump was then closed causing No. 12 Steam Generator level to shrink to the low-low level setpoint.
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16 07/16/90 TYPE NUMBER DATE APPLICABLE SALE!1 SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED
( Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE Af~S 3.5 SIM TEST
====================================================================================================
S-1-86-017 N
S-1-86-018 I I N
S-1-86-019 I I N
S-1-86-020 I I-N S-1-86-021 I I 1 - Fire Door C-8-1 Inoperable - Failure to Enter L.C.O.
Action Statement.
1 - Environmental Qualification of Limitorque Motor Valve Operators.
1 - i.S. 3.7.11 Non-Compliance -
Fire Barrier Wall ImPariment Discovered.
1 - Technical Specification Surveillance 4.7.7.1.A (ensure operability of the Auxiliry Building ventilation charcoal filter train) was not completed within the required time
[late by 3 days).
1 - T.S.
Surveillance 4.3.3.9
- Detector 1R41C Functional Test was not Performed in Specified time due to Personnel Error.
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17 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEt1 SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-87-001 I I N
S-1-87-002 I I N
S-1-87-003 I I N
S-1-87-004 I I N
Unit 1 during a periodic review of chemistry log book data, it was discovered that on January 30, 1987, the Boron concentration limit for the Refueling Water Storage Tank
( RWST) had been exceeded.
Unit No. 1 Loss of Control of a High Radiation Area Locked Door due to Personnel Error a Radiation Protection Technician found a Plastic shoe cover wedged to block open a normally locked door to the Unit 1 bioshield.
Unit 1 Containment Pressure/Vacuum Relief Valves Open Beyond 1000 Hour Limit due to Procedural Inadequacy.
Unit 1 Diesel Generator Missed Surveillance due to Inadequate Post
- Maintenance Testing N
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18 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM Sii1ULAiOR Significant Plant Operating Events Sorted by iype Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE Ar~S 3.5 SIM TEST
====================================================================================================
S-1-87-005 I I N
S-1-87-006 I I y
Caused by Personnel Error.
Unit 1 lf Group Bus Underf requency Protection Inoperable due to Mispositioned Knife Switch because these switches are vital components located outside of a controlled area, locks have been installed on the doors to the group bus electrical panels which contain them to limit access to these switches.
Both Trains of High Head SI Declared Inoperable - T.S.
3.0.5 entered it was observed that the spring charging motor, for No. 11 Centrifugal Charging Pump breaker was not attached to the breaker framework.
The pump was subsequently declared inoperable.
Prior to discovery No. 12 CCP emergency power supply had been made unavailable.
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19 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by iype Number IS SIMULATOR SIM MOD MODIFICATION COMPLEiE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGil CAUSE ANS 3.5 SHI TEST
====================================================================================================
S-1-87-007 I I N
S-1-87-008 I I N
S-1-87-009 I I N
S-1-87-010 I I N
S-1-87-011 I I N
1 - Turb. Trip/Rx.
irip from 100% -
5021 Deams Line Cross Trip Scheme -
Lighting Strike.
Salem Soecial Report 87-7.
Unit 1 Failure to Implement Portions of the Inservice Testing (IST)
Program.
Salem - Appendix R Criteria Non-Conformance.
1 - Non-Compliance with 10CFR 50 Appendix A Critieria a discovery was made that the control circuitry for the Diesel Generator Fuel Oil Transfer Pumps No.21 11 and 12 did not meet the signal criteria as specificed by the Code of Federal Regulations.
Salem Generating Station - Unit 1 Potentially Inadquate Breaker Coordination it was N
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20 07/16/90 SALEM SIMULA TOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD TYPE DATE APPLICABLE MODIFICATION COMPLETE PLANT ANS 3.5 NUMBER ENTERED (Y/Nl DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE SIM TEST
====================================================================================================
S-1-87-012 I I N
S-1-87-013 I I N
S-1-87-0H N
determined that breaker coordination could not be shown to be documented for several voltage levels in either Salem Unit 1 or Unit
- 2.
Salem Generating Station - Unit 1 Technical Specification Non-Compliance -
Reactor Trip System Instrumentation.
The root cause of this event was inadequate Relay Department procedures.
Salem Generating Station - Unit 1 Reactor Trip from 0%
Power due to Inadequate Design, Source Range Detector N31 was found to contain water resulting in an increase of leakage current between the detector tube wall and the detector housing wall caused the source range meter to peg high.
Salem Generating M
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21 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-87-015 I I N
S-1-87-016 I I N
S-1-87-017 I I N
Station - Unit 1 Loss of Control of a Locked High Radiation Area Door due to Personnel Error. Radiation Protection Technician found a plastic shoe cover wedged to block open a locked High Radiation Area door into the Unit 1 bioshield.
Unit 1 Technical Specification 3.8.1.2.b -
Non-Compliance due to Personnel Error.
Fuel was moved from the core to the Spent Fuel Pool, in support of refueling activities, with two of three Diesel Generators (D/G) inoperable.
Unit 1 Power Operated Relief Stop Valves Cabling Found Degraded - Inad.
Design Rev.
Salem Generating Station - Unit 1 Discovered Leakage Paths (23) AFW Pump Compartment -
Control of Design N
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22 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM Sif1ULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGft CAUSE
- ANS 3.5 SIM TEST
====================================================================================================
S-1-87-018 I I N
S-1-88-001 I I N
S-1-88-002 I I N
S-1-88-003 I I y
Req'ts.
Unit 1 - Lead/Lag Derivative AmPli fiers Improperly Calibrated due to Procedural Inadequacy.
Salem Generating Station - Unit 1 Diesel Generator Day Tanks do not Meet Seismic Criteria due to Inadequate Design
& Review.
Salem Generating Station - Unit 1 T-avg Deviation on all Channels due to Design.
Investigation revealed the resistance compensation circuit for the field wiring lenght (RTD to the Low Level Amp) was jumpered out not in accordance with design.
Salem Unit 1 Rx.
Trip on a False Intermediate Range -
High Flux Signal due to Personnel Error.
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23 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULA TOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
====================================================================================================
S-1-88-004 I I N
S-1-88-005 I I N
S-1-88-006 I I N
S-1-88-007 I I N
Upon repeating the procedure, the technician oerf ormed the procedure out of sequence.
The outout trip signal was not bypassed Prior to pulling the channel fuses.
Salem Unit 1 - Mode Change with Hydrogen Analyzer InoP. due to Missed EQ Surveillance.
Salem Unit 1 -
Technical Specification Surveillance 4.7.8.1.2a, sealed source leak checks, was not performed within six months from the prior surve i 11 ance attributed to inadequate administrative controls.
Salem Generating Station - Unit 1 10CFR 50 Appendix R Cable Design Deficiency due to Design Error.
Salem Generating N
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24 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SUlULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
====================================================================================================
S-1-88-008 I I N
S-1-88-009 I I y
S-1-88-010 I I N
Station - Unit 1 Fire Barrier Dampers due to Inadequate Review of Procurement Documents.
Unit 1 - T. S.
Action Statement 3.7.11 Noncompliance
- Hourly Roving Fire Watch Late.
Salem Generating Station Unit 1 -
Unit 1 Manual Reactor Trip due to loss of EH Pumps 11 and 12 - Poor Communication in Conjunction with Equipment Failure.
No. 12 EH Pump tripped and No. 11 EH Pump failed to automatically start.
With the loss of the EH pumps, and decreasing pressure in the control oil system, the Trubine Governor Valves began to drift.
Salem Generating Station - Unit 1 Potential Loss of D/G Areas Ventilation due to Seismic Concern -
Inad. Design Review.
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25 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEl1 SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-88-011 I I N
S-1-88-012 I I N
S-1-88-013 I I N
Unit 1 ~lissed Technical Specification Surveillance due to Personnel Error.
It was discovered that the monthly Technical Specification Surveillance 4.0.5 for April had not been performed for valve 12SW39.
When a valve surveillance stroke time is greater than 25% of the last surveillance stroke time, the surveillance frequency is required to be increased from quarterly to monthly.
Unit 1 - Tech. Spec.
3.7.9 Non-compliance due to Inadequate Administrative control a snubber was discovered missing from.the Safety Injection (SI) line just downstream of the 1SJ13.
Unit 1 - T.S.
Non-Compliance; Fire N
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26 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-88-0H I I N
S-1-88-015 I I N
S-1-88-016 I I N
Barrier Pene.
Inoperable due to Inad. Procedural Gui dance.
It was identified that several penetration seals did not conform to the correct color or cell structure as recommended by the silicone foam manufacturer.
Salem Generating Station Unit 1 -
Quality Assuace (QA) review of Technical Specification Surveillances identified several missied fire barrier penetrations surveillances.
Unit 1 - A Reactor Trip/Turbine Trip occurred due to low Auto Stop Oil System pressure. Apparent cause of this event has been attributed to an equipment problem. A pressure
!'educing 1/ 32' orifice probably clogged during functional testing.
Unit 1 - T.S.
Surveillance 4.7.11 N
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07/16/90 TYPE NUMBER 27 DATE APPLICABLE SALEM SIMULAiOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-88-017 I I N
S-1-88-017 I I N
S-1-88-018 N
S-1-88-019 I I N
Non-compliance -
Fire Dampers not Surveilled - Inad.
Adrnin. Con.
Salem Generating Station Unit 1 -
i.S. 3.0.3 Entry; Five Stearn Flow Channels Read Low.
The root cause investigation of the low steam channel readings is continuing, Unit 1 - T.S. 3.0.3 Entry; Five Stearn Flow Channels Read Low.
I and II for Nos. 12 and 14 Stearn Generators and channel II for No.
13 S/G were declared inoperable. These channels were all reading -5% low.
Unit 1 - T.S. Action Statement 3.7.11 Non-Compliance; Hourly Roving FW Patrol Late; Equip.
Prob.
Unit 1 - T.S. Action Statement 3.7.11 Non-Compliance -
Hourly Roving FW N
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28 07/16/90 TYPE NUMBER DATE APPLICABLE SALEl1 SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM NOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE
/~NS 3.5 SIM TEST
====================================================================================================
S-1-88-020 I I y
S-1-89-001 I I y
S-1-89-002 I I y
Patrol Late; Pers.
Error.
Salem Generating Station Unit 1 - Two Trains of an Eng.
Safety System Made Inoperable by Common Mode.
The Containment "Air Particulate Detector Trouble" alarm annunciated.
Investigation revealed that the sample flow was high.
Salem Generating Station Unit 1 -
Tech. Spec. 3.0.3 Entry; 3 Groups of CFCUs Inoperable due to Equipment Problems.
Two groups 14 15 CFCUs.
were inoperable due to a No. 12 Service Water Header outage.
The third group, No. 11 CFCU, became inoperable when it developed a SW leak.
Unit 1 - the 1R41C channel began spiking high.
This resulted in an actuation signal for Containment Ventilation N
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29 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEf1 SIMULATOR Significant Plant Operating Events Sorted by iype Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-89-003 I I N
S-1-89-004 I I N
S-1-89-005 I I y
Isolation *as well as automatic closure of the Waste Gas 1WG41.
The root cause of this event has been attributed to an equipment problem.
Tech. Spec. Action Statement Non-Compliance - due to Personnel error.
The root cause of this event has been attributed to personnel error. A data entry clerk had not entered the impairment date for the subject impaired penetration seals into the computer.
T. S. Surveillance not Performed Historically due to Inda. Admin.
Controls.
The surveillance requires functional testing of the Reactor Coolant Pump breaker position trip every refueling.
Unit 1 - TS Action Statement 3.0.5 Entered; Both Trains of ECCS due to Equipment.
No. 11 N
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30 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-39-006 I I y
S-1-89-007 I I y
(SI l Pump ( CCP l were declared inoperable due to inoperability of 12 Service Water header 12 CCP declared inoperable due to inoperability of the lC Deisel Generator.
Unit 1 - 1R41C (RMS) channel began spiking high. This resulted inan actuation signal for Containment Purge/Pressure Vacuum Relief System
{BF) isolation.
The root cause of this event has been attributed to failed electrical components.
Unit 1 - Reactor Trip on No. 14 Steam Generator (S/G) Low Level concurrent with Steam Flow/feed flow mismatch.
At the time of the event, No. 14 S/G steam pressure channel I functional surveillance was in progress.
The root cause of this event has been attributed to personnel error.
(NCO) did not select the correct channel during performance N
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31 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
of I&C procedures.
S-1-89-008 I I y
Unit 1 ESF Actuation y
Blackout Loading on lC Vital Bus due to Personnel Error.
At the time of the event, No. 13 Reactor Coolant PumD was in the process of being started.
The NCO depressed 2nd Level undervoltage defeat for A & B Vital Buses but not for C Bus.
S-1-89-009 I I N
Unit 1 - T.S.
N I I 3.3.3.9 Action 36 Non-compliance - due to Personnel Error.
The root cause of this event has been attributed to inadequate procedures.
Functional testing of the 1R41A channel causes the R41 pump to turn off, however, this is not speci ti call y identified in the functional testing procedure.
S-1-89-010 I I y
Unit 1 - 2 Trains of N
I I an ESF System Made Inoperable by Common Mode.
1R11A, filter j
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32 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/Nl SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SHI TEST
====================================================================================================
paper was changed.
Upon completion of the filter paper change out, the monitor sample pump would not start.
Subsequently, 1R11A, 1R12A, and 1R12B remained inoperable.
S-1-89-011 I I N
T. S. Action N
I I Statement 3.7.lla Non-Compliance due to Personnel Error.
Plant personnel identified 24 impaired fire barrier penetration seals, impaired due to design change work, which had not been reported to the Nuclear Regulatory Commission.
S-1-89-012 I I N
Unit 1 - Reactor N
I I Trip on Turbine Trip due to Personnel Error.
At the time of the event, reactor power was at 10-8 Amps.
The trip signal was the result of a Turbine Trip with power above permissive P-7.
The root cause of this event has been attributed An I&C Technician performing the transmitter functional test.
contrary to the l
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33 07/16/90 iYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-1-89-013 I I S-1-89-014 I I S-1-89-016 I I N
N N
requirements of the procedure.
Missed Tech. Spec.
Surveillance 4.0.5 -
V & P due to Inadequate Administt-ative Control.
Containment Purge/Pressure-Vacuu m Relief Isolation Due To An Equipment Problem.
1R41A channel intermittently spiked.
T.S.3.11.2.5.a Non-Compliance - 02 Concentration in WGDT ) 2% for } 48 Hours.
The root cause of this event has been attributed to system design.
During refueling outages hydrogen proxide is used to remove crud from the internal surfaces of Reactor Coolant System (RCS) and reactor components.
This results in oxygenation of the RCS water. During RCS drain down a significant volume of water is drained to the eves Holdup N
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34 07/16/90 SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD TYPE DATE APPLICABLE MODIFICATION COMPLETE PLANT ANS 3.5 NUMBER ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE SIM TEST
====================================================================================================
i anl<s.
S-1-89-020 I I N
1 - Tech Spec Surv.
N I I 4.5.2h non-compliance; max charging pump SI flow rate exceeded (SI) line flow metering orifices were installed backwards. During
. the last refueling outage, the indicated flowrate was 549 gpm for No.
11 (CCP) 540 gpm for No. 12 CCP.
With the meters correctly installed, the indicated flowrate was 15% higher.
S-2-84-001 I I N
I I System-RTD Bypass Line - 22RC17 &
23RC24 Valve Failu'res radiography results of the Reactor Coolant System Resistance Temperature detector bypass Line valves (in Unit 2) indicated that the stems were detached from the disks on 22RC17 and 23RC24.
S-2-84-002 I I y
Unit 2 Residual Heat N
I I 100 Removal common 6
suction valve (2RH1) inadvertently shut while testing was being performed on the Pressurizer Overpressure Protection System.
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35 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-84-003 I I y
S-2-84-004 I I N
S-2-84-005 I I N
The breakers for the RHR common suction valves were not tagged, as required, prior to POPS testing.
Unit 2 Pressurizer Overpressure Protection System (Power Operated Relief Valves 2PR-1 and 2PR-2) actuated due to the induced pressure transient caused by starting a reactor coolant
- pump, The reactor coolant pump was started as part of the procedure.
Unit 2 28 Diesel Generator Test Failure.
2A Diesel Generator Test Failure.
2A Diesel Generator, the generator output breaker tripped on overcurrent.
The trip occurred following the starting of a reactor coolant pump and an auxiliary feed pump (approximately five minutes apart, and one minute after the auxiliary feed pump was started).
2A y
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36 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3. 5 SIM TEST
====================================================================================================
S-2-84-006 I I N
S-2-84-007 I I N
Diesel Generator was declared inoperable the occurrence was attributed to an isolated case of the overcurrent relay not fully resetting following the reactor coolant pump start.
Unit 2 Electrical Power Systems - Loss of 28 4KV Vital*Bus occured while paralle 1 ing 2B Emergency Diesel Generator with the grid.
The Bus differential Protection Relay actuated, which, in turn, actuated the Multi-Trip Relay and tripped 2B Emergency Diesel Generator Breaker and the 4KV Vital Bus Infeed Breaker.
The event was attributed to paralleling the generator out-of-phase.
Unit 2 while performing routine surveillance on the rod control assembilies, it was discovered that four rods were missing from the surveillance N
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37 07/16/90 TYPE NU!1BER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant pperating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-84-008 I I y
S-2-84-009 I I N
check-off sheet.
Due to a typographical error.
Unit 2 during routine power operation, a turbine/reactor trip occurred as a result if a false low condenser vacuum signal. ihe low vacuum singal (which caused the turbine trip) was due to a water slug entering the turbine trip block, via an instrument line, while troubleshooting the problem with the low vacuum alarm.
Unit 2 Non-Representative Sample of No. 23 Gas Decay Tank Prior to Release of Contents.
The pre-release samples were obtained as required; although, unknown to the chemist, the remote operated sample isolation valve remained shut.
Consequently, the samples were drawn on a dead leg of PiPing, and were not representative of the tank contents.
N I I N
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38 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-84-010 I I y
S-2-84-011 I I N
S-2-84-012 I I
'{
Unit 2 On April 23, 1984, a turbine trip and reactor trip occurred during unit startup operations, dut to high-high level in No. 23 Steam Generator.
The event was attributed to sluggish response of the Feedwater Level Control System during low power operation.
Unit 2 reactor power level at six percent testing was being performed on No. 23 Steam Generator Water Level Control System. Test results revealed that No. 23 Steam Generator Feedwater Flow indication channels were not responding, Both channels were declared inoperable.
Unit 2 Reactor Trip From 5% Due to Steam Flow/Feed Flow Mismatch unit startup operations were in progress.
Upon latching the turbine, No. 21 N
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39 07/16/90 iYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEi1 SIMULA TOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SHl TEST
====================================================================================================
S-2-84-013 I I y
S-2-84-014 I I y
Turbine Stop Valve failed to open.
Personnel were immediately dispatched to investigate.
The stop valve was subsequently opened, resulting in an increase in turbine speed.
Unit 2 Reactor Trip From 100% Due to Personnel Error While Testing due to steam flow/feed flow mismatch, coincident with No. 22 Steam Generator low water trip ~ccurred while troubleshooting and calibrating No. 22 Steam Generator Narrow Range Level Recorder.
Unit 2 Containment Ventilation Isolation -
Inoperable a containment pressure relief was performed utilizing the Plant vent Gaseous Activity Monitor (2R41C) in lieu of the containment Gaseous Activity Monitor (2R12A).
This occurrence was attributed to the failure to follow the operating N
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40 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-84-015 I I y
S-2-84-016 y
S-2-84-017 I I N
S-2-84-018 I I N
procedure as written.
Unit 2 Reactor Trip From 10% During Unit Shutdown Operations.
This reactor trip was initiated by (I.R.) high flux circuitry attributed to a conservative LR. high flux trip bistable setpoint combined with an excessively large bistable ' reset deadband'.
Unit 2 Controlled Shutdown Due to Charging Line Leak.
Unit 2 Impingement of Sea Turtle in the Circulating Water Intake.
Unit 2 while performing the final steps of the Pressurizer Overpressure Protection System functional test, Reactor Coolant System pressure rapidly decreased upon opening POVR block valve 2PR6.
The depressurization was caused by the y
I I Add Bistable Trip.
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I I 2PR47 Removed.
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SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-84-019 I I N
S-2-84-020 I I N
S-2-84-021 I I y
inadvertent opening, and failure to reseat, of POPS relief valve 2PR47.
Investigation of 2PR6 revealed a broken wire in the valve operator circuit.
Unit 2 a controlled cooldown to Mode 5 was being performed.
both Containment Spray Pumps were inadvertently cleared and tagged while the Unit was in Mode 4.
Unit 2 Component Cooling System -
Missed Surveillance during a periodic audit of the Component Cooling System valve lineup, it was discovered (2CC37) which is a normally locked valve, was not locked as required.
The routine surveillance is not performed on this valve because of its normally "locked' status.
Unit 2 Reactor Trip From 100% Due to Low N
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42 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-84-022 I I y
S-2-84-023 I I y
Low Level #24 Steam Generator during normal power operation, a trip of No. 21 Steam Generator Feed Pump resulted in a steam flow/feed flow mismatch.
The feed oumo trip was not immediately recognized due to the failure to receive the bezel alarm.
Unit 2 a reactor trip from fifty-four percent power occurred due to steam flow/feed flow mismatch coincident with low water level in No. 24 Steam Generator.
The root cause of the event was a sheared shaft on No. 22 Condensate Pump, caused by fracture of the lower pump bearing support. Air entrainment into the system caused speed oscillations of No.
22 Steam Generator Feed Pump resulted in the pump tripping on oversoeed.
Unit 2 the Plant Vent Samole Pumo was found to be N
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43 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
5-2-84-024 I I y
S-2-84-025 I I N
S-2-84-026 I I N
S-2-85-001 I I N
inoperable due to the remotely located switch being in the off position.
Salem Unit 2 Reactor Trip From 100% Due to Turbine Generator Failure. ihe ground fault occured on generator.
Salem Unit 2 WeeklY Plant Vent Particulate Sample Not Analyzed Within Time Required By Technical Specifications.
Unit 2 Radioactive Liquid Release Not Continuosuly Recorded.
Unit 2 2 A Diesel Generator - Valid Test Failure.
During an operational retest, following completion of repairs to 21SW39 (2A Diesel Generator Service Water Control Valve), the diesel generator tripped as the result of a high jacket water temperature signal.
It was discovered that 21SW39 valve N
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H 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG~ CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-85-002 I I N
S-2-85-003 I I y
S-2-85-004 I I y
actuator had been installed incorrectly, Unit 2 2A Diesel Generator - Valid Test Failures.
2A Diesel Generator failed to automatically start and accelerate during surveillance testing.
Unit 2 during Reactor Coolant System (RCS) fill and vent operations, with RCS pressure at 325 psig, Pressurizer Overpressure Protection System (POPS) Channel II actuated when a reactor collant pump was started.
Unit 2 Reactor Trip From 25% During Startup Operations.
The trip occurred after the genrator was synchronized to the grid, and was the result of No. 24 Steam Generator steam flow/feed flow mismatch, coincident with a low steam generator water N
y N
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200 9
RCS IJpgrade.
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45 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-85-005 I I N
level signal.
24 Steam Generator Steam Flow Channel bistables were in a tripped condition which essentially reduced the anticipatory trip for loss of feedwater to a low steam generator water level trip alone.
Unit 2 Reactor Trip From 17.5% Power During Startup Operations a.
water hammer. noise was heard in the reheat steam line associated with No.
21 Steam Generator Feed Pump.
An equipment operator opened a steam trap drain valve, a solid stream of water issued from the
- trap, While in the process of draining water from the reheat steam line, No. 21 Steam Generator Feed Pump speed and discharge pressure decreased sharply. This was followed by decreasing steam generator water levels.
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46 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/Nl SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-85-006 I I N
S-2-85-007 I I y
S-2-85-008 I I N
S-2-85-009 I I y
Unit 2 Reactor Trip From 54% - Turbine Trip and P-7 during routine power operations, a reactor trip was initiated by a turbine trip, when a partially filled Main Turbine Lube Oil Cooler was placed in service.
Number 22 SG Safety Valves inoperable.
During plant heatup 9 of 5 safeties prematurely lifted.
Salem Unit 2 Reactor Trip From 69% - Main Generator 'Loss of Field' Relay Acutation.
Investigation revealed that the
'Loss of Field' relay (Relay #10 -
Type CEH-11A) was installed incorrectly.
The relay was wired according to the electrical schematic; however, the electrical schematic was not correct.
Unit 2 Reactor Trip N
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47 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-85-010 I I N
From 100% - Dropped Control Rod.
The root cause was attributed to a high resistance connection in Rod 2C~ Control Rod Drive Mechanism cable connector which Prevented the stationary grippers from energizing, resulting in the dropped control rod.
Unit 2 Technical Specification 3.1.3.2.2, applicable in Modes 3,
~ and 5, requires the Reactor Trip System breakers to be opened in the event of an inoperable Individual Rod Position Indication (IRPI) with the unit in Mode 2, it was discovered that this action requirement was not complied with earlier in the day (while in Mode
- 3) when the IRPI for Control Rod 2SA2 was found to be diviating from the group demand indication by greater than twelve steps.
M I I T/S Changed.
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48 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-85-011 I I y
S-2-85-012 I I y
Unit 2 Reactor Trip From 33% - High-High Level No. 21 Steam Generator/Turbine Trip during unit startup operations, a reactor trip occurred from thirty-three percent reactor power level.
The cause of the event was personnel error, with the root cause being attributed to the failure to follow procedures.
Specifically, the Nuclear Control Operator relied on memory to Place the Steam Generator Water Level Control System in automatic operation, rather than utilizing the approved procedure.
Unit 2 Reactor Trip From 10% Due to Low-Low Water Level in No. 23 Steam Generator during unit startup operations, a reactor trip occurred from ten percent power level.
The root cause was attributed to the lack of coordination between operators and supervisors in the control room.
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I I 200 IOP-3 Changed.
4 200 9
200 5
200 9
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49 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED
( Y / N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST*
====================================================================================================
S-2-85-013 I I N
S-2-85-014 I I N
S-2-85-015 I I N
S-2-85-016 I I N
During the Main Steam System warmup operations.
Unit 2 2B Diesel Generator Test Failure Unit 2 2B diesel Generator Test Failure due to Fuel Oil Leak Unit 2 Reactor Coolant System Unidentified Leakage Greater Than T/S Limit.
Investigation revealed the packing glands on 2PR9 (Pressurizer Safety Valve Loop Seal Drain Valve) and 2PS8 (Pressurizer Instrumentation Tap) to be the source of the leakage, The leaks were terminated by adjustment of the packing on both valves.
Unit 2 Boric Acid Tanks & Boron Injection Tank Boron Concentration Below Spec. This event was attributed to an equipment malfunction; i.e.,
2CV173 ( a check N
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50 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
======~=============================================================================================
S-2-85-017 I I y
valve located between the boric acid blender and the BASTS's and the BIT) apparently leaked by while performing Reactor Coolant System dilution operations during the recent startup, This resulted in the addition of approximately 1000 gallons of water to the tanks, and their inadvertant dilution.
Unit 2 Reactor Trip From 100% During Solid State Protection System Testing when Reactor Trip Breaker 'A' automatically opened while testing Solid State Protection System Train "B'.
The reactor trip was caused by a loose wire which supplies power to the undervoltage trip mechanism associated with Reactor Trip Breaker 'A'.
Following the trip, the Atomspheric (MSlO valves) did not open automatically upon increasing steam Pressure, resulting in Main Steam Code Safety Valve N
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51 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE
. SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-85-018 I I y
S-2-85-019 I I y
S-2-85-020 I I y
actuations.
The MS10 valves were opened manually, and the safety valves reseated.
Unit 2 No. 22 Component Cooling Water Heat Exchanger (CCHX) Service Water Outlet Valve (22SW356) failed to the closed position.
Attempts to jack the valve open failed to adequately restore service water flow to the heat exchanger.
Because the redundant CCHX (No.
- 21) was out of service for maintenance at the time, Techncial Specification 3.0.3 was entered.
Unit 2 Service Water Leak in Containment discovered on No. 23 Containment Fan Coil Unit Unit 2 Reactor Trip Manually Initiated Due to Lowering Pressurizer Pressure Valve 2PS3 was suspected to be leaking by and N
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103 5
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52 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
causing the pressure decrease. Attempts to fully seat the valve using the manual control failed. A Unit load reduction and charging were initiated to aid in maintaining pressure; however, pressure continued to decrease.
With pressurizer pressure at 1915 psig and reactor power was initiated, a manual reactor trip.
S-2-85-021 I I N
Unit 2 2B Diesel Generator Test Failure S-2-85-022 y
2 - Reactor Trip/Safety Injection - Voltage Spike on 2C Vital Instruemnt Bus.
The voltage spike ccurred when an I&C technician, who was troubleshooting a problem with the Doric Recorder, inserted the test leads of a frequency counter into a vital bus receptacle.
Subsequent investigation revealed that the receptacle 'hot' and
" neutral' leads were N
y I I I I RCP Brkr Trip logic Charged.
Page No.
53 07/16/90 SALEM SIMULATOR Significant Plant Operating Events Sorted by Tyoe Number IS SIMULATOR SIM MOD TYPE DATE APPLICABLE MODIFICATION COMPLETE PLANT ANS 3.5 NUMBER ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE SIM TEST
====================================================================================================
S-2-86-001 I I N
S-2-86-002 y
reversed.
Unit 2 Waste Gas Holdup System Not Continuously Sampled For Oxygen due to the gas analyzer sampling selector switch being in the No. 21 Waste Gas Decay Tank position.
Unit 2 Reactor Trip/Turbine Trip 50% - No. 23 S/G High-High Level No.
22 Steam Generator Feedwater Pump (SGFP) tripped.
Steam generator water levels shrank to approximately ten percent, recovered following a turbine load reduction and then began oscillating.
Operators attempted to stabilize the levels utilizing manual control of the feedwater regulating valves.
However, before any effective corrective action could be taken, the level in No. 23 Generator reached the high-high level setpoint.
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54 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG~ CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-86-003 I I S-2-86-004 I I y
y 2 Reactor Trip/Safety Injection From 5%
During Controlled Shutdot.1n steam generator t.1ater level oscillations began occurring at approximately 15 percent power.
To preclude a reactor trip in the event that a turbine trip occurred due to high water level signals, control rods were inserted to reduce reactor pot.1er below 10 percent. With reactor power at 5 percent, the Main Generator output breakers were opened.
Approximately 3 seconds later a safety injection (SI) occurred.
The SI was caused by high steam flow coincident t.1ith low average reactor coolant temperature (Tave).
The low Tave was the direct result of operator action to reduce power below 10 percent.
2 Reactor Trip From 100% Power - Loss of 2B Inverter.
The Reactor Protection N
I I y
I I IOP Changed TAVE on program before Turb. Trio.
RCP Brkr open logic changed.
Page No.
07/16/90 TYPE NUMBER 55 DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-86-005 I I N
S-2-86-006 I I y
System sensed a fals*e Reactor Coolant Pump breaker open condition due to the three (3) second time delay for the inverter to transfer to the backup power supply.
2 Reactor Trip During Startup -
Voltage Spike on 2B Inverter.
The technician performing the surveillance connected the test leads incorrectly, which resulted in a voltage spike.
The spike reinstated the P-7 permissive interlock.
2 Reactor Trip From 52% Power - 23 Steam Generator High-High Level root cause to be a combination of equipment malfunction and personnel error.
Sediment in the governor actuator for No. 21 SGFP caused a sluggish response by the pump from an increase in demand signal.
The sluggish response by the pump from an increase in demand N
I I N
I I 200 5
201 7
I I
J
Page No.
56 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/Nl SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-86-007 I I y
signal.
The Operator reaction to mitigate the Steam Generator level transient by increasing No. 22 SGFP speed too quickly caused Steam Generator level to increase rapidly, 2 Reactor Trip/Safety Injection From 100%
& Loss of Offsite Power Indication reactor trip and safety injection were initiated by the spurious actuation of bistables and contacts resulting from a voltage spike on the 2C Vital Instrument Bus.
The
'loss of offsite power ' signal was apparently the result of a sustained undervoltage condition on the Station Power Transformers (SPTsl.
The undervoltage condition was apparently caused by the transfer of the group busses from the Auxiliary Power Transformer to the SPT's and aggravated by multiple transfers between y
I I 104 5
Page No.
57 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-86-008 N
S-2-86-009 I I y
the Station Power Transformers.
2 Technical Specification 3.7.11 Non-compliance -
Impairment of the 122' /100' Elevation Floor Hatch in Auxiliary Building -
No Fire l~atch.
2 Reactor Trip from 74% - Loss of No. 22 Station Power Transformer cause of this event was an electrical fault in 2F 4160/230V Transformer.
The fault resulted in the operation of overcurrent relay protection which opened the 4KV breaker 2F5D supplying the 2F 4160/230V Transformer.
Simultaneously, No.
22 Station Power Transformer (SPT) 13/4KV, was isolated due to Phase A and Phase B differential relay protection operation.
Loss of these busses caused the loss of No. 23 and No. 24 Reactor Coolant Pumps, N
I I N
I I
Page No.
58 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-86-010 I I N
S-2-86-011 I I N
S-2-86-012 I I N
S-2-86-013 I I -
y 2 T. S. 3.7.11 NON COMPLIANCE - FIRE BARRIER PENETRATION DISCOVERED IMPAIRED 2 T.S. Surveillance
- 4. 9. 7. - Not performed Within Specificed Time Due to Personnel Error.
Personnel did not comply with Procedural requirements to test the operability of the Fuel Handling Crane {DB} overload cutoff within seven days prior to crane use.
2 Containment Systems - Type B and C Leak Rate out-of-Specification Due to Valve 2PR25 Excessive Leakage.
2 Turbine Reactor Trip From 8% Power on P-7 Interlock Due To Personnel Error While Controlling Speed Using Governor Valve Position Limit.
The turbine control was in
'manual' due to problems with the Electro-Hydraulic N
I I N
I I N
I I N
I I
_J
Page No.
59 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE 1
ANS 3.5 SIM TEST
====================================================================================================
S-2-86-0a I I y
S-2-87-001 I I N
S-2-87-002 I I y
Control (EHC) {TG)
System.
While stablizing turbine speed with the valve position limiter, sufficient steam was admitted to the turbine to reset the P-7 Interlock and cause the turbine overspeed, resulting in a Turbine/Reactor trip.
2 Reactor Trip from 77% Power On Steam Flow/Feed Mismatch &
23 Steam Generator Low Level Due to Valve 23BF19 Control Problems.
2 Loss of RHR Injection Capability to Two Cold Legs Due to Technical Specification Misinterpretation.
2 Reactor Trip From 3% Power on Erroneous High Neutron Flux Signal Due to Personnel Error.
The Reactor was at three percent (3%) power and the P10 Interlock which blocks trips from Intermediate Range indication, was not N
N N
I I I I I I 201 7
J
Page No.
07/16/90 TYPE NUMBER 60 DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGU CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-87-003 I I N
S-2-87-004 I I y
S-2-87-005 I I y
in effect when the Technician pulled the control power fuse for the channel, resulting in an erroneous
'High Neutorn Flux' signal and subsequent Reactor Trip.
2 Unti 2 Fuel Handling Crane Missed Surveillance Due to Personnel Error.
Unit No. 2 Generator-Turbine/Re actor Trip Due to Loss of Field On The Main Generator.
The investigation results suggest that the loss of Generator excitation occurred due to a bumped transfer when the voltage regulator was shifted from auto to manual.
When this occurred, Generator field current was observed to drop from 5500 amps to approximately 3000
- amps, 2 Turbine/Reactor Trip from 85% Power N
I I y
I I N
I I
Page No.
61 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGfi CAUSE ANS 3.5 SHI TEST
====================================================================================================
S-2-87-006 I I N
S-2-87-007 N
S-2-87-008 I I N
S-2-87-009 I I S-2-87-010 I I N
due to Loss of DC Control Power to the Turbine Electro Hydraulic Control System Caused by a Failed Servo Card.
2 T. S. 3.7.10.3 Non-Compliance -
Inadequate Fire l~atch Due to Personnel Error.
2 T. s. 3.7.11 Non-Compliance -
Discovery of Fire Barrier Impairment.
2 Missed T. S.
Surveillance 4.5.2.b Due to Personnel Error - T.S. 3.0.3 Entered.
Unit 2 Appendix R Criteria Non-Conformance.
Salem Generating Station Unit 2 T.S.
3.7.11 - Fire Barrier Impairment Non-Conpliance due to Personnel Error.
N I I N
I I N
I I N
I I N
Page No.
62 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SG~ CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-87-011 I I y
S-2-87-012 I I N
S-2-87-013 I I N
S-2-87-014 I I N
Salem Generating Station Unit 2 Reactor Trip - No.
24 Steam Generator High-High Level, Personnel Error.
At the time of the triP, a No. 24 S/G Level Channe 1 II functional test was in progress.
The root cause of the trip has been attributed to personnel error.
The Nuclear Control Operator during the functional test did not correct a feed-steam flow deviation.
Salem Generating Station Unit 2 RHR Pump Room Flood Curb Missing due to Personnel Error.
Salem Generating Station Unit 2 Technical Specification Surveillance 4.8.1.1.3 a Missed Due to Inad.
Procedural Control.
Unit 2 Incorrect Diesel Generator Infeed Breaker N
N N
N I I I I I I I I 200 5
201 8
201 DA 103 8
Page No.
63 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-87-015 I I N
S-2-87-016 I I N
S-2-87-017 I I N
Setpoint Due to Inadequate Documentation Control. This resulted in a lack of breaker coordination between the SW Pump {BI}
overcurrent protection relays and the upstream D/G overcurrent protection relays.
Unit 2 and Engineering review of Salem electrical systems revealed that with a degraded grid condition, a LOCA and assuming the 13.8 KV to ' KV station power tap changer fails to function, certain MCC control circuits
{ED} may not have adequate voltage to pick-up their respective MCC starter coil.
Salem Unit 2 2A Diesel Generator Surveillance Missed due to Personnel Error.
Salem Unit 2 T.S.
Non-Compliance due to Procedural N
I I N
I I N
I I
___J
Page No.
64 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by iype Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGU CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-87-018 I I N
S-2-88-001 I I N
inadequacy Unit 2 vital 28 Volt and 125 Volt batteries were declared inoperable when it was found that some of the present measured specific gravities did not meet Technical Specification Surveillances
- L8.2.3.2 and 4.8.2.5.2.
Salem - the required functional test for both Unit 1 and Unit 2 Waste Gas Oxygen Monitors were not performed within their required time frame.
Technical Specification surveillance 4.3.3.9.1.b requires the performance of this functional test every 31 days.
The root cause of this event has been attributed to inadequate administrative controls.
Salem Generating Station - Unit 2 at 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br /> Chemistry was required to draw a sample in accordance with Tehcnical N
I I N
I I
Page No.
65 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIM ULA TOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-002 I I y
S-2-88-003 I I N
S-2-88-004 I I N
Specification Table 3.3-12 Action 28.c.
However, the sample was not taken until 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br /> that day contrary to the requirements of the Action Statement.
Salem Generating Station - Unit 2 T.S. 3.0.3 Entry - 4 SW Pumps Inoperable due to an Equipment Failure.
Salem Unit 2 - On 2/08/88, it was identified that Radiation Monitoring System Channel (RMS) 2Rl 9C {Ill was inoperable since 02/02/88 and the required periodic samples were not taken. during preventative maintenance of the 2R19C Steam Generator Blowdown sample line pressure regulators, flow through the sample line stopped due to the failure of the low pressure regulator.
The Safeguards Equipment Control N
I I N
I I N
__J
Page No.
66 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULA TOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
SGti CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-88-005 I I N
S-2-88-006 I I N
S-2-88-007 I I y
(SEC) System 'output Test & Interface Cabinet' {JG l slave relay testing was not performed on a staggered test basis as required by Tehcnical Specifications Surveillance 4.3.2.1.1, Table 4.3-2.
Salem Unit 2 T.S.
Action Statement 3.7.11.a Non-Compliance -
Hourly Rove late due to Personnel Error.
Rx. Trip from 100%
Power - False No. 23 RC Loop Flow Signal due to Personnel Error. Maintenance technician was repairing a leak on the low pressure side of No. 23 RC Loop Flow transmitter.
Salem Generating Station Unit 2 a Turbine Trip occurred as a result (S/G) High-High Level. A review of the event revealed that just prior to the Turbine Trip N
N N
I I I I I I Added Time delay to pick UP load in CANA.
Page No.
67 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by iYPe Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-008 I I N
S-2-88-009 I I y
S-2-88-010 I I N
reactor power increased from approximately 12% to 18%.
This power increase apparently caused level in No.
23 S/G apparent cause of this event has bee attributed (EHC) problems.
Unit 2 2FP147 was not surveiled on February 10, 1988, as required by Technical Specification Surveillance 4.0.5.
em Caused by A Salem Generating Station Unit 2 Reactor Trip from 97% Power - Control Rod Dropped.
Investigations revealed it is highly probable that Control Rod No. 103 had dropped resulting in the negative rate trip.
Testing to determine why the control rod dropped was inconclusive.
Salem Generating Station Unit 2 Lack of Backup Over current Protection for 37 N
I I N
N I I
Page No.
68 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED
[Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-011 I I N
S-2-88-012 I I y
S-2-88-013 I I y
Electrical Circuits Penetrating Containment due to Inadequate Design Review.
Unit 2 Technical Specification Surve i 11 ance 4.7.10.1.1.c was late.
The root cause of this event has been attributed to inadequate communications between Nucelear Fire and Safety Department supervision and station management.
Unit 2 T.S. Action 3.0.3 Entry - Both Centrifugal Charging Pumps Delcared Inop.
due to Eqiup. No. 21 CCP.
High vibration on the outboard bearing No. 22 CCP could not be placed in service due to its 4KV breaker failing to close.
The breaker's springs had failed to charge after the breaker was last closed.
Unit 2 Tech. Spec.
3.3.2.1.b Action 13 N
I I N
I I N
I I l
Page NQ.
69 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-014 I I y
S-2-88-019 I I N
S-2-88-020 I I N
Non-Compliance SEC Inoperable due to Equip. Problems.
'SEC Trouble Alarm" annunciated cause of the alarm to be associated withthe 2B SEC Auto Test Failure circuitry.
The circuit would not reset.
The Operating Shift did not declare 2B SEC inoperable and subsequently did not enter Tech. Spec.
3.3.2.1.b Action 13.
the "SEC Trouble Alarm" Alarm Response Procedure (ARP) has been revised.
Salem Generating Station Unit 2 Reactor Trip/Safety Injection due to Failure of the C Vital Instrument Bus Inverter.
Salem Generating Station Unit-2 Nos.
22 & 24 Steam Generators - Several Tubes found Degraded; Categorization C-3.
Unit 2 Tech. Spec.
- 3. 9.12 y
I I N
I I N
I I I
Page No.
70 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGti CAUSE ANS 3.5 srn TEST
====================================================================================================
Non-Compliance due to Personnel Error fuel assembly insert changeouts in the Fuel Handling was being conducted with the No. 21 FHB Exhaust Fan {VG}
The Updated Final Safety Analysis Report (UFSAR), requires both FHB Exhause Fans to be operable to consider the FHB Ventilation System operable.
S-2-88-021 I I N
Salem Generating N
I I Station Unit 2 T.S.
Action Statement
- 3. 7.11 Non-Compliance -
Hourly Roving FW Patrol Late due to Personnel Error.
Page No.
07/16/90 TYPE NUMBER 71 DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-02~
I I y
S-2-88-025 I I N
Monitors -
Inadequate Administrative Controls.
Unit 2 Rx. Trip due to Inadequate Procedural Guidance Resulting in an Equipment Problems during turbine startup, No. 23 Steam Generator (S/G) level increased to 67%
resulting in a Turbine Trip.
Following the Turbine Trip, a Reactor Trip occurred as a result of No. 22 S/G low-low level.
Maintenance Procedures do not adequately define how to set the regulator for the BF19 Feedwater Control valve positioner.
Subsequently, the 23BF19 positioner regulator was set high causing the valve to lock open momentarily.
This led to the Turbine Trip on hi~h-high S/G level.
Salem Generating Stat ion Unit 2 N
N I I
Page No.
07/16/90 TYPE NUMBER 72 DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-88-026 I I y
Technical Specification Surveillance 4.3.2.1.1 Table 4.3-2 item BF has not been performed as required.
The surveillance requires functional testing of the (AFW)
System actuation on the trip of the Main Feedwater Pumps every startup if not performed within the previous 92 days.
Unit 2 - Plant Vent effluent release composite iodine and radioactive particulate samples were not obtained as required by Technical Specification 3.3.3.9 Table 3-13 Action Statement.
The root cause of this event h~s been attributed to inadequate administrative control. Operations Department personnel were not made aware of the necessity for entry into action statement 36 upon the loss of flow through the 2R41 plant vent monitor.
N l
J
Page No.
73 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE (Y/N)
SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number DESCRIPTION IS SIMULATOR SIM MOD MODIFICATION ~OMPLETE PLANT REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 SIM TESi
====================================================================================================
S-2-89-001 I I y
Unit 2 - Tech. Spec.
N I I 3.0.3 Entry - SW Headers Inoperable due to Equipment Problems No. 2C SEC was declared inoperable due to an Auto Test Fault.
This SEC will start Nos. 25 and 26 SW No. 22 SW Pump was cleared and tagged to No. 23 SW Pump had failed its
~.0.5-P surveillance due to low SW flow.
S-2-89-002 I I y
Unit 2 - The 2R11A N
I I monitor began spiking high. This resulted in an actuation signal for Containment Purge/Pressure-Vacuu m Relief System (BF}
isolation.
The root cause of this event has been. attributed to an equipment feed problem.
S-2-89-003 I I y
Unit 2 - following y
I I power reduction from 90% to 60%, the Unit experienced a Rx.
Trip on No. 23 (S/G)
Low Level concurrent with Steam Flow/Feed mismatch.
The root cause of this event has been attributed to inadeciuate procedures
Page No.
74 07/16/90 TYPE NUMBER DATE ENTERED APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y/N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 SIM TEST
====================================================================================================
S-2-89-004 I I y
S-2-89-005 I I y
S-2-89-006 I I y
S-2-89-007 I I N
associated with operating the plant with Gire. Water System reduced capacity concurrent with an inoperable Heater Drain Pump.
Procedure AOP-COND-2 has been revised.
2 - the Containment Purge/Pressure-Vacuu m Relief System valves isolated as a result of the failure of (RMS)
Monitor, 2R11A.
2 - Rx Trip/SI from 100% due to deenergization of D-Vital bus.
Both SG feed pumps reduced their speed to idel Rx tripped in SG levels; SI on Hi Stm Flow Low Stm Press.
2 - On March 29, 1989 a Unit shutdown was required to comply with TSAS 3.7.7.b in support of replacement (ABVS) charcoal filter.
With the Unit in Mode 3, a reactor trip signal 21 (S/G) occurred.
The root cause of this event has been N
I I y
I I N
I I N
I I J
r I
I
- 1*
I I
I I
I I
I I
I I
I I I I
I I
I
~
-~-
Page No.
75 07/16/90 TYPE NUMBER DATE APPLICABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT ENTERED (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SG# CAUSE ANS 3.5 srn TEST
====================================================================================================
S-2-89-010 I I N
S-2-89-013 07/26/89 N
S-1-89-024 07/26/89 N
attributed to an equipment problem.
The steamline flow transmitters were found to be out of calibration.
2 - (RMS) 2RUC, channel spiking result in a Containment Purge/
Pressure-Vacuum Relief System (CP/PVRS) isolation signal and a Waste Gas Decay Tank Vent Control Valve isolation.
The root cause of this event has been attributed to a design/equipment problem.
MANUAL REACTOR TRIP FROM 15% POWER; LOSS OF 5 OR 6 CIRCULATINGWATER SYSTEM CIRCULATOR PUMPS DUE TO EXTERNAL CAUSES AND INADEQUATE CORRECTIVE ACTION FROM A SIMILAR PRIOR EVENT.
UNIT 1 - ON JUNE 9, 1989 AT 1641 HOURS, WITH THE UNIT IN HOTSTANDBY A SAFETY INJECTION OCCURRED.
THE TRIP SIGNAL AND SIWERE THE RESULT OF HIGH STEAMLINE DELTA N
N I I N
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Page No.
76 07/16/90 TYPE NUMBER DATE ENTERED APPLiCABLE SALEM SIMULATOR Significant Plant Operating Events Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT (Y /N)
DESCRIPTION REQUIRED DATE
RESPONSE
SGtt CAUSE ANS 3.5 srn TEST
====================================================================================================
P CAUSED BY THE NO.
113MS15 MAIN STEAM (MS) SAFETY VALVE
[SB] LIFTING.
S-1-89-025 07/26/89 N
UNIT 1 - JUNE 9, N
I I 1989, IT WAS DISCOVERED BY OPERATIONSPERSONNEL THAT THE MAIN STEAMLINE MONITOR (R46) CHANNESL DRAINLINE WAS ISOLATED DUE TO CLOSURE OF VALVE
- 1MS217, INVESTIGATION REVEALED THAT THE VALVE HAS PROBABLY BEEN CLOSEDSINCE INSTALLATION COMPLETEION OF A DESIGN MODIFICATION(PACKAGE lEC-2136) IN JULY 1988.
THIS DESIGN CHANGE CONNECTEDTHE R46 CHANNEL STEAM DISCHARGE TO THE CONDENSER.
S-1-89-026 07/26/89 N
UNIT 1 - ON JUNE 18, N
I I 1989 AT 1615 HOURS, A CONTROLLED SHUTDOWNWAS COMPLEiED iN
- ACCORDANCE T. S.
3.3.2.1 TABLE 3.3-3 ACTION 13.
lA SEC HAD BEEN DECLARED INOPERABLE ON JUNE 18 AT 1112 HOURSDUE TO AN AUTO TEST FEATURE LOCAL ALARM.
THE ALAR~l WASDISCOVERED BY A
r Page No.
77 07/16/90 TYPE NUMBER SALEM SIMULATOR Significant Plant Operating Events DATE APPLICABLE ENTERED (Y/N)
DESCRIPTION Sorted by Type Number IS SIMULATOR SIM MOD MODIFICATION COMPLETE PLANT REQUIRED DATE
RESPONSE
. SGtt CAUSE ANS 3.5 SIM TEST
====================================================================================================
MAINTENANCE ELECTRICIAN DURING THE 1C VITAL BUSUNDERVOLTAGE/UNDE RFREQUENCY SURVEILLANCE TESTING.
S-1-89-027 07/26/89 N
ON JUNE 19, 1989 AT N
I I 2100 HOURS, A REACTOR TRIP ON NO.
13 STEAMGENERATOR (S/G) 'LOW-LOW LEVEL' OCCURRED.
13MS167 HAD CLOSED.
AT THE TIME OF THE EVENT, A POST MAINTENANCE OPERABILITY RETEST(PROCEDURE SP(0)4.0.5-V) FOR THE 12MS18 WAS IN PROGRESS.
THE ROOT CAUSE OF THIS EVENT HAS BEEN ATTRIBUTED TO INADEQUATEDESIGN OF THE CONTINUITY CHECK CIRCUITRY FOR THE MS167 VALi/ES.
,,-----------~------------------
Page No.
07/20/90 SALEM SIMULATOR Operator Feedback Sorted by Feedback Number s
D p FEEDBACK DATE I H S 0 R DATE CORRECTIVE NUMBER INITIATOR WRITTEN POSITION SYSTEM DESCRIPTION M W W C 0 FIXED ACTIONS
==============================================================================
S-FB-88-002 BEST, R.
10/25/88 TEST DIRECTOR N510 WASTE GAS SYSTEM IS NOT Y N Y N I I Given low DYNAMIC EXPAND THE SCOPE priority during OF SIMULATION.
1989 annual review. Wil 1 be implemented as time permits.
S-FB-88-005 BEST, R.
10/25/88 TEST DIRECTOR IAOO EXPAND IA OVDO & OVDI TO y N y y I I Given low AT LEAST 16 ITEMS EACH.
Priority during 1989 annual review. Will be implemented as time permits.
S-FB-88-011 MOORE, K.
12/15/88 INSTRUCTOR N530 DURING A LARGE SW LEAK Y N Y N I I Given low (20,000 GPM) IN THE MECH.
priority during PENE.AREA THE EQ GO TO 1989 annual THE AUX. BLDG.
SUMP TANK review. Will be AND RHR SUMP, BOTH OF implemented as WHICH ARE PUMPED TO THE time permits.
WHUT/WMHUT.
THE RECORDER ON RP-1 FOR WHUT/WMHUT.
AND AUX BLDG. SUMP TANKLEVEL NEVER CHANGE.
S-FB-88-013 LOWENSTEN, 12/19/88 NTS MCOO NO. 2 STATION AIR N N N N I I IJ.
COMPRESSOR AMMETER SHOULD INDICATE A STABLE 125 AMPS WHILE OM SERVICE.
S-FB-88-01~ LOWENSTEN, 12/19/88 INSTRUCTOR
!ADO NEED PRESSURIZER LEVEL y N y y I I Given low
- v.
CH. I DMAL Priority during 1989 annual review. Will be implemented as time permits.
S-FB-89-008 HOFFMAN, D.
02/02/89 TEST ENGINEER
- R100 ADD MALFUNCTION(S) WHICH y N y y I I Given low WILL ENABLE THE SG Priority during LEVELTRANSMITTERS WHICH 1989 annual DO NOT INPUT TO SGWLL TO review. l.Jill be BE FAILEDFROM 0 - 100%.
implemented as THIS WILL BE BENEFICIAL time permits.
TO TRAINING,ESPECIALLY IN THE INITIAL TRAINING OF PLANT OPERATORS. THIS WILL
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ALLOW FOR A SG LEVEL TRANSMITTER FAILURE WHICHWILL NOT EFFECT SGWLL BUT WILL REQUIRE RESEARCH INTECHNICAL SPECIFICATIONS.
J
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S-FB-89-015 HOFFMAN, D.
02/28/89 TEST ENGINEER S-FB-89-017 HOFFMAN, D.
03/09/89 TEST ENGINEER S-FB-89-018 HOFFMAN, D.
03/09/89 TEST ENGINEER N100 ADD MALFUNCTION TO FAIL Y N Y Y OPEN THE 600~ RELIEF VALVE (2CV6)AND PREVENT RECLOSURE.
THIS SHOULD PASS FULL LETDOWNFLOW TO THE PRT UNTIL THE 2CV2/2CV277 VALVES ARE CLOSED.
N600 ADD MALFUNCTION TO CAUSE Y N Y Y AN ELECTRICAL TRIP OF THE N0.21, 22 SAFETY INJECTION PUMPS.
THE CONDITIONS WHICHCOULD CAUSE THE ELECTRICAL TRIP ARE:
- 1.
U6DV VITAL BUS DIFFERENTIAL
- 2.
4160V VITAL BUS OVERLOAD
- 3.
DA, DC GROUND OVERCURRENT (INFORMATION TAKEN FROM LOGIC DIAGRAM 239925-B-9648-25/2D/86)
NlOO ADD MALFUNCTION TO CAUSE Y N Y Y ELECTRICAL TRIPS OF THE NO. 21,22 CHARGING PUMPS.
THE CONDITIONS WHICH COULD CAUSE THEELECTRICAL TRIP ARE:
- 1.
4160V VITAL BUS DIFFERENTIAL
- 2.
4160V VITAL BUS OVERLOAD
- 3.
DA, DC & GROUND OVER CURRENT I I Given low priority during 1898 annual review. Will be implemented as time permits.
I I Given low priority during 1989 annual review.
~Jill be implemented as time permits.
I I Given low Priority during 1989 annual review. Will be implemented as time permits.
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S-FB-89-019 HOFFMAN, D.
03/20/89 TEST ENGINEER S-FB-89-020 HOFFMAN, D.
03/20/89 TEST ENGINEER S-FB-89-021 HOFFMAN, D.
04/04/89 TEST ENGINEER S-FB-89-023 NICHOLS, J. 05/06/89 INSTRUCTOR S-FB-89-026 HOFFMAN, D.
05/16/89 TEST ENGINEER RlOO R100 (INFORMATION TAKEN FROM LOGIC DIAGRAM 224414 B95834/17/86)
ADD NEW MALFUNCTION TO ALLOW FOR FAILURE OF PT-505 OVERITS FULL RANGE.
ADD NEW MALFUNCTION TO ALLOW FOR FAILURE OF PT-506 OVERITS FULL RANGE.
y N y y y N y y N210 ADD SIMULATOR MALFUNCTION Y N Y Y TO CAUSE A VARIABLE SIZE LEAKIN THE COMPONENT COOLING l4ATER MISCELLANEOUS HEADERRETURN LINE.
E150 REACTOR BUS XFER SWITCH N N N N IN CDIG FOR IRPI POWER SUPPLY TOIT'S ALTERNATE POWER F300 ADD MALFUNCTION 122, TO Y N Y Y FAIL LA 1018 (HTR DRN TANK 2A) LA1020 (HTR DRN TANK 2B) LA 1022 (HTR DRN TANK 2C)
LEVELCONTROLLERS.
THIS WILL CAUSE THE 21, 22 OR 23HD VALVE TOREPOSITION I I Given high oriority at 1989 ann1Jal review.
Scheduled forcompletion in 1990.
I I Given high priority at 1989 annual review.
Scheduled forcompletion in 1090.
I I Given low Priority during 1989 annual revie 1iJ. Wi 11 be implemented as loo' oime perm11.s.
I I I I Given low Priority during 1989 annual revie1J.
l~ill be implemented as time permits.
_________ J
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S-FB-89-039 HOFFMAN, D.
06/08/89 TEST ENGINEER R600 CHANGE MALFUNCTION 0 TO y N y y I I Given low ALLOW SPRAY VALVE TO BE priority during FAILEDFROM 0 - 100% VALVE 1989 annual POSITION.
review. Will be implemented as time permits.
S-FB-89-040 HOFFMAN, D.
06/08/89 TEST ENGINEER R600 CHANGE MALFUNCTION 0 TO y N y y I I Given low ALLOW SAFETY VALVE TO BE priority during FAILEDFROM 0 - 100% VALVE 1989 annual POSITION.
review. Will be implemented as time permits.
S-FB-89-042 HOFFMAN, D.
06/13/89 TEST OPERATOR N210 ADD NEW SIMULATOR y N y y I I Given low MALFUNCTIONS TO CAUSE priority during OUTLEAKAGE FROMVARIOUS 1989 annual POINTS IN THE CCW SSYTEM.
review. Will be THIS ~JILL FOR implemented as MORETRAINING FLEXIBILITY time permits.
IN AOP-CC-1 USE.
S-FB-89-043 HOFFMAN, D.
06/13/89 TEST ENGINEER R500 ADD P/A CONVERTER TO y y y y I I Given high ENABLE TRAINING ON priority at 1989 RESETTING annual review.
DURINGPERFORMANCE OF Scheduled AOP-ROD-2, -3 AND -4 forcomoletion in 1990.
S-FB-89-044 HOFFMAN, D.
06/13/89 TEST OPERATOR N100 ADD CV83, 89 AND 95 TO Y N N Y I I Given low ALLOW FOR SLOWY RE Priority during ESTABLISHINGSEAL 1989 annual INJECTION FLOW.
review. !.Jill be ACTUALLY, WHAT IS NEEDED implemented as IS THEABILITY TO DECREASE time permits.
SEAL WATER OUTLET TEMPERATURE BY loFPER MINUTE OR LESS S-FB-89-046 NICHOLS, J. 06/09/89 INSTRUCTOR R200 PLEASE MAKE RCP SEAL y N y y I I Given low FAILURES VARIABLE.
priority during 1989 annual review.
~Ji 11 be implemented as time permits.
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S-FB-89-047 NICHOLS, J.
06/09/89 INSTRUCTOR S-FB-89-048 NICHOLS, J.
06/09/89 INSTRUCTOR S-FB-89-052 HOFFMAN, D.
06/23/89 TEST OPERATOR S-FB-89-055 MOORE, K.
07/24/89 SIM. INSTRUCTOR S-FB-89-056 MOORE, K.
01/19/89 SIM. INSTRUCTOR S-FB-89-058 NICHOLS, J.
09/18/89 INST.
S-FB-89-060 LOONSBURY, 10/03/89 NSS D.
N600 STARTING SI PUMP WITH SUCTION VALVE CLOSED -
PUMP SHOWSNO CAVITATION AND THE SAME AMPS AS A PUMP ON RECIRC.
IAOO PLEASE MAKE ALL STEAM LINE MALFUNCTIONS (BREAKS) VARIABLE.
M600 ADD MALFUNCTION TO CAUSE LEAK IN SERVICE WATER TURBINEHEADER, TO SUPPORT AOP-SW-2.
TlOO WOULD LIKE AN ADDITIONAL EHC MALFUNCTION. EHC AUX SPEEDCHANNEL FAILURE.
y N y y y N y y Y N N N y N y y R400 IT WOULD BE USEFUL FOR US Y N Y Y TO HAVE A MALFUNCTION THATWOULD PREVENT THE SR DETECTORS FROM AUTO.
ENTER,FOLLOWING A RX.
TRIP IAOO WOULD LIKE A PROTECTED N N N N I&C FOR BOLD SID RODS OUT XENONFREE WITH CRITICAL RED POSITION OF CAT 58 STEPS E500 GENERATOR CONDITION y y y y MONITOR IN UNIT #2 CONTROL ROOM !SNOT IN I I I I I I I I I I I I Given high Priority at 1989 snnual review.
Scheduled forcompletion in 1990.
Given high priority at 1989 annual revie1..i.
Scheduled forcompletion in 1990.
Given low priority during 1989 annual revie11. Wi 11 be implemented as time permits.
Given low priority during 1989 annual review. Will be implemented as time permits.
FIXED IN CONJUNCTION WITH DR-88-0~6.
Given 1011 priority during 1989 annual
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SIMULA TOR.
THE GOULD COMPUTER THAT RUNS SPDS ALSORUNS THIS MONITOR.
PLEASE INVESTIGATE THE POSSIBILITY OFUTILIZING THE GOULD COMPUTER IN THE SIMULATOR TO ALSO RUNTHIS MONITOR.
S-FB-89-061 LOONSBURY, 10/03/89 NSS HWOO SIMULATOR DOES NOT HAVE A N N N N RADIO, AS DOES BOTH D.
S-FB-89-062 MOORE, K.
S-FB-89-063 MOORE, K.
S-FB-89-06~ MOORE, K.
S-FB-89-065 MOORE, K.
CONTROLROOMS.
IN THE PLANT RADIOS ARE USED FREQUENTLY BY FIELDOPERATORS.
10/10/89 LEAD SIM. INSTR.
0000 WOULD IT BE POSSIBLE TO N N N N SOMEHOW ENCASE OR REDUCE THENUMBER OF WIRES/CABLES 10/10/89 INSTRUCTOR THAT EXISTS UNDER THE TABLETOP ONTHE SIMULATOR FLOOR.
I BELIEVE THIS IS A SAFETY HAZARD.
R500 Ai THE OPS MANAGER'S Y Y Y Y REQUEST, IT WOULD ENHANCE SIMULATORTRAINING IF THE ROD CONTROL P/A CONVERTER WAS INSTALLEDIN THE EQUIPMENT ROOM. A MOCK-UP, AT A MINIMUM, l~OULDSUFFICE.
10/10/89 LEAD SIM. INSTR.
P250 IT WOULD BE BENEFICIAL TO Y N Y N HAVE THE P-250 VISUAL 10/ 10/ 89 INST.
TRENDPOINT TO BE I.C.
DEPENDENT.
C310 THE FOLLOWING AUX ALARMS N N N N SHOULD BE IN.
SINCE ONLY ONE OFTWO BATTERY CHARGES ARE IN SERVICE.
55, 58, 59, 17~,280, 319, 02 &
918 review. Will be implemented as time permits.
RADIO INSTALLED, NOT YET OPERATIONAL.
I I Given high Priority at 1989 annual review.
Scheduled forcompletion in 1990.
I I Given low Priority during 1989 annual review. Will be implemented as time permits.
I I
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S-FB-89-067 MOORE, K.
10/18/89 LEAD SIM. INSTR.
N510 ADD MALFUNCTION FOR GDT Y N Y N I I LEAK (VARIABLE) IN AUX.
BLDG.
S-FB-89-068 MOORE, K.
10/30/89 LEAD SIM. INSTR.
IAOO NEED CANA FOR 21-22 Y N Y N I I Given high HDIG'S (HDP OUTLET V'S) orioritY during TO COMPLYWITH PLANT 1989 annual PHILOSPHY OF THROTTLING review. Scheduled DISCH. VALVES for completion in 1990.
S-FB-89-069 NICHOLS, J.
10/31/89 INSTRUCTOR F300 NEED CANA FOR HOOD SPRAY y N y y I I Given low BYPASS VALVE PER IOP-4/5.
Priority during 1989 annual review.
l~ill be implemented as time permits.
S-FB-89-070 MOORE, K.
11/01/89 LEAD. srn. INSTR. IAOO ADD MONITOR VARIABLE y N y y I I Given 10 1*
POINTS FOR RHR SYSTEM priority during BORON SO WECAN SAMPLE PER 1989 annual IOP-5.
review. Wi 11 be implemented as time permits.
S-FB-89-071 MOORE, K.
11/01/89 LEAD SIM. INSTR.
G220 ADD CANA OR MALFUNCTION y N y y I I Given low TO ALLOW MODELING OF priority during SECONDARYPLANT STEAM 1989 annual LEAKAGE (VARIABLE review.
~Jill be SEVERITY OF 0 - 10%).
implemented as PREFER IT TO BE INSIDE time permits.
TURB BLDG WITH DYNAMIC RESPONSEFROM TURB SUMP /PUMPS, ETC.
S-FB-89-072 MOORE, K.
11/01/89 LEAD SIM. INSTR.
T900 PROVIDE CANA'S FOR TAC TO Y N Y Y I I Given loil H2 COOLERS SO WE Priority during CANADJUST/CLOSE PER IOP &
1989 annual TURB S/D PROCEDURE review. Will be implemented as time permits.
S-FB-89-073 MOORE, K.
11/03/89 LEAD SIM. INSTR.
F300 NO MALFUNCTION y N y y I I Given low DESCRIPTION EXISTS FOR priority during FAILURE OF 26 HTRLVL 1989 annual
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S-FB-89-074 MOORE, K.
CHANNEL 11/03/89 LEAD SIM. INSTR.
M600 ADD MALFUNCTION TO FAIL Y N Y Y PRESS SENSOR CONTROLLING 2ST1.
S-FB-89-075 MOORE, K.
11/03/89 LEAD SIM. INSTR.
IAOO NEED DMON'S FOR FEEDWATER Y N Y Y S-FB-89-076 MOORE, K.
5-FB-89-078 MOORE, K.
S-FB-89-079 MOORE, K.
FLOW IN INCHES /\\P TO PERFORMMANUAL CALORMETRIC 11/03/89 LEAD SIM. INSTR.
E570 CANNOT VENT MAIN GEN TO Y N Y Y REDUCE GEN GAS PRESS ( 60 PSIGPER IOP-3 11/13/89 LD SIM. INSTR.
R400 WE NEED A CDIG THAT Y N Y Y 11/27/89 INSTRUCTOR ALLOWS US TO TRIP IR HI FLUX TRIPBISTABLE.
IAOO NEED THE CAPABILITY TO Y N Y Y LOCALLY OPERATE MOTOR-OPERATEDSAFEGUARDS VALVES (I.E., SJ4,5,12,13
& C\\168, 69)
FROMINSTRUCTOR AIDS UNDER CDIG OR CANA REASON:
AFTER SI WAS RESET A BLACKOUT OCCURRED IN WHICHB HAS FAILED TO ENERGIZE.
iHE EOP HAD NORMAL CHARGINGREESTABLISHED BUT CV68 MUST BE LOCALLY I
I I
I I
review.
\\~ill be implemented as time Permits.
I Given low Priority during 1989 nannual review. Wi 11 be implemented as time permits.
I Given low Priority during 1989 annual review. Will be implemented as time permits.
I Given low oriority during 1989 annual review.
~Ji 11 be implemented as time permits.
I Given low priority during 1989 annual review. wi il be implemented as time Permits.
I Given high priority at 1989 annual reviei.1*
Scheduled forcompletion in 1990.
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OPENED.
I WASUNABLE TO DO THIS.
CHARGING COULD NOT BE ESTABLISHEDWITHOUT REENERGIZING B I/ITAL BUS.
S-FB-90-002 MOORE, C.
12/26/89 LD SIM. INSTR.
0000 THE SIMULATOR NEEDS TO N Y N Y I I INSTALL BATTERY OPERATED LIGHTSABOl/E RP-8.
THIS WILL PROVIDE THE OPERATORS WITHADEQUATE LIGHTING DURING A BLACKOUT SCENARIO.
S-FB-90-003 CURRAN, L 01/02/90 OPS ENGINEER NlOO DURING SCENARIO OF y N y y I I BLACKOUT WITH NO SI OR CHARGING ORCOMPONENT COOLING, INTACT RCP SEAL PACKAGES TEMPERATURERESPONDED EXTREMELY SLOWLY, ESPECIALLY SINCE CV104 VALVESWERE OPEN PROVIDING FLOW PATH THRU SEAL RETURN.
RELIEFVALVE
( 2CV115) TO PRT.
m1P SHOT OFFSCALE HIGH WHENCl/985 WERE CLOSED.
S-FB-90-006 MOORE, C.
01/03/90 SIM. INSTR.
E300 WHEN THE EDG's ARE N N Y Y I I SECURED iHE FREQUENCY METER FOR ALLTKE EDG BEZELS SHOULD READ NEAR 60 HZ, NOT PEGGED LOW.
ICALLED UNIT 2 CONTROL ROOM AND THE FOLLOWING ARE CURRENTREADINGS:
2A 28 2C 59.9 60.2 60.~
S-FB-90-007 SAMPSON, R.
01/16/90 SIM. INSTR.
M600 DURING MODE III OPS. (SI YfJ y '(
I I
& BLACKOUT) BOTH SW122's SHUT. THE EOP's REQUIRE THEM TO BE RE-OPENED
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AFTER SI IS RESET. AS OF NOW, THE 122's OPEN AS SOON AS THE SI IS RESETWITH NO OPERATOR ACTION.
PLEASE MODIFY THE CDIG TOINCLUDE CONTROL FOR THE SW122's.
S-FB-90-008 MOORE, C.
01/17/90 LEAD SIM. INSTR.
0000 OD-12-RELATED TECH.
SPECS. DO NOT HAVE THECROSS-REFERENCE NOTE ON THE DOCUMENT SLEEVE.
N N N N I I S-FB-90-010 MOORE, C.
01/18/90 LEAD SIM. INSTR.
HWOO LABLE FOR LP FW HTR INLET N Y N Y
/ /
VALVES IS INCORRECT SHOULDREAD.
21CN27 21CN22 21A-22A OPEN 23A-25A OPEN 21CN27 21CN27 21A-22A CLOSED 21A-22A CLSD S-FB-90-011 MOORE, C.
02/05/90 LEAD SIM. INSTR.
IAOO SNAPSHOT NEEDS TO BE N N N N S-FB-90-012 LOUNSBURY, D 02/05/90 SRO REPONE TO INCLUDE:
x 22 CN~8 CLOSED x 22 CC PUMP ---} AUTO.
x 23 SW PUMP ---} AUTO.
x PUMP ROOM COOLERS ---)
AUTO.
0000 NEED TO INSTALL A SECOND N Y N Y PHONE BY SPDS FOR USE BY STATIONAND SHIFT SUPERVISOR WHEN BOARD NCO PHONE IS TIED UP - ORWHEN BOTH NCO's ARE ON THE PANEL AND PHONE CALLS NEED TOBE MADE FROM STA I I I I
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S-FB-90-013 LOUNSBURY, D 02/05/90 SRO S-FB-90-014 NICHOLS, J. 02/12/90 SIM. INSTR.
S-FB-90-015 NICHOLS, J. 02/12/90 SIM. INSTR.
S-FB-90-016 PETERSON, K. 02/27/90 NCO S-FB-90-017 HOUSE, ALEX 03/14/90 SNTS S-FB-90-018 FOEHNER, A.
06/04/90 TEST ENGINEER I l
~---
STAiION.
DODD WHEN PERFORMING EXERCISES N N N N ON THE SIMULATOR A COPY OF THEIOP IN EFFECT SHOULD BE MADE FOR THE BOARD NOC WliH THEAPPLICABLE STEPS SIGNED OFF.
THIS WOULD LEVEL THEPROCEDURES IN THE RACK AVAILABLE FOR THE SRO-SSS TOREVIEW AND MAKE IT LESS CUMBERSON FOR NCO.
R200 INCREASE THE SIZE OF RCS y N y y LEAK INSIDE CONTAINMENT TO 2000GPM TO ENCOMPASS FULL RANGE OF FSAR BREAK SIZES.
NlOO MOVE LETDOWN LINE LEAK Y N Y N FROM INSIDE CONTAINMENT TO OUTSIDECONTAINMENT TO ALLOW USE OF EOP-LOCA-5.
KEEP LEAKAGERATE THE SAME SIZE.
F400 AFW PUMP LOW SUCTION N N N N PRESSURE TRIP.
WHEN ENERGIZING THETRIP, YOU SHOULD GET AN ALARM ON AUX ALARM TYPEWRITER.
THIS DOESN'T HAPPEN.
F310 PLEASE PROVIDE A DMAL TO y N y y FAIL A CONTROLLING CHANNEL FORSTEAM PRESSURE.
PRESSURECHANNEL WE CAN FAIL WILLNOT EFFECT SWWLC.
R600 CHANGE LR-D CHANGES 2PS1 y N y y
& 2PS3 STROKE TIMES TO I I I I I I I I I I I I
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45SECONDS FROM 10 SECONDS.
PER TELECOM WITH C. LASHKARI OFPLANT ENGINEERING VALVES NOW STROKE AT 21 SEC.
PLEASECHANGE STROKE TIME ON SIMULATOR.
S-FB-90-019 NICHOLS, J. 06/13/90 SIM. INSTR.
F300 ADD CANA TO VARY INTEGRAL Y N Y Y I I VALUE OF LEVEL ERROR/FLOW ERRORPI CONTROLLER FOR BF19s FROM 200 TO INFINiTE SECONDS S-FB-90-020 NICHOLS, J. 06/13/90 SIM. INSTR.
F300 ADD CANA TO INCREASE
'I N Y N I I INSTRUMENT NOISE ON SG STEAM ANDFEED FLOW DURING STARTUP.
S-FB-90-021 NICHOLS, J.
06/15/90 SIM. INSTR.
RrnD POWER RANGE NIS GAIN POTS Y N Y Y I I SHOULD BE FUNCTIONAL TO ALLOWADJUSTMENT WHEN NECESSARY.
PRESENTLY, NIS READ 2% HIGHERTHAN CALIROMETRIC AND SHOULD BE ADJUSTED.
S-FB-90-022 NICHOLS, J. 07/06/90 SIM. INSTR.
E140 NEED MALFUNCTION TO TRIP y N y y OPEN GROUP & VITAL 4KV
--) 480VOLT TRANSFORMER IN FEED BREAKERS TO SIMULATE LOSS OF 480V iRANSFORMER S-FB-90-023 NICHOLS, J. 07/06/90 SIM. INSTR.
E153 NEED CDIG TO ALLOW TOGGLE Y N Y Y BETWEEN REGULAR AND BACK-UPPOWER SUPPLY FOR 480 VOLT.
AND MAC PANELS.
CROSS TIE BETWEEN 23 MAC 'E' 460 REG/'G' 460 B/U 2G - 2E 460 I/, INTER-
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24 MAC 'F' 460 REG/'H' 460 B/U LOCKED WITH INFEEDS.
, Intentionally left blank