IR 05000272/1991014
| ML18096A180 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/17/1991 |
| From: | Eselgroth P, David Silk NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18096A179 | List: |
| References | |
| 50-272-91-14-OL, 50-311-91-14-OL, NUDOCS 9108060227 | |
| Download: ML18096A180 (68) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING REQUALIFICATION PROGRAM EVALUATION REPORT REQUALIFICATION PROGRAM EVALUATION REPORT NO /91-14 (OL-RQ)
50-311/91~14 (OL-RQ)
FACILITY DOCKET NO FACILITY LICENSE NO LICENSEE:
FACILITY:
EXAMINATION DATES:
CHIEF EXAMINER:
50-272/311 DPR-70/75 Public Service Electric and Gas Company 244 Chestnut Street Salem N. Salem Generating Station Units 1 and 2 June 11 - 13, 1991 David Silk, Operations Engineer
~D'ate PWR Section, Operations Branch, DRS APPROVED BY:
/Oat SUMMARY:
The licensed operator requalification training program was rated as satisfac-tor The evaluation was conducted in accordance with Revision 6 to NUREG-1021; Operator Licensing Examiner Standard Requalification examinations were administered to six senior reactor operators (SROs) and six reactor operators (ROs).
The examinations were graded concurrently and independently by the NRC and the facility training staf As graded by the NRC and the facility, ten of twelve operators passed all portions of the examination and two of three crews were evaluated as satisfactory during the dynamic simulator examinatio PDR ADOCK 05000272 V
I DETAILS TYPE OF EXAMINATIONS:
Requalification EXAMINATION RESULTS:
NRC RO I
SRO TOTAL Grading Pass/Fail I Pass/Fail Pass/Fail I
I Written 6/0 I
6/0 12/0 I
I Simulator 5/1 I
5/1 10/2 I
I Walk-through 6/0 I
6/0 12/0 I
I Overa 11 5/1 I
5/1 10/2 I
Facility RO I
SRO I
TOTAL Grading Pass/Fail I Pass/Fail I Pass/Fail I
I I
I Written 6/0 I
6/0 I
12/0 I
I I
I Simulator 5/1 I
5/1 I
10/2 I
I I
I Walk-through 6/0 I
6/0 I
12/0 I
I I
I Overall 5/1 I
5/1 I
10/2 I
I 1. 0 PERSONNEL CONTACTED DURING THE EXAMINATION/EVALUATION NRC PERSONNEL:
David Silk, Sr Operations Engineer (Chief Examiner) (1,2,3,4)
Tom Johnson, Senior Resident Inspector (4)
Tim Bardell (Pacific Northwest Laboratory - PNL)
(1,2)
Nancy Maguire-Moffitt (PNL)
(1,2)
PSE&G:
Craig Bersak, Sr. Nuclear Instructor/Supervisor Benson Binggeli, Sr. Nuclear Instructor/Supervisor
. P.atrick Casey, Evaluator Larry Curran, Operating Engineer/Salem Station Frank Johnson, Training Marios Kafanataris, Nuclear Shift Supervisor J. K. Lloyd, Principal Training Supervisor Greg Mecchi, Principal Trainer - Ops Training Brian O'Brien, Evaluator A. Orticelle, Manager - Nuclear Training V. J. Polizzi, Operations Manager Mike Reese, Sr Training Instructor/Supervisor (1,2,3,4)
(1,2,3,4)
(2)
(4)
(1,2,4)
(1,2,3,4)
(2,3,4)
(3,4)
(2)
(3,4)
(1,2,4)
(2,4)
LEGEND:
(1) Participated in examination development (2) Participated in examination administration (3) Attended entrance meeting on Monday June 10, 1991 (4) Attended exit meeting on Friday June 14, 1991 PROGRAM EVALUATION RESULTS The program for licensed operator requalification training at the Salem Generating Station Units 1 and 2 was rated as satisfactory in accordance with the criteria established in NUREG-1021, ES-60 Those criteria are: The facility grading must be as conservative as the NRC grading on at least 90% of the pass/fail decisions for each section of the examinatio NRC and facility grading determined that all twelve operators passed the written examination, thus satisfying Criterion NRC and facility evaluators determined that ten operators passed the operating examination, thus also satisfying Criterion At least 75% of all operators pass the examinatio NRC grading is the only consideration for this criterio Ten operators passed the examination overall (83%), thus meeting Crite-rion Failure of. no more than one-third of the crews during the simulator portion of the operating examinatio NRC grading is the only consideration for this criterio Only one of the three crews failed thus meeting Criterion There are six other criteria stated in ES-6IT1 of which if the facility has three or more applicable to the program, then that program shall be determined to be unsatisfactor The Salem requalification program met none of the following criterja: The facility evaluators do not concur with the NRC evaluators on all UNSATISFACTORY crew evaluations The facility failed to train and evaluate operators in all positions permitted by their individual licenses More than one facility evaluator is determined to be unsatisfactory A lack of administrative controls to preclude an RO or SRO from performing licensed duties without satisfying the requirements of 10 CFR 55.53 to restore.the license to active status A lack of quality control of the facility's examination bank The facility's failure rate is excessive relative to the NRC's failure rate SIMULATOR SCENARIOS The following items were noted regarding the dynamic simulator portion of the operating examination This information is being provided to aid the licensee in upgrading licensed operator and operator requalification training program For the simulator examinations, the NRC requested that the Shift Technical Advisor (STA) not be ~resent in the control room until after the reactor tri The NRC rationale was that, since the STA's ISCTs (Individual Simulator Critical Tasks) were all post-trip, the absence of the STA minimized the possibility of him unintentionally intercepting other crew members' ISCTs and, thus, necessitate running more scenarios to ensure all individuals were evaluated on two ISCT The NRC recognized that this crew configuration was not how the facility trained the crews but for evaluation purposes it was decided to remove the STA from the control roo The only observable impact of the absence of the STA prior to the reactor trip on the crew was that the one SRO in the control room had to evaluate Technical Specification In one instance, an SRO entered the incorrect action statement for RCS activity exceeding the limit; and, for the same condition, another SRO (from a different crew) wavered until finally choosing the correct action statemen This demonstrated indi-vidual weaknesses of SROs to evaluate infrequently entered Technical Specification Another request of the NRC was to disable the SPDS (Safety Parameter Display System) since it monitors and displays the status tree Disabling the SPDS assisted in the evaluation the STAs' abilities to
monitor the status tree In four of the six scenarios, the STAs correctly monitored the status tree In two scenarios (different crews but the same scenario), the STAs did not identify a purple path (i.e.,
orange path in standard Westinghouse terminology) regarding possible RCS thermal shoc Following a large break LOCA, due to RCS cold leg temp-eratures and RCS pressure, entry conditions were met for the functional recovery procedure addressing thermal shoc The reason that the NRC did not consider these individuals unsatisfactory was that given the plant condition (i.e., post-LOCA), there were no adverse safety significant con-sequences for not entering that procedure since repressurization of the RCS was an impossibilit The NRC did, however, downgrade those crews on procedural implementation for not identifying the purple path Thus, disabling the SPDS revealed weaknesses in the STAs 1 abilities to monitor the status trees in some instance Crew communications were, in some instances, barely audible or muffled during scenario drill Repeatbacks were not always stated and had contributed to the failure of a crew to properly perform a swapover to cold leg recirculation following a LOC There was improved definition of operator responsibilities inside and outside of the 11 horse shoe 11 area from the last requalification examina-tio This expedited crew implementation of the emergency operating procedure During the previous requalification examination, the SROs would sometimes shadow the board operator as he performed manipulations while the desk operator read the Emergency Operating Procedures (EOPs).
Since the SRO is now the EDP proGedure reader after completion of the immediate actions of EOP-TRIP-1 (Reactor Trip or Safety Injection), the SRO would, therefore, be away from the controls and could better maintain a broad perspective of plant status during emergency evolution The reporting of equipment status was much more consistent in this exam-ination than the last on During the last examination, the operators would report the same equipment status in many different ways (e.g., pump status - secured, not running, out of service).
During the evaluation of the third crew, it was noted that a procedural inconsistency existed within the Abnormal Operating Procedures (AOPs)
regarding Reactor Coolant Pump (RCP) bearing temperatures (i.e., RCP trip criteria at 175°F versus 185°F).
The third crew met the entry conditions and implemented an AOP that was not expected by the facilit This AOP was more conservative than the intended ADP and directed the crew to trip the reactor, This action by the operators disrupted the evaluation worth of the scenario; and, thus, it was stopped and a substitute scenario was use The licensee initiated an investigation into the procedural dis-crepancy and determined that the alarm setpoint should be at 175°F and the RCP trip criteria should be 185° Actions taken regarding the AOPs will be done under the Procedures Upgrade Progra.0 WRITTEN EXAMINATION The following paragraphs contain items that were noted from various phases of the written examinatio This information is being provided to aid the licensee in upgrading licensed operator and operator requalification training program During the preexamination visit, the NRC noted that there were several questions in the Section B portion that belonged in Section Section A questions are to be designed to evaluate an operator's knowledge of plant systems, integrated plant operations, instrument~
and controls and the recognition of Technical Specification Limiting Conditions for Operatio Section B questions are to be designed to test an operator's knowledge of and use of plant procedures and administrative control The licensee incorporated the NRC's comments and modified the written examination by transferring ques-tions to the proper section and adding or deleting questions as necessary prior to examination administratio There were no grading discrepancies between the facility and the NR Based on the grading of the written examinations, the following deficiencies were note (A deficiency is considered to exist when one-third or more of the licensed operators missed the same question).
Bank Number (SECTION A)
STATE-05 6 (Static - Emergency)
STATA-01 9 (Static - Abnormal)
STATA-01 15 STATA-01 16 Knowledge Conditions under which a ruptured steam generator is considered to be isolated Control rod speed based on temperature error when transferring from manual to automatic Method of RCS depressurization for a steam generator tube rupture following an event in which a pressurizer spray valve failed and stuck open Volume Control Tank (VCT) level response following a VCT level indica-tion failure and makeup controls subsequently placed in manual
(SECTION B)
NRC-SF02 14 NRC-SF03 20
Determine when saturation will be reached in the RCS for a given set of conditions during Mode 5 operations Identify the cond1tion following a reactor trip and safety injection when ECCS equipment is required PLANT WALK-THROUGH (JOB PERFORMANCE MEASURES)
The following items were noted regarding the Job Performance Measure (JPM)
portion of the operating examination This information is being provided to aid the licensee in upgrading licensed operator and operator requal-ification training program During the preexamination visit, the NRC walked down and reviewed all of the JPMs and question Several changes needed to be mad Designated critical steps in some of the JPMs were scrutinized (as well as other steps within the tasks) to accurately identify steps that were indeed critica The criteria in ES-603, paragraph C. (4), was used to determine criticaJ step Several questions in the JPMs required minor wording changes to add clarit The facility evaluators generally performed well during JPM adminis-tratio There were several instances where evaluator responses could have been construed as 11positive feedback 11 (e.g., nodding the head or saying 11 0K 11 ).
Evaluators need to be more conscientious about maintaining a 11 poker face 11 while interacting with the operator Also, after hearing the operator answer a question, the evaluators were to paraphrase the operator 1s answer to ensure that the evaluator understood the operator 1 s answe There were several instances where the evaluators more closely paraphrased the answer key answer than the operator's answe In the instances where this was observed, the operator 1 s response would have been acceptabl To prevent this paraphrasing from providing an inappropriate cue, evaluators should exercise caution to accurately articulate operator response During the JPM administration, the operators performed the JPMs well, and no generic performance deficiencies were note The following items are instances where the NRC observed incorrect responses or operator difficulty in clearly articulating the correct respons Question Number 118 501 05 01 Jl 01 Knowledge Cause of pressurizer level stabilization following an RCS depressurization to refill the pressurizer
Question Number 010 514 04 01 Jl 01 010 514 04 01 Jl 02
Knowledge Knowledge of heat sink availability criteria for terminating safety injection flow Basis for tripping the RCP as per the Continuous Action Summary for EOP-TRIP-1 The overall comparison of NRC and facility grading of operator responses to JPM questions only differed on four (out of 120) question On three of the four questions, the facility grading was more restrictive than that of the NR.0 SUMMARY OF COMMENTS MADE AT EXIT MEETING ON JUNE 14, 1991 Appreciation was expressed for the cooperation of all those involved in the examination proces The NRC communicated to the licensee the contents of Sections 2.0 through 5.0 ?f this repor The NRC stated that its preliminary evaluation of the facility's requalification program was satisfactory and reiterated the numbers of unsatisfactory individuals and crew.0 FACILITY COMMENTS The facility stated the corrective actions it would be taking regarding the unsatisfactory crew and individual Namely, the crew will be removed from licensed dutie Remedial training would begin the following week with emphasis on communications and team wor The individuals would then be reevaluated at the appropriate tim Based on other observations from the examination week, the facility would be providing EOP Status Tree and STA retraining for two individuals who demonstrated weakness in those area Also, the facility will be emphasizing infrequently used Technical Specifications in future li*censed operator requalification trainin ATTACHMENTS Simulator Examinations (summary page only) Job Performance Measures (cover sheet only) Written Examination and Answer Key Licensee Report of the Requalification Examination Simulator Fidelity Report
ATTACHMENT 1 SIMULATOR EXAMINATIONS ESG-002 Total Loss of Feedwater ESG-011 Large-Break LOCA ESG-001 Total Loss of Feedwater (For third crew in lieu of ESG-002)
SALEM SIMULATOR EXAMINATION SCENARIO SCENARIO TITLE
- TOTAL LOSS OF FEEDWATER SCENARIO NUMBER
- ESG -
002 EFFECTIVE DATE NOVEMBER 2, 1990 SCENARIO DURATION: 1 HOUR REVISION NUMBER
PROGRAM PREPARED BY:
REVISED BY:
APPROVED BY:
APPROVED BY:
.. -=X__ LO REQUAL
--~ INITIAL LICENSE
___ OTHER J. ELLIS (INSTRUCTOR)
C. BERSAK (INSTRUCTOR)
(SALEM TRAINING SUPERVISOR)
/1/fo.2---j.f.c llf (SAL~ERATIONS MANAGER)
PAGE 1 OF 16 ESG-002 11[02L9o (DATE)
05[30[91
{DATE)
{DATE)
rill1h1
I (DATE)
ESG-002 MAJOR EVENTS A. 22SW122 AUTO CONTROL FAILS B. SGFP SILENT TRIP DUE TO LINKAGE FAILURE C. MAIN TURBINE OVERSPEED (GRID FREQUENCY TRANSIENT)
D. *TOTAL LOSS OF AUX. FEEDWATER I SCENARIO SUMMARY A. THE CREW WILL RECEIVE THE TURNOVER AT 85% POWER, WITH SEVERAL PIECES OF EQUIPMENT OUT OF SERVIC TWO WILL HAVE AN IMPACT ON THE SCENARIO; 23 AFWP AND 21 CCW HEAT EXCHANGE POWER IS BEING INCREASED AFTER A VAR ALAR THE VAR ALARM IS CLEA THE FUEL IS CONDITIONED TO FULL POWE B. AFTER THE TURNOVER 22SW122 AUTO CONTROL FAILS HIGH, CAUSING CCW TEMPERATURE TO INCREAS C. ONCE 22SW122 IS RE-OPENED IN LOCAL/MANUAL, 21 SGFP LINKAGE FAILS, CAUSING 21 SGFP TO RUN BACK TO 0 RP D. ONCE THE CORRECTIVE ACTIONS HAVE BEEN PERFORMED, A GRID FREQUENCY TRANSIENT CAUSES THE MAIN TURBINE TO OVERSPEED, BUT THE TURBINE FAILS TO TRI A MANUAL TURBINE TRIP WILL BE SUCCESSFU E. DURING THE REACTOR TRIP RESPONSE, 21 & 22 AFWP'S TRI FRHS-1 WILL BE ENTERED AND THE SCENARIO WILL TERMINATE WHEN FEEDWATER FLOW IS RESTORE F. CONTROL RODS IN MANUA G. SPDS TURNED OF PAGE 2 OF 16
SCENARIO TITLE
..
SALEM SIMULATOR EXAMINATION SCENARIO LARGE-BREAK LOCA SCENARIO NUMBER ESG -
011 EFFECTIVE DATE NOVEMBER 2, 1990 SCENARIO DURATION:
1 HOUR REVISION NUMBER PROGRAM PREPARED BY:
REVISED BY:
APPROVED BY:
APPROVED BY:
..
..
x LO REQUAL INITIAL LICENSE OTHER R. SAMPSON l J. ELLIS (INSTRUCTOR)
C BERSAK (INSTRUCTOR)
(SAL~INING SUPERVISOR)
JM ~~11 (SALEM OPERATIONS MANAGER)
PAGE 1 OF 29 ESG -
011 11£02£90 (DATE)
05£30£91 (DATE)
(DATE)
tflo'/JtE J
--- MAJOR EVENTS A. "B" SEC FAILURE B. HIGH RCS ACTIVITY C. SPEED REFERENCE CHANNEL FAILS ON MAIN TURB D. "LARGE-BREAK LOCA I SCENARIO SUMMARY ESG -
011 A. THE CREW WILL RECEIVE A SHIFT TURNOVER WITH THE UNIT AT 100%
POWER FOR THE LAST 30 DAY A 1 GPD TUBE LEAK WAS IDENTIFIED IN 21 S/G 20 DAYS AGO AND HAS NOT CHANGE D4 ROD DROPPED INTO THE CORE 2 HOURS AGO WHEN ITS STATIONARY GRIPPER COIL FUSE FAILE THE FUSE WAS REPLACED AND THE ROD HAS BEEN RESTORED TO ITS NORMAL POSITIO THE ACTIONS OF AOP-ROD-3 HAVE JUST BEEN COMPLETED AND ALL APPLICABLE ACTION STATEMENTS HAVE BEEN EXITE SHORTLY AFTER SHIFT TURNOVER, 2B SEC WILL FAI THE SRO WILL SEND AN OPERATOR TO THE SEC CABINETS TO DETERMINE THE CAUSE OF THE FAILUR A/S 3.3.2.1 ACTION #13 IS ENTERE THE RCS SPECIFIC ACTIVITY BEGINS TO SLOWLY INCREAS THE CREW ENTERS AOP-RAD-3 & AOP-RCS-2 AND REQUESTS CHEMISTRY TO DP~~W ~
RX COOLANT SAMPLE AND ANALYZE FOR DOSE EQUIVALENT I-131 AND GROSS BETA-GAMM CHEMISTRY REQUESTS THE OPERATORS TO MAXIMIZE LD FLO CHEMISTRY SAMPLE RESULTS OF 75 MICROCURIES/CC DOSE-EQUIVALENT I-131 FORCES A POWER REDUCTION IAW TECH. SPEC DURING THE POWER REDUCTION A FAILURE OF THE SPEED REFERENCE CHANNEL REQUIRES POWER TO BE REDUCED USING TURBINE MANUA SHORTLY AFTER THE POWER REDUCTION IS RE-INITIATED, A LARGE-BREAK LOCA OCCURS ON 21 COLD LE THE REACTOR TRIPS AND AN SI IS ACTUATE THE CREW ENTERS TRIP-1 AND PERFORMS THE REQUIRED ACTION DIAGNOSTICS INDICATE A LOCA INSIDE CONTAINMEN THE OPERATORS TRANSITION TO LOCA-1 TO TAKE ACTIONS FOR THE LOC THE SCENARIO WILL TERMINATE FOLLOWING COMPLETION OF THE TRANSITION TO COLD LEG RECIRCULATION IAW EOP-LOCA-SPDS OUT OF SERVIC PAGE 2 OF 29
ESG-001 SALEM SIMULATOR EXAMINATION SCENARIO SCENARIO TITLE
- TOTAL LOSS OF FEEDWATER SCENARIO NUMBER ESG -
001 EFFECTIVE DATE SCENARIO DURATION: 1 HOUR REVISION NUMBER
- 5 PROGRAM
-=X__ LO REQUAL INITIAL LICENSE
___ OTHER PREPARED BY:
J. ELLIS 10[16[89 (INSTRUCTOR)
(DATE)
REVISED BY:
C. Bersak 04[22[91 (INSTRUCTOR)
(DATE)
APPROVED BY:
(SALEM TRAINING SUPERVISOR)
(DATE)
APPROVED BY:
( SAL<f//£o(;; !.AGER) -
PAGE 1 OF 16
ESG-001 MAJOR EVENTS A. 2A 4KV VB TRIPS ON DIFF. PROTECTION B. CONTINUOUS ROD INSERTION C. TOTAL LOSS OF AUX. FEEDWATER I SCENARIO SUMMARY A. THE CREW WILL RECEIVE THE TURNOVER AT 100% POWER, EOL WITH SEVERAL PIECES OF EQUIPMENT OUT OF SERVIC TWO WILL HAVE AN IMPACT ON THE SCENARIO, 23 AFWP AND 22 CHARGING PUM B. AFTER THE TURNOVER, 2A VITAL BUS IS LOST DUE TO DIFF PROTECTIO C. ONCE THE CORRECTIVE ACTIONS HAVE BEEN PERFORMED, A CONTINUOUS ROD INSERTION OCCURS FORCING THE CREW TO TRIP THE UNI D. DURING THE REACTOR TRIP RESPONSE THE REMAINING AFWP (22) TRIP FRHS-1 WILL BE ENTERED AND THE SCENARIO WILL TERMINATE WHEN FEEDWATER FLOW IS RESTORE PAGE 2 OF 16
Simulator JPMs ATTACHMENT 2 JOB PERFORMANCE MEASURES (Cover Sheets Only) (RO)
Perform a Reactor Startup Using Control Rods (SRO) Classify Emergency/Non-Emergency Events TCAF a Power Range Channel Malfunction TCAF a LOCA - Depressurize to Refill the Pressurizer TCAF a PORV Stuck Open Determine Required Hydrogen Recombiner Power Setting In-plant JPMs Startup Vital Instrument Inverter with Bypass Alternate in Service Startup a Diesel Generator Locally Reset the Steam Driven AFW Pump TCAF a Control Room Evacuation - Establish Charging from Outside the Control Room 10. Energize the Alternate Shutdown Panel
- TCAF - Take Corrective Action For
e Station:
Name_:
System:
SALEM OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Date Used:
ROD CONTROL Task, Task #,' and PERFORM A REACTOR STARTUP USING CONTROL RODS, Remote(R) or Control Room (CR): 001 507 01 01 Jl, (CR)
~ SRO 0 Importance Factor: 3.7 / ( RO ) I (-S--'R'--0-) -
Approval:
Grade:
Pr ncipal Trai ( perations/Si or designee SATISFACTORY Reason, if unsatisfactory:
O r
s Manager -
HC/SAL (or de ignee) as appropriate UNSATISFACTORY Evaluator's Signature=~~~~~~~~~~~~~-
DATE:
Location(s) of Test:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM: 20 MINS Start Time Stop Time References:
IOP-3
-***
Tools and Equipment:
NONE Comments:
Align simulator with Control Bank D approximately 100 steps below critical rod heigh Boron concentration may be varied to vary critical rod height and stay within ~ 500 pcm ban S-JPM/1:25 001 507 01 01 Jl Grade/Comments Approved:
NTC-207 DATE: 04/11/90 REV. :
e Station:
- Name: -
System:
SALEM OPERATOR TRAINING PROGRAM..
JOB PERFORMANCE MEASURE INVENTORY FORM Date Used:
EMERGENCY PLAN Task, Task #,_.and. CLASSIFY EMERGENCY/NON-EMERGENCY* EVENTS Remote(Rl or
- ~
Control Room (CR): Task #122 501 05 02 JlS EO D RO D SRO El K/A Number: 194001 Al.16 Importance Factor: 3.1/ (RO)/(SRO)
Approval:
Principal Training Supervisor Operations Manager -
HC/SAL (Operations/Simulator)
(or designee) as or designee appropriate Grade:
SATISFACTORY UNSATISFACTORY Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM:
10 min. Start Time Stop Time References:
ECG Tools and Equipment:
None Comments:
Grade/Comments Approved:
305S-JPM/1:25 NTC-207 122 501 05 02 JlS DATE: 04/11/90 REV.:
- Station:
Name:
System:
SALEM OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Date Used:
NUCLEAR INSTRUMENTATION Task, Task #, and TAKE CORRECTIVE ACTION FOR A POWER RANGE MALFUNCTION Remote(R) or Control Room (CR): 015 530 05 01 Jl, (CR)
~ SRO
~-
Approval:
K/A Number: 015 999 SG13 Principal Training Supervisor (Operations/Simulator)
or designee Importance Factor:
Operations Manaqer ~ HC/SAL (or designee) a approp_r ia te Grade:
SATISFACTORY UNSATISFACTORY
~ Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM:
15 MINS Start Time Stop Time References: AOP-NIS-1, OP IV-10. Tools and Equipment: NONE Comments:
From any MODE 1 IC set activate Loss of N41 control power malfunction:
DMAL, NI, JPM/1:37 015 530 05 01 Jl Grade/Comments Approved:
NTC-207 DATE: 04/11/90 REV.:
Station:
Name:
System:
Salem OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Date Used:
Emergency Response Task, Task #, and TCAF A LOCA - Depressurize to Refill Pressurizer, Remote(R) or Control Room (CR): 118 501 05 01 Jl, (CR)
E.O. D RO Approval:
SRO 0 K/A Number: 000 009 A6 Principal Training
- (Operations/Simulator)
or designee Importance Factor:
Operations Manager -
HC/SAL (or designee) as appropriate Grade:
SATISFACTORY UNSATISFACTORY
Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM:
10 Min Start Time Stop Time References:
EOP-LOCA-2 Tools and Equipment:
None Comments:
The evaluator will read the steps of EOP-LOCA-2 to the operato Ensures Pzr. level "on-scale" before JPM commence S-JPM:38/2 118 501 05 01 Grade/Comments Approved:
NTC-207 DATE: 12/18/89 REV.:
OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM e Station:
SALEM Name:
Date Used:
System:
REACTOR COOLANT SYSTEM (EMERGENCY PLANT EVOLUTIONS)
Task, Task #, and TCAF A PORV STUCK OPEN Remote(R) or Control Room (CR): 010-514-04-01 Jl (CR)
EO Approval:
P "ncipal Train ng Supervisor (Operations/Simulator)
or designee Importance Factor: 3.9/ ~
~ s Operatl s Manager -
HC/SAL (or designee) as appropriate Grade:
SATISFACTORY UNSATISFACTORY Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Location(s) of Test:
Method of Performance:
SIMULATED PERFORMED Estimated Time-* to Complete JPM:
Start Time Stop Time References:
EOP-TRIP-1 Tools and Equipment:
None Comments:
Initialize the simulator at 100% powe Fail 2PR2 ope Perform EOP-TRIP-1 up to step 2 JPM:l0/2 010 514 04 01 Jl Grade/Comments Approved:
NTC-207 DATE: 05/20/91 REV. :
-
Station:
Narne:
System:
SALEM OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Date Used:
HYDROGEN RECOMBINERS Task, Ta,sk #, and DETERMINE REQUIRED H2 RECOMBINER POWER SETTING Remote(R) or Control Room (CR): 022 527 05 01 Jl (CR)
~ SRO ~
K/A Number: 028 000 A2.0l Approval:
Importance Factor:3.4 / (RO)/( SRO)
HJ~~
Ope~otSMaiiager;z-HC/SAL (or designee) as appropriate Grade:
SATISFACTORY UNSATISFACTORY Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Location(s) of Test:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM:
15 Mins. Start Time
~- Stop Time ~-
References:
OP-II-15. Tools and Equipment:
CALCULATOR Comments:
305S-JPM:25/l 022 527 05 01 Jl Grade/Comments Approved:
NTC-207 DATE: 04/11._ g()
REV. :
9 Station:
SALEM OPERAT~R TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Name:
Date Used:
System:
115 VAC VITAL INSTRUMENTATION Task, Task #,. and STARTUP VITAL INSTRUMENT INVERTER WITH BYPASS ALTERNATE Remote(R) or Control Room (CR): IN SERVICE 062 514 0*1 01 Jl EO D RO 0 SRO 0 Approval:
K/A Number:
062 S/G A9 P1t}hc1p Tra1,, ing Supervisor
!{-Operations/ mulator)
or designee
- .
Importance Factor: 3.2/ (RO)/(SRO)
Operatic Manager -
HC/SAL (or designee) as appropriate Grade:
SATISFACTORY UNSATISFACTORY 41t.Reason, if unsatisfactory:
Evaluator's Signature:
Location(s) of Test:
Method of Performance:
SIMULATED
- Estimated Time to Complete JPM:
20 Min. Start Time Ref'erences: OPERATING PROCEDURE IV-4. Tech Specs Tools and Equipment: none Comments:
DATE:
PERFORMED Stop Time Grade/Comments Approved:
305S-JPM/1:37 062 514 01 01 Jl NTC-207 DATE:
~l.l,~90 REV. :
.,
j. r l
'
OPERATOR TRAINING PROG~
JOB PERFORMANCE MEASURE INVENTORY FORM e Station:
SALEM Name:
Date Used:
System:
EDG Task, Task #,,and STAR'l' UP A DIESEL GENERATOR LOCALLY Remote(R) or Control Room (CR): 120 507 04 01 Ji, (R)
~ SRO ~
K/A Number: 064 000 A4.06 Approval:
- incipal T in1ng Supervisor (Operation /Simulator)
or designee Importance Factor: 3.9/ (RO)/( SRO)
~
Operat o Manager -
HC/SAL (or designee) as appropriate G*rade:
SATISFACTORY UNSATISFACTORY Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Location(s) of Test: D/G Control Room
---'-~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM: 20 Min. Start Time ~~-
Stop Time References: S2.0P-SO.DG-000l(Q) 2A Diesel Generator Operation Rev. O
. -- -*
-0002 (Q) 2B
-.0003(0) 2C Tools and Equipment:
None Comments:
This JPM will be started in a D/G Control Roo S-JPM/1:37 120 507 04 01 Jl Grade/Comments Approved:
NTC-207 DATE: 04/11/90 REV.:
e Station:
SALEM OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Name:
Date Used:
System:
AFW Task, Task #,, and RESET THE STEAM DRIVEN AFW PUMP, Remote(k) or
.
.
Control Room (CR): 120 502 04 oi Jl, (R)~
~ SRO ~
K/A Number: 061 000 A4.09 Approval:
or Importance Factor: 3.8/ (RO)/(SRO)
Operations anager -
HC/SAL (or designee) as appropriate SATISFACTORY UNSATISFACTORY
~
Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Location(s) of Test:
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM: 10 Mi Start Time Stop Time References:
EOP-FRHS 1, OP-III-10. Tools and Equipment:
None Comments:
305S-JPM:l0/l 120 502 04 01 Jl Grade/Comments Approved:
NTC-207 DATE: 04/11/90 REV.:
I I,
OPERATOR TRAINING PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Station:
SALEM Name:
Date Used:
System:
MISC AOPS Task, Task#~ and TCAF A CONTROL ROOM EVACUATION -.ESTABLISH CHARGING FROM Remote(R) or Control Room (CR):OUTSIDE THE CONTROL ROOM 120 514 04 01 Jl EO D RO Approval:
Grade:
SRO 0 Principal Trai (Operations/Si or designee SATISFACTORY Reason, if unsatisfactory:
Evaluator's Signature:
068 EAl.13 Importance Factor: 4.1 / ;fff#~RO) I (_S_R_O_) -
Optri=afions Manager -
HC/SAL (or designee) as appropriate UNSATISFACTORY DATE:
Location(s) of Test: AUXILIARY BUILDING
~~~~~~~....;;_;;;_-=.;;,.~~~~~~~~~~~~~~~~~~~~
Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM: 15 Mi Start Time Stop Time References:
AOP-EVAC-2 Fire Related Alternate SD Equipment Operating Instructions Tools and Equipment:
Panel 216, 213 keys Comments:
This task is performed in several different location The JPM is written with the operator at Panel 21 JPM:lO/l 120 514 04 01 Jl Grade/Comments Approved:
NTC-207 DATE: 12/18/89 REV. :
OPERATOR TRAINING-PROGRAM JOB PERFORMANCE MEASURE INVENTORY FORM Station:
SALEM Name:
Date Used:
System:
OUTSIDE THE CONTROL ROOM Task, Task #, and ENERGIZE THE ALTERNATE SHUTDOWN PANEL (R)
Remote(R) or Control Room (CR): 120 516 04 01 Jl
=..;;;;....;;_...=.;::=..~;;__;;;...=..'--.....;;;...=-~~~~~~~~~~~~~~~~~~~~
K/A Number: 062 SG 9
~~~____,;,~~~~~
E incipal Trai.* n~ Supervisor
- (Operations;s** mulator)
or designee Importance Factor: 3.2/ (RO)/(SRO)
~*-
O~tiOilSManager -
HC/SAL (or designee) as appropriate Grade:
SATISFACTORY UNSATISFACTORY
~
Reason, if unsatisfactory:
Evaluator's Signature:
DATE:
Location(s) of Test: AUXILIARY BLDG, ELEV 64 RELAY ROOM Method of Performance:
SIMULATED PERFORMED Estimated Time to Complete JPM:
10 Min. Start Time Stop Time References:
OPERATING PROCEDURE IV-4.3.11 Tools and Equipment: NONE Comments:
Grade/Comments Approved:
305S-JPM/1:37 120 516 04 01 Jl NTC-207 DATE: 04/11/90 REV. :
ATTACHMENT 3 WRITTEN EXAMINATION AND ANSWER KEY Bank Number Title Section A STATE-05 SGTR with Instrument Failures STATA-01 2PS1 ( PZR spray valve) Failed Open Section 8 TES TI N/A
USE BLACK INK OR NO. 2 PENCIL 00 NOT WRITE ON RED BORDER Page ___ of D
D D
NUCLEAR TRAINING CENTER EVALUATION INSTRUMENT COVER SHEET Job Performance Measure {JPM)
DATE: _0_-_l.......;;;:'J~---&..9_1 __ _
Lab/shop Practical Exercise Knowledge Test COURSE/MOD.NAME~~~-d~'~~~~l-*-~~-~~-e~~~u~a~I-------~
TEST NUMB ER: _f
___
/ __
l-1__.e: __
DATE:
~-S-- 9!
DATE:
~~/1f I
INSTRUCTOR/EVALUATOR/OBSERVER-----------------
AVAILABLE POINTS I " TOTAL POINTS ___ _
GRADE ____ _
PASSING GRADE
%
GRADER'S SIGNATURE __________ DATE:
JPM RESULTS SATISFACTORY UNSATISFACTORY COMMENTS: __________________ _
GRADER'S SIGNATURE:
DATE:
TRAINEE NAME:-------------- (PRINn SSN:
All work done on this examination is my own. I have neither given nor received ai Trainee Signature: _________ _
I have had the opportunity to review the questions and answers on my exa Trainee Signature:-------------
Date:
NTC-62 DATE:
11/16/89
,...,...,,
'"'
.Form.....
1.
Name
OPERATING SUMMARY The plant has sustained a SGTR in No. 23 SG, resulting in a Reactor Trip and ECCS actuatio The operating crew has completed actions to stabilize the plant through step 4.1 of EOP-SGTR-The crew has begun actions to isolate the ruptured generator in accordance with step Prior to the trip, the plant was operating at 100% RTF, EOL, Equilibrium Xenon condition Boron concentration was 100 pp Attempts to close 23MS167 per EOP-SGTR-1 have been unsuccessful. No other actions have been completed at this tim Using this information, as well as the Control Board indications and Control Room references, answer the following question Z SijIFT SECTION A-E JUNE 13, 1991 Form A
~age_ 1 QUESTION:
For the current plant indications, what is the required core exit thermocouple temperature for terminating the rapid RCS cooldown? °F °F QUESTION:
Auxiliary Feedwater flow is lower to 23 and 24 Steam Generators due to: a failure of 21 AFW pump to star a failure of the runout protection on 22 AFW pum a pump shaft failure of 22 AFW pum a governor valve failure of 23 AFW pum Z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 2 QUESTION:
Step 16 of EOP-SGTR-1 asks if the ruptured steam generator has been isolate Assuming all attempts fail at closing 23MS167, 23 S/G Main Steamline Isolation Valve,... the operators must continue efforts to isolate 23 steam generator per steps 4, 5, and 6 and not proceed with the RCS cooldow as long as 21, 22, and 24MS167s, MS18s and MS7s are closed on the intact steam generators, the affected steam generator is considered isolated and the RCS cooldown block of steps is begu The SRO should direct the step be circled due to the S/G not being isolated and continue on to the RCS cooldown block of step The SRO should direct the step be circled and order transition to EOP-SGTR-.
QUESTION:
Select the group of statements which best describes the status of the ECCS equipmen CCP has failed to star CS pump has failed to star A EDG output breaker has failed to clos and 22SW122 have failed to clos l Z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 3 QUESTION:
Prior to step 17 of EOP-SGTR-1, a continuous caution states that ttWHEN TAVG IS LESS THAN OR EQUAL TO 543 DEGREES IN TWO OUT OF 4 LOOPS (P-12) THEN BLOCK LOW STEAMLINE PRESSURE SI.
When this is performed, the "STEAMLINE ISOLATION SI BLOCKED TRAINS A&B" lamp illuminates 6n RP-4 and the Hi Steam Flow with Low Tavg/Low Steam Pressure SI and Stearnline Isolation are blocked The "AUTOMATIC SAFETY INJECTION BLOCKED" lamp illuminates on RP-4 and the Low Steamline Pressure SI is blocke The "STEAMLINE ISOLATION SI BLOCKED TRAINS A&B lamp illuminates on RP-4 and the Hi Steam Flow with Low Tavg/Low Steam Pressure SI is blocke The "STEAMLINE ISOLATION SI BLOCKED TRAINS A&B" lamp illuminates on RP-4 and the Hi Steam Flow with Low Tavg SI is blocke Hi Stearn Flow with Low Steam Pressure SI remains operabl.
QUESTION:
For this event, the cooldown of the reactor coolant system will have to be accomplished using the condenser steam dump syste using the MS10s on the unaffected S/G using the safety valves of the unaffected S/G using the feed and bleed procedure for the reactor coolant syste.
\\
Z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 4 QUESTION:
Before the reactor trip/SI what condition caused the OT-DELTA-T bistable for protection channel III to be energized? A Hot Leg RTD in Loop 23 failing hig A Cold Leg RTD in Loop 23 failing lo The Pressurizer pressure input to OT-DELTA~T failing lo Power Range N43 failing, imposing an OT-DELTA-T penalt.
QUESTION:
The Nuclear Shift Supervisor has directed you to open 23SS94, 23 SIG sample valv To accomplish this... an EO must locally open the valv the 23SS94 OPEN pushbutton only needs to be presse all S/Gs must be restored to normal operating levels and the 23SS94 OPEN pushbutton needs to be presse Restore SG narrow range levels >16%, depress the start pushbuttons on the AUX FW PUMPS and then depress the 23SS94 OPEN pushbutto.
QUESTION:
At step 14 of EOP-SGTR-1 RCS pressure suddenly drops to 1400 psi SI flow is approximately 500 gp What is the appropriate action to take? Ensure 25 degrees subcooling exists prior to initiating the rapid cooldown A concurrent LOCA has occurre Transition to EOP-LOCA-Stop all RCPs Ensure the Charging pump rniniflow valves are ope Z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 5 1 QUESTION:
Assume that the plant has had a Steam Generator Tube Rupture and a Loss of Offsite Powe Both Pressurizer PORV Block Valves have been cleared and tagged prior to this even It is desired to depressurize the RCS as the required core exit thermocouple readings have been reache Depressurization may not proceed -
transition to EOP-SGTR-not proceed until letdown is established to preclude exceeding the 320° delta-T Technical Specification requirement across the spray nozzl proceed using auxiliary spra proceed once offsite power is restored and an RCP is starte.
QUESTION:
Step 4.1 of EOP-SGTR-1 has you reset the ruptured S/G's MS10, Atmospheric Relief Valve, to 1035 psi This action is taken t limit the potential RCS cooldow maintain secondary to primary delta-P within limit ensure the intact S/Gs are the only steam supply to the condenser steam dump minimize the potential for an unisolable release path if RCS pressure is not reduced rapidl Z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 6 1 QUESTION:
If 22 Auxiliary Feedwater pump were to trip, which of the following actions should be initiated? Trip 23 Auxiliary Feedwater pump, close 23MS45 then restart 23 Auxiliary Feedwater pum Continue running 23 Auxiliary Feedwater pump, dispatch an operator with key to close 23MS4 Trip 23 Auxiliary Feedwater pum Continue efforts to restart 21 or 22 Auxiliary Feedwater pump Trip 23 Auxiliary Feedwater pump and transition to EOP-FRHS-z SHIFT SECTION A-E JUNE 13, 1991 Form A Page 1 Z-E5 TEST KEY Answer Bank #
QT Fig Di ff Disc Category 1 Category 2
1. A STATE-05 3 MC
2. A STATE-05 5 MC 118 Em erg Resp 000 Emr Pl 3. B STATE-05 6 MC 118 Em erg Resp 000 Emr Pl $TATE-05 8 MC 118 Em erg Resp 000 Emr Pl.
STATE-05 9 MC 010 Pzr p & Lvl 000 Emr Pl 6. B STATE-05 14 MC
7. c STATE-05 15 MC
8. D STATE-05 16 MC
9. c STATE-05 17 MC
10. c STATE-05 21 MC
11. D STATE-05 18 MC
12. B STATE-05 19 MC
USE BLACK INK OR NO. 2 PENCIL DO NOT WRITE ON RED BORDER Page of
---
NUCLEAR TRAINING CENTER EVALUATION INSTRUMENT COVER SHEET Job Performance Measure (JPM)
DA TE: -~--_/'._J
____ -_....9.....,/ __
_
Lab/shop Practical Exercise Knowledge Test D
D g]
couRs~MOD.NAME~~f_a_L~~~L--~~-~~~~~~~/~~~~~~-
TEST NUMBER: L/- I f1 fl DATE:
(::. -S-
~I DATE:
13/~~
I 7 INSTRUCTOR/EVALUATOR/OBSERVER -------------------------------
AVAILABLE POINTS
TOTAL POINTS GRADE
PASS ING GRADE ____ %
GRADER'S SIGNATURE DATE:
JPM RESULTS SATISFACTORY UNSATISFACTORY
COMMENTS: ___________________ _
GRADER'S SIGNATURE:----------- DATE:----
TRAINEE NAME:
(PRINn
~
SSN:
All work done on this examination is my own. I have neither given nor received ai Trainee Signature: _________ _
I have had the opportunity to review the questions and answers on my exa Trainee Signature:------------------
Date:
NTC-62 DATE:
11/16/89 n r I
.
('"\\
Form ___.~-=---.
3...
8..
1.
1 Name
OPERATING SUMMARY (A-1)
The unit is at 100% power, 1150 MW All systems and controls have been operating normally, with the exception of VCT Level Transmitter LT-112 failing as is, up to the receipt of the current alarm ~ased on this information, control board indications and control room references answer the following question Z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 1 QUESTION:
If a reactor trip occurs, OHA F-46, "RX TRIP" should come i What condition(s) must exist for this alarm to actuate? Either Reactor Trip Breaker and its associated Bypass Breaker are ope All IRPis indicate less than 20 step Reactor Trip status light on 2RP4 is li Reactor Trip Breakers "A" and "B" are ope.
QUESTION:
Assume the operators perform the appropriate actions for a failed open pressurizer spray valve. (Unable to close valve.)
After the plant is stabilized, if a SGTR were to occur, the RCS depressurization would be performed by: spraying via PS 1 and opening both PORV spraying via PS opening one POR.
QUESTION:
Just prior to this event, 2PR6 (PORV Block Valve) was determined to be stuck ope Which of the following actions is required to be performed? Within the next 30 minutes, valv close and remove power from the B *
Within the next 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, place the plant in HOT STANDB Within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, place the plant in HOT STANDB Within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> restore the valve to operabl Z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 2 QUESTION:
Assume that the transient continues and that prior to a reactor trip a total of two (2) Overtemperature Delta T Turbine Runbacks occu The reduction in turbine load, just prior to the trip, would be approximately * 50 MW MW MW MW QUESTION:
Tf the cause of the low pressure condition had been a failed Master Pressure Controller, which of the following alarms/components would not have been directly effected? PR1 (PORV) PR2 (PORV) Pzr Backup Heaters PS1 (Spray Valve)
Z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 3 QUESTION:
Appendix 1 to 2-AOP-PZR-1 provides a procedure to align pressurizer heaters to emergency power supplie Assume that steps 4.1. through 4.1.2.h of this procedure have been complete What effect, if any, would cycling of pressurizer level below and above 17% have on heater operation? ?\\o effec The emergency feeder breaker (2C23X) would trip ope The emergency feeder breaker (2C23X) would cycle open and closed as level oscillated below and above 17%. The emergency feeder breaker (2C23X) would remain closed, however, the individual heater breakers would trip ope.
QUESTION:
Based on temperature error only, if rod control were placed in AUTO, the control rods would begin to move at steps per minut ~...,
I 6 8.
QUESTION:
The low RCP seal injection flow is due to: charging pump tri C\\"71 has failed ope charging flow has decreased due to Pzr level erro C'~arging flow has der;reasen due ti) r.edU<;PG RCS 11rec::sur Z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 4 QUESTION:
Actual VCT level is:. between 14% and 24 % due to auto makeup being in servic between 77% to 87% due to letdown flow diverting to the HU greater than 87.2% due to letdown flow diverting to the HU less than 14% but greater than 1. 4% due to auto makeup being in service and swapover to the RWST not occurrin.
QUESTION:
If seal injection flow to the RCPs continues to be reduced, the RCPs must be immediately trippe must be tripped if RCP Seal Water Inlet temperature exceeds 225°F may continue to be operated as long as seal leakoff remains greater than 2.1 gp must be tripped if proper seal flow cannot be restored within 5 minute.
QUESTION:
In response to the failed VCT level indication, the operators take manual control of the makeup system and secure auto makeu If no additional actions are taken, CV35 will not return flow to the VCT. (i.e., flow will remain diverted to the HUT) The operating charging pump(s) will trip on low suction pressur CV 40 and 41 will close switching Charging pump suctions to the RWST when actual VCT level is 1.4% the VCT will eventually reach Lo-Lo level without receiving a HI/LO LEVEL alarm or auto swapover to the RWS Z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 5 1 THIS QUESTION IS NOT BASED UPON THE CURRENTLY SIMULATED CONDITION QUESTION:
Auxiliary Feedwater Pump Discharge pressure is 1285 psi The operator decides to increase auxiliary feedwater flow to the affected'steam generator by going to full-open on the OPEN pushbutton on the AFWP discharge control valve 21AF21. Select the correct statement below regarding the valve respons The steam generator inlet control valve, 211AF21, will open fully, but the PRESS OVERRIDE light will come o No chang The PRESS OVERRIDE light will come on immediately, but the steam generator inlet control valve, 211AF21, will not change positio No chang The pressure signal is auctioneered, and as a result, the steam generator inlet control valve, 21AF21, will not change positio The steam generator inlet control valve, 21AF21, will open until the operator is overriden by the discharge pressure requirements of the pumps or the valve is full ope.
THIS QUESTION IS NOT BASED UPON THE CURRENTLY SIMULATED CONDITIONS QUESTION:
While operating at 100% reactor power, Turbine First Stage Pressure Transmitter, PT-505, fails lo The condenser steam dumps... will all trip open in response to an indicated 100% loss of turbine loa will receive an arming signal but remain closed until a 5°F temperature error is generate £.
will not open until an arming signal is generate will not operate in the Tavg mode if "RESET is selected to clear the loss of load bistabl..
.I-z SHIFT SECTION A-A JUNE 13, 1991 Form A Page 1 Z-Al TEST KEY lmswer Bank #
QT Fig Dif f Disc Category 1 Category 2
1. A STATA-01 7 MC (/)
118 Em erg Resp 000 Emr Pl 2 * c STATA-01 15 MC 010 PZR p 3. c STATA-01 3 MC 114 Administrat 010 PZR p
.J * B
- STATA-01 5 MC 045 Mn Gen 5. A STATA-01 6 MC 000 Emr Pl 6. A STATA-01 8 MC 010 Pzr p & Lvl 000 Emr Pl 7. STATA-01 9 MC 001 CRDS 8. c STATA-01 10 MC 004 eves 9. B STATA-01 13 MC 010 PZR p 10. B STATA-01 14 MC 010 PZR p 11. D STATA-01 16 MC 010 PZR p 12. D STATA-01 17 MC 012 RPS 13. c STATA-01 18 MC
USE BLACK INK OR NO: 2 PENCIL DO NOT WRITE ON RED BORDER Page ___ of __
NUCLEAR TRAINING CENTER EVALUATION INSTRUMENT COVER SHEET Job Performance Measure (JPM)
DATE: --=b----"-t;:--"':3;;...._-__..L...9,.t,_'/ __
_
Lab/shop Practical Exercise Knowledge Test D
D
~
COURSE/MOD. NAME~~~a~~~~~-~-~-~--~~1~~~a;;...._/ ________ ~
TEST NUMBER:
~
- /
DATE:
{p -S-- °II
~
DATE: b/)/f/
~~7r--~7~<-+---~
APPROVAL BY:
INSTRUCTOR/EVALUATOR/OBSERVER
~-----------------
AVAILABLE POINTS 35"
TOT AL POINTS GRADE
PASS ING GRADE
%
GRADER'S SIGNATURE DATE:
JPM RESULTS SATISFACTORY UNSATISFACTORY COMMENTS:
~--__...,.'------------------~
GRADER'S SIGNA TRAINEE NAME:-------------- (PRINT)
SSN:
All work done on this examination is my own. I have neither given nor received ai Trainee Signature: _________ _
I have had the opportunity to review the questions and answers on my exa Trainee Signature:------------
Date:
NTC-62 DATE:
11/16189 REV.:
Form A Name..
3.
3.
3.
3.
3.
3. (i).
1.
1.
1.
1.
1.
2.
2.
2.
2.
Z SHIFT SECTION B JUNE 13, 1991 Form A Page 1 QUESTION:
When performing/reviewing safety tagging request/releases, which of the following is true? When performing a tagging request the alignment must be performed in the order indicated on the TRIS f?r Written permission of all persons on the tagging request is required before a temporary release may be mad If performing a tagging request in a contaminated area, the original is to be left at the control point and initialed upon the operator's departure from the are When returning a system to service the electrical alignment is to be performed prior to the mechanical alignmen.
QUESTION:
Which of the following require the direct supervision of an SRO? Liquid and gaseous release Load Change Reactivity manipulation Core alteration.
QUESTION:
Select the answer *below which best describes a condition in which the Emergency Coordinator would authorize entry into an area which has been evacuate To allow Health Physics to complete radiological survey To perform repairs on equipment damaged as a result of the inciden To conclude scheduled maintenance or repair To perform operations to mitigate the effect of the emergenc Z SHIFT SECTION B JUNE 13, 1991 Form A Page 2 QUESTION:
Which of the following REQUIRE activation of both the TSC and OSC? Any emergency plan classificatio ~ny emergency condition classified higher than unusual even Only an emergency classification of site area emergency or general emergenc Only emergency conditions classified as a general emergenc.
QUESTION:
In the event the Control Room becomes uninhabitable and control of shutdown systems have not been established locally within 15 minutes, the ECGs refer you to the Site Area Emergency attachmen At what point (which procedure} do you specifically address the control room evacuation? Attachment 3 refers you to EPIP 103S which addresses the control evacuation by going to Section 3.2 -
Radiological/Operational/Personnel Emergencie Attachment 3 refers you to EPIP 103S which addresses the control evacuation by going to Section 3.3 - Fire/Natural Event/Security Emergencie Attachment 3 refers you to EPIP 103S which addresses the control evacuation by going to Section 3.4 - Miscellaneous Emergencie Attachment 3 refers you to EPIP 103S which addresses the contr61 evacuation by going to Section 3.5 - Termination and De-escalatio Z SHIFT SECTION B JUNE 13, 1991 Form A Page 3 QUESTION:
An Alert has been declared at Hope Creek Generating Station. From the following, select the response NOT required of the Salem SNS Maintain communications with the Hope Creek Generating Statio Implement personnel assembl Staff Emergency Response Facilities with minimum personne Monitor station for habitabilit.
QUESTION:
The secondary activity limit is based on a micro-curies per gram and is.00, Failed open Steam Line Safety Valve.10, S/G Tube Leak with a concurrent Steamline Rupture.00, S/G Tube Leak with a concurrent Stearnline Rupture.10, Design Basis Steam Generator Tube Rupture QUESTION:
The plant has experienced a reactor trip and safety injectio The SRO has made the transition to 2EOP-TRIP-Under which of the*
following conditions would continued operation of ECCS equipment be required?
RCS PRESSURE SUBCOOLING PZR LEVEL 96 Stable
15 Increasing
13 Decreasing
10 (.Adverse) Stable
6
Z SHIFT SECTION B JUNE 13, 1991 Form A Page 4 QUESTION:
EOP-SGTR-1, "Steam Generator Tube Rupture, step 6, requires feed to the ruptured SIG be maintained until its narrow range level is >5%
[15% adverse], then stoppe What is the reason for this requirement? Ensure adequate inventory for RCS backfil Prevent depressurization of the SIG by keeping the U-tubes covere Ensure U-tubes are covered for maximum heat transfer capabilit Prevent an adverse chemical environment from impacting on intact U-tube.
QUESTION:
The plant has experienced a loss of all AC powe During recovery from this event, the NCOs begin a controlled depressurization of the intact SIGs to: prevent lifting of the Pressurizer Code Safetie prevent. lifting of the Stearn Generator Code Safetie increase Auxiliary Feedwater Flow to the intact S/G reduce RCP seal leakage and minimize RCS inventory los Z SHIFT SECTION B JUNE 13, 1991 Form A Page 5 1 QUESTION:
A startup is in progress on Unit All conditions are normal, except that 21BF40 will not close in accordance with step 5.51 of IOP-The valve is stuck fully ope Which one of the following actions is appropriate? Continue the power ascensio Decrease power, place the AFW system in service and close 21BF21, manual stop valve downstream of 21BF4 All actions must be completed within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Place the unit in Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and Mode 5 within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Place the unit in ~ode 3 within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
QUESTION:
Feed and Bleed cooling has been initiated in response to a loss of heat sin A review of core exit thermocouples shows temperatures have continued to trend u What flow rate is called for when any feed capability is restored? At least 22E4 lbm/h no more than 5E4 lbm/h maximum available flow to all steam generator maximum available flow to the selected steam generato Z SHIFT SECTION B JUNE 13, 1991 Form A Page 6 1 QUESTTON:
A Unit startup is in progress with reactor power at 50% and all conditions norma The Equipment Operator reports that a Steam Generator Code Safety Valve has lifte Based on these conditions, which one of the following best describes the appropr~ate operator action? Trip the reactor and enter EOP-TRIP-Enter AOP-STM-1 because reactor power will exceed turbine load by more than 10%. Declare the Code Safety valve inoperable and reduce the PR NI high flux trip setpoint to 87% within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Place the unit in mode 4 within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
QUESTION:
Which ONE of the following will result in an initial positive reactivity addition to the Reactor?
~-
Decrease in main feedwater flo Decrease in main feedwater temperatur Increase in reactor coolant system temperatur Decrease in main steam flo Z SHIFT SECTION B JUNE 13, 1991 Form A Page 7 1 QUESTION:
The response of the Containment Fan Cooler Units (CFCU) to a Loss of Coolant Accident is:
(Assume offsite power available.) Running CFCUs tri All OPERABLE CFCUs are started in slow speed by the SE Running CFCUs immediately shift to slow spee sequenced on in slow speed by the SE Idle CFCU are Running CFCUs tri Previously idle CFCUs are sequenced on in slow speed by the SE Running CFCUs trip. Previously running CFCUs are sequenced on in slow speed by the SE.
QUESTION:
The plant is operating at 100% power when a fire on 2A 115 VAC Vital Instrument Bus results in an overcurrent trip of the associated inverter output breake The plant does not tri After the fire is out, which of the following actions must be taken? Manually trip the reacto Loss of A Vital Bus should result in an automatic reactor trip. Reenergize the effected bus within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; restore the inverter to OPERABLE within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s* The remaining AC buses and inverters shall be determined OPERABLE and energized within 8 hou Begin shutdown to bring the plant to HOT STANDBY within 1 hou Z SHIFT SECTION B JUNE 13, 1991 Form A Page 8 1 QUESTION:
Control Air pressure has decreased to 65 psi Which one of the following gives the correct valve response? fail closed:
Pressurizer PORVs (PR-1 & PR-2). fail closed:
Letdown isolation valves (CV-2 & CV-277) and Charging line stop valves (GV-71 & CV-77). fail to VCT position:
Deborating demin inlet/bypass valve (CV-27), and Letdown flowpath demin divert valve (CV-21). fail open:
RCP seal leakoff stop valves (CV-114 & CV-104s)
and VCT vent valve (CV-280).
1 QUESTION:
A failure of a Power Range Channel requires the installation of a jumper when it is placed in the tripped conditio This jumper: ensures the availability of the power range high flux (low setpoint) and intermediate range high flux trips upon the failure of a second PR channe ensures that a reactor trip will occur if power is decreased to less than 10% of RT prevents a spurious reactor trip if power is decreased to less than 10~6 of RT ensures the continued availability of the at-power reactor trips (P-7).
Z SHIFT SECTION B JUNE 13, 1991 Form.;
Page 9 1 QUESTION:
A reactor startup is in progress with reactor power at 1.0E-8 amp If intermediate range channel N35 fails low, WHAT operational limit must be observed? Power must be reduced below P-6 setpoint within 1 hou The inoperable channel must be placed in the tripped condition within 1 hou Power must remain below 5% until the inoperable channel is repaire All operations involving positive reactivity additions must be suspended immediatel.
QUESTION:
Unit 2 is being refueled (Cavity level at elevation 128' RH-R pump is in service controlling RCS temperature at 100° A EDG is cleared and tagged for scheduled maintenanc An electrical
supervisor calls to ask if 22 RHR Pump can be cleared and tagged to perform an electrical inspection of the moto Can the 22 RHR pump be cleared and tagged? ~o, two independent loops are required at all time No, since the 2A EDG is cleared and tagged, no credit can be taken for the 21 RHR pump being operable and the 22 RHR pump cannot be removed from servic Yes, emergency power supplies for the RHR pumps are not required in Mode Yes, as long as one RHR loop is in service, the second RHR loop can be removed from service for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> perio Z SHIFT SECTION B
,JUNE 13, 1991 Form A Page 10 2 QUESTION:
The "Fir~ in Control 2\\rea" mode of operation of the Control Area Air Conditioning System is initiated:* automatically by the. Fire Protection System, drawing air from the outside through the control area, then back insid automatically by the Fire Protection System, drawing air from storage bottles through the dampers, then back outsid manually by the operator, drawing air from the outside through the control area, then back outsid manually by the operator, drawing air from storage bottles through the dampers, then back outsid.
QUESTION:
Which statement below is correct if the power range instruments have been adjusted to 100% based on a calculated calorimetric? If the feedwater temperature used in the calorimetric calculation was higher than actual feedwater temperature, actual power will be less than indicated powe If the reactor coolant pump heat input used in the calorimetric calculation is omitted, actual power will be less than indicated powe If the steam flow used in the calorimetric calculation was lower than actual steam flow, actual power will be less than indicated powe If the steam pressure used in the calorimetric calculation is lower than actual steam pressure, actual power will be greater than indicated powe Z SHIFT SECTION B JUNE 13, 1991 Form."!\\
Page 11 2 QUESTION:
Unit 2 is at full power, all conditions norma At 0200 on May 5th ' a brief fire in the 2C Vital Instrument Inverter causes the 2C Vital Instrument Bus to swap to the AC Regulato Assuming that the Inverter is not repaired, Unit 2 must be placed in Mode 3 by 0800 on May 5 and Mode 5 by 1400 on May Mode 3 by 1600 on May 5 and Mode 5 b~' 2200 on May Mode 3 by 0200 on May 6 and Mode 5 by 0800 on May 7. Mode 3 by 0800 on May 6 and Mode 5 by 1400 on May *.
2 QUESTION:
During routine operations at full power, OHA A-27, "."!\\MSAC BYP.;SSED",
alarm Upon investigation, it is revealed that the alarm is valid and that AMSAC is, in fact, bypasse Due to these conditions, '{OU should: notify I&C that AMSAC has failed and continue to operate normall begin a controlled rampdown of power to less than 40% power and maintain power below 40% until AMSAC is restored to OPERABLE statu return AMSAC to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or reduce power to less than 40% of RT return A~SAC to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or place the plant in HOT STANDBY within the following 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />...
Z SHIFT SECTION B JUNE 13, 1991 Form A Page 12 2 QUESTION:
Following a control rod drop and the subsequent recovery of the rod to its previous position, the Flux Rate Trip on the NIS cabinet is rese Why is the rate trip reset after recovering the rod rather than before the recovery? The rate trip cannot be reset until the control rod is returned to its original positio The rate trip cannot be reset until the control rod is returned to +/- 12 steps of its bank positio If the rate trip is reset and the rod falls again, the second rate trip may result in a reactor tri The rate trip is maintained to assure a reactor trip will occur in the event another control rod drop occurs during the recovery proces,J Z SHIFT SECTION B JUNE 13, 1991 Form A Page 13 2 QUESTION:
At 1:00 a.m., during a down power evolution to 80% for maintenance, steam was observed blowing by one of the steam generator code safety valve The steam leak-was verified and the code safety valve was declared inoperable at 3:00 a.m. Repair time was estimated to be 2 day At approximately 3:30 p.m. the same day, during turnover, the maintenance activity requiring the power reduction was completed, and reactor power was now back to 90% and increasin You have just completed the turnover and you have the control Which of the following actions are to be taken? Continue the power ascension to 100% powe There was, and is, no Tech. Spec. violation - at least one steam generator is available with all associated code safety valves operable and the affected steam generator will have the code safety valve repaired in under 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Power is to be reduced to below 87% power at which time the Power Range Neutron Flux High trip setpoint is reduced to 87%.
With the reduced power setpoint, the plant is permitted to operate indefinitely; no Tech. Spec. has been violate Because the valve cannot be repaired within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor power must be reduced to HOT STANDBY within four (4) hour Additionally, because the time period for power reduction has been exceeded, the NRC must be notified in writing within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> The applicable Tech. Spec. has been violate Action 3.0.3 must be implemented until the Power Range Neutron Flux High trip setpoint is reduced to less than 87%.
2 QUESTION:
Auxiliary Feedwater System Operating Procedures require that the alternate suction header be drained when not in us This requirement is designed to avoid leakage onto safety related switchgea avoid leakage of non-primary grade water into the AFW Syste minimize corrosion of the pipin minimize pipe stress from the static loa "*
.)
Z SHIFT SECTION B JUNE 13, 1991 Form A Page 14 2 QUESTION:
Which of the following statements regarding EOP-LOSC-1, Loss of Secondary Coolant, is true? The procedure assumes an increased risk of a SGTR due to high differential pressure across the faulted S/G U-tube The procedure assumes that the RCS will no longer be subcooled once the faulted S/G has completely cooled dow The procedure assumes that the complete blowdown of one SIG will cool down the RCS to less than 350° The procedure assumes an increased risk of a LOCA due to Pressurized Thermal Shoc.
QUESTION:
Unit 2 is presently in a refueling outage with fuel movement in progres The Unit 1 Primary Equipment Operator is sent to tag q,ut lA Diesel Generator but inadvertently tags out 2A Diesel Generator (4 KV breaker racked down and tags properly hung). The Unit 2 Desk NCO identified the mistake and tags were removed, breaker racked up and control power was restored on 2A Diesel Generato The 2A diesel operability surveillance was verified to be current and was not reperforme Following shift turnover, 28 Diesel Generator is tagged in preparation for planned maintenance activitie Based on these conditions, what action(s) must be taken to comply with Tec Specs.? No action is necessar C Diesel Generator must be demonstrated to be OPERABLE within one hour and subsequently once every 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> CORE ALTERATIONS must be suspende CONTAINMENT INTEGRITY must be established within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Z SHIFT SECTION B JUNE 13, 1991 Form A Page 15 3 QUESTION:
The Unit is operating at full powe A 125 VDC Bus is being supplied with power through 2A2 Battery Charge Al Battery Charger has been inoperable since 1300 on December 20 th.
In order to comply with Technical Specifications, no action is necessar restore 2A 125 VDC Bus to OPERABLE status by 1300 on December 23 rd, or be in HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> restore 2A 125 VDC Bus to OPERABLE status by 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> on December 20 th,or be in HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> return 2Al Battery Charger to operation by 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on December 27 th, or be in HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
QUESTION:
A reactor trip has occurred due to a loss of all feed pumps. After completing EOP-TRIP-1, "Reactor Trip or Safety Inje'ction", EOP-FR... HS-1, "Response to Loss of Secondary Heat Sink", has been entered due to inadequate auxiliary feedwater flo While performing this procedure a complete loss of AC power occur Which of the following is the ciorrect action to take? Go to EOP-LOPA-1, "Loss of All AC Power". Continue with EOP-FRHS-Continue with EOP-FRHS-1 until a higher order red path occur Continue with EOP-FRHS-1; review EOP-LOPA-1 for appropriate actions to perform in paralle Z SHIFT SECTION B JUNE 13, 1991 Form A Page 16 3 QUESTION:
Unit 2 is in Mode 5 following refueling with 21 RHR Pump in servic Plant temperature is being controlled at 100°F with RCS water level at mid-loo The NCO notices erratic RHR pump amp indication, followed by the OHA E-21, DECREASING VESSEL LEVE After reducing flow in an attempt to correct the cavitation probl~m, the operator stops the affected pum The reactor has been shutdown for 10 day Saturation will be reached in the RCS in approximately:
l\\.
12 minutes minutes minutes minutes 3 QUESTION:
A reactor startup is in progress on Unit Inverse Count Rate Ratio data is as follows:
o Reference Counts = 212 cps o
Bank C @ 100 steps; count rate = 312 cps o
Bank C 125 steps; count rate = 342 cps Based on this data, the ICRR predicts criticality at ? Bank D @ 20 steps Bank D @ 100 steps Bank D @ 180 steps Bank D @ greater than 228 steps
Z SHIFT SECTION B JUNE 13, 1991 Form A Page 17 3 QUESTION:
A small break LOCA has occurred on Unit 2. All systems have functioned as designe Charging Pump has just been taken out of service per step 20 of EOP-LOCA-Assuming that RCS subcooling is sufficient, the next ECCS pump to be taken out of service should be SI Pum SI Pum Charging Pum Charging Pump.
. 3 QUESTION:
If power and pressurizer pressure are held constant, how does tripping a reactor coolant pump while at power affect DNBR? No effec DNBR increase DNBR decrease DNBR decreases slightly, but eventually increase ~*-
z SHIFT SECTION B JUNE 13, 1991 Form A Page 1 TESTl TEST KEY Answer Bank #
QT Fig Diff Disc Category 1 Category 2
1. B NRC-GEN 5 MC
2. D NRC-GEN 2 MC 005 RHR 3. D NRC-GEN 6 MC
4. B NRC-GEN 9 MC
5. c NRC-GEN 12 MC
6. c NRC-GEN 17 MC
7. B NRC-SF03 19 MC 113 Misc Cond 8. A NRC-SF03 20 MC 012 RPS 9. B NRC-SF03 23 MC
10. D NRC-SF04 18 MC 012 RPS 11. B NRC-SF05 30 MC
12. D NRC-SF05 45 MC
13. c NRC-SF05 13 MC
14. B NRC-SF05 11 MC
15. A NRC_:SF06 6 MC
16. B NRC-SF07 7 MC 005 RHR 17. c NRC-SF08 1 MC 117 Misc Emerge 18. A NRC-SF09 7 MC 075 Circ Water 19. c
- NRC-SF09 29 MC
20. c NRC-SFll 5 MC
21. c NRC-SFll 11 MC
22. B NRCTHRMO 7 MC 005 RHR 23. D NRC-SF07 33 MC
24. A NRC-SF01 3 MC 075 Circ Water 25. D NRC-SF01 8 MC 039 MS & BS 26. D NRC-SF05 2 MC 005 RHR 27. A NRC-SF05 14 MC 110 Dem in Water 28. A NRC-SF05 49 MC 035 SGs 29. c NRC-SF07 5 MC 008 ccw 30. D NRC-SF07 9 MC 005 RHR 31. A NRC-SF07 29 MC
32. c NRC-SF02 14 MC 016 Detectors 33. D NRCRXFND 12 MC 055 Cond A 34. A NRC-SF03 36 MC
35. c NRC-.SF04 26 MC 035 SG & SGBD
-~**
Contents:
ATTACHMENT 4 LICENSEE REPORT OF THE REQUALIFICATION EXAMINATION June 24, 1991, letter to T. Timothy Martin, Regional Administrator, USNRC from Arthur Orticelle, Manager - Nuclear Training, PSE& NOTE:
In the Tabular Summary of Individual and Team Results (ATTACHMENT 1), the license numbers have been erased and page 3 of 3 (which correlated numbers with individual operators) has been removed from the licensee report for operator privacy considerations.