ML20056E484
| ML20056E484 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/09/1993 |
| From: | Aiello R, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20056E465 | List: |
| References | |
| 50-369-93-300, 50-370-93-300, NUDOCS 9308240124 | |
| Download: ML20056E484 (150) | |
See also: IR 05000369/1993300
Text
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! AT LANT A. GEORGI A 30323 ,s...../ Report Nos.: 50-369/93-300 and 50-370/93-300 Licensee: Duke Power Company 1 422 South Church Street Charlotte, NC 28242 i Facility Docket Nos.: 50-369 and 50-370 Facility License Nos.: NPF-9 and NPF-17 Examination Conducted: July 12-16,,1993 1 Chief Examiner: /
RonirTd F. Aiello Dat6 S(gned Examiners: B. Haagensen, Sonalysts G. Weale, Sonalysts Approved by: . b- ' LiwFence L. Lawyer, Chief [/ Date Signed Operatcr Licensing Section 1 Operations Branch Division of Reactor Safety SUMMARY Scope.: HRC examiners conducted regular, announced operator licensing requalification examinations and associated inspection activities during the periods June 28- July 2 and July 12-16, 1993. Examiners administered examinations under the guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7. Seven Senior Reactor Operators (SR0s) and four Reactor Operators (R0s) received written and operating examinations. For the simulator portion of the examination, operators comprised three crews. Four stand-in operators were used in the simulator portion of the examination to complete the crew complement and satisfy the minimum numbers requirement for program evaluation in accordance with NUREG-1021, Revision 7. The examiners conducted technical interviews and reviewed the facility's evaluation of previously administered written and operating remedial training (1991 to the pracant) under the guidelines of TI 2515/117 to verify that the licensee had adequately addressed l licensed operator and crew performance weaknesses. The examination team also l reviewed licensed operators' medical records in accordance with ANSI /ANS-3.4- 1983. 9308240124 930812 PDR ADOCK 0500036? ! V PDR El , 1
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, 2 l , Results: Operator Pass / Fail: SR0 R0 Total Percent Crews Percent j
Pass 9 5 14 93.3 3 100- l Fail 0 1 1 6.7 0 0 ! Examiners judged that the McGuire Nuclear Station requalification program was satisfactory based on the results of the examinations. Examiners identified weaknesses in Job Performance Measures (JPM) construction
(paragraph 2.c), JPM administration (paragraph 2.e), crew communications I (paragraph 2.e), operator knowledge and Emergency Preparedness (paragraph 2.1 i and Appendix B) and Requalification training and remediation (paragraph 2.i ! and Appendix D). ! '
l Examiners identified a strength in the area of Anticipated Transient Without ! Trip (ATWT) versus Reactor Protection System (RPS) failure discrimination i (paragraph 2.1 and Appendix B). ! - ! One inspector follow-up item (IFI) in the area of communications was ! identified (paragraph 2.e). ! 1' One cited violation in the area of procedures was identified (paragraph 2.h l l and Appendix A). One cited violation in the area of medical records was identified (paragraph f 2.1 and Appendix C). i i Three non-cited violations (NCVs) were identified. Two were in the area of ! medical records, (paragraph 2.1 and Appendix C) and one was in the area of l training (paragraph 2.i and Appendix D).
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l REPORT DETAILS 1. Persons Contacted Licensee Employees
- J. Alexander, Supervisor, Occupational Health
- D. Baxter, Operations Support Manager
- A. Beaver, Shift Operations Manager
- B. Caldwell, Training Manager
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- M. Geddie, Station Manager
- B. Hamilton, Superintendent of Operations
- D. McGinnis, Operations Training Director
- T. McMelkin, Vice President, McGuire Nuclear Station
- P. Weaver, Assistant Shift Supervisor
l Other licensee employees contacted included instructors, engineers, . technicians, operators, and office personnel. '
' NRC Personnel
- K. VanDoorn, Senior Resident Inspector
T. Cooper, Resident Inspector ,
- Attended exit interview
, 2. Discussion a. Scope l NRC examiners conducted regular, announced operator licensing requalification examinations and associated inspection activities during the periods June 28-July 2 and July 12-16, 1993. Examiners administered examinations under the guidelines of the Examiner Standards, NUREG-1021, Revision 7. Seven SR0s and four R0s received , written and operating examinations. For the simulator portion of the examination, operators comprised three crews. In the simulator portion of the examination, four stand-in operators completed the crew complement and satisfied the minimum numbers requirement for program evaluation in accordance with NUREG-1021, Revision 7. The examiners conducted technical interviews and reviewed the facility's evaluation of previously administered written and simulator remedial training (1991 to the present) under the guidelines of TI 2515/117 to verify , that the licensee had adequately addressed licensed operator and crew performance weaknesses. The examination team also reviewed licensed operators' medical records against the requirements 10 CFR 55. ' b. Reference Material The examination team reviewed the reference material and determined , that the reference material was adequate to support the examination. The examiners also reviewed the licensee's 1992 Licensed Operator Requalification Program sample plan. The sample plan was compared to NUREG-1021, ES-601, Attachment 2. The sample plan met all the guidelines for NRC administered examinations. l
) Report Details 2 ' c. Examination Development The examination team conducted a comparison of the facility's proposed written, walk-through, and dynamic simulator examinations to the guidance of NUREG 1021. The NRC substituted new material for about 10 . percent of each section of the examination. The team approved the ' remainder of the examination with the following changes. (1) Written i The team either changed or replaced several distractors to make the questions more plausible. The examination team replaced one question on Part A and two on Part B. One new question for the , written Part A examination was developed by the examination team
to investigate a weakness that was suspected by the examiners during the preparation week. (2) JPMs . ' All six JPMs for this examination contained instances of verification-only steps being designated as critical steps. Verification-only steps, as indicated in Examiner Standard (ES-603), are not critical steps. The examination team made significant revisions to the JPMs to bring the critical step designations in line with the guidelines of ES-603. The examiners walked down two in-plant JPMs (JPM-DG-DG-ll and JPM-EL-EPL-51) during the preparation week. Neither one of the i JPMs was ready for use in an examination setting. In addition to the problem of improperly designated critical steps described above, neither JPM contained sufficient prepared evaluator cues on panel indications and equipment responses. JPM-DG-DG-ll also contained a designation " MANUAL STOP" which appears in enclosure 4.2 of procedure OP/1/A/6350/02, Diesel Generator, but does not match the " DIESEL CONTROL" designation for the applicable switch on the local control panel. The licensee was informed of this discrepancy. (3) Simulator Scenarios , The exam team reviewed all simulator scenarios that were selected for the exam week. These scenarios were generally well constructed, were in keeping with the guidelines of the examiner standards, and ran properly in the simulator. The examiners identified no significant concerns, and made no major substi- tutions to the facility developed scenarios. However, they did . include minor modifications to improve the discrimination of the j simulator examination. One passive failure (failure of a - . -
Report Details 3 charging valve to close on containment isolation) was added and
one crew critical task standard (trip the turbine during an ATWT) l was modified for accuracy. The licensee agreed with all of the l changes. d. Examination Administration , l The licensee's administration of the JPMs was not well planned and
coordinated. The facility did not provide an extra operator in the , l simulator to handle nuisance and unrelated alarms, as normally done ! during facility administered requalification examinations, thus causing the operators to be unnecessarily distracted. The control room was not available during the administration of JPMs which caused i a " log jam" during the performance of the simulator JPMs. This caused 1 the operator stress factor to go up slightly due to the extended hours. The written and simulator portion of the examination closely followed the schedule. The examination administration satisfactorily met all the recommendations delineated in NUREG 1021, ES-601, and ' ES-604, Paragraph D, respectively. The team determined that all facility evaluators were satisfactory. c. Operator Performance Eleven licensed operators took a written examination and an operating test. The operating test consisted of a walk-through examination and , I a dynamic simulator examination. This examination was developed and i ! administered in accordance with NUREG 1021, ES-602, ES-603, and ES- 604, respectively. Four stand-in operators were used in the simulator portion of the examination to complete the crew complement and satisfy the minimum numbers requirement of NUREG-1021, Revision 7. l (1) Written ! SR0 performance was unsatisfactory on one question in the Part A l examination that was inserted by the examination team to test a l previously identified facility knowledge weakness (paragraph 2.1). Three of the five SR0s missed the question. Overall, one R0 failed the written portion of the examination. The R0/SR0 average for the Part A and Part B combined was 83.8/89.2 percent respectively. The previous R0/SRO 1993 requalification written examination averages were 86.1/87.7 percent respectively. The examination team determined that the written examination was sufficiently discriminating. (2) JPM Performance The facility operators were very thorough during JPM performance. The examination team noted that the operators were making every effort to apply the facility STAR program (Stop, Think, Act, and Review). One operator missed a critical task in one of the JPMs
Report Details 4 t l due to improper cuing from the facility evaluator resulting in a failure of that JPM. The cue was distracting as a result of excessive " window dressing" causing the operator to miss its intent. This was identified as a weakness. During the performance of JPM-EL-EPL-51, Start Vital Inverter . 2EVIA, several operators had difficulty with step 2.1.8 of ! Enclosure 4.5 to OP/0/A/6350/01A. Step 2.1.8 required the operators to " verify correct 2EVIA static inverter output voltage, frequency, and zero output amps." The operators tried - to perform this verification by looking at the system output meters on the Manual Bypass Panel rather than the inverter output meters on 2EVIA. ' During the performance of JPM-MT-MT-88, Recover From a Turbine Runback, many operators had difficulty with subsequent action step 2 of AP/1/A/5500/03, Case II. Step 2 required the operators to " verify condenser Dump valves - OPEN." At this point in the recovery, some or all of the dump valves may be closed. The AP procedure ambiguity caused one operator to take the steam dump controller to manual and subsequently, he unnecessarily opened i all of the condenser dump valves. The facility was informed of l this procedure deficiency. (3) Simulator , Crew performance in the simulator Part of the examination was ' satisfactory. However, vital information was not relayed to l plant personnel via the plant paging system during key phases of , emergencies as required by OMP 1-12, " Operations Communication l Standard". The Training Department does NOT enforce the use of the plant paging system during Active Simulator Exercises (ASEs). The procedure clearly states that the page announcing system should be used to announce emergencies or unexpected events, to ! relay information regarding plant status, and to direct actions in the plant. This was considered negative training and was identified as a training department weakness. This open item will be tracked as IFI 50-369,370/93-300-01: Failure to incorporate the use of the plant paging system into simulator training exercises. f. Program Evaluation Based on the examination results, McGuire Nuclear Station meets the criterion established in NUREG-1021, ES-601. The examiners judged that the McGuire Nuclear Station requalification program was satisfactory. } !
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, l Report Details 5 I g. Simulator Facility NRC examiners reviewed the licensee's record of simulator usage during calendar year 1992. The inspectors found that the simulator was ' available for a total of 6,240 hours based on a Monday through Friday usage. During these hours of availability, the simulator was used for operator training related functions 59.7 percent of the time. It was > used for nonoperator training functions 1.3 percent of the time, and it was idle 39 percent of the time. The examiners concluded that the simulator availability for operator training usage was satisfactory. The examination team identified one configuration deficiency in the simulator. In the actual control room, the plant vent flow integrator has a fixed conversion value plaque as an operator aid. This conversion value plaque was not installed in the simulator. This deficiency was corrected during the examination week. The simulator froze up twice; once during the preparation week and once during the examination week. The licensee stated that this was not typical of the facility. These freeze-ups did not significantly contribute to excmination delays. h. General Observations (71707) The licensee failed to adhere to the requirements of 10 CFR 50, Appendix B, regarding the availability and use of an uncontrolled document (Guidelines for Inoperability) by SR0s in the control room. This finding which resulted in a cited violation is discussed in further detail in Appendix A. l No problems were encountered with either security or health physics ! inprocessing. , i. Training and Qualification Effectiveness (41500) (1) Technical Interviews The examiner identified several weaknesses in operator knowledge. The examiners conducted technical interviews with licensed operators in accordance with NRC inspection procedure 41500, to determine whether personnel have qualifications commensurate with the performance requirements of their jobs and to ensure training improvement programs were effective. During the interviews and performance observations, the examiners determined that the overall quality of work was consistently meeting job performance requirements. However, many operators displayed knowledge i weaknesses. Knowledge weaknesses in these areas could prevent proper corrective actions during casualties. Strengths and weaknesses are identified and discussed in detail in Appendix B. , , , - . - -,---r-, - - -
_- - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l Report Details 6 l (2) Medical Records The examiners reviewed all of the licensed operators' medical certifications in accordance with the requirements of ANSI /ANS-3.4-1983, " Medical Certification and Monitoring of Personnel Requiring Operator licenses for Nuclear Power Plants." The findings, which resulted in one cited violation and two non- cited violations, are discussed in detail in Appendix C. (3) Remedial Training ] The examiners reviewed the facility's evaluation of previously administered written and simulator remedial training (1991 to the present) in accordance with TI 2515/117 to verify that the training department had adequately addressed licensed operator and crew performance weaknesses. The McGuire Nuclear Station Training Department did not maintain adequate records to be used in determining whether operator remediation had been correctly specified, characterized, or completed as required by 10 CFR 55.59. The findings, which resulted in one non-cited violation and a training department weakness, are discussed in detail in Appendix D. j. Procedures The examiner inspected the procedures listed in Appendix E for quality and usefulness and found them to be satisfactory with the exception of the discrepancy identified in paragraph 2.e above regarding the subsequent action step 2 of AP/1/A/5500/03, Case II. 5. Exit Interview At the conclusion of the site visit, the examiners met with those representatives of the plant staff, indicated in paragraph I above, to discuss the results of the examinations and inspection findings. The licensee did not identify as proprietary any material provided to or reviewed by the examiners. The examiners further described the areas inspected and discussed in detail the inspection findings listed below. Item Number Description / Para 50-369, 370/93-300-01 IFI - Failure to incorporate the use of the plant paging system into simulator training exercises (paragraph 2.e) 50-369,370/93-300-32 NOV - Failure to adhere to the requirements of 10 CFR 50, Appendix B, regarding the use of procedures (paragraph 2.h and Appendix A) o____--_ .. . _
! , i Report Details 7 i i 50-369,370/93-300-03 NOV - Failure to report the medical status of , certain licensed individuals (paragraph 2.i and i Appendix C) c 50-359,370/93-300-04 NCV - Failure to properly correct an error made on a licensee medical form (paragraph 2.i and Appendix C) ! 50-369,370/93-300-05 NCV - Failure to obtain a Nuclear Operator biennial medical examination within the required time frame (paragraph 2.i and Appendix C) 50-369,370/93-300-06 NCV - Failure to adhere to the requirements of 10 CFR 55.59 regarding the evaluations and documentation of operating tests for operators who have exhibited deficiencies (paragraph 2.i and Appendix D). i 1 i 2 ,
_ l , ENCLOSURE 3 ! SIMULATOR FACILITY REPORT
i Facility Licensee: McGuire Nuclear Station facility Docket Nos.: 50-369 and 50-370 Operating Tests Administered During: The Week of July 12-16, 1993 This form is used only to report observations. These observations do not constitute, in and of themselves, audit or inspection findings and are not, l without further verification and review, indicative of noncompliance with l 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required solely in response to these observations. ' r The examination team identified no simulator fidelity items. The examination team identified one configuration item. In the actual control room the plant vent flow integrator had a fixed conversion value plaque as an operator aid. This conversion value plaque was not installed in the simulator during the examination preparation week but was corrected before the examination began. i !
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! APPENDIX A i USE OF AN UNCONTROLLED DOCUMENT ! During a simulator scenario, the examiners observed the operators using a book i entitled " Guidelines for Inoperability" to determine which bistables should be
tripped for a pressurizer pressure instrument failure. An additional bistable j had been added to the list by hand-writing the name in pencil. This bistable, l Over Temperature Delta Temperature (0 TDT), was required to be tripped to properly place the failed instrument in a tripped condition. In the i " Guidelines for Inoperability" there was a statement to the affect that the remaining " Guidelines for Inoperability" would be developed by July 1,1987, and maintained as a controlled document either as part of the technical l specification (TS) reference manual or as a new Operations Management i Procedure (0MP). This document had never been made a controlled document but' , had been used by operators in the control room to determine operability. j
10 CFR 50, Appendix B, in part, states that the pertinent requirements of this appendix apply to all activities affecting the safety-related functions of i those structures, systems, and components which includes operating procedures. Furthermore, it states that quality assurance includes quality control, which i comprises those quality assurance actions related to the physical i characteristics of a material, structure, component, or system which provide a i means to control the quality of the material, structure, component, or system l to predetermined requirements. j 10 CFR 50, Appendix B, Criterion V, states that activities affecting quality l l shall be prescribed by documented instructions, procedures, or drawings, of a l l type appropriate to the circumstances and shall be accomplished in accordance i with these instructions, procedures, or drawings. Instructions, procedures, i or drawings shall include appropriate quantitative or qualitative acceptance
i criteria for determining that important activities have been satisfactorily accomplished. ] The use of this uncontrolled document by SR0s in the control room is identified as a failure to adhere to the requirements of 10 CFR 50, Appendix B, and is identified as Violation 50-369,370/93-300-02: Failure to adhere to the requirements of 10 CFR 50, Appendix B, regarding the use of procedures. l l
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APPENDIX B ! KNOWLEDGE STRENGTHS AND WEAKNESSES l Five of six SR0s stated that a plant shutdown in accordance with TS 3.7.1.2
was required for a loss of all Auxiliary Feedwater (CA) pumps. Several SR0s- ' incorrectly stated that action statement (b) of TS 3.7.1.2 which states "With , i two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours" applied. Several SR0s
stated that TS 3.0.3 applied. The' correct action was Action Statement (c) which states "With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to ( operable status as soon as possible." Most SR0s did not understand the problems associated with shutting down without a reliable CA supply. Several SR0s did not understand that only one TS action statement could be entered at a time for one TS. Many operators did not understand that a controlled , shutdown to mode 3 was not possible without CA flow. Three of six SR0s could not explain the flow reversal phenomenon associated with the loss of one of four RCPs. Several SR0s did not understand that flow
would drop to zero and reverse in the loop. Several SR0s could not predict i the performance of the Reactor Coolant (NC)-loop flow indications. , Four of six SR0s could not describe how the NC loop flow detectors sensed loop fl ow. They did not understand the principle behind the flow detectors. They r did not understand why the pressure sensing taps were physically configured with three low pressure taps and one high pressure tap. All SR0s interviewed were not able to accurately and rapidly complete . enclosure 4.2 of RP-10, " Checklist for Significant Event Notifications", which ! is used to provide information for the NRC Headquarters Operations Officer (H00). Most SR0s indicated that they were not familiar with the information required on the form and had not been trained on its use. Furthermore, the , procedure, RP-10, does not clearly define the parameters desired nor the l criteria for making data selections. Many of the blocks on the second page of - the form contained information that would have been confusing to the NRC , Operations Center H00s and had the potential to mischaracterize the severity l
of the event to NRC Senior Management. - Most SR0s did not understand how to complete many of the data blocks on ' ' the second page of enclosure 4.2 to RP-10. RP-10 does not provide guidance to the SR0s regarding the definition of the information that is ! needed by the NRC Incident Response Center (IRC) Response Team. Misunderstandings could occur if the licensee's interpretation of the . requested data is different than the H00. NUREG-0845 defines the l information that is needed by the NRC IRC Response Teams and its-input i to the various computer codes used to determine offsite dose projection, source term quantification and accident consequence assessment for NRC senior management. Examples of misinterpretation are as follows- , i e i
l l Appendix B 2 - None of the SR0s understood that release " duration" was the total estimated time that the release would continue before termination. Most of the SR0s listed the release duration time as the time from onset of the release to the present time. 1 ' None of the SR0s understood that a steam generator tube rupture - (SGTR) is not considered to be a loss of coolant accident (LOCA) by NRC IRC procedures. All of the SR0s completed both the I " Reactor Coolant Leak" and the " Steam Generator Tube Leak" sections of enclosure 4.2. This would be understood by the IRC Teams to mean that both a SGTR and a LOCA had occurred at the site. l - Several SR0s did not understand the criteria for selecting a l liquid versus gaseous release. Several indicated that both liquid and gaseous releases were occurring for a SGTR with a stuck steam generator (SG) power operator relief valve (PORV). - Several SR0s did not understand the decision criteria for selecting the " Development" of the event. Choices were " sudden" or "long term." One SR0 selected both choices. This information is intended to indicate if the event occurred suddenly or if it had been slowly getting worse over a period of time. - Most SR0s incorrectly thought that the " Volume" block on page 2 referred to the total volume of water release from the start of the event to the present time. The NRC IRC procedures currently use this data for the total volume of the reactor coolant system and SG. l l - Several SR0s reported reactor coolant (NC) activity as the vaius of NC activity prior to the event (i.e., normal NC activity). The SR0s did this even though they recognized that NC activity was significantly elevated over pre-event levels based on process and area radiation monitor readings. . - Most SR0s took a prolonged amount of time (30 to 45 minutes) to complete ) RP-10 enclosure 4.2 and notify the NRC IRC H00. While the maximum time i requirement to notify the NRC is one hour from event declaration, it is i , l unlikely that the crew can afford to have the Shift Supervisor dedicated j l to completing this form during an accident for such a prolonged period of time (30 to 45 minutes). - None of the SR0s telefaxed the form to the NRC IRC as required by McGuire RP-10 step 3.4.1. NRC I&E Information Notice 89-89 states that licensees should telefax the information to the NRC H00 to 4 reduce the time spent verbally passing the data. l - One SRO used the state form for NRC notification in place of the NRC l immediate notification form contrary to RP-10 Subsequent Action 3.4. l The state notification form does not contain the required event ' information to support NRC IRC Teams. t l - . ._ _ _ _ . I
-. .- b Appendix B 3 - One SR0 over-classified the event as a General Emergency on the basis of a failure of two out of three fission product barriers with a potential for the failure of the third barrier. He based his decision on Emergency Action Level (EAL) GE #1 from RP-00 Event 4.1.2, Fuel Damage. He incorrectly concluded that the clad barrier had failed. Although coolant activity levels were elevated over normal activity levels by a factor of X10, the magnitude of the increase in activity was several orders of magnitude less than required to determine that the clad boundary had failed (per RP-00 Event 4.1.2, Emergency Action Level (EAL) for Fuel Damage) and there was no indication of potential for further degradation (all engineered safety feature (ESF) systems had properly actuated and all critical safety functions (CSFs) were satisfied). - Two SR0s stated that offsite Protective Action Recommendations (PARS) were appropriate at the SAE level contrary to the direction provided by RP-03, " Site Area Emergency", sections 2.2 and 3.3. Offsite PARS are only recommended at the General Emergency classification level. ' - Several SR0s did not provide the correct estimation of RCS to SG leakage rate. SR0 estimates varied from 500 gpm (which was close to actual leakage) to under 50 gpm which was significantly lower than the actual leak rate. - Two SR0s did not complete the radiological monitor section of RP-10 , Enclosure 4.2, " Checklist for Significant Event Notifications" as required by RP-10 section 3.3. This information is needed by the NRC Protective Measures Team. All of the SR0s were asked to explain the difference between an ATWT and a failure of RPS. All six SR0s were able to correctly differentiate between the two. Furthermore, the Operations Department issued a memorandum to all licensed operators on March 25, 1988, regarding the ATWT definition. This memorandum clearly states that instrument failures, by themselves, (one bistable in a trip condition while another channel failed) are not necessarily transients. In this case, if the reactor failed to trip, this would be a failure of the reactor protection system and not an ATWT condition. This is in keeping with the ATWT definition in 10 CFR 50.62 as an anticipated operational occurrence followed by the failure of the reactor trip portion of the protection system. This was considered a Training Department strength. . --- . - .
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! ! APPENDIX C MEDICAL RECORDS REVIEW , On August 12, 1992, October 11, 1991, and March 16, 1989, the facility licensee's physician made the determination that each of three operator's i eyesight no longer met the minimum standards required by 55.33 (a)(1) as ' measured by the standards of ANSI /ANS-3.4-1983. Also, on November 27, 1991, ' the facility licensee's physician made the determination that a fourth operator should have been "no solo" because the operator did not meet the standards mentioned above. This failure to notify the Commission of the change in medical status of these operators within 30 days as required by 10 CFR 55.25 j and 10 CFR 50.74 is identified as Violation 50-369,370/93-300-03: Failure to . 2 adhere to the requirements of 10 CFR 55.33 regarding the reporting of the
" medical status of certain licensed individuals. !
" White out" correction fluid was used to alter medical form 08312 (eye i examination results). This is a violation of Nuclear System Directive 702, , " Document Control," which clearly states " Documents should be free of ! corrections whenever practical." When it is necessary to correct a document, one line shall be placed through the information being corrected using black i ink. The initials of the person making the correction and the date shall be i placed at the sta.t or end of the correction. Documents shall not be corrected using opaaueing iluid or correction tape. This discrepancy has been ' corrected. This NRC identified violation is not being cited because the I criteria specified it: Section VII.B of the NRC Enforcement Policy were
satisfied. This non-cited violation is identified as 50-369,370/93-300-04:
q Failure to properly correct an error made on a licensee medical form. This ! ' non-cited violation is also being closed out due to the facility's prompt ! corrective action while the examiners were on site. 1 ! One operator was two months late on getting his biennial medical examination as required by ANSI /ANS-3.4-1983. This violation was licensee identified on August 2, 1989, and prompt corrective action was taken. This violation will - not be subject to enforcement action because the licensee's efforts in e identifying and correcting the violation meet the criteria specified in
Section VII.B of the Enforcement Policy. This non-cited violation is i identified as 50-369,370/93-300-05: Failure to obtain a biennially nuclear ! ' operator medical examination within the required timeframe. This non-cited,
licensee identified violation is also being closed out due to the facility's ! prompt corrective action at the time the facility identified this discrepancy. ' , i
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APPENDIX D REMEDIAL TRAINING REVIEW 10 CFR 55.59 states that records must contain copies of written examinations administered, the answers given by the licensee, and the results of evaluations and documentation of operating tests and any additional training administered in areas in which an operator or senior operator has exhibited deficiencies. A review of all operators (ten individuals) who had failed the last facility administered annual requalification examination showed that the licensee was not adequately documenting individual areas of deficiencies nor adequately documenting the remedial training administered. The facility had less than adequate procedural guidance regarding the conduct and documentation of remedial training. Under the previous requalification examination cycle, the relevant procedure, McGuire Operation Training " Conduct of Nuclear Generation Department (NGD) Administered Annual Operating Requalification Examinations" (MC-0P-TG-16), required that individual requalification examination results be documented on Attachment 7 and the remediation training be documented using a form TSR-10. The facility instructors frequently did not use Attachment 7 to document examination results. They used an outdated Attachment E from an earlier revision to TG-16. Additionally, they did not use TSR-10 to document the remedial training but rather described the training under the "remediation" block of Attachment E. The description of remedial training was too cursory to understand what had been accomplished. In one case where five crew members failed a simulator scenario, the description on all five Attachment E forms consisted of the entry: " Crew failed ASE-15. Critique on ASE 15 and 22. Retook ASE 19 SAT." In each of the five individual cases, the individual passed the simulator evaluation but the crew ft":d the simulator evaluation. No record was provided that indicated the crew competency evaluation, the individual simulator critical tasks (ISCT) missed, or the specific discrepancies that justified a crew failure. ASE-15 and ASE-19 were similar scenarios for a steam generator tube rupture. Interviews with the training staff and operators who had failed an annual examination indicated that remediation was generally limited to a discussion of the specific deficiency that caused the failure. No analysis was conducted to determine whether the failure was due to a general or specific knowledge weakness, lack of mastery of a skill or ability, or a one-time human performance error. Remediation was limited to the task or competency that caused the failure and frequently consisted of self study followed by one or more simulator scenarios that were similar to the failed scenario. Operator participation in this remedial program was required to be documented usirg a " Training Attendance Sheet" [ Form 18886]. In most cases, this form wn not used. When the form was used, the entry consisted of a single line entry for each operator with a total number of hours listed. The block that requires the operator to initial completion of the training was not initialled -
_ _ _ Appendix D 2 by any of the remediated operators and the " Actual Hours" entry was frequently a large number of hours that could not be accomplished during one session (such as 20 hours). This form was not being used to keep track of actual training conducted and the operator was not required to verify that he had trained for the time required. The licensee recognized that the remediation program documentation procedures required improvement as evidenced by a recent set of written guidelines entitled " Guideline for Documenting and Communicating Annual Requalification Examination Results." This memorandum was issued for comment in the Training Department and provided an additional level of detail and administrative requirements to ensure that documentation of the requalification examination results were clear. The memorandum stated that a " Proposed Remediation Plan" will be submitted and approved by the Director of Operations Training. The - requirements for this remediation plan were not specified in this memorandum. , This failure to adhere to the requirements of 10 CFR 55.59 regarding the evaluations and documentation of operating tests for operators who have exhibited deficiencies is identified as non-cited violation 50-369,370/93- 300-06: Failure to adhere to the requirements of 10 CFR 55.59 regarding the evaluations and documentation of operating tests for operators who have
exhibited deficiencies. This NRC identified violation is not being cited because the criteria specified in Section VII.B of the NRC Enforcement Policy were satisfied. However, the licensee's corrective action will be reviewed during a future inspection. The facility restored two failed crews to shift duties without passing an ASE retake examination consisting of at least two scenarios. The requirements of OP-MC-TG-16 stated that operators failing the simulator portion of their i annual requalification examination were to be remediated and were required to l pass another ASE examination. The procedure does not define what constitutes l an ASE. Some of the ASE retakes included two scenarios and others only one. l Many of the single ASE retakes did not involve weaknesses identified on the previous examination. This inconsistency was identified as a Training Department weakness. . - , -
APPENDIX E LIST OF PROCEDURES REVIEWED PROCEDURE TITLE REV. NO. AP/1/A/5500/07 Loss of Normal Power to Either 2 Case II through IETA or 1ETB Step 13
AP/1/A/5500/01 Steam Leak 3 AP/1/A/5500/10 NC System Leakage Within the 1 Capacity of Both NV Pumps AP/1/A/5500/ll Pressurizer Pressure Anomalies 0 AP/1/A/5500/03 50% Load Rejection 0 Case II AP/1/A/5500/17 ASP Alignment 0 0P/1/A/6350/02 Diesel Shutdown 52 Encl 4.2 OP/0/A/6350/01A Static Inverter 2EVIA Operation 27 Encl 4.5 EP/1/A/5000/01 Reactor Trip or Safety Injection 11 EP/1/A/5000/03 Steamline Break Outside Containment 14 EP/1/A/5000/04 Steam Generator Tube Rupture 10 EP/1/A/5000/3.1 SI Termination Following 14 Excessive Cooldown EP/1/A/5000/ll.1 Response to Nuclear Power 1 Generation /ATWT EP/1/A/5000/13.1 Response to Loss of Secondary 4 through Step 13 Heat Sink RP/00 Classification of Emergency 0 RP/03 Site Area Emergency 0 RP/04 General Emergency 0 RP/10 NRC Immediate Notification 0 Requirements RP/11 Conducting a Site Assembly 0 or Evacuation RP/15 Notifications to States and Counties 0 _. _ _ I
O ge cv,:. 93-sw - jf93 Alu Pakr ,A, ,, , ' ' ' . RD EX A M . , mu 0F SPECi,iC,,,,, QUESTIONS GROUPED BY TOPIC /LV6ED 4 3D 3
CNT CUESTIONS . QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOC IMPORTANCE TASK ID MINUTES MAT ONLY? CATE PLAN COVERED SYS MODE No. R0 SRO ID ACNT-VD RC1 1.0 3 MC 05/11/93 CNT-VQ 91-1 103 000 A1.01 3.7 4.1 3340 TOTAL QUESTIONS FOR CNT TOPIC: 1 TOPIC PolNTS 1.00 TOPIC TIME 3.00 CP QUESTIONS QUESTION PolNTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? CATE PLAN COVERED SYS MODE No. RO SRO ID ACP-AD-RD1 1.0 5 MC 06/30/93 CP-AD 89-1 000 068 EA2.09 4.1 4.3 5307 l TOTAL QUESTIONS FOR CP TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 5.00 ECC QUESTIONS
QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT G CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID ' AECC-NS-R02 1.0 3 MC 06/30/93 ECC-NS 91-2 026 000 A1.01 3.9 4.2 2313 TOTAL QUESTIONS FOR ECC TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 IC QUESTIONS i 1 l QUESTION POINTS TIME FOR- SRO REV. LESSoh SEGMENT KA CATALOG IMPORTANCE TASK ID MlWUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID A!C-ICM-R01 1.0 3 MC 06/30/93 IC-ICM 91-3 000 074 EK1.01 4.3 4.7 2318 AIC-IRE R06 1.0 4 MC 06/30/93 IC-!RE 92 4 001 050 A2.01 3.7 3.9 2315 Alt-IRX ROS 1.0 3 MC 06/30/93 1C-1Rx 93-2 002 020 KS.D9 3.6 3.9 3310 l TOTAL QUESTIONS FOR IC TOP!C: 3 TOPIC POINTS 3.00 TOPlc TIME 10.00 PSS QUESTIONS 1
QUESTION PolNTS TIME FOR- SRO REY. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID APSS-KC-RO1 1.0 3 MC 06/30/93 PSS-KC 93 2 008 000 K4.01 3.1 3.3 3313 TOTAL QUESTIONS FOR PSS TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 I ) PAGE(1) j
__ _ _ _ .-- . - _ _ _ . . __ . O - ' . . ll > l l i . RT QUESTIONS l CUESTION PolWTS TIME FOR- SRO REV. l LESSON SECMENT KA CATALOG IMPORTANCE TASK j ID MINUTES MAT ONLY? DATE l PLAN COVERED SYS MODE No. R0 SRO ID ART-RB-RO1 1.0 3 MC 06/30/93 RT RB 93-2 001 010 A4.04 3.5 4.1 ! 2309 i , , j 3 TOTAL QUESTIONS FOR RT TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 i. l SSE02 QUESTIONS QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID , f SSE-02-R01 1.0 3 MC 06/30/93 IC4NB 92-1 015 000 A2.01 3.5 3.9 j 2302 f SSE-02-R03 1.0 3 MC 06/30/93 STM-IDE 91-2 041 020 A3.02 3.3 3.4 6309 , a SSE-02-R04 1.0 3 MC 06/30/93 CF-CF 92-1 059 000 A2.01 3.4 3.6 f 5310 i SSE-02-R05 1.0 3 MC 06/30/93 CF-LF 93-1 059 000 A2.07 3.0 3.3 l f 6312 i SSE-02-R06 1.0 3 MC 06/30/93 CF-CA 93-1 061 000 A2.04 3.4 3.8 l ] 3314 i ] SSE-02-R07 1.0 3 MC 06/30/93 PS-!LE 93-2 011 000 A2.11 3.4 3.6 , f i 6305 SSE-02-R09 1.0 3 MC 06/30/93 PS-IPE 93 2 010 000 K3.01 3.8 3.9 6303 I j SSE 02-R11 1.0 3 MC 06/30/93 PS-WV 93-1 000 009 EK3.19 3.6 3.9 , 3305
L SSE-02-R13 1.0 3 MC 06/30/93 CF-IFE 93-2 035 010 A1.01 3.6 3.8 E i 6308 ! SSE-02-R15 1.0 3 MC 06/30/93 PSS-RN 91-3 076 000 A3.02 3.7 3.7 i 3333 SSE-02 R17 1.0 3 MC 06/30/93 IC-!PE 93-1 012 000 K4.02 3.9 4.3 3
! 2301 , 1
! SSE-02 R18 1.0 4 MC 06/30/93 IC-EDA 93-2 014 000 K4.06 3.4 3.7 2307 TOTAL QUESil0NS FOR SSE02 TOPIC: 12 TOPIC PolNTS- 12.00 TOPIC TIME 37.00 TOTAL QUESTIONS 20 TOTAL Po!NTS 20.00 TOTAL TIME 64.00 3
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- _______________ ___ _____ _ . , PRINT NAME PROG. ID: OP-MC-LRQ SIGNATURE 1993(NRC)(RO)F (A) , SSN Prepared by:_ . . SHIFT DATE REVIEW / INITIAL LICENSED REQUAL ANNUAL EXAMINATION TEST # 1993(NRC)(RO)PART(A) Total Points: 20.00 Total Questions: 20
- INSTRUCTIONS ***
I 1. Use NO. 2 PENCIL only. 2. Fill in the appropriate information on the TEST COVER SHEET. 3. Ask the test monitor instructor about any questions which are not clear to you. 4 Circle the correct answer on each question of this exam. 5. When finished with exam, read the "non-compromise statement" and sign the TEST COVER SHEET. NON-COMPROMISE STATEMENT: "My Signature on this form is my declaration that the responses given on the attached test or exam are entirely my own. 1 It further declares that I am aware that I am subject to termination from the training program immediately and in addition, will be subject to further disciplinary action up to and including discharge from the company for cheating and/or compromising on exams / tests / quizzes."
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SCENARIO
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1 PART "A" AIR ADD & RELEASE SYS ACNTVQR01
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1 Pt(s) Given:
i 1. A VQ Release has been made. 2. The VQ Release was started at 0705. 3. The VQ Release was stopped at 0730. 4. Initial Containment Pressure was 0.20 psig. ' 5. Final Containment Pressure was 0.12 psig. l 6. The VQ Totalizer is inoperable. i Vnat was the TOTAL VOLUME released? A. 185 Cubic Feet B. 6505 Cubic Feet C. 6690 Cubic Feet i D. 6875 Cubic Feet
! . ANSVER: D ' MISCINFO: , t REFERENCES: OP/1/A/6450/17
LESSON: OP-MC-CNT-VQ TASK: MO-3340 ' OBJECTIVES: LPRO OBJ: 5 TIME: 4 MINUTES LPSO OBJ: 5 l KA103000 A1.01 3.7/4.1 REV. DATE: 05/11/93 1 . i . - . . - ,. - . _ -
O O . , e i " 2 PART "A" STBY SHUTDOWN FAC. ACPADR01 ! .
I 1 Pt(s) Civen the following information: 1) You have completed transfer of control to the Standby Shutdown Facility during a Fire Event. 2) You are in the process of verifying Natural Circulation. 3) These are the current Incore T/C readings:
T/C #1 - 562*F T/C #2 - 559'F " T/C #3 - 564*F T/C #4 - 560*F T/C #5 - 562*F 4) The Reference Junction Temp. Deviation - - 3*F i Determine the " Corrected Incore Temperature" at the SSF in , order to determine the status of subcooling. l A. 556*F B. 561*F C. 564*F D. 567'F ANSWER: B MISCINFO: i . REFERENCES: AP/1/A/5500/24 . LESSON: OP-MC-CP-AD TASK: MO-5307 OBJECTIVES : LPRO OBJ: 3 TIME: 5 MINUTES LPSO OBJ: 3 KA000068 EA2.09 4.1/4.3 REV. DATE: 06/30/93 . _ ,
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, i 1 Pt(s) A LOCA has occurrel and containment pressure reached a
maximum of 4.5 psig. Containment spray initiated, as > designed, and containment pressure has been reduced to 0.2 psig. Containment Spray has NOT been " Reset". , Which one of the following most accurately describes the ! present status-of containment spray? I i A. Containment Spray is off and will NOT automatically restart until it has been " Reset" i i r B. Containment Spray is off and will automatically restart if containment pressure rises to greater than 0.8 psig C. Containment Spray is off and will NOT automatically restart, unless containment pressure rises to greater j than 3.0 psig , , D. Containment Spray should still be in operation ANSWER: B MISCINFO: ! REFERENCES: OP-MC-ECC-NS ! LESSON: OP-MC-ECC-NS TASK: MO-2313 , OBJ ECTIVES: LPRO OBJ: 3 TIME: 3 MINUTES LPSO OBJ: 3 l KA026000 A1.01 3.9/4.2 REV. DATE: 06/30/93 i i
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j AICICMR01 l .
1 Pt(s) If the NCPs are tripped following a LOCA and the break has
been isolated, which one of the following situations provides
' the highest subcooling? PZR PRESSURE HOT LEG TEMP COLD LEG TEMP , , ' A. 600 500 470 . ! ' j B. 800 535 510 .i { C. 1000 525 520 i 4 D. 1200 570 560 , ANSWER: C J MISCINFO: , , 1 , 4 i REFERENCES: Data Book ! 4 LESSON: OP-MC-IC-ICM TASK: MO-2318 , , l OBJECTIVES: LPRO OBJ: 11 TIME: 3 MINUTES j LPSO OBJ: 11 l d , , KA000074 EKl.01 4.3/4.7 REV. DATE: 06/30/93 , . - E & 4 4 j , - . . , _ . , _ . . - . _ , , . . . . - _ - . ..-..,_,,ii
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5 PART "A" ROD CONTROL AICIRER06 ' ' . ! ! 1 Pt(s) Unit 1 is at 86% RTP, increasing load at 2 MWe/ min, when i the " Rod Control Urgent Failure" (LAD 2 A-10) annunciator alarms. Which one of the following would prevent moving the ! Controlling Bank in the " Bank Select" mode in order to maintain T-ave - T-ref? A. The " Urgent Failure" exists in Power Cabinet " LAC" B. The " Urgent Failure" exists in Power Cabinet "2AC" C. The " Urgent F'ailare" exists in Power Cabinet "lBD" D. The " Urgent Failure" exists in Power Cabinet "SCDE" l ANSWER: C MISCINFO: REFERENCES: OP-MC-IC-IRE LESSON: OP-MC-IC-IRE TASK: MO-2315 , OBJECTIVES: LPRO OBJ: 10 TIME: 4 MINUTES
j LPSO OBJ: 10 I KA001050 A2.01 3.7/3 9 REV. DATE: 06/30/93 l t . I
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_ _ ~ O O ! ' 6 REACTOR CONTROL AICIRXR05 I ~ 1 Pt(s) Given the following conditions: 1) Reactor power is 60% , 2) Loop "C" Delta-T indicates LOW 3) Loop "C" Tave indicates HIGH k'hich one of the following RTD failures caused these indications? A. T-cold failed high
B. T-cold failed low i C. T-hot failed high D. T-hot failed low i ANSk'ER: A & MISCINFO: REFERENCES: OP-MC-IC-IRX , LESSON: OP-MC-IC-IRX TASK: MO-3310 ! l OBJECTIVES: LPRO OBJ: 3 TIME: 3 MINUTES l l LPSO OBJ: 3 l t KA002020 K5.09 3.6/3.9 REV. DATE: 06/30/93 ! l l N , - . - - - - - ~.-...,~.-_s-... . - . - . - - - . . ~ .
[- 1 r i O O . . 7 PART "A" COMPONENT COOLING APSSKCR01 j . i
1 Pt(s) Given the following conditions: ] 1) Reactor power is 100% l 2) KC pumps B1 and B2 are supplying system loads ' 3) Safety injection signal is received i Which one of the following describes the response of the KC , system? , 4 A. KC pumps 31 and B2 will continue to run, then KC pumps , Al and A2 will be sequenced on. 4 B. KC pumps B1 and B2 will be load shed, then KC pumps Al, , A2, B1 and B2 will be sequenced on l C. KC pumps B1 and B2 will be load shed, then KC pumps Al and A2 will be sequenced on, KC pumps B1 and B2 will not be sequenced on . D. KC pumps B1 and B2 will continue to run, KC pumps Al and A2 will not be sequenced on ANSWER: A l MISCINFO: . I REFERENCES: OP-MC-PSS-KC \\ LESSON: OP-MC-PSS-KC TASK: MO-3331 s OBJECTIVES : LPRO OBJ: 5 TIME: 3 MINUTES i LPSO OBJ: 5 i i KA008000 K4.01 3.1/3.3 REV. DATE: 06/30/93
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t 8 PART "A" REACTIVITY BALANCES ! ARTRBR01 i ! ! - i 1 Pt(s) Unit 1 is tripped due to a Loss of All NC Pumps. The i following conditions exist: > 1) A Cooldown to 250 degrees F must be conducted 2) Current T-ave is 557 degrees F 3) Current Core Burnup is 100 EFFD 4) The NC Pumps will NOT be available for the Cooldown 5) Current boron concentration is 1125 ppm Determine the minimum boron concentration required to 3 l cooldown to 250 degrees F. A. 1125 ppm , B. 1235 ppm l
C. 1301 ppm - > D. 1411 ppm , ANS'w'ER : D MISCINFO: f I REFERENCES: Data Book EP/1/A/5000/1.1 , LESSON: OP-MC-RT-RB TASK: MO-2309 -, t OBJECTIVES: LPRO OBJ: 10 TIME: 3 MINUTES l LPSO OBJ: 10 l ! l KA001010 A4.04 3.5/4.1 REV. DATE: 06/30/93 { , b , 1 1 ! ! ! i _ ,
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. O O . [ 9 SCENARIO QUESTIONS ' l SSE02R01 , ,
! ' l 1 Pt(s) Based on present plant conditions, which one of the following l describes how the Protection System would respond? l l i A. If power were lost to EKVA, a Reactor Trip would occur , B. If Intermediate Range 1NI-36 were to fail high, a Reactor l Trip would NOT occur ! C. If only the "A" Train " Source Range Select Switch" were turned to the " Reset" position, a Reactor Trip would NOT
occur
I , D. If the control power fuses to Source Range Instrument N31 ! I were to blow, a Reactor Trip would occur I l ANSWER: A MISCINFO: l REFERENCES: OP-MC-IC-ENB . t ,
LESSON: OP-MC-IC-ENB TASK: MO-2302 i 5 OBJ ECTIVES : LPRO OBJ: 11 TIME: 3 MINUTES l LPSO OBJ: 11 i ! KA015000 A2.01 3.5/3.9 REV. DATE: 06/30/93 ! t ! l t ! ! I l
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10 SCENARIO QUESTIONS j SSE02R03 < 4 , 1 Pt(s) Based on present plant conditions, which one of the following ! vould occur if the Steam Dump Select switch were placed in j the "T-avg Mode" of operation? A. The condenser steam dumps would modulate to maintain , , j T-avg - T-ref ! i B. The condenser steam dumps would modulate to maintain , j T-avg five degrees higher than T-ref
J l C. The condenser steam dumps would close, T-avg would ', increase until the SM PORV's open and maintain T-avg )
D. The condenser steam dumps would continue to modulate to maintain SM pressure at approximately 1092 psig ' ANS'.'ER : C MISCINFO: 4 t . REFERENCES: OP-MC-STM-IDE j i i i j LESSON: OP-MC-STM-1DE TASK: MO-6309
t ! l OBJ ECTIVES: LPRO OBJ: 1 TIME: 3 MINUTES ! LPSO OBJ: 1
i i e KA041020 A3.02 3.3/3.4 REV. DATE: 06/30/93 i
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.. . - . . . - . - - O O l 11 SCENARIO QUESTIONS - SSE02R04 , ! b . ' 1 Pt(s) Based on present plant conditions, which one of the following would occur if the AMSAC for CF Valves " Unblock" pushbutton were depressed? (Consider short term effects only, within 1 minute)
A. The Main Turbine would trip; the Reactor would NOT trip p B. The Main Turbine would trip; the Reactor would trip ! ! C. The Main Turbine would NOT trip; the Reactor would NOT trip . ! D. The Main Turbine would NOT trip; the Reactor would trip ,
s ANSER : A l MISCINFO: REFERENCES: OP-MC-CF-CF > LESSON: OP-MC-CF-CF TASK: MO-5310 OBJECTIVES: LPRO OBJ: 15 TIME: 3 MINUTES
LPSO OBJ: 15 ' KA059000 A2.01 3.4/3.6 REV. DATE: 05/30/93 l m* g -. g ,.,-im.r- w - .* y-1,-
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' a , , . 12 SCENARIO QUESTIONS , ' SSE02R05 i I ' > ! 1 Pt(s) Based on present plant conditions, which one of the following ' would occur if a Feed Pump lube oil leak necessitated , ' immediately stopping both CF Pump Turbine MOP Pumps for "lB" CF Pump? A. The Main Turbine would trip; both Motor Driven CA Pumps l would start; the Reactor would trip; the "lB" CF Pump , suction valve ICM-272 would close ' i B. The Main Turbine would trip; both Motor Driven CA Pumps would start; the Reactor would NOT trip; the "lB* CF l Pump suction valve ICM-272 would close j C. The Main Turbine would NOT trip; both Motor Driven CA i Pumps would start; the Reactor would NOT trip; the "lB" CF Pump suction valve ICM-272 would close j l D. The Main Turbine would trip; the Motor Driven CA Pumps , would NOT start; the Reactor would trip; the lB" CF Pump " suction valve ICM-272 would NOT close i i ANS*JER: B I MISCINFO: REFERENCES: OP-MC-CF-LF 1 1 LESSON: OP-MC-CF-LF TASK: MO-6312 OBJECTIVES: LPRO OBJ: 20 TIME: 3 MINUTES LPSD OBJ: 20 ' KA059000 A2.07 3.0/3.3 REV. DATE: 06/30/93 1 j 1 , y _pm- v. , , _ ,. ~ . . . e:. +%-
~ O O i . 13 SCENARIO QUESTIONS SSE02R06 . , , 1 Pt(s) '*hich one of the following describes how the CA System will
respond to present plant conditions? A. CA flow to S/G "C" can NOT be throttled using CA-48, unless both Train "A" & "B" "CA Modulating Valve Reset" pushbuttons for the Turbine and Motor Driven Pumps are depressed B. CA flow to S/G "D" can NOT be throttled using CA-36, unless both Train "A" & "B" "CA Modulating Valve Reset; , Turbine" pushbuttons are depressed C. CA flow to S/G "A" can NOT be throttled using CA-64, unless the Train "A" "CA Modulating Valve Reset; Turbine" pushbutton is depressed 1 , D. CA flow to S/G "B" can be throttled using CA-52, without ! any of the "CA Modulating Valve Reset" pushbuttons being depressed
! ANS*JER: D MISClNFO: ! REFERENCES: OP-MC-CF-CA LESSON: OP-MC-CF-CA TASK: MO-3314
I , OBJECTIVES : LPRO OBJ: 13 TIME: 3 MINUTES LPSO OBJ: 13 KA061000 A2.04 3.4/3.8 REV. DATE: 06/30/93 i , j - - . .
,_ .. . O O > J 14 SCENARIO QUESTIONS . SSE02R07 ' ' , 1 Pt(s) Based on present plant conditions, which one of the following would occur if the " Pressurizer Level Control Select" switch l were placed in the "3 & 2" position? (Assume no other ! operator action is taken.) t A. The letdown isolation valves NV-2 and NV-458 will close; ' the reactor will NOT trip B. The letdown isolation valves NV-1 and NV-458 will close; , the reactor will eventually trip due to a pressurizer t high pressure reactor trip signal C. The letdown isolation valves NV-1 and NV-458 will close: the reactor will NOT trip D. The letdown isolation valves NV-2 and NV-458 will close; the reactor will eventually trip due to a pressurizer high level reactor trip signal . , ! l ANS'JER: A ! l , MISCINFO: , REFERENCES: OP-MC-PS-ILE , LESSON: OP-MC-PS-ILE TASK: MO-6305 - OBJECTIVES : LPRO OBJ: 1 TIME: 3 MINUTES t , LPSO OBJ: 1 - ' ! ! KA011000 A2.11 3.4/3.6 REV. DATE: 06/30/93
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O O . 15 SCENARIO QUESTIONS j SSE02R09 l ' i 1 Pt(s) Based on present plant conditions, which one of the following i would occur, if the output of the " Pressurizer Pressure Master" were to fail to the "100%" position? (Assume no j operator action is taken.) A. All pressurizer heaters will go off; PORV NC-34 will open; pressurizer sprays will open; the reactor will i trip due to a low pressurizer pressure reactor trip ' signal of 1945 psig B. All pressurizer heaters will go off; PORV NC-34 will open and then will reclose; pressurizer sprays will open; the reactor will not trip until pressurizer pressure decreases to less than 1845 psig due to the Safety Injection signal C. All pressurizer heaters will go on; all. PORV's will be blocked from opening; pressurizer sprays will close; the reactor will trip due to a high pressurizer pressure reactor trip signal of 2385 psig D. All pressurizer heaters will go on; PORV NC-32 and NC-36 will cycle open and closed; pressurizer sprays will close; the reactor will not trip due to any pressurizer pressure reactor trip signal ANS'JER: B i , MISCINFO: r REFERENCES: OP-MC-PS-IPE i ' , LESSON: OP-MC-PS-IPE TASK: MO-6303 OBJECTIVES: LPRO OBJ: 3 TIME: 3 MINUTES LPS0 OBJ: 3 KA010000 K3.01 3 8/3.9 REV. DATE: 06/30/93 , . - - _ , . -, . - .
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16 SCENARIO QUESTIONS - l SSE02R11 0 t I .
- ! 1 Pt(s) Based on present plant conditions, which one of the following , , j would occur if "A" Train Safety Inj ection were manually } initiated? (Assume no operator action is taken.)
A. CA flow to the "A" S/G would increase to greater than 400 gpm
, i ' B. The "A" D/G would start, but the ETA Emergency Breaker would NOT close 1 C. The "B" NV Pump would remain aligned to the VCT D. The orifice isolations would close, isolating Letdown , ANSWER: D ' MISCINFO- i i i i
i REFERENCES: OP MC-PS-NV
i i - LESSON: OP-MC-PS-NV TASK: MO-3305 j ! i OBJECTIVES : LPRO OBJ: 3 TIME: 3 MINUTES LPSO OBJ: 3 ) J
KA000009 EK3.19 3.6/3.9 REV. DATE: 05/30/93 1 i d < l 1 i !. . . - . . -. . . .
) . O O . . 17 SCENARIO QUESTIONS SSE02R13
, . ! 1 Pt(s) Based on present plant conditions, which one of the following correctly evaluates the S/G NR Level-Instrumentation status? A. All S/G Narrow Range Level Channels are operable , B. S/G Narror Range Level Channel 2 of "B" S/G must be declared inoperable C. S/G Narrow Range Level Channtl 2 of "C" S/G must be declared inoperable i D. Both S/G Narrow Range Level Channel 2 of "B" S/G and S/G Narrow Range Level Channel 2 of "C" S/G must be declared inoperable ! ANSk'ER: C t MISCINFO: l REFERENCES: OP-MC-CF-IFE , i i LESSON: OP-MC-CF-IFE TASK: MO-6308
! OBJECTIVES: LPRO OBJ: 13 TIME: 3 MINUT.?S I LPS0 OBJ: 13 KA035010 A1.01 3.6/3.8 REV. DATE: 06/30/93 i i i l 1 l - - . . . . ._. _ _. I
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. 18 SCENARIO QUESTIONS SSE02R15 "t i f l . ' 1 Pt(s) Based on present plant conditions, which one of the following , correctly describes the RN System response if PCB's 8, 9, ll, " and 12 were to open? (Assume no operator action is taken.) A. The "A" Train KC Heat Exchanger vould be provided with RN flow [ B. The "B" Train of RN would remain aligned to the Lower Level Intake C. The RN AB Non-essential header would not lose RN flow D. The NC Pump Motor Air Coolers would lose RN Flow . ANS 'ER: D l r MISCINFO:
! REFERENCES: OP-MC-PSS-RN t i LESSON: OP-MC-PSS-RN TASK: MO-3333 ' l OBJECTIVES : LPRO OBJ : 9 TIME: 3 MINUTES LPSO OBJ: 9 KA076000 A3.02 3.7/3.7 REV. DATE: 06/30/93 [ t 4
- . . . - - - - _ - - , O O 19 SCENARIO QUESTIONS , SSE02R17 ' ! . 1 Pt(s) Based on present plant conditions, which one of the following . correctly describes plant response? i . A. If stator cooling water differential pressure decreases to under 15 psid for over 45 seconds, the main turbine will trip B. If the "A" CM Booster Pump were stopped, a reactor trip would NOT occur within 60 seconds ' C. If turbine impulse pressure Channel I were to fail high, ' j I actual pressurizer level will increase l 1 i D. If the ITB and 1TC mode select switches are placed in manual and then the normal breaker " Trip" pushbuttons ' were depressed, a reactor trip would occur within ! 60 seconds ANS'JER : B ' ! MISCINFO: ' , REFERENCES: OP-MC-IC-IPE { l t LESSON: OP-MC-IC-IPE TASK: MO-2301 t OBJECTIVES : LPRO OBJ: 1 TIME: 3 MINUTES LPSO OBJ: 1 KA012000,K4.02 3.9/4.3 REV. DATE: 06/30/93 l l -
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. . 20 SCENARIO QUESTIONS c l SSE02R18 l
l ! . l 1 Pt(s) Based on present plant conditions, which one of the following
correctly describes the position of control rod H-87 A. Control Rod H-8 is Fully Withdrawn or Ejected B. Control Rod H-8 is aligned with the other Bank "D" rods C. Control Rod H-8 is Fully Inserted l D. Control Rod H-8's position cannot be determined with the ' l information displayed i l ANS'.'ER : D MISCINFO: ! REFERENCES: OP-MC-IC-EDA !
LESSON: OP-MC-IC-EDA TASK: MO-2307 . I i OBJECTIVES : LPRO OBJ: 6 TIME: 4 MINUTES l LFSO OBJ: 6 l i KA014000 K4.06 3.4/3. REV, DATE: 06/30/93
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m aa& f* Ar 8 O yn # c- > i AdC 28 A A1 i TABLE OF SPECIFICATIONS
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QUESTIONS GROUPED BY TOPIC [[ g/) 3 f p g7 g Q 07/01/93 ' AP QUEST 10h5 QUESTION POINTS TIME FOR- SRO REY. LESSON SEGMENT -KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. RO- SRO ID AP-AP08-RG4 1.0 4 MC 05/12/93 PS-NCP 92-1 000 015 EA1.23 3.1 3.2 7307 i AP-AP17-R01 1.0 3 MC 07/01/93 CP-SS 92-1 000 005 EA2.03 3.5 4.4 7315 AP-AP19-R02 1.0 4 MC 05/12/93 PS-ND 93-1 000 025 EA1.01 3.6 3.7 i ! 7317 TOTAL QUESTIONS FOR AP TOPIC: 3 TOPIC POINTS 3.00 70ptC TIME '11.00 i l CNT QUESTIONS QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? CATE PLAN COVERED SYS MODE No. RD SRO ID l CNT-VX-RDS 1.0 4 MC 05/12/93 CNT VX 91 2 028 000 A2.01 3.4 3.6 , 3341 I TOTAL QUESTIONS FOR CNT TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 EP QUESTIONS j QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID MINUTES MAT ONLY? CATE PLAN COVERED SYS MODE NO. RD SRO ID -l l EP-EPF-RO7 1.0 4 MC 07/01/93 EP-EPF 92-2 000 011 G.11 4.3 4.5 i ! 8309- TOTAL QUESTIONS FOR EP TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 EP01 QUESTIONS QUEsil0N POINTS TIME FOR- SRP REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. R0 SR0 ID j f EP-EP01-R02 1.0 4 MC 07/01/93 EP-EP1 92-5 000 017 EK1.01 4.4 4.6 l 8301 I EP-EP01-R18 1.0 5 MC 07/01/93 EP-EP1 92-5 000 007 G.12 3.8 3.9 I 8301 l ! TOTAL QUESil0NS FOR EP01 TOPIC: 2 TOPlc PolNTS 2.00 TOPIC TIME 9.00 ! l EP02 QUESTIONS , ! QUESTION POINTS TIME FOR- SR0 REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK , ID MINUTES MAT ONLY? DATE PLAN COVERED STS MODE NO. R0 SRO ID l l EP-EP02-R10 1.0 4 MC 07/01/93 EP EP2 92-5 000 011 -G.12 4.0 4.1 8306 4610 TOTAL QUESTIONS FOR EP02 TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 PACE (1) ! . . - - -, a
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EP03 QUESTIONS ! QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l l ID MINUTES NAT ONLY? DATE PLAN COVERED SYS MODE NO. RD SRO ID
i j EP-EP03-R02 1.0 3 MC 05/12/93 EP EP3 93 2 000 040 G.12 3.8 - 4.1
i 8303 l EP-EPC3 R03 1.0 3 MC 05/12/93 EP EP3 93 2 004 000 A4.04 3.2 3.6 8303 3305 TOTAL QUESTIONS FOR EP03 TOPIC: 2 TOPlc POINTS 2. 0t, TOPIC TIME 6.00 .. EP09 QUESTIONS i QUESTION POINTS TIME FOR-l SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK [ ID MINUTES MAT lONLY? DATE PLAN COVERED SYS MODE NO. RO SRO ID EP-EP09 R01 1.0 4 MC 06/21/93 EP EP09 92-4 000 055 G.12 3.9 4.0 ' 8308 l ' TOTAL QUESTIONS FOR EP09 TOPIC: 1 TOPIC POINTS 1.00 TOPlc TIME 4.00 l 1 EP12 QUESTIONS l QUEST 10W PolNTS TIME FOR. SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK k ID MINUTES MAT ONLY? DATE PLAN COVERED STS MODE NO. R0 SR0 ID l { EP-EP12 R03 1.0 3 MC 05/12/93 EP-EP12 92-2 000 074 G.12 4.3 4.4 8311 j TOTAL QUESTIONS FOR EP12 TOPIC: 1 TOPlc POINTS 1.00 TOPlc TIME 3.00 j EP13 QUESTIONS QUESTION PolNTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? CATE PLAN COVERED SYS MODE NO. R0 SRO ID
! EP EP13-R01 1.0 4 MC 05/12/93 EP-EP13 93 1 000 0 74 G.12 4.3 4.4 i 8312 { TOTAL QUES 110NS FOR EP13 TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 ' i PS QUESTl0NS QUESTION PolWTS TIME FOR- SRO REV. l LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE g PLAN COVERED SYS MODE No. RO- SRO ID PS-NC-R03 1.0 4 MC 07/01/93 PS-NC 92-1 002 000 C.11 3.3 4.0 4332 ' , 5304 6310 PS-ND R04 1.0 4 MC 05/12/93 PS-ND 92 2 005 000 c.11 3.1 3.8 3311 TOTAL QUESTIONS FOR PS TOPIC: 2 TOPIC PolNTS 2.00 TOPIC TIME 8.00 PACE (2) l l - . , _ - _ . - _ . . . - , - . . . - - . - . . . ~ . . . . -. , . - .
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RT QUESTIONS l t QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLT7 DATE PLAN COVERED STS MODE NO. RO SRO ID , ! RT-RP ROS 1.0 4 MC 07/01/93 RT-RP 90 2 001 010 A3.02 4.1 3.7
! 6302
I TOTAL QUESTIONS FOR RT TOPIC: 1 TOPIC PolNTS 1.00 TOPIC TIME 4.00 l TA QUESTIONS j , QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG- IMPORTANCE TASK i ID MINUTES MAT ONLT7 DATE PLAN COVERED. STS MODE NO. R0 SRO 10 l v TA-PTS-R02 1.0 3 MC 05/12/93 TA-PTS 92-4 000 040 C.12 3.8 4.1 t 8313 ! 4615 i TOTAL QUESTIONS FOR TA TOPIC: 1 TOPIC POINTS 1.00 TOPlc TIME 3.00 i TOTAL QUESTIONS 17 TOTAL POINTS 17.00 TOTAL TIME 64.00
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. . l PRINT NAME PROG. ID: OP-MC-LRQ ! l SIGNATURE 1993(NRC)(RO)P ) ! l SSN Prepared by: SHIFT DATE REVIEW / INITIAL , LICENSED REQUAL ANNUAL EXAMINATION TEST # 1993(NRC)(RO)PART(B) Total Points: 17.00 Total Questions: 17
- INSTRUCTIONS ***
, ' l. Use NO. 2 PENCIL only. 2. Fill in the appropriate information on the TEST COVER SHEET. 3. Ask the test monitor instructor about any questions which are not clear to you. 4 Circle the correct answer on each question of this exam. 5. When finished with exam, read the "non-compromise statement" and sign the TEST COVER SHEET. , l i l 1 NON-COMPROMISE STATEMENT:
"My Signature on this form is my declaration that the responses given l on the attached test or exam are entirely my own. It further declares that I am aware that I am subject to termination from the training program immediately and in addition, will be subject to further disciplinary action up to and including discharge from the company for cheating and/or compromising on exams / tests / quizzes." l l 1 l I _ _ _ _
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1 EP 09 SERIES EPEP09R01 l , 1 Pt(s) Given the following: ! 1) lETA and lETB both have a zero bus voltage 2) The Unit was at 100% just prior to the event
3) Current core burnup is 280 EFFD 4) Pressurizer level is slowly decreasing 5) Intact S/G's are being depressurized to minimize NC System inventory loss Which of the following set of conditions would permit continuing the S/G depressurization process? A. An NC Pressure of 210 psig, "C" S/G Pressure of 155 psig. NC System Cold Leg Temperature of 375 degrees F. All S/G Levels are 3% and slowly decreasing, Total CA flow is 1099 gpm B. An NC Pressure of 300 psig, "B" S/G Pressure of 175 psig, NC System Cold Leg Temperature of 370 degrees F. All S/G Levels are 27% and being maintained constant, Total CA fic is 900 gpm C. An NC Pressure of 190 psig, "C" S/G Pressure of 175 psig, NC System Cold Leg Temperature of 373 degrees F, 1 All S/G Levels are 21% and being maintained constant, Total CA flow is 850 gpm D. An NC Pressure of 225 psig, "B" S/G Pressure of 190 psig, ' NC System Cold Leg Temperature of 375 degrees F, j All S/G Levels are 7% and being maintained constant. 1 Total CA flow is 850 gpm ANSWER: D MISCINFO: REFERENCES: EP/1/A/5000/09 LESSON: OP-MC-EP-EP09 TASK: MD-8308 OBJECTIVES : LPRO OBJ: 2 TIME: 4 MINUTES i LPS0 OBJ: 2 l ' KA000055 C.12 3.9/4.0 REV. DATE: 06/21/93 I l l l _
. O O l* l l l ' l 2 RESIDUAL HEAT REMOVAL SYSTEM - ! PSNDR04 l l ! < . 1 Pt(s) Unit 1 is in Mode 6 and core alterations are in progress. There is > 23 feet of water above the reactor vessel flange. Train "lB" of ND is providing decay heat removal. The RO notes NC System temperature increasing with ND flowrate stable at 3000 gpm.
i Which statement best describes actions that will help to stabilize the NC System temperature? A. Throttle back total ND flowrate as necessary to stabilize ' NC System temperature B. Increase KC flow to ND heat exchanger as necessary to stabilize NC System temperature C. Start additional RN pump to increase flow through KC heat ! exchanSers and throttle back on KC to ND heat exchangers , D. Increase NC System water level to increase suction pressure to the ND pump thereby increasing ND flow rate l l ANSWER: B l [ MISCINFO: 1 REFERENCES: OP/1/A/6200/04 PT /A/60b/03 ~ ' LESSON: OP-MC-PS-ND TASK: MO 3311 OBJECTIVES : LPRO OBJ: 9 TIME: 4 MINUTES LPSO OBJ: 9 KA005000 C.11 3.1/3.8 REV. DATE: 05/12/93
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3 EP 12 SERIES EPEP12R03 , 1 P (s) Unit 1 is operating at 100% RTP when a small break LOCA While responding to a high energy line break inside occurs. of the containment the following conditions are observed: 1) All Condenser Cire Water (CCW) pumps are off 2) The five (5) highest Core Exit T/C Temperatures are all greater than 700 degrees F 3) Containment Pressure is 3.2 psig 4) RVLIS lower range level is 37% 5) All NC pumps are off Based on these indications, the operating crew should:
A. Cooldown the NCS using the S/G PORV's, disregarding the 100 degree F/hr cooldown limit, and maintain NC Pressure constant in an attempt to establish NCS Subcooling i B. Cooldown the NCS using the S/G PORV's, staying within the 100 degree F/hr cooldown limit, this will result in a subsequent decrease in NCS Pressure allowing increased ECCS flow C. Cooldown the NCS as quickly as possible using the Steam Dump Valves to the condenser, this will result in a subsequent decrease in NCS Pressure allowing increased ECCS flow D. Cooldown the NCS as quickly as possible using the S/G PORV's, this will result in a subsequent decrease in NCS Pressure allowing increased ECCS flow ANSWER: D MISCINFO: REFERENCES: EP/1/A/5000/12.1 LESSON: OP-MC EP-EP12 TASK: MO-8311 OBJECTIVES : LPRO OBJ: 3 TIME: 3 MINUTES LPSD OBJ: 3 KA000074 C.12 4.3/4.4 REV. DATE: 05/12/93
_ _ _ - _ _ ___ - _ - . _ , !. O O l .l l 4 MALFUNCTION OF NC PUMP APAP08R04 . 1 Pt(s) The Unit is at 57% Reactor Power when the following data was collected: 1) "B NC Pump Seal Leakoff Hi" flow indicates 4 gpm and is stable 2) "B NC Pump L/B Temp" is 200 degrees and increasing slowly 3) "B NC Pump Seal Tmp" is 205 degrees and increasing slowly 4) "B" NC Pump Seal D/P is 400 psid 5) "B" NC Pump Motor Frame Vibration is 3 mils increasing l slowly l 6) "B" NC Pump " Motor Shaft Vibration" is 19 mils l 7) "B" NC Pump " Pump Shaf t Vibration" is 17 mils 8) "B" NC Pump " Motor Axial Vibration" is 15 mils 9) "B" NC Pu=p " Motor Flywheel Vibration" is 21 mils Based on these conditions, which of the following actions should the operator take? A. Close NV-50 (B NC Pump Seal Return Isol), reduce unit ! l load to below P-8 and then stop "B" NC Pump B. Continue to monitor the "B" NC Pump Motor Frame Vibration and if it exceeds 5 mils trip the Reactor and then stop "B" NC Pump t 1 l C. Reduce Reactor Power to less than P 8 and then stop "B" NC Pump - D. Trip the Reactor and then stop "B" NC Pump ANS*JER: D MISCINFO: REFERENCES: AP/1/A/5500/08 , ' LESSON: OP-MC-PS-NCP TASK: h 107 OBJECTIVES : LPRO OBJ: 14 TIME: 4 MINUTES LPSO OBJ: 14 KA000015 EA1.23 3.1/3.2 REV. DATE: 05/12/93
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l S. LOSS OF CONTROL ROOM 1 APAPl7R01 , I 4; ' i ] j 1 Pt(s) The Control Room has become uninhabitable and has been evacuated. The following data was' collected from OAC ' , l General Program 76: 1) Three of the Control Bank "D" Croup 2 Rods D12, M4 and H8 are NOT fully inserted 2) All other rods are fully inserted 3) Current boron concentration is 400 PPM i 1 1 Based on these conditions, we must Emergency Borate to a ! minimum value of PPM to account for the rods not fully inserted. A. 2000 - . B. 850 , C. 700 t D. 550 i , ANSWER: B MISCINFO: REFERENCES: AP/1/A/5500/17
i LESSON: OP MC-CP-SS TASK: MO-7315 OBJECTIVES: LPRO OBJ: 2 TIME: 3 MINUTES LPSO OBJ: 2
KA000005 EA2.03 3.5/4.4 REV. DATE: 07/01/93 ! 1 . - - -
1 O O 9 i ' 6 EP 01 SERIES EFEP0lR02 . 1 Pt(s) A Natural Circulation Cooldown is in progress and it is desired to depressurize the NC System. The NC System is presently: 1. 1500 psig 2. HOTTEST Loop Th - 402* F l 3. HOTTEST Upper Head Tc - 419'F , 4 All VR Fans are running , At what approximate NC System pressure must the ' depressurization be stopped to insure the MINIMUM allowable subcooling is maintained? A. 535 psig j , B. 435 psig I C. 335 psig D. 200 psig ANSWER: A MISCINFO: ? REFERENCES: EP/1/A/5000/1.1, 2-22-91 LESSON: OP-MC-EP-EPl TASK: MO-8301.2 - OBJECTIVES : LPRO OBJ : 2 TIME: 4 MINUTES LPSO OBJ: 2 KA000017 EKl.01 4.4/4.6 REV. DATE: 07/01/93 , I l
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7 REACTOR POISONS RTRPR05 , 1 Pt(s) Given the following Unit 1 conditions: , 1. NC VR Pressure: 400 psig 2. NC T-ave: 300 degrees F . 3. Pzr Level: Water Solid
Determine the amount of Boric Acid required to increase the Reactor Coolant boron concentrat~on from 920 ppm to 1250 ppm. A. 3407 gal { B. 4088 gal ' C. 4599 gal t D. 18711 gal l t ANSWER: C MISCINFO: REFERENCES: OP/1/A/6100/22 Sect. 5.1
LESSON: OP-MC-RT-RP TASK: MO-6302 i OBJECTIVES : LPRO OBJ: 4 TIME: 4 MINUTES LPSO OBJ: 4 ,
KA001010 A3.02 4.1/3.7 REV. DATE: 07/01/93 i om
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8 EP 03 SERIES EPEP03R03 . . 1 Pt(s) While terminating safety injection following a steam line break, the RO is directed to ensure that the VCT makeup , control system is set for a boron concentration greater than the NC system shutdown boron concentration. The RO notes the following: 1) NC System Pressure 800 psig < 2) NC System Temperature 380'F 3) Boric Acid Tank Concentration 7500 ppm 4) Total Blender Flow Rate 90 gpm 5) Shutdown Boron Concentration 1100 ppm What is the minimum that the potentiometer should be set? i A. 1.7 I B. 3.3 C. 6.8 D. 13.2 , ANSWER: B MISCINFO: l
REFERENCES: OP/1/A/6100/22 Enc. 4.3 Table 5.2 - i LESSON: OP-MC-EP-EP3 TASK: MO-8303 { MO-3305 ! ! CLJECTIVES: LPRO OBJ : 3b TIME: 3 MINUTES LP50 OBJ: 3b KA004000 A4.04 3.2/3.6 REV. DATE: 05/12/93 - ..
.. ! O O i i l i l; , I 9 PRESSURIZED THERMAL SHOCK
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TAPTSR02 ) , i 1 Pt(s) You are in the process of recovering from a steam line break , inside containment where excessive cooldown has occurred. 3 ] You have completed the soak required while responding to an imminent pressurized thermal shock condition and are about to 1 initiate a cooldown to Mode 5.
i ) Select the allowable NC System pressure based on the l following conditions: The NCP's are NOT running. i i a Loop A B C D j l t I WR Th 360*F 350*F 360*F 360*F WR Tc 342*F 340*F 341*F 342*F , l A. 200 psig B. 1025 psig i C. 1200 psig , D. 1600 psig . ANSWER: B j MISCINFO:
" REFERENCES: EP/1/A/5000/14.1 . ! LESSON: OP-MC-TA-PTS TASK: MO-8313 MO-4615 OBJ ECTIVES : LPRO OBJ: 24 TIME: 4 MINUTES LPSO OBJ: 24 r KA000040 C.12 3.8/4.1 REV. DATE: 05/12/93
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10 EP 02, 06, 08 SERIES EFEP02R10 , ' 1 Pt(s) A reactor trip and SI occurred at 0100. Initial reactor { power was 100%. The control room operators were unable to align the ECCS into Cold Leg Recirculation due to the l inability to open NI-184 and NI-185 (Containment Sump Suction ) Valves). The following conditions exist at 0500: j 4 1. Total SI Flow Rate: 500 GPM 2. ECCS Systems aligned for injection mode 3. FWST level: 110 inches 4 Subcooling: - 5'F 5. All NC pumps are off 6. Attempts are being made to open NI 184 & 185 locally i Based on these conditions the control room operators should: A. Establish SI flow to maintain P:r level and RVLIS indications stable B. Reduce total SI flow to approximately 105 GPM C. Maintain 500 GPM flow and proceed to the next step D. Reduce total SI flow to approximately 215 GPM ANSWER: D , t ' MISCINFO:
REFERENCES: EP/1/A/5000/06
LESSON: OP-MC-EP-EP2 TASK: MO-8306 MO-4610 \\ OBJECTIVES: LPRO OBJ : 7 TIME: 4 MINUTES i LP50 OBJ: 7 KA000011 G.12 4.0/4.1 REV. DATE: 07/01/93 , I ! _ _
. . - . - O O ~ 11- EP 03 SERIES < EPEP03R02 i ' l 1 Pt(s) Following a steam line break outside of the containment on "lB" S/G, the RO has terminated safety injection following an excessive cooldown and establishes normal charging flow. Once VI is restored to the containment the RO observes the following: NOTE: The fault on "1B" S/G occurred 40 minutes prior to this with the unit at 50% RTP. l a. NC System UR Pressure 1750 psig b. NC Loop Cold Leg WR Temp "1A" 500*F "lB" 350*F "1C" 500*F "1D" 500*F c. NC Loop Hot Leg UR Terp "lA" 505'F "1B" 400*F "1C" 505'F "lD" 505*F d. "1B" S/G VR Level 0 e. "lB" S/G Pressure O psig Based on this information the RO should decrease NC System pressure. Which of the following would be an acceptable , pressure? ' A. 1600 psig B. 1500 psig C. 1000 psig i D. 300 psig ANS'.'ER : C MISCINFO: REFERENCES: EP/1/A/5000/3.1 8/6/90 LESSON: OP-MC-EP-EP3 TASK: MD-8303 OBJECTIVES: LPRO OBJ: 3b TIME: 3 MINUTES 1 LP50 OBJ: 3b KA000040 G.12 3.8/4.1 REV. DATE: 05/12/93 . . - . -. .- . - . . - . . - _. -
1 O O l 12 EP 01 SERIES ! EFEP0lR18 9 1 Pt(s) Given the following conditions:
1 -. Reactor Trip has occurred 2. EP/1.3 has been implemented 3. All control rod bottom lights lit 4 NCS T-hot 500*F 5. NCS T-cold 490*F ' 6. Burnup 150 EFPD 7. Present Boron Concentration 750 ppm How many gallons cf boric acid must be added to the NCS to obtain the required boron concentration to achieve an adequate Shutdown Margin?
t A. O gallons . B. 30 gallons , ! C. 330 gallons l ! D. 1830 gallons ANSWER: C ! t MISCINFO: F REFERENCES: EP/1/A/5000/1.3 DATA BOOK CURVE 1.14 LESSON: OP-MC-EP-EPl TASK: MO-8301 t OBJECTIVES: LPRO OBJ: 4 TIME: 5 MINUTES LP50 OBJ: 4 KA000007 C.12 3.8/3.9 REV. DATE: 07/01/93 , ,
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O O - i I 13 EP 13 SERIES
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EPEP13R01 1 , s 1 Pt(s) While checking to see if an immediate NC System Feed and i Bleed should be initiated while responding to a loss of ! Secondary Heat Sink the following information is noted: . 1) NC Loop T-Hot 595'F slowly increasing ! 2) NC System Pressure 2330 psig l 3) Containment Pressure 3.7 psig , 4) Avg of 5 Highest Core Exit T/C's 610* F slowly increasing 5) PORV's INC-32 and 1NC-36 Open j 6) S/G Wide Range Levels 20% slowly decreasing l Based on this information, which course of action should be ! l followed? i A. A feed and bleed IS warranted since S/G wide range , levels are < 22% and core exit T/C's are increasing B. A feed and bleed IS warranted since containment , pressure is greater than 3 psig and the pressurizer PORV's are open C. A feed and bleed IS NOT warranted since S/G wide range levels are greater than 9% even though core exit T/C's are increasing D. A feed and bleed IS NOT warranted since NC System subcooling is greater than 0*F ! ANSWER: A
MISCINFO: REFERENCES: EP/1/A/5000/13.1 LESSON: OP-MC-EP-EPl3 TASK: MO-8312 , OBJECTIVES: LPRO OBJ : 2 TIME: 4 MINUTES i LPSO OBJ: 2 KA000074 G.12 4.3/4.4 REV. DATE: 05/12/93 .-.
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I 14 H-2 SKIMMER & AIR RETURN SYS. CNTVXR05 1
1 Pt(s) Following a Unit 1 LOCA, the H2 Recombiners must be placed in f service. Determine the power setting for H2 recombiner "lA". ', Containment pressure is 9.1 psig. , A. = 324.6 K'J l 53. 5 K'J B. = 13. 7 K'4 C. = 1.5 K'J D. = ANS'JER: B - MISCINFO:
t REFERENCES: OP/1/A/6100/22 Curve 1.8 OP/1/A/6450/10 Enc. 4.2 LESSON: OP-MC-CNT-VX TASK: MO-3341 OBJECTIVES : LPRO OBJ: llA TIME: 4 MINUTES , LPSO OBJ: llA ., i KA028000 A2.01 3.4/3.6 REV. DATE: 05/12/93 , h t i
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15 REACTOR COOLANT SYSTEM PSNCR03 ,
9 1 , 1 Pt(s) 'w'hile performing a heatup on Unit 1 from Mode 4 the following 1 , temperatures were logged. 4 TIME NC PRESS NC TEMP PIR LIQ SPACE TEMP l ,
4 1015 450 psi 355 460*F i I 4 1045 742 psi 392 512*F i > 1 1115 1260 psi 420 575'F -
Based on this information. Select the correct response:
i 1 A. The Tech. Spec. heatup rate limits were exceeded on both ! ' the Reactor Coolant System and the Pressurizer B. The Tech. Spec. heatup rate limits were NOT exceeded on i .
neither the Reactor Coolant System nor the Pressurizer 4 C. The Tech. Spec. heatup rate limit was exceeded on the i Reactor Coolant System, but was NOT exceeded on the Pressurizer i D. The Tech. Spec. heatup rate limit was NOT exceeded on
i the Reactor Coolant System, but was exceeded on the ! Pressurizer ANS'.'ER : A MISCINFO: REFERENCES: Tech. Spec. 3.4.9.1, 3.4.9.2 OP/1/A/6100/01 PT/1/A/4600/08 , Data Book Curve 1.5 LESSON: OP-MC-PS NC TASK: MO-4332 MO-5304 MO-6310 OBJECTIVES: LPRO OBJ: 12 TIME: 4 MINUTES LESO OBJ: 12 KA002000 C.11 3.3/4.0 REV. DATE: 07/01/93
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.- 16' EP FORMAT, RULES FOR USE EPEPFR07 I . 1 Pt(s) With a temperature decrease in loop "B" NC Cold Leg having . exceeded 100 DECREES F in the last 60 minutes, which of the ' following combinations of loop "B" NC Cold Leg temperature and pressure will result in the most severe challenge to the Reactor Coolant Integrity CSF? A. 240 degrees F and 800 psig B. 250 degrees.F and 2100 psig C. 290 degrees F and 2200 psig D. 330 degrees F and 2400 psig ANSWER: B MISCINFO: ! ! REFERENCES: OP-MC-EP-EPF EP/1/A/5000/10 LESSON: OP-MC-EP-EPF TASK: MO-8309
OBJECTIVES: LPRO OBJ: 9e TIME: 4 MINUTES ' LP50 OBJ: 9e KA0000ll G.ll 4.3/4.5 REV. DATE: 07/01/93 l
_ _ _ _ _- m . , , O O . ' . . 17 LOSS OF ND SYSTEM APAP19R02 , i . 1 Pt(s) The following conditions exist: ) ' 1) The Unit is Mode 5 2) NC System level is 80% and stable on the Pressurizer - Cold Cal 3) The S/G manways are installed and the Reactor Vessel
Head is set 4) Both ND Pumps were stopped due to ND-1 being , inadvertantly closed 5) Natural Circulation flow exists !' 6) The ND System venting is complete 7) The Pressurizer PORV's are closed and are providing
over-pressure protection 8) ND-1 is reopened . ' 9) You are instructed to restart the ND Pumps and then to cool the NC Temperature T-Cold to 148 degrees 10) The current NC Temperature T-Cold is 178 degrees What is the Maximum Administrative cooldown rate permitted in this situation? , A. 100 Degrees Per Hour B. 50 Degrees Per Hour ' C. 20 Degrees Per Hour j ( D. 10 Degrees Per Hour ANSWER. C , 4 l l MISCINFO: j REFERENCES: AP/1/A/5500/19 . ! ' LESSON: OP-MC-PS-ND TASK: MO-7317 l \\
l OBJECTIVES: LPRO OBJ: 11 TIME: 4 MINUTES l LPSO OBJ: 11 ! l e l KA000025 EA1.01 3.6/3.7 REV. DATE: 05/12/93 l l - - ..
_. . ._ _ _ _ - -_ _ _ - _. . . nu n.m _ _. . O O m 3 dec Pn W A *. ' ' ~' , sRD EXA M \\ TABLE OF SPECIFICATIONS .:' QUESTIOh5 CROUPED BY TOPIC ff l CF QUEST]DNS ' QUEST]ON PO:NTS TIME FOR* SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES NAT ONLY7 DATE PLAN COVERED SYS MODE No. R0 SRO ID l ACF-CA-R05 1.0 3 MC Y 07/01/93 CF CA 93-1 061 000 G.11 3.4 4.1 ! i 3314 ' TOTAL QUESTIONS FOR CF TOPIC: 1 TOPIC POINTS 1.00 TOP!C TIME 3.00 CNT QUESTIONS QUESTION P0]NTS TIME FOR- SR0 REV. LES$0N SEGMENT KA CATALOG IMPORTANCE TASK , ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID ! ! ACNT-VQ R01 1.0 3 MC 05/11/93 CNT-VQ 91 1 103 000 A1.01 3.7. 4.1 ! 3340 j TOTAL QUESTIONS FOR CNT TOPIC: 1 TOPlc POINTS 1.00 TOPIC TIME 3.00 , CP QUESTIONS l QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY7 DATE PLAN COVERED SYS MODE No. RD SRO ID , ACP-AD-RD1 1.0 $ MC 06/30/93 CP-AD 89-1 000 068 EA2.09 4.1 4.3 -[ 5307 TOTAL QUEST]ONS FOR CP TOPICI 1 TOPIC PotNTS 1.00 topic TIME 5.00 g ! EL QUESTIONS ! QUESTION PolNTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK f ID MINUTES MAT ONLT? DATE PLAN COVERED SYS MODE NO. RO SRO ID . i AEL-EP-RO1 1.0 3 MC 06/30/93 EL-EP 91-4 062 000 G.11 3.1 3.7 l 1 i 3326 i l _
! TOTAL QUESTIONS FOR EL TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 , IC QUESTIONS QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID WINUTES MAT ONLT7 CATE PLAN COVERED SYS MODE NO. RD SRO ID - Alt-!CM-RD1 1.0 3 MC 06/30/93 IC-!CM 91-3 000 074 EK1.01 4.3 4.7 . l 2318 j AIC-IRE-RDS 1.0 4 MC Y 06/30/93 IC-!RE 92-4 000 005 EK3.03 3.6 4.1 l l 1309 ! f ! AIC-!Rx-R05 1.0 3 MC 06/30/93 IC-IRx 93-2 002 020 KS.09 3.6 3.9 I 3310 TOTAL QUESTIONS FOR IC TOPIC: 3 TOPIC PolNTS 3.00 TOPIC TIME 10.00 t i I $ I I I PACE (1) l i ! .. - - . . - - . -. . . - . - - . . - - . - - - . . . . . - .
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. . . 1 ' PSS QUESTIONS , QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID M]NUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. R0 SRO ID APSS KC-R01 1.0 3 MC 06/30/93 PSS-KC 95-2 008 000 K4.01 3.1 . 3.3 3313 , o' TOTAL QUESTIONS FOR PSS TOPIC: 1 TOPIC POINTS 1.00 TOPlc TIME 3.00 RT QUESTIONS !
QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK , ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. RO SRO ID ART-RB RO1 1.0 3 MC 06/30/93 RT RB 93-2 001 010 A4.04 3.5 4.1 ' 2309 { TOTAL DUESTIONS FOR RT TOPIC: 1 TOPIC POINTS 1.00 topic TIME 3.00 l SSEC2 QUESTIONS i i QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ! ID MINUTES NAT ONLY? DATE PLAN COVERED SYS MODE NO. RO SRO ID l SSE-02-P01 1.0 3 MC 06/30/93 -IC-ENB 92 1 015 000 A2.01 3.5 3.9 2302 { SSE-02-R03 1.0 3 MC 06/30/93 STM-IDE 91-2. 041 020 A3.02 3.3 3.4 [ ! 6309 f SSE-02-R04 1.0 3 MC 06/30/93 CF-CF 92-1 059 000 A2.01 3.4 3.6 5310 I . , SSE-02-R05 1.0 3 MC 06/30/93 CF-LF 93-1 059 000 A2.07 3.0 3.3 ! 6312
SSE-02 R06 1.0 3 MC 06/30/93 CF-CA 93 1 061 000 A2.04 3.4 3.8 { 3314 SSE-02-R09 1.0 3 MC 06/30/93 PS-IPE 93-2 010 000 0 .01 3.8 3.9 . 6303 SSE-02-R11 1.0 3 MC 06/30/93 PS-NV 93-1 000 009 EK3.19 3.6 3.9
3305 i SSE-02-R12 1.0 3 MC Y 06/30/93 PS-ILE 93-2 011 000 G.11 3.2 3.9 6305 SSE-02-R14 1.0 3 MC Y 06/30/93 CF-IFE 93-2 035 000 G.11 2.9 3.7 7 6300 SSE-02-R15 1.0 3 MC 06/30/93 PSS-RW 91-3 076 000 A3.02 3.7 3.7 3333 SSE-02 R17 1.0 3. MC 06/30/93 IC-IPE 93-1 012 000 K4.02 3.9 4.3 2301 TOTAL QUESTIONS FOR SSE02 TOPIC: 11 TOPIC Po!NTS 11.00 TOPIC TIME 33.00 TOTAL QUESTIONS 20 TOTAL POINTS 20.00 TOTAL TIME- 63.00 PAGE(2) . . - - - . . . .- . .
~ O O - . PRINT NAME PROG. ID: OP-MC-LRQ SIGNATURE 1993(NRC)(SRO)P ( ) ' ^ SSN Prepared by: . SHIFT DATE REVIEW / INITIAL LICENSED REQUAL ANNUAL EXAMINATION TEST # 1993(NRC)(SRO)PART(A) Total Points: 20.00 Total Questions: 20
- INSTRUCTIONS ***
1. Use NO. 2 PENCIL only. 2. Fill in the appropriate infor=ation on the TEST COVER SHEET. 3. Ask the test monitor instructor about any questions which are not clear to you. 4 Circle the correct answer on each question of this exam. 5. When finished with exam, read the "non-compromise statement" and sign the TEST COVER SHEET. i NON-COMPROMISE STATEMENT: "My Signature on this form is my declaration that the responses given on the attached test or exam are entirely my own. It further declares that I am aware that I am subject to termination from the training program immediately and in addition, will be subject to further disciplinary action up to and including discharge from the l ! company for cheating and/or compromising on exams / tests / quizzes." i I
, -. l- O O , 1 1 J . ,! 4 3 ' i s, j ! SCENARIO ~ .
! . i . , ! SPECIFIC ,
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i ! 2 1 ! QUESTIONS . 1 t '
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- o o ~ ! 1 SCENARIO QUESTIONS SSE02R01
! 1 Pt(s) Based on present plant conditions, which one of the following j describes how the Protection System would respond? A. If power were lost to EKVA, a Reactor Trip would occur l B. If Intermediate Range 1NI-36 were to fail high, a Reactor ! Trip would NOT occur C. If only the "A" Train " Source Range Select Switch" were turned to the " Reset" position, a Reactor Trip would NOT occur l D. If the control power fuses to Source Range Instrument N31 ! were to blow, a Reactor Trip would occur ANS'a'ER : A MISCINFO: REFERENCES: OP-MC-IC-ENB LESSON: OP-MC-IC-ENB TASK: MO-2302 OBJ ECTIVES : LPRO OBJ: 11 TIME: 3 MIh'UTES I LPSO OBJ: 11 KA015000 A2.01 3.5/3.9 REV. DATE: 06/30/93 ! ! I f i .
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' 2 SCENARIO QUESTIONS . SSE02R03 . 9 $ 1 Pt(s) Based on present plant conditions, which one of the following would occur if the Steam Dump Select switch were placed in the "T-avg Mode" of operation? t. . The condenser steam dumps would modulate to maintain T-avg - T-ref B. The condenser steam dumps would modulate to maintain T-avg five degrees higher than T-ref C. The condenser steam dumps would close, T-avg would increase until the SM PORV's open and maintain T; avg D. The condenser steam dumps would continue to modulate to maintain SM pressure at approximately 1092 psig ANSWER: C MISCINFO: REFERENCES: OP-MC-STM-IDE LESSON: OP-MC-STM-IDE TASK: MO-6'09 OBJECTIVES : LPRO OBJ : 1 TIME: 3 MINUTIS LPS0 OBJ: 1 i KA041020 A3.02 3.3/3.4 REV. DATE: 06/30/93 1
_ __ , . l ~ O O > 1 i ' 3 SCENARIO QUESTIONS . SSE02R04 j ' , 1 Pt(s) Based on present plant conditions, which one of the following would occur if the AMSAC for CF Valves " Unblock" pushbutton 1 were depressed? (Consider short term effects only, within 1 minute) A. The Main Turbine would trip; the Reactor would NOT trip B. The Main Turbine would trip; the Reactor would trip f C. The Main Turbine would NOT trip; the Reactor would f NOT trip D. The Main Turbine would NOT trip; the Reactor would trip ! ! ANSk'ER: A i MISCINFO: ' REFERENCES: OP-MC-CF-CF ! i l LESSON: OP-MC-CF-CF TASK: MO-5310 I 1 i OBJECTIVES: LPRO OBJ: 15 TIME: 3 MINUTES , LPSO OBJ: 15 i KA059000 A2.01 3.4/3.6 REV. DATE: 06/30/93 l i ! 1 l 1 1 < l . - . - -
___ O O 4 SCENARIO QUESTIONS . , SSE02R05 ! ' , 1 Pt(s) Based on present plant conditions, which one of the following , would occur if a Feed Pump lube oil leak necessitated = immediately stopping both CF Pump Turbine MOP Pumps for "lB" CF Pump? , A. The Main Turbine would trip; both Motor Driven CA Pumps would start; the Reactor would trip; the "1B" CF Pump suction valve ICM-272 would close - B. The Main Turbine would trip; both Motor Driven CA Pumps I would start; the Reactor would NOT trip; the "1B" CF Pump suction valve ICM-272 would close C. The Main Turbine would NOT trip; both Motor Driven CA Pumps would start; the Reactor would NOT trip; the "lB" CF Pump suction valve ICM-272 would close 1 D. The Main Turbine would trip; the Motor Driven CA Pumps would NOT start; the Reactor would trip; the "1B" CF Pump ' suction valve ICM-272 would NOT close ANSWER: B MISCINFO: P REFERENCES: OP-MC CF-LF LESSON: OP-MC-CF-LF TASK: MO-6312
OBJECTIVES : LPRO OBJ: 20 TIME: 3 MINUTES LPSO OBJ: 20 & , KA059000 A2.07 3.0/3.3 REV. DATE: 06/30/93 f - , . 1 ) i . . . - -
- - ~ O O !- s i 1 5 SCENARIO QUESTIONS
SSE02R06 ' - , '~hich one of the following describes how the CA System will 1 P (s) - respond to present plant conditions?
j A. CA flow to S/G "C" can NOT be throttled using CA-48,
unless both Train "A" & "B" "CA Modulating Valve Reset" , 4 pushbuttons for the Turbine and Motor Driven Pu=ps are ' depressed , B. CA flow to S/G "D" can NOT be throttled using CA-36, unless both Train "A" & "B" "CA Modulating Valve Reset:
) Turbine" pushbuttons are depressed J t C. CA flow to S/G "A" can NOT be throttled using CA-64, unless the Train "A" "CA Modulating Valve Reset;-Turbine" pushbutton is depressed D. CA flow to S/G "B" can be throttled using CA-52, without any of the "CA Modulating Valve Reset" pushbuttons being depressed ANSER: D i MISCINFO: ! !
REFERENCES: OP-MC-CF-CA ^ , k LESSON: OP-MC-CF-CA TASK: MO-3314 i . OBJECTIVES: LPRO OBJ: 13 TIME: 3 MINUTES ' LPSO OBJ: 13
! KA061000 A2.04 3.4/3.8 REV. DATE: 06/30/93 . I
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O O , ' 6 SCENARIO QUESTIONS
SSE02R09 , 1 Pt(s) Based on present plant conditions, which one of the following would occur, if the output of the " Pressurizer Pressure Master" were to fail to the "100:" position? (Assume no operator action is taken.) A. All pressurizer heaters will go off; PORV NC-34 vill open; pressurizer sprays will open; the reactor will trip due to a low pressurizer pressure reactor trip signal of 1945 psig B. All pressurizer heaters will go off; PCRV NC-34 will open and then will reclose; pressurizer sprays will open; the reactor will not trip until pressurizer pressure decreases to less than 1845 psig due to the Safety Injection signal C. All pressurizer heaters will go on; all PORV's will be blocked from opening; pressurizer sprays will close; the reactor will trip due to a high pressurizer pressure reactor trip signal of 2385 psig , D. All pressurizer heaters will go on; PORV NC-32 and NC-36 will cycle open and closed; pressurizer sprays will close; the reactor will not trip due to any pressurizer pressure reactor trip s! nal ANSWER: B MISCINFO: REFERENCES: OP-MC-PS-IPE , , LESSON: OP-MC-PS-IPE TASK: MO-6303 ' OBJECTIVES: LPRO OBJ: 3 TIME: 3 MINUTES LPSO OBJ: 3 KA010000 K3.01 3.8/3.9 REV. DATE: 06/30/93
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' , 7 SCENARIO QUESTIONS . SSE02R11 , 1 Pt(s) Based on present plant conditions, which one of the following would occur if "A" Train Safety Injectier.were manually j initiated? (Assume no operator action is taien.) A. CA flow to the "A" S/G would increase to greater than 400 gpm B. The "A" D/G would start, but the ETA Emergency Breaker would NOT close C. The "B" NV Pump would remain aligned to the VCT D. The orifice isolations would close, isolating Letdown l l AN SER: b
MISCINFO: REFERENCES: OP-MC-PS-NV ! LESSON: OP-MC-PS-NV TASK: MO-3305 l , > OBJECTIVES : LPRO OBJ: 3 TIME: 3 MINUTES ' LPS0 OBJ: 3 . ? KA000009 EK3.19 3.6/3.9 REV. DATE: 06/30/93
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O O 8 SCENARIO QUESTIONS i
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SSE02R12 ! , 1 Pt(s) Based on present Pressuri::er Water Level instrumentation conditions, which one of the following is correct concerning an evaluation of this instrumentation? ' A. Power may not exceed 5% RTP until the Channel 3 bistables are placed in a tripped condition within 6 hours B. Power may not exceed 5% RTP, even if the Channel 3 bistables are placed in a tripped condition C. Power is permitted to exceed 5% RTP, but the Channel 3 bistables must be placed in a tripped condition within 6 hours . D. Power is permitted to exceed 5% RTP, but power may not exceed P-7, regardless of Channel 3's bistable condition ANSWER: C MISCINFO: REFERENCES: OP-MC-PS-ILE l LESSON: OP-MC-PS-ILE TASK: MO-6305 OBJECTIVES: LPRO OBJ : 1 TIME: 3 MINUTES LPSO OBJ: 1 KA011000 G.11 3.2/3.9 REV. DATE: 06/30/93 l l 4 i ! 1 l l l 1
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9 SCENARIO QUESTIONS SSE02R14 . 1 Pt(s) Based on present plant conditions, which one of the following correctly evaluates the S/G NR Level Instrumentation status? A. All S/G Narrow Range Level Channels are operable B. S/G Narrow Range Level Channel 2 of "B" S/G must be declared inoperable and its associated low low and high high bistables must be tripped within,B' hours l SsL C. S/G Narrow Range Level Channel 2 of "C" S/G must be declared f.noperable and its associated low low bistables must be tripped within 1 hour and its associated high high bistables must be tripped within f' hours j $+ D. Both S/G Narrow Range Level Channel 2 of "B" S/G and S/C Narrow Range Level Channel 2 of "C" S/G must be declared inoperable and their associated low low bistables must be tripped within 1 hour and their associated high high bistables must be tripped within ,( hours I 9 AN S'w'ER : C MISCINFO: REFERENCES: OP-MC-CF-IFE LESSON: OP-MC-CF-IFE TASK: MO-6308 , OBJECTIVES : LPRO OBJ : 13 TIME: 3 MINUTES LPS0 OBJ: 13 KA035000 G.11 2.9/3.7 REV DATE: 06/30/93 i . - -
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10 SCENARIO QUESTIONS SSE02R15 , 1 Pt(s) Based on present plant conditions, which one of the following correctly describes the RN System response if PCB's 8, 9, 11 and 12 were to open? (Assume no operator action is taken.) A. The "A" Train KC Heat Exchanger would be provided with RN flow B. The "B" Train of RN would remain aligned to the Lower Level Intake i C. The RN AB Non-essential header would not lose RN flow l l D. The NC Pump Motor Air Coolers would lose RN Flow l l l AN SER : D i MISCINFO: REFERENCES: OP-MC PSS-RN, LESSON: OP-MC-PSS-RN TASK: MO-3333 OBJECTIVES : LPRO OBJ: 9 TIME: 3 MINUTES ' LPSO OBJ: 9 KA076000 A3.02 3.7/3.7 REV. DATE: 06/30/93
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' O O 11 SCENARIO QUESTIONS . SSE02R17 . 1 Pt(s) Based on present plant conditions, which one of the following correctly describes plant response? A. If stator cooling water differential pressure decreases to under 15 psid for over 45 seconds, the main turbine will trip I B. If the "A" CM Booster Pump were stopped, a reactor trip would NOT occur within 60 seconds C. If turbine impulse pressure Channel 1 were to fail high, actual pressurizer level will increase D. If the 1TB and 1TC mode select switches are placed in manual and then the normal breaker " Trip" pushbuttons were depressed, a reactor trip would occur within 60 seconds i AN S'*'ER : B , MISCINFO: REFERENCES: OP-MC-IC-IPE , LESSON: OP-MC-IC-IPE TASK: M0-2301 ! OBJECTIVES : LPRO OBJ : 1 TIME: 3 MINUTES I LPSO OBJ: 1 , KA012000 K4.02 3.9/4.3 REV. DATE: 06/30/93
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! l QUESTIO'NS UNRELATED TO THIS SCENARIO I
O O ~ , 4 12 PART "A" AUX. FEEDWATER ACFCAR05 , l 1 Pt(s) Given the following plant conditions: 1) The Unit is operating at 50% RTP 2) The Turbine Driven CA pump was taken 00S 12 hours ago because of a ruptured casing and discharge piping 3) A fire occurred 30 minutes ago which damaged both Motor ' Driven CA pump motors to the point where neither motor will operate 4) No other plant damage occurred in the fire 5) Corrective action has been initiated to restore a CA Pump to operable status as soon as possible, estimated time of restoration is 48 hours What other action (s) concerning power operations are required to be taken? A. None; maintain the plant at power B. Apply Action Statement (b) of T.S. 3.7.1.2; be in at least HOT STANCBY within 6 hours of the time of the fire ' C. Apply T.S. 3.0.3 and be in at least HOT STANDBY within 7 hours of the time of the fire D. Apply T.S. 3.0.3 and be in at least HOT STANDBY within 7 hours from now since it has just been determined that no CA pumps will be operable for 48 hours 1 ANSUER: A ' MISCINFO: REFERENCES: Tech. Spec. 3.0.3 Tech. Spec. 3.7.1.2 , h LESSON: OP-MC-CF-CA TASK: MO-3314 6 r OBJECTIVES: LPRO OBJ: N/A TIME: 3 MINUTES F LPSO OBJ: 9 , KA061000 G.11 3.4/4.1 REV. DATE: 07/01/93 .
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O O , l ! l l 13 PART "A" AIR ADD & RELEASE SYS
ACNT"QR01
l , I l 1 Pt(s) Given: . ' ' l. A VQ Release has been made. 2 The VQ Release was started at 0705. 3. The VQ Release was stopped at 0730. , 4 Initial Containment Pressure was 0.20 psig, l 5. Final Containment Pressure was 0.12 psig. 6. The VQ Totalizer is inoperable. hat was the TOTAL VOLUME released? . l A. 185 Cubic Feet , B. 6505 Cubic Feet ' C. 6690 Cubic Feet f D. 6875 Cubic Feet ANS'JER: D l l MISCINFO: l REFERENCES: OP/1/A/6450/17 i LESSON: OP-MC-CNT-VQ TASK: MO-3340 OBJECTIVES: LPRO OBJ: 5 TIME: 4 MINUTES LPSO OEJ: 5 l
l KA103000 A1.01 3.7/4.1 REV. DATE: 05/11/93 i l . i I l , ! ! ! l l s --- , . _ _ - . . -
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' 14 PART "A" STBY SHUTDC7N FAC. ACPADR01 . 1 Pt(s) Given the following information: 1) You have completed transfer of control to the Standby Shutdown Facility during a Fire Event. 2) You are in the process of verifying Natural Circulation. 3) These are the current Incore T/C readings: T/C #1 - 562*F T/C #2 - 559'F T/C #3 - 564*F T/C (14 - 560*F T/C #5 - 562*F 4) The Reference Junction Temp. Deviation - - 3*F Determine the " Corrected Incore Temperature" at the SSF in order to determine the status of subcooling. A. 556*F B. 561*F C. 564*F D. 567'F ANSWER: B MISCINFO: REFERENCES: AP/1/A/5500/24 LESSON: OP-MC-CP-AD TASK: MO-5307 OBJECTIVES: LPRO OBJ: 3 TIME: 5 MINUTES LPS0 OBJ: 3 KA000068 EA2.09 4.1/4.3 REV. DATE: 06/30/93 l l l l 1
^ - O O 15 PART "A" MAIN POWER DIST. - AELEPR01 . 1 P (s) Civen the following information: 1) The Unit is at 3% power. 2) The standby breaker for ITD (6900V Bus) is closed. 3) The normal breaker for ITD (6900V Bus) is open and cannot ' be closed due to a breaker failure. ' 4) All other 6900V and 4160V breakers are in a normal alignment. 5) Both D/G's are operable. Which one of the following correctly explains the status of the AC electrical power system? A. The AC power system is inoperable since both lETA and lETB are both supplied power from one offsite trans. mission network (Busline 1A) B. The AC power system is operable since both lETA and 1ETB are still energized C. The AC power system is operable as long as lETA and IETB are energized and 13 diesel generator remains operable D. The AC power system is inoperable since 1TA is supplying power to lETA and ITD is supplying power to lETB ANSWER: A MISCINFO: REFERENCES: MNS Tech. Spec. 3.8.1.1 LESSON: OP MC-EL-EP TASK: MO-3326 OBJECTIVES: LPRO OBJ : 38 TIME: 3 MINUTES LPSO OBJ: 38 KA062000 C.11 3.1/3.7 REV. DATE: 06/30/93
- ' O O ~ 16 PART "A" ICCM AICICMR01 , 1 Pt(..) If the NCPs are tripped following a LOCA and the break has been isolated, which one of the following situations provides the highest subcooling? PZR PRESSURE HOT LEG TEMP COLD LEG TEMP A. 600 500 470 B. 800 535 510 C, 1000 525 520 D. 1200 570 560 ANS'n'ER : C MISCINFO: REFERENCES: Data Book LESSON: OP-MC-IC-ICM TASK: MO 2318 OBJECTIVES: LPRO OBJ : 11 TIME: 3 MINUTES LPSO OBJ: 11 KA000074 EKl.01 4.3/4.7 REV. DATE: 06/30/93
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i 17 PART "A" ROD CONTROL AICIRER05 . i 1 P (s) The plant is at 80% and increasing in power when a rod in , l Control Bank "D" is discovered to be 18 steps below the other ] Bank "D" rods. ISE investigates and reports that ALL power to that rod is deenergized. < Which one of the following actions is required as a result of - this condition? . A. Operation may continue with no action required since the j Accident Analysis assumes one stuck rod 1 B. Operation may continue, provided AFD is maintained within limits, thermal power is reduced to less than 75% within the next hour and the neutron flux trip setpoint is reduced to 85% within the following four hours ' C. The rod is untrippable since it should have dropped with power removed. Tech. Specs. require that SDM is verified within one hour and the plant be in hot standby in } 6 hours 4 4 D. Operation may continue provided other rods in Control Bank D are positioned within 12 steps of the misaligned rod within 1 hour, while maintaining proper rod sequence and insertion limits. The rod must be restored to operable status within 72 hours . ! ANSWER: C I MISCINFO: ! REFERENCES: Tech. Specs. 3.1.3.1 LESSON: OP-MC-IC-IRE TASK: MO-3310 OBJECTIVES : LPRO OBJ: 14 TIME: 4 MINUTES LPSO OBJ: 14
KA000005 EK3.03 3.6/4.1 REV. DATE: 06/30/93 i 1 ]
- O O - 18 REACTOR CONTROL , AICIRXROS , 1 Pt(s) Given the following conditions: + 1) Reactor power is 60% 2) Loop "C" Delta-T indicates LOW , 3) Loop "C" Tave indicates HIGH . Which one of the following RTD failures caused these indications? A. T-cold failed high B. T-cold failed low i C. T-hot failed high D. T-hot failed low i ANSWER: A i MISCINFO:
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REFERENCES: OP-MC IC-IRX LESSON: OP-MC-IC-IRX TASK: MO-3310 i , OBJECTIVES: LPRO OBJ: 3 TIME: 3 MINUTES LPSO OBJ: 3 KA002020 K5.09 3.6/3.9 REV. DATE: 06/30/93 - . . . . - . -. .. --. . - - . . .~. - . - _ . - -
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19 PART "A" COMPONENT COOLING APSSKCR01 . 1 Pt(s) Given the following conditions: 1) Reactor power is 100% 2) KC pumps B1 and B2 are supplying system loads 3) Safety injection signal is received Which one of the following describes the response of the KC system? A. KC pumps B1 and 32 will continue to run, then KC pumps Al and A2 will be sequenced on B. KC pumps B1 and B2 will be load shed, then KC pumps Al, A2, B1 and B2 will be sequenced on C. KC pumps B1 and B2 will be load shed, then KC pumps Al and A2 will be sequenced on, KC pumps B1 and B2 will not be sequenced on D. KC pumps B1 and B2 will continue to run, KC pumps Al and A2 will not be sequenced on ANSWER: A MISCINFO: REFERENCES: OP-MC-PSS-KC LESSON: OP-MC-PSS-KC TASK: MO-3331 OBJECTIVES: LPRO OBJ: 5 TIME: 3 MINUTES LPSO OBJ: 5 KA008000 K4.01 3.1/3.3 REV. DATE: 06/30/93 t
, O O - - . 20 PART "A" REACTIVITY BALANCES '
ARTRBR01 _ 0 1 Pt(s) Unit 1 is tripped due to a Loss of All NC Pumps. The following conditions exist: 1) A Cooldown to 250 degrees F must be conducted 2) Current T-ave is 557 degrees F 3) Current Core Burnup is 100 EFPD 4) The NC Pu=ps will NOT be available for the Cooldown 5) Current boron concentration is 1125 ppm Determine the minimum boren concentration required to cooldown to 250 degrees F. A. 1125 ppm B. 1235 ppm r C. 1301 ppm D. 1411 ppm ANS'.'ER : D ! MISCINFO: ! t REFERENCES: Data Book ! EP/1/A/5000/1.1 r ! LESSON: OP-MC-RT-RB TASK: MO-2309 i OBJECTIVES: LPRO OBJ: 10 TIME: 3 MINUTES LPSO OBJ: 10 KA001010 A4.04 3.5/4.1 i REV. DATE: 06/30/93
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TABLE OF SPECIFICATIONS QUESTIONS CROUPED BY TOPIC
- 6
/0W/ 5 Lp & 7 u 93
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AP QUESTIONS QUESTION POINTS TIME FOR- SRO REV. LESSON SEDMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. RO SRO ID , AP-AP17-R01 1.0 3 MC 07/01/93 CP-SS 92-1 000 005 EA2.03 3.5 4.4 , i 7315 i ] AF-AP19-R02 1.0 4 MC 05/12/93 PS-ND 93-1 000 025 EA1.01 3.6 3.7 7317 j TOTAL QUESTIONS FOR AP TOPIC: 2 TOPIC POINTS 2.00 TOPIC TIME 7.00 ! CTH QUESTIONS i DUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID l CTH-CP-R01 1.0 3 MC Y 07/01/93 CTH-CP 91-4 001 000 K5.10 3.9 4.1 ! $303 i TOTAL QUESTIONS FOR CTH TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 DG QUESTIONS l QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MOCE No. R0 SRO . ID r s DG-DG-R06 1.0 4 MC Y 05/12/93 DG-DG 91-4 064 000 c.11 3.4 3.9 I 3328
4314
TOTAL QUESTIOkS FOR DG TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 i ECC QUESTIONS ! l l QUESTION PolNTS TIME FOR- SRO REV. LESSON SEGMENT -KA CATALOG IMPORTANCE TASK [ ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. RD SRO ID ! { l ECC-NS-R02 1.0 4 MC Y 07/01/93 ECC-NS 91-2 026 000 G.11 3.2 4.1
i 3327 , TOTAL QUESTIONS FOR ECC TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 ! EP QUESTIONS ~ QUESTION PolWTS TIME FOR- SRO REV. LESSON SECMENT KA CATALOG IMPORTANCE TASK- ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. R0 SRO ID l EP EMP-R13 1.0 4 MC Y 07/01/03 EP EMP 93 2 000 055 G.2 2.9 4.1 l 9300 TOTAL DUESTIONS FOR EP TOPIC: 1 TOPlc PolNTS 1.00 TOPIC TIME 4.00 l ' f EP02 OUESTIONS f f QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG !MPORTANCE TASK ] ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. RO SRO ID ' EP-EP02-R10 1.0 4- MC 07/01/93 EP-EP2 92-5 000 011 G.12 4.0 4.1 ) 8306- 1 PAGE(1) ' i i I
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! ! 1 ) I l EPC2 QUESTIONS
QUESTION POINTS TIME FOR* SRO REY. LESSON SEGMENT KA CATALOG IMPORTANCE TASK \\ t ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID ( ! ' 4610
, TOTAL QUESTIONS FOR EPC2 TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00
! EP03 OUESTIONS QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE Wo. R0 SRO ID i a. EP-EP03 R02 1.0 3 MC 05/12/93 EP-EP3 93-2 000 040 G.12 3.8 4.1 .I ! ] 8303 l EP-EP03 RC3 1.0 3 MC 05/12/93 EP-EP3 93-2 004 000 A4.04 3.2 3.6 8303 ! 3305 l ' j TOTAL QUESTIONS FOR EP03 TOPIC: 2 TOPIC POINTS 2.00 TOPIC TIME 6.00 l i EPO4 QUESTIONS QUESTION POINTS TIME FOR. SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK i ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. RO SRO ID -t i EP-EPO4-R07 1.0 6 MC 07/01/93 EP-EP4 91-4 000 038 G.12 3.8 4.0 [ ] j 8305 ! 4608 [
TOTAL QUESTIONS FOR EPO4 TOPIC: 1 TOPIC PolNTS 1.00 TOPIC TIME 6.00 l I > EP09 QUEST!DNS ! ! QUESTION PolNTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ! ID MINUTES MAT ONLY? .DATE PLAN COVERED SYS MODE NO. RO SRO ID EP-EP09-R01 1.0 4 MC 06/21/93 EP-EP09 92-4 000 055 c.12 3.9 4.0 i i 8308 l t TOTAL QUESTIONS FOR EP09 TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 1 EP12 QUESTIONS i QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK l ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. R0 SRO ID ) EP-EP12 R03 1.0 3 MC 05/12/93 EP-EP12 92-2 000 074 G.12 4.3 4.4 8311 TOTAL QUESTIONS FOR EP12 TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 3.00 EP13 QUESTIONS QUESTION Po!NTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE No. R0 SRO ID EP EP13-R01 1.0 4 MC 05/12/93 EP-LP13 93-1 000 074 G.12 4.3 4.4 1 8312
TOTAL QUESTIONS FOR EP13 TOPIC: 1' TOPIC POINTS 1.00 TOPlc TIME 4.00 , PAGE(2) , - . - . . . . _ - . . , . .
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i FM QUESTIONS
QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ' ID MINUTES MAT ONLT? CATE PLAN COVERED SYS M3DE NO. RO SRO ID , FN-FC-406 1.0 4 MC Y 07/01/93 FM-FC 90-5 015 000 G.11 3.1 3.8 I 9313 4326 i 4327
i TOTAL ouESTIONS FOR FM TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 l l IC ouESTIONS i ! QUESTION POINTS TIME FOR- SRO REV. LESSON SECMENT KA CATALOG IMPORTANCE TASK ID MINUTES NAT ONLY7 DATE PLAN COVERED SYS MODE NO. RO SRO ID IC-ENA-RO1 1.0 8 MC Y 07/01/93 IC ENA 88 5 015 000 A1.04 3.5 3.7
i 2316 i TOTAL QUESTIONS FOR IC TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 8.00 ! PS QUEST!DNS I QUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ID MINUTES MAT ONLY7 DATE PLAN COVERED SYS MODE Wo. RO SRO ID PS-ND RO1 1.0 4 MC Y 05/12/93 PS-ND 92 2 005 000 G.5 3.2 3.8 i
3311 i i TOTAL QUESTIONS FOR PS TOPIC: 1 TOPIC POINTS 1.00 TOPIC TIME 4.00 l ' l TA QUESTIONS i < r i OUESTION POINTS TIME FOR- SRO REV. LESSON SEGMENT KA CATALOG IMPORTANCE TASK ! ID MINUTES MAT ONLY? DATE PLAN COVERED SYS MODE NO. R0 SRO ID q TA-PTS-R02 1.0 3 MC 05/12/93 TA-PTS. 92-4 000 040 G.12 3.8 4.1 8313 4615
TOTAL QUESTIONS FOR TA TOPIC: 1 TOPIC PolNTS 1.00 TOPIC TIME 3.00 i ! TOTAL QUESTIONS 17 TOTAL POINTS 17.00 TOTAL TIME 68.00
J l i l i i I i !
I , ) - PACE (3) 1
O- O> - ~ 9 . PRINT NAME PROG. ID: OP-MC-LRQ SIGNATURE 1993(NRC)(SRO)P T(B) SSN Prepared by: SHIFT l DATE REVIEW / INITIAL
LICENSED REQUAL ANNUAL EXAMINATION , TEST # 1993(NRC)(SRO)PART(B) Total Points: 17.00 Total Questions: 17 !
- INSTRUCTIONS ***
1. Use No. 2 PENCIL only. l 2. Fill in the appropriate information on the TEST COVER SHEET, 3. Ask the test monitor instructor about any questions which are i not clear to you. 4 Circle the correct answer on each question of this exam. 5. When finished with exam, read the "non-compromise statement" and sign the TEST COVER SHEET. NON-COMPROMISE STATEMENT: "My Signature on this form is my declaration that the responses given on the attached test or exam are entirely my own. It further declares that I am aware that I am subject to termination from the training program immediately and in addition, will be subject to further disciplinary action up to and including discharge from the company for cheating and/or compromising on exams / tests / quizzes." r i k i
a O O , . I l ! 1 EP 04, 05 SERIES . EPEP04R07 , 1 Pt(s) DURING a SGTR the operator has isolated S/G "A". The following conditions / indications exist: 1) NCS subcooling 23*F , 2) NCS pressure 1080 psig stable ! 3) Ruptured S/G "A" pressure 480 psig decreasing j 4) Ruptured S/G "A" NR level 67% 5) FWST level 290 inches , 6) Containment Sump Level 0 ft [
Select the procedure the operator will be using to perform , the NCS cooldown . ! A. EP/04 SGTR B. EP/5.1 SGTR with Continuous NC System Leakage - Subcooled Recovery - C. EP/5.2 SGTR with Continuous NC System Leakage - , Saturated Recovery - D. EP/06 Loss of Emergency Coolant Recirculation i ANSWER: B MISCINFO: REFERENCES: EP/1/A/5000/04 EP/1/A/5000/5.1 , LESSON: OP-MC-EP-EP4 TASK: MO-8305 MO-4608 OBJECTIVES : LPRO OBJ : 7 TIME: 6 MINUTES > LPSO OBJ: 7 i ! KA000038 G.12 3.8/4.0 REV. DATE: 07/01/93 ! I
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i ' O O i l l 2 INCORE INSTRUMENTATION - ICENAR01 ' . 1 Pt(s) Unit 1 is at 100% power. Power Range Channel 42 has been l declared inoperable due to detector failure. The Reactor ' , G oup has attempted to run the Incore Instrument System to l determine QPTR. They report that 30 of the 58 thimbles are ' INOPERABLE. Determine which of the following is the minimum action , required to be taken. A. Thermal Power is restricted to < 75% RTP and the P.R. Neutron Flux High Trip setpoint is reduced to 5 85% RTP
within 4 hours B. Monitor QPIR at least once per 12 hours when above 75% RTP using Incore Detectors ' C. With the number of operable channels one-less than the minimum channels operable requirement, restore the inoperable channel to operable status within 48 hours or be in Hot Standby within the next 6 hours. D. Reduce thermal power to < 50% RTP within the next 2 hours , and reduce P.R. neutron flux high trip setpoint to 5 55% ' RTP within the next 4 hours ANSWER: A ! ! MISCINFO: , REFERENCES: Tech. Spec. 3.3.3.2 Tech. Spec. 3.3.1
LESSON: OP-MC-IC-ENA TASK: MO-2316 OBJECTIVES: LPRO OBJ: 7 TIME: 8 MINUTES LPS0 OBJ: 7 KA015000 A1.04 3.5/3.7 REV. DATE: 07/01/93
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. . - . . l O O l 3 CONTAINMENT SPRAY SYSTEM - . l l ECCNSR02 l l - . ! l 1 Pt(s) During testing of the Containment Pressure Control System
(CPCS). NS pump 1A pressure transmitter channel (INSPT-5520) l setpoint is .24 psig. Unit 1 is at 100% RTP. How does this . affect Containment Spray System operability and what action I should be taken if any? A. The NS System (Train A) is operable since the value is I within the allowable range of pressure B. The NS System (Train A) is inoperable. Apply the action statements of Tech. Spec. 3.6.5.6 (Cont. Air Return /H2 ,
Skimmer) and 3.6.2 (Cont. Spray) C. The NS System (Train A) is inoperable, so we must place the inoperable channel in the start permissive mode within one hour and apply the action statement of Tech. ! Spec. 3.6.2 (Cont. Spray) ! D. The NS System (Train A) is operable because more than one l pressure failure is required to make it inoperable
ANSER : C MISCINFO: REFERENCES: Tech. Spec. 3.3.2 Tech. Spec. 3.6.2 LESSON: OP-MC-ECC-NS TASK: MO-3327.5
OBJECTlVES: LPRO OBJ: 7 TIME: 4 MINUTES l LPS0 OBJ: 7 . KA026000 C.ll 3.2/4.1 REV. DATE: 07/01/93 i , f f . - - - - - - - - - - - - - . - - - - - - - - - - - - e ~ ,
. l O O ' 4 PRESSURIZED THERMAL SHOCK . TAPTSR02 ' 1 Pt(s) You are in the process of recovering from a steam line break inside containment where excessive cooldown has occurred. i You have completed the soak required while responding to an J imminent pressurized thermal shock condition and are about to initiate a cooldown to Mode 5. t Select the allowable NC System pressure based on the following conditions: ! ! The NCP's are NOT running.
Loop A B C D j WR Th 360*F 350*F 360*F 360*F l l i l WR Tc 342*F 340*F 341*F 342*F i A. 200 psig l l B. 1025 psig
l l C. 1200 psig l
l D. 1600 psig i ANSWER: B MISCINFO: REFERENCES : EP/1/A/5000/14.1 . LESSON: OP-MC-TA-PTS TASK: MO-8313 MO-4015 OBJECTIVES: LPRO OBJ: 24 TIME: 4 MINUTES LPS0 OBJ: 24 l KA000040 G.12 3.8/4.1 REV. DATE: 05/12/93 ,
. O O 5 LOSS OF ND SYSTEM . APAP19R02 , 1 Pt(s) The following conditions exist: l 1) The Unit is Mode 5 2) NC System level is 80% and stable on the Pressurizer Cold Cal 3) The S/C manways are installed and the Reactor Vessel
Head is set 4) Both ND Pumps were stopped due to ND-1 being inadvertantly closed 5) Natural Circulation flow exists 6) The ND System venting is complete 7) The Pressurizer PORV's are closed and are providing , over-pressure protection
8) ND-1 is reopened 9) You are instructed to restart the ND Pumps and then to cool the NC Temperature T-Cold to 148 degrees 10) The current NC Temperature T-Cold is 178 degrees , ! What is the Maximum Administrative cooldown rate permitted in this situation? A. 100 Degrees Per Hour , B. 50 Degrees Per Hour C. 20 Degrees Per Hour D. 10 Degrees Per Hour . ANSWER: C MISCINFO: REFERENCES: AP/1/A/5500/19 LESSON: OP-MC-PS-ND TASK: MO-7317 OBJECTIVES: LPRO OBJ: 11 TIME: 4 MINUTES LPSO OBJ: 11 KA000025 EA1.01 3.6/3.7 REV. DATE: 05/12/93 [ t 1 i 4 l i l
_ _ . - _ -- ... I i 6 EP 12 SERIES . EPEP12R03 l r 1 Pt(s) Unit 1 is operating at 100% RTP when a small break LOCA occurs. While responding to a high energy line break inside , of the containment the following conditiona are observed: { ' 1) All Condenser Cire Water (CCW) pumps are off 2) The five (5) highest Core Exit T/C Temperatures are all greater than 700 degrees F , 3) Containment Pressure is 3.2 psig t 4) RVLIS lower range level is 37% 5) All NC pumps are off
i Based on these indications, the operating crew should: ' A. Cooldown the NCS using the S/G PORV's, disregarding the , 100 degree F/hr cooldown limit, and maintain NC Pressure constant in an attempt to establish NCS Subcooling
i B. Cooldown the NCS using the S/G PORV's, staying within the 100 degree F/hr cooldown limit, this will result in a subsequent decrease in NCS Pressure allowing increased ECCS flow t C. Cooldown the NCS as quickly as possible using the Steam ' Dump Valves to the condenser, this will result in a i subsequent decrease in NCS Pressure allowing increased - ECCS flow
D. Cooldown the NCS as quickly as possible using the S/G 'e PORV's, this will result in a subsequent decrease in NCS Pressure allowing increased ECCS flow , f ANSWER: D MISCINFO: t REFERENCES: EP/1/A/5000/12.1 LESSON: OP-MC-EP-EPl2 TASK: MC-8311 OBJECTIVES: LPRO OBJ : 3 TIME: 3 MINUTES i LPS0 OBJ: 3
l KA000074 G.12 4.3/4.4 REV. DATE: 05/12/93 , l
i O O l l 7 DIESEL GENERATOR .' DCDGR06 i
' - 1 Pt(s) While performing the "lA" diesel generator (D/G) operability l test, the "1A" D/G heat exchanger supply isolation valve ' (lRN-70A) did NOT automatically open, but was opened
manually. All other components operated as they should have during the operability test. , Assuming 1RN-70A will NOT automatically open, which of the following is correct? l A. If 1RN-70A is left open, with power removed, the "lA" D/G l is operable B. Until IRN-70A is repaired and can again automatically open (regardless of its position) "1A" D/G must be declared inoperable C. With 1RN-70A closed and incapable of automatically opening, but capable of being manually opened, the "lA" D/G is operable D. The operation of 1RN-70A has no effect on the operability of "lA" D/G , ANSWER: A MISCINFO: *-Need to provide DBD T.A.C. Vol.II as reference material. REFERENCES: PT/1/A/4350/02A TAC LESSON: OP-MC-DG-DG TACK: MO-3328.5 , MO-4314 i i OBJECTIVES: LPRO OBJ: 10 TIME: 4 MINUTES LPSO OBJ: 10 KA064000 G.11 3.4/3.9 REV. DATE: 05/12/93 i
1
O O ~ l ' 8 EP 13 SERIES . EPEP13R01
' 1 Pt(s) While checking to see if an immediate NC System Feed and Bleed should be initiated while responding to a loss of Secondary Heat Sink the following information is noted: 1) NC Loop T-Hot 595'F slowly increasing 2) NC System Pressure 2330 psig 3) Containment Pressure 3.7 psig 4) Avg of 5 Highest Core Exit T/C's 610*F slowly increasing 5) PORV's INC-32 and 1NC-36 Open 6) S/G Wide Range Levels 20% slowly decreasing Based on this information, which course of action should be followed? A. A feed and bleed IS warranted since S/G wide range levels are < 22% and core exit T/C's are increasing B. A feed and bleed IS warranted since containment pressure is greater than 3 psig and the pressurizer PORV's are open C. A feed and bleed IS NOT warranted since S/G wide range levels are greater than 9% even though core exit T/C's are increasing l D. A feed and bleed IS NOT warranted since NC System l subcooling is greater than O'F . l ANSWER: A
MISCINFO: REFERENCES: EP/1/A/5000/13.1 LESSON: OP-MC-EP-EP13 TASK: MO-8312 OBJECTIVES: LPRO OBJ: 2 TIME: 4 MINUTES LPSO OBJ: 2 KA000074 G.12 4.3/4.4 REV. DATE: 05/12/93 . >
_. O O . 9 LOSS OF CONTROL ROOM , APAPl7R01 ! ! l
l
- 1 Pt(s) The Control Room has become uninhabitable and has been evacuated. The following data was collected from OAC General Program 76: 1) Three of the Control Bank "D" Group 2 Rods D12, M4 and H8 are NOT fully inserted ! 2) All other rods are fully inserted 3) Current boron concentration is 400 PPM Based on these conditions, we must Emergency Borate to a minimum value of PPM to account for the rods not fully inserted. A. 2000 B. 850 , C. 700 D. 550 ANSWER: B MISCINFO: REFERENCES: AP/1/A/5500/17 l LESSON: OP-MC-CP-SS TASK: MO-7315 { l l OBJECTIVES: LPRO OBJ : 2 TIME: 3 MINUTES l LPSO OBJ: 2 1 KA000005 EA2.03 3.5/4.4 REV. DATE: 07/01/93 1 l . ! i I
.. .- - ! t O O 10 EP 02, 06, 08 SERIES .. EPEP02R10 i , i . i l 1 Pt(s) A reactor trip and SI occurred at 0100. Initial reactor
power was 100%. The control room operators were unable to l align the ECCS into Cold Leg Recirculation due to the l inability to open NI-184 and NI-185 (Containment Sump Suction i i Valves). The following conditions exist at 0500: 1 1. Total SI Flow Rate: 500 GPM I ! ' l 2. ECCS Systems aligned for injection mode ( 3. FWST level: 110 inches ! 4. Subcooling: - 5'F 5. All NC pumps are off r 6. Attempts are being made to open NI 184 & 185 locally Based on these conditions the control room operators should: A. Establish SI flow to maintain Por level and RVLIS indications stable j I B. Reduce total SI flow to approximately 105 GPM l C. Maintain 500 CPM flow and proceed to the next step D. Reduce total SI flow to approximately 215 GPM ' ANSWER: D , MISCINFO: REFERENCES: EP/1/A/5000/06 t l LESSON: OP-MC-EP-EP2 TASK; MO-8306 l MO-4610 , OBJECTIVES: LPRO OBJ: 7 TIME: 4 MINUTES
LPSO OBJ: 7 ' KA000011 G.12 4.0/4.1 REV. DATE: 07/01/93 1 I i < -
i O O . l I , ' 11 CORE PERFOPJ'.ANCE
CTHCPR01 ' i 1 Pt(s) While performing a Unit I startup with Turbine load at 47% RTP, the operator notices that the QPTR is 1.03. The power escalation: ' A. May continue provided the QPTR is calculated to be s 1.09 at least once per hour and reactor power is s 97% B. May continue with the approval of the Operations Staff Manager (Currently D. Bumgardner), no other limitations need be imposed C. May continue, but not to exceed 50% power, until the QPTR , out of limit condition has been identified and corrected D. Must be terminated and power reduced at least 3% for each 1% of indicated QPTR in excess of 1.0 ANSWER: A MISCINFO: REFERENCES: OP/1/A/6100/03 Tech. Spec. 3.2.4 Tech. Spec. Int. QPTR 3.2.4 (6-6-85) LESSON: OP-MC-CTH-CP TASK: MO-5303.2 OBJ ECTIVES : LPRO OBJ: 38 TIME: 3 MINUTES LPSO OBJ: 38 KA001000 K5.10 3.9/4.1 REV. DATE: 07/01/93
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. _ , 1 . 12 EP 03 SERIES . EPEP03R02 , 1 . " 1 Pt(s) Following a steam line break outside of the containment on "1B" S/G, the RO has terminated safety injection following an < l excessive cooldown and establishes normal charging flow. Once VI is restored to the containment the RO observes the following: , NOTE: The fault on "1B" S/G occurred 40 minutes prior to } this with the unit at 50% RTP j a. NC System WR Pressure 1750 psig - b. NC Loop Cold Leg WR Temp "1A" 500*F "1B" 350*F ' "1C" 500*F "1D" 500*F ,
c. NC Loop Hot Leg WR Temp "lA" 505'F , "1B" 400*F - ' "1C" 505'F "1D" 505'F d. "1B" S/G WR Level 0%
e. "1B" S/G Pressure 0 psig
Based on this information the RO should decrease NC System pressure. Which of the following would be an acceptable ,
pressure? , A. 1600 psig i , B. 1500 psiS '
I [ C. 1000 psig D. 300 psig . . ANSWER: C , MISCINFO:
, REFERENCES: EP/1/A/5000/3.1 8/6/90 LESSON: OP-MC-EP-EP3 TASK: MO-8303
! > OBJECTIVES: LPRp OBJ: 3b TIME: 3 MINUTES ! LP50 OBJ: 3b I KA000040 G.12 3.8/4.1 REV. DATE: 05/12/93 e
. O O 4 1 13 EP 03 SERIES
EPEP03R03 2 d . 1 Pt(s) While terminating safety injection following a steam line break, the RO is directed to ensure that the VCT makeup
j control system is set for a boron concentration greater than , the NC system shutdown boron concentration. The RO notes the
following: 1) NC System Pressure 800 psig 2) NC System Temperature 380*F ' 3) Boric Acid Tank concentration 7500 ppm
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4) Total Blender Flow Rate 90 gpm ! 5) Shutdown Boron Concentration 1100 ppm ! , '
What is the minimum that the potentiometer should be set?
A. 1.7 B. 3.3 , , C. 6.8
D. 13.2 , ANSWER: B MISClNFO: REFERENCES: OP/1/A/6100/22 Enc. 4.3 Table 5.2 LESSON: OP-MC-EP-EP3 TASK: MO-8303 MO-3305 ' OBJECTIVES: LPRO OBJ: 3b TIME: 3 MINUTES LPSO OBJ: 3b KA004000 A4.04 3.2/3.6 REV. DATE: 05/L2/93 . i
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14 EMERGENCY Pl>.N EPEMPR13 + ' 1 Pt(s) The following conditions occurred with Unit 1 at 1% kTP: 1. The 1A D/G is under going repairs due to a fuel injection failure discovered during Surveillance Testing 2. While running D/G 1B to meet the Tech. Spec. required Surveillance Test the 1B D/G loses oil pressure and throws a rod 3. A lightning strike in the switchyard causes PCB's 8, 9, 11, and 12-to open If it takes 10 minutes to re-start the 1A D/G, what course of action, concerning Emergency Classifications, should be taken to address the situation? A. You must declare a Notification of Unusual Event B. You must declare an Alert C. You must declare a Site Area Emergency
D. No Emergency Classification is warranted ANSWER: B MISCINFO: REFERENCES: RP/0/A/5700/00 LESSON: OP-MC-EP-EMP TASK: MO-9300 OBJECTIVES: LPRO OBJ: 8 TIME: 4 MINUTES LPSO OBJ: 10 KA000055 G.2 2.9/4.1 REV. DATE: 07/01/93 a _
_ l. O O ' r . l ' ! 15 RESIDUAL HEAT REMOVAL SYSTEM ., PSNDR01 , 1 Pt(s) Unit 1 is in Mode 5. The pressurizer is being maintained at 95% level on the cold calibration instrument while preparations are being made to drain the NC System. ND Train > "13" is in operation. All four S/G's are being drained with narrow range levels of 4-5%. r I&E requests permission to perform a surveillance test on the ND pump "lA" supply breaker. The breaker will be out of service for at least 2.5 hours. Select the correct action concerning this condition. A. The SRO should allow I&E permission to perform the surveillance test i B. The surveillance test can not be performed until all S/G l levels are increased to 101 l ! ! C. The SRO should not allow I&E to perform the surveillance , ' ! test D. Surveillance testing can be allowed if at least ono S/G's ' i level is increased > 12% 1 ANS'JER: C , MISCINFO: REFERENCES: Tech. Spec. 3.4.1.4.1 I PT/1/A/4600/03A LESSON: OP-MC-PS-ND TASK: MO-3311 OBJ ECTIVES : LPRO OBJ : 9 TIME: 4 MINUTES LPS0 OBJ: 9 KA005000 G.5 3.2/3.8 REV. DATE: 05/12/93
_. -, . O O . ^ 16 FUEL HANDLING FHFCR06 . ! 1 Pt(s) The following conditions exist: 1) Unit 1 is in Mode 6 2) Defueling in progress with one half of the assemblies removed ! 3) "A" ND Pump operating with 3200 gpm flow 4) "B" ND Pump is inoperable ' 5) NC WR level meter is inoperable 6) Refueling cavity water level is 25 inches below the cavity window Evaluate the above information and select the proper action based an that evaluation: . ' A. Immediately suspend core alterations and begin Emergency Boration of refueling cavity to ensure Keff is 2.95 B. Immediately suspend all operations involving movement of fuel assemblies or control rods within the reactor vessel and take action to immediately restore ND pump 1B to operable or restore refueling canal level to 2 23 f t. above reactor vessel C. Continue to defuel the unit while monitoring ND Train "A" temperature to be s 140*F, ND Train "A" flow to be 21000 gpm and radiation levels within the reactor building to be stable or decreasing D. Suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operable and operating status as soon as possible. ANSWER: B MISCINFO: REFERENCES: Tech. Spec. 3.9.8.2, 3.9.9 ' PT/1/A/4600/03A ' PT/1/A/4600/03B j LESSON: OP-MC-FH-FC TASK: MO-9313 MO-4326 MO-4327 OBJECTIVES: LPRO OBJ: 14 TIME: 4 MINUTES LPSO OBJ: 14 KA015000 C.11 3.1/3.8 REV. DATE: 07/01/93 . > . . . .
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17 EP 09 SERIES EPEP09R01 , 1 Pt(s) Given the following: 1) 1 ETA and lETB both have a zero bus voltage 2) The Unit was at 100% just prior to the event i 3) Current core burnup is 280 EFPD 4) Pressurizer level is slowly decreasing 5) Intact S/G's are being depressurized to minimize NC System inventory loss ' Which of the following set of conditions would permit
continuing the S/G depressurization process? A. An NC Pressure of 210 psig, "C" S/G Pressure of 155 psig, J NC System Cold Leg Temperature of 375 degrees F. ' All S/G Levels are 3% and slowly decreasing, Total CA flow is 1099 gpm t B. An NC Pressure of 300 psig, i "B" S/G Pressure of 175 psig, 7 NC System Cold Leg Temperature of 370 degrees F, i l All S/G Levels are 27% and being maintained constant, , Total CA flow is 900 gpm - ! C. An NC Pressure of 190 psig, [ "C" S/G Pressure of 175 psig, 1 NC System Cold Leg Temperature of 373 degrees F,
All S/G Levels are 21% and being maintained constant. Total CA flow is 850 gpm ' D. An NC Pressure of 225 psig, "B" S/G Pressure of 190 psig, NC System Cold Leg Temperature of 375 degrees F, All S/G Levels are 7% and being maintained constant, Total CA flow is 850 gpm
ANSWER: D ! MISCINFO: i i REFERENCES: EP/1/A/5000/09 , LESSON: OP-MC-EP-EP09 TASK: MO-8308 OBJECTIVES: LPRO OBJ: 2 TIME: 4 MINUTES LPSO OBJ: 2 KA000055 C.12 3.9/4.0 REV. DATE: 06/21/93 i 5 t .
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i 1 LICENSED REQUAL ANNUAL EXAMINATION 4 1 i QUICK ANSWER KEY l - 3 . PART "B" RO EXAM , )
ANS. ! i 1 ] 1 D ) 2 B
3 D 1 4 D ' i
5 B
6 A i , 7 C , 8 B 9 B i 10 D ' 11 C i , 12 C 13 A ' 14 B i 15 A 16 B i 17 C - - . . .
. . . . . . . - - . - , O O ' - -
t , l LICENSED REQUAL ANNUAL EXAMINATION - ! f QUICK ANSWER KEY . PART "B" SRO EXAM - .
ANS. , 1 B 2 A 3 C i i 4 B 5 C 6 D , 7 A ' 8 A , 9 B ! ' 10 D 11 A I l 12 C 13 8 1 14 B 15 C 1 ' 16 B 17 D l l l l I , , , ,- .. .. ... .. -
, _ _ _ . _ _ _ _ ._ 0 O - - _ , LICENSED REQUAL ANNUAL EXAMINATION
QUICK ANSWER KEY , ' PART "A" RO EXAM , i
ANS. ! . 1 D 2 B , i , 3 B i 4 C
l 5 C ' l 6 A i 7 A I 8 D ' 9 A i 10 C 11 A 12 B i 13 D 14 A 15 B i 16 D 1 1 17 C l ~ 18 D 19 B 20 D i , , . . - . - . . . . - - - _ . -
i ( O O ' - 1 - I LICENSED REQUAL ANNUAL EXAMINATION r 1 l ' QUICK ANSWER KEY , PART "A" SRO EXAM i 1
ANS. , ! 1 A ! f 2 C 3 A . 4 8 ! 5- D 6 B . 7 D I 8 C l 9 C 10 D ! l l 11 B 12 A 13 D i 14 B 15 A 16 C 17 C ' 18 A 19 A 20 D i ! -- .
, ' OP-MC-SSE-02 . ' Rev. 04/06-30-93/WGH/ygg Page 1 of 6 . PROGRAM: Operations Training ' MODULE: - License Requalification TOPIC: Static Simulator Scenario #2 (SSE-2) ' EXERCISE: Pressurizer Code Safety LOCA
' OVERVIEW- This Exercise Guide will test the candidates ability to correctly diagnose transient / accident conditions, evaluate plant and response and/or specify the correct course of action for severalinstrumentation and equipment malfunctions based on control room indications and appropriate control room reference material. PREREQUISITE KNOWLEDGE LEVEL: Previously licensed at the RO or SRO level. , t ! ! March 10,1989/JRP
._ n o i . ' v O P-MC-SSE-02 ' Page 2 of 6 { , i 1.0 SYNOPSIS l ' This scenario presents the candidate with the TD CA Pump running due to a valve failure, a failed pressurizer level transmitter and a deenergized D/G load ! sequencer. The candidates ability to diagnose the ' event, evaluate plant l response and conditions, and specify a course of action is tested using j selected questions associated with the exam scenario. , ! 2.0 INITIAL CONDITIONS i 2.1 Reset Simulator to IC-30 with Rod Step Counters ON. [ t i 2.2 Plant Conditions , i Unit One ! Power History 3.3%,144 EFPDS ' . Boron 1010 ppm ! Tave 559 deg. F ! Xenon 8 pcm
Samarium difference -327 pcm j i 2.3 Select: ICCM-A = RVLIS ! ICCM-B = Core Cooling j 2.4 Check Rod Step counters for proper indications. 2.5 Rotate chart recorders to clear paper. 3.0 PROCEDURE 3.1 Turn Chart Recorder Power On i 3.2 Run Simulator i ' 3.3 Set up OAC as follows: ! A. Select Display Group 069 to display on Monitor Video
B. Select Nuclear 06 to display on Utility Video ' C. Acknowledge OAC Alarm Video !
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.. - . . . . --. . .- - - . .. . 'O P-MC-SS E-02 ' Page 3 of 6 1 .i k j 3.4 Acknowledge Control Room Annunciators ) , y- ! ' 3.5 Freeze Simulator , 4 , 3.6 insert the following Malfunction: ! i A. Insert MALF EQB2 (Sequencer Control Power Failure) l
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I 1. Select D/G = 1 = D/G "A" 2. Set Time Delay = 5 secs 3. Activate B. Insert MALF EDA 3 (DRPI Data "A" > Data "B" by > 12 steps) 1. Select Rod =~ H8 ' 2. Set Time Delay = 10 secs 3. Activate 3.7 Insert the following Overrides: A. Insert OVR SWITCHES / CONTROLLERS CA22 (TD AFWP SEL SW) 1. Set Value = 1 = Hi 2. Set Time Delay = 10 secs 3. Activate B. Insert OVR INDICATORS SA2A (Stop Viv Line "1C" Pos ind Light) 1. Set value = 2 = Low 2. Set Time Delay = 15 secs ' 3. Activate
0 - 'OP-MC-SSE-02 Page 4 of 6 . C. Insert OVR INDICATORS SA2B (Stop Viv Line "18" Pos ind Light) , 1. Set Value = 1 = Hi 2. Set Time Delay = 20 secs 3. Activate D. Insert OVR INDICATORS SG18 ("1C" N/R Lvl Ch.1 Meter) 1. Set Value = 48% 2. Set Time Delay = 25 secs 3. Activate E. Insert OVR INDICATORS SG39 ("1C" N/R Liv Ch. 2 Meter) 1. Set value = 40% 2. Set Time Delay = 30 secs 3. Activate l F. Insert OVR INDICATORS SG40 ("1C" N/R Lvl Ch. 3 Meter) 1. Set value = 40% 2. Set Time Delay = 35 secs 3. Activate i l i G. Insert OVR INDICATORS SG41 ("1C" N/R Lvl Ch. 4 Meter) j 1. Set value = 40% 2. Set Time Delay = 40 secs 1 3. Activate ' l 1 l l { ! .
(~ ' ~ % p.MC-SSE-02 Page 5 of 6 . H. Insert OVR INDICATORS SG31 ("1B" N/R Lvl Ch. 2 Meter) ' 1. Set value = 44% l l 2. Set Time Delay = 45 secs ! 3. Activate l 1. Insert OVR TRANSMITTER ILE3 (Pzr Lvl Ch. 3) 1. Set value = 0% 2. Set Time Delay = 50 secs 3. Activate 3.8 Perform the following manual actions once the simulator is in its 60 sec run: A. Type in "RUN60" and depress Enter B. Trip the "1 A" MFWP C. Reset "A or B" CF Pump Recirc Valve Closure Circuit D. Acknowledge Control Room Annunciators E. Acknowledge OAC Alarm Video 3.9 When the simulator " Freezes", verify the General Warning LED for Rod H8 is lit. If not, toggle between Run & Freeze until the GW light is lit in Freeze. 4.0 TURNOVER SHEET 4.1 See Attachment 1
l p/ , . SOP-MC-SSE-02 w Page 6 of 6 . TURNOVER SHEET Attachment 1 . Conditions Existing at the time of the Incident: A. Plant Conditions
l - Power History 3.3%,144 EFPDS l - Boron 1010 ppm l - Tave 559 deg. F l - Xenon 8 pcm - Samarium difference -327 pcm B. Turnover information: 1. The "1 A" MFWP is ready to be reset and returned to service. 2. SA-49 (SM FRM S/G B to TD CA PMP ISOL) has failed open. l&E has reported that the failure is NOT an auto start and that NO auto start exists. 3. Pressurizer level channel 3 has just failed and no action has been taken as of yet. 4. "1 A" D/G sequencer control power was incorrectly removed and will be reenergized within the next 15 - 20 minutes.
C. Current Procedure: OP/1/A/6100/03 Step 3.2. 89 , Simulator and chart recorders are " Frozen", the OAC and STA Monitors cannot be changed. }}