IR 05000483/1997001
| ML20137G136 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/26/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20137G122 | List: |
| References | |
| 50-483-97-01, 50-483-97-1, NUDOCS 9704010333 | |
| Download: ML20137G136 (200) | |
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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.: 50-483 License No.: NPF 30 Report No.: 50-483/97-01 Licensee: Union Electric Company i
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Facility: Callaway Plant Location: Junction Hwy. CC and Hwy. O l Fulton, Missouri ;
Dates: February 24-28, 1997
! Inspectors: H. Bundy, Chief Examiner R. Lantz, Examiner
! T. McKernon, Examiner l
M. Murphy, Examiner l
l Approved By: J. L. Pellet, Chief, Operations Branch l l Division of Reactor Safety l
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l ATTACHMENTS:
- i Attachment 1: Supplemental Information '
Attachment 2: Simulation Facility Report -
Attachment 3: Final Written Examination and Answer Key .
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9704010333 970326 3 PDR ADOCK 0500 G l
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-2-EXECUTIVE SUMMARY Callaway Plant NRC Inspection Report 50-483/97-01 l
NRC examiners evaluated the competency of six senior reactor operator and five reactor ;
operator license applicants for issuance of operating licenses at the Callaway Plant facilit The licensee developed the initial license examinations using the pilot process program ;
guidance contained in Generic Letter 95-06 and NUREG 1021, Supplement 1, " Operating ;
Licensing Examiners Standards." NRC examiners reviewed, approved, and administered the examinations. The initial written examinations were administered to all 11 applicants on February 24,1997, by facility proctors in accordance with instructions provided by the chief examiner. The NRC examiners administered the operating tests on February 25-27, 1997. All of the senior reactor operator applicants and two of the five reactor operator applicants displayed the requisite knowledge and skills to satisfy the requirements of 10 CFR 55 and were issued the appropriate licenses. Orie reactor operator applicant displayed marginal knowledge and skills, and the licensing decision is undergoing further revie Two of the reactor operator applicants failed the written examination and were denied license Operations
- The control room operators exhibited professional demeanor and the shift turnover briefing observed was effective and comprehensive (Section 01).
- All six applicants passed the senior reactor operator written examination. Two of five applicants passed the reactor operator written examination. Another reactor operator applicant demonstrated marginal knowledge on the written examination ;
and the finallicensing decision on his application was delayed pending further I review. Overall, the reactor operator applicants demonstrated a marginal knowledge level on the written examination. No generic broad knowledge or training weaknesses were identified as a result of evaluation of the graded examinations (Section 04.1).
- All 11 applicants passed the operating test. Minor performance and procedure deficiencies were identified for licensee and applicant consideration and corrective action as appropriate (Section 04.2).
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- The licensee submitted an acceptable examination outline (Section 05.1.1).
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- The licensee submitted examination package was of hig5 quality and adequate for administration. The licensee staff was responsive in proviuing enhancements identified during the review process (Section 05.1.2).
- The simulator supported the examinations well. One minor malfunction impacted examination administration, but did not affect examination validity (Section 05.2).
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- The chief examiner concluded that the Updated Final Safety Analysis Report wording was consistent with the observed plant practices, procedures, and/or parameters (Section E2.1).
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Reoort Details Summarv of Plant Status !
The plant operated at essentially 100 percent power for the duration of this inspectio l. Operations
01 Conduct of Operations Inspection Scope During the in-plant main control room portion of the operating test walkthroughs, the examiners observed the on-shift operators during routine operations of the facility, Observations and Findinog l
The demeanor of the operators was professional and crew communications were 1 effective. One shift turnover briefing observed was effective and comprehensive and consistent with briefings given by applicants for senior reactor operator licenses during the dynamic simulator section of the operating tes Conclusions, The control room operators exhibited professional demeanor and the shift turnover briefing observed was effective and comprehensive.
03 Operations Procedures and Documentation An apparent weakness in Procedure OTO-ZZ-OOOO1, " Plant Control from ASP with ,
Control Room Fire," Revision 14, is discussed in Section 0 )
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04 Operator Knowledge and Performance 04.1 Initial Written Examination insoection Scooe On February 24,1997, the facility licensee proctored the administration of the written examination appivved by the chief examiner and NRC Region IV supervision to five individuals who had applied for initial reactor opera.or licenses, three individuals who had applied for initial instant senior reactor operetor licenses, and
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three individuals who had applied for initial upgrade senior reactor operator license The licensee graded the written examinations and the staff reviewed its result The license 6 also performed a post-examination question analysis which was reviewod by the examiners, Observations and Findno:,
Th3 minimum passing score was 80 percent. The scores for senior reactor operator applicants ranged from 83 to 93 percent, with an average score of 89 percent. The scores for reactor operator applicants ranged from 73 to 92 percent, with an average score of 81.2 percent. Two reactor operator applicants failed the written examination with scores of 73 and 77 percent. A third reactor operator applicant passed the written examination with a marginal score of 80 percent. Overall, the reactor operator applicants demonstrated e marginal level of knowledge on the written examination. More than half of all applicants missed the following questions which had the same number on both examinations: 3, 4, 55, 64, 84, and 90. Also, more than half the applicants missed C.uestion 33 on the senior reactor operator l
examination. All of the above questions were determined by the licensee to be
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valid and the chief examiner concurred with this determination. No broad training or knowledge weaknesses were identified. Reasons for missing these questions ;
appeared to be related to question difficulty and isolated training weaknesses. The i
- licensee initiated appropriate actions to upgrade candidate specific knowledge and l correct specific training weaknesses.
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l Conclusions All six applicants passed the senior reactor operator written examination. Two of five applicants passed the reactor operator written examination. Another reactor
! operator applicant demonstrated marginal knowledge on the written examination l and the final licensing decision on his application was delayed pending additional I
review in response to challenges of the proposed licenne denials. Overall, the reactor operator applicants demonstrated a marginal krtowledge level on the written examination. However, no broad knowledge or training, weaknesses were identified as a result of evaluation of the graded examination .2 Initial Ooeratino Test Insrection Scope i
The examination team administered the various portions of the operating
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examination to the 11 applicants on February 25-27, 1997. Each applicant participated in two dynamic simulator scenarios. Each also received a walkthrough test which consisted of ten system tasks and four administrative areas, except the upgrade senior reactor operator licenses were tested on only five system tasks with four administrative area l 6-
b. Observations and Findinas
All applicants passed all portions of the operating test. The applicants perforr ad wellin the dynamic simulator scenario l However, during Scenario 2 following a large break loss-of-coolant accider:t with a I l subsequent failure of the emergency diesel generator loss-of-coolant accident electricalload controller (sequencer), all of the crews experienced difficulty in l realigning Train B control room ventilation isolation system components in j l accordance with Procedure E-0, " Reactor Trip or Safety injection," Revision 1B2, i Step 15b. The " response not obtained" column required performing Attachment 11 l (
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to the procedure. To perform Attachment 11 successfully, resetting the safety injection actuation system and containment spray actuation system was required.
l None of the crews reset both signals before performing Attachment 11. This failure l was of minimal safety significance because the Train A ventilation components l adequately accomplished the safety function. The licensee issued a training field report dated February 26,1997, to upgrade training materials.
I in the same scenario an automatic swap of the residual heat removal pump suction from the refueling water storage tank to the recirculation sump f ailed to occur. In attempting to perform a manual swap, the applicant in the control room supervisor !
position on one crew misread the procedure and gave orders which would have i resulted in both residual heat removal pump suction valves being open. Valve interlocks prevented this from occurring. Eventually, he discovered his error and j properly read the procedure to accomplish the manual swap. The applicant in the
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reactor operator position attempted to accomplish the swap incorrectly twice without challenging the validity of the initial directions. These actions delayed
! placing the plant in a stable conditio Most of the applicants performed well on the walkthrough portion of the tes However, a few generic minor performance weaknesses were observed. Several of the applicants experienced uifficulty in controlling steam generator water levels from the auxiliary shutdown panel in accordance with Procedure OTO-ZZ-00001, " Plant Control From ASP With Control Room Fire," Revision 14. The procedure sacked some specific instructions, such as how to control auxiliary feedwater flow to l i
Steam Generator A. Also, the applicants displayed unfamiliarity with this specific j l
task. The licensee issued a training field report to include this procedure in i Requalification Training Cycle 97-3 to accomplish additional procedure and operator
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When adding blended water to the refueling water storage tank in accordance to ,
Procedure OTN-BG-00002, Reactor Makeup Control and Boron Thermal l Regeneration System," Revision 13, several applicants failed to perform Step 4.5.10 ;
correctly, in that they did not immediately turn the reactor coolant makeup water i control switch to OFF after insertion of the designated amount of blended water in accordance with Step 4.5.10. The examiners perceived that the applicants thought
7-that flow would stop when the designated amount of blended water had been pumped into the refueling water storage tank. However, the flow of concentrated boric acid continued until the switch was placed in OFF. The safety consequences of this delayed action were minimalin that it resulted in a slight over boration of the refueling water storage tank. However, it inoicated an applicant system operation weaknes Although it did not result in any test f ailures, several of the applicants exhibited weakness in answering the job performance measure system questions. In many instances, they were able to correctly answer only one of the two system followup questions. This resulted in a marginally satisfactory grade for epplicants on several system Conclusions All 11 applicants passed the operating test. Performance ~and procedure deficiencies not sufficient for license denial were identified for licensee and applicant consideration and corrective action as appropriate.
05 Operator Training and Qualification 05.1 Initial Licensina Examination Devetooment The facility licensee developed the initiallicensing examination in accordance with guidance provided in Generic Letter 95-06, " Changes in the Operator Licensing Program."
05.1.1 Examination Outline insoection Scope The f acility licensee submitted the initial examination outline on December 17, 1997. The chief examiner reviewed the submittal against the requirements of NUREG-1021, " Licensed Operator Examiner Standards" Revision 7, Supplement 1, and NUREG/BR-0122, " Examiner's Handbook for Developing Operator Licensing Written Examinations," Revision Observations and Findinas The staff determined that the initial examination outline satisfied all requirements and the chief examiner advised the licensee to proceed with examination development. The licensee authors of the outline had communicated informally with the chief examiner concerning the contents of the outline on several occasions prior to the formal submitta !
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8 Conclusions l
The licensee submitted an high quality examinaHon outline on December 17,199 .1.2 Examination Packaae i Inspection Scoce
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The facility licensee submitted the completed examination package on January 24, l 1997. The chief examiner reviewed the submittal against the requirements of NUREG-1021, " Licensed Operator Examiner Standards" Revision 7, Supplement 1, and NUREG/BR-0122, " Examiner's Handbook for Developing Operator Licensing Written Examinations," Revision Observations and Findinas The draft examination was transmitted by the licensee to the NRC by a letter dated January 24,1997. The draft written examination contained 127 questions, 73 were designated to be included on both reactor operator and senior reactor operator examinations. Most of the questions were developed for this examinatio Only four questions from the facility examination question bank were used on both !
examir:ations and an additional question from the examination bank was used on the senior reactor operator examination only. The draft examination was considered technically valid, to discriminate at the proper level, and responsive to the knowledge and abilities sample plan submitted by the licensee on December 17, 1996. However, the chief examiner provided enhancement suggestions for 14 questions which appeared on both the reactor operator and senior reactor operator examinations,4 questions which appeared only on the reactor operator examination, and 4 questions which appeared only on the senior reactor operator examination. The suggestions generally related to specific questions with regard to clarity of wording in the stem, use of inadvertent cues, plausibility of distractors, or levet of knowledge required. After extensive discussion of the chief examiner's suggestions, the licensee modified or rewrote 11 questions appearing on both examinations, 2 questions appearing only on the reactor operator examination, and 2 questions appearing only on the senior reactor eperator examination. The chief examiner concurred with the resolution of his suggestions and the final produc The licensee performed a postexamination analysis and recommended that no further changes be made to the written examinations. The examiners concurred with this analysis and recommendatio The licensee submitted three dynamic scenarios, including one backup scenario which was not used during the examination. The chief examiner noted several errors on Examiner Standards Form 301-5, " Transient and Event Checklist," which were corrected by the licensee. These evolution assignment errors did not invalidate any of the dynamic scenarios. The chief examiner made a generic 1
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comment that the expected operator / plant response forms were a compilation of expected actions, which did not indicate which applicant was expected to perfor The licensee revised the forms for all three scenarios to indicate specific expected applicant response. Other comments, which the licensee incorporated, included editorial and enhancements to facilitate administration. The licensee initiated a few minor editorial enhancements to the scenarios to facilitate administration during the chief examiner's preparation week onsit To support the systems walkthrough section of the operating test, the facility licensee provided job performance measures developed to evaluate selected operator tasks that contained well written task elements, performance standards, and comprehensive evaluator cues. Eleven job performance measure tasks with
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two followup questions each were submitted. One job performance measure was for backup and was not used. The chief examiner provided comments concerning enhancement of the walkthrough test outline, which were incorporated. The chief examiner questioned the critical step assignments for three job performance measures and the licensee made additional critical step assignments for these job
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performance measures. Also, the licensee rewrote three job performance measure
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questions in response to the chief examiner's enhancement suggestions. Other
. chief examiner comments, which the licensee incorporated, concerned use of j references on specific job performance measure questions. A minor editorial change a was made to one job performance measure to correct a typographical error, which
- the chief examiner discovered when examining the first applican The licensee submitted both job performance measures and questions to cover the administrative section of the walkthrough test. One set was submitted for reactor operator applicants and another set was submitted for senior reactor operator applicants. The job performance measures submitted were acceptable. However, to f acilitua administration some minor changes were made to one job performance measure during preparation week. Also, the expected accuracy requirement for this job performance measure was upgraded in response to a chief examiner commen After reviewing the chief examiner's comments, the licensee made minor changes to the administrative questions as follows. The chief examiner evaluated Question 1 in Section 3 of the reactor operator test as discriminating at too low a level and the licensee replaced it. Also, the chief examiner evaluated both questions in Section 4 to be beyond the scope of reactor operator responsibilities and the licensee revised them to be more appropriate to operator level required knowledg On the senior reactor operator set, the licensee tsm
- :ved what the chief examiner considered a cue for Question 1 in Section Conclusions Overall, tP.e written examination and operating test materials submitted were of high quality, discriminated at the appropriate license level, and were adequate for administration. Further, licensee staff were highly responsive in responding to enhancement suggestions developed during the review process. No significant changes to examination materials were required as a result of administration.
05.2 Simulation Facility Performance Inspection Scope The examiners observed simulator performance with regard to fidelity during the examination validation and administratio Observations and Findinas Only one simulator performance problem affected examination administratio During performance of job performance measures, the annunciator reset switch at the primary console failed to function. This caused a slight delay in administration while the simulator support personnel corrected the proble A few simulator performance problems were observed by the cNef examiner during examination validation and are listed in Attachment 2. All of these problems had been previously identified by the licensee, and none affected examination validity, Conclusions The simulator supported the examinations well. One minor malfunction impacted examination administration, but did not affect examination validit Ill. Enaineerina E2 Engineering Support of Facilities & Equipment E Review of the Uodated Final Safety Analysis Report Commitments A recent discovery of a licensee operating their f acility in a manner contrary to the Updated Final Safety Analysis Report description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the Updated Final Safety Analysis Report descriptions. While performing the inspection discussed in this report, the inspector reviewed the applicable portions of the Updated Final Safety Anahsis Report that related to the areas inspected. The inspector verified that the Updated Final Safety Analysis Report wording was consistent with the observed plant practices, procedures, and/or parameter . . . . . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ . . _ . . . . . . . . _ _ . _ _ . . . . _ . . . . _ _
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V. Manaaement Meetinas j i
X1 Exit Meeting Summary !
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The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on February 28,1997. The ,
licensee acknowledged the findings presente l The licensee did not identify as proprietary any information or materials examined during the inspectio l t
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ATTACHMENT 1 '
SUPPLEMENTAL INFORMATION l
PARTIAL LIST OF PERSONS CONTACTED I Licensee '
R. Affolter, Plant Manager l R. Barton, Operating Supervisor, Training i F. Biermann, Operating Supervisor, Training G. Czeschin, Superintendent, Training i J. Dampf, Shift Supervisor Operations, Training '
J. Davis, Engineer, Quality Assurance S. Halverson, Senior Training Supervisor, Simulator i S. Henderson 11, Operating Supervisor, Training R. Lamb, Superintendent, Operations J. Neher, Engineer, Quality Assurance R. Nelson, Operating Supervisor, Training D. Neterer, Assistant Superintendent, Operations G. Randolph, Vice President, Nuclear NRC l F. Brush, Resident inspector i
D. Passehl, Senior Resident inspector l
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ATTACHMENT 2 SIMULATION FACILITY REPORT Facility Licensee: Union Electric Company l
Facility Docket: 50-483 Operating Examinations Administered at: Callaway Plant i Operating Examinations Administered on: February 24-28,1997
These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 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 in response to these observation Only one simulator performance problem affected examination administration. During performance of job performance measures, the annunciator reset switch at the primary conr. ole failed to function. This caused a slight delay in administration while the simulator support personnel corrected the proble The following simulator deficiencies were identified during examination validation and did not impact the examination:
- Two computer points were available to monitor the ultimate heat sink level in the f
plant. Although the computer points were available in the simulator, they were not driven by the simulation system. The instructor had to insert a high level alarm to cause the applicant to pump down the ultimate heat sink level. The applicant had j to request local monitoring of leve !
- A modification had been installed in the plant to monitor N-16 detectors on the plant computer. The modification had not been completed on the simulator and was scheduled for corapletion in April 1997. The examiners had to cue the applicants that the N-16 monitors were out of service and alternate radiation monitoring instruments were used for performance of the dynamic scenarios and simulator job performance measure .i- A w a - s ATTACHMENT 3 FINAL WRITTEN EXAMINATION AND ANSWER KEY
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ES-401 Written Examination Cover Sheet Form ES-401-1
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION APPLICANT INFORMATION Name: Region: IV Date: February 24,1997 Facility / Unit: Callaway License Level: SRO Reactor Type: Westinghouse INSTRUCTIONS:
Use the answer sheet provided to document your answers. Staple this cover sheet on top of the answer sheet. Each question is worth one point. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
j All work done on this examination is my own. I have neither given nor received aid.
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Applicant's Signature l
RESULTS Exami .ation Value 100 Points Applicant's Score Points
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ES-402 Policies and Guidelines Attachment 2 for Taking NRC Written Examinations l
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l l l Cheating on the examination will result in a denial of your application and could result in more severe penalties.
l After you complete the examination, sign the statement on the cover sheet i
indicating that the work is your own and you have not received or given assistance in completing the examination.
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! To pass the examination, you must achieve a grade of 80 percent or greater.
i Each question is worth 1 poin !
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! There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the l
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answer shee ! Mark your answers on the answer sheet provided and do not leave any question I blan i l l
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' If the intent of a question is unclear, ask questions of the examiner onl l
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1 Restroom trips are permitted, but only one applicant at a time will be allowed to
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leave. Avoid all contact with anyone outside the examination room to eliminate 1 l even the appearance or possibility of cheatin . When you complete the examination, staple the examination cover sheet on top of the answer sheet and give it to the examiner or proctor. Remember to sign the statement on the examination cover sheet.
l 1 After you have turned in your examination, leave the examination area as defined l by the examiner.
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QUESTION #001
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When performing a boration to the reactor coolant system for a down power transient, the ,
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PZR heaters should be tumed on in manual to:
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A. Maintain PZR pressure in the normal operating range during the down powe P B. Allow an increased ramp rate for the down powe C. Equalize the reactor coolant system and PZR boron concentration D. Ensure positive PZR controlis established prior to starting the down powe ANSWER:
C. Equalize the reactor coolant system and PZR boron concentration RO #19 SRO #21 K/A #004000K601 OBJECTIVE #003AA4B1 1 REFERENCES: OTN-BG-00002, " Reactor Makeup Control and Boron Thermal ,
Regeneration System" l
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QUESTION #002 The plant experiences a sustained loss of all AC powcr.
Which ONE of the below would be used to makeup to the spent fuel pool due to low l spent fbel poollevel?
A. Pressurize VCT and use Reactor Makeup B. Diesel Fire Pump and fire hose :
l C. Gravity drain condensate storage tank D. Essential service water emergency makeup ANSWER- '
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B. Diesel Fire Pump and fire hose RO #41 SRO #44 K/A #033000G11 OBJECTIVE #003D220Z ;
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REFEBENCES: ECA-0.0, Step 23 i
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SRO Test QUESTION #003
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Which ONE of the below computer data quality codes indicates that the alarm function is
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still operable?
A.DALM B. DEL a
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ANSWER-
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K/A #194001 A115
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OBJECTIVE #003 A02D4 REFERENCES: OOA-RJ-00001
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SRO Test
, QUESTION #004 Preventive Maintenance is scheduled on the ' A' Condensate Pump Motor and its supply breaker PB0304. Which ONE of the following locations MUST be tagged in accordance with the Workman's Protection Assurance Program?
A. Breaker PB0304 local handswitch B. Condensate Pump Discharge Valve
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C. Racking Mechanism for Breaker PB0304 D. Main Control Board Switch AD-HIS-1 ANSWER:
C. Racking Mechanism for Breaker PB0304 RO #4 SRO #4 K/A #19400lK107 OBJECTIVE #003 A330F REFERENCES: APA-ZZ-00310 Page 20 i
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QUESTION #005 l l
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A Reactor Startup is in progress with Control Bank B at 50 steps and Reactor Power at l
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Which ONE of the following is required if Source Range Nuclear j
, Channel N32 fails high?
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l A. Place N32 in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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ft B. Verify all Rod Bottom Lights li ;
i l i C. Verify Shutdown Margin within one hou i D. Insert all Control Banks and repair channel N3 I r
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B. Verify all Rod Bottom Lights li !
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K/A #000032G11 OBJECTIVE #0110280E l L REFERENCES: OTO-SE-00001 i j E-0 l
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QUESTION #006 i
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The reactor tripped 5 minutes ago.
Which one of the following completes the statement concerning the heat transfer relationship between the RCS and Steam Generators?
The heat transfer rate between the RCS and the S/Gs will:
A. decrease as RCS temperature increases and AFW flow increases.
B. decrease as AFW temperature decreases and AFW flow increases.
C. increase as AFW temperature increases and RCS flow decreases.
D. increase as RCS temperature increases and AFW flow increases.
ANSWER:
D. increas6 as RC3 temperature increases and AFW flow increases.
RO #33 SRO #33 K/A #061000K501 OBJECTIVE #003D260R REFERENCES: T61.003 l i
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SRO Test QUESTION #007 Which of the following are allowable relaxations for Independent Verification when restoring a system requiring IV?
1. Comparing the tagout control sheet to current plant reference material (flow diagrams, procedures, etc.) to ensure adequacy of the tagou . Verifying status lights, annunciators, meter indications, etc. on the main control board that unequivocally depicts the equipment statu . Performing a functional test that verifies that the component is in the specified configuratio . When the concept of ALARA would be violate A. 2, 3, 4 B. 1, 2, 4 C. 1, 2, 3 D. 1, 3, 4 ANSWER *
A. 2, 3, 4 RO #1 SRO #1 K/A #194001K101 OBJECTIVE #003 A33 A6 REFERENCES: APA-ZZ-00310
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SRO Test QUESTION #008 The Callaway Plant is entering MODE 4 from MODE 3 with the following conditions:
. RCS pressure is being controlled at 650 psi * All wide range Cold Leg temperatures are 350 . Cold Overpressure Protection is in ARME * Loop 1 Wide Range Cold Leg temperature sensor, TE413B, fails lo Which ONE of the following describes the plant response to this failure? l l
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A. Only PORV 455A will ope B. Only PORV 456A will ope C. Both PORV 455A and 456A will open.
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D. Neither PORV 455A or 456A will open.
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B. Only PORV 456A will open.
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SRO #51 K/A #010000K403 OBJECTIVE #0110300C REFERENCES: DWG 8756D37 Sheet 6 l
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SRO Test QUESTION #009
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OTO-72-00001, Control Room Inaccessibility, requires operation of three ' Control Room j Isolation Transfer' switches on the Auxiliary Shutdown Panel, which isolate control and t indication of the associated devices from the control roo !
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Which ONE of the following describes the reason for operating these switches? ;
A. Prevent inadvertent actuation of components which are necessary to safely shutdown the plan B. Initiates a reactor trip and transfer control of the plant to the auxiliary shutdown pane C. Required by Technical Specifications action to ensure that auxiliary shutdown Operability is satisfie D. Transfers alarm and control of pressurizer heaters from the Control Roo ANSWER:
A. Prevent inadvenent actuation of components which are necessary to safely shutdown the plan RO #71 SRO #69 K/A #000067K304 j OBJECTIVE #0110480D :
REFERENCES: T61.0110.6 LP-#48 l
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! QUESTION #010 i i i 1 !
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j Reactor coolant system pressure will be controlled by: [
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- A. Steady state heaters and pressurizer spra !
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l C. Steady state heater only.
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j D. Backup heaters and pressurizer spra !
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l SRO #92 K/A #000007A103 l OBJECTIVE #0110090J
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SRO Test QUESTION #011
The following conditions exist:
- Containment pressure tammitter PT-937 declared inoperable
- Required Technical Specification Actions have been taken for channel 937 l
Which ONE of the following statements describes the coincidence for a Containment Spray Actuation to occur and the actions that will result in this coincidence?
A. 2/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A) in the TEST positio B. 2/3 coincidence after the channel is placed in the BYPASS condition, by placing bistable (PB-937A) in the TEST positio C. 1/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A) in the TEST posi'. io D. 1/3 coincidence after the channet is placed in the BYPASS condition, by placing bistable (PB-937A) in the TEST positio l ANSWER:
B. 2/3 coincidence after the channel is placed in the BYPASS condition, by placing bistable (PB-937A) in the TEST positio l l
RG A23 SRO #24 K/A #013000K502 OBJECTIVE #003A0212 REFERENCES: T/S 3.3.2 ACTION c, Table 3.3-3 FU 2.c ACTION 16 PRINT 7250D64 S008
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SRO Test
i QUESTION #012 ;
i Following a LOCA, hydrogen concentration in the containment has increased slowly over several days, reaching 1.0 volume per cen Which ONE of the following actions should be taken? !
A. One train of the electric hydrogen recombiner system should be placed in servic B. Electric hydrogen recombiners should be placed in service when hydrogen concentration reaches 4.0 volume per cen i C. Electric hydrogen recombiners cannot be placed in service. Heater operating temperature on the recombiner exceeds ignition temperature for hydrogen at this concentratio }
D. Both trains of electric hydrogen recombiners should be placed in service in conjunction i with a containment purg i ANSWER:
A. One train of the electric hydrogen recombiner system should be placed in servic RO #63 SRO #42 K/A #028000K501 OBJECTIVE #0110400J REFERENCES: OTN-GS-00001 E-1 i
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SRO Test f QUESTION #013
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i Which ONE (1) of the following groups ofindications has revised limits during adverse containment?
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, A. S/G wide range level, RCS subcooling, S/G pressure
i B. RCS subcooling, S/G pressure, Pressurizer level i
C. S/G pressure, Pressurizer level, S/G wide range level
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D. Pressurizer level, S/G wide range level, RCS subcooli'n g
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ANSWER:
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D. Pressurizer level, S/G wide range level,' RCS subcooling i
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l SRO Test QUESTION #014 The following plant conditions exist:
. Reactor Power at 100%.
. RCS pressure 2235 psig.
. Tavg is 584 F.
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Thermal bearing cooling water inlet temperature is 104'F.
. SealInjection flow is lost.
Which ONE (1) of the following describes a condition which would require tripping a RCP?
t. #1 sealleakoff rate 5.5 gpm B. Shaft Vibration 14 mils
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C. #2 Seal Delta P of 35 psid D. #1 Seal and Bearing Inlet temperature 239 F ANSWER:
D. #1 Seal and Bearing Inlet temperature 239 F SRO #74 K/A #000015A210 l
OBJECTIVE #003B150B
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REFERENCES: OTO-BB-00002
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SRO Test QUESTION #015 Which ONE of the following Area Radiation Monitors is required by Technical Specifications?
A. Containment Area Radiation Monitor SDRE0041 B. New Fuel Storage Area Radiation Monitor SDRE0035 C. Control Room Area Radiation Monitor SDRE0033 l
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D. Cask Handling Area Radiation Monitor SDRE0034 ANSWER: ,
B. New Fuel Storage Area Radiation Monitor SDRE003 i RO #36 j I
SRO #34 K/A #072000K302 l OBJECTIVE #0110360G l REFERENCES: T/S 3.3.3.1, Table 3.3-6 FU 2.b.(2)
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QUESTION #016 i
I Which ONE of the following class lE 125VDC Electrical System Lineups can be ;
performed to satisfy MODE 1 Technical Specification LCO? j I
A. LC NG01 to Swing Charger NK25 to bus NK04 i B. LC NG04 to Swing Charger NK26 to bus NK02 C. LC NG01 to Swing Charger NK26 m bus NK03 D. LC NG04 to Swing Charger NK25 to bus NK01 I
ANSWER:
B. LC NG04 to Swing Charger NK26 to bus NK02 SRO #36 K/A #063000K402 OBJECTIVE #0110060A REFERENCES: OTN-NK-00001 l
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SRO Test QUESTION #017 The crew implemented FR-C.1, Response to Inadequate Core Cooling.
Which one of the following combinations of core exit thermocouples and indicated temperatures would require starting RCP's, even if the normally required support conditions could not be met?
- of TC's Indicated Temp *F F *F F ANSWER: 4
) F
'I RO #27 SRO #27 K/A #017020A402 OBJECTIVE #003D250E REFERENCES: FR-C.1 Background l
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SRO Test QUESTION #018 Callaway Plant is preparing for Reactor Core Offload with Refueling Pool Level at 391 inches (2046 n. level). The polar crane operator inadvertently lins the Reactor Vessel Upper Internals out of the water and causes a Hi Hi alarm on Containment Building Area Radiation Monitor SDRE004 Which ONE of the following is a required Immediate Action?
A. Close ECV0995, Fuel Transfer Tube Isolation Valv B. Initiate a Containment Purge Isolation Signal (CPIS).
C. Transfer the Charging Pump suction to the RWST and increase flo D. Evacuate personnel from containmen ANSWER:
D. Evacuate personnel from containmen RO #94 SRO #91 K/A #000061G09 OBJECTIVE #003E05I4 REFERENCES: OTO-KE-00001 OTA-RL-RK062, Att. A l
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SRO Test QUESTION #019 FR-S.1 " Response to Nuclear Power Generation /ATWS" Step 2 requires a turbine tri Why would it be desirable to trip the turbine if a reactor trip had not been achieved?
(Choose ONE)
A. The reactor will be subcritical due to manual rod insertion before the turbine is trippe B. Tripping the turbine will conserve SG inventory and limit the pressure transient that would result from a loss of all feedwa e C. Tripping the turbine will insert negative reactivity from moderator temperature coefficient, thus assisting in reactor shutdaw D. Tripping the turbine will generate an additioaal reactor trip signal and suppress core void formation by increasing RCS pressur ANSWER:
B. Tripping the turbine will conserve SG inventory and limit the pressure transient that would result from a loss of all feedwate I RO #86 SRO #61 K/A #000029K312 OBJECTIVE #003D290C REFERENCES: T61.003D.6 LP-#29 i
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QUESTION #020 Which ONE (1) of tl.e following is the HIGHEST RCS pressure at which the Safety Injection Pumps will deliver water to the RCS?
A.1050 psig B. 1250 psig C. 1450 psig D. 1650 psig ,
ANSWER:
C. 1450 psig RO #43 SRO #38 K/A #006000K603 OBJECTIVE #0110170A REFERENCES: E-0 T61.0110.6 LP-#17
SRO Test QUESTION #021 While performing actions in E-3, " Steam Generator Tube Rupture" the Control Room Supervisor asks the Balance of Plant Operator to check intact Steam Generator narrow range levels greater than 4%. Which ONE of the following BOP responses would satisfy Callaway Plant Communication Guidelines?
A. Yes, intact Steam Generator narrow range levels are greater than 4%.
B. Yes, intact Steam Generator narrow range levels are 50% and steble.
C. Yes, intact Steam Generator narrow range levels are increasing.
D. Yes, intact Steam Generator narrow range levels are 10%.
ANSWER:
B. Yes, intact Steam Generator narrow range levels are 50% and stable.
RO #8 SRO #7 K/A #194001A105 OBJECTIVE #003A060H REFERENCES: UEND-COMMUNICATIONS-01, Page 4 of 5 l
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QUESTION #022 l
Given the following:
- The Main Turbine tripped from 95% powe All systems responded normally to the tri Which ONE (1) of the following is the expected position of the steam dump valves with Tavg at 575'F7 Full Open Modulating Full Closed l l j i
B, 9 3 ^ j l ANSWER:
' RO #57 SRO #55 K/A #041020K418 OBJECTIVE #0110200J REFERENCES: T61.0110.6 LP-#20
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SRO Test QUESTION #023
A plant startup is in progress with power indicating IE-6% on both IR channels. Which one of the following will occur ifIR channel N36 fails to 21%7 A. IR high flux reactor trip B. Manual and automatic rod stop C. PZR low pressure reactor trip is unblocked l D. PR low flux reactor trip ANSWER:
B. Manual and automatic rod stop RO #95 SRO #87 K/A #000033A202 OBJECTIVE #0110260J REFERENCES: OTO-SE-00002 l
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l SRO Test QUESTION #024 Given the following conditions:
- Pressurizer Pressure = 1710 psig
- RCS C.L. Temperature = 560'F
- Core Exit TC = 568 F Which one of the following is the correct amount of subcooling for the above conditions?
A. 38
B. 41 C. 47 D. 49 ANSWER:
l E. 41 l
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RO #39
- SRO #37 l
K/A #002000K509 OBJECTIVE #003D070S
! REFERENCES: Steam Table l
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j QUESTION #025
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i A permit required confined space entry is to be conducted at the Water Treatment Plant !
blowdown line manhole.
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) A. The attendant may enter the space if necessary, to rescue the entrant,
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B. The work supervisor must be present whenever personnel are in the confined spac C. Each entrant shall use a chest or full body hames D. The Medical Emergency Response Team will perform any emergency rescue if necessar ANSWER:
I C. Each entrant shall use a chest or full body harnes SRO #15 -
K/A #194001K113 OBJECTIVE #003A30G3 REFERENCES: APA-ZZ-00802
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l SRO Test QUESTION #026 I
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With the plant in MODE 1, AND one safety related CCP INOPERABLE, RCP Seal i Injection should be provided by the which will maintain seal cooling in 1 the event of a l l l A. Non-safety related charging pump, CCW thermal barrier leak.
l l B. Non-safety related charging pump, loss of a single electrical bus.
! C. Opposite train safety related CCP, CCW thermal barrier lea l i D. Opposite train safety related CCP, loss of a single electrical bu !
I ANSWER:
B. Non-safety related charging pump, loss of a single electrical bus.
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RO #20 SRO #22 K/A #004000K202 OBJECTIVE #003 A04A1 REFERENCES: OTN-BG-00001 l
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SRO Test =
QUESTION #027 i
Which ONE of the follow'mg describes the tagott control used for the temporary operation of equipment that is protected under a Hold oft.
A. The tags shall be cleared prior to operation then a new tagout written and new tags hun ;
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B. The tags may be lined and reused aner operation providing a briefing is held and the individual signed on the WPA is present at the component to be checke )
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C. With Shin Supervisor and Requester approval, equipment may be operated without clearing the tags, if the requester is in the equipment area and operation completed in the same shin.
D. The tags which must be cleared to allow for the operation can be temporarily cleared, replaced with Caution Tags until the operation is complete, then the Caution Tags replaced with new Hold Off Tags.
ANSWER:
B. The tags may be lined and reused after operation providing a briefing is held and the individual signed on the WPA is present at the component to be checked.
RO #2
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SRO #2 K/A #194001K102 OBJECTIVE #003 A330L REFERENCES: ODP-ZZ-00310 Page 10
SRO Test QUESTION #028 During operations at 95% power and pressurizer level at 48%, the Tave input to the pressurizer level controller fails low. What INDICATIONS does the operator have that the Tave input failed low?
A. Backup heaters are energized, charging flow control valve slowly closes, high level deviation alarm actuates.
B. Backup heaters are deenergized, charging flow control valve slowly opens, low level deviation alarm actuates.
C. Backup heaters are energized, charging flow control valve slowly opens, low level deviation t.larm actuates.
D. Backup heaters are deenergized, charging flow control valve slowly closes, high level deviation alarm actuates.
ANSWER:
A. Backup heaters are energized charging flow control valve slowly closes, high level
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deviation alarm actuate j RO #40 l SRO #39 ;
K/A #011000A203 OBJECTIVE #0110090C REFERENCES: OTO-BB-00004
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SRO Test QUESTION #029 Plant condit'ons:
. Operating in MODE 1, at 100% powe . SJ-RE-01, CVCS Letdown Monitor, Alarming Hi/Hi
. SD-RE-20, AB 2000 Area, Alarming Hi/Hi Which ONE of the following operator actions is required per OTO-BB-00005, RCS High Activity?
A. Reduce power B. Isolateletdown C. Increase letdown to 120 gpm D. Initiate hourly sampling of the RCS ANSWER:
C. Increase letdown to 120 gpm RO #76 SRO #73 K/A #000076G008 OBJECTIVE #003B180A REFERENCES: OTO-BB-00005 l
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SRO Test QUESTION #030
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Given the following conditions:
. Tavgis 576 F e Pressurizer Pressure is 2240 psig Charging Flow is being controlled in MANUAL
. The BACKUP HEATERS havejust ENERGIZED Which ONE of the following is the ~ actual pressurizer level?
A. 37%
B. 42%
C. 47%
D. 52%
ANSWER:
D. 52%
RO #6 SRO #98 K/A #000028A201 OBJECTIVE #0110300K REFERENCES: T61.0110.6 LP-#30 Y
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SRO Test
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QUESTION #031
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A Ruptured Steam Generator has been cooled down and depressurized. ECCS pumps have been secured and Normal Charging and Letdown have been established, .
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Plant Conditions: ;
! * PZR Level 30% and DECREASING e Ruptured S/G NR LevelINCREASING Which ONE of the following is required to balance inventory?
A. Depressurize the RCS !
j B. Increase RCS Makeup Flow j i
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! D. Decrease RCS Makeup Flow l ANSWER- !
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l A. Depressurize the RCS l
i RO #85 l SRO #90 L K/A #000038K306 OBJECTIVE #003D17JJ REFERENCES: T61.003D.6 LP-#17 E-3, SGTR i
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l QUESTION #032 Which of the following is NOT an event the MSiVs are used to protect against?
A. Steam Line Break inside Containment B. Feedwater Line Break upstream of check valve C. Steam Line Break outside Containment D. Steam Generator Tube Rupture ANSWER:
B. Feedwater Line Break upstream of check valve SRO #52 K/A #035010K601 OBJECTIVE #0110200A REFERENCES: T61.0110.6 LP-#20 i
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SRO Test QUESTION #b33 With the plant in MODE 1 the Shift Supervisor is notified by Security that a confirmed penetration has occurred by unauthorized personnel into the NB01 switchgear room. The Plant Emergency Alarm is sounded and a CODE RED is announced over the Gai-tronics.
Which ONE of the below may be performed during the initial response by Control Room personnel?
A. Evacuate all unnecessary personnel, shut the Control Room Missile Door, and notify the NRC of 10CFR50.54(x) implementation within ONE hour.
B. Trip the Reactor, commence RCS cooldown at the Technical Specification limit, and declare an Unusual Event.
C. Shut the Control Room Missile Door, have all Equipment Operators report to the Field Office, and declare an ALERT.
D. Declare an ALERT, trip the Reactor, and notify the NRC of 10CFR50.54(x)
implementation within ONE hour.
ANSWER:
D. Declare an ALERT, trip the Reactor, and notify the NRC of 10CFR50.54(x)
implementation within ONE hour.
SRO #13 K/A #194001 Al16 OBJECTIVE #003B280B REFERENCES: EIP-ZZ-00102, At OTO-SK-00001
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SRO Test QUESTION #034 A normal plant heatup is in progress pen OTG-ZZ-00001 with the following plant conditions:
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- RCS pressure 1835 psig
- RCS pressurization rate 15 psig/ min j l - RCS temperature 485'F l
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l - RCS heat up rate 10 F/hr
- S/G pressure 575 psig If the current trend continues, which ONE of the following occur FIRST7 A. Main Steam Isolation Valves clos B. Pressurizer PORV's ope C. Low Pressurizer Pressure Safety Injectio D. First group of steam dumps throttle ope ANSWER:
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A. Main Steam Isolation Valves clos RO #21 i SRO #25 K/A #013000K403 OBJECTIVE #0110520B REFERENCES: OTG-ZZ-00001," Plant Heatup Cold Shutdown to Hot Standby" Page 25 i
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QUESTION #035 !
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Which ONE of the following conditions satisfies the Technical Specification 3.5.5, l
" Refueling Water Storage Tank", requirement for an operable RWST in MODE 17 l
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Borated Water Volume Boron Concentration Solution Temperature - ,000 gallons 2400 ppm 80* ;
I ,000 gallons 2000 ppm 95'F j
- ,000 gallons 2400 ppm 40 F l
! ,000 gallons 2500 ppm 105 F i i
ANSWER: j
, ,000 gallons 2400 ppm 40 F
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SRO Test l QUESTION #036 l
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A surveillance to be performed on a piece of equipment having a contact reading of ( 50 R/hr in a room with a general area radiation reading of 125 mR/hr, would require entry intoa:
A. Danger High Radiation Area B. Caution High Radiation Area C Danger High Radiation Area Radiological Exclusion Area D. Very High Radiat>. :n Are ANSWER:
B. Caution High Radiation Area ,
RO #3 SRO #3 K/A #19400lK103 OBJECTIVE #003A3IF3 REFERENCES: APA-ZZ-01000 Page 6 l
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SRO Test QUESTION #037 Technical Specification 3/4.2.4 " Quadrant Power Tilt Ratio" (QPTR) lists required actions that must be accomplished if QPTR exceeds specified limits for more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Which of the following is the basis for the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time limit?
A. To allow time for identification and correction of a dropped / misaligned control ro B. To allow time for identification and correction of a malfunctioning power range instrumen C. To allow time for testing, identification and correction of power cabinet multiplexing circuit D. To allow sufIicient time for control rod response time testing of the malfunctioning solid state protection circuit ANSWER:
A. To allow time for identification and correction of a dropped / misaligned control ro SRO #76 K/A #000003G07 OBJECTIVE #003AA3A2 REFERENCES: TS 3/4.2.4 Bases l
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SRO Test QUESTION #038 Which ONE of the following is the basis for the 's echnical Specification limit on total steam generator tube leakage of 600 gpd for all steam generators?
A. A limited amount ofleakage is expected and this threshold value is sufficiently low to ensure early detection of additional leakag B. To ensure that the dosage contribution from the tube leakage will be acceptable in the event of either a steam generator tube rupture or steam line brea C. This is a known source which can be readily detected by radiation monitors on steam generator blowdown so it will not interfere with detection ofleakage from other source D. To ensure that the steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA condition ANSWER:
i B. To ensure that the dosage contribution from the tube leakage will be acceptable in the event of either a steam generator tube rupture or steam line brea SRO #93 K/A #000037G05 OBJECTIVE #003 AA213 REFERENCES: TS 3/4.4.6.2 Bases
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SRO Test QUESTION #039 During a loss of all AC while performing ECA-0.0, Loss of All A.C. NK11 battery ;
discharge amps is at 300 amp J Which ONE of the following is the MAXIMUM time that NK01 could be predicted to be j Operable assuming the battery was fully charged initially?
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A. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> C. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
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D. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> I ANSWER:
B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> RO #67 SRO #65 K/A #000055K101 OBJECTIVE #003D220V REFERENCES: E21NK01 i
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QUESTION #040 I
i A Reactor Trip has just occurred. The following conditions are found while performing Step 3 ofE-0, Reactor Trip or Safety Injection:
. NB01 energized from Emergency Diesel NE-01 I
. NB02 deenergized (no lockout)
Which ONE of the following descri'oes the required action and basis for that action?
A. Transition to ECA-0.0, Loss of all AC Power because E-0 assumes that Offsite Power is Availabl B. Attempt to restore power to NB02 while continuing with E-0 because it is desirable to l have power to all AC Emergency buse j C. Attempt to restore Off Site Power to BOTH NB buses because E-0 assumes that Off Site Power is Availabl D. Do no rnake attempts to restore NB02 because it will delay the operator action and I
only one NB bus is assumed energized by E- !
ANSWER:
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B. Attempt to restore power to NB02 while continuing with E-0 because it is desirable to have power to all AC Emergency buse RO #99 SRO #99 K/A #000056K302 OBJECTIVE #003D040E REFERENCES: T61.003D.6 LP-#4
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QUESTION #041 )
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i A periodic load test is being performed on NE02, Standby Diesel Generator 'B' in accordance with OSP-NE-0001B. NE02 has been paralleled with 4160V Bus NB02 and ,
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is carrying 6 MW of real load. A Main Steamline break occurs and containment pressure increases to 20 (twenty) psi I Which ONE of the following describes the response of the Load Shedding Emergency Load Sequencing System (LSELS)?
A. The LOCA Sequencer starts the Containment Spray Pumps at Step 3 (Time 15 seconds). l B. The Shutdown Sequencer stans the 'A' Essential Service Water Pump at Step 5 (Time 25 seconds).
C. The LOCA Sequencer starts the Safety Injection Pumps at Step 1 (Time 5 seconds). I i
D. The Shutdown Sequencer starts the Residual Heat Removal Pumps at Step 2 ;
(Time 10 seconds).
l ANSWER:
! C. The LOCA Sequencer starts the Safety Injection Pumps at Step 1 (Time 5 seconds).
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I QUESTION #042 !
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WHICH of the following red paths is MOST LIKELY to occur for a steam line break on a single S/G outside containment, resulting in a reactor trip and Sl? (Assume that all safeguards equipment functions as designed.)
A. Response to Inadequate Core Cooling (FR-C.1)
B. Response to Loss of Secondary Heat Sink (FR-H.1)
l C. Response to Imminent Pressurized Thermal Shock Condition (FR-P.1) )
D. Response to High Containment Pressure (FR-Z.1)
ANSWER:
C. Response to Imminent Pressurized Thermal Shock Condition (FR-P.1)
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RO #70 SRO #63 K/A #000040K101 OBJECTIVE #003D280A REFERENCES: T61.003D.6 l
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SRO Test QUESTION #043 A plant cooldown is initiated following a reactor trip using the AUX FEED system and S/G PORV's. The CST level is initially at 87% (407,000 gal).
Which ONE of the following is the time available until CST level decreases to the MODE 3 Technical Specification limit with AUX feed flow at 300,000 lbm/hr. (8.345 lbm/ gal)
A. 3.5 hr.
B. 4,0 hr, C. 4.5 hr.
D. 5,0 hr.
ANSWER:
A. 3.5 hr.
RO #34 SRO #31 K/A #061000A104 OBJECTIVE #0110250E REFERENCES: T/S 3.7. Tank Book TDB-001
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SRO Test QUESTION #044'
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Which ONE of the following events is required to be recorded in the RO Narrative Logs? l I
A. Chemical addition to the condensate syste l B. Security Event due to Security System (SAS) malfunction.
C. Annunciator switchyard carrier potential / tone loss, alarms.
D. Unexpected ESFAS alarm on ESW system.
ANSWER:
D. Unexpected ESFAS alarm on ESW system.
RO #9 SRO #8 K/A #194001 A106 OBJECTIVE #003 A02B1 REFERENCES: ODP-ZZ-00006, Section l
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m SRO Test QUESTION #045 Given the following information:
. Train A Emergency Diesel Generator became inoperable one hour ag . 92% Power Operatio Which ONE (1) of the following statements describes the operability of the other A train equipment?
A. All systems, equipment, components, or devices which normally receive emergency power from the train A Emergency Diesel Generator are also inoperabl B. All systems, equipment, components, or devices which normally receive emergency power from the train A Emergency Diesel Generator are also inoperable, except those which are powered by an operable batter C. The operability of the remaining train A equipment is not impacted, but the train B equipment and the TDAFP are required to be verified operable per Technical Specification 3.8. D. The operability of the remaining train A equipment is not impacted, except for the ESF electrical bus that the Emergency Diesel Generator support ANSWER:
C. The operability of the remaining train A equiptr nt is not impacted, but the train B
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equipment and the TDAFP are required to be verified operable per Technical Specification 3.8. '
SRO #53 K/A #062000G008 OBJECTIVE #0110060G REFERENCES: TS 3.8. TSI 48
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SRO Test QUESTION #046 !
Which ONE of the following is the preferred method ofinjecting highly borated water into the RCS during an ATWS7 A. Manually align Charging Pump suction to the RWST.
B. Borate through BGV0177, Alternate Immediate Boration Valv i C. Manually initiate a Safety Injection from RL00 I D. Borate through BG-HV-8104, Emergency Borate to Charging Pumps Suction Valve.
ANSWER:
D. Borate through BG-HV-8104, Emergency Borate to Charging Pumps Suction Valve.
SRO #80 K/A #000029G11 l OBJECTIVE #003D290B !
REFERENCES: FR- ;
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SRO Test i
l QUESTION #047 -
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, The plant has experienced a large break RCS loss of coolant accident.
i j Which ONE of the following must be reset to allow opening KAHV0029, Instrument Air
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Ctmt Isolation?
A. CISA l
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B. CISB
{ C. SIS D. FBVIS
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A. CISA i-1-
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- RO #24
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SRO #23
!. K/A #013000A201 ';
j OBJECTIVE #003B480A
- REFERENCES: E-0, Reactor Trip / Safety Injection i M22KA01 e
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QUESTION #048 (
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Callaway Plant is in MODE 1,30% Reactor Power on a Chemistry hol l
Annunciator 70B, "RCP VIB/SYS ALERT" alarms. The Reactor Operator checks j
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vibrations on RP312 and finds 'C' RCP shaft vibration indicating 15 mils and stead l Which one of the following is the required actio l l
A. Trip the Reactor, Trip 'C' RCP and go to E-0, Reactor Trip or S !
l B. Continue to monitor vibration on the 'C' RC C. Trip the 'C' RCP and declare the Loop 3 RTD channel inoperabl D. Increase Component Cooling Water temperature to reduce 'C' RCP vibratio ANSWER:
. B. Continue to monitor vibration on the 'C' RC RO #18 SRO #20 K/A #003000G10 OBJECTIVE #003B150B REFERENCES: OTO-BB-00002 l
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SRO Test
l QUESTION #049 )
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l Which one of the following areas does NOT have restricted access as part ofRCS l l
Reduced Inventory Controls?
' A. Electrical Penetration Rooms on the AB 2026'. I l
. B. Switchyard :
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C. Around the MA cabinets on TB 2033' level D. NB Switchgear Rooms e
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A. Electrical Penetration Rooms on the AB 2026'
SRO #14 K/A #19400lK105 OBJECTIVE #003EE20B >
REFERENCES: OTN-BB-00002, Attachment 10 ;
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SRO Test QUESTION #050 In FR-H.5, Response to Steam Generator Low Level, AFW flowrate is procedurally restricted to 50,000 lbm/hr when recovering a steam generator level if the level has fallen below 24% wide range indication?
Which ONE of the following indicates why?
A. Minimize thermal stress conditions on steam generator component B. Minimize RCS cooldown rate and prevent resultant thermal stress on RCS component C. Ensure RCS inventory demand does not exceed normal charging pump capacit D. Ensure pressuiizer level transient does not result in pressure transient that would actuate S ANSWER:
' A. Minimize therrnal stress conditions on steam generator component SRO #94 K/A #000054K102 OBJECTIVE #003D260S REFERENCES: T61.003D.6, LP-#26
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' SRO Test I i
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QUESTION #051
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A 30 gpm leak has developed on the charging line between BG-HCV-182 (CVCS CHG PMPS TO REGEN HX HCV) and the regenerative heat exchanger. When the Control Room isolates the leak and completes the applicable Off-Normal procedures, the reactor makeup flowpath will be via . and the reactor letdown flowpath will be via'
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Choose ONE of the following to fill in the blank A. alternate charging; normalletdown B. alternate charging; excess letdown C. sealinjection; excessletdown
' D.'sealinjection; normalletdown ANSWER:
C. sealinjection; excessletdown RO #90 ,
SRO #84 K/A #000022A101 OBJECTfVE #003B220B REFERENCES: OTO-BG-00002
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OTO-BB-00003 I
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SRO Test QUESTION #052 t
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l Given the following conditions: ;
- RCS at NOP/NOT for 100% RTP,
- PORV 456A has seat leakage to the PRT,
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. Which ONE of the following is the approximate tailpipe temperature?
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I i A. 212*F B. 228'F '
C. 248'F D. 258'F ANSWER: i D. 258'F ,
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. K/A #007000A201 1 OBJECTIVE #0070130B i REFERENCES: Steam Table
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SRO Test
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- QUESTION #053 i
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l The Callaway Plant is in MODE 3 at NOP and NOT. An earthquake ruptures the l Condensate Storage Tank and causes a steam break on 'C' S/G. The following conditions
exist:
} SG A, B & D NR Level 45 %
! SG C NR Level 10%
) SG A, B, & D Press 900 psig i SG C Press 300 psig AFW Suction Press 4 psig i Which one of the following describes the resulting flowpath of feedwater to the Steam
,- Generators?
j A. 'B' ESW Pump to 'B' MDAFP to 'C' S/G
1 B 'A' ESW Pump to 'A' MDAFP to 'P' S/G l
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C. 'B' ESW Pump to 'B' MDAFP to 'B' S/G J-j D. 'A' ESW Pump to 'A' MDAFP to 'D' S/G
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ANSWER:
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j B. 'A' ESW Pump to 'A' MDAFP to 'B' S/G
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RO #35 i
SRO #32 K/A #061000A303 OBJECTIVE #0110250D i REFERENCES: T61.011 OTA-RL-RK127A
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, SRO Test QUESTION #054
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J j With the plant at 40% power which one of the below would be TRUE regarding operation of the ATWS Mitigation Actuation Circuitry (AMSAC)?
A. If S/G Levels decrease to less than 5% on 2 of 3 AMSAC logic circuits, then a Turbine Trip and MD AFAS, are actuated 25 seconds late B. If S/G Levels decrease to less than 5% on 1 of 2 AMSAC logic circuits, then a Turbine Trip and MD AFAS, are actuated 232 seconds late C. If S/G Levels decrease to less than 14.8% on 2 of 3 AMSAC logic circuits, then a Turbine Trip and MD AFAS, are actuated 25 seconds late D. If S/G Levels decrease to less than 14.8% on 1 of 2 AMSAC logic circuits, then a Turbine Trip and MD AFAS, are actuated 232 seconds late ANSWER:
A. If S/G Levels decrease to less than 5% on 2 of 3 AMSAC logic circuits, then a Turbine Trip and MD AFAS, are actuated 25 seconds late RO #22 SRO #30 K/A #001000GK04 OBJECTIVE #0110540B REFERENCES: OTA-RL-0083A E23ACll l
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-...I SRO Test QUESTION #055 Liquid Radwaste Discharge Monitor (HDRE18) alarms on the RM-11 in dark blue condition.
Which ONE of the below could be the cause?
A. Loss of Sample Flow B. Loss of Process Flow C. Monitor Purging D. ChannelNo Pulses Received ANSWER:
D. Channel No Pulses Received I
RO #97 SRO #68 K/A #000059A201 OBJECTIVE #0110360B REFERENCES: OTN-SP-00002 OTA-SP-RM011 l
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SRO Test !
QUESTION #056 f
Following a safety injection due to a RCS leak in containment, plant conditions are established that meet the SI termination criteria ofE-1, Loss ofReactor or Secondary Coolan I Which ONE of the below is true regarding these plant conditions?
A. All safety related equipment is Operable as required by Technical Specification i B. Reactor core decay heat is being removed by the steam generator I C. Containment pressure is below the safety injection actuation setpoin D. Steam Generator pressure are approximately equal to RCS pressur ANSWER:
B. Reactor core decay heat is being removed by the steam generator l RO #82 SRO #83 I K/A #000009K324 OBJECTIVE #003D090J REFERENCES: ES-1.1 SI Termination l
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SRO Test
' QUESTION #057 l
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i Which ONE of the following valves fail open on a loss ofinstrument air?
f A. Steam Generator Atmospheric Relief !
I B. Main Feed Regulating Bypass Valves (
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C. Main Feed Pump Recirc Valve l
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D. Heater Drain Pump Recirc Valve l ANSWER: !
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D. Heater Drain Pump Recirc Valve RO #64 i
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SRO #56 i l K/A #078000K302 l OBJECTIVE #003B330A l REFERENCES: OTO-KA-00001 '
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SRO Test I i
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QUESTION #058 l
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l An automatic preaction sprinkler syste, " trouble" alarm would indicate: .l
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A. a deluge valve actuation l B. an alarm check valve operation
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C. a fire detector in alarm condition D. an open sprinkler head ANSWER:
l D. an open sprinkler head RO #47 SRO #50 K/A #086000A402 OBJECTIVE #0110350C REFERENCES: T61.0110.6 LP-#35 Callaway Bank l
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SRO Test !
QUESTlON //059 i
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i Given the following conditions:
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- Containment pressure is at 10 psig and increasing at I psig/ minut * Normal Feeder breaker NB0209 was inadvertently opened causing a loss of power on l
' ESF bus NB02, t
4 ESF bus NB01 has' remained energized from Normal Feeder NB0112
- The original SI signal has not been rese '
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. AT THE SAME TIME that breaker NB0211 closed m, reenergizing bus NB02 from NE02 diesel generator, a containment spray (CS) actuation signal was generate j i
Assuming all interlocks are met, WHICH ONE of the following combinations states the
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l times at which the CS pumps will start? i A CS Pump B CS Pump
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. Immediately Immediately l j' Immediately 15 seconds
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L l seconds 15 seconds seconds 40 seconds l-ANSWER: In1rnediately 15 seconds l SRO #35 K/A #026000A301
.OBJEC11VE #0110510F REFERENCES: E22NF01
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I SRO Test QUESTION #060 Which ONE of the following should be performed by any individual discovering a fare?
A. Notify Control Room, then use any available fire fighting equipment, then report to Fire Brigade Leade . B. First attempt extinguishment using closest available extinguisher, then call Control Room ifunsuccessfu C. First atterm extinguishment using closest available extinguisher then report to Fire Brigade S: ng Are D Notify Control Room, then use closest available extinguisher, if practical, then report l to Fire Brigade Leade ANSWER:
! D Notify Control Room, then use closest available extinguisher, if practical, then report
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to Fire Brigade Leade RO #5 l SRO #5 i K/A #19400lK116 ;
l OBJECTIVE #003A30F3 i
REFERENCES: EIP-ZZ-00226, Att. 2 l
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SRO Test QUESTION #061 Which ONE of the below shows the correct speed settings for the TD AFW pump?
IDLE SPEED NORMAL OPERATING SPEED OVERSPEED rpm 3850 rpm 4235 rpm rpm 3550 rpm 4435 rpm rpm 3850 rpm 4235 rpm rpm 3550 rpm 4435 rpm ANSWER: rpm 3850 rpm 4235 rpm RO #38 SRO #45 K/A #039000A404 OBJECTIVE #0110250C REFERENCES: OSP-AL-P0002 i
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SRO Test QUESTION #062 i
The plant is in MODE 3 at Normal operating pressure and temper:iure, Train ' A' COPS has inadvertently been left ARMED for Cr'd Overpressure Ps;tection.
The selected pressurizer pressure channel, BBPT455 subsequently fails high.
With no operator actions, which ONE of the following is TRUE?
A. PORV 455 initially opens, then closes when actual PZR Pressure decreases to <2185 psig.
B. PORV 455 stays closed initially but will function as required for COPS.
C. PORV 455 initially opens and stays open when actual PZR pressure decreases to
<2185 psi :
D. PORV 455 stays closed initially and PORV BLOCK VALVE (8000A) closes when actual PZR pressure decreases to <2185 psig.
ANSWER: i i
i A. PORV 455 initially opens, then closes when actual PZR Pressure decreases to <2185 '
psig.
RO #74 SRO #86 K/A #000027A101 OBJECTIVE #003B190A REFERENCES: 7250D64 Sheet 17
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. SRO Test QUESTION #063 i
Which ONE of the following is the reason for depressurizing the Steam Generators at the ,
maximum rate during ECA-0.0, " Loss of All AC Power"? '
- A. To allow feeding S/G's from Die el Driven Fire Water Pum B. To minimize RCS inventory los C. To enhance restoration of SG level from TD AFW Pum D. To prevent lif).ing PZR PORV ANSWER:
B. To minimize RCS inventory los RO #68 SRO #66 K/A #000055K302 OBJECTIVE #003D220S REFERENCES: T61.003 ._ _ _ . . S R O T cst QUESTION #064 Given the following:
- Callaway is operating at 30% steady state reactor powe I&C technician receives permission to perform a calibration on Power Range Channel N-4 The I&C technician mistakenly pulls the control power fuses on N-42; then, realizing his mistake, he reinserts the fuses for N-42 and pulls the control power fuses for the correct channel, N-41, causing a reactor tri '
Which ONE (1) of the following describes the reason for the reactor trip?
A. PR neutron flux low setpoint tri j I
B. Overpower Delta T tri C. PR neutron flux high setpoint tri D. PR positive rate tri ANSWER:
D. PR positive rate tri RO #53 SRO #41 K/A #012000K603 OBJECTIVE #0110270D REFERENCES: T61.n110.6 LP-#27 T61.0110.6 LP-#28
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SRO Test
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QUESTION #065 ,
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Which ONE of the below conditions would require containment coolers to be operated in j SLOW speed?
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A. Service Water Temperature <60'F :
I B. ESW Supplying Contamment -
C. Emergency Diesel Supplying NB Bus D. Containment Temperature <80 F ANSWER:
A. Service Water Temperature <60 F RO #28 SRO #29 K/A #022000A101 OBJECTIVE #003A2001 REFERENCES: OTN-GN-00001 i
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1 SRO Test l
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QUESTION #066 j
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i Prior to opening the Reactor Trip Breakers during a plc.it shutdown, the crew is directed to reduce the inservice MFP speed to 3650 RPM in anticipation of a Feedwater Isolation i Signa I Using the attached graph, determine which ons of the following is the minimum flowrate l required to provide pump protection for this spee j
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! A.1500 Kibm/hr B, 1750 Klbm/hr C. 2000 Klbm/hr
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D. 2250 Klbm/hr l ANSWER: ;
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l C. 2000 Klbm/hr l
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- K/A #194001 A108 OBJECTIVE #003A040E REFERENCES
- OTN-AE-00001, Att. 4.
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MAIN FEED PUMP MINIMUM FLOW (LBMlHR VS. RPM)--MINIMUM FLOW AT DESIGN
~; SPEED OF 5000 RPM IS 6000 GPM OR APPROX. 2800 K LBMlll ,
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5000 <
+ b i O 4000
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m W h3000 - 'l C a :
C-o a ?
w / a-W 2000 4 -
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k / -
i E 1000
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0 570 1140 1710 2280 2850 MAIN FEED PUMP MINIMUM FLOW (K LBMlllR) FOR GIVEN SPEED
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SRO Test i
I QUESTION #067 '
The plant experienced a Primary LOCA due to an earthquake. Both the CCW and ESW l systems are Inoperable. All CCP and SI pumps are in operation in response to the Safety l Injectio Which ONE of the following describes the operation of the CCP and SI pumps?
i l A. Continued operation of all CCP and SI pumps is acceptable.
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B. Secure all CCP and SI pumps until CCW or ESW is restore '
C. Alternate CCP and Si pumps so that only ONE train is injecting.
l D. Operate CCPs only while securing the SI pumps.
l ANSWER: '
l-C. Alternate CCP and SI pumps so that only ONE train is injecting.
l SRO #78 l K/A #000011G007 *
OBJECTIVE #003A0100 l REFERENCES: OTN-BG-00001, 2.21 OTN-EM-00001
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SRO Test
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QUESTION #068
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l WHICH of the following groups of parameters read out at the Auxiliary Shutdown Panel? l A. RCS WR pressure, S/G pressure, S/G level, containment pressure
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B. RCS Tavg, S/G pressure, S/G level, containment pressure l
C. RCS hot leg temp, S/G level, TDAFWP flow, containment pressure l
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D. RCS cold leg temp, RCS hot leg temp, S/G level, S/G pressure ANSWER:
D. RCS cold leg temp, RCS hot leg temp, S/G level, S/G pressure RO #72 SRO #70 K/A #000068K201 OBJECTIVE #0110480B !
REFERENCES: T61.011 !
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SRO Test QUESTION #069
The signal from the 'A' train SSPS to cause a reactor trip will:
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I A. open the 'A' reactor trip breaker and the 'A' reactor trip bypass breake !
B. open the 'B' reactor trip breaker and the 'B' reactor trip bypass breake C. open the 'A' reactor trip breaker and the 'B' react' or trip bypass breake D. open the 'B' reactor trip breaker and the . . reactor trip bypass breake '
ANSWER:
C. open the 'A' reactor trip breaker and the 'B' reactor trip bypass breake RO #54 SRO #40
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K/A #012000A403
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i OBJECTIVE #0110270C '
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REFERENCES: T61.0110.6 LP-#27 Callaway Bank
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SRO Test l i
QUESTION //070 !
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During a refueling outage welding is being performed in a high radiation area. No fire l watch will be used due to ALARA consideration ,
Which ONE of the below would approve the hot work permit as the designated I management representative?
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I A. Shift Supervisor i
B. Outage Shift Manager C. Maintenance Work Supervisor D. Health Physics Supervisor ANSWER:
A. Shift Supervisor SRO #17 K/A #194001 All6 OBJECTIVE #003A30A4 REFERENCES: APA-ZZ-00010, 4.3. APA-ZZ-00742, 3. _ _
SRO Test l
I QUESTION #071 ;
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A spurious SI causes a plant trip and St. Which one of tie below actions is acceptable to !
be performed while performing E-0 steps I through 147 ;
A. Securing NE01 due to ESW pump A tripping.
B. Securing RHR Train 'A' due to RCS pressure at 223 l l
C. Stopping one CCP to minimize injection to RCS.
D. Starting a SFP pump to restore Fuel Pool Cooling.
ANSWER:
A. Securing NE01 due to ESW pump A tripping.
RO #6 SRO #6 K/A #194001A102 OBJECTIVE #003A29C4 REFERENCES: ODP-ZZ-00025 l
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SRO Test l
i QUESTION #072 l l
Both trains of Essential Service Water (ESW) are placed into service to reduce containment temperature. Shortly after placing ESW into service, reactor power is noted to be slowly increasin Which ONE of the following is the probable cause of the power increase?
A. Change in containment air temperature affecting operation of the power range detector B. Change in main feedwater temperature due to flow variations in the S/G Blowdown i
syste C. Change in the CVCS letdown temperature causing deboration in the letdown demineralizer D. Change in main condenser vacuum causing increasing main steam flow through the main turbin ANSWER:
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C. Change in the CVCS letdown temperature causing deboration in the letdown l demineralizer I
RO #42 SRO #48 K/A #075000A401 OBJECTIVE #003 A09Al REFERENCES: OTN-EF-00001 OTN-EG-00001
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SRO Test
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4 ' QUESTION #073 ,
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Given the following plant conditions: l 4 i
! 4 i e SAFETY INJECTION ACTUATED l
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l e PZR PRESSURE 1800 PSIG Slowly Decreasing
- e RCS TEMPERATURE 550'F Slowly Decreasing
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- S/G NR LEVELS ' 1% Slowly Increasing
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. PRT Pressure ' 3 psig Stable ;
e S/G PRESSURE 1000 PSIG STABLE j j e PZR Level 28% INCREASING
! e RM-11 GTRE31 & 32 Alarming )
j e CTMT Temperature 140*F Slowly Increasing j l
- CTMT Pressure 8 psig
- e CTMT Humidity Increasing i
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- Which ONE of the following could be the cause of the above conditions?
A. ' Steam Generator Safety Valve failed open.
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,. B. Pressurizer PORV failed open.
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C. RCS Leak from a cold le D. Pressurizer steam space lea ANSWER:
D. Pressurizer steam space lea l RO #81 SRO #82 K/A #000008A106 OBJECTIVE #003D030F REFERENCES: E-0 Reactor Trip / Safety Injection l
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SRO Test QUESTION #074 i
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The Callaway Plant is operating at 94% power with all four containment cooling fans '
running in fast spee A simultaneous Safety Injection and loss of the normal power supply to NB01 occurs. All systems function as designe l Which one of the following describes the response of the Containment Cooling fans?
A. Fans A and C start in FAST speed, B & D continue to run in FAST spee B. Fans A.& C start in SLOW speed, fans B & D shift to SLOW spee C. Fans A & C start in FAST speed, fans B & D shin to SLOW spee D. Fans A & C start in SLOW speed, fans B & D continue to run in FAST spee ANSWER: l
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B. Fans A & C start in SLOW speed, fans B & D shin to SLOW spee RO #29 SRO #28 K/A #022000A301 OBJECTIVE #0110400D REFERENCES: E21005 E21001 l
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SRO Test QUESTION #075
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i Which one of the following describes the operation of 7.5 KVA Inverter NN12 when the 125VDC supply from NK0211 is intermpted?
A. The Static Transfer switch will AUTOMATICALLY transfer to the Bypass !
Transformer and will AUTOMATICALLY transfer back to the invener when '
125VDC is restore B. The Static Transfer switch will AUTOMATICALLY transfer to the Bypass I Transformer, but must be MANUALLY transferred back to the inverter when I 125VDC is restore l C. The Static Transfer switch must be MANUALLY transferred to the Bypass Transformer, but will AUTOMATICALLY transfer back to the inverter when I 125VDC is restore D. The Static Transfer switch must be MANUALLY transferred to the Bypass Transformer and MANUALLY transferred back to the inverter when 125VDC is restore l l
ANSWER: l B. The Static Transfer switch will AUTOMATICALLY transfer to the Bypass Transformer, but must be MANUALLY transferred back to the invener when 125VDC is restore ,
l SRO #81 K/A #000057A101 OBJECTIVE #0110060E REFERENCES: OTN-NN-00001 l
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SRO Test QUESTION #076 The plant is in the injection phase of Safety Injection due to a RCS LOCA. Containment Pressure has reached a maximum of 25 psi Which ONE of the following indicates ONLY loads being cooled by CCW?
A. RHR Pumps, RHR Heat Exchangers, Sample systems B. Fuel Pool, Reactor Coolant Pumps, Excess Letdown Heat Exchangers C. Containment Spray Pumps, Charging Pumps, Reactor Coolant Pumps D. Reactor Coolant Pumps, Charging Pumps, RHR Pumps ,
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D. Reactor Coolant Pumps, Charging Pumps, RHR Pumps RO #77 SRO #60 K/A #000026K302 OBJECTIVE #011010CC REFERENCES: M22EG01 !
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j SRO Test
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- QUESTION #07 I t
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The Callaway Plant is in a Reduced inventory condition and has suffered a Loss of RHR ,
Cooling.
l Which ONE of the following would cause a reduction in T-Boil (Time to Boil)?
A. Fewer Effective Full Power Days (EFPD)
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- B. Longer Time since Shutdown .
. C Lower Steam Generator Level l
D. Lower RCS Loop Level ANSWER:
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D. Lower RCS Loop Level
.i RO #91
- SRO #85 K/A #000025G10
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OBJECTIVE #003EE20B
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REFERENCES: OTN-BB-00002 T-Boil Calc-Theory '
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SRO Test QUESTION #078 The plant is in MODE 6, performing CRDM drag testing when Source Range Channel N-31 fail CRDM drag testing may continue:
A. For 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with only Source Range Channel N-32 Operable.-
B. Using Gamma . Metrics Flux Monitor and Source Range N-3 C. After determining the Reactor Coolant System boron concentratio D. Only for those CRDMs that are adjacent to Source Range N-3 ANSWER:
B. Using Gamma Metrics Flux Monitor and Source Range N-3 SRO #100 K/A #000036K101 OBJECTIVE #003E040A REFERENCES: TS 3.9.2 Int. #42
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SRO Test QUESTION #079 Plant startup is in progress with main turbine roll commencing and reactor power at 6%.
Power range N-44 is out of service due to a failed detector.
Which one of the below is UNBLOCKED under these conditions?
A. Intermediate Range High Flux Reactor Trip B. Pressurizer Low Pressure Reactor Trip C. Reactor Trip from Turbine Trip D. Pressurizer High Level Reactor Trip.
ANSWER:
A. Intermediate Range High Flux Reactor Trip RO #25 SRO #26 '
K/A #015000A303 OBJECTIVE #003A24A2 REFERENCES: OTG-ZZ-00003 OTO-SA-00001 I
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SRO Test .
QUESTION #080 L
l Use the attached Figure 7-5 to answer the following questio The plant is in MODE 3, 557 F,2235 psig. Which one of the following is the amount of j water needed to reduce the RCS boron concentration from 1150 ppm to 1100 ppm?
i l A. I167 ga B. I195 gal.
C. 2688 ga 'D. 2752 gal.
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! D. 2752 ga RO #14 SRO #19 K/A #001010K521
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OBJECTIVE #003AA40E REFERENCES: Plant Curve Book i
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i FIGURE 7-5 Rev. 001 I J
REACTOR MAKEUP CONTROL SYSTEF. NOMOCRAPHS l
BORON DILUIION
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M at 0% power = 515676 ' Ibm * .superin' ent. Engineering
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M at 100% power = 503624 lbm *
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ion g target ban _
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SRO Test
QUESTION #081
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The plant is in MODE 5 with the containment purge exhaust fan operating and containment purge supply off. The Containment Coordinator identifies a positive air flow ,
condition from containment to outside atmosphere through the equipment hatch with the !
! containment personnel hatch ope *
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j Which ONE of the below actions should be performed for this condition? l
A. Activate a Containment Purge Isolation [
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B. Start either Fuel Bldg / Aux Bldg Emergency Exhaust train l l i t C. Activate a Control Room Ventilation Isolation
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D. Shift the Aux Building Normal Exhaust to FAST l
ANSWER: '
D. Shift the Aux Building Normal Exhaust to FAST l RO MS SRO #43 -
K/A #029000K103 t OBJECTIVE #003 A120B REFERENCES: OTN-GT-00001
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SRO Test QUESTION #082 Given the following plant conditions:
. Operating at 100% power at MO . All systems are operabl . While in AUTO rod control, Control Bank "D" starts stepping in slowly, but at a noticeable rate.
. Which ONE of the following events will cause this response?
A. A tube leak in the Regenerative Heat Exchange B. A tube leak in the Seal Water Heat Exchange C. A tube leak in the Letdown Heat Exchange D. A tube leak in the Excess Letdown Heat Exchange ANSWER:
B. A tube leak in the Seal Water Heat Exchange SRO #57 K/A #008010A303 OBJECTIVE #0110100H REFERENCES: T61.0110.6, LP-#10
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SRO Test QUESTION #083 A Hi Hi Radiation signal from SJ-RE-02, Steam Generator Blowdown System Radiation Monitor, will automatically close which ONE of the following valves?
A. BM-HV-21, S/G 'C' Blowdown Nuclear Sampling System Upper Isolation Valv B. BM-FV-54, S/G Blowdown Discharge Pumps Discharge Flow Control Valv C. BM-HV-6, S/G 'B' Blowdown Nuclear Sampling System Line Downstream Isolation Valv D. BM-HV-38, S/G 'D' Blowdown Nuclear Sampling System Lower Isolation Valv ANSWER:
C. BM-HV-6, S/G 'B' Blowdown Nuclear Sampling System Line Downstream Isolation Valv RO i!56 l
SRO 447 '
K/A #073C00K101 OBJECTIVE #0110120D REFERENCES: T61.0110.6 LP-#12 OTO-SA-00001
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SRO Test ;
QUESTION #084 l
i Given the following plant conditions- '
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- Steam Break in AREA 5 !
e All MSIVs closed j
e 'A', 'B', and 'D' Steam Generator Pressures Stable e l
'C' Steam Generator Pressure Decreasino j e Performing actions ofE-2," Faulted Stun Generator Isolation" ,
. TD AFW pump is the only AFW pump available i Which ONE c /the following actions would be performed during completion of E-27
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A. Close ABHV0006, 'C' Steam Supply to the TD AFW pum B. Open all S/G Common Sample Isolation Valves, BMHV0065 through 6 C. Reduce Aux Feedwater flow to 15,000 lbm/hr to each Steam Generato D. Close ABLV0007, Main Steam Low Point Drain SG 'C'.
l ANSWER:
B. Open all S/G Common Sample Isolation Valves, BMHV0065 through 68,
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RO #66
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SRO #62 f K/A #0000404103 My OBJECTIVE #003D150C REFERENCES E-2 I i
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Y SRO Test QUESTION #085 i
Main Turbine exhaust pressure is 4" Hga and increasing at a rate of 0.5" Hga per minut !
Which of the following is the minimum amount of time that could elapse before an automatic low vacuum turbine trip occurs?
A. 5 minutes B. 7 minutes C. 9 minutes ;
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D. 12 minutes ANSWER: I
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B. 7 minutes l
RO #73 SRO #64 K/A #000051 A202 OBJECTIVE #003BB90A )
REFERENCES: OTO-AD-00001 l
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SRO Test !
E QUESTION #086 .j i
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Which one of the following situations will require completing a " Request to Exceed NRC Overtime Restrictions" form?
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A. An I&C Computer Technician is called out to work the OWL shin immediately preceding his scheduled AM shif <
B. An Operating Supervisor works 7 a.m. to 3 p.m. in Training, then starts the Night Shift at 6 p.m. the same day and works until 6 a.m. the following mornin C.. An Equipment Operator works 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period due to a change from Daylight Savings Tim D. A Rad-Chem Technician works 7 a.m. to 7 p.m. for six continuous day ANSWER:
B. An Operating Supervisor works 7 a.m. to 3 p.m. in Training, then starts the Night Shia at 6 p.m. the same day and works until 6 a.m. the following mornin SRO #12 K/A #194001 A103 OBJECTIVE #003A290E REFERENCES: APA-ZZ-00905 i
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SRO Test I i
QUESTION #087
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t Which one of the following is an entry condition for OTO-72-00003, Loss of Shutdown i Margin? t A. Mode 3, following Reactor Trip at 0950 and RCS Tavg 545 F at i11 :
B. Mode 2, with Reactor Power at 5% and Control Bank C at 35 step !
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C. Mode 3, with RCS temperature decrease of 100'F in 20 minutes with ECCS operating in the Injection phas l;
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D. Mode 5, with Shutdown Margin Calculation indicating the core net reactivity at j-1100 pcm i
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ANSWER: !
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B. Mode 2, with Reactor Power at 5% and Control Bank C at 35 step !
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K/A #000024G10 OBJECTIVE #003B610A i
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REFERENCES: T61.003B.6 LP-#B-61 OTO-ZZ-00003 l Plant Cune Book l
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SRO Test i
QUESTION #088
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A working copy of a procedure is taken from the " Working File" in the Field Office on 9/21/96 at 0900. Which one of the following would allow this procedure to be used in the plant on 9/25/96 at 1700?
A. The procedure copy was verified to be the correct revision and signed by the Operating Supervisor on 9/21/96.
B. The procedure is marked " Controlled Copy" and was signed and dated by the Shin I Supervisor on 9/23/96.
C. The procedure is marked " Working File" and has been initialed and dated on each shin j since issu ;
D. The procedure is marked " Working Copy" and was signed by the Shin Clerk on 9/24/96 at 2359.
ANSWER: ,
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D. The procedure is marked " Working Copy" and was signed by the Shin Clerk on 9/24/96 at 2359.
SRO #16 K/A #194001 A101 OBJECTIVE #003 AA6B2 l REFERENCES: ODP-ZZ-00009 Modified from 1994 NRC Exa . . _ __
SRO Test QUESTION #089
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The secondary equipment operator notes that Annunciator 6E, D.C. Control Power Failure Alarm" is on for diesel NE01 local alarm panel. On Panel KJ121 IL1 and IL2 lights are OFF, IL3 and IL4 lights are O Which ONE of the following describes the effect on the diesel generator?
A. NE01 is OPERABLE if staning air pressure is maintained 610 to 640 psi B. NE01 is INOPERABLE since diesel start circuits are disable C. NE01 is OPERABLE as long as outside air temp is less than or equal to 65 D. NE01 is INOPERABLE since the fuel oil transfer pump is disable ANSWER:
B. NE01 is INOPERABLE since diesel start circuits are disable SRO #95 K/A #000058A201 OBJECTIVE #011003DD REFERENCES: OTA-KJ-00121 l
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SRO Test i i I
i QUESTION #090
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! A void exists in the reactor vessel durin3 natural circulation cooldown. Which ONE of the following actions is used to collapse a~ excessive void, according to ES.O.3," Natural Circulation Coo!down with Steam Voids"?
A. Decrease RCS temperature while maintaining RCS pressure constan )
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- C. Increase RCS pressure using pressurizer heaters while maintaining pressurizer level.
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l D. Start an Si pump to increase RCS pressure while maintaining temperature constant.
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ANSWER:
C, increase RCS pressure using pressurizer heaters while maintaining pressurizer leve RO #69 SRO #72 K/A #000074A101 OBJECTIVE #003D070K REFERENCES: T61.003 ES-0.3
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QUESTION #091 i
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The plant is in MODE 1 with all systems in normal except that 1&C is performing corrective maintenance in the Rod Control Power Cabinet IBD. Group 1 ofControl Bank D is being energized from the DC Hold Bu ;
Breaker PGl902, Motor Circuit Breaker to Rod Drive Motor-Generator SF01, is i inadvertently opened. All plant systems respond as designe Which ONE of the below is true regarding power to the control rods?
A. Power continues to all control rod I J
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B. Power is interrupted to all control rod l
C. Power is interrupted to all rods except Control Bank D, Group D. Power continues to all rods except Control Bank D, Group ANSWER: l l
A. Power continues to all rod RO #15 SRO #18 K/A #001000K202 OBJECTIVE #0110260G REFERENCES: T61.0110.6 LP-#26 I i
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t QUESTION #092 j
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j- Which one of the following could be a direct result of a loss of Vital AC Instrument bus !
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f A. Charging Pump suction swaps to the RWST j i
B. Source Range Hi Flux Reactor Trip C. Intermediate Range High Flux Reactor Trip !
r D. CVCS Letdown Isolation
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ANSWER:
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D. CVCS Letdown Isolation RO #79 SRO #67 !
K/A #000057A219 l OBJECTIVE #003B450A REFERENCES: OTO-NN-00001 i
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SRO Test QUESTION #093
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A reactor trip has occurred and the operating crew is responding in accordance with ES-0.1, Reactor Trip Respons i
. Reactor trip and bypass breakers open a NIS power is 1% and decreasing e
Bank D, Group 2 rods indicate 188 steps withdrawn. All other rods are fully inserted Which one of the following is TRUE for the above conditions?
A. An emergency boration of 450 ppm must be performed to ensure the minimum shutdown margin is maintaine B. An emergency boration of 150 ppm must be performed to limit fission gas release and maintain fuel pellet temperature within design limit i C. No immediate action is required since the core is designed for these conditions, and the reactor has been verified tripped by diverse indication D. A safety injection signal (SIS) must be actuated to maintain the reactor core in a safe shutdown conditio i ANSWER:
A. An emergency boration of 450 ppm must be performed to ensure the minimum shutdown margin is maintaine RO #65 SRO #58 K/A #000005K301 OBJECTIVE #003D060C REFERENCES: ES-0.1
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SRO Test ;
i QUESTION #094 l
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i l During a Reactor Startup, the Reactor Operator verifies one decade of overlap between the source and Intermediate Range Nuclear Instruments. This verification is defined as ,
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A. Source Check I l
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B. Analog Channel Operational Test i l
C. Channel Calibration :
D. Channel Check i ANSWER:
D. Chemel Check i
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RO #11 SRO #10 K/A #194001 Al13 OBJECTIVE #003A0211
' REFERENCES: Tech Spec Definitions l
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SRO Test
QUESTION #095 l
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I Which ONE of the following components is manually (or automatically) isolated and remains isolated for a Faulted 'B' Steam Generator, but NOT necessarily for a 'B' Steam Generator Tube Rupture? (NOTE: Assume all equipment actuated as required.)
A. Main Steam isolation Valve (AB-HV-17)
B. Main Feedwater Isolation Valve (AE-FV-40) :
C. Auxiliary Feedwater Flow Control Valve (AL-HV-10)
D. Main Steam Supply Valve to T/D AFW Pump (AB-V085)
ANSWER:
C. Auxiliary Feedwater Flow Control Valve (AL-HV-10)
RO #84 I SRO #89 l K/A #000038A132 i
OBJECTIVE #003D17NN l REFERENCES: T61.003D.6 LP-#17 E-3, SGTR E-2, Faulted S/G Isolation l
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SRO Test QUESTION #096 Given the following:
- Unit 1 is operating at 100% powe * All c=:rels are in the normal power operation lineu * Pressurizer levelis DECREASIN I e VCT level is INCREASING.-
e SEAL INJECTION TO RCP FLOW LO alarm is li {
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e REGEN HX HI TEMP alarm is li e l
LETDN HX DISCHARGE HI TEMP alarm is li )
e CHARGING LINE FLOW Hl/LO alarm is li I Which ONE of the following procedures should be implemented?
A. OTO-BG-00001, Loss ofLetdown B. OTO-BG-00002, Loss of Charging
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C. OTO-BB-00003, RCS Excessive Leakage D. OTO-BB-00001, Steam Generator Tube Leak ANSWER:
B. OTO-BG-00002, Loss of Charging SRO #97 K/A #000022A201 OBJECTIVE #003B220A REFERENCES: OTA-RL-RK042, Att. A j i
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SRO Test QUESTION #097 r
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The Technical Specification bases for observing that the RCCAs are positioned above their respective insertion limits during normal operation include which one of the following?
A. Ensures that the moderator temperature coefficient is within its analyzed rang B. Ensures that the trip instrumentation is within its normal operating rang .
C. Ensures that the pressurizer is capable of being Operable with a steam bubbl D. Ensures that acceptable power distribution limits are maintaine ANSWER:
- D. Ensures that acceptable power distribution limits are maintaine SRO #75 ;
K/A #000001K302 i OBJECTIVE #003 AA3E2 REFERENCES: TS 3/4.1.3 Bases
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SRO Test QUESTION #098 Which ONE of the following situations violates a requirement for containment integrity or containment closure?
A. A containment vent is performed with the plant operating at 100% power.
B. The plant is in refaeling mode with the refueling cavity flooded. Steam generator safeties have been removed; secondary manways are also removed. No fuel movement is in progress.
C. The plant is in refueling mode with fuel movement in progress. Containment Shutdown purge is initiated.
D. The plant is in hot standby. The "A" steam generator blowdown isolation valve BM-HV-1 is stuck open.
ANSWER:
D. The plant is in hot standby. The "A" steam generator blowdown isolation valve BM-HV-1 is stuck open.
RO #78 SRO #71 K/A #000069A202 OBJECTIVE #003E014A REFERENCES: TS 3. TS 3.6. TS 3. ~
l SRO Test l l
QUESTION #099 )
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The Callaway Plant is operating at 100% power, with 'B' CCP in service. Assuming no operator action, which ONE of the following components could suffer a sustained loss of Instn2 ment Air and NOT cause an AUTOMATIC Reactor Trip? Consider each component individually.
A. BGLCV0459, RCS Loop 3 letdown to regen hx level control valve I
B. BGHV8141B, RCP B #1 seal water outlet isolation valve C. BGFCV121, CVCS CCP A & B discharge to regen heat exchanger flow control valve D. KAFV0029 Reactor Building instrument air supply flow control valve ANSWER:
B. BGHV8141B, RCP B #1 seal water outlet isolation valve l I
l SRO #96 K/A #000065A206 OBJECTIVE #003B330A REFERENCES: OTO-KA-00001
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SRO Test QUESTION #100
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Which ONE of the following components has its air supply AUTOMATICALLY isolated if air pressure decreases to 108 psig?
l A. Closed Cooling Water Temperature Controller B. First Stage RHDT Level Control Valves C. Main Feedwater Reg Valve Bypass Valves D. Auxiliary Feedwater Pump Room Sump Pumps ANSWER i
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I D. Auxiliary Feedwater Pump Room Sump Pumps RO #46 SRO #49 K/A #079000K101 ;
OBJECTIVE #0110140C l REFERENCES: OTO-KA-00001 l l
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CALLAWAY PLANT
, EXAMINATION COVER SHEET TRAINING DEPARTMENT l
l COURSE TITLE: SRO INITIAL LICENSE EXAMINATION DATE: 2/24/97 I
NAME (Print): SCORING:
{
] SIGNATURE: Points Possible: 100 Points Missed:
Grade:
l DIRECTIONS: DLACK OUT CORRECT ANSWERS 1. l A l l B l E l D l 26. l A l E l C l l D l 51. l A l l B j E l D l 76. l A l l B l l C l E j 2. l A l E l C l l D j 27. W E l C l l D l 52. l A j l B l l C l E 77. l A l l B l l C l E l 3. l A l l B l l C l E 28. E l a l l c i l o l 53. R E I c I I o l 78. R E I c l l o l
- 4. l A j j B l E y 29. l A l l B l E l D l 54. E l B j % % 79. E % l C l l D l S.lAl E.lCl % 30. l A j j B l [ E 55. % l B l l C l E 80. l A l l B j [C] E 6. % l B l l C l E 31. E l B j l C l % 56. l A l %y 81. W l B l l C l E 7. E l B l l C l l D l 32. l A l E l C l l D j 57. l A l l B l l C l l 82. % E W l D j 8. l A l E l C l l D l 33. l A j j B l % E 58. % B lCl E 83. % l B l E j D j 9. E [ B ) l C l l D l 34. E [ B ) l C l l D ] $9. l A l E l C l l D l 84. l A l E W l D l 10. l A l E.l C l l D l 35. l A l l B l E y 60. % l B l l C l E 85. % E l C l %
11. l A l E W l D l 36. l A l E l C l % 61. E l B l W l D l 86. l A l E l C l @
12. E l B j l C l l D l 37. E l B l l C l l D l 62. E % % l D l 87. % ~ lCllDl 13. l A l W W E 38. l A l E % l D l 63. l A l E l C l l D l 88. l A l % l C l E 14. l A j j B l l C l E 39. hl E W l D j 64. l A l l B l l C l 89. [ E l C l %
15. l A l E,l C l l D l 40. l A l E l C l [ D ) 65. E l B l l C l l D j 90. hl l B l E %
16. l A l E l C l l D l 41. W l B j E l D l 66. l A l l B l E l D j 91. E l B l l C l l D l 17. [ A } l B l lDl 42 l A l l B l E l D l 67. l A l l B l E y 92. l A l l B l l C l E 18. j A l l B l % E 43. E l B l l C l l D l 68. l A l l B l l C l E 9 %lCllDl 19. l A] E l C l l D l 44. hl [ B ) l C l E 69. l A l l B l E l D l 94. l A l l B l l C l E 20. l A l {BJ lDl 45. l A l l B l E l D j 70. E l B j j C l l D l 95. W l B l E l D j 21. W E l C l l D j 46. l A l (I) % E 71. E l B l l C l @ 96. l A l E l C l l D l 22. W l B l E l D j 47, E [ B ) l C l l D l 72. l A l W E l D l 97. l A l l B l l C l E 23. l A l E l C l l D l 48. l A j E l C l l D l 73. W l B l l C l E 98. l A l l B l l C l E 24. l A j E l C l [ D) 49. E l B j j C l l D j 74. l A l E l C l l D l 99. l A l E % l D j 25. l A l l B l E l D l 50. E % l C l l D j 75. % E l C l l D l 100. l A l W l C l E
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CHIEF EXAMINER WRITTEll EXAM RESULTS ANALYSIS - CALLAWAY 2/24/97 i
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Each exam had 100 questions valued at one point eac l
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SRO: High - 93; Low - 83; Average - 89 l l
RO: High - 92; Low - 73; Average - 8 '
Analysis.
For the same questions, the same question numbers were used on either exam. The chief ;
l examiner concurs with the licensee's analysis attached. More than half of the applicants missed joint questions 3,4,55,64,84, and 90. More than half of the applicants also
- missed SRO question 3 All of the above questions were determined to be valid. No generic training or knowledge deficiencies were identified. Reasons for missing these questions appeared to be related to question difficulty and isolated training weaknesses. The licensee initiated appropriate actions to upgrade candidate specific knowledge and correct specific training weaknesse I
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Review ofInitial NRC Written Examinations Callaway Plant - 2/24/97
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The Reactor Operator and Senior Reactor Operator examinations were graded by F.X. Biermann and checked by R.A. Nelson. Both examinations were reviewed using the guidance contained in ES-403, Grading Site Specific Examinations at Power Reactors. This review is documented on the attached completed QA Checkoff Sheets, ES-403- This review revealed that seven (7) questions were missed by greater than 50% of the candidate Below is a summary of actions taken for each specific question:
Ouestion # Tonic Action 3 Plant Computer Alarm Operation item not specifically covered by lesson plan objectives. Submitted CA-#1031 to change objective WPA Tagging Requirements TFR written to emphasize method of tagging to be used on 4160V and above power block breakers when work is to be perfonned on downstream component SRO #33 Plant Security Event Question beyond objective oflesson plan. TFR written to evaluate if actions should be included into lesso Liquid Process Monitor Failure TFR written to include system operation of Liquid Process Monitors. Stress differences between liquid and atmosphere monitor PR Nuclear Instrumentation Failure Question evaluated, valid and correc No action require Cover with candidat EOP E-2 Actions Question evaluated, valid and correc No action require Cover with candidat EOP ES-0.3 Pressure Control Question evaluated, valid and correc No action require Cover with candidat In addition the subject questions above were examined for any common deficiencies regarding systems, types of operations imiolved, or safety system functions. No common deficiencies were note ES-401 Written Examination Cover Sheet Form ES-401-1 U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION APPLICANT INFORMATION Name: Region: IV Date: Febmary 24,1997 Facility / Unit: Callaway License Level: RO Reactor Type: Westinghouse INSTRUCTIONS:
Use the answer sheet provided to document your answers. Staple this cover sheet on
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top of the answer sheet. Each question is wonh one point. The passing grade requires a '
final grade of at least 80 percent. Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination start All work done on this examination is my own. I have neither given nor received ai Applicant's Signature RESULTS Examination Value 100 Points
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Applicant's Score Points
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Applicant's Grade Percent
_ . __ _ __ _ . - _ . _ . _ _ . . _ . _ _ . _ . _ _ _ _ _ __ .. ES-401 Written Examination Cover Sheet Form ES-401-1
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION APPLICANT INFORMATION l
Name: Region: IV Date: February 24,1997 Facility / Unit: Callaway
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License Level: RO Reactor Type: Westinghouse !
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l INSTRUCTIONS:
l Use the answer sheet provided to document your answers. Staple this cover sheet on l
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top of the answer sheet. Each question is worth one point. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination start j l
All work done on this examination is my own. I have neither given nor received ai j i
Applicant's Signature RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent
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ES-402 Policies and Guidelines Attachment 2 for Taking NRC Written Examinations
l Cheating on the examination will result in a denial of your application and could i result in more severe penaltie ! After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given ;
assistance in completing the examinatio l To pass the examination, you must achieve a grade of 80 percent or greate . Each question is worth 1 poin ! There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examinatio I l Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the l answer shee l l Mark your answers on the answer sheet provided and do not leave any question j blan I
! If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyons outside the examination room to eliminate even the appearance or possibility of cheatin !
1 When you complete the examination, staple the examination cover sheet on top I of the answer sheet and give it to the examiner or proctor. Remember to sign the ,
statement on the examination cover shee l 1 After you have turned in your examination, leave the examination area as defined by the examine ;
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i l' QUESTION #001
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When performing a boration to the reactor coolant system for a down power transient, the
- PZR heaters should be turned on in manual to
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A. Maintain PZR pressure in the normal operating range during the down power.
, B. Allow an increased ramp rate for the down powe C. Equalize the reactor coolant system and PZR boron concentrations.
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- D. Ensure positive PZR control is established prior to starting the down power.
I j ANSWER:
C. Equalize the reactor coolant system and PZR boron concentration ;
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SRO #21 i K/A #004000K601 OBJECTIVE #003AA4B1 REFERENCES: OTN-BG-00002, " Reactor Makeup Control and Boron Thermal l Regeneration System"
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QUESTION #002 :
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The plant experiences a sustained loss of all AC power.
- Which ONE of the below would be used to makeup to the spent fuel pool due to low
- spent fuel poollevel?
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. A. Pressurize VCT and use Reactor Makeup I
B. Diesel Fire Pump and fire hose l
C. Gravity drain condensate storage tank D. Essential service water emergency makeup
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ANSWER: '
B. Diesel Fire Pump and fire hose i
RO #41 SRO #44 K/A #033000G11 OBJECTIVE #003D220Z REFERENCES: ECA-0.0, Step 23
RO Test QUESTION #003 Which ONE of the below computer data quality codes indicates that the alarm function is still operable?
A.DALM B. DEL C.SUB D.LRL ANSWER:
D.LRL RO #12 SRO #11 K/A #194001Al15
- OBJECTIVE #003A02D4 REFERENCES: OOA-RJ-00001 l l
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RO Test QUEST!'iN #004 Preventive Maintenance is scheduled on the 'A' Condensate Pump Motor and its supply breaker PB0304. Which ONE of the following locations MUST be tagged in accordance with the Workman's Protection Assurance Program?
A. Breaker PB0304 local handswitch B. Condensate Pump Discharge Valve C. Racking Mechanism for Breaker PB0304 D. Main ControlBoard Switch AD-HIS-1 ANSWER:
C. Racking Mechanism for Breaker PB0304 RO #4 SRO #4 K/A #194001K107 OBJECTIVE #003A330F REFERENCES: APA-ZZ-00310 Page 20
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RO Test QUESTION #005 l
A Reactor Startup is in progress with Control Bank B at 50 steps and Reactor Power at 102 CPS.
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Which ONE of the following is required if Source Range Nuclear l Channel N32 fails high?
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l A. Place N32 in the tripped condition within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. Verify all Rod Bottom Lights li C. Verify Shutdown Margin within one hou D. Insert all Control Banks and repair channel N3 ANSWER:
B. Verify all . Rod Bottom Lights li RO #96 SRO #88 K/A #000032G11 l OBJECTIVE #0110280E !
REFERENCES: OTO-SE-00001 l E-0 Tech Spec 3.3.1 l
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RO Test -
- QUESTION #006 The reactor tripped 5 minutes ag l Which one of the following completes the statement concerning the heat transfer !
- relationship between the RCS and Ste an Generators?
The heat transfer rate between the RCS and the S/Gs will:
' A. decrease as RCS temperature increases and AFW flow increase B. decrease as AFW temperature decreases and AFW flow increase C. increase as 'AFW temperature increases and RCS flow decrease D. increase as RCS temperature increases and AFW flow increase ANSWER:
D. increase as RCS temperature increases and AFW flow increase RO #33 SRO #33 K/A #061000K501 OBJECTIVE #003D260R REFERENCES: T61.003D.6
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RO Test j
QUESTION #007 Which of the following are allowable relaxations for Independent Verification when restoring a system requiring IV?
1. Comparing the tagout control sheet to current plant reference material (flow l
diagrams, procedures, etc.) to ensure adequacy of the tagou I i
l 2. Verifying status lights, annunciators, meter indications, etc. on the main control board that unequivocally depicts the equipment statu . Performing a functional test that verifies that the component is in the specified l configuratio . When the concept of ALARA would be violated.
I I A. 2, 3, 4 B.: 1, 2, 4
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C. 1, 2, 3 i l
D. 1, 3, 4 ANSWER:
A. 2, 3, 4 i RO#1 SRO #1 K/A #19400lK101 j OBJECTIVE #003A33 A6 REFERENCES: APA-Z7 00310 i
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QUESTION #008 j
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I Which statement describes loss of CCW flow to a RCP in the emergency procedures? !
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A. CCW flow is low for >10 minutes or RCP motor bearing temperature is >l95'F l
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B. CCW flow is low for >10 minutes or RCP motor bearing temperature is <!95' !
C. CCW flow is interrupted for >10 minutes or RCP motor bearing temperature is
> l95' D. CCW flow is intermpted for >10 minutes or RCP motor bearing temperature is
<195* ANSWER:
C. CCW flow is interrupted for >10 minutes or RCP motor bearing temperature is
>195' RO #75 I K/A #000015A210 OBJECTIVE #003D040H REFERENCES: E-0, Foldout l
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RO Test QUESTION #009 OTO-ZZ-00001, Control Room Inaccessibility, requires operation of three ' Control Room Isolation Transfer' switches on the Auxiliary Shutdown Panel, which isolate control and indication of the associated devices from the control roo Which ONE of the following describes the reason for operating these switches?
i A. Prevent inadvertent actuation of components which are necessary to safely shutdown the plant.
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B. Initiates a reactor trip and transfer control of the plant to the auxiliary shutdown !
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C. Required by Technical Specifications action to ensure that auxiliary shutdown l Operability is satisfie !
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D. Transfers alarm and control of pressurizer heaters from the Control Room.
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l A. Prevent inadvertent actuation of components which are necessary to safely shutdown j the plant.
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RO #71 SRO #69 K/A #000067K304 OBJECTIVE #0110480D REFERENCES: T61.0110.6 LP-#48
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RO Test QUESTION #010 Containment Spray actuates, and is still required, following a large break LOCA in containment. Cold Leg Recirculation alignment per ES-1.3 has been completed earlier for 1 the ECCS pumps. The "RWST LO/LO 2" annunciator alarms with RWST level at 9% l
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and decreasin Which ONE of the following actions should be performed on the Containment Spray i System?
A. Stop the Containment Spray Pumps at 5% RWST level if the pump suctions do nat automatically swap from the RWST to the containment recirc sumps at 9% leve B. Open containment spray suctions from the containment sumps, reset the CSAS actuation, close pump suctions from the RWST while allowing the Containment Spray j Pumps to continue to ru C. Stop the Containment Spray Pumps, open containment spray suctions from the !
{ containment sumps, reset the CSAS actuation, close pump suctions from the RWST, i and then restart the Containment Spray Pump !
D. Immediately reset the CSAS actuation and stop one Containment Spray Pump, verify all containment coolers in service, then stop the other Containment Spray Pump at 5%
RWST leve ANSWER:
B. Open containment spray suctions from the containment sumps, reset the CSAS actuation, close pump suctions from the RWST while allowing the Containment Spray Pumps to continue to m RO #44 K/A #026000K401 OBJECTIVE #0110180F REFERENCES: ES- a RO Test QUESTION #011
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The following conditions exist:
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Containment pressure transmitter PT-937 declared inoperable
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Required Technical Specification Actions have been taken for channel 937 Which ONE of the following statements describes the coincidence for a Containment Spray Actuation to occur and the actions that will result in this coincidence?
A. 2/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A)in the TEST positio B. 2/3 coincidence after the channel is placed in the BYPASS condition, by placing bistable (PB-937A) in the TEST positio C. 1/3 coincidence after the channel is placed in the TRIP condition, by placing bistable (PB-937A)in the TEST positio D.1/3 coincidence after the channelis placed in the BYPASS condition, by placing 1 bistable (PB-937A) in the TEST positio ANSWER:
B. 2/3 coincidence after the channelis placed in the BYPASS condition, by placing bistable (PB-937A) in the TEST positio RO #23 SRO #24 K/A #013000K502 OBJECTIVE #003A0212 REFERENCES: T/S 3.3.2 ACTION c, Table 3.3-3 FU 2.c ACTION 16 i PRINT 7250D64 S008 i
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QUESTION #012 l
Following a LOCA, hydrogen concentration in the containment has increased slowly over several days, reaching 10 volume per cen j l
Which ONE of the following actions should be taken?
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A. One trair, of the electric hydrogen recombiner system should be placed in servic !
l B. Electric hydrogen recombiners should be placed in service when hydrogen i concentration reaches 4.0 volume per cen l C. Electric hydrogen recombiners ca 'not be placed in service. Heater operating temperature on the recombiner exceeds ignition temperature for hydrogen at this concentratio l
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D. Both trains of electric hydrogen recombiners should be placed in service in conjunction with a containment purge.
ANSWER:
A. One train of the electric hydrogen recombiner system should be placed in service.
RO #63 SRO #42 K/A #028000K501 OBJECTIVE #0110400J REFERENCES: OTN-GS-00001 E-1
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RO Test t QUESTION #013 Which ONE of the below is designed to protect the reactor from an uncontrolled RCCA bank withdrawal from a suberitical condition?
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A. C-5 Low Power Interlock B. Boron Dilution Flux Doubling Actuation !
C. Source Range High Flux Trip D. High Positive Flux Rate Trip
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ANSWER:
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RO #87 i K/A #00000lK103 e
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OBJECTIVE #0110270D ,
l REFERENCES: T61.0110.6 LP-#27 i i l
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R O T cst QUESTION #014 With the plant in MODE 1 the Shift Supervisor is notified by security that a confined penetration has occurred by unauthorized personnel into the NB01 switchgear room. The Plant Emergency Alarm is sounded and the CODE RED is announced over the Gai-tronics.
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Which ONE of the below may be included in the initial response by Control Room personnel?
A. Trip the reactor, perform Control Room evacuation, and commence RCS cooldown from the Aux Shutdown Pane B. Shut the Control Room missile door, trip the reactor, and commence RCS cooldown from the Control Roo C. Shut the Control Room missile door, increase monitoring ofMCB indications, and have all Equipment Operators report to the Field Offic D. Trip the reactor, commence RCS cooldown from the Control Room, and evacuate all non-essential personne ANSWER:
B. Shut the Control Room missile door, trip the reactor, and commence RCS cooldown from the Control Roo RO #13 K/A #194001 Al16 OBJECTIVE #003B280B REFERENCES: EIP-ZZ-00102, At OTO-SK-00001 l
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RO Test QUESTION #015
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l Which ONE of the following Area Radiation Monitors is required by Technical ;
Specifications? .
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A. Containment Area Radiation Monitor SDRE0041 )
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B. New Fuel Storage Area Radiation Monitor SDRE0035 j
i C. Control Room Area Radiation Monitor SDRE0033 )
D. Cask Handling Area Radiation Monitor SDRE0034 ANSWER: i
. B. New Fuel Storage Area Radiation Monitor SDRE003 RO #36 SRO #34 K/A #072000K302 OBJECTIVE #0110360G REFERENCES: T/S 3.3.3.1, Table 3.3-6 FU 2.b.(2)
Callaway Bank l
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RO Test j QUESTION #016 Which of the following flow paths correctly describes how power is normally supplied to a t typical reactor protection instrument bus?
A. 480V AC from the safeguard bus, rectified to 125V DC, inverted to 120V AC, and supplied to the instrument bu B. 480V AC from the safeguard bus, transformed to 120V AC, and supplied to the instrument bu C. 125V DC from the battery, supplied to the battery bus, and supplied to the instrument bu D. 480V AC from the safeguards bus, rectified to 120V DC, and supplied to the instrument bu ANSWER:
A. 480V AC from the safeguard bus, rectified to 125V DC, inverted to 120V AC, and supplied to the instrument bu j RO #48 K/A #062000K201 OBJECTIVE #0110060A REFERENCES: T61.0110.6 LP-#6 i
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RO Test QUESTION #017 !
l The crew implemented FR-C.1, Response to Inadequate Core Cooling.
Which one of the following combinations of core exit thermocouples and indicated temperatures would require starting RCP's, even if the normally required support conditions could not be met? i
SRO #27 K/A #017020A402 OBJECTIVE #003D250E REFERENCES: FR-C.1 Background
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RO Test QUESTION #018 Callaway Plant is preparing for Reactor Core Offload with Refueling Pool Level at 391 inches (2046 fl. level). The polar crane operator inadvertently lifts the Reactor Vessel Upper Internals out of the water and causes a Hi Hi alarm on Containment Building Area Radiation Monitor SDRE004 Which ONE of the follow *mg is a required Immediate Action?
A. Close ECV0995, Fuel Transfer Tube Isolation Valv B. Initiate a Containment Purge Isolation Signal (CPIS).
C. Transfer the Charging Pump suction to the RWST and increase flo D. Evacuate personnel from containmen ANSWER:
D. Evacuate personnel from containmen RO #94 SRO #91 K/A #000061G09 OBJECTIVE #003E05I4 REFERENCES: OTO-KE-00001 OTA-RL-RK062, Att. A l
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- QUESTION #019
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FR-S.1 " Response to Nuclear Power Generation /ATWS" Step 2 requires a turbine tri ; Why would it be desirable to trip the turbine if a reactor trip had not been achieved?
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(Choose ONE)
J A. The reactor will be suberitical due to manual rod insertion before the turbine is tripped.
i B. Tripping the turbine will conserve SG inventory and limit the pressure transient that j would result from a loss of all feedwater.
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) C. Tripping the turbine will insert negative reactivity from moderator temperature
! coefficient, thus assisting in reactor shutdown.
D. Tripping the turbine will generate an additional reactor trip signal and suppress core void formation by increasing RCS pressure. ANSWER:
J l B Tripping the turbine will conserve SG inventory and limit the pressure transient that . 1 j would result from a loss of all feedwate '
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- SRO #61
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K/A #000029K312 j OBJECTIVE #003D290C
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REFERENCES: T61.003D.6 LP-#29 a
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I QUESTION #020 ;
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Which ONE (1) of the following is the HIGHEST RCS pressure at which the Safet l Injection Pumps will deliver water to the RCS7 ,
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A. 1050 psig B. 1250 psig C.1450 psig l
D. 1650 psig l ANSWER:
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C. 1450 psig ]
RO #43 SRO #38 K/A #006000K603 OBJECTIVE #0110170A REFERENCES: E-0 T61.0110.6 LP-#17 i
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l QUESTION #021 While performing actions in E-3, " Steam Generator Tube Rupture" the Control Room )
Supervisor asks the Balance ofPlant Operator to check intact Steam Generator narrow i range levels greater than 4%. Which ONE of the following BOP responses would satisfy Callaway Plant Communication Guidelines?
A. Yes, intact Steam Generator narrow range levels are greater than 4%.
B. Yes, intact Steam Generator narrow range levels are 50% and stabl C. Yes, intact Steam Generator narrow range levels are increasin D. Yes, intact Steam Generator narrow range levels are 10%.
ANSWER:
B. Yes, intact Steam Generator narrow range levels are 50% and stabl RO #8 SRO #7 K/A #194001 A105 OBJECTIVE #003A060H REFERENCES: UEND-COMMUNICATIONS-01, Page 4 of 5 l
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RO Test QUESTION #022
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Given the following: 1
- The Main Turbine tripped from 95% powe All systems responded normally to the tri i Which ONE (1) of the following is the expected position of the steam dump valves with ;
Tavg at 575'F?
Full Open Modulating Full Closed ANSWER: , RO #57 SRO #55 K/A #041020K418 OBJECTIVE #0110200J REFERENCES: T61.0110.6 LP-#20 l
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RO Test QUESTION #023
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A plant startup is in progress with power indicating IE-6% on both IR channels. Which !
one of the following will occur ifIR channel N36 fails to 21%7 i A. IR high flux reactor trip B. Manual and automatic rod stop i
C. PZR low pressure reactor trip is unblocked l
D. PR low flux reactor trip l
ANSWER:
B. Manual and automatic rod stop RO #95 SRO #87 K/A #000033A202 OBJECTIVE #0110260J REFERENCES: OTO-SE-00002
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QUESTION #024
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f Given the following conditions:
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- Pressurizer Pressure = 1710 psig l
- RCS C.L. Temperature = 560 F l
- Core Exit TC = 568*F l l
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Which one of the following is the correct amount of subcooling for the above conditions?
A. 38 B. 41 C. 47 D. 49 ANSWER: 1 B._ 41
K/A #002000K509 l
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' REFERENCES: Steam Table
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RO Test QUESTION #025 The Callaway Plant is performing a Plant Startup following a Refueling Outage. While
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transferring Feedwater Control to the Main Feedwater Reg Valves, a Reactor Trip occurs on Low S/G Level. The resulting Aux Feedwater Actuation has caused RCS Tavg to decrease to 475 F. All systems operate as designed.
Which one of the following components would be cooled by the Service Water System?
A. 'A' Class IE Air Conditioner B. 'B' Containment Spray Pump Room Cooler C. 'A' Component Cooling Water Heat Exchanger D. 'B' Closed Cooling Water Heat Exchanger ANSWER:
D. 'B' Closed Cooling Water Heat Exchanger RO #60 K/A #076000K119 OBJECTIVE #0110040G 0110040H REFERENCES: T61.0110.6 LP-#4 l
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RO Test QUESTION #026 With t he plant in MODE 1, AND one safety related CCP INOPERABLE, RCP Seal Injection should be provided by the which will maintain seal cooling in the event of a i
A. Non-safety related charging pump, CCW thermal barrier lea B. Non-safety related charging pump, loss of a single electrical bu C. Opposite train safety related CCP, CCW thermal barrier lea D. Opposite train safety related CCP, loss of a single electrical bu ANSWER:
B. Non-safety related charging pump, loss of a single electrical bu RO #20 SRO #22 I K/A #004000K202 OBJECTIVE #003A04A1 REFERENCES: OTN-BG-00001
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RO Test
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QUESTION #027 i
i Which ONE of the following describes the tagout control used for the temporary .
operation of equipment that is protected under a Hold Of A. The tags shall be cleared prior to operation then a new tagout written and new tags hun B. The tags may be lifted and reused afler operation providing a briefing is held and the individual signed on the WPA is present at the component to be checke C. With Shift Supervisor and Requester approval, equipment may be operated without clearing the tags, if the requester is in the equipment area and operation completed in the same shif D. The tags which must be cleared to allow for the operation can be temporarily cleared, replaced with Caution Tags until the operation is complete, then the Caution Tags replaced with new Hold OffTag ANSWER:
B. The tags may be lifted and reused after operation providing a briefing is held and the individual signed on the WPA is present at the component to be checke RO #2 l
SRO #2 !
K/A #194001K102 OBJECTIVE #003A330L REFERENCES: ODP-ZZ-00310 Page 10
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RO Test
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- QUESTION #028 i
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During operations at 95% power and pressurizer level at 48%, the Tave input to the pressurizer level controller fails low. What INDICATIONS does the operator have that the Tave input failed low?
A. Backup heaters are energized, charging flow control valve slowly closes, high level I deviation alarm actuate .
B. Backup heaters are deenergized, charging flow control valve slowly opens, low level deviation alarm actuate C. Backup heaters are energized, charging flow control valve slowly opens, low level
- deviation alarm actuate D. Backup heaters are deenergized, charging flow control valve slowly closes, high level deviation alarm actuate ANSWER:
A. Backup heaters are energized, charging flow control valve slowly closes, high level deviation alarm actuate RO #40 SRO #39 K/A #011000A203 OBJECTIVE #0110090C REFERENCES: OTO-BB-00004 i l
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RO Test i QUESTION #029 l
l Plant conditions:
- Operating in MODE 1, at 100% powe {
. SJ-RE-01, CVCS Letdown Monitor, Alarming Hi/Hi j
e SD-RE-20, AB 2000 Area, Alarming Hi/Hi l
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Which ONE of the following operator actions is required per OTO-BB-00005, RCS High l Activity?
A. Reduce power
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B. Isolateletdown C. Increase letdown to 120 gpm j
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D. Initiate hourly sampling of the RCS ANSWER:
C. Increase letdown to 120 gpm
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RO #76 i SRO #73 l K/A #000076G008 i OBJECTIVE #003B180A l
REFERENCES: OTO-BB-00005 I
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'I RO Test QUESTlON #030
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- Given the following conditions:
a e Tavg is 576 F 1 i
e Pressurizer Pressureis 2240 psig
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. Charging Flow is being controlled in MANUAL e The BACKT JP HEATERS havejust ENERGlZED
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J F Which ONE of the following is the actual pressurizer level?
A. 37%
B. 42%
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D. 52%
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D. '52% !
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K/A #000028A201 !
OBJECTIVE #0110300K REFERENCES: T61.0110.6 LP-#30 i i
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RO Test QUESTION #031 A Ruptured Steam Generator has been cooled down and depressurized. ECCS pumps have been secured and Normal Charging and Letdown have been established.
Plant Conditions:
- PZR Level 30% and DECREASING
- Ruptured S/G NR LevelINCREASING Which ONE of the following is required to balance inventory?
A. Depressurize the RCS B. Increase RCS Makeup Flow C. Turn on Pressurizer Heaters D. Decrease RCS Makeup Flow ANSWER:
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A. Depressurize the RCS I l
RO #85 SRO #90 K/A #000038K306 OBJECTIVE #003D17JJ REFERENCES: T61.003D.6 LP-#17 E-3, SGTR
RO Test QUESTION #032
I&C Technicians are troubleshooting a Rod Control Urgent Failure alarm that was received during Physics testing. When the technicians pull a Stationary Gripper Firing Card in Power Cabinet IBD, the Control Bank D, Group 1 Control Rods drop to the bottom of the cor Which ONE of the following describes the required action of the Control Room Operators?
A. Adjust Turbine Load to maintain TAVG and TREF AT less than 3* B. Trip the Reactor and proceed to E-0, Reactor Trip or Safety Injectio C. Recover the dropped Control Rods within one hour or be in HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. Initiate boration to restore SHUTDOWN MARGIN to greater than or equal to 1.3%
AK/ ANSWER:
B. Trip the Reactor and proceed to E-0, Reactor Trip or Safety Injectio i RO #55 l K/A #014000A203 OBJECTIVE #003B260D 1 REFERENCES: LER 95-01 !
OTO-SF-00003 j i
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. RO Test QUESTION #033 A complete loss of all circulating water pumps occurs from 100% steady state powe Assuming no operator actions and all systems function as designed, which ONE of the following corresponds to plant conditions 10 minutes after the loss of all circulating water pumps?
RCS TAVG S/G Pressures F 1092 psig F 1125 psig *F 1092 psig *F 1125 psig ANSWER: *F i125 psig '
RO #37 K/A #035010K301 OBJECTIVE #0110200E REFERENCES: Steam Table ES-0.1
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RO Test QUESTION #034
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A normal plant heatup is in progress per OTG-ZZ-00001 with the following plant conditions:
- RCS pressure 1835 psig
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- RCS pressurization rate 15 psig/ min
- RCS temperature 485'F
- RCS heat up rate 10 F/hr
- S/G pressure 575 psig If the current trend continues, which ONE of the following occur FIRST7 A. Main Steam Isolation Valves close.
B. Pressurizer PORV's open.
C. Low Pressurizer Pressure Safety Injection.
D. First group of steam dumps throttle ope i ANSWER:
A. Main Steam Isolation Valves close.
RO #21 SRO #25 K/A #013000K403 OBJECTIVE #0110520B REFERENCES: OTG-ZZ-00001," Plant Heatup Cold Shutdown to Hot Standby" Page 25 l
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j RO Test QUESTION #035 :
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Which ONE of the following will occur upon a decreasing Instrument Air System pressure due to a break at the condensate polishers?
A. The Lag air compressor loads at 117 psig; and all compressors " Fail-Safe" start at 115 psi l B. The Standby compressor loads at 117 psig; the Service Air Header Isolation valve i
KA-PV-11 " Fail-Safe" close at 110 psi ;
C. The Standby air compressor loads at 117 psig; and all compressors will be running at 110 psi l D. Service Air header isolation valve KA-PV-11 will close at 117 psig; the Lag air compressor loads at 115 psi ANSWER:
i A. The Lag air compressor loads at 117 psig; and all compressors " Fail-Safe" start at 115 i psi i RO #100 K/A #000065G10 OBJECTIVE #0110140D REFERENCES: OTO-KA-00001 l Callaway Bank l
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RO Test
- QUESTION #036 A surveillance to be performed on a piece of equipment having a contact reading of 50 R/hr in a room with a general area radiation reading of 125 mR/hr, would require entry intoa:
A. Danger High Radiation Area B. Caution High Radiation Area
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' C. Danger High Radiation Area Radiological Exclusion Area D. VeryHighRadiation Are ANSWER:
B. Caution High Radiation Area RO #3 SRO #3 K/A #19400lK103 OBJECTIVE #003A3IF3 REFERENCES: _ APA-ZZ-01000 Page 6 l
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RO Test QUESTION #037 Which ONE of the following statements describes the effect of a loss of DC control power
' to 4160 VAC breaker NB0112, NB01 MN FDR BKR FROM XNB0l? (Assume that the breaker is the only component affected by the loss of DC power.)
A. The breaker will fail in its current position and cannot be tripped or closed from the MC . B. The breaker will fail in its current position and can be tripped but not closed from the MC C. The breaker will trip and can be closed but not tripped from the MC D. The breaker will trip and cannot be tripped or closed from the MC ANSWER:
A. The breaker will fail in its current position and cannot be tripped or closed from the MC RO #52
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OBJECTIVE #0110060E REFERENCES: T61.0110.6 LP-#6 E-23NB12
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RO Test QUESTION #038 The Callaway Plant is operating at 60% Reactor Power, increasing at 3% per hour. MCB Annunciator 106A, "Cond Hotwell Lvl Lo Lo" alarms. The Lo Lo level condition is verified on MCB indicator AD-LI-l14.
Which one of the following is a required immediate action for this plant condition?
A. Run the remaining feed pump speed to the Hi Speed Stop to restore S/G level.
B. Start the Motor Driven Auxilianj Feedwater Pumps PALOl A and PALOl I C. Drive Control Rods to reduce Reactor Power to less than 2%.
D. Trip the Reactor and refer to E-0, Reactor Trip or Safety Injection.
ANSWER:
D. Trip the Reactor and refer to E-0, Reactor Trip or Safety Injection.
RO #30 K/A #056020G10 OBJECTIVE #0110220M REFERENCES: OTA-RL-RK106, Att. A i
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RO Test QUESTION #039
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During a loss of all AC while performing ECA-0.0, Loss of All A.C. NK11 battery discharge amps is at 300 amp Which ONE of the following is the MAXIMUM time that NK01 could be predicted to be i
Operable assuming the battery was fully charged initially?
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A. 2 hcurs
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B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
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C. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
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D. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ANSWER:
B. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> RO #67 SRO #65 K/A #000055K101 OBJECTIVE #003D220V REFERENCES: E21NK01 l
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RO Test QUESTION #040 i
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A Reactor Trip hasjust occurred. The followini, conditions are found while performing Step 3 ofE-0, Reactor Trip or Safety injection:
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. NB01 energized from Emergency Diesel NE-01
. NB02 deenergized (no lockout)
Which ONE of the following describes the required action and basis for that action?
A. Transition to ECA-0.0, Loss of all AC Power because E-0 assumes that Offsite Power is Availabl B. Attempt to restore power to NB02 while continuing with E-0 because it is desirable to have power to all AC Emergency buse C. Attempt to restore Off Site Power to BOTH NB buses because E-0 assumes that Off Site Power is Availabl D. Do no make attempts to restore NB02 because it will delay the operator action and i only one NB bus is assumed energized by E- I l
I ANSWE B. Attempt to restore power to NB02 while continuing with E-0 because it is desirable to have power to all AC Emergency buse ;
RO #99 SRO #99 K/A #000056K302 OBJECTIVE #003D040E REFERENCES: T61.003D.6 LP-#4
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RO Test QUESTION #041 A periodic load test is being performed on NE02, Standby Diesel Generator 'B' in accordance with OSP-NE-0001B. NE02 has been paralleled with 4160V Bus NB02 and is canying 6 MW of realload. A Main Steamline break occurs and containment pressure increases to 20 (twenty) psi Which ONE of the follow'mg describes the response of the Load Shedding Emergency Load Sequencing System (LSELS)?
A. The LOCA Sequencer starts the Containment Spray Pumps at Step 3 (Time 15 seconds).
B. The Shutdown Sequencer starts the 'A' Essential Service Water Pump at Step 5 .
l (Time 25 seconds). I C. The LOCA Sequencer starts the Safety Injection Pumps at Step 1 (Time 5 seconds).
D. The Shutdown Sequencer starts the Residual Heat Removal Pumps at Step 2 (Time 10 seconds).
ANSWER:
C. The LOCA Sequencer stans the Safety Injection Pumps at Step 1 (Time 5 seconds).
l RO #50 ,
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SRO #46 j l K/A #064000A307 !
l OBJECTIVE #0110510F 1 j REFERENCES: T61.0110.6 LP-#51 l
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J RO Test QUESTION #042 WHICH of the following red paths is MOST LIKELY to occur for a steam line break on a j
single S/G outside containment, resulting in a reactor trip and SI? (Assume that all l safeguards equipment functions as designed.)
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l A. Response to Inadequate Core Cooling (FR-C.1)
B. Response to Loss of Secondary Heat Sink (FR-H.1)
C. Response to Imminent Pressurized Thennal Shock Condition (FR-P.1)
D. Response to High Containment Pressure (FR-Z.1)
ANSWER:
C. Response to Imminent Pressurized Thennal Shock Condition (FR-P.1)
RO #70 l SRO #63 K/A #000040K101 OBJECTIVE #003D.?80A I REFERENCES: T61.003 !
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QUESTION #043 i
A plant cooldown is initiated following a reactor trip using the AUX FEED system and .
S/G PORV's. The CST level is initially at 87% (407,000 gal). ;
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Which ONE of the following is the tirne available until CST level decreases to the MODE '
3 Technical Specification limit with AUX feed flow at 300,000 lbm/hr. (8.345 lbm/ gal)
l A. 3.5 h B. 4.0 h C. 4.5 h I D. 5.0 h ANSWER: 1 A. 3.5 h l RO #34 SRO #31 i K/A #061000A104
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OBJECTIVE #0110250E REFERENCES: T/S 3.7. Tank Book TDB.001 l
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i QUESTION #044
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I Which ONE of the following events is required to be recorded in the RO Narrative Logs?
! A. Chemical addition to the condensate system.
1 B. Security Event due to Security System (SAS) malfunctio )
1 i i C. Annunciator switchyard carrier potential / tone loss, alarm l 4- 1
ANSWER:
i D. Unexpected ESFAS alarm on ESW syste i
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} PO #9 i SRO #8 i K/A #194001 A106 I OBJECTIVE #003A02B1
! REFERENCES: ODP-ZZ-00006, Section .
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j QUESTION #045 a
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- Conditions: 1
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- Reactor Power = 100% i
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- CCW Pump .'D' Running and 'B' in Standby j - A Lockout occurs on the Startup Transformer
Which one of the following describes the design response of the CCW System?
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A. 'D' CCW Pump continues to run and 'B' CCW Pump does not start.
i B, 'D' CCW Pump is shed and 'B' CCW Pump is started by the Shutdown Sequencer.
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C, ' 'D' CCW Pump continues to run and 'B' CCW Pump is started by the Shutdown i Sequence D. 'D' CCW Pump is shed and 'B' CCW Pump does not start.
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- RO #61 K/A #008010A301
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OBJECTIVE #0110100E
REFERENCES: T61.0110.6 LP-#10
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~ QUESTION #046 i
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l The plant is at 6 % power with feed system contruiin automatic on the bypass feed .l reg valves. Main turbine chest and shell warming are in progress. Steam header pressure ;
transmitter ABPT507 fails hig ,
Which ONE of the following describes the INITIAL plant response? i A. Steam dumps CLOSE, MFW Pump Speed INCREASES, Bypass FRVs OPEN :
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B. Steam dumps OPEN,. MFW Pump Speed UNCHANGED, Bypass FRVs CLOSE l C. Steam dumps CLOSE, MFW Pump Speed UNCHANGED, Bypass FRVs OPEN l
D. Steam dumps OPEN, MFW Pump Speed INCREASES, Bypass FRVs CLOSE - !
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D. Steam dumps OPEN, MFW Pump Speed INCREASES, Bypass FRVs CLOSE i
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K/A #059000K104 OBJECTIVE #0110230F REFERENCES: OTO-AB-00004 8756D37 S025 i
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.f QUESTION #047
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' The plant has experienced a large break RCS loss of coolant acciden !
Which ONE of the following must be reset to allow opening KAHV0029, Instrument Air ;
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Ctmt Isolation?
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4 A. CISA B, CISB
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C. SIS l
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i ANSWER:
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A. CISA
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! K/A #013000A201 j OBJECTIVE #003B480A
] REFERENCES: E-0, Reactor Trip / Safety Injection M22KA01
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RO Test QUESTION #048 Callaway Plant is in MODE 1,30% Reactor Power on a Chemistry hol Annunciator 70B,"RCP VIB/SYS ALERT" alarms. The Reactor Operator checks vibrations on RP312 and finds 'C' RCP shaft vibration indicating 15 mils and stead l
Which one of the following is the required actio A. Trip the Reactor, Trip 'C' RCP and go to E-0, Reactor Trip or S B. Continue to monitor vibration on the 'C' RC C. Trip the 'C' RCP and declare the Loop 3 RTD channelinoperabl D. Increase Component Cooling Water temperature to reduce 'C' RCP vibratio l l
ANSWER:
B. Continue to monitor vibration on the 'C' RC RO #18 SRO #20 K/A #003000G10 OBJECTIVE #003B150B REFERENCES: OTO-BB-00002 l
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QUESTION #049 l
A Reactor Operator (normally working a 12-hour shifl) has worked the following hours (excluding turnover) on the dates indicated:
i Date Hours Worked 2/13/94 0600 through 2000 2/14/94 0600 through 1900 2/15/94 0600 through 2200 2/16/94 0600 through 2000 2/17/94 0600 through 2400 Which one of the following lists the date on which this operator FIRST violated the overtime requirements of APA-ZZ-00905, Limitations of Callaway Plant Staff Working Hours?
l A. 2/13/94 B. 2/14/94 ;
C. 2/15/94 I D. 2/17/94 4 ANSWER:
B. 2/14/94
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RO #7 K/A #194001 A103 OBJECTIVE #003A390E REFERENCES: APA-ZZ-00905, Page 2 l
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RO Test QUESTION #050 Which ONE of the following sets of conditions will permit the Standby Diesel Generators to continue to run following an Emergency Start?
A. Lube Oil pressure 57 psig and Jacket Water temperature 192'F B. Crankcase pressure 8 psig and Engine speed 54 rpm C. Lube Oil pressure 57 psig and Crankcase pressure 8 psig D, Engine speed 541 rpm and Jacket Water temperature 192 F ANSWER:
D. Engine speed 541 rpm and Jacket Water temperature 192*F RO #49 K/A #064050G07 OBJECTIVE #0110030J REFERENCES: T61.0110.6 LP-#3 l
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i RO Test QUESTION #05)
A 30 gpm leak has developed on the charging line between BG-HCV-182 (CVCS CHG PMPS TO REGEN HX HCV) and the regenerative heat exchanger. When the Control Room isolates the leak and completes the applicable Off-Normal procedures, the reactor makeup flowpath will be via . and the reactor letdown flowpath will be via Choose ONE of the following to fill in the blank A. alternate charging; normalletdown B. alternate charging; excess letdown C. sealinjection; excessletdown D. sealinjection; normalletdown ANSWER:
C. sealinjection; excessletdown i
l RO #90 SRO #84 K/A #000022A101 OBJECTIVE #003B220B REFERENCES: OTO-BG-00002 OTO-BB-00003
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RO Test
QUESTION #052
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Given the following conditions: ;
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- RCS at NOP/NOT fo: 100% RTP, '
- PORV 456A has seat leakage to the PRT,
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Which ONE of the following is the approximate tailpipe temperature?
A. 212"F B. 228'F C. 248'F D. 258*F ANSWER:
D. 258*F RO #58 SRO #54 K/A #007000A201 1 OBJECTIVE #0070130B '
REFERENCES: Steam Table I
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RO Test QUESTION #053 The Callaway Plant is in MODE 3 at NOP and NOT. An earthquake ruptures the Condensate Storage Tank and causes a steam break on 'C' S/G. The following conditions exist:
SG A, B & D NR Level 45 %
SG C NR Level 10 %
SG A, B, & D Press 900 psig SG C Press 300 psig AFW Suction Press 4 psig Which one of the following describes the resulting flowpath of feedwater to the Steam Generators?
l A. 'B' ESW Pump to 'B' MDAFP to 'C' S/G l
l B. 'A' ESW Pump to 'A' MDAFP to 'B' S/G i
C. 'B' ESW Pump to 'B' MDAFP to 'B' S/G D. 'A' ESW Pump to 'A' MDAFP to 'D' S/G ANSWER: l l
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B. 'A' ESW Pump to 'A' MDAFP to 'B' S/G RO #35 SRO #32 K/A #061000A303 OBJECTIVE #0110250D REFERENCES: T61.011 OTA-RL-RK127A l
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QUESTION #054 t
i With the plant at 40% power which one of the below would be TRUE regarding operation !
of the ATWS Mitigation Actuation Circuitry (AMSAC)?
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A. If S/G Levels decrease to less than 5% on 2 of 3 AMSAC logic circuits, then a '!
Turbine Trip and MD AFAS, are actuated 25 seconds late l t
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B. If S/G Levels decrease to less than 5% on 1 of 2 AMSAC logic circuits, then a j Turbine Trip and MD AFAS, are actuated 232 seconds late C. If S/G Levels decrease to less than 14.8% on 2 of 3 AMS AC logic circuits, then a
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Turbine Trip and MD AFAS, are actuated 25 seconds late !
t D. If S/G Levels decrease to less than 14.8% on 1 of 2 AMSAC logic circuits, then a j Turbine Trip and MD AFAS, are actuated 232 seconds late .
i ANSWER: !
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'A If S/G Levels decrease to less than 5% on 2 of 3 AMSAC logic circuits then a
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l Turbine Trip and MD AFAS, are actuated 25 seconds late ;
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RO #22 .
SRO #30 !
K/A #001000GK04 i
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! QUESTION #055 i I I l .
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Liquid Radwaste Discharge Monitor (HDRE18) alarms on the RM-11 in dark blue j conditio Which ONE of the below could be the cause?
A. Loss of Sample Flow
,. B. Loss of Process Flow l
I l C. Monitor Purging l
l D. Channel No Pulses Received
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ANSWER:
D. ChannelNo Pulses Received
RO #97 SRO #68 i K/A #000059A201 OBJECTIVE #0110360B REFERENCES: OTN-SP-00002 OTA-SP-RM011 !
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I RO Test QUESTION #056 l
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Following a safety injection due to a RCS leak m containment, plant conditions are established that meet the SI termination criteria of E-1, Loss of Reactor or Secondary l Coolan Which ONE of the below is true regarding these plant conditions?
A. All safety related equipment is Operable as required by Technical Specification B. Reactor core decay heat is being removed by the steam generator C. Containment pressure is below the safety injection actuation setpoin D. Steam Generator pressure are approximately equal to RCS pressur i ANSWER:
B. Reactor core decay heat is being removed by the steam generator l l
RO #82 SRO #83 K/A #000009K324 ;
OBJECTIVE #003D090J
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REFERENCES: ES-1.1 SI Tennination l
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QUESTION #057 l
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Which ONE of the following valves fail open on a loss ofinstrument air?
A. Steam Generator Atmospheric Relief B. Main Feed Regulating Bypass Valves C. Main Feed Pump Recire Valve D. Heater Drain Pump Recirc Valve ANSWER:
D. Heater Drain Pump Recire Valve RO #64 SRO #56 K/A #078000K302 OBJECTIVE #003B330A REFERENCES: OTO-KA-00001 I
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QUESTION #058 l
l An automatic preaction sprinkler system " trouble" alarm would indicate: (
A. a deluge valve actuation
B. an alarm check valve operation )
C. a fire detector in alarm condition D. an open sprinkler head ,
ANSWER:
D. an open sprinkler head
K/A #086000A402 l OBJECTIVE #0110350C
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REFERENCES: T61.0110.6 LP-#35
- Callaway Bank i
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RO Test I QUESTION #059 i
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! The Callaway Plant is in MODE 3, NOT, NOP, performing a plant shutdown. Steam Generator levels are being maintained by the 'A' Main Feedwater Pump and the AFP l ESFAS BLOCK switches are in PERMI j l
While making preparations to open the Reactor Trip Breakers, the Main Feedwater Pump i Discharge pressure increases to 1980 psi l Which one of the following describes the immediate plant response? ,
A. 'A' MFP Trip, MDAFAS, SGBSIS B. 'A' MFP Trip, MDAFAS, TDAFAS C. 'A' MFP Trip, TDAFAS, SGBSIS D. MDAFAS, TDAFAS, SGBSIS ANSWER l
A. ' A' MFP Trip, MDAFAS, SGBSIS I
RO #32 K/A #059000K302 OBJECTIVE #0110230D REFERENCES: LER 96-02 OTO-SA-00001
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RO Test QUESTION #060 Which ONE of the follow'mg should be performed by any individual discovering a fire?
A. Notify Control Room, then use any available fire fighting equipment, then report to Fire Brigade Leade B. First attempt extinguishment using closest available eietinguisher, then call Control Room ifunsuccessfu C. First attempt extinguishment using closest available extinguisher then report to Fire Brigade Staging Are D. Notify Control Room, then use closest available extinguisher, if practical, then report ,
to Fire Brigade Leade I ANSWER:
D. Notify Control Room, then use closest available extinguisher, if practical, then report to Fire Brigade Leade RO #5 SRO #5 K/A #194001K116 OBJECTIVE #003A30F3 REFERENCES: EIP-ZZ-00226, Att. 2 l
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RO Test QUESTION #061
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Which ONE of the below shows the correct speed settings for the TD AFW pump?
IDLE SPEED NORMAL OPERATING SPEED OVERSPEED rpm 3850 rpm 4235 rpm rpm 3550 rpm 4435 rpm rpm 3850 rpm 4235 rpm rpm 3550 rpm 4435 rpm ANSWER: rpm 3850 pm 4235 rpm RO #38 SRO #45 K/A #039000A404 OBJECTIVE #0110250C REFERENCES: OSP-AL-P0002
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RO Test QUESTION #062 The plant is in MODE 3 at Normal operating pressure and temperature, Train 'A' COPS l has inadvertently been left ARMED for Cold Overpressure Protectio The selected pressurizer pressure channel, BBPT455 subsequently fails hig With no operator actions, which ONE of the following is TRUE7 A. PORV 455 initially opens, then closes when actual PZR Pressure decreases to <2185 l psi B. PORV 455 stays closed initially but will function as required for COPS.
! C. PORV 455 initially opens and stays open when actual PZR pressure decreases to
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<2185 psi l D. PORV 455 stays closed initially and PORV BLOCK VALVE (8000A) closes when ;
actual PZR pressure decreases to <2185 psi l ANSWER:
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A. PORV 455 initially opens, then closes when actual PZR Pressure decreases to <2185
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QUESTION #063 Which ONE of the following is the reason for depressurizing the Steam Generators at the maximum rate during ECA-0.0," Loss of All AC Power"?
A. To allow feeding S/G's from Diesel Driven Fire Water Pum B. To minimize RCS inventory los C. To enhance restoration of SG level from TD AFW Pum D. To prevent lifling PZR PORV ANSWER:
B. To minimize RCS inventory los RO #68 SRO #66 K/A #000055K302 OBJECTIVE #003D220S REFERENCES: T61.003 __ _. ___ . _ - _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ . _ _ _ . - . _ _ _ . _ _ . _ _ .
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QUESTION #064 :
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Given the followin ,
- Callaway is operating at 30% steady state reactor powe I&C technician receives permission to perform a calibration on Power Range l Channel N-4 l l - The I&C technician mistakenly pulls the control power fuses on N-42; then, realizing !
his mistake, he reinserts the fuses for N-42 and pulls the control power fuses for the '
correct channel, N-41, causing a reactor tri 'I l
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Which ONE (1) of the following describes the reason for the reactor trip?
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A. PR neutron flux low setpoint in B. Overpower Delta T tri C. PR neutron flux high setpoint tri l D. PR positive rate tri l
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ANSWER * !
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D. PR positive rate tri l l
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t RO #53 !
SRO #41 i K/A #012000K603 !
OBJECTIVE #0110270D i REFERENCES: T61.0110.6 LP-#27 !
T61.0110.6 LP-#28
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l QUESTION #065 !
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Which ONE of the below conditions would require containment coolers to be operated in SLOW speed? -l P
A. Service Water Temperature <60*F l i
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C. Emergency Diesel Supplying NB Bus i D Containment Temperature <80 F i
ANSWER:
A. Service Water Temperature <60*F !
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QUESTION #066 I
Prior to opening the Reactor Trip Breakers during a plant shutdown, the crew is directed j to reduce the inservice MFP speed to 3650 RPM in anticipation of a Feedwater Isolation !
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Using the attached graph, determine which one of the following is the minimum flowrate required to provide pump protection for this spee ;
l A.1500 Klbm/hr l
B.1750 Kibm/hr
- C. 2000 Klbm/hr i
D. 2250 Klbm/hr i ANSWER-C. 2000 Kibm/hr r
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- K/A #194001 A108 OBJECTIVE #003A040E l REFERENCES: OTN-AE-00001, Att. 4 1
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MAIN FEED PUMP MINIMUM FLO$ (LBMlHR VS. RPM)--MINIMUM FLOW AT i
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RO Test QUESTION #067 ES-1.1, SI Termination, Step 1 directs that the SI be reset using BOTH SI reset switches, ,
SBHS42A and SBHS43A.
Which ONE of the following describes the effect of operating only ONE switch at this step instead of both?
A. SI actuate light on SB069 would extinguish and automatic SI would reinitiate after 60 seconds.
B. SI actuate light on SB069 would extinguish since either switch resets both SI trains.
C. SI actuate light on SB069 would blink and automatic SI would reinitiate after 60 ;
seconds.
D. SI actuate light on SB069 would blink since reset switches are train specific.
ANSWER:
l D. SI actuate light on SB069 would blink since reset switches are train specifi l
t RO #93 K/A #000007K203 OBJECTIVE #0110270C REFERENCES: E-0 Step 4
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QUESTION #063 l
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l WHICH of the following groups of parameters read out at the Auxiliary Shutdown Panel? [
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l A. RCS WR pressure, S/G pressure, S/G level, containment pressure B. RCS Tavg, S/G pressure, S/G level, containment pressure !
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C. RCS hot leg temp, S/G level, TDAFWP flow, containment pressure j D. RCS cold leg temp, RCS hot leg temp, S/G level, S/G pressure l ANSWER: !
D. RCS cold leg temp, RCS hot leg temp, S/G level, S/G pressure
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K/A #000068K201 ;
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OBJECTIVE #0110480B REFERENCES: T61.011 .
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QUESTION #069 .
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The signal from the 'A' train SSPS to cause a reactor trip will:
A. open the 'A' reactor trip breaker and the 'A' reactor trip bypass breake l i
B. open the 'B' reactor trip breaker and the 'B' reactor trip bypass breake i
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l C. open the 'A' reactor trip breaker and the 'B' reactor trip bypass breake .
D. open the 'B' reactor trip breaker and the 'A' reactor trip bypass breake l r
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i l C. open the 'A' reactor trip breaker and the 'B' reactor trip bypass breake i l -
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! ' OBJECTIVE #0110270C l
REFERENCES: T61.0110.6 LP-#27 Callaway Bank
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RO Test QUESTION #070 During a reactor startup, the Intermediate Range Rod Stop is blocked when two of the four power range channels exceeds the setpoint.
A. Manually, C-5 B. Manually, P-10 C. Automatically, C-5 D. Automatically, P-10 ANSWER:
B. Manually, P-10 RO #26 K/A #015000K402 OBJECTIVE #003 A23A4 REFERENCES: OTG-ZZ-00003 OTO-SA-00001
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A spurious Si causes a plant trip and SI. Which one of the below actions is acceptable to j be performed while performing E-0 steps 1 through 147 ;
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l A. Securing NE01 due to ESW pump A trippin l t
B. Securing RHR Train 'A' due to RCS pressure at 223 ]
C. Stopping one CCP to minimize injection to RCS.
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D. Starting a SFP pump to restore Fuel Pool Coolin )
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A. Securing NE01 due to ESW pump A tripping.
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- SRO #6
- K/A #194001 A102
] OBJECTIVE #003 A29C4 REFERENCES: ODP-ZZ-00025
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RO Test QUESTION #072 Both trains of Essential Service Water (ESW) are placed into service to reduce containment temperature. Shortly after placing ESW into service, reactor power is noted to be slowly increasin Which ONE of the following is the probable cause of the power increase?
A. Change in containment air temperature affecting operation of the power range detector B. Change in main feedwater temperature due to flow variations in the S/G Blowdown syste C. Change in the CVCS letdown temperature causing deboration in the letdown demineralizer D. Change in main condenser vacuum causing increasing main steam flow through the main turbin ANSWER:
C. Change in the CVCS letdown temperature causing deboration in the letdown demineralizer RO #42 SRO #48 K/A #075000A401 OBJECTIVE #003A09A1 REFERENCES: OTN-EF-00001 OTN-EG-00001 i
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RO Test QUESTION #073 Given tlie following plant conditions:
. SAFETYINJECTION ACTUATED e - PZR PRESSURE 1800 PSIG Slowly Decreasing
- RCS TEMPERATURE 550*F Slowly Decreasing
- S/G NR LEVELS 1% Slowly Increasing
. PRT Pressure
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3 psig Stable
- S/G PRESSURE 1000 PSIG STABLE e ' PZR Level 28% INCREASING e .RM-11 GTRE31 & 32 Alarming
- CTMTTemperature 140*F Slowly Increasing
- CTMT Pressure 8 psig
- CTMTHumidity Increasing Which ONE of the following could be the cause of the above conditions?
A. Steam Generator Safety Valve failed ope B. Pressurizer PORV failed open.
- C. RCS Leak from a cold le D. Pressudzer steam space lea ANSWER:
D. Pressudzer steam space lea RO #81 SRO #82 K/A #000008A106 OBJECTIVE #003D030F REFERENCES: E-0 Reactor Trip / Safety Injection
RO Test QUESTION #074 The Callaway Plant is operating at 94% power with all four containment cooling fans running in fast speed.
'A simultaneous Safety Injection and loss of the normal power supply to NB01 occurs. All systems function as designed.
Which one of the following describes the response of the Containment Cooling fans?
A. Fans A and C start in FAST speed, B & D continue to run in FAST speed.
B. Fans A & C start in SLOW speed, fans B & D shia to SLOW speed.
C. Fans A & C start in FAST speed, fans B & D shift to SLOW speed.
D. Fans A & C start in SLOW speed, fans B & D continue to run in FAST speed.
ANSWER:
B. Fans A & C start in SLOW speed, fr.n- B & D shia to SLOW speed.
RO #29 SRO #28 K/A #022000A301 OBJECTIVE #0110400D REFERENCESi E21005 E21001
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QUESTION #075
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The plant is at 190'F and 350 psig with BOTH RHR trains in service in the cooldown mod j J
With both RHR suction reliefs and Pressurizer PORVs lined up for COPS, which ONE of j the following describes RCS overpressure control on increasing pressure? !
A. Pressurizer PORVs open sequentially first, then BOTH RHR suction reliefs would lif ,
B. Both RHR suction reliefs would lift first, then Pressurizer PORVs open sequentially.
C. Pressurizer PORVs and RHR suction reliefs would lift at the same tim ;
D. BOTH RHR suction reliefs would lift first, then both Pressurizer PORVs open simultaneousl '
ANSWER: j
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B. BOTH RHR suction reliefs would lift first, then Pressurizer PORVs oper. sequentiall i i
RO #62 l
K/A #005000A202 !
OBJECTIVE #003A210A I REFERENCES: Curve Book Fig 1 j OSP-BB-00003 !
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RO Test QUESTION #076 I
The plant is in the injection phase of Safety Injection due to a RCS LOCA. Containment Pressure has reached a maximum of 25 psi Which ONE of the following indicates ONLY loads being cooled by CCW7 i A. RHR Pumps, RHR Heat Exchangers, Sample systems B. Fuel Pool Reactor Coolant Pumps, Excess Letdown Heat Exchangers C. Containment Spray Pumps, Charging Pumps, Reactor Coolant Pumps D. Reactor Coolant Pumps, Charging Pumps, RHR Pumps ANSWER:
D. Reactor Coolant Pumps, Charging Pumps, RHR Pumps RO #77 l SRO #60 )
K/A #000026K302 OBJECTIVE #0110100C REFERENCES: M22EG01 !
E210010 l
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QUESTION #077 i
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The Callaway Plant is in a Reduced Inventory condition and has suffered a Loss of RHR Coolin j Which ONE of the following would cause a reduction in T-Boil (Time to Boil)?
A. Fewer Effective Full Power Days (EFPD)
B. Longer Time since Shutdown C. Lower Steam Generator Level D. Lower RCS Loop Level ANSWER:
D. Lower RCS Loop Level l
RO #91 1 SRO #85 K/A #000025G10 OBJECTIVE #003EE20B REFERENCES: OTN-BB-00002 T-Boil Calc-Theory j l
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_ . . RO Test QUESTION #078 i
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Which one of the following describes the effect of an LSELS Load Shed signal on Pressurizer Pressure Control?
A. Both Backup and Proportional Heaters are shed upon receipt of a Safety Injectio !
B. Only Backup Heaters are shed on an NB Bus undervoltage condition.
C. Both Backup and Proportional Heaters are shed on an NB Bus undervoltage condition.
D. Only Proportional Heaters are shed upon receipt of a Safety Injection.
ANSWER:
B. Only Backup Heaters are shed or. an NB Bus undervoltage conditio !
RO #51 K/A #010000K102 OBJECTIVE #0110510A REFERENCES: T61.0110.6 LP-#51 E210010
RO Test QUESTION #079 Plant startup is in progress with main turbine roll commencing and reactor power at 6%.
Power range N-44 is out of service due to a failed detector.
Which one of the below is UNBLOCKED under these conditions?
A. Intermediate Range High Flux Reactor Trip B. Pressurizer Low Pressure Reactor Trip C. Reactor Trip from Turbine Trip D. Pressurizer High Level Reactor Trip.
ANSWER:
A. Intermediate Range High Flux Reactor Trip RO #25 i SRO #26 K/A #015000A303 OBJECTIVE #003A24A2 REFERENCES: OTG-ZZ-00003 OTO-SA-00001 l
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RO Test QUESTION #080 l
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I Use the attached Figure 7-5 to answer the following questio !
l The plant is in MODE 3,557'F,2235 psig. Which one of the following is the amount of l water needed to reduce the RCS boron concentration from 1150 ppm to 1100 ppm?
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A. I167 gal.
B. I195 ga l C. 2688 gal.
D. 2752 gal.
ANSWER:
D. 2752 ga l RO #14 SRO #19 K/A #001010K521 OBJECTIVE #003AA40E REFERENCES: Plant Curve Book i
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FICURE 7-5 Rev. 001 i
REACTOR MAKEUP CONTROL SYSTEM NOMOCRAPHs BORON DILUTION
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M at 0% power = 515676 lbm * superin M at 100% power = 503624 lbm *
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RO Test QUESTION #081 The plant is in MODE 5 with the containment purge exhaust fan operating and containment purge supply off. The Containment Coordinator identifies a positive air flow :
condition from containment to outside atmosphere through the equipment hatch with the containment personnel hatch open.
Which ONE of the below actions should be performed for this condition?
A. Activate a Containment Purge Isolation B. Start either Fuel Bldg / Aux Bldg Emergency Exhaust train C. Activate a Control Room Ventilation Isolation D. Shift the Aux Building Normal Exhaust to FAST ANSWER:
D. Shift the Aux Building Normal Exhaust to FAST RO #45 SRO #43 K/A #029000K103 OBJECTIVE #003 A120B REFERENCES: OTN-GT-00001
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a RO Test QUESTION #082 Which one of the following will prevent outward control rod motion in both automatic and manual control?
A. Selected Turbine Impulse Pressure channelis reading 13% equivalent powe B. Two AT channels are within 3% of the overtemperature AT trip setpoin C. Control Bank D rods are positioned at 224 step D. One Power Range NI is reading 102%.
ANSWER:~
B. Two AT channels are within 3% of the overtemperature AT trip setpoin RO #16 K/A #001000K402 OBJECTIVE #0110260H REFERENCES: OTO-SA-00001, Table II
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RO Test QUESTION #083 A Hi Hi Radiation signal from SJ-RE-02, Steam Generator Blowdown System Radiation Monitor, will automatically close which ONE of the following valves?
A.' BM-HV-21, S/G 'C' Blowdown Nuclear Sampling System Upper Isolation Valv B. BM-FV-54, S/G Blowdown Discharge Pumps Discharge Flow Control Valv C.- BM-HV-6, S/G 'B' Blowdown Nuclear Sampling System Line Downstream Isolation Valv D. BM-HV-38, S/G 'D' Blowdown Nuclear Sampling System Lower Isolation Valv ANSWER:
C. BM-HV-6, S/G 'B' Blowdown Nuclear Sampling System Line Downstream Isolation Valve.
RO #56 SRO #47 K/A #073000K101 i OBJECTIVE #0110120D I REFERENCES: T61.0110.6 LP-#12 OTO-SA-00001
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QUESTION #084
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. Given the following plant conditions:
- Steam Break in AREA 5 j
. All MSIVs closed )
- 'A', .'B', and 'D' Steam Generator Pressures Stable l
- 'C' Steam Generator Pressure Decreasing a Performing actions of E-2, " Faulted Steam Generator Isolation"
. TD AFW pump is the only AFW pump available Which ONE of the following actions would be performed during completion of E-27 i
A. Close ABHV0006, 'C' Steam Supply to the TD AFW pum B. Open all S/G Common Sample Isolation Valves, BMHV0065 through 6 C. Reduce Aux Feedwater flow to 15,000 lbm/hr to each Steam Generato D. Close ABLV0007, Main Steam Low Point Drain SG 'C'.
ANSWER:
. B. Open all S/G Common Sample Isolation Valves, BMHV0065 through 68.
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K/A #000040E103 OBJECTIVE #003D150C REFERENCES: E-2
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RO Test QUESTION #085 Main Turbine exhaust pressure is 4" Hga and increasing at a rate of 0.5" Hga per minute.
Which of the following is the minimum amount of time that could elapse before an automatic low vacuum turbine trip occurs?
A. 5 minutes B. 7 minutes C. 9 minutes D.12 minutes ANSWER:
B. 7 minutes RO #73 SRO #64 K/A #000051 A202 OBJECTIVE #003BB90A REFERENCES: OTO-AD-00001 l
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QUESTION #086 Which one of the following describes the operation of the Main Turbine Steam Valves during Control Valve Chest Warming?
A. Main Stop Valve #2 Bypass is Open B. ' AllIntermediate Stop Valves are Shut C. Control Valves #1, #2, and #3 are Open D. All Main Stop Valves are Open ANSWER:
A. Main Stop Valve #2 Bypass is Open RO #59 K/A #045000A401 ;
OBJECTIVE #0110380E !
REFERENCES: T61.0110.6 LP-#38, Pg. 63 OTN-AC-00001 t
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RO Test QUESTION #087 Which one of the following is an entry condition for OTO-ZZ-00003, Loss of Shutdown Margin?
A. Mode 3, following Reactor Trip at 0950 and RCS Tavg 545'F at 111 B. Mode 2, with Reactor Power at 5% and Control Bank C at 35 steps i
C. Mode 3, with RCS temperature decrease of 100 F in 20 minutes with ECCS l operating in the Injection phase.
l D. Mode 5, with Shutdown Margin Calculation indicating the core net reactivity at-1100 pcm ANSWER:
B. Mode 2, with Reactor Power at 5% and Control Bank C at 35 step RO #80 SRO #59 K/A #000024G10 OBJECTIVE #003B610A REFERENCES: T61.003B.6 LP-#B-61 OTO-ZZ-00003 Plant Curve Book i
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QUESTION #088
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Which one of the following containment conditions would require the use of Adverse I
! Containment values when responding to a Large Break LOCA7 l
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A. Temperature had been 180 F and has decreased to 150 i B. Radiation had been 2.0E5 R/HR and has decreased to 500 R/HR l l
C. Pressure had been 30 psig and has decreased to 5 psi ,
I D. Recirculation Sump Level is greater than 138 inche ANSWER:
B. ' Radiation had been 2.0E5 R/HR and has decreased to 500 R/H RO #83 K/A #000011G11 OBJECTIVE #003D040R REFERENCES: T61.003D.6 LP-#4 i
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RO Test
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r f QUESTION #089 l
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- A steam generator tube leak causes a high radiation alarm on condenser air removal. Data i is taken to determine the steam generator leakrate.
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Time =0 Time =1 minute Time =2 minute j Reactor Power 99 99 99 i Tave 58 .3 588.3 j Charging Flowrate 100 100 100 Letdown Flowrate 80 80 80
- Total SealInjection Flowrate 33 33 33 j Pressurizer Level 55 % 54.8 % 54.6 %
i Total Seal LeakoffFlowrate 12 12 12
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c (Assume 1% Pressurizer Level = 60 gallons)
l Which ONE of the following is the approximate steam generator leakrate?
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. A. 5 gpm
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B. 10 gpm
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. C. 15 gpm
D. 20 gpm )
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l ANSWER:
D. 20 gpm
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- K/A #000037A212 OBJECTIVE #0110110P
. REFERENCES: T61.0110.6 LP-#11 OTO-BB-00001 l
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' RO Test QUESTION #090'
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A void exists in the reactor vessel during natural circulation cooldown. Which ONE of the following actions it used to collapse an excessive void, according to ES-0.3, " Natural Circulation Cooldown with Steam Voids"?
A. Decrease RCS temperature while maintaining RCS pressure constan B. Fill the Pressurizer solid and vent the reactor vessel hea !
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C. Increase RCS pressure using pressurizer heaters while maintaining pressur.12 er level.
D. Start an SI pump to increase RCS pressure while maintaining temperature constan '
ANSWER:
C. Increase RCli pressure using pressurizer heaters while maintaining pressurizer level.
RO #69 SRO #72 K/A #000074A101 OBJECTIVE #003D070K-REFERENCES: T61.003 ES 0.3 -
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l l QUESTION #091 t-
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. r The plant is in MODE 1 with all systems in normal except that I&C is performing l I,. corrective maintenance in the Rod Control Power Cabinet IBD. Group 1 of Control !
j Bank D is being energized from the DC Hold Bus l
Breaker PGl902, Motor Circuit Breaker to Rod Drive Motor-Generator SF01, is l inadvertently opened. All plant systems respond as designe j Which ONE of the below is true regarding power to the control rods? l
l A. Power continues to all control rod B. Power is intermpted to all control rod C. Power is intermpted to all rods except Control Bank D, Group D. Power continues to all rods except Control Bank D, Group ANSWER:
A. Power continues to all rod i
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RO #15 SRO #18 K/A #001000K202 OBJECTIVE #0110260G REFERENCES: T61.0110.6 LP-#26 i
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RO Test QUESTION #092 Which one of the following could be a direct result of a loss of Vital AC Instrument bus NNO37
. A. Charging Pump suction swaps to the RWST l B. Source Range Hi Flux Reactor Trip C. Intermediate Range High Flux Reactor Trip ,
i D. CVCS Letdown Isolation l
ANSWER:
l l l D. CVCS Letdown Isolation l
RO #79 SRO #67 K/A #000057A219 OBJECTIVE #003B450A REFERENCES: OTO-NN-00001 i
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RO Test QUESTION #093 i
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l A reactor trip has occurred and the operating crew is responding in accordance with !
ES-0.1, Reactor Trip Respons l
. Reactor trip and bypass breakers open
. NIS power is 1% and decreasing
. Bank D, Group 2 rods indicate 188 steps withdrawn. All other rods are fully inserted Which one of the following is TRUE for the above conditions?
A. An emergency boration of 450 ppm must be performed to ensure the minimum i shutdown margin is maintaine i B. An emergency boration of 150 ppm must be performed to limit fission gas release and maintain fuel pellet temperature within design limit C. No immediate action is required since the core is designed for these conditions, and the reactor has beer vanfied tripped by diverse indication D. A safety injection signal (SIS) must be actuated to maintain the reactor core in a safe shutdown conditio ANSWER:
l A. An emergency boration of 450 ppm must be performed to ensure the minimum shutdown margin is maintained.
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RO #65 SRO #58 i K/A #000005K301
! OBJECTIVE #003D060C l REFERENCES: ES- ___ - . . _ _ . - _ . . _ _ .. _ _ . . ~ . . _ - . . . ___ _ . . . _ _ . . . _ . _ . . _ .
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RO Test QUESTION #094 During a Reactor Startup, the Reactor Operator verifia one decade of overlap between the source and Intermediate Range Nuclear Instruments. This verification is defined as a(n) .
A. Source Check B. Analog Channel Operational Test C. ChannelCalibration D. ChannelCheck ANSWER:
D. Channel Check RO #11 SRO #10 j K/A #194001 Al13 1
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RO Test QUESTION #095
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Which ONE of the following components is manually (or automatically) isolated and remains isolated for a Faulted 'B' Steam Generator, but NOT necessarily for a 'B' Steam Generator Tube Rupture? (NOTE: Assume all equipment actuated as required.)
A. Main Steam Isolation Valve (AB-HV-17)
B. Main Feedwater Isolation Valve (AE-FV-40)
C. Auxiliary Feedwater Flow Control Valve (AL-HV-10) !
D. Main Steam Supply Valve to T/D AFW Pump (AB-V085)
ANSWER:
C. Auxiliary Feedwater Flow Control Valve (AL-HV-10) l RO #84 SRO #89 K/A #000038A132 OBJECTIVE #003D17NN REFERENCES: T61.003D.6 LP-#17 !
E-3, SGTR E-2, Faulted S/G Isolation
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RO Tcst QUESTION #096 Which one of the following sets of conditions should have resulted in a LoLo S/G Level Reactor Trip?
S/G NR CTMT LOOP TIME Level (%) Press (psig) AT(%) (sec) h 8 10 .5 23 180 ANSWER: RO #88 K/A #000054G09 OBJECTIVE #0110270D REFERENCES: T61.0110.6 LP-#27
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QUESTION #097 f
Which of the following should be performed if a 125VDC Vital Battery Charger fails?
A. Place the swing battery charger in service to replace the normal battery charger's functio B. Declare that train 125VDC Vital system inoperable and commence plant shutdow C. Align the maintenance supply to power that trains vital 120V AC instrument loads !'
directly.
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D. Align that trains inverter rectifier to perform the required battery charger functio l l
ANSWER: j l
A. Place the swing battery charger in service to replace the normal battery charger's l functio ;
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RO #92 K/A #000058A103 OBJECTIVE #0110060A REFERENCES: OTN-NK-00001 OTO-NK-00001
RO Test QUESTION #098 Which ONE of the following situations violates a requirement for containment integrity or containment closure?
A. A containment vent is performed with the plant operating at 100% powe B. The plant is in refueling mode with the refueling cavity flooded. Steam generator safeties have been removed; secondary manways are also removed. No fuel movement is in progres C. The plant is in refueling mode with fuel movement in progress. Containment Shutdown purge is initiate D. The plant is in hot standby. The "A" steam generator blowdown isolation valve BM-HV-1 is stuck ope ANSWER:
D. The plant is in hot standby. The "A" steam generator blowdown isolation valve BM-HV-1 is stuck ope RO #78 SRO #71 K/A #000069A202 OBJECTIVE #003E014A REFERENCES: TS 3. TS 3.6. TS 3.6.3
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RO Test QUESTION !;099
l l The Callaway Plant is operating at 30% power and it is necessary to secure the 'B' ,
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l Reactor Coolant Pump due to high vibration. After the RCP is tripped, the 'B' Loop AT and the other Loop AT's . (Assume unit load is held constant.)
A. Increases; Decrease I l B. Increases; Increase
C. Decreases; Decrease
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D. Decreases; Increase l
l RO #17 l K/A #003000A107 OBJECTIVE #01100901 REFERENCES: OTO-BB-00002 l
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RO Test QUESTION #100 l
l Which ONE of the following components has its air supply AUTOMATICALLY isolated if air pressure decreases to 108 psig?
A. Closed Cooling Water Temperature Controller
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B. First Stage RHDT Level Control Valves C. Main Feedwater Reg Valve Bypass Valves D. Auxilian Feedwater Pump Room Sump Pumps ANSWER:
D. Auxiliary Feedwater Pump Room Sump Pumps RO #46 SRO #49 K/A #079000K101 OBJECTIVE #0110140C REFERENCES: OTO-KA-00001 i
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CHIEF EXAMINER WRITTEN EXAM RESULTS ANALYSIS - CALLAWAY 2/24/97 l i
Scores: \
Each exam had 100 questions valued at one point eac )
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i SRO: High - 93; Low - 83; Average - 89 RO: High - 92; Low - 73; Average - 8 j Analysis:
I For the same questions, the same question numbers were used on either exam. The chief examiner concurs with the licensee's analysis attached. More than half of the applicants missed joint questions 3,4,55,64,84, and 90. More than half of the applicants also missed SRO question 3 All of the above questions were determined to be valid. No generic training or knowledge deficiencies were identified. Reasons for missing these questions appeared to be related to question difficulty and isolated training weaknesses. The licensee initiated appropriate actions to upgrade candidate specific knowledge and correct specific training weaknesses, j I
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1 CALLAWAY PLANT I
EXAMINATION COVER SHEET TRAINING DEPARTMENT l
COURSE TITLE: RO INITIAL LICENSE EXAMINATION DATE: 2/24/97
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NAME (Print): SCORING:
SIGNATURE: Points Possible: 100 Points Missed:
! Grade:
DmECDONS BLACK OLTI CORRECT ANSWERS f
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Review ofInitial NRC Written Examinations Callaway Plant - 2/24/97 The Reactor Operator and Senior Reactor Operator examinations were graded by F.X. Biermann and checked by R.A. Nelson. Both examinations were reviewed using the guidance contained in ES-403, Grading Site Specific Examinations at Power Reactors. This review is documented on the attached completed QA Checkoff Sheets, ES-403- This review revealed that seven (7) questions were missed by greater than 50% of the candidate Below is a summary of actions taken for each specific question:
Ouestion # Tonic Action 3 Plant Computer Alarm Operation item not specifically covered by lesson plan objectives. Submitted CA-#1031 to change objective WPA Tagging Requirements TFR written to emphasize method of tagging to be used on 4160V and above power block breakers when work is to be performed on downstream component Plant Security Event Question beyond objective oflesson i SRO #33 plan. TFR written to evaluate if actions ;
should be included into lesso j 55 Liquid Process Monitor Failure TFR written to include system operation ;
of Liquid Process Monitors. Stress differences between liquid and atmosphere monitor PR Nuclear Instrumentation Failure Question evaluated, valid and correc No action require l Cover with candidat EOP E-2 Actions Question evaluated, valid and correc No action require Cover with candidat '
90 EOP ES-0.3 Pressure Control Question evaluated, valid and correc No action require Cover with candidat In addition the subject questions above were examined for any common deficiencies regarding systems, types of operations involved, or safety system functions. No common deficiencies were noted.
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