IR 05000280/1993300
| ML20059C359 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/20/1993 |
| From: | Lawyer L, Edwin Lea NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059C351 | List: |
| References | |
| 50-280-93-300, 50-281-93-300, NUDOCS 9311010126 | |
| Download: ML20059C359 (200) | |
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%g# fg NUCLEAR REGULATORY COMMISSION Y o REGION 11 ..* * g 101 MARIETTA STREET, f ATLANTA, GEORGI A 30323 %q.....so Report Nos.: 50-280/93-300 and 50-281/93-300 Licensee: Virginia Electric and Power Company 5000 Dominion Boulevard Glen Allen, VA 23060 Docket Nos.: 50-280 and 50-281 License Nos.: 'DPR-32 and DPR-37 Facility Name: Surry 1 and 2 Examination Conducted: September 20 - 24, 1993 Inspector: k/
Edwin Lea 4 d c '
,o /p - /9- 93 Date Signed [
Accompanying Personnel: K. Faris, PNL
, ./ L. Sherfey, PN,L -
Approved by: C "@ D Lawerence L. Lawyer', Chief Date Signed Operator Licensing Section [/ ! J Operations Branch l Division of Reactor Safety SUMMARY Scope: NRC examiners conducted regular, announced operator licensing initial j examinations during the period of September 20-24, 1993. Examiners l administered examinations under the guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7. Three senior reactor operator (SRO) and four reactor operator (RO) candidates received written and operating examination Results: Candidate Pass / Fail: SR0 R0 Total Percent Pass 3 3 7 86% Fail 0 1 7 14% No violations or deviations were identifie ; I 9311010126 931021 PDR V ADOCK 05000280 pyn
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REPORT DETAILS Persons Contacted Licensee Employees
*A. Brown, Supervisor of Nuclear Training *D. Delamorton, Supervisor Nuclear Training *A. Friedman, Superintendent *L. Gardner, Senior Simulator Instructor * Henry, Shift Supervisor * Marshall, Lead License Class Instructor *H. McCallum, Supervisor Operator Training Other licensee employees contacted included instructors, engineers, operators, and office personne NRC Personnel *M. Branch, NRC, Senior Resident Inspector * Attended exit interview Discussion Results Initial licensing examinations were administered to three senior reactor operator (SRO) and four reactor operator (RO) candidates who had applied for licenses to operate the Surry Power Station. All
" three SR0 candidates passed each section of the examinations. Three of the four Ros passed all sections of the examinations. One R0 failed the simulator section of the examinatio Reference Material The licensee supplied all examination material requested within the allotted time period. The material provided was well organized. The examiners noted several instances in which the information in the lesson plan did not match information in technical specifications (TS) ; or procedures. The discrepancies in the materials' information were resolved during prep-wee ] Simulator Facility NRC examiners observed simulator activities during the prep week (September 7-10,1993) and the examination week (September 20-24, 1993). The licensee provided corporate and site simulator personnel ; to support examination activities. The support provided by the simulator personnel was beneficial in minimizing problems that could ; have occurred during actual simulator examinations and job performance ! measures (JPMs). The simulator and simulator personnel performed well during the actual examination i I
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Report Details 2 The examiners noted that the facility's simulater did not adequately model mid-loop evolutions associated with the Residual Heat Removal System (RHR). The JPMs associated with mid-loop activities had to be performed during static simulator conditions instead of the desired dynamic simulator conditions. The examiners and the licensee discussed the inability of the simulator to model mid-loop activitie The licensee staff stated that future simulator enhancement will include improving the simulator mid-loop modeling capability, Training and Qualification Effectiveness The examiners and the senior resident inspector observed a demonstration of the licensee's classroom simulator computer. The licensee demonstrated how the computer would be used to aid in the training of licensed and non-licensed personnel. The examiners and the senior resident inspector commended the licensee on its effort to incorporate the classroom simulator computer in training licensee personne Medical Record Review The examiners reviewed a random sample of licensed operators' medical files. The review was performed to determine if the licensee had notified the NRC of all applicable changes to licensed operators' medical conditions as specified in 10 CFR and the Examiners Standard The examiners determined, from the review of the' medical records, that all appropriate information concerning changes in the operators' medical history was provided to the NRC as required by 10 CFR and the Examiners Standard i
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3. Exit Interview At the conclusion of the site visit, the examiners met with representatives of the plant staff listed in paragraph one to discuss activities associated with the examination and inspection performe The licensee did not identify as proprietary any material provided to, or reviewed by the examiner Dissenting comments were not received from the license l j
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ENCLOSURE 2 SIMULATOR FACILITY REPORT Facility Licensee: Surry Power Station Facility Docket Nos.: 50-280 and 50-281 License Nos. DPR-32 and DPR-37 Operating Tests Administered On: September 21-23, 1993 This form is to be used only to report observations. These observations do - not constitute, in and of themselves, 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 that may be used in future evaluations. No licensee action is required solely in response to.these observation While conducting the simulator portion of the operating tests, the following , items were observed (if none, so state): i ITEM DESCRIPTION . r Mid-loop operation The simulator did not adequately model mid-loop operation / evolutions associated with RH : i L l l l
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. . w w September 27,1993 ENCIOSURE 3 Regional Administrator vocket No U. S. Nuclear Regulatory Commission 50-281 Region II License No DPR-32 101 Marietta Street, N. DPR-37 Suite 2900 Atlanta, Georgia 30323 Attention: Mr. Thomas A. Peebles Mr. Peebles:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 WRITTEN LICENSE EXAMINATION COMMENTS In accordance with NUREG-1021, Section ES-402, the following comments are submitted concerning the Reactor Operator and Senior Reactor Operator written examinations administered at Surry on September 20,199 OUESTION: SRO 093 The following conditions exist:
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Operator "A" is 25 years old and has received 21 Rem whole body lifetime exposure prior to 199 * 11is accumulated whole body dose through 09/20/1993 is 3500 mre * Ilis present quarterly whole body exposure is 100 mre * Ile is needed to perform a highly specialized job that would give him an additional 700 mre Which ONE (1) of the following is the MINIMUM action required before Operator "A" can go to work on this job? An extension of his quarterly limit approved by IIcalth Physic An extension of his yearly limit approved by the Station Manage An extension of his lifetime limit approved by the Vice President - Nuclear Operation Operator "A" must be denied access due to his exposure histor . ._ - .- - - . - - - _ -_ - - . - - . - - - .
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ANSWER: Reference: VPAP-2101, " Radiation Protection Plan", page 2 C0515fENTS: .
(a) is also a correct answer. All extensions will require IIealth Physics approval. See
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attached VPAP-2101, page 25. Recommend accepting both'(a) and (c) as correct answers.
, OUESTION: RO 049/SRO 038 Which ONE (1) of the following pieces of fuel handling equipment is in an OVERLOAD condition when lifting 500 pounds? Spent Fuel IIandling Tool ; RCC Change Tool New Fuel IIandling Tool Burnable Poison Rod Assembly Tool
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ANSWER: ' Reference: ND-92.5-LP-3, " Fuel Handling Tools," Objective E, page 1 CO31NTENTS: > l (d)is also a correct answer. The weight of the heaviest BPRA is 71 pounds. The BPRA tool is a hand crank tool. A load of 500 pounds on this tool would definitely be an , overload condition. Because the BPRA weighs less than a control rod, the BPRA tool ' should also be considered to be in an overload- condition when lifting 500 pound Recommend accepting both (b) and (d) as correct answer ;
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. w w OUESTION: RO 020/SRO 012 Which ONE (1) of he following is a DIRECT source of water to the Low Level Liquid Waste Ileader? Decon Building Resin Mix tank Turbine Building sump S/G Blowdown System Service Water System ANSWER: Reference: ND-92.4-LP-4, " Liquid Waste System," Objective B, page COMMENTS: (c) is also a correct answer. The Steam Generator can be released to the Low Level Liquid Waste Ileader by using the S/G Blowdown and Recire Transfer Systems. See the attached reference drawings for the system flowpaths. The Recire and Transfer System is a subsystem of the S/G Blowdown System. Therefore, answer (c) is also a correct answer. Recommend accepting both (a) and (c) as correct answer ;
QUESTION: SRO 027 Which ONE (1) of the following RVLIS readings indicates core voiding? The Full Range reading 95% with no RCPs runnin The Upper Range reading 50% with no RCPs runnm I l The Dynamic Range reading 35% with a single RCP running., The Full Range reading 25% with a single RCP runnin ANSWER: ! I Reference: ND-95.3-LP-54, "FR-I.3, Respond to Voids in Reactor Vessel," Objective E, page 3 i l
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(b) is also a correct answer. The answer on the answer key (c) was taken from item (2)
on top of page 30 on the attached reference. On page 29 of the same reference, item (1) states that an upper range level < 95% is indication of a void in the reactor vessel. A reading of 50% on the upper range as given in answer (b) is definitely an indication of core voiding. Recommend accepting both (b) and (c) as correct answer . OUESTION: RO 056/SRO 045 Which ONE (1) of the following is the reason charging and letdown are SECURED following a loss of Component Cooling Water (CCW)? With no CCW there is no cooling for letdown flow and it is too hot for the Non-Regenerative IIeat Exchanger; and with no letdown, there is no heating flow in the regenerative heat exchange With no CCW there is no cooling for charging pumps; and with no charging there is no cooling for letdown flow which becomes too hot for the Non-Regenerative Ileat Exchange With no CCW there is no cooling for letdown flow and it is too hot for the Ion Exchangers; and with no letdown, there is no heating Dow in the regenerative heat exchange . With no CCW there is no seal water for the charging pumps; and with no charging letdown must be secured to stabilize pressurizer leve ANSWER: l Reference: ND-88.5-LP-1, " Component Cooling Water," Objective F, page 2 I l COMMENTS:
(a) is also a correct answer. Without CC Dow, letdown temperature will increase to Reactor Coolant System cold leg temperature of 543 F. This temperature is greater than the design temperatures for the NRllX as indicated on the attached referenc Recommend accepting both (a) and (c) as the correct answer Very truly yours, /, n Wh . '
4 - M. R. psi II. F. McCallum Station ager Supervisor Operations Training Attachments
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cc: Mr. M. .W. Branch . NRC Senior Resident Inspector Surry Power Station ,.
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bc: Mr. L. M. Girvin - IN-2SE
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Mr. R. F. Saunders - IN-2SE Mr. D.' A. Christian - SPS Mr. J. H. McCarthy - SPS Dr. T. M. Williams - IN-2NE' ' Dr. A. II. Friedman - STC
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. -. _ . . . . ._ VPAP.2101 VIRGINIA asvislos 4
. POWER O rd PAGE 25 OF 86 i
. b. Administrative whole body quarterly dose levels for a worker may be extended up to either 1750 or 2750 mrem. Extensions to 1750 will require approval by a l I
Superintendent cognizant of worker duties, while extensions to 2750 will also
/F require Station Manager approval. All extensions will require acknowledgment by the affected worker and Health Physics approva LI c. Administranve whole body monthly dose limits for a Declared Pregnant Woman //
may be extended up to 50 mrem with approval by a Superintendent cognizant of the Declared Pregnant Woman's duties. Extensions will require acknowledgment by the affected Declared Pregnant Woman and approval by Health Physic d. Requests for extension of lifetime limits will require approval of both the Station Manager and either the Vice President-Nuclear Operations or the Assistant Vice President-Nuclear Operation e. More than one worker may be included within a single request, however, each worker must be individually listed. A reason justifying the extension will be required. Acceptable reasons include: unique qualifications of an individual, a higher individual dose will allow lower collective dose, and other qualified individuals with lower doses are not availabl f. Health Physics shall provide forms necessary for extending dose limits and provide - assistance for the process upon reques g. Health Physics shall provide summary reports of worker dose for use by Station management and supervision to assist in maintaining cognizance of worker dose for planning and exposure tracking. Reports may include dose estimates pending reading of TLDs for dose of record determination .
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VPAP-2101 VIRGINIA g REVISION 4
POWER w bq PAGE 21 OF 86 2. If a worker's dose exceeds the limits provided in this Section, the worker shall be advised of the dose receive . The following reports, as available, shall be provided to a worker upon request from the worker:
. Worker's annual exposure
. Worker's exposure during assignment at the Station
. Worker's exposure for a quarter as indicated, either known or estimated exposure 6.3.2 Administrative Dose Limits The purpose of establishing administrative dose limits is to minimize the pote itial for exceeding federal limits. Exceeding administrative dose limit values shall not be considered a violation of either 10 CFR 20 or Technical Specifications: however, exceeding administrative limits shall require a radiological incident investigatio Results of any such investigation shall be used to determine reporting applicability and shall become a part of Station records, a. Quarterly Administrative Exposure Limits 1. Whole Body 750 mrem per calendar quarter
~ Extremities 15,000 mrem per calendar quarter 3. Skin of Whole Body 5000 mrem per calendar quarter b. Yearly Adrninistrative Exposure Limit
. Whole Body 5000 mrem per calendar year c,J.ifetime Exposu e Limit j NOTE: The lifetime exposure limit app!!c ar.!y W he t individual's quarterly dose is to excecc 750 mrem per calendar quarte An individual may receive a lifetime cu nulative whole body dose equal to the lesser of:
. One tem per each year of age: or
. Five rem per year for each year after the eighteenth year of age (i.e.,5(N-18))
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- & w A w Note that actual venting time is dependent on RCS pressure, the panicular gas / vapor material composition, containment hydrogen concentration, and the size of the void. It is possible that venting may have to be terminated with voids remaining in the reactor vesse l l I During venting operations, if any RCPs stop, the venting should continue i until a vent termination criterion is exceeded while allowing natural circulation flow to be established. This minimizes the amount of gas l which could collect in the SG U-tube During a subsequent RCS depressurization, any voids existing in the SG U-tubes will expand, which may result in anomalous pzr level readings and degradation of SG heat transfer. (rk) 2 STEP 21: CIIECK RVLIS INDICATES VESSEL FUL The purpose of this step is to determine whether the venting has eliminated the void from the reactor vesse Following the vent operation, the RVLIS level indication would be checked to determine if the void has been eliminated. If the void still exists, the RCS is repressurized to the original pressure recorded in step 15 and another venting cycle is initiated by returning to step 1 Note that proper RVLIS indication for determining whether a full reactor vessel exists depends on the status of the RCP (1) If the vessel level indicates full with no RCPs running, a RVLIS upper range greater than 95% (the value corresponding to a full upper plenum level indication, including allowance for normal channel accuracy), the void has been successfully remove ND-95.3-LP 54 Page 29 Revision 3
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- w (2) With one RCP running, a Dynamic Range of greater than 42% is indication of a full reactor vesse . STEP 22: CIIECK PRZR LEVEL - STABL The purpose of this step is to stabilize pressurizer level prior to leaving this guidelin This step establishes a stable pressurizer level after the vent operation has been completed, prior to returning to the procedure and step in effec . STEP 23: RETURN TO PROCEDURE AND STEP IN EFFEC The purpose of this step is to direct the team to the proper guideline
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following successful completion of the steps in this guidelin Now that the guideline steps have been completed and venting has been completed, the team should continue plant recovery operations by returning to the guideline and step in effect at the time FR-I.3 was entere Summa ry FR-1.3, Response to Voids in Reactor Vessel, provides the operating team with guidance to address the presence of voids in the reactor vessel upper head area. There are numerous conditions which could allow for void formation in the vessel. Elimination of vessel voids using this guideline should not be attempted until a stable, subcooled RCS exists with SI terminate ND-95.3-LP-54 Page 30 Revision 3
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TABLE 41-2 CVCS SYSTEM AND COMPONENT DATA (Continued) Ooeratine Parameters Maximum Heatup Normal Purification (Desien) Shell-side 29,820 lbm/hr 59,640 lbm/hr 59,640 lbm/hr Flow Inlet Temperature 543.5*F 543.5'F 547'F Outlet Temperature 285* 300*F 340*F Tube-side 22,366 lbm/hr 52,185 lbm/hr 36,480 lbm/hr Flow Inlet Temperature 130.0*F 130.0*F 130.0*F Outlet Temperature 496*F 432*F 491*F NON-REGENERATIVE HEAT EXCHANGER Desien Data Shell-side Tube-side Component Cooling Water Boric Acid Solution Fluid Circulated 494,000 lbm/hr 59,700 lbm/hr Fluid Flow 150 psig 600 psig Pressure 250*F 400*F Temperature
.0005 hr-ft *F 2 .0003 hr-ft2 Fouling Factor Btu Btu Page 5 07-31-92 NCRODP-41-S
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i ENCLOSURE 4 NRC RESOLUTION OF FACILITY COMMENTS The post exam facility comments (attached) proposed the following revisions to the answer key: Question SRO 012 /R0 020 - Answers "A" or "C" be accepted as-correc . Question SRO 027 - Answers "B" or "C" be accepted as correc . Question SR0 038 / R0 049 - Answers "B" or "D" be accepted as correc . Question SR0 045 / R0 056 - Answers "A" or 'C" be accepted as correc . Question SR0 093 - Answer "A" or "C" be accepted as correc NRC made a complete impartial review of all questions that were shown to be incorrect during the initial grading of the examinations. This review , included the following: l Technical accuracy Question wording for clarity Statistical summary of all missed questions on all exams administered at Surry (both R0 and SRO) for question validity Sufficient KA value to support the question Facility Learning objective to support the question Validity of facility post examination comment The results of this review find all questions to be technically accurate and valid with sufficie'it importance rating to be maintained as a portion of the examination with the following exceptions and recommendations: Question SR0 012 / R0 020 - The question has two (2) correct' answers as written. Accept the facility comment and revise the answer key to show distractors "A" or "C" as correc . Question SR0 027 - As shown on ND - 93.4, " Inadequate Core Cooling Monitor System", H/T-3.7, the lowest valid reading on the RVLIS Summary Page is 60%. A 50% reading on the Upper Range as l specified in distractor "B" is not a valid reading. Distractor
"C" is the only correct answe . Question SR0 038 / R0 049 - Reference'ND-92.5-LP-3, " Fuel Handling Tools", states that the Burnable Poison Rod Assembly Tool is of similar design to the Thimble Plug Tool which employs a 1000 l scale. Distractor "B" is the only correct answe . -.
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/ Enclosure 4 2 Question SR0 045 / R0 056 - Prior to Charging being secured, letdown flow is cooled in the Regenerative Heat Exchanger to maintain temperature to less than the design temperature. Only after charging is secured for inventory control due to a loss of letdown does the possibility exist for temperature in the Non-regenerative Heat Exchanger to increase to greater than design temperature. Distractor "C" is the only correct answe . Question SR0 093 - The question has N0 correct answers as writte The question has been deleted form the examinatio P h
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N'T 21- .-. . - ~ ~ - ~~ ~ Nuclear Regulatory Commission Operator Licensing Examination l l This document is removed from ' Official Use Only category on .
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date of examinatie I NRC Official Use Only l I r_
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l l U. S. NUCLEAR REGULATORY COMMISSION ;
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SITE-SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION
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Name: Region: II l l Date: 09-20-93 Facility / Unit: Surry License Level: RO Reactor Type: W INSTRUCTIONS Use the answer sheets provided to document your answer Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours after the examination start All work done on this examination is my ow I have neither given nor received ai Applicant's Signature RESULTS Examination Value __100 Points Applicant's Score Points Applicant's Grade Percent __ -
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Page 2 i EEACTO,R.0PERATOR ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d ___ 016 a b c d 039 a b c d 017 a b c d 040 a b c d 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d
#4EACT0,R. 0PERATOR Page_ 3 ANSWER SHEET l Hultiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan a b c d 069 a b- c- d 047 a b c d 070 a b c d 048 a b c d 071 a b c d ' 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d 052 a b c d 075 a b c d 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d 081 a b c d 059 a b c d 082 a b c d 060 a b c d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b c d OSS a b c d 088 a b c d 066 a b c d 089 a b c d 067 a b c d 090 a b c d 068 a b c d 091 a b c d
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SIACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan a b c d 093 a b c d 094 a b c d 095 a b c d 096 a b c d l 097 a b c d 098 a b c d 099 a b c d 100 a b c d l l (********** END OF EXAMINATION **********) ,
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Page 5 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the administration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
, After the examination has been completed, you must 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 examination. This must be done after you complete the exam- ! ination, j i Restroom trips are to be limited and only one candidate at a time l may leave. You must avoid all contacts with anyone outside the i examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible repro- ' duction . Print your name in the blank provided in the upper right-hand corner of the examination cover shee . Fill in the date on the cover sheet of the examination (if necessary). Print your name in the upper right-hand corner of the first page of each section of your answer sheet . Before you turn in your examination, consecutively number each answer sheet, including any additional paces inserted when writing your answers on the examination question pag . The point value for each question is indicated in parentneses after the questio . Partial credit will NOT be give . If the intent of a question is unclear, ask questions of the examiner onl . When you are done and have turned in your examination, leave the examination area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke . _ ,
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. REACT 0,R OPERATOR Page 7 QUESTION: 001 (1.00)
The following plant conditions exist:
* The Reactor is at 100% powe , * A R00 CONTROL SYSTEM URGENT FAILURE alarm has just been received due to a problem in the Logic Cabinet for the Bank "C" rod ,
WHICH ONE (1) of the following will be the affect of this condition on the movement of Bank "D" rods? , ONLY Bank "0" OUTWARD rod motion is possible in BANK SELEC , ONLY Bank "0" INWARD rod motion is possible in BANK SELEC , N0 Bank "0" rod motion is possibl Movement of individual Bank "D" rods is possible in BANK SELEC ,
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QUESTION: 002 (1.00) The following plant conditions exist:
* The Reactor is at 100% powe * The Rod Control System is in AUTOMATI WHICH ONE (1) of the following is the response of the Rod Control System if Tavg is five (5) degrees F. greater than Tref? Rods step IN at forty (40) steps per minut Rods step OUT at forty (40) steps per minut Rods step IN at seventy two (72) steps per minut Rods step OUT at seventy two (72) steps per minut !
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.REACTOP<0PERATOR Page 8 QUESTION: 003 (1.00)
The following plant conditions exist:
* Unit 1 is in COLD SHUTDOW * RHR is in service with a solid Reactor Coolant Syste * RCS pressure is 350 psi * RCS temperature is 190 degrees F. and stabl *
Steam Generator "A" pressure is 5 psi * Reactor Coolant pump "A" was started at 0800 hours and secured at 0900 hours to repair an oil lea * Reactor Coolant pump "B" has been running for one 1
* At 0920 hours Maintenance personnel report that the(1) hour.to repairs Reactor Coolant pump "A" are complet I WHICH ONE (1)' of the following actions, per 1-0P-5.2.1, " Starting Any Reactor Coolant Pump , should be taken following a request from maintenance personnel to start RCP "A"? The RCP can be started immediately with the present condition The RCP can be started with the present conditions at 0930 hours i following the required 30 minute rest perio , The RCP can be started immediately after establishing a bubble in l the pressurize The RCP can be started after establishing a bubble in the pressurizer following the required 30 minute rest perio i i
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.REACTQR. OPERATOR QUESTION: 004 (1.00)
The following plant conditions exist:
* Operators are starting Reactor Coolant Pump * RCS temperature is 350 degrees * RCS pressure is 450 psi * The Bearing Lift Pump for "A" RCP has been running for one (1)
minut WHICH ONE (1) of the following conditions is the reason the WHITE (oil pressure) indicating light is NOT lit for the Bearing Lift Pump? Bearing Lift Pump discharge is directed to the upper shoes, Bearing Lift Pump discharge is directed to the lower shoe Bearing Lift pump discharge oil pressure is less than 500 psig, The Bearing Lift Pump has only been running for one (1) minut l l l l
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r j. . REACT 0P4 OPERATOR O O Page 10 QUESTION: 005 (1.00) The following conditions exist:
* RCS temperature is 280 degrees F as indicated by In-core Thermocouple * Pressurizer level indicates 30%. * RCS pressure is 350 psi * Train "A" Residual Heat Removal (RHR) is operating and all RCPs are OF *
Loops 1 and 2 cold leg temperatures are 275 degrees * Steam Generator secondary side water temperatures are 330 degrees WHICH ONE (1) of the following statements describes the anticipated RCS - pressure response and reason for that response when a Reactor Coolant Pump is started in loop "B"? l Decreases due to higher temperature loop water being cooled as it passes through the cooler core region.
! ' Decreases due to greater cooling flow through the Shutdown Cooling Heat Exchangers caused by the RCP pressure differential.
, Increases due to heating RCS fluid as it passes through the Steam Generators.
l ' I Increases due to reduced cooling flow through the Shutdown Cooling l Heat Exchangers caused by the RCP pressure differential.
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' QUESTION: 006 (1.00) WHICH ONE (1) of the following will occur if VCT level channel LT-lll2 fails HIGH7 Assume N0 operator action is taken.
l Actual VCT level will DECREASE until charging pump suction transfers L to the RWST.
I Actual VCT level will DECREASE until AUTOMATIC makeup to the VCT is l initiate LT-1115 will IMMEDIATELY divert letdown flow to the PD d.- Charging pump suction will IMMEDIATELY transfer to the RWST.
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. ,R.EACTO,R, OPERATOR Page 11 QUESTION: 007 (1.00)
WHICH ONE (1) of the following is a response of the Chemical and Volume . Control System (CVCS) to a Safety Injection System actuation?
. Loop Fill lines isolate, Excess letdown isolates, _ Charging Pump Recirc lines isolat ' Seal Water injection isolate ,
QUESTION: 008 (1.00) WHICH CANNOT be one BLO(1)CKED?of the following Safety Injection System Actuation Signals HIGH STEAM HEADER /LINE DIFFERENTIAL PRESSURE HI CLS LOW PRESSURIZER PRESSURE HIGH STEAM LINE FLOW WITH LOW Tavg QUESTION: 009 (1.00) WHICH ONE (1) of the following Engineered Safety Features System trip relays ENERGIZE to actuate? HI CLS subsystem only HI-HI CLS subsystem only HI CLS subsystem and RMT HI-HI CLS subsystem and RMT
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'"' . R.EACTO,R OPERATOR Page 12 QUESTION: 010 (1.00)
WHICH ONE (1) of the following will cause AUTOMATIC Control Rod WITHDRAWAL? Tavg/ Tref error signal increases from 0.5 to 1.3 degrees Pimp fails HIG NI Channel N44 fails LO NI Channel N44 fails HIG QUESTION: 011 (1.00) WHICH ONE (1) of the following describes the result of OVERCOMPENSATION of BOTH Intermediate Range Nuclear Instrumentation Detectors? The Source Range Nuclear Instruments will automatically energize at a higher power level than normal, Permissive P-6 will not automatically de-energize on a reactor shutdow The High Flux at Shutdown alarm will be automatically BLOCKE Permissive P-10 will not automatically de-energize on a reactor shutdow ;
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. REACTOR OPERATOR O O Page 13 QUESTION: 012 (1.00) )
'WHICH ONE (being indicated by the core exit thermocouples (CETCs)? '1) of the following is th capable of degrees degrees degrees H degrees j
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u I I QUESTION: 013 (1.00) j The following plant conditions exist: -
* Unit I was at 100% power when a HI-HI CLS signal was receive * ALL three (3) Containment Air Recirc fans were operating at the time of the CLS signa WHICH ONE (1) of the following describes the response of the Containment Air 1 Recirc fans to the CLS signal? l
, "A" and "B" fans trip on the HI HI CLS signal and sequence back on their Emergency buses. "C" fan trips and does not come back on unless started manually by operator "A","B" "C" fans trip on the HI HI CLS signal and sequence back on theil and Emergency buse "A","B", and "C" fans trip on the HI HI CLS signal and do NOT come back on unless started manually by operator ' "A" and "B" fans trip on the HI HI CLS signal and do NOT sequence back on their Emergency buses. "C" fan remains on unless stopped j
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Page 14 QUESTION: 014 (1.00) WHICH ONEt (1)he isolation of feedwater heater IA7of the following actions should be taken should a leak necessitate Reduce load 10% due to reduced condensate flo Reduce power 10% to prevent uneven FW flow and resultant quadrant power tilt abnormalitie Isolate feedwater heater IB to equalize FW flow and prevent quadrant power tilt abnormalitie Isolate feedwater heater IB to prevent excessive condensate flow that could damage the "B" side heater QUESTION: 015 (1.00) WHICH ONE (Unit 1 is operating at 100% power?1) of the following will AUTOMATICALLY trip both pumps when MFW pump lube oil pressure is 5 psi MFW pump suction header pressure is 70 psi MFW and Recirc flow is 3000 gp Inadvertent SI signal on "A" trai J
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, REACTOR, OPERATOR h Page 15 QUESTION: 016 (1.00)
WHICH ONE (1) of the following conditions will AUTOMATICALLY close Main Feedwater Pump P-1A Discharge Motor Operated (M0V) valve? Steam generator level is 80% in "A" Steam Generator, Steam generator level is 75% in "A" and "B" Steam Generator l Main Feedwater pump motor P-1A1 feeder breaker from 15A5 is OPE Main Feedwater pump P-1A suction pressure is 70 psi ) I
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QUESTION: 017 (1.00) l l WHICH ONE (1) of the following sources of Auxiliary Feedwater is the LEAST l desirable? l Firewater makeup supply l Unaffected unit's Auxiliary Feedwater supply CN-TK-3 Horizontal Emergency Makeup tank via booster pumps CN-TK-2 Normal Condensate Storage tank makeup i l l t l
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. REACTOR OPERATOR QUESTION: 018 (1.00) , The following plant conditions exist:
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Unit I has tripped due to a HI-HI CLS even * Offsite power was lost following the reactor trip and the #1 Diesel Generator has failed to star * Steam Generator levels are 21% narrow rang WHICH ONE (1) of the following is the status of the Auxiliary Feedwater System sixty (60) seconds after loss of offsite power? , BOTH Motor Driven pumps are 0FF and the Turbine Driven pump is supplying Auxiliary Feedwater, BOTH Motor Driven pumps are ON supplying Auxiliary Feedwater and the i Turbine Driven pump is 0F The 1-FW-3A Motor Driven and the Turbine Driven pumps are OFF and the 1-FW-3B Motor Driven pump is supplying Auxiliary Feedwate :) The 1-FW-3A Motor Driven pump is 0FF and the 1-FW-3B Motor Driven and Turbine Driven pumps are supplying Auxiliary Feedwate i l
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, REACT 0,R, OPERATOR Page 17 QUESTION: 019 (1.00)
The following condition exists:
* Unit I has just lost "H" Bu WHICH AuxiliaryONE Fee (1)dwater System (AFW)?of the following is the affect of this loss on the Unit 1 AFW automatically actuate AFW Turbine Driven pump EHC control is los AFW radiation monitoring power supply AUTOMATICALLY swaps to the Unit 2 suppl AFW radiation monitoring may be los QUESTION: 020 (1.00)
WHICH ONE (1) Header?of the following is a DIRECT source of water to the Low Level Liquid Waste
, Decon Building Resin Mix tank Turbine Building sump S/G Blowdown System Service Water System
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OPERATOR Page 18 QUESTION: 021 (1.00) WHICH ONE (1) of the following is the MAXIMUM administrative release rate for a Waste Gas Decay Tank (WGDT)? .0 cfm .0 cfm .0 cfm ! .0 cfm QUESTION: 022 (1.00) WHICH ONE (1) of the following sets of Waste Gas Tank parameters will place Unit 1 in a Technical Specification Limiting Condition for Operation (LCO)? pressure Hydrogen Oxygen psig 5.0% 1.5% psig 3.0% 2.5% psig 3.0% 2.0% psig 5.0% 1.0%
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. REACTOR OPERATOR Page 19 QUESTION: 023 (1.00) WHICH ONE (1) of the following is indicated by a BLINKING GREEN Fail / Reset light on a Victoreen Area Radiation Monitor? a. Detector has faile b. The check source is exposed to the detecto , c. Counts are below a minimum preset leve d. Monitor is ready to be rese ' -QUESTION: 024 (1.00) The following plant conditions exist:
* Loop 1 & 3 Tavg meters indicate 574 degrees * Loop 2 Tavg meter indicates offscale LO * Loop 1 & 3 Delta T meters indicate 100%. *
Loop 2 Delta T meter indicates 0%. WHICH ONE (1) of the following is the cause of these indications? a. Loop 2 Tcold failed LO b. Loop 2 Tcold failed HIG , c. Loop 2 Thot failed LO d. Loop 2 Thot failed HIG ; i
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, QUESTION: 025 (1.00) WHICH of the following ranges of RVLIS is/are valid ONLY with RCPs RUNNING? FULL & UPPER ranges - ONLY UPPER range ONLY FULL range ONLY DYNAMIC range t QUESTION: 026 (1.00) WHICH ONE (1) of the following is an INPUT to the Inadequate Core Cooling Monitor (ICCM)? Train "A" narrow range hot leg temperature Reactor Coolant Pump breaker status
, Auctioneered high CETC temperature , Train "B" narrow range RCS pressure
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Page 21 QUESTION: 027 (1.00) The following plant conditions exist:
* Unit I has tripped from 100% power due to a loss of coolant accident (LOCA) with a loss of offsite powe * Core Exit Thermocouples (CETCs) indicate 330 degrees * RCS pressure is 160 psi * RWST level is 20%. *
Emergency Diesel Generator #1 has failed to star * N0 flow is indicated for Low Head Safety Injection (LHSI) Pump "B".
WHICH ONE (1) of the following is the reason LHSI Pump "B" has N0 flow indication? RCS pressure is above LHSI Pump shutoff hea J 480v Emergency Bus has no powe SI Cross-Tie valves TV-SI-102A/B are OPE LHS1 Pump Containment Sump Suction Valves have failed to open automatically on low RWST leve l
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l QUESTION: 028 (1.00) The following conditions exist:
* Unit I has tripped from 100% power due to a large break Loss Of ; *
Coolant Accident (LOCA). ! SI has been MANUALLY actuate j conditions must be satisfied to CLOSE the "lA" SI , WHICHofthefollowin$ve? Accumulator Outlet va > l The key switch must be placed to NORMAL AND RCS pressure must be ' below 2000 psig with no SI signal presen The key switch must be placed to BYPASS with SI signals RESE The key switch must be placed to NORMAL AND RCS pressure must be below 2000 psig with the SI signal BLOCKE The key switch must be placed to BYPASS with the SI signal BLOCKE QUESTION: 029 (1.00) The following plant conditions exist:
* Pressurizer Master Controller M/A Station range is 1700 to 2500 psi * The potentiometer is a 0-10 turn po WHICH ONE (1) of the following is the pot setting corresponding to a desired RCS pressure of 2000 psig? .50 .75 .50 .25
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Page 23 .REACTQR OPERATOR QUESTION: 030 (1.00) ; WHICH ONE (1) of the following is used to generate the Pressurizer level setpoint? Auctioneered low Tcold Auctioneered high Thot Median Auctioneered Tavg Median Auctioneered Delta T l l QUESTION: 031 (1.00) The following plant conditions exist:
* Unit 1 is at 100% powe *
The Pressurizer Level LOWER Control Channel has failed LO WHICH ONE (1) of the following will result from this failure? Assume N0 operator actions, Actual Pressurizer level will increase due to minimum charging flow with letdown isolated and the Reactor will trip on HIGH Pressurizer leve ! Actual Pressurizer level will decrease due to reduced charging flow i with increased letdown and the Reactor will trip on LOW Pressurizer pressur Actual Pressurizer level will initially decrease until letdown isolates, then increase due to seal injection until the Reactor . trips on HIGH Pressurizer leve l
! Actual Pressurizer level will initially increase until PORVs open, l then decrease due to loss of RCS inventory until the Reactor trips j on LOW Pressurizer pressur !
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QUESTION: 032 (1.00) WHICH ONE (1) of the following will be the affect on the Anticipatory Mitigation Actuation Circuitry (AMSAC) from a loss of the 125 vdc 31ac Battery? AMSAC actuates resulting in a reactor tri AMSAC TROUBLE annunciator actuates due to initiation of battery backup power supply.
i ! There is no affect since the normal supply to AMSAC is from 181-3 l 480v MCC.
1 AMSAC will NOT actuate without 125 vdc Black Battery powe : l QUESTION: 033 (1.00) The following conditions exist: l
* Unit 1 is operating at 100% power.
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* Tavg is 574 degrees F.
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* Pressurizer pressure is 2200 psi *
Delta "I" is +1 * Delta T is 60 degrees F.
l l WHICH ONE (1) of the following plant parameter changes will DECREASE the OT Delta T Trip setpoint? l Delta "I" decreases to + Tavg increases to 578 degrees Delta T decreases to 55 degrees F with Tavg remaining constant, Pressurizer pressure increases to 2250 psig.
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. REACTOR OPERATOR Page 25 QUESTION: 034 (1.00)
The following plant conditions exist:
* Unit 1 is at 100% powe * Power has been lost to the Individual Rod Position Indication System (IRPI) Output Modul t!HICH ONE (d?1) of the following will result from this failure and HOW might it be diagnose An IRPI Dropped Rod Runback with N0 deviations on the four (4) PRNIS readings An IRPI Dropped Rod Runback in. conjunction with a NIS Dropped Rod signa An AUTO R0D WITHDRAWAL BLOCK signal and all Rod Bottom Lights are LI An AUTO ROD WITHDRAWAL BLOCK signal in conjunction with a NIS Dropped Rod signa !
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The following plant condition exists: Unit 1 power accension is in progres " Unit 1 is at 40% powe WHICH ONE (1) of the following Control Rods should be considered INOPERABLE per Technical Specifications 3.12.C., " Inoperable Control Rods"? a. A Bank "D" rod bottom light remains LIT even though actual control rod withdrawal occurre b. IRPI has been out of service for a Bank "D" rod for the past forty five (45) minute c. A Bank "C" rod has been misaligned from its Group Step Demand Position by 10 steps for the last sixty (60) minute A Bank "C" rod has been misaligned from its Group Step Demand Position by 28 steps for the last thirty (30) minute QUESTION: 036 (1.00) WHICH ONE (1) of the following cor.ditions may exist in order to CLOSE Containment Spray Discharge valve MOV-CS-101C? a. CS pump 1A breaker may be racked ou b. CS pump IB breaker may be racked ou c. CS pump 1A breaker may be in TES CS pump 1B breaker may be in TES .c
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Page 27 QUESTION: 037 (1.00) WHICH ONE (1) of the following variables does NOT cffect the Containment Spray to depressurize the containment in the event of a Design Systems capacity (DBA)? Basis Accident Containment temperature Containment pressure RWST temperature Component Cooling Water temperature QUESTION: 038 (1.00) WHICH ONE (1) of the following is the IMMEDIATE affect of a loss of the Station Instrument Air System on the Containment Purge to the Auxiliary Ventilation System Isolation Trip dampers? No affect, dampers are motor operate No affect, dampers have accumulator backup supply ai Dampers fail CLOSE Dampers fail OPEN.
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. REACTOR OPERATOR h Page 28 QUESTION: 039 (1.00) WHICH ONE (Pit per 0-AP-22.02, " Loss Of Spent Fuel Pit Level", when a high1) of the following i Spent Fuel volume of makeup is required? Component Cooling Water Service Water Fire Main Water Demin Water QUESTION: 040 (1.00) MHICH ONE (1) of the following conditions will CLOSE Main Feed Regulating valves? Following a reactor trip, Tavg is 557 degrees Safety Injection System actuation occur Pimp fails LO Steam Generator Level Channel 2 fails HIG a
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l . REACTOR OPERATOR Page 29 QUESTION: 041 (1.00) WHICH ONE (1) of the following is an indication of a "10 K positive ground" on a 125 VDC bus? Bus voltage indication is LOW, current indication is normal, and the control board ground indicating light is G Bus voltage indication is HIGH, current indication is LOW, and the control board ground indicating light is O Bus voltage indication is normal, current indication is normal, and the control board ground indicating light is 0F Bus voltage indication is LOW, current indication is LOW, and the control board ground indicating light is 0F QUESTION: 042 (1.00) WHICH ONE (1) of the following is the correct setting and reason for that setting on the speed droop for an Emergency Diesel running alone supplying power to its vital bus? Set at MINIMUM to enable the unit to change load without changing speed, Set at MINIMUM to enable the unit to decrease speed as load increase Set at MAXIMUM to enable the unit to change load without changing spee Set at MAXIMUM to enable the unit to decrease speed as load increase O
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* .REACTQR OPERATOR Page 30 QUESTION: 043 (1.00)
WHICH one (1) of the following valves AUTOMATICALLY closes when a HIGH alarm is received on Component Cooling Water Radiation Monitor, RM-CC-105? Thermal Barrier Heat Exchanger Isolation valve, TV-CC-120 Surge Tank Vent valve, HCV-CC-10 "CC RTN HDR A/B OTSD VLVs", TV-CC-10 "RCP's THERMAL BARRIER CC OUTLET FLOW INSIDE AND OUTSIDE TRIP VLV", TV-CC-14 i QUESTION: 044 (1.00) The following plant conditions exist: I
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Unit 2 is in INTERMEDIATE SHUTDOW ;
* RHR is in service, i * RCS pressure is 350 psig and INCREASIN * RCS temperature is 340 degrees * ALL system lineups are in a normal configuratio WHICH ONE 1) of the following will act FIRST to prevent overpressurizing the RHR System PORV PCV-1456 will OPEN at 365 psig as indicated on PT-45 RHR Suction Isolation Valve (M0V-1700) will automatically CLOSE at 460 psig as indicated by PT-140 RHR Suction Isolation Valve (MOV-1701) will automatically CLOSE at 460 psig as indicated by PT-145 RHR relief valve will relieve to the PRT at 600 psig.
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QUESTION: 045 (1.00) . The following plant conditions exist: 2
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Unit 2 is in INTERMEDIATE SHUTDOW * Operators are attempting to warm up the RHR Syste * RCS temperature is 340 degrees * RCS pressure is 350 psig
* RHR flow, other than recirculation flow, can NOT be established even though BOTH RHR pumps are runnin t!HICH ONE (1) of the following could result in these conditions? -
a. - A disconnected electrical lead to the RHR Heat Exchanger. Bypass Valve (HCV-1605) has caused the valve to fail CLOSE ' RHR Suction Isolation Valves (MOV-1700/1701) are interlocked CLOSE A broken air line to the RHR Heat Exchanger Outlet Valve (HrV-1758) has caused the valve to fail CLOSE A broken air line to RHR to CVCS Flow regulator Valve (HCV-1142) has caused the valve to fail CLOSE QUESTION: 046 (1.00) WHICH ONE (1) of the following discharges DIRECTLY to the Pressurizer Relief Tank (PRT)? RCP #2 Seal Leakoff Regen Heat Exchanger Relief Valve SI Accumulator Vent Valve Vessel Head Seal Leakoff
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e~ a~ . REACTOR OPERATOR Page 32 l QUESTION: 047 (1.00) WHICH ONE (level in the Component Cooling Water Surge Tank?1) of the following leaking comp INCREASING Excess Letdown Heat Exchanger Bearing Cooling Water Cooler Regenerative Heat Exchanger Component Cooling Water Heat Exchanger QUESTION: 048 (1.00) The following plant conditions exist:
* Unit I was operating at 100% power when it tripped due to a Loss Of f *
Coolant Accident (LOCA).
Containment pressure is now 16 psig and increasing i
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Hydrogen Recombiners are being placed in service while performing 1-E-1, Loss Of Reactor Or Secondary Coolant".
WHICH ONE (1) of the following indicates proper operation of the Hydrogen Recombiners during a loss Of Coolant Accident (LOCA)? Recombiners are placed in service when Containment hydrogen concentration is 5.0% Containment pressure decreases after Recombiners are placed in servic Hydrogen Recombiner power is immediately adjusted to 40 k Hydrogen Recombiner thermocouple temperature indicates 1175 degrees F. after power adjustments have been mad . REACTOR OPERATOR e
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QUESTION: 049 (1.00) WHICH ONE (1) of the following pieces of fuel handling equipment is in an DVERLOAD condition when lifting 500 pounds? Spent Fuel Handling Tool RCC Change Tool New Fuel Handling Tool Burnable Poison Rod Assembly Tool QUESTION: 050 (1.00) WHICH ONE (1) of the following conditions will ARM the Steam Dump System? Steam Dump Mode Selector switch placed in the "Tavg" position AND 2/4 Turbine 5 top valves are CLOSE Steam Dump Mode Selector switch placed in the "Tavg" position AND Turbine First Stage Impulse Pressure (PT-447) decreases 15% over a two (2) minute perio Steam Dump Control switch momentarily placed to the BYPASS-INTERLOCK position and Tavg increases to 557 degrees Steam Dump Control switch momentarily placed to the BYPASS-INTERLOCK position and a Turbine trip occur ! l I l
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. REACTOR OPERATOR Page 34 QUESTION: 051 (1.00) WHICH ONE (ESW)-pump 1-S (l') W-P-1 C7of the following conditions will TRIP Emergency Service Water Loss of 4 V0 VAC bus 1G HIGH oil pressure HIGH water temperature LOW ESW Pumphouse temperature l l
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QUESTION: 052 (1.00)
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WHICH ONE exists per 1- (1)AP-39.0, " Natural Circulation Of RCS"?of the following is an indication that natural ci i RCS Subcooling S/G pressure WR Cold Leg Temperature ,
;j degrees F 400 psig 440 degrees F , degrees F 500 psig 475 degrees F degrees F 600 psig 490 degrees F degrees F 700 psig 550 degrees F -
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QUESTION: 053 (1.00) The following plant conditions exist:
* Unit 1 is operating at 30% powe * The "A" Reactor Coolant Pump has just trippe WHICH ONE (1) of the following is the overall plant response? The reactor trips on a LOW RCS FLOW conditio Unit power is reduced to approximately 20% power (2/3 of original power level). Unit power remains the same with steam flow increasing on the other
, two steam generators.
' The reactor trips on HIGH steam generator level when "A" Steam generator level swells.
i l QUESTION: 054 (1.00) i WHICH ONE (1) of the followin 1 -AP-3.00, " Emergency Boration"g ? conditions requires EMERGENCY B0 RATION per 1- Shutdown Margin of 2.00% delta k/k while operating at 100% powe : One (1) shutdown bank rod remaining out of the core following a i reactor tri j Bank "D" Control Rods inserted below Insertion Limits i ppm boron concentration in the RCS while performing Refueling i Operations l l _
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l . REACTOR OPERATOR Page 36 QUESTION: 055 (1.00) .
WHICH ONE (1) of the following indicates the MAXIMUM Emergency Boration flow l rate that can be achieved to the RCS? Flow through MOV-1350 as indicated on FR-1-ll3.
, Flow through CH-228 as indicated on FR-1-11 Flow through MOV-1350 as indicated on 1-CH-FI-111 Flow through CH-228 as indicated on 1-CH-FI-ll1 , l l l QUESTION: 056 (1.00) WHICH ONE (1) of the following is the reason charging and letdown are SECURED following a loss of Component Cooling Water (CCW)? With no CCW there is no cooling for letdown flow and it is too hot i for the Non-Regenerative Heat Exchanger; and with no letdown, there is no heating flow in the regenerhtiv7 heat exchanger, With no CCW there is no cooling for charging pumps; and with no charging there is no cooling for letdown flow which becomes too hot for the Non-Regenerative Heat Exchange With no CCW there is no cooling for letdown flow and it is too hot for the Ion Exchangers; and with no letdown, there is no heating flow in the regenerative heat exchange With no CCW there is no seal water for the charging pumps; and with no charging letdown must be secured to stabilize pressurizer levt.1.
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a* Page 37 QUESTION: 057 (1.00) ! WHICH ONE (1) of the following requires an IMMEDIATE trip of all operating RCPs? Component Cooling Thermal Barrier Return valve TV-CC-120A fails CLOSE Component Cooling Water is lost to all Reactor Coolant Pump motor Seal Injection is lost to all RCP RCP Bearing temperatures indicate 170 degrees QUESTION: 058 (1.00) The following plant conditions exist:
* Unit 1 is operating at 100% powe *
Pressurizer spray valve PCV 455B is stuck OPE * RCS pressure is decreasing RAPIDL * All efforts to close PCV 4558 have faile WHICH ONE open spray va (1)lve?of the following actions must be taken to mitigate the stuck Trip RCP "B", then trip the reacto Trip RCP "C", then trip the reacto Trip the reactor, then trip RCP "B". Trip the reactor, then trip RCP "C".
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! l QUESTION: 059 (1.00) WHICH ONE (1) of the following is the reason the Design Basis Accident (UFSAR) for a Main Steam Line break is analyzed for a Hot Shutdown Condition (HSD)? Emergency Core Cooling System requirements are less restrictive at HSD.
l Steam Generator temperatures are lower at HSD.
1 Steam Generator inventories are greater at HS Minimum Shutdown Margin requirements are less restrictive at HS QUESTION: 060 (1.00) l WHICH one (1) of the following means of communication is available following a l l complete loss of ALL AC power to the site for greater than one (1) eight hour l l shift? J 1 Station radio I Plant PBX
! Gai-tronics ! Sound powered phones
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.REACTQR OPERATOR - h Page 39 QUESTION: 061 (1.00)
WHICH ONE (1) of the following component failures determines the severity of an accident resulting from the loss of ALL AC power? Pressurizer PORVs RCP seals Steam Generator safety valves RCP thermal barrier heat exchangers QUESTION: 062 (1.00) WHICH ONE (1) of the following MAJOR fires does NOT require a MANUAL reactor trip? Turbine Building fire Auxiliary Building fire Mechanical Equipment Room fire Main Switchyard fire .
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QUESTION: 063 (1.00) VHICH 0NE (1) of the following is the initial Control Room oxygen level which , requiras Operations personnel to INITIATE use of Self Contained Breathing ' Apparatuses (SCBA)? .5%
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QUESTION: 064 (1.00) The following plant conditions exist:
* Both Units are being shutdown due to a fire in the Main Control Roo * The Main Turbine has failed to trip using the MANUAL pushbutton * The Main Steam Trip Valves (MSTVs) will A0T close from the main control boar !
WHICH ONE (1) of the following alternate methods should be used to TRIP the Main Turbine per 0-FCA-1.00, Limiting MCR Fire"? Open the Main Generator Output breaker l Dis >atch an operator to trip the Main Turbine locally at the tur.)in I Close the MSTVs using FIRE EMERG CLOSE switch on APP R Panel in ESG Dispatch an operator to trip the Auto Stop 011 pump locally at the ; pum I l I
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dIIb w l ,REACTQROPERATOR Page 41 l l l l QUESTION: 065 (1.00) The following conditions exist:
* Unit 2 has experienced a small break Loss of Coolant Accident (LOCA).
- Voids are present in the reactor vesse * Operators are preparing to vent the reactor vesse * Containment pressure is 5.0 psi *
Containment temperature is 230 degrees * Containment hydrogen concentration is 2.5%.
* RCS pressure is 800 psi *
CETCs indicate 522 degrees WHICH ONE (1) of the following is the correct venting time per 1-FR-I.3, ; l " Response To Voids In Reactor Vessel"? (Attachments 1, 2 and 3 are attached.) .5 min .1 min ; 1 .6 min .1 min I l I I
b ~ O w Page 42 .REACTQR OPERATOR QUESTION: 066 (1.00) The following conditions exist:
* Unit I has tripped from 100% power due to a large break Loss of Coolant Accident (LOCA).
- Pressurizer pressure is 450 psi * Thot is 430 degrees * Tcold is 400 degrees *
Tavg is 415 degrees * CETCs indicate 435 degrees WHICH ONE (1) of the following is the RCS Subcooling Margin? degrees degrees degrees degrees QUESTION: 067 (1.00) WHICH ONE (1) of the following methods should be used to collapse a void in the RCS per 1-FR-I.3, " Response To Voids In Reactor Vessel"? Decrease RCS temperature by dumping steam while maintaining RCS pressure constan Increase RCS pressure using pressurizer heaters while maintaining pressurizer level constan Start a High Head Safety Injection (HHSI) pump and increase RCS pressure while keeping temperature constan Start all Charging pumps, fill pressurizer solid, then vent the reactor vessel hea , REACTOR OPERATOR Page.43
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QUESTION: 068 (1.00) The following conditions exists:
* Reactor is 30% increasing at 3% per hou *
Control rods. are in AUTOMTI * Bank "D" rods are at 110 steps and stepping out with N0 demand signa WHICH ONE (1) of the following actions are required to be performed FIRST per AP-1.00, " Rod Control System Malfunction"? Place the BANK SEL switch to the CONTROL BANK D position, Place the BANK SEL switch to the MAN positior,. TRIP the reactor and enter 1-E-0, " Reactor Trip or Safety injection". Enter 1-FR-S.1, " Response To Nuclear Power Generation /ATWS", to ensure the reactor is TRIPPE I i l
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QUESTION: 069 (1.00) The following conditions exist:
* Unit I was o erating at 100Y. power when an automatic turbine runback was initiate * The Unit is currently operating at 707. power following recognition that a Bank "C" control rod has dropped into the cor * The rod has been in the core for twenty (20) hour WHICH ONE (1) of the following is the proper method of recovery and the reason this method is necessary? Reactor power must be held constant below 75'/. while the rod is withdrawn at two (2) steps per hour to prevent Xenon oscillation , Reactor power is increased to 100Y while the rod is withdrawn at two l (2) steps per hour to prevent Xenon oscillation Reactor power must be held constant below 75Y. while the rod is withdrawn at ten (10) steps per hour to prevent rapid changes in local power densities that could cause DN Reactor power is increased to 100Y. while the rod is withdrawn at ten (10) steps per hour to prevent rapid changes in local power densities that could cause DN .
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j% O w .REACTQR OPERATOR Page 45 QUESTION: 070 (1.00) of the following indications distinguishes between an OPEN WHICil ONEP(1)ORV and an OPEN Pressurizer Code Safety valve while operating at Pressurizer 100% power? PRT HIGH TEMPERATURE alarm Pressurizer Code Safety valve position indicator Acoustic monitor Tail pipe temperature . l
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l QUESTION: 071 (1.00) l
!!HICH ONE (1) of the following is the Reactor Coolant Pump Trip Criteria for a small break Loss of Coolant Accident (LOCA)? HHSI flow indicated to cold legs AND RCS Subcooling less than 30 degrees ALL Chargin pumps operating as indicated by pump amps and RCS Subcooling ess than 30 degrees HiiSI flow indicated to cold legs AND RCS pressure less than 1350 psi ALL Charging pumps operating as indicated by pump amps and RCS pressure less than 1350 psi . . - - .- - _ _ _ . - ___ _ .REACTQR OPERATOR ~O O Page 46 -
t QUESTION: 072 (1.00) WHICH ONE (1)(of the following provides MAXIMUM cooling to the core for the12) hours during first twelve
' (LOCA)? Reflux boiling Break flow cooling , Natural Circulation Fallback cooling QUESTION: 073 (1.00)
WHICH ONE (1) of the following is the reason for establishing hot leg recirculation following a large break LOCA? To quench steam in the hot legs and to prevent formation of stratification layers in the cor To quench steam in the hot legs and to prevent boron precipitation.
. To quench steam in the core and to prevent formation of stratification layers in the cor To quench steam in the core and to prevent boron precipitatio ,
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O w-Page 47 l QUESTION: 074 (1.00) WHICH of the following parameters affects vapor entrainment in the RHR suction piping? RHR flow rate AND RCS level RHR flow rate AND RCS pressure I Number of RHR pumps running AND RCS pressure Number of RHR pumps running AND RCS level QUESTION: 075 (1.00) The following conditions exist:
*
Unit I has been shutdown for forty (40) hours from an extended operating run at 100% powe * Loop Stop valves are CLOSE * RCS level is at mid-loo * Pressurizer PORVs are OPE * RCS temperature is 140 degrees WHICH expectedONE fo (llowing a complete loss of RHR cooling?1) AssumeofN0the following is the MINIMUM time b alternate l cooling method is established. (1-AP-27.00," Loss of Decay Heat Removal ! Capability", Attachment 4 is attached) l .5 hours .5 hours .0 hours .5 hours l l l l l ,
_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - O w Page 48 f.REACTOROPERATOR
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_ QUESTION: 076 (1.00) WHICH ONE (1) of the following is the MAXIMUM time allowed to establish conditions which will assure acceptable" consequences following an Anticipated Transient Without Trip (ATWT) Event? Turbine is tripped within ten (10) seconds and Auxiliary Feedwater is established within ninety (90) second Turbine is tripped within thirty (30) seconds and Auxiliary . Feedwater is established within sixty (60) second j Turbine is tripped within sixty (60) seconds and Auxiliary Feedwater ) is established within sixty (60) second : i Turbine is tripped within ninety (90) seconds and Auxiliary l
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Feedwater is established within ninety (90) second l l l QUESTION: 077 (1.00) The following conditions exist:
* A eam Generator Tube leak has been confirmed in Steam Generator *
Steam Generator "A" pressure is 530 ps * Steam Generator "B" pressure is 750 ps l
*
Steam Generator "C" pressure is 530 ps WHICH ONE (1) of the following is the MAXIMUM RCS temperature that should be maintained in the RCS following the RCS cooldown to maintain a 50 degrees F subcooling margin after RCS depressurization? degrees degrees degrees degrees . - . . -_- . REACTOR OPERATO O' O Page 49'
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i I QUESTION: 078 (1.00) The following conditions exist:
*
The Reactor has been MANUALLY tripped following a rupture in the "B" Steam Generato * SI has automatically actuate !
* All RCPs are secure * RCS pressure is 1000 psi *
RCS is 60 degrees F subcoole :
*
Pressurizer level indicates 0%. l RVLIS Full Range indicate 85%.
' * ~*
Operators desire to restart "C" RCP per 1-E-3, " Steam Generator Tube Rupture".
WHICH ONE (1) of the following actions must be performed prior to resuming RCP
"C" operation? RCS pressure must be increased to collapse voids in the Reactor Vesse , RCP "A" must be started first to establish Pressurizer Spray flo RCS subcooling must be increased to within allowable RCP restart limit Pressurizer level must be increased to within allowable RCP restart limit QUESTION: 079 (1.00)
WHICH ONE (1) of the following differentiates bettan a unisolable feedline break and an unisolable steam line break of the same size? RCS heat removal would be greater from the steam line brea Containment pressure would be greater for a feedline break, Containment radiation levels would be higher from a steam line brea RCS depressurization would be greater from a feedline brea _ _ _ _ __
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iREACTOR OPERATOR Page 50-
. ( b LQUESTION:080 (1.00) :The following plant. conditions exist: j * :Ur.it ~1 has been manually tripped due to a loss of ALL Feedwate * . Operators are initiating step 3 of 1-FR-H.1, " Response To Loss Of ,
Secondary Heat Sink" and are tripping ALL RCP .j
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;WHICH ONE (1)this step in the procedure?-of the following is the basis for tripping all Reactor Coo
--.Pamps as.per To reduce decay heat removal in the steam generator . To reduce Reactor Coolant System pressure, To reduce thermal stress to the steam generator To reduce heat input into the. Reactor Coolant Syste ;
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' QUESTION: 081 (1.00) ,
.The following plant conditions exist: * Unit 2 has tripped from 100% power when ALL Feedwater is los * The "3B" Auxiliary Feedwater pump can.be returned to servic * ALL Steam Genarators indicate 5% NR leve '
WHICH ONE (1) of the following describes HOW feedwater flow should be re-established to the Steam Generators per 1-E-1, " Loss Of Reactor Or Seconda y Coolant"? Initiate feedwater ONLY to "B" Steam Generator at a maximum flow l rate of 350 gp Initiate feedwater to "A", "B", and "C" Steam Generators at a .' maximum flow rate of 350 gp Initiate feedwater DNLY to "B" Steam Generator at the maximum flow-rate possible (700 gpm). Initiate feedwater to "A" "B", and "C" Steam Generators at the maximum flow rate possible (70') gpm).
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. REACTOR OPERATOR h Page 51 QUESTION: 082 (1.00) WHICH ONE (1) of the following is an indication of the loss of lA DC Bus? #3 EDG automatically start MS-S0V-102B has failed OPE RX Trip Breaker "B" and Bypass Breaker "A" indicating lights are 0F Annunciator Panels A-E de-energiz QUESTION: 083 (1.00) WHICH ONE (1) of the following is an Operator Action following a HIGH alarm trip of Process Vent Radiation Monitor Rl-GW-101/102? Stop Containment Vacuum (CV) pumps to prevent blowing off discharge hos Stop Containment Vacuum (CV) pumps to prevent collapsing suction hoses, Place filter selector switches for areas tripped to FILTER position to provide filtered releas ' Place filter selector switches for areas tripped to CLOSE to isolate all release path . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ i I .REACTQR OPERATOR e
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QUESTION: 084 (1.00) WHICH ONE (1) of the following is the affect of the failure of the Victoreen Sample pump on the radiation monitoring capability for the Ventilation Vent l Stack flow? l None, redundant sampling is provided by the Kaman sample pump under all condition None, redundant sampling is provided by the HP Accountability Sample Rig pump under all conditions.
, The Kaman sample pump provides redundant sampling provided stack flow is reduced to less than S4,000 CFM and the HP Accountability Sample Rig pump is secure j 1 The HP Accountability Sample Rig pump provides redundant sampling l for all stack flow rates and the Kaman sample pump is secure ) l l l l QUESTION: 085 (1.00) The following conditions exist:
* Unit 1 is performing Refueling Operation * Fuel movement is in progres * Refueling Cavity level is DECREASING rapidl kHICH ONE Refueling Cav(1)ity ?of the following is the FIRST choice of makeup water to the Charging pump with suction from the VC Diesel Fire pump with suction from the FP-TK-1 HHSI pump with suction from the RWS LHSI pump with suction from the RWST.
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. REACTOR OPERATOR a* O Page 53 QUESTION: 086 (1.00) WHICHONE(1)tepower?of loss of offsi the following loads is the FIRST to sequence ON following a Pressurizer heater loads Outside Recirc Spray pumps 1-RS-P-2A and 2B AFW pumps 1-FW-P-3A and 3B Filter Exhaust fans 1-VS-F-58A and 58B QUESTION: 087 (1.00) WHICH ONE (1)le loss of Instrument Air?of the following is the reason ALL RCPs are TRIPPED following a non-recoverab < To prevent overpressure of Main Steam System if Main Steam Trip valves clos To reduce decay heat removal as steam generators overfee Component Cooling Water will be lost to the RCP Seal Injection will be lost to the RCP . __ _. . . . .
. REACTOR OPERATOR Page 54 QUESTION: 088 (1.00)
WHICH ONE (1) of the following designates an IMMEDIATE OPERATOR ACTION statement in an Emergency Procedure? A star precedes the ste An asterisk precedes the ste Step number is in bracket Step number is BOL QUESTION: 089 (1.00) WHICH ONE (1) of the following individuals by title is the MINIMUM authorization that must be obtained before Containment entry during subatmospheric conditions? Station Manager Shift Supervisor H.P. Superviror Immediate Supervisor
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QUESTION: 090 (1.00) l ' The following conditions exist:
* Unit 1 is at 100% powe * A small leak has developed on the charging line to the RCS inside containmen * A repair team has been formed and repairs have been ongoing for the past four hour *
It is now 1300 hours.
l UHICH ONE (1) of the following individuals is allowed to work on the repair team? I An operator who has been on site since midnigh A HP Technician who worked the job earlier in the day and left the containment for lunch at 1100 hour A maintenance person who has returned to the site after having left at 0800 hours following a sixteen (16) hour shif An untrained Contract Vendor who will be escorted by a station maintenance employee who will assist in the repair QUESTION: 091 (1.00) WHICH ONE (1) of the following methods should be used to control the status of an annunciator alarm that has been caused by tagging out a component? Include the reason for the annunciator as part of the control room lo Submit a Request For Engineering Assistance to clear the alar Submit a Work Re clear the alarm. quest to have the annunciator jumpered so as to Include the annunciator as part of the Tag-Out.
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. REACTOR OPERATOR Page 56 i QUESTION: 092 (1.00) WHICH ONE (1) of the following is the MAXIMUM number of times that an operator may attempt to RESET a thermal overload device? one ; I two i three l four QUESTION: 093 (1.00) The following conditions exist:
* A Tag-Out is in progres * An administratively controlled locked open valve located in a contaminated area must be CLOSE ,
WHICH ONE (1) of the following methods should be used to reposition this component? The valve shall be locked in the CLOSED position following authorization from the Unit SR The valve shall be locked in the CLOSED position following authorization from the Superintendent of Operations, The valve shall be CLOSED following authorization from the Unit SRO, and the lock shall be locked to the valve body without locking the valv The valve shall be CLOSED following authorization from the Superintendeat of Operations, and the lock shall be locked to the valve body without locking the valv . _ _ _ _. _. - , . _ . _ . - _ _ . __ . REACTOR OPERATOR Page 57 QUESTION: 094 (1.00) The following conditions exist:
* An Operator is performing an approved procedure when he encounters a " BLUE" tag signed by a Shift Supervisor giving instructions contradictory to those in the procedur WHICH ONE (1) of the following actions should be taken by the operator? Deviate from the procedure and follow instructions of the BLUE ta Destroy the tag, this type of tag is no longer approved for us Perform the procedure as written and disregard instructions on the BLUE ta Notify Shift Supervisor of the BLUE tag and await instructions before proceedin QUESTION: 095 (1.00)
The following plant conditions exist:
*
Unit 1 is in Intermediate Shutdow * A plant operating procedure calls for operating a motor operated containment isolation valv * The valve motor is tagged out for maintenance, so the valve must be MANUALLY operate WHICH ONE (1) of the following actions should be taken? Operate the valve as instructed as it is authorized by a procedur DO NOT operate the valve because MANUAL operation of containment isolation valves is prohibite Operate the valve Ot!LY after approval is obtained from the Superintendent of Operation Operate the valve ONLY after approval is obtained from the Station Nuclear Safety and Operating Committee (SNSOC).
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QUESTION: 096 (1.00) WHICH ONE (1) of the following color combinations is used to identify drums I for separation and disposal of radioactive and clean waste? CLEAN WASTE CONTAMINATE WASTE Green Red White Red Green Yellow White Yellow
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QUESTION: 097 (1.00) The following conditions exist:
* Operator "A" is Independently verifying the position of a component repositioned by Operator "B".
- Operator "A" finds the component in the WRONG positio )
i WHICH ONE (1) of the following actions is the MINIMUM required of Operator "A" per VPAP-1405, " Independent and Simultaneous Verification"? Have Operator "B" reposition the component while watching his l action Have Operator "B" reposition the component independently then verify the components position indopendentl Reposition the component to the correct position then report the discrepancy to the cognizant supervisor, Report the discrepancy to the cognizant supervisor and await further ! instruction before continuin _, .- -
e* & Page 59 , REACT,0R OPERATOR QUESTION: 098 (1.00) WHICH ONE (1) of the following methods should be used to Independently Verify the position of a LOCKED CLOSED valve? Remove the lock and attempt to move the valve in the CLOSED directio Attempt to move the valve in the CLOSED direction without removing the loc Remove the lock, partially 0 PEN the valve, then CLOSE and RELOCK the valv Partially OPEN the valve without removing the lock, then CLOSE the valv QUESTION: 099 (1.00) The following conditions exist:
* A Site Evacuation has been ordere * * One This (1)dividual in is known to be onsite. individual has not reported for accountabilit WHICH of the the missing following?are the preferred individuals to conduct the search for individual Security staff members Health Physics staff members Licensed Operation staff members Fire Team members
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.REACTQR OPERATOR QUESTION: 100 (1.00)
WHICH ONE (1) of the following is the MINIMUM number of plant personnel required to be onsite to man the Fire Brigade per Technical Specifications 6.1.B.7, " Organization, Safety, and Operation Review"? i Four (4) which CANNOT include those required for safe shutdow ' I Four (4) which CAN include those required for safe shutdown.
, Five (5) which CANNOT include those required for safe shutdown.
1 Five (5) which CAN include those required for safe shutdow : l l I l l l
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(********** END OF EXAMINATION **********) i l
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ANSWER: 001 (1.00) l [+1.0]
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REFERENCE: ND-93.3-LP-3, " Rod Control System", Objective H, page 2 , KA 001050K401 (3.4/3.8) l
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001050K401 ..(KA's) ANSBER: 002 (1.00) [+1.0] ;
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REFERENCE: 1 ND-93.3-LP-3, " Rod Control System", H- . KA 001000K402 (3.8/3.8)
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001000K402 ..(KA's) i ANSBER: 003 (1.00) [+1.0] l REFERENCE: l P-5.2.1, " Starting Any Reactor Coolant Pump", page . 1-GOP-1.1, " Unit Startup, RCS Heatup From Ambient To 195 Degrees F", page 2 . KA 003000G005 (3.4/3.8), 003000G010 (3.3/3.6) ; l l 003000G005 003000G010 ..(KA's) l l l L
.REACTQR OPERATOR
'- ) h Page 62 ANSWER: 004 (1.00) (+1.0]
REFERENCE: ND-88.1-LP-6, " Reactor Coolant Pumps", Objective C, page 2 . KA 003000A403 (2.8/2.5) 003000A403 ..(KA's) ANSWER: 005 (1.00) [+1.0] REFERENCE: P.S.2.1, " Starting Any Reactor Coolant Pump", page . KA 003000A107 (3.4/3.4) 003000A107 ..(KA's) l l ANSWER: 006 (1.00) [+1.0] REFERENCE: ND-88.3-LP-2, " Charging and Letdown", Objective C, page 1 . KA 004020A305 (3.2/3.0) 004020A305 ..(KA's)
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ANSWER: 007 (1.00) [+1.0] l REFERENCE: ND-88.3-LP-10, " Chemical and Volume Control System", Objective F, pages 18-1 ; KA 004010A407 (3.4/3.3) 004010A407 ..(KA's)
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ANSWER: 008 (1.00) ! P [+1.0] REFERENCE: ND-91-LP-3, " Safety Injection System Operation", Objective B, page , KA 013000X412 (3.7/3.9) 4 013000K412 ..(KA's)
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ANSWER: 009 (1.00) i [+1.0] P e i
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. REACTOR OPERATOR REFERENCE: ND-91-LP-4, " Consequence Limiting Safeguards", Objective D, page . ND-91-LP-3, " Safety Injection System Operations", Objective F, page 2 . KA 013000K409 (2.7/3.1)
013000K409 ..(KA's) ANSBER: 010 (1.00) [+1.0] REFERENCE: ND-93.3-LP-3, "" Rod Control System", Objective C, page . KA 015000K302 (3.3/3.5) 015000K302 ..(KA's)
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ANSBER: 011 (1.00) [+1.0] REFERENCE: ND-93.2-LP-3, " Intermediate Range NIS", Objective D, page 8 & . KA 015000A303 (3.9/3.9) .
015000A303 ..(KA's) l l ANSWER: 012 (1.00) [+1.0] _
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. REACTOR OPERATOR REFERENCE: ND-93.4-LP-3, " Inadequate Core Cooling Monitor", Objective F, page 2 . KA 017020A401 (3.8/4.1) 017020A401 ..(KA's) ANSWER: 013 (1.00) [+1.0] REFERENCE: ND-88.4-LP-6, " Containment Ventilation", Objective B, page . KA 022000A301 (4.1/4.3) 022000A301 ..(KA's) ANSWER: 014 (1.00) [+1.0] I REFERENCE: ND-89.0-LP-2, " Main Condensate System", Objective F, page 1 . KA 059000G001 (3.1/3.2) 059000G001 ..(KA's) ANSMER: 015 (1.00) [+1.0]
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.REACTQR OPERATOR Page 66 REFERENCE: ND-89.3-LP-3, " Main Feedwater System", Objective G, page 1 . KA 059000K416 (3.1/3.2) 059000K416 ..(KA's) ANSWER: 016 (1.00) [+1.0] REFERENCE: ND-89.3-LP-3, " Main Feedwater System", Objective J, page . KA 059000K419 (3.2/3.4) 059000K419 ..(KA's) ANSWER: 017 (1.00) a .- [+1.0] REFERENCE: ND-89.3-LP-4, " Auxiliary Feedwater System", Objective D, page 1 . KA 061000K107 (3.6/3.8) 061000K107 ..(KA's) ANSWER: 018 (1.00) [+1.0]
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. REFERENCE: ND-89.3-LP-4, " Auxiliary Feedwater System", Objective H, page 1 . KA 061000K402 (4.5/4.6) 061000K402 ..(KA's) l I ANSWER: 019 (1.00) l
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REFERENCE: )
^l ND-89.3-LP-4, " Auxiliary Feedwater System", Objective C, page 1 ' KA 061000K109 (2.6/2.8)
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061000K109 ..(KA's) ANSWER: 020 (1.00) a.st[+1.0] 14E /s-cay REFERENCE: . ND-92.4-LP-4, " Liq /2.9)uid Waste System", Objective B, page KA 068000K107 ( K107 ..(KA's) ANSWER: 021 (1.00) [+1.0)
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REFERENCE: ND-92.4-LP-2, " Gaseous Waste System", Objective D, page 1 I KA 071000G005 (2.4/3.1) l 071000G005 ..(KA's) i I ANSWER: 022 (1.00) ( l l , [+1.0]
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REFERENCE:
' ND-92.4-LP-2, " Gaseous Waste System", Objective B, page 18.
, KA 071000K504 (2.5/3.1), 071000A429 (3.0/3.6) l l I 071000K504 071000A429 ..(KA's)
!
l ANSBER: 023 (1.00) [+1.0] REFERENCE: ND-93.5-LP-1, "Victoreen Area Monitoring System", Objective D, page . KA 072000A401 (3.0/3.3) I ! ! 072000A401 ..(KA's) l ANSWER: 024 (1.00) [+1.0]
l l \ l t _ - - _ - - _ _ _ _ ____ --_- - - -
, .REACTQR OPERATOR .Page 69-REFERENCE: ND-93.3-LP-2, " Delta T/Tavg Instrumentation System", Objective E, page . KA 016000A201 (3.0/3.1) ;
l 016000A201 ..(KA's) .
-1 ANSWER: 025 (1.00) [+1.0] l
!
REFERENCE:
! ND-93.4-LP-3, " Inadequate Core Cooling Monitor System", pages 7- I KA 002000K107 (3.5/3.7) !
I I 002000K107 ..(KA's) l ANSWER: 026 (1.00) . 1 [+1.0] 1
:l REFERENCE- ! ND-93.4-LP-3, " Inadequate Core Cooling Monitor System", pages 11-1 . KA 002000G007 (3.3/3.6) ;
002000G007 ..(KA's) ANSWER: 027 (1.00) [+1.0] i
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.REACTQR OPERATOR Q Page 70 l
REFERENCE: ND-91-LP-2, " Safety Injection System", Objective E, page 1 . KA 006020K603 (2.8/3.1) ;
!
006020K603 ..(KA's) l ANSWER: 028 (1.00) [+1.0) i I REFERENCE: j i ND-91-LP-2, " Safety Injection System", Objective E, page 1 !' KA 006000K410 (3.6/3.7)
006000K410 ..(KA's) ANSWER: 029 (1.00) [+1.0] REFERENCE: ND-93.3-LP-5, " Pressurizer Pressure Control", Objective C, page . KA 010000G009 (3.6/3.5) I 010000G009 ..(KA's)
;
ANSWER: 030 (1.00) [+1.0] l l
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.REACTQR OPERATOR Page 71 REFERENCE: ND-93.3-LP-7, " Pressurizer Level Control System", Objective B, page . KA 011000K404 (3.0/3.3) l 011000K404 ..(KA's) i ANSWER: 031 (1.00) [+1.0]
REFERENCE: ND-93.3-LP-7, " Pressurizer Level Control", Objective D, page 1 . KA 011000A101 (3.5/3.6) 011000A101 ..(KA's) ANSWER: 032 (1.00) [+1.0) REFERENCE: ND-93.3-LP-17, " Anticipatory Mitigation Actuation Circuitry (AMSAC)", Objective F, page . KA 012000K201 (3.3/3.7)
!
012000K201 ..(KA's) ANSWER: 033 (1.00) [+1.0]
:
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.REACTQR OPERATOR Page 72 REFERENCE: ND-93.3-LP-14, " Overpower /0vertemperature Delta T", Objective H, ND-93.3-H/T-1 . KA 012000K611 (2.9/2.9) 012000K611 ..(KA's) ANSWER: 034 (1.00) [+1.0] REFERENCE: ND-93.3-LP-4, " Rod Position Indication System", Objective C, page 7, KA 014000A202 (3.1/3.6) 014000A202 ..(KA's) ANSWER: 035 (1.00) [+1.0] REFERENCE: ND-93.3-LP-3, " Rod Control System", Objective M, page 2 . Technical Specifications 3.12.C., " Inoperable Control Rods", page 3.12-1 . KA 014000G0ll (3.0/3.9) 014000G011 ..(KA's)
._ _ -__ __ _____ ____ _ - . _
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O w Page 73 l
.REACTQR. OPERATOR l
, ANSWER: 036 (1.00) \ l l [+1.0] l , REFERENCE:
1 l ND-91-LP-5, " Containment Spray", Objective B, page 11.
I KA 026000A401 (4.5/4.3) l 026000A401 ..(KA's)
ANSWER: 037 (1.00) [+1.0] l l REFERENCE: ND-91-LP-5, " Containment Spray", Objective D, page 16.
, KA 026000A101 (3.9/4.2)
026000A101 ..(KA's) ANSMER: 038 (1.00) [+1.0] REFERENCE: ND-88.4-LP-6, " Containment Ventilation", Objective E, page . KA 029000A301 (3.8/4.0) 029000A301 ..(KA's) l l
1
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g O REACTOR OPERATOR U - Page 74 ANSWER: 039 (1.00) [+1.0] REFERENCE: AP-22.02, " Loss Of Spent Fuel Pit Level", page . KA 033000A203 (3.1/3.5), 033000K107 (2.4/2.5) 033000A203 033000K107 ..(KA's) ANSWER: 040 (1.00) [+1.0] REFERENCE: ND-93.3-LP-8, "S/G Water Level Control System", Objectives C & D, page . KA 035010K401 (3.6/3.8) 035010K401 ..(KA's) ANSWER: 041 (1.00) [+1.0] l
l
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, REACTOR OPERATOR O O Page 75'
'j REFERENCE:
1 ND-90.3-LP-6, "125 VOC Distribution", Objective F, page 1 ! KA 063000G008 (3.1/3.2)
!
!
l 063000G008 ..(KA's) J i ANSWER: 042 (1.00) [+1.0] REFERENCE: j ND-90.3-LP-1, " Emergency Diesel Generator-Mechanical", Objective E, page i 2 '
- KA 064000G007 (3.4/3.6) l
!
l 064000G007 ..(KA's) l l l ANSWER: 043 (1.00) ; l [+1.0] i l
!
REFERENCE: 1 ND-88.5-LP-1, " Component Cooling Water System", Objectives E, pages 7- l KA 073000K401 (4.0/4.3) 073000V,401 ..(KA's)
!
!
ANSWER: 044 (1.00) [+1.0]
!
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.REACTQR OPERATOR h Page 7G REFERENCE: ND-88.2-LP-1, " Residual Heat Removal System", Objective C, page 1 . ND-93.3-LP-6, " Overpressure Mitigation System", Objective E, pages 6- . KA 005000K401 (3.0/3.2) 005000K401 ..(KA's) ANSWER: 045 (1.00) [+1.0] REFERENCE: ND-88.2-LP-1, " Residual Heat Removal System", Objective 8, page 1 . KA 005000K407 (3.2/3.5) 005000K407 ..(KA's) ANSWER: 046 (1.00) [+1.0] REFERENCE: l l ND-88.1-LP-3, " Pressurizer And Pressure Relief", Objective D, Pressurizer l Relief Tank Diagra ' KA 007000A301 (2.7/2.9) 007000A301 ..(KA's) ANSWER: 047 (1.00) [+1.0]
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.REACTQR OPERATOR Page 77 REFERENCE: ND-88.5-LP-1, " Component Cooling". Objective C & F , page . KA 008000K104 (3.3/3.3)
008000K104 ..(KA's)
ANSWER: 048 (1.00) [+1.0] REFERENCE: ND-88.4-LP-8, " Hydrogen Recombiners", Objective D, page . 1-E-1, " Loss Of Reactor Or Secondary Coolant", Attachment . KA 028000A401 (4.0/4.0) 028000A401 ..(KA's) ANSWER: 049 (1.00) [+1.0] REFERENCE: ND-92.5-LP-3, " Fuel Handling Tools", Objective E, page 13, KA 034000G005 (2.6/3.5) , 034000G005 ..(KA's) ANSWER: 050 (1.00) [+1.0] J l
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..REACTQR OPERATOR Page 78 ' REFERENCE: ND-93.3-LP-9, " Steam Dump Control System", Objective D, page 1 . KA 041020K418 (3 4/3.6), 041020K414 (2.5/2.8) i 041020K418 041020K414 ..(KA's) l ' . ANSWER: 051 (1.00) [+1.0]
REFERENCE: , ND-89.5-LP-2, " Service Water System", Objective D, page 8.
, KA 076000A401 (2.9/2.9), 076000G005 (2.8/3.2) ]
l i 076000A401 076000G005 ..(KA's)
!. i j ANSWER: 052 (1.00) i j [+1.0] i- ] REFERENCE:
' AP-39.0, " Natural Circulation Of RCS", Attachment 1, page 1.
,' KA 000015Al21 (4.4/4.5) ;
l 000015Al21 ..(KA's) ANSWER: 053 (1.00) [+1.0) i
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.REACTQR OPERATOR Page 79 REFERENCE: ND-95.1-LP-3, " Partial Loss Of RCS Flow", Objective 8, pages 9-1 . l KA 000015K104 (2.9/3.1) , i 000015K104 ..(KA's) l ANSWER: 054 (1.00) [+1.0) REFERENCE:
, AP-3.00, " Emergency Boration", page l
' KA 000024K301 (4.1/4.4)
000024K301 ..(KA's) ANSBER: 055 (1.00) [+1.0) REFERENCE: : ND-88.3-LP-9, " Blender Control Subsystem", Objective D, page 2 ; KA 000024All7 (3.9/3.9) )
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000024All7 ..(KA's) ANSSER: 056 (1.00) [+1.0]
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REACTQR OPERATOR Page 80 REFERENCE: ND-88.5-LP-1, " Component Cooling Water", Objective F, page 2 . KA 000026K303 (4.0/4.2) l
!
000026K303 ..(KA's) ANSWER: 057 (1.00) i [+1.0) l REFERENC ; !-AP-15.00, " Loss Of Component Cooling Water", page I KA 000026G010 (3.6/3.5) 000026G010 ..(KA's) ) ANSWER: 058 (1.00) l [+1.0] REFERENCE: I ND-95.1-LP-13, " Stuck Open Pressurizer Spray Valve", Objective . 1-AP-31.00, " Increasing or Decreasing RCS Pressure", page . KA 000027G010 (3.7/3.8) 000027G010 ..(KA's) ANSWER: 059 (1.00) [+1.0] ;
.
P .REACTQR 0PERATOR Page 81 REFERENCE: ND-95.2-LP-3, "Secondar ' ~KA 000040K106 (3.7/3.8)y Breaks", Objective D, page 3.21 i
.
000040K106 ..(KA's) ..
(
ANSWER: 060 (1.00) [+1.0] , REFERENCE: ND-95.2-LP-8, " Loss Of All AC Power", page . KA 000055G006 (3.8/4.1) ! i 000055G006 ..(KA's)
-:
ANSWER: 061 (1.00) ,
' [+1.0]
REFFRENCE: l
! ND-95.2-LP-8, " Loss Of All AC Power", Objective A, page . KA 000055G007 (3.6/3.7)
l 000055G007 ..(KA's) ANSWER: 062 (1.00) [+1.0] !
! .,
i .REACTQR OPERATOR Page 82
.
REFERENCE:
' FCA-9.00, " Limiting Intake Structure Fire", page . KA 000067G010 (3.3/3.7)
000067G010 ..(KA's)
.
ANSWER: 063 (1.00) , [+1.0] REFERENCE: , AP-20.01 " Main Control Room Oxygen Monitor - Alarm Or Malfunction", page . KA 000068G007 (3.4/3.5) 000068G007 ..(KA's)
;
l ANSWER: 064 (1.00) '. I [+1.0] _i
l
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REFERENCE: FCA-1.00, " Limiting MCR Fire", page . KA 000068A123 (4.3/4.4) 000068A123 ..(KA's) ANSWER: 065 (1.00) [+1,0]
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O w Page 83 REACTQR 0PERATOR REFERENCE: FR-I.3, " Response To Voids In Reactor Vessel", Attachment . KA 000069G012 (3.5/3.5) 000069G012 ..(KA's) ANSWER: 066 (1.00) [+1.0] l REFERENCE: F-2, " Core Cooling", Drawing Number CB38 . Steam Tables KA 000074A201 (4.6/4.9) 000074A201 ..(KA's)
!
ANSUER: 067 (1.00) [+1.0] REFERENCE: FR-I.3, " Response To Voids In Reactor Vessel", page i KA 000074K311 (4.0/4.4) l
l 000074K311 ..(KA's) ANSWER: 068 (1.00) [+1.0] l
:
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Page 84 i REFERENCE: ND-95.2-LP-2, " Rod Withdrawal Accident", AP-1.00, " Rod Control System Malfunction"page 15., page . KA 000001G010 (3.9/4.0) 000001G010 ..(KA's) ANSWER: 069 (1.00) [+1.0] > REFERENCE: ND-95.1-LP-5, " Dropped / Misaligned Control Rod Recovery", Objective , pages 7-1 . KA 000003G007 (3.4/3.6)
000003G007 ..(KA's) l l 1 l l i ANSWER: 070 (1.00) ! 1 [+1.0] REFERENCE: l ND-88.1-LP-3, " Pressurizer and Pressure Relief", Objective D, pages 3.24-3.2 . KA 000008A203 (3.9/3.9) 000008A203 ..(KA's) l l l
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.REACTQR OPERATOR Page 85 ANSWER: 071 (1.00) [+1.0] REFERENCE: ND-95.2-LP-7, " Loss of Reactor Coolant Accident", Objective G, page AIA- . KA 000009A223 (2.8/3.3) 000009A223 ..(KA's) ANSWER: 072 (1.00) [+1.0] REFERENCE: ND-95.2-LP-7, " Loss Of Reactor Coolant Accident", Objectives C & D, pages 18-19 & 2 . KA 000009K101 (4.2/4.7) 000009K101 ..(VA's) ANSUER: 073 (1.00) [+1.0]
.
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.REACTQR 0PERATOR REFERENCE: ND-95.3-LP-II, "ES-1.4, Transfer To Hot Leg Recirculation", Objective B, page . G 0000llK313 (3.8/4.2)
000011K313 ..(KA's) ANSWER: 07' (1.00) [+1.0; REFERENCE: l ND-95.2-LP-12, " Loss Of RHR Events", Objective B, pages 24-2 l KA 000025K101 (3.9/4.3) ;
000025K101 ..(KA's) l ANSUER: 075 (1.00) [+1.0] REFERENCE: ND-95.2-LP-12, " Loss Of RHR Events", Objective B, pages 28-2 . KA 000025G007 (3.4/3.6) 000025G012 ..(KA's) ANSWER: 076 (1.00) [+1.0]
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REACTQR'0PERATOR O O Page 87
!
l I REFERENCE: ND-95.1-LP-11, " Anticipated Transient Without Trip (ATWT)", Objective C, l page 11.14 KA 000029G007 (3.8/4.0), 000029G010 (4.5/4.5) I j
:
000029G007 000029G010 ..(KA's) )
:
l i ANSWER: 077 (1.00) : [+1.0)
I REFERENCE: 1, 1-AP-24.01, "Large Steam Generator Tube Leak", page . KA 000037A216 (4.1/4.3) l
.I 000037A216 ..(KA's) -l l
,
ANSWER: 078 (1.00) , [+1.0]
)
'
REFERENCE: E-3, " Steam Generator Tube Rupture", page 3 . KA 000038A217 (3.8/4.4) 000038A217 ..(KA's) l
!
ANSWER: 079 (1.00) l [+1.0] H l
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O Page 88 REFERENCE: , ND-95.2-LP-3, " Secondary Breaks", Objective E, page 3.27.
I KA 000054K101 (4.1/4.3) 000054K101 ..(KA's) ANSWER: 080 (1.00) [+1.0] REFERENCE: ND-95.3-LP-41, "FR-H.1, Response To loss Of Secondary Heat Sink", Objective E, page 12, KA 000054K304 (4.4/4.6).
000054K304 ..(KA's) ANSWER: 081 (1.00) [+1.0] REFERENCE: E-1, " Loss Of Reactor Or Secondary Coolant", page . KA 000054A203 (4.2/4.3), 000054K102 (3.6/4.2) 000054A203 000054K102 ..(KA's) ANSWER: 082 (1.00) [+1.0] l l l l l
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REACTOR 0PERATOR Page 89 j REFERENCE: AP-10.06, " Loss of DC Power", Attachment , KA 000058A203 (3.5/3.9) l l 000058A203 ..(KA's) ANSWER: 083 (1.00) [+1.0] REFERENCE: ND-93.5-LP-4, "Xaman Process Radiation Monitoring System", Objective D, page 1 . KA 000060G010 (3.8/3.8) 000060G010 ..(KA's) ANSWER: 084 (1.00) [+1.0] REFERENCE: AP-5.21, " Radiation Monitor System Ventilation Monitor Malfunction", i page . KA 000060K201 (2.6/2.9) l 000060K201 ..(KA's) l l ANSWER: 085 (1.00) . 1 [+1.0] ; I
.REACTQR 0PERATOR Page 90 REFERENCE: AP-22.01, " Loss of Refueling Cavity", page . KA 000036G010 (3.7/3.8) 000036G010 ..(KA's) ANSWER: 086 (1.00) [+1.0] REFERENCE: 1. . ND-90.3-LP-7, " Emergency Distribution Protection and Control", Objective E, pop ND-90.3/H- . KA 000056A247 (3.8/3.9) 000056A247 ..(KA's) ANSWER: 087 (1.00) [+1.0] REFERENCE: ND-95.1-LP-9, " Loss Of Instrument Air", Objective C, page 9.23 KA 000065K308 (3.7/3.9) 000065K308 ..(KA's) ANSWER: 088 (1.00) [+1.0]
REACTOR 0PERATOR Page 91 REFERENCE: ND-95.3-LP-2, " Emergency Procedure Writer's Format", Objective C, page 12.
l KA 194001A102 (4.1/3.9) l 194001A102 ..(KA's) . l l' ANSWER: 089 (1.00)
! [+1.0]
i REFERENCE: SUADM-0-19, " Guidelines, Procedures, and Limitations For Containment Entry", page . KA 19400lKil3 (3.3/3.6) I" 19400lK113 ..(KA's) ANSWER: 090 (1.00) [+1.0] REFERENCE: l SUADM-0-19, " Guidelines, Procedures, and Limitations For Containment Entry", page . KA 19400lKll3 (3.3/3.6) 19400lK113 ..(KA's) l l l
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! REACTQR 0PERATOR h Page 92 l ! ANSWER: 091 (1.00) [+1.0] RLftdENCE: l OPAP-0006, " Shift Operating Practices", page 1 . KA 19400lK102 (3.7/4.1) 19400lK102 ..(KA's) I ANSBER: 092 (1.00) [+1.0] REFERENCE: OPAP-0006, " Shift Operating Practices", page 1 . KA 19400lK107 (3.6/3.7) 19400lK107 ..(KA's) ANSWER: 093 (1.00) [+1.0] REFERENCE: OPAP-0008, " Administrative Control of Keys and Locked Valves and Switches", page . KA 19400lK102 (3.7/4.1) 19400lK102 ..(KA's) l l l
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REACTOR *0PERATOR Page 93 ANSWER: 094 (1.00) [+1.0] REFERENCE: OPAP-00ll, "Special Order Tag Control", pages 4 and . KA 19400lK102 (3.7/4.1) 19400lK102 ..(KA's) ANSUER: 095 (1.00) [+1.0] REFERENCE: OPAP-0012, " Valve Operations", page . KA 194001A110 (2.9/3.9) 194001A110 ..(KA's) ANSWER: 096 (1.00) [+1.0] REFERENCE: VAP-2101, " Radiation Protection Plan", page 5 . KA 19400lK104 (3.3/3.5) 19400lK104 ..(KA's)
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b) REACTOR' OPERATOR Page 94 ANSWER: 097 (1.00) [+1.0] REFERENCE: VPAP-1405, " Independent and Simultaneous Verification", page . KA 194001K101 (3.6/3.7) 19400lK101 ..(KA's) ANSWER: 098 (1.00) [+1.0] REFERENCE: VPAP-1405, " Independent and Simultaneous Verification", page 1 KA 19400lK101 (3.6/3.7) 19400lK101 ..(KA's) ANSWER: 099 (1.00) [+1.0]
.
REFERENCE: EPIP-5.03, " Personnel Accountability", page . KA 194001A116 (3.1/4.4) 194001All6 ..(KA's)
e v h Page 95 REACTOR 0PERATOR l l ANSWER: 100 (1.00) [+1.0] REFERENCE: Technical Specification 3.21.B.1, " Fire Protection Features", page TS , ' 3.21- . KA 194001A103 (2.5/3.4) 194001A103 ..(KA's) l I
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l (********** END OF EXAMINATION **********)
._ _ . . _ _ .REACTGR* 0PERATOR Page 1 ANSWER KEY l
I I l MULTIPLE CH0 ICE 023 c l 001 c 024 c 002 c 025 d 003 b 026 b l 004 b 027 a l 005 c 028 b I 006 b 029 b 007 b 030 c 008 b 031 a j i 009 d 032 b i 010 b 033 b 011 a 034 a 012 b 035 d 013 d 036 c 014 c 037 d 015 d 038 b l 016 c 039 c 017 a 040 b 018 a 041 c
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019 d 042 a 020 a or c y ,. .% 043 b
'021 b 044 a l 022 b 045 d
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ANSWER KEY l 046 b 069 a ! 047 a 070 c
048 d 071 a 049 b 072 c 050 b 073 d 051 c 074 a 052 c 075 b 053 c 076 b 054 c 077 b 055 c 078 d 056 c 079 a 057 b 080 d 058 d 081 b 059 c 082 d 060 d 083 a I 061 b 084 c 062 d 085 d
'
063 b 086 c 064 c 087 c i 065 c 088 c I 066 a 089 a 067 b 090 b 068 b 091 d l l l
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Page 3 REACTOR'0PERATOR ANSWER XEY 092 a 093 c 094 d 095 d 096 d 097 d 098 b 099 d 100 c l
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(********** END OF EXAMINATION **********)
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EST CROSS REFERENCE Page 1 R0 Exam PWR Reactor 0rganized by Question Number QUESTION VALUE REFERENCE 001 1.00 8000007 - 002 1.00 8000009-003 1.00 8000014 004 1.00 8000015 < 005 1.00 8000057 006 1.00 8000005 - 007 1.00 8000058 - 008 1.00 8000019 - 009 1.00 8000034 010 1.00 8000035 - 011 1.00 8000037 - 012 1.00 8000039 013 1.00 8000041 014 1.00 8000054 015 1.00 8000052 016 1.00 8000053 - 017 1.00 8000050 - 018 1.00 8000051 019 1.00 8000059 - 4 020 1.00 8000060 l 021 1.00 8000055 l 022 1.00 8000056 023 1.00 8000003 024 1.00 8000011-025 1.00 8000012* 026 1.00 8000013 ; 027 1.00 8000020 : 028 1.00 8000022- 1 029 1.00 8000024 : 030 1.00 8000025 ; 031 1.00 8000027 - ! 032 1.00 8000032 1 033 1.00 8000033 034 1.00 8000030 035 1.00 8000031 036 1.00 8000043 - 037 1.00 8000044 I
'
038 1.00 8000045 039 1.00 8000046 l 040 1.00 8000028 041 1.00 8000049 042 1.00 8000047 043 1.00 8000002 - 044 1.00 8000017 4 045 1.00 8000018 046 1.00 8000023 - 047 1.00 8000006 048 1.00 8000042 049 1.00 8000048 e
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l .- . IEST CROSS REFERENCE Page 2 i R0 Exam PWR Reactor 0rganized by Question Number QUESTION VALUE REFERENCE 050 1.00 8000029 051 1.00 8000061 052 1.00 8000074 053 1.00 8000092 054 1.00 8000071 055 1.00 8000088 - 056 1.00 8000072 057 1.00 8000073 058 1.00 8000095 059 1.00 8000075 060 1.00 8000078 061 1.00 8000079 062 1.00 8000086 063 1.00 8000084 064 1.00 8000085 1 065 1.00 8000077 066 1.00 8000069 067 1.00 8000070 068 1.00 8000063 069 1.00 8000064 070 1.00 8000098 071 1.00 8000082 072 1.00 8000089 073 1.00 6000091 074 1.00 8000096 075 1.00 8000097 076 1.00 8000067 077 1.00 8000101 078 1.00 8000099 079 1.00 8000076 080 1.00 8000093 081 1.00 8000094 082 1.00 8000102 083 1.00 8000080 084 1.00 8000081 085 1.00 8000103 086 1.00 8000104 087 1.00 8000066 088 1.00 8000107 , 089 1.00 8000108 l 090 1.00 8000109 091 1.00 8000110 092 1.00 8000111 - 093 1.00 8000113 094 1.00 8000115 - 095 1.00 8000116* 096 1.00 8000119* 097 1.00 8000123 ' 098 1.00 8000125- , I , l l l l i k
.- .
e TEST CROSS REFERENCE O Page 3 R0 Exam PWR Reactor Organized by Question Number QUESTION VALUE REFERENCE 099 1.00 8000128 100 1.00 8000130 kbb$bb
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. * . '"IESTCROSSREFERENCE Page 4 R0 Exam PWR Reactor ,
0r9anized by KA Group i PLANT WIDE GENERICS QUESTION VALUE KA j ,
088 1.00 194001A102 100 1.00 194001A103 095 1.00 194001A110 099 1.00 194001A116 , 097 1.00 194001K101 l 098 1.00 19400lK101 091 1.00 194001K102 093 1.00 19400lK102 094 1.00 19400lK102 i 096 1.09 19400lK104 ' 092 1.00 19400lK107 ' 089 1.00 19400lK113 090 1.00 19400lKll3 ) PWG Total ib$bb . PLANT SYSTEMS Group I QUESTION VALUE KA 002 1.00 001000K402 001 1.00 001050K401 005 1.00 003000A107 004 1.00 003000A403 003 1.00 003000G005 007 1.00 004010A407 006 1.00 004020A305 009 1.00 Ol3000K409 008 1.00 013000K412 011 1.00 015000A303 010 1.00 015000K302 012 1.00 017020A401 013 1.00 022000A301 014 1.00 059000G001 015 1.00 059000K416 016 1.00 059000K419 017 1.00 061000K107 019 1.00 061000K109 018 1.00 061000K402 , ' 020 1.00 068000K107 021 1.00 071000G005 1 022 1.00 071000K504 023 1.00 072000A401
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i .- . O TEST CROSS REFERENCE O Page 5 . R0 Exam PWR Reactor 0rganized by KA Group
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PLANT SYSTEMS Group I QUESTION VALUE KA PS-I Total 23.00 Group II QUESTION VALUE KA 026 1.00 002000G007 025 1.00 002000K107 028 1.00 006000K410 027 1.00 006020K603 029 1.00 010000G009 031 1.00 011000A101 030 1.00 011000X404 032 1.00 012000K201 033 1.00 012000K611 034 1.00 014000A202 035 1.00 014000G0ll : 024 1.00 016000A201 1 l 037 1.00 026000A101 036 1.00 026000A401 ) 038 1.00 029000A301 039 1.00 033000A203 ) 040 1.00 035010K401 1 041 1.00 063000G008 1 042 1.00 064000G007 ') 043 1.00 073000K401 PS-II Total 2b$bb Group III OUESTION VALUE KA 044 1.00 005000K401 045 1.00 005000K407 046 1.00 007000A301 047 1.00 008000K104 048 1.00 028000A401 049 1.00 034000G005 - I 050 1.00 041020K418 051 1.00 076000A401 j PS-III Total b$bb
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'"IESTCROSSREFERENCE e '"'
Page 6 R0 Exam PWR Reactar 0rganized by KA Group PLANT SYSTEMS QUESTION VALUE KA PS Total 51.00 EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 052 1.00 000015A121 053 1.00 000015K104 055 1.00 000024All7 054 1.00 000024K301 057 1.00 000026G010 056 1.00 000026K303 058 1.00 000027G010 059 1.00 000040K106 060 1.00 000055G006 061 1.00 000055G007 062 1.00 000067G010 064 1.00 000068A123 063 1.00 000068G007 065 1.00 000069G012 066 1.00 000074A201 067 1.00 000074K311 EPE-I Total ib$bb Group II QUESTION VALUE KA 068 1.00 00000lG010 069 1.00 000003G007 070 1.00 000008A203 071 1.00 000009A223 072 1.00 000009K101 073 1.00 0000llK313 075 1.00 000025G012 074 1.00 000025K101 076 1.00 000029G007 077 1.00 000037A216 078 1.00 000038A217 081 1.00 000054A203 079 1.00 000054K101 080 1.00 000054K304 082 1.00 000058A203
.
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.' 4, O TEST CROSS REFERENCE O Page 7 l
R0 Exam PWR Reactor I 0rganized by KA Group i V, EMERGENCY PLANT EVOLUTIONS l Group II QUESTION VALUE KA ! 083 1.00 000060G010 084 1.00 000060K201 EPE-II Total 'ih$bb Group III ; QUESTION VALUE KA 085 1.00 000036G010 086 1.00 000056A247 087 1.00 000065K308 EPE-III Total b5bb
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EPE Total bb$bb ______ W66666 Weemme l u
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_ __ _ _ s-REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (circle or X your choice) If you change your answer, write your selection in the blan HULTIPLE CHOICE 023 a b f( d a
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for the Ion Exchangers; and with no letdown, there is no heating i flow in the regenerative heat exchange With no CCW there is no seal water for the charging pumps; and with :l no charging letdown must be secured to stabilize pressurizer leve ;
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e O SENIOR REACTOR OPERATOR Page 32 , QUESTION: 046 (1.00) WHICH ONE (1) of the following requires an IMMEDIATE trip of all operating RCPs? Component Cooling Thermal Barrier Return valve TV-CC-120A fails l CLOSE Component Cooling Water is lost to all Reactor Coolant Pump motor Seal Injection is lost to all RCP RCP Bearing temperatures indicate 170 degrees QUESTION: 047 (1.00) The following plant conditions exist:
*
Unit 1 is operating at 100% powe * Pressurizer spray valve PCV 455B is stuck OPE * RCS pressure is decreasing RAPIDL * All efforts to close PCV 455B have faile WHICH ONE open spray va (1)lve?of the following actions must be taken to mitigate the stuck Trip RCP "B", then trip the reacto Trip RCP "C", then trip the reacto Trip the reactor, then trip RCP "B". Trip the reactor, then trip RCP "C".
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* ', SENTOR REACTOR OPERATOR Page 33 QUESTION: 048 (1.00)
WHICH ONE (1) of the following is the reason the Design Basis Accident (UFSAR) for a Main Steam Line break is analyzed for a Hot Shutdown Condition (HSD)? Emergency Core Cooling System requirements are less restrictive at HS Steam Generator temperatures are lower at HS Steam Generator inventories are greater at HSD, Minimum Shutdown Margin requirements are less restrictive at HS QUESTION: 049 (1.00) The following plant condition exists:
* Unit 1 is operating at 100% powe WHICH ONE (1) of the following chemistry conditions is indicative of a Main Condenser tube leak? Increased oxygen Increased Sodium Lecreased conductivity Decreased pH l
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SENIOR REACTOR OPERATOR Page 34
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l l QUESTION: 050 (1.00) i WHICH one (1) of the following means of communication is available following a J complete loss of ALL AC power to the site for greater than one (1) eight hour i shift? Station radio j
i Plant PBX l Gai-tronics ! Sound powered phones l l '
!
QUESTION: 051 (1.00) WHICH ONE (1) of the following component failures determines the severity of ) an' accident resulting from the loss of ALL AC power? Pressurizer PORVs I' RCP seals Steam Generator safety valves RCP thermal barrier heat exchangers i i l i
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i l ' * SENIDR REACTOR OPERATOR O O Page.35
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QUESTION: 052 (1.00) WHICH ONE (1) of the following MAJOR fires does NOT require a MANUAL reactor trip? Turbine Building fire Auxiliary Building fire ; Mechanical Equipment Room fire Main Switchyard fire l l ^l
!
QUESTION: 053 (1.00) WHICH ONE (1) of the following is the initial Control Room oxygen level which requires Operations personnel to INITIATE use of Self Contained Breathing Apparatuses (SCBA)?
' .5% .5% .5% .5%
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- SENIOR REACTOR OPERATOR O O Page 36
, QUESTION: 054 (1.00) j i The following plant conditions exist: l
*
Both Units are being shutdown due to a fire in the Main Control Roo * The Main Turbine has failed to trip using the MANUAL pushbutton ,
*
The Main Steam Trip Valves (MSTVs) will NOT close from the main ! control boar WHICH ONE (1) of the following alternate methods should be used to TRIP the Main Turbina per 0-FCA-1.00, limiting MCR Fire"? , Open the Main Generator Output breakers, Dispatch an operator to trip the Main Turbine locally at the turbin Close the MSTVs using FIRE EMERG CLOSE switch on APP R Panel in ESG ' Dispatch an operator to trip the Auto Stop 011 pump locally at the pum ! I
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A e SENIOR REACTOR OPERATOR Page 37 QUESTION: 055 (1.00) The following conditions exist:
* Unit 2 has experienced a small break Loss of Coolant Accident (LOCA . * Voids)are pre.. 9t in the reactor vesse * Operators are preparing to vent the reactor vesse * Containment pressure is 5.0 psi * Containment temperature is 230 degrees * Containment hydrogen concentration is 2.5%. * RCS pressure is 800 psi * CETCs indicate 522 degrees WHICH ONE (1) of the following is the correct venting time per 1-FR-I.3, " Response To Voids In Reactor Vessel"? (Attachments 1, 2 and 3 are attached.) .5 min :
1 .1 min .6 min l .1 min i l
- - * *
- . NUMBER ATTACHME!IT TITLE REVZSIO!1
- *
1-FR- ! H2 VENT TIME CALCUIATION ATTACHMENT PAGE 1 1 of 2
1 _DETERM1NE CONTAINMENT ! VOLUME (STP): I a) Calculate CTMT temperature in 'R I CTMT Temp 'R - CTMT Temp 'F + 460 l CTMT Temp 'R- l b) CTMT VOL (STP) - 1.75x10 6 cu.ft x CTMT Pressw x 492*R 14.7 psia CTMT Temp 'R c) CTMT VOL (STP) - 1.75x10 6 cu.ft x ( ) psia x 492'R ,
'
14.7 psia ( )*R d) CTMT VOL (STP) -
*H CTMT pressure less than 14.7 psia, THEN input actual valu E greater than 14.7 psia, THEN input 14.7 psia for conservatis ]
2 _DETERMIllE CONTAINMENT H2 CONCENTRATION: l a) Obtain H2 analysis b) CTMT H2 - % by Volum _, DETERMINE MAXIMUM H2 VOLUME TO BE VENTED: a) !MX H 2 VOL - ( 3 % - CTMT H2) x CTMT VOL (STP) 100% b) MAX H2 VOL - (3% - %) x( ) CU FT 100% c) 1%X H2 VOL - CU FT
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. . NUMBER ATTAC11 MENT TITLE REVISION
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* *
1-IR H2 VENT TIME CALCULATION ATTACIIMENT PAGE 1 2 of 2 4 _CALCUIATE MAXIMUM VENT PERIOD: a) Determine flowrate from Attaclunent 2, i b) H2 Flow Rate - SCFM, c) Determine vent period 1) VENT PERIOD - MAX H2 VOL -( )CU FT 112 FLO'JRATE ( ) SCFM d) VENT PERIOD - MIN l l l l l l l l l l l
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ATTACHMENT TITLE * REVZSION . . NUMBER TECH SPEC FIGURE 3.1-2 6
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- SENIOR REACTOR OPERATOR Page 38 QUESTION: 056 (1.00)
'
The following conditions exist:
* Unit I has tripped from 100% power due to a large break Loss of Coolant Accident (LOCA).
- Pressurizer pressure is 450 psi * Thot is 430 degrees * Tcold is 400 degrees * Tavg is 415 degrees * CETCs indicate 435 degrees WHICH ONE (1) of the following is the RCS Subcooling Margin? degrees degrees degrees degrees QUESTION: 057 (1.00)
WillCH ONE (1) of the following methods should be used to collapse a void in the RCS per 1-FR-I.3, " Response To Voids In Reactor Vessel"? Decrease RCS temperature by dumping steam while maintaining RCS pressure constan Increase RCS pressure using pressurizer heaters while maintaining pressurizer level constan Start a High Head Safety Injection (HHSI) pump and increase RCS pressure while keeping temperature constan Start all Charging pumps, fill pressurizer solid, then vent the reactor vessel hea .
"
Page 39 SENIOR REACTOR OPERATOR . . QUESTION: 058 (1.00) WHICH ONE (1) of the following is the basis for reducing RCS temperature to less than 500 degrees F. when RCS activity exceeds Technical Specification limits? Slows coolant / fuel reaction rate thereby reducing the RCS source term activit Prevents the release of activity should a Steam Generator tube ruptur Prolongs the life of CVCS demineralizers while RCS cleanup is in progres Minimizes the iodine spiking which contributes to the RCS source term activit l QUESTION: 059 (1.00)
!!HICH OfiE (1) of the following Automatic Safety Features is designed to act to TERMINATE a continuous rod assembly withdrawal from a SUBCRITICAL condition? Power Range High Flux Rod Stop Overpower Delta T Rod Stop Startup High Flux Rate Trip Overtemperature Delta T Trip
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. SENIOR REACTOR OPERATOR
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Page 40 7 I
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QUESTION: 060 (1.00) The following conditions exists:
* Reactor is 30% increasing at 3% per hou * Control rods are in AUTOMATI * Bank "D" rods are at 110 steps and stepping out with N0 demand signa WillCH ONE (1) of the fellowing actions are required to be performed FIRST per AP-1.00, " Rod Control System halfunction"? Place the BANK SEL switch to the CONTROL BANK D positio Place the BANK SEL switch to the MAN positio TRIP the reactor and enter 1-E-0, " Reactor Trip or Safety Injection". Enter 1-FR-S.1, " Response To Nuclear Power Generation /ATWS", to ensure the reactor is TRIPPE .__ _ ~ ~
SENIOR REACTOR OPERATOR Page 41 QUESTION: 061 (1.00) The following conditions exist:
* Unit I was operating at 100% power when an automatic turbine runback was initiate * The Unit is currently operating at 70% power following recognition that a Bank "C" control rod has dropped into the cor * The rod has been in the core for twenty (20) hour WHICH ONE (1) of the following is the proper method of recovery and the reason this method is necessary? Reactor power must be held constant below 75% while the rod is withdrawn at two (2) steps per hour to prevent Xenon oscillation Reactor power is increased to 100% while the rod is withdrawn at two (2) steps per hour to prevent Xenon oscillation Reactor power must be' held constant below 75% while the rod is withdrawn at ten (10) steps per hour to prevent rapid changes in local power densities that could cause DN Reactor power is increased to 100% while the rod is withdrawn at ten 10) steps per hour to prevent rapid changes in local power ensities that could cause DN _ - _ _ _
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* SENIOR REACTOR OPERATOR Page 42
.. . QUESTION: 062 (1.00) WHICH ONE (1) of the following indications distinguishes between an OPEN Pressur.izer PORV and an OPEN Pressurizer Code Safety valve while operating at 100% power? PRT HIGH TEMPERATURE alarm Pressurizer Code Safety valve position indicator Acoustic monitor Tail pipe temperature QUESTION: 063 (1.00) WHICH ONE (1) of the following is the Reactor Coolant Pump Trip Criteria for a small break Loss of Coolant Accident (LOCA)? HHSI flow indicated to cold legs AND RCS Subcooling less than-30 degrees ALL Charging pumps operating as indicated by pump amps and RCS Subcooling less than 30 degrees HHSI flow indicated to cold legs AND RCS pressure less than 1350-psig, ALL Charging pumps operating as indicated by pump amps and RCS pressure less than 1350 psi _
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SENIOR REACTOR OPERATOR Page 43 I QUESTION: 064 (1.00) WHICH ONE (1)(of the following provides MAXIMUM cooling to the core for the12) hours during first twelve (LOCA)? Reflux boiling Break flow cooling Natural Circulation ; Fallback cooling l QUESTION: 065 (1.00) The following conditions exist:
* In response to a large break LOCA a transition from 1-E-0, " Reactor Trip or Safety Injection" to 1-E-1, " Loss of Reactor or Secondary Coolant" has been performe * Due to a RED path on the CORE COOLING status tree, a transition to 1-FR-C.1, " Response to Inadequate Core Cooling" has been performe * During performance of 1-FR-C.1, you observe that the CORE COOLING status tree has changed from a RED to a YELLOW condition, while you identify a RED path on the CONTAINMENT status tre WHICH ONE (1) of the following is the proper procedural transition, and why? l l Complete 1-FR-C.1; since it was entered due to a RED path, it must be completed unless a higher priority path occur Immediately transition to 1-FR-Z.1, " Response to Containment High Pressure", since a RED path is a higher priority than a yellow pat Complete 1-FR-C.1; since once ANY FR is entered, it must be completed before any other transition can be mad Perform the actions of 1-FR-C.1 and 1-FR-Z.1 simultaneously, since FR procedures of the same priority can be executed together.
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Page 44 - SENIOR REACTOR OPERATOR
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QUESTION: 066 (1.00) WHICH ONE (1) of the following is the reason for establishing hot leg recirculation following a large break LOCA? To quench steam in the hot legs and to prevent formation of stratification layers in the cor To quench steam in the hot legs and to prevent boron precipitatio To quench steam in the core and to prevent formation of stratification layers in the cor To quench steam in the core and to prevent boron precipitatio QUESTION: 067 (1.00) WHICH of the following parameters affects vapor entrainment in the RHR suction piping? RHR flow rate AND RCS level RHR flow rate AND RCS pressure Number of RHR pumps running AND RCS pressure Number of RHR pumps running AND RCS level
(\_).
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l - SENIOR REACTOR OPERATOR Page 45 ,
I QUESTION: 068 (1.00) The following conditions exist:
* Unit I has been shutdown for forty (40) hours from an extended operating run at 100% aowe * Loop Stop valves are C 0 SED.
l
* RCS level is at mid-loo *
l Pressurizer PORVs are OPEN.
! * RCS temperature is 140 degrees WHICH expectedONE fo (llowing a complete loss of RHR cooling?1) Assumeof the NOfollowing alternate is the MINIMUM time befo cooling method is established. (1-AP-27.00," Loss of Decay Heat Removal Capability", Attachment 4 is attached) i 1 .5 hours ! .5 hours
; .0 hours .5 hours
l l
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/ tMiBER '" , ATTACIMENT TITLE ,'
REVISION E f-AP-27,00
* *
ATTACIMENT CORE HEAT UP TIMES FOR SURRY
LOSS OF PJLR AT MID-LOOP OPEP.ATION PAGE 140*F INITIAL TEMPERATURE 1 of 1 TIME AFTER LOSS OF RHR (MINUTES) 1000 -
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', SENIOR REACTOR OPERATOR Page 46 QUESTION: 069 (1.00)
WHICH ONE (1) of the following is the MAXIMUM time allowed to establish conditions which will assure acceptable" consequences following an Anticipated Transient Without Trip (ATWT) Event? Turbineistrippedwithinten$0'l is established within ninety ( ) seconds seconds.and Auxiliary Feedwater Turbine is tripped within thirty (30) seconds and Auxiliary Feedwater is established within sixty (60) second Turbine is tripped within sixty (60) seconds and Auxiliary feedwater is established within sixty (60) second Turbine is tripped within ninety (90) seconds and Auxiliary Feedwater is established within ninety (90) second QUESTION: 070 (1.00) WHICH ONE (1) of the following actions should be taken if a Safety injection (SI) signal occurs while performing 1-FR-S.1, " Response To Nuclear Power Generation /ATWS"? Perform 1-FR-S.1 until PGW is isolated then enter 1-E-0, " Reactor Trip Or Safety injection". Perform 1-FR-S.1 and 1-E-0, " Reactor Trip Or Safety injection", simultaneously until Reactor Trip Breakers are OPEN then exit 1-FR- Perform 1-FR-S.1 until Reactor Trip Breakers are OPEN then enter 1-E-0, " Reactor Trip Or Safety Injection". Perform 1-FR-S.1 Immediate Action Steps and then perform 1-E-0,
" Reactor Trip Or Safety Injection", Immediate Action Steps simultaneously with the 1-FR-S.1 procedure.
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O b l *,-SENIOR REACTOR OPERATOR Page 47 l QUESTION: 071 (1.00) The following conditions exist:
* A Steam Generator Tube leak has been confirmed in Steam Generator
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"B".
Steam Generator "A" pressure is 530 ps * Steam Generator "B" pressure is 750 ps * Steam Generator "C" pressure is 530 ps WHICH ONE maintained in (1)the of the RCSfollowing followingisthe theRCS MAXIMUM RCSto temperature maintain a 50that shouldFbe , l cooldown degrees l subcooling margin after RCS depressurization? degrees degrees degrees degrees F.
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SENIOR REACTOR OPERATOR Page 48 QUESTIONi 072 (1.00) The' following conditions exist:
*
The Reactor has been MANUALLY tripped following a rupture in the "B" Steam Generato * SI has automatically actuate * All RCPs are secure *
*
RCS pressure is 1000 psi * RCS is 60 degrees F subcoole * Pressurizer level indicates 0%. RVLIS Full Range indicate 85%.
*
Operators desire to restart "C" RCP per 1-E-3, " Steam Generator Tube Rupture".
WHICH ONE (1) of the following actions must be performed prior to resuming RCP
"C" operation? RCS pressure must be increased to collapse voids in the Reactor Vesse RCP "A" must be started first to establish Pressurizer Spray flo RCS subcooling must be increased to within allowable RCP restart limits, Pressurizer level must be increased to within allowable RCP restart limit i
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O O . Page 49 ]SENIORREACTOROPERATOR QUESTION: 073 (1.00) WHICH ONE (1) of the following is the basis for maintaining Steam Generator-level greater than 11% for a RUPTURED Steam Generator per 1-E-3, " Steam Generator Tube Rupture"? Ensures adequate volume in the Steam Generator for backfil Reduces thermal stresses associated with uncovering tubes in the Steam Generato Prevents ruptured Steam Generator depressurization by establishing a thermal stratification layer above the top of the tube Prevents corrosion of tubes in the ru maintaining a level above the tubes. ptured Steam Generator by QUESTION: 074 (1.00) WHICH ONE (1) of the following differentiates between a unisolable feedline break and an unisolable steam line break of the same size? RCS heat removal would be greater from the steam line brea Containment pressure would be greater for a feedline brea Containment radiation levels would be higher from a steam line brea RC3 depressurization would be greater from a feedline brea _ ___
a~ A~ Page 50 SENIOR REACTOR OPERATOR
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l QUESTION: 075 (1.00)
The following plant conditions exist:
* Unit I has been manually tripped due to a loss of ALL Feedwate ;
* Operators are initiating step 3 of 1-FR-H.1, " Response To Loss Of Secondary Heat Sink" and are tripping ALL RCP WHICH ONE (1) of the following is the basis for tripping all Reactor Coolant '
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Pumps as per this step in the procedure? To reduce decay heat removal in the steam generator ! To reduce Reactor Coolant System pressur To reduce thermal stress to the steam generator To reduce heat input into the Reactor Coolant System.
QUESTION: 076 (1.00) The following plant conditions exist:
* Unit 2 has tripped from 100% power when ALL Feedwater is los * The "3B" Auxiliary Feedwater pump can be returned to servic *
ALL Steam Generators indicate 5% NR leve WHICH ONE (1) of the following describes HOW feedwater flow should be re-established to the Steam Generators per 1-E-1, " Loss Of Reactor Or Secondary Coolant"? Initiate feedwater ONLY to "B" Steam Generator at a maximum flow rate of 350 gp Initiate feedwater to "A", "B", and "C" Steam Generators at a maximum flow rate of 350 gp Initiate feedwater ONLY to "B" Steam Generator at the maximum flow rate possible (700 gpm). Initiate feedwater to "A", "B", and "C" Steam Generators at the maximum flow rate possible (700 gpm).
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, QUESTION: 077 (1.00) WHM ONE (1) of the following is an indication of the loss of 1A DC Bus? #3 EDG automatically start MS-S0V-102B has failed OPE RX Trip Breaker "B" and Bypass Breaker "A" indicating lights are 0F Annunciator Panels A E de-energiz QUESTION: 078 (1.00) WHICH ONE (1) of the following is an Operator Action following a HIGH alarm trip of Process Vent Radiation Monitor Rl-GW-101/1027 Stop Containment Vacuum (CV) pumps to prevent blowing off discharge hos Stop Containment Vacuum (CV) pumps to prevent collapsing suction hose Place filter selector switches for areas tripped to FILTER position to provide filtered releas Place filter selector switches for areas tripped to CLOSE to isolate all release path ,
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QUESTION: 079 (1.00) WHICH ONE (1) of the following is the affect of the failure of the Victoreen Sample pump on the radiation monitoring capability for the Ventilation Vent Stack flow? None, redundant sampling is provided by the Kaman sample pump under all condition * None, redundant sampling is provided by the HP Accountabiiity Sample Rig pump under all condition The Kaman sample pump provides redundant sampling provided stack ' flow is reduced to less than 54,000 CFM and the HP Accountabilit Sample Rig pump is secure The HP Accountability Sample Rig pump provides redundant sampling for all stack flow rates and the Kaman sample pump is secure : QUESTION: 080 (1.00) The following conditions exist:
* Unit 1 is performing Refueling Operation * Fuel movement is in progres * Refueling Cavity level is DECREASING rapidl of the following is the FIRST choice of makeup water to the WHICH RefuelingONE Cav (1)ity ? Charging pump vi i th suction from the VC Diesel Fire pump with suction from the FP-TK-18, HHSI pump with suction from the RWS LHSI pump with section from the RWS I
C O ,SENIdRREACTOROPERATOR Page 53 i l QUESTION: 081 (1.00) WHICH ONE (l;i of the following loads is the FIRST to sequence ON following a loss of offs 1te power? Pressurizer heater loads Outside Recirc Spray pumps 1-RS-P-2A and 2B AFW pumps 1-FW-P-3A and 3B Filter Exhaust fans 1-VS-F-58A and 58B QUESTION: 082 (1.00) The following conditions exist: 1
* A loss of ALL AC power has occurre * The STA reports the status of the CFSs are as follows: -
Subcriticality - RED <
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Core Cooling - RED l
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Heat Sink - GREEN
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Integrity - GREEN
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Containment - GREEN
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Inventory - YELLOW WHICH ONE (1) of the following procedures should be used to mitigate these conditions? FR-C.1, " Response To Inadequate Core Cooling" FR-S.1, " Response To Nuclear Power Generation /ATWS" E-0, " Reactor Trip Or Safety Injection" ECA-0.0, " Loss Of All AC Power"
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* SENIOR REACTOR OPERATOR O O Page 54 a
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l QUESTION: 083 (1.00) WHICH ONE (1) of the following is the reason ALL RCPs are TRIPPED following a non-recoverable loss of Instrument Air?
' To prevent overpressure of Main Steam System if Main Steam Trip valves clos ' To reduce decay heat removal as steam generators overfee Component Cooling Water will be lost to the RCP l l Seal Injection will be lost to the RCP !
l i . QUESTION: 084 (1.00) ) During WHICH ONE (1) of the following conditions can ES-0.0, "Rediagnosis", be used? During the performance of 1-E-0.2, " Natural Circulation Cooldown", when a twenty (20) gpm Steam Generator tube leak is detecte During the performance of 1-E-1, " Loss of Reactor Or Secondary , Coolant", when a RED path is detected in Heat Sin j After transition to 1-ES-0.1, " Reactor Trip Response", when SI was l actuated erroneousl )
, After transition to 1-E-3, " Steam Generator Tube Leak", when ALL AC l power is los '
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SENIOR REACTOR OPERATOR Page 55 . . QUESTION: 085 (1.00) WHICH ONE (1) of the following individuals by title is the MINIMUM authorization that must be obtained before Containment entry during i subatmospheric conditions? 1 Station Manager Shift Supervisor H.P. Supervisor Immediate Supervisor QUESTION: 086 (1.00) The following conditions exist:
* Unit 1 is at 100% powe * A small leak has developed on the charging line to the RCS inside containmen * A repair team has been formed and repairs have been ongoing for the past four hour *
It is now 1300 hour WHICH ONE (1) of the following individuals is allowed to work on the repair team? An operator who has been on site since midnight, A HP Technician who worked the job earlier in the day and left the containment for lunch at 1100 hour ; A maintenance person who has returned to the site after having left at 0800 hours following a sixteen (16) hour shif An untrained Contract Vendor who will be escorted by a station maintenance employee who will assist in the repair .
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- SENIOR REACTOR OPERATOR Page 56 QUESTION: 087 (1.00)
WHICH ONE (1) of the following methods should be used to control the status of an annunciator alarm that has been caused by tagging out a component? Include the reason for the annunciator as part of the control room lo Submit a Request For Engineering Assistance to clear the alar Submit a Work Request to have the annunciator jumpered so as to clear the alar Include the annunciator as part of the Tag-Ou l QUESTION: 088 (1.00) . l WHICH ONE (1) of the following key lockers control keys to Vital Areas? I Security Department Key Locker AND Emergency Key Locker Security Department Key Locker AND Vital Area Key Locker Operation Key Locker AND Emergency Key Locker Operation Key Locker AND Vital Area Key Locker i
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_,SENIOk REACTOR' OPERATOR Page 57 -l QUESTION: 089 (1.00) The following conditions exist:
*
Cubicle 1503 on "D" Bus for Unit 1 is Tagged-Ou * A grounding device was installed by an electrician and independently , verified by a Qualified Operator assigned to perform the Tag-0ut.
l
* The Ground Placement Tag was prepared by a Qualiff ed Operator and ;
j installed by an electricia .j WHICH ONE (1) of the above actions was a VIOLATION of Station Electrical Practices? , The grounding device was installed by an electricia l The grounding device was independently verified by an Operato The Ground Placement Tag was prepared by an Operato The Ground Placement Tag was installed by an electricia QUESTION: 090 (1.00) WHICH ONE (1) ion of a THROTTLED valve?of the following is the PREFERRED method of verifying correct posit I Close and re-open the valve the prescribed number of turn Observe expected flow rate of the syste Open and re-close the valve the prescribed number of turn Observe control room position indication ligh ..
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OV O ' SENIOR REACTOR OPERATOR Page 58 l QUESTION: 091 (1.00) The'fo110 wing conditions exist: ;
* Unit 1 is performing Refueling Operation * The Shift Supervisor needs a control room operator for the u) coming evening shift on a Saturday night to replace an ill crew mem)e WHICH ONE (1) of the following is the PREFERRED operator to fill this vacancy per VPAP-0103, " Working Hours And Limitations"? who has worked his normal dayshift, is willing to hold Operator"A"$eeveningshiftbuthestatesthathecameinonehour over on to t early to relieve an operator for a doctors appointmen Operator "B", who has worked his normal dayshift is willing to hold over on to the evening shift but h(8e hours),
states that he worked eight (8) hours on the evening shift the day before after working day shif Operator "C" who works relief dayshift, is willing to work the
' evening shift but states that he worked an overtime shift two (2)
days ago and it would be his third overtime shift this wee Operator "D", who works the midnight shift, is willing to come in and work the evening shift previous shift for a one (i)but states hour he held training class.over. from his QUESTION: 092 (1.00) , WHICH ONE (1) of the following is required for work conducted in an Extreme - High Radiation Area? RWP briefing by the Shift Superviso Direct audio communication between the control room and the worker at the job sit A RWP approved by the Operations Superintendent, A procedure approved by Station Nuclear Safety and Operating Committee (SNSOC).
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QUESTION: 093 (1.00) The following conditions exist:
* Operator "A" is 25 years old and has received em whole body lifetime exposure prior to 199 * His accumulated whole body dose thgugh 9 /1993 is 3500 mre * His present quarterly whole bod (xpo e is 100 mre * He is needed to perform a high yl aligpd f job that would give him an additional 700 mre // )
WHICH ONE (1) of the followin * 'hr Nilg' action required before Operator
"A" can go to work on t i An extens 6n of yaf limit approved by Health Physic An extension iis yearly limit approved by the Station Manage An extens on of his lifetime limit approved by the Vice President-Nucle (Operation erator "A" must be denied access do to his exposure histor QUESTION: 094 (1.00)
WHICH ONE (1) of the following represents a NON-INTENT Change in a procedure? A change in initial condition A modification to setpoint Deleting a hold poin Correction to step sequenc ; o
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QUESTION: 096 (1.00) WHICH ONE (1) of the following defines an Item Control Area (ICA)? An area established for the purpose of controlling contaminated tools and equipment and is defined by a radiation boundar An area established for the purpose of controlling contaminated tools and equipment and is required to be LOCKE An area established for the purpose of controlling Nuclear Material and is defined by a radiation boundar An area established for the purpose of controlling Nuclear Material' and is required to be LOCKE l l I l l l
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QUESTION: 097 (1.00) WHICH ONE (1) of the following defines MINIMUM reporting requirements to State and Local Governments during an emergency? Notify within fifteen (15) minutes then report conditions ONLY when there is a chang Notify within thirty (30) minutes then report conoitions ONLY when there is a chang Notify within fifteen (15) minutes then report conditions every thirty (30) minute Notify within thirty (30) minutes then report conditions every thirty (30) minute QUESTION: 098 (1.00) The following conditions exist:
*
A Site Evacuation has been ordere *
* One This (1)dividual in is known to be onsite. individual has not reported for accountabilit WHICH of the following are the preferred individuals to conduct the search for the missing individual? Security staff members Health Physics staff members Licensed Operation staff members l Fire Team members i
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k QUESTION: 099 (1.00) . WHICH ONE (1) of the following is the MINIMUM approved method for alternat fire monitoring for containment if detection instruments are INOPERABLE per Technical Specification 3.21.B.1, " Fire Protection Features"? Continuous fire watch Inspection every twenty four (24) hours J Monitor containment air temperatures once per hour Monitor containment humidity once per hour QUESTION: 100 (1.00) WHICH ONE (1) of the following is the MINIMUM number of plant personnel required to be onsite to man the Fire Brigade per Technical Specifications 6.1.B.7, " Organization, Safety, and Operation Review"? , 1 Four (4) which CANNOT include those required for safe shutdow Four (4) which CAN include those required for safe shutdow Five (5) which CANNOT include those required for safe shutdow i Five (5) which CAN include those required for safe shutdow (********** END OF EXAMINATION **********) .
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ANSWER: 001 (1.00) .! [+1.0] l I
. REFERENCE: ND-93.3-LP-3, " Rod Control System", H- ! KA 001000K402 (3.8/3.8)
001000K402 ..(KA's)
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I ANSWER: 002 (1.00) j [+1.0] l REFERENCE:
! P-5.2.1, " Starting Any Reactor Coolant Pump", page . 1-G0P-1.1, " Unit Startup, RCS Heatup From Ambient To 195 Degrees F", page 'i 2 ; KA 003000G005 (3.4/3.8), 003000G010 (3.3/3.6) l
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003000G005 003000G010 ..(KA's) j ANSWER: 003 (1.00) [+1.0] REFERENCE: P.5. " Starting Any Reactor Coolant Pump", page . KA003000Al07(3.4/3.4) 003000A107 ..(KA's) .
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ANSWER: 004 (1.00) [+1.0] REFERENCE: ND-88.3-LP-9, " Blender Control Subsys em", Objective F, page . ND-88.3-LP-10, " Operation Of CVCS", Objective G, page 1 . KA 004020A305 (3.2/3.0) 004020A305 ..(KA's) ANSWER: 005 (1.00) [+1.0] REFERENCE: ND-88.3-LP-10, " Operation Of CVCS", Objective E, page 2 . KA 004010A204 (3.6/4.2) 004010A204 ..(KA's) ANSWER: 006 (1.00) [+1.0] REFERENCE: ND-93.3-LP-3, "" Rod Control System", Objective C, page . KA 015000K302 (3.3/3.5) 015000K302 ..(VA's)
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. . ANSWER: 007 (1.00) [+1.0) REFERENCE: ND-93.4-LP-3, " Inadequate Core Cooling Monitor", Objective F, page 2 . KA 017020A401 (3.8/4.1) 017020A401 ..(KA's) l l ANSWER: 008 (1.00) [+1.0] l REFERENCE: ND-88.4-LP-6, " Containment Ventilation", Objective B, page 6, KA 022000A301 (4.1/4.3) 022000A301 ..(KA's) i ANSWER: 009 (1.00) [+1.0] REFERENCE: ND-89.3-LP-2, " Main Condensate System", Objective F, page 1 . VA 059000G001 (3.1/3.2) 059000G001 . . (VA's )
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ANSWER: 010 (1.00) - [+1.0] l
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ANSWER: 011 (1.00) [+1.0] i REFERENCE: ND-89.3-LP-4, " Auxiliary Feedwater System", Objective H, page 1 . KA 061000K402 (4.5/4.6) ,
-j 061000K402 ..(KA's)
ANSWER: 012 (1.00) a . > c [+1. 0] W o.4-93 l REFERENCE: . ND-92.4-LP-4," KA 068000K107 ( Liq /2.9)uid Waste System", Objective B, page l 068000K107 ..(KA's)
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, REACTOR OPERATOR ANSWER: 013 (1.00) [+1.0] REFERENCE: ND-92.4-LP-2, " Gaseous Waste System", Objective D, page 1 . KA 071000G005 (2.4/3.1) 071000G005 ..(KA's) l ANSWER: 014 (1.00) [+1.0] REFERENCE: ND-92.4-LP-2, " Gaseous Waste System", Objective B, page 1 . KA 071000K504 (2.5/3.1), 071000A429 (3.0/3.6) 071000K504 071000A429 ..(KA's) ANSWER: 015 (1.00) [+1.0] REFERENCE: ND-93.5-LP-1, "Victoreen Area Monitoring System", Objective D, page . KA 072000A401 (3.0/3.3) 072000A401 ..(KA's)
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l l ANSWER: 016 (1.00) ! [+1.0] l l REFERENCE: l ND-93.3-LP-2, " Delta T/Tavg Instrumentation System", Objective E, page 7.
l KA 016000A201 (3.0/3.1)
016000A201 ..(KA's)
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' ANSWER: 017 (1.00) i [+1.0] REFERENCE: ! l ND-93.4-LP-3, " Inadequate Core Cooling Monitor System", pages 11-1 I KA 002000G007 (3.3/3.6) l 002000G007 ..(KA's) l j l ANSWER: 018 (1.00) [+1.0] ' REFERENCE: 1 ND-91-LP-2, " Safety Injection System", Objective E, page 1 ; KA 006020K603 (2.8/3.1)
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l l 006020K603 ..(KA's) l ' I
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', SENIOR', REACTOR OPERATOR ANSWER: 019 (1.00) [+1.0]
REFERENCE: ND-91-LP-2, " Safety Injection System", Objective 0, page 11, Technical Specification 3.3, " Safety Injection System , page 3.3- . KA 006000G005 (3.5/4.2), 006000G0ll (3.6/4.2) 006000G005 006000G0ll ..(KA's) ANSWER: 020 (1.00) [+1.0] REFERENCE: ND-93.3-LP-5, " Pressurizer Pressure Control", Objective C, page . KA 010000G009 (3.6/3.5) 010000G009 ..(KA's) ANSWER: 021 (1.00) [+1.0]
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'SENIO$REACTOROPERATOR ,
REFERENCE:
- ND-93.3-LP-7, " Pressurizer Level Control System", Objective B, page 4.
' KA 011000K404 (3.0/3.3) :; I
)
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011000K404 ..(KA's) l i ANSWER: 022 (1.00) ] [+1.0] REFERENCE: HD-93.3-LP-7, " Pressurizer Level Control", Objective D, page ) KA 011000A101 (3.5/3.6)
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011000A101 ..(KA's) )
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ANSWER: 023 (1.00) [+1.0] j I
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REFERENCE: ND-93.3-LP-17, " Anticipatory Mitigation Actuation Circuitry (AMSAC)", Objective F, page . KA 012000K201 (3.3/3.7) 012000K201 ..(KA's) ANSWER: 024 (1.00) [+1.0]
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REFERENCE: ND-93.3-LP-14, " Overpower /0vertemperature Delta T", Objective H, ND-93.3-H/T-1 . KA 012000K611 (2.9/2.9) 012000K611 ..(KA's) ANSWER: 025 (1.00) [+1.0] REFERENCE: ND-93.3-LP-4, " Rod Position Indication System", Objective C, page . KA 014000A202 (3.1/3.6) 014000A202 ..(KA's) ANSWER: 026 (1.00) [+1.0] , i REFERENCE: ! ND-93.3-LP-3, " Rod Control System", Objective M, page 2 . Technical Specifications 3.12.C., " Inoperable Control Rods", page 3.12-
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1 . KA 014000G011 (3.0/3.9)
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014000G0ll . . (VA's) i
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ANSWER: 027 (1.00) [+1.0] REFERENCE: ND-95.3-LP-54, "FR-1.3, Response To Voids in Reactor Vessel", Objective E, page 3 . KA 016000G015 (3.6/3.8) 016000G015 ..(KA's) AMSWER: 028 (1.00) [+1.0] l REFERENCE: ND-91-LP-5 " Containment Spray", Objective D, page 1 l KA026000Al01(3.9/4.2) 026000A101 ..(KA's) ANSWER: 029 (1.00) [+1.0] REFERENCE:
" ND-88.4-LP-6 KA029000A301(ContainmentVentiletion",ObjectiveE,page .8/4.0) ,
029000A301 ..(KA's)
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hs SENIOR [REACTOROPERATOR Page 73 ANSWER: 030 (1.00)
' [+1.0]
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REFERENCE: 1 AP-22.02, " Loss Of Spent Fuel Pit level", page . KA 033000A203 (3.1/3.5), 033000K107 (2.4/2.5) 033000A203 033000K107 ..(KA's) ANSWER: 031 (1.00) [+1.0] REFERENCE: ND-93.3-LP-8, "S/G Water Level Control System", Objectives C & D, page . KA 035010K401 (3.6/3.8) i 035010K401 ..(KA's) ANSWER: 032 (1.00) [+1.0]
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REFERENCE: ND-90.3-LP-6, "125 VDC Distribution", Objective F, page 11.
l KA 063000G008 (3.1/3.2) 063000G008 ,.(KA's) r ANSWER: 033 (1.00) [+1.0) REFERENCE: ND-90.3-LP-1, " Emergency Diesel Generator-Mechanical", Objective E, page-2 . KA 064000G007 (3.4/3.6)
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064000G007 ..(KA's) ANSWER: 034 (1.00) [+1.0] REFERENCE: ND-93.5-LP-4, "Kaman Process Monitoring System", Objectives D & . Annunciator Response Procedure 1-RM-M7, Condenser Air Ejector - HI", aage . (A 073000K401 (4.0/4.3) 073000K401 ..(KA's) ANSWER: 035 (1.00) [+1.0]
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a O SENIOR REACTOR OPERATOR Page 75
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KEFERENCE: ND-88.2-LP-1, " Residual Heat Removal System", Objective B, page 1 . KA 005000K407 (3.2/3.5) 005000K407 ..(VA's) ANSWER: 036 (1.00) [+1.0) REFERENCE: ND-88.5-LP-1, "Comoonent Cooling", Objective C & F , page . KA 008000K104 (3.3/3.3) 008000K104 . . (VA's) ANSWER: 037 (1.00) [+1.0] REFERENCE: ND-88.4-LP-8, " Hydrogen Recombiners", Objective D, page . 1-E-1, " Loss Of Reactor Or Secondary Coolant", Attachment . KA 028000A401 (4-0/4.0)
. l
028000A401 ..(KA's) ; ANSWER: 038 (1.00) , i [+1.0]
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, SENIOR REACTOR OPERATOR Page 76 l
REFERENCE: ND-92.5-LP-3, " Fuel Handling Tools", Objective E, page 13, KA 034000G005 (2.6/3.5) l
034000G005 ..(KA's) : ANSWER: 039 (1.00) [+1.0]
REFERENCE:
ND-93.3-LP-9, " Steam Dump Control System", Objective D, page 1 ' . KA 041020K418 (3.4/3.6), 041020K414 (2.5/2.8) ,
041020K418 041020K414 ..(KA's) ANSWER: 040 (1.00) [+1.0] i I REFERENCE: ND-89.5-LP-2, " Service Water System", Objective D, page . KA 076000A401 (2.9/2.9), 076000G005 (2.8/3.2) i l 076000A401 076000G005 ..(KA's) ANSWER: 041 (1.00) [+1.0) ,
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O O ' S.ENIOR JtEACTOR OPERATOR
~Page 77 L
REFERENCE:
. AP-39.0, " Natural Circulation Of RCS", Attachment 1, page t KA 000015Al21 (4.4/4.5)
;
000015A121 ..(KA's) l
.;
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I , ANSWER: 042 (1.00) [+1.0) .j
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l- REFERENCE: ND-95.1-LP-3, " Partial loss Of RCS Flow", Objective B, pages 9-1 . KA 000015K104 (2.9/3.1)
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000015K104 ..(KA's) j ANSWER: 043 (1.00) [+1.0] REFERENCE: AP-3.00, "Emer y Boration", page . KA000024K301(4?$4.4) ; i 000024K301 ..(KA's) ANSWER: 044 (1.00) I [+1.0] l
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O O SENIOR REACTOR OPERATOR Page 78
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REFERENCE: FR-S.1, " Response To Nuclear Power Generation /ATWS", page . KA 000024A205 (3.3/3.9) , 000024A205 ..(KA's) ANSWER: 045 (1.00) [+1.0) . REFERENCE: ND-88.5-LP-1, " Component Cooling Water", Objective F, page 2 . KA 000026K303 (4.0/4.2) 000026K303 ..(KA's) t ANSWER: 046 (1.00) [+1.0)
,
REFERENCE:
' AP-15.00, " Loss Of Component Cooling Water", page . KA 000026G010 (3.6/3.5)
..
.000026G010 ..(KA's)
ANSWER: 047 (1.00) [+1.0)
.
l l
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O l LO 5ENIOR REACTOR OPERATOR Page'79 ;
, ,
REFERENCE: ND-95.1-LP-13 " Stuck Open Pressurizer Spray Valve", Objective .' l-AP-31.00 "IncreasingorDecreasingRCSPressure",page4- i ; KA000027Gd10(3.7/3.8) 000027G010 ..(KA's) . ANSWER: 048 (1.00) [+1.0] j REFERENCE: ND-95.2-LP-3, " Secondary Breaks", Objective D, page 3.21 , KA 000040K106 (3.7/3.8) ,
,
000040K106 ..(KA's)
ANSWER: 049 (1.00) [+1.0]
!
REFERENCE: ND-95.1-LP-6, " Loss Of Condenser Vacuum", Objective C, page . KA 000051A201 (2.4/2.7)
N 000051A201 ..(KA's)
:
ANSWER: 050 (1.00) [+1.0] , t
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O O-SENIOR REACTOR OPERATOR Page 80
. ..
' REFERENCE: ) ND-95.2-LP-8, " Loss Of All AC Pcwer", page . KA 000055G006 (3.8/4.1) l i
!
' 000055G006 ..(KA's)
ANSWEP.: 051 (1.00) I [+1.0] REFERENCE: ND-95.2-LP-8, " Loss Of All AC Power", Objective A, page . KA 000055G007 (3.6/3.7) j 000055G007 ..(KA's) :
,
ANSWER: 052 (1.00) [+1.0] REFERENCE: FCA-9.00, " Limiting Intake Structure Fire", page . KA 000067G010 (3.3/3.7) 000067G010 ..(KA's)
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ANSWER: 053 (1.00) [+1.0] _
:
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O SENIOR * REACTOR OPERATOR Page 81 REFERENCE: AP-20.01 " Main Control Room Oxygen Monitor - Alarm Or Malfunction", aage . (A 000068G007 (3.4/3.5) 000068G007 ..(KA's) ANSWER: 054 (1.00) [+1.0) l REFERENCE:
) FCA-1.00, " Limiting MCR Fire", page . KA 000068A123 (4.3/4.4)
000068A123 ..(KA's) ANSWER: 055 (1.00) [+1.0) REFERENCE: FR-I.3, "Res onse To Voids In Reactor Vessel", Attachment . KA 000069G012 3.5/3.5) 000069G012 ..(KA s) ANSWER: 056 (1.00) [+1.0]
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O O . 5lENIORhEACTOROPERATORl Page 82 . REFERENCE: 1.- F-2, " Core Cooling", Drawing Number CB38 . Steam Tables
' KA 000074A201 (4.6/4.9)
000074A201 ..(KA's) ANSWER: 057 (1.00)- [+1.0] -REFERENCE: FR-I.3, " Response To Voids In Reactor Vessel", page . KA 000074K311 (4.0/4.4) 000074K311 ..(KA's)
.
ANSWER: 058 (1.00) i [+1.0] ; REFERENCE: 3 I Technical Specification Bases 3.1.D, page 3.1-1 . KA 000076G004 (2.1/3.7) ; 000076G004 ..(KA's) I
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ANSWER: 059 (1.00) [+1.0]
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a_ O 5EN10R,'REACTOROPERATOR Page 83 REFERENCE: ND-95.2-LP-2, " Rod Withdrawal Accident", Objective A, page . KA 00000lG004 (2.8/3.8) 00000lG004 ..(KA's) ANSWER: 060 (1.00) [+1.0] REFERENCE: ND-95.2-LP-2, " Rod Withdrawal Accident", page 1 . AP-1.00, " Rod Control System Malfunction", page . KA 00000lG010 (3.9/4.0)
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000001G010 ..(KA's) ANSWER: 061 (1.00) [+1.0) REFERENCE: ND-95.1-LP-5, " Dropped / Misaligned Control Rod Recovery", Objective , aages 7-1 . (A 00J003G007 (3.4/3.6) 000003G007 ..(KA's) ANSWER: 062 (1.00) [+1.0]
O ([) SENIOR * REACTOR OPERATOR Page 84 REFERENCE: ND-88.1-LP-3, " Pressurizer and Pressure Relief", Objective D, pages 3.24-3.2 . KA 000008A203 (3.9/3.9) 000008A203 ..(KA's) ANS!!ER: 063 (1.00) [+1.0]
'
REFERENCE: ND-95.2-LP-7, " Loss of Reactor Coolant Accident", Objective G, page AIA- Kkb00009A223(2.8/3.3) 000009A223 ..(KA's) . I ANS!!ER: 064 (1.00) [+1.0] REFERENCE: ND-95.2-LP-7, " Loss Of Reactor Coolant Accident", Objectives C & D, pages 18-19 & 2 . KA 000009K101 (4.2/4.7)
000009K101 ..(KA's)
l l l l
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O O Page 85 SINIOR REACTOR OPERATOR
!
l A!1SWER: 065 (1.00) (+1.0) l REFEREllCE: j 11D-95.3-LP-26, " Critical Safety Function Status Trecs", Objective D, l
' pages 9-1 . KA 000011G012 (4.0/4.1)
0000llG012 ..(KA's) ! ANSWER: 066 (1.00) , I
[+1.0]
REFEREf1CE: f10-95.3-LP-ll, "ES-1.4, Transfer To Hot leg Recirculation", Objective B, page , KA 0000llK313 (3.8/4.2)
0000llK313 ..(KA's) I AliSWER: 067 (1.00) [+1.0] 4
REFERENCE: " Loss Of RHR Events", Objective B, pages 24-2 . 110-95.2-LP-12,(3.9/4.3) KA 000025K101 ! 000025K101 ..(KA's)
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* SENIOR REACTOR OPERATOR Page 86
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ANSWER: 068 (1.00) ,
~ [+1.0) .
REFERENCE: NO-95.2-LP-12, " Loss Of RHR Events", Objective B, pages 28-2 . KA 000025G007 (3.4/3.6) ,
!
l l 000025G012- ..(KA's) i l ANSWER: 069 (1.00) .
'
l [+1.0] l~ REFERENCE: ND-95.1-LP-11, " Anticipated Transient Without Trip (ATWT)", Objective C, page 11.14 KA 000029G007 (3.8/4.0), 000029G010 (4.5/4.5) l 000029G007 000029G010 ..(KA's) ANSWER: 070 (1.00) l [+1.0] l
' REFERENCE: ' FR-S.1, " Response To Nuclear Power Generation /ATWS", page . KA 000029G012 (4.1/4.2)-
000029G012 ..(KA's)
<
y - , - . . , - ,
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S'EN10R hEACTOR OPERATOR Page 87 l
ANSWER: 071 (1.00) [+1.0] REFERENCE: AP-24.01, "Large Steam Generator Tube Leak", page . KA 000037A216 (4.1/4.3) 000037A216 ..(KA's) i ANSWER: 072 (1.00) [+1.0] REFERENCE: E-3, " Steam Generator Tube Rupture", page 30, KA 000038A217 (3.8/4.4) 000038A217 ..(KA's) ANSWER: 073 (1.00) [+1.0] REFERENCE: ND-95.3-Lp-13, "E-3 - Steam Generator Tube Rupture", Objective E, page 2 . KA 000038K306 (4.2/4.5) 000038K306 ..(KA's)
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Page 88 S,ENIOR , REACTOR OPERATOR ANSWER: 074 (1.00) [+1.0] REFERENCE: ND-95.2-LP-3, " Secondary Breaks", Objective E, page 3.2 . KA 000054K101 (4.1/4.3) 000054K101 ..(KA's) ANSWER: 075 (1.00) 1 [+1.0] REFERENCE: ND-95.3-LP-41, "FR-H.1, Response To Loss Of Secondary Heat Sink", Objective E, page 1 . KA 000054K304 (4.4/4.6). i 000054K304 ..(KA's) ANSWER: 076 (1.00) [+1.0] REFERENCE: E-1, " Loss Of Reactor Or Secondary Coolant", page . KA 000054A203 (4.2/4.3), 000054K102 (3.6/4.2) 000054A203 000054K102 ..(VA's)
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~S,ENIOR ," REACTOR OPERATOR Page.89
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ANSWER: 077 (1.00) [+1.0] REFERENCE: AP-10.06, " Loss of DC Power", Attachment . KA 000058A203 (3.5/3.9) 000058A203 ..(KA's)
!
ANSWER: 078 (1.00) [+1.0]
,
REFERENCE: ND-93.5-LP-4, "Kaman Process Radiation Monitoring System", Objective D, > page 1 ' M 000060G010 (3.8/3.8)
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000060G010 ..(KA's)
:
ANSWER: 079 (1.00) [+1.0]
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O O S'ENIOR REACTOR OPERATOR Page 90
]
REFERENCE-l AP-5.21, " Radiation Monitor System Ventilation Monitor Malfunction", page . KA 000060K201 (2.6/2.9)
!
l i 000060K201 ..(KA's) l ANS!!ER: 080 (1.00) [+1.0] l l
REFERENCE: AP-22.01, " Loss of Refueling Cavity", page . KA 000036G010 (3.7/3.8) ; i l 000036G010 ..(KA's) l
i l ANSWER: 081 (1.00) ) [+1.0]
.
'
REFERENCE: ND-90.3-LP-7, " Emergency Distribution Protection and Control", Objective E, page ND-90.3/H- . KA 000056A247 (3.8/3.9) l l 000056A247 ..(KA's)
ANSWER: 082 (1.00) [+1.0) i
a Page 91 SENIOR llEACTOR OPERATOR REFERENCE: ECA-0.0, " Loss Of All AC Power", page . KA 000056G012 (3.4/3.6) 000056G012 ..(KA's) ANSWER: 083 (1.00) [+1.0] REFERENCE: ND-95.1-LP-9, " Loss Of Instrument Air", Objective C, page 9.23 KA 000065K308 (3.7/3.9)
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000065K308 ..(VA's) ANSWER: 084 (1.00) [+1.0] REFERENCE: ND-95.3-LP-2, " Emergency Procedure Writer's format", Objective E, page 1 . KA 194001A102 (4.1/3.9) 194001A102 . . (VA's ) ANSWER: 085 (1.00) [+1.0]
i e A SENIOR REACTOR OPERATOR Page 92 ,
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REFERENCE: j SUADM-0-19, " Guidelines, Procedures, and Lituitations for Containment i Entry", page I KA 19400lKll3 (3.3/3.6) I 19400lKll3 ..(KA's) ,
1 ANSWER: 086 (1.00) [+1.0] l REFERENCE: SUADM-0-19, " Guidelines, Procedures, and Limitations for Containment Entry", page . KA 19400lKll3 (3.3/3.6) l 19400lKll3 ..(KA's) ANSWER: 087 (1.00) [+1.0] REFERENCE: OPAP-0006, " Shift Operating Practices", page 1 . KA 19400lK102 (3.7/4.1) 19400lK102 ..(KA's) ANSWER: 088 (1.00) [+1.0]
. _ . __ - _
__ -. O O Page 93-SENIOR llEACTOR OPERATOR REFERENCE: OPAP-0008, " Administrative Control of Keys and Locked Valves and Switches", page . - KA 19400lK105 (3.1/3.4) 19400lK105 ..(KA's) ANSWER: 089 (1.00) [+1.0] REFERENCE: OPAP-0010, "Ta -Outs", pages 14-1 . KA 19400lK107 3.6/3.7)
.
19400lK107 ..(KA's) ANSWER: 090 (1.00)
; [+1.0]
;
REFERENCE: OPAP-0012, " Valve Operations", page 1 i KA 19400lK101 (3.6/3.7)
!
19400lK101 ..(KA's)
ANSWER: 091 (1.00) [+1.0]
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O O
'SE NIOR, REACTOR OPERATOR Page 94 REFERENCE:
' VPAP-0103, "Workin KA 194001A103 (2.5/g3.4) Hours And Limitations", pages 9-1 A103 ..(KA's)'
:
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ANSWER: 092 (1.00) [+1.0] REFERENCE: ; VPAP-2101, " Radiation Protection Plan, page 3 . KA 19400lK104 (3.3/3.5) 194001K104 ..(KA's) ANSWER: 093 (1.00) i [+1.0] p r f p-O c 1 ; REFERENCE: 3 b(5I , " Radiation Protection Plan", page 2 !
'
19400lK103 (2.8/3.4) 19400lK103 ..(KA's) 1' ANSWER: 094 (1.00) 4 [+1.0] j i _ _
. _ _ _ _ _ .
b) () Page 95 " SENIOR [REACTOROPERATOR REFERENCE: VPAP-0502, " Procedure Process Control", page 9, KA 194001A101 (3.3/3.4) 194001A101 ..(KA's) ANSWER: 095 (1.00) [+1.0] REFERENCE: VPAP-1405, " Independent and Simultaneous Verification", page 1 . KA 19400lK101 (3.6/3.7) 19400lK101 ..(KA's) ANSWER: 096 (1.00) [+1.0] REFERENCE: VPAP-1406, " Nuclear Material Control", page 1 . KA 194001K105 (3.1/3.4) 19400lK105 ..(KA's) ANSWER: 097 (1.00) [+1.0]
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$'ENIOR llEACTOR OPERATOR Page 96 REFERENCE: EPIP-1.02, " Response To Notification Of Unusual Event", page . KA 194001A116 (3.1/4.4) 194001All6 ..(KA's) ANSWER: 098 (1.00) [+1.0] REFERENCE: EPIP-5.03, " Personnel Accountability", page . KA 194001All6 (3.1/4.4) 194001All6 ..(KA's) ANSWER: 099 (1.00) [+1.0] REFERENCE: Technical Specification 3.21.B.1, ' Fire Protection Features", page TS 3.21- . KA 194001K116 (3.5/4.2) 19400lKil6 ..(KA's) ANSWER: 100 (1.00) [+1.0]
Aw e w 5,ENIOR[REACTOROPERATOR Page 97 REFERENCE: Technical Specification 3.21.B.1, " Fire Protection Features", page TS 3.21- . KA 194001A103 (2.5/3.4) 194001A103 ..(KA's) l
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(********** END OF EXAMINATION **********)
. . _ - . . .
SENIOR' REACTOR OPERATOR O O Page 1 ANSWER KEY MULTIPLE CHOICE 023 b
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001 a 024 b 002 b 025 a 003 c 026 d 004 c 027 c 005 c- 028 d 006 b 029- b 007 b 030 c 008 d 031 b l 009 c 032 c 010 d 033 a
011 a 034 a 012 a o r- c. g a -o-e 035 d 013 b 036 a 014 b 037 d .015 c 038 b 016 'b 039 b 017 b 040 c
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018 a 041 c 019 a 042 c
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-020 b 043 c
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021 c 044 a
'022 c 045 c
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SENIOR BEACTOR OPERATOR Page 2 ANSWER KEY 046 b 069 b 047 d 070 d 048 c 071 b 049 b 072 d 050 d 073 c 051 b 074 a 052 d 075 d 053 b 076 b 054 c 077 d 055 c 078 a 056 a 079 c 057 b 080 d 058 b 081 c 059 a 082 d 060 b 083 c 061 a 084 c 062 c 085 a 063 a 086 b 064 c 087 d 065 a 088 b 066 d 089 b 067 a 090 b 068 b 091 c
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$ENIOR[REACTOROPERATOR Page 3 ANSWER KEY ,
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m ~092 d 003 : A<.l d 4 w m-9 43 . 094 d 095 c 096- d 097 c 098 d 099 c 100 c (********** END OF EXAMINATION **********)
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e S TEST CROSS REFERENCE Page 1 SR0 Exam PWR Reactor 0rganized by Question Number QUESTION VALUE REFERENCE 001 1.00 8000008 002 1.00 8000014 003 1.00 8000057 004 1.00 8000004 005 1.00 800001 .00 8000036 007 1.00 8000039 008 1.00 8000041 009 1.00 8000054 010 1.00 8000052 011 1.00 8000051 012 1.00 8000060 013 1.00 8000055 014 1.00 8000056 015 1.00 8000003 016 1.00 8000010, 017 1.00 8000013 018 1.00 8000020 019 1.00 8000021< 020 1.00 8000024 021 1.00 8000025 022 1.00 8000026 023 1.00 8000032 024 1.00 8000033 025 1.00 8000030 026 1.00 8000031 027 1.00 8000038 - 028 1.00 8000044 . 029 1.00 8000045 l 030 1.00 8000046 031 1.00 8000028 032 1.00 8000049 033 1.00 8000047 034 1.00 8000001 - 035 1.00 8000018 036 1.00 8000006 037 1.00 8000042 038 1.00 8000048 039 1.00 8000029 040 1.00 8000061 041 1.00 8000074 042 1.00 8000092 043 1.00 8000071 l
,
044 1.00 8000087-045 1.00 8000072 046 1.00 8000073 047 1.00 8000095 048 1.00 8000075 l 049 1.00 8000065< l
l
' ,' ,
TEST CROSS REFERENCE Page 2 SR0 Exam PWR Reactor 1 0r9anized by Question Number
*
l . QUESTION VALUE REFERENCE l j 050 1.00 8000078 l 051 1.00 8000079 1 052 1.00 8000086 4 053 1.00 8000084 1 054 1.00 8000085 055 1.00 8000077 056 1.00 8000069 057 1.00 8000070 058 1.00 8000083-059 1.00 8000062 - 060 1.00 8000063 061 1.00 8000064 062 1.00 8000098 l
'
063 1.00 8000082 064 1.00 8000089 065 1.00 8000090 - 066 1.00 8000091 067 1.00 8000096 068 1.00 8000097 069 1.00 8000067 070 1.00 8000068-071 1.00 8000101 072 1.00 8000099 073 1.00 8000100 - 074 1.00 8000076 075 1.00 8000093 076 1.00 8000094 077 1.00 8000102 078 1.00 8000080 079 1.00 8000081 080 1.00 8000103 081 1.00 8000104 082 1.00 8000105-083 1.00 8000066 084 1.00 8000106 085 1.00 8000108 086 1.00 8000109 087 1.00 8000110 088 1.00 8000112 - 089 1.00 8000114 090 1.00 8000117 091 1.00 8000118 - 092 1.00 8000120-093 1.00 8000121 094 1.00 8000122+ 095 1.00 8000124 096 1.00 8000126 - 097 1.00 8000127' 098 1.00 8000128
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O TEST CROSS REFERENCE O Page 3
,
'
SR0 Exam PWR Reactor 0 r g a n i z b y- Question Number QUESTION VALUE REFERENCE
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099 1.00 8000129 100 1.00 8000130 ibb$bb ______
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1
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.
I i
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b,n Page 4 i TEST CROSS REFERENCE SR0 Exam PWR Reactor 0rganized by KA Group PLANT WIDE GENERICS QUESTION VALUE KA 094 1.00 194001A101 084 1.00 194001A102 091 1.00 194001A103 100 1.00 194001A103 097 1.00 194001All6 098 1.00 194001All6 090 1.00 19400lK101 095 1.00 19400lK101 087 1.00 19400lK102 093 1.00 19400lK103 092 1.00 19400lK104 096 1.00 19400lK105 088 1.00 19400lK105 089 1.00 19400lK107 085 1.00 19400lK113 086 1.00 19400lKll3 099 1.00 19400lK116 PWG Total kh$bb i PLANT SYSTEMS Group I QUESTION VALUE KA 001 1.00 001000K402 003 1.00 003000A107 002 1.00 003000G005 1.00
'
005 004010A204 004 1.00 004020A305 025 1.00 014000A202 026 1.00 014000G0ll 006 1.00 015000K302 007 1.00 017020A401 008 1.00 022000A301 028 1.00 026000A101 009 1.00 059000G001 010 1.00 059000K416 011 1.00 061000K402 032 1.00 063000G008 012 1.00 068000K107 013 1.00 071000G005 014 1.00 071000K504 015 1.00 072000A401 ______
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O TEST CROSS REFERENCE O Page 5
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SR0 Exam PWR Reactor Organized by KA Group PLANT SYSTEMS Group I QUESTION VALUE KA i PS-I Total 19.00 Group II
'
QUESTION VALUE KA
,
017 1.00 002000G007 019 1.00 006000G005 018 1.00 006020K603 020 1.00 010000G009 022 1.00 011000A101 021 1.00 011000K404 023 1.00 012000K201 024 1.00 012000K611 016 1.00 016000A201 027 1.00 016000G015 037 1.00 028000A401 '
'
029 1.00 029000A301 030 1.00
'
033000A203 038 1.00 034000G005 031 1.00 035010K401 033 1.00 064000G007 034 1.00 073000K401 PS-II Total ih$bb 1 Group III .; VALUE KA l QUESTION 035 1.00 005000K407 036 1.00 008000K104 039 1.00 041020K418 040 1.00 076000A401
......
l PS-III Total 4.00 l
......
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PS Total 4b$bb ; I EMERGENCY PLANT EVOLUTIONS
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Group I
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$ TEST CROSS REFERENCE Page 6 l SR0 Exam PWR Reactor 0rganized by KA Group EMERGENCY FLAMT FVOLUTIONS ,
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Group I QUESTION VALUE KA 059 1.00 00000lG004 060 1.00 00000lG010 061 1.00 000003G007 065 1.00 0000llG012 066 1.00 0000llK313 041 1.00 000015Al21 04% 1.00 000015K104 044 1.00 000024A205 043 1.00 000024K301 046 1.00 000025G010 045 1.00 000026K303 069 1.00 000029G007 070 1.00 000029G012 048 1.00 000040K106 049 1.00 000051A201 050 1.00 000055G006 051 1.00 000055G007 052 1.00 000067G010 054 1.00 000068Al23 053 1.00 000068G007 055 1.00 000069G012 056 1.00 000074A201 057 1.00 000074K311 058 1.00 000076G004 EPE-I Total 24$bb Group 11 QUESTION VALUE KA 062 1.00 000008A203 ) 1.00
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063 000009A223 064 1.00 000009K101 068 1.00 000025G012 1 067 1.00 000025K101 047 1.00 000027G010 071 1.00 000037A216 072 1.00 000038A217 l 073 1.00 000038K306 l
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076 1.00 000054A203 074 1.00 000054K101 075 1.00 000054K304 077 1.00 000058A203 078 1.00 000060G010
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) , () e f , , TEST CROSS REFERENCE Page 7 SR0 Exam PWR Reactor 0rganized by KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 079 1.00 000060K201 083 1.00 000065K308 EPE-II Total ib$bb Group III QUESTION VALUE KA 080 1.00 000036G010 081 1.00 000056A247 082 1.00 000056G012 EPE-III Total b$bb ,
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ANSWER SHEET i Hultiple Choice (Circle or X your choice) i If you change your answer, write your selection in the blank.
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