IR 05000354/1998301
| ML20151V815 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151V805 | List: |
| References | |
| 50-354-98-301OL, NUDOCS 9809150081 | |
| Download: ML20151V815 (116) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No: 50 354
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t No: 50-354/98-301 License No: NPF-57 Licensee: Public Service Electric & Gas Company Facility: Hope Creek Nuclear Generating Station Location: Hancocks Bridge, New Jersey 08038 Dates: August 10-12,1998 Examiners: J. Williams, Senior Operations Engineer / Examiner J. Caruso, Operations Engineer /Examinar T. Fish, Operations Engineer / Examiner Approved By: Richard J. Conte, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety l
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98091500C1 9so904 i
(DR ADOCK 05000354 3 PDR ,
EXECUTIVE SUMMARY Hope Creek Generating Station Inspection Report No. 50-354/98-301 Ooerations The four senior reactor operator candidates performed well on both the written and operating portions of the examination, and thus were issued licenses. The candidates were well prepared for the examination, indicating that the facility had thoroughly evaluated the -
knowlsdge and abilities of each candidate in an effort to determine their readiness to sit for an initiil NRC, SRO examination. Crew communications, control board awareness, and crew bilefings were very good.
During th9 review of the applications it was determined that two applicants had medical exams (NRC FORM 396) that were over six months old. When informed of this, the facility promptly completed new medical exams and provided the results to the NRC.
The facility informed NRC of exam overlap between the NRC written operating test and the facility written audit exam. The NRC exam was revised to eliminate the duplication and this resulted in a more valid measure of the candidate's knowledge and abilities, ii
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Report Details 1. Operations 05 Operator Training and Qualifications 05.1 Senior Reactor Operator initial Examinations
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s. Scope The NRC examiners prepared all portions of the exam in accordance with the guidelines in interim Revision 8, of NUREG-1021, " Examiner Standards," and Revision 1 of NUREG-1123, " Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling Water Reactors." The NRC examiners administered all portions of the examination to four senior reactor operator (SRO) candidates. The I applications were reviewed for satisfactory completion of the educational, I experience and medical requirements. The specific qualification requirements stated l in the updated final safety analysis report, the technical specifications and the facility procedures were reviewe b. Observations and Findinas The results of the SRO examination are summarized below:
SRO Pass / Fail l
Written 4/O l Operating 4/0 ;
Overall 4/0 The written examination was reviewed by the facility during the week of July 27, 1998 and administered on August 10,1998. Several changes were made to the exam as a result of this review. The exam consisted of 100 multiple choice questions. An NRC analysis of the exam results did not identify any generic ;
weaknesses in the candidates knowledge or the training program. The candidates I performance on the written exam demonstrated that they were well prepared. The licensee did not make formal comments on the written exam following the administration of the exa The operating portion of the examination was reviewed by the facility the week of July 27,1998 and was administered August 11-12,1998, and consisted of three simulator scenarios and either five or ten JPMs depending upon the applicants status (i.e., upgrade or instant). All JPMs were followed up with two system-related questions. All candidates were also examined using JPMs and/or questions to evaluate the administrative requirement portion of the examination. During the facility exam review, the facility informed the NRC that four JPMs were identical or very similar to ones used on the facility audit exam. These JPMs were changed on the NRC exam to avoid duplicatio . . _
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l Simulator and JPM performance by the candidates was good. Communications were good, including the use of repeat backs. The examiners noted that crew briefings were routinely performed by the SROs. Control board awareness by all of ;
the candidates was evident throughout each of the three scenarios. Control board awareness by the upgrade candidates was excellen A half core ATWS was planned for one scenario, but it did not occur. After the exam it was determined that inserting the malfunction 'CD20' (SDV drain valves fail to close on scram) canceled malfunction 'RPO6'(half core ATWS). The licensee evaluated this problem and found a computer code problem which was correcte The "as run" scenario was completely adequate for examination purpose Hope Creek procedure HC.OP-AP.ZZ-0014(Q), " Personnel Qualification and Training," implements the requirements from the technical specifications and the updated final safety analysis report (UFSAR). These requirements include compliance with ANSI /ANS 3.1-1981 and USNRC Regulatory Guide 1.8. Revision Each candidate met the educational and experience requirements described in the facility procedur During the review of the applications it was determined that two applicants had medical exams (NRC FORM 396) that were over six months old contrary to the Examiner Standards. When informed of this, the facility completed new medical exams and provided the results to the NR c. Conclusions The four senior reactor operator candidates performed well on both the written and operating portions of the examination, and thus were issued licenses. The candidates were well prepared for the examination, indicating that the facility had thoroughly evaluated the knowledge and abilities of each candidate in an effort to determine their readiness to sit for an initial NRC, SRO examination. Crew communications, control board awareness, and crew briefings were very goo During the review of the applications it was determined that two applicants had medical exams (NRC FORM 396) that were over six months old. When informed of this, the facility promptly completed new medical exams and provided the results to the NR The facility informed NRC of exam o.verlap between the NRC written operating test and the facility written audit exam. The NRC exam was revised to eliminate the duplication and this resulted in a more valid measure of the candidate's knowledge and abilities.
l E8 Review of the FSAR l
l While performing the preexamination activities discussed in this report, the examiners reviewed applicable portions of the UFSAR, that related to the selected
- examination questions or topic areas. No discrepancies were identifie . _ _ _ _ . _ _ _ - - _ - - . . - ..
V. Manaaement Meetinos X1 Exit Meeting Summary On August 19,1998 the NRC discussed their observations regarding the examination with
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Hope Creek operations and training management representatives. This discussion was conducted over the telephone. The results of the exam were provided to the licensee at that tim The examiners expressed their appreciation for the cooperation and assistance that was provided during the review and administration of the exam by licensed operator training personnel and operations personnel. The following participated in the exit meetin PARTIAL LIST OF PERSONS CONTACTED HOPE CREEK K. Krueger, Acting Operations Manager A. Faulkner, Operations Superintendent-Training Liaison D. Rein, Operator Training Instructor B. Havens, Operator Training Supervisor (not at exit)
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Attachments:
1. Hope Creek SRO Written Examination w/ Answer Key 2. Simulation Facility Report
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HOPE CREEK SRO WRITTEN EXAMINATION W/ ANSWER KEY I
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U. S. NUCLEAR REGULATORY COMMISSION !
SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1 APPLICANT'S NAME: N ,
FACILITY: HOPE CREEK
REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: Aunust 10,1998 l
INSTRUCTIONS TO APPLICANT:
Use the answer sheets provided to document your answers. Stapte this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the l question. The passing grade requires a final grade of at least 80.00%. Examination papers will be picked up five (5) hours after the examination start l l
w-TEST VALUE APPLICANT'S SCORE FINAL GRADE % l i
100.00 All work done on this examination is my own. I have neither given nor received ai Applicant's Signature
1 HOPE CREEK - SRO EXAM - ANSWER KEY - 8/98 QUES ANS QUES ANS C 2 A A 2 C B 2 A D 2 A D 3 D C 3 C 6s D 3 B C 3 C C 3 C 1 B 3 D 1 C 3 D 1 D 3 A 1 A 3 C 14. gh 3 B 1 D 4 C 1 B 4 D 1 C 4 D 1 D 4 C 1 A 4 A 2 C 4 A 2 D 4 D 2 D 4 B 2 A 4 B 2 C 4 C 2 C 5 D
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. HOPE CREEK - SRO EXAM - ANSWER KEY - 8/98 QUES ANS QUES ANS l 5 B 7 D 5 C 7 D 5 D 7 B 5 B 7 C 5 D 8 B 5 A 8 B 5 C 8 C 5 A 8 D l 5 B 8 B 6 B 8 B 6 B 8 A 6 D 8 B 6 A 8 C 6 B 8 D 6 C 9 A 6 A 9 B 6 A 9 D 6 B 9 C 6 D 9 D 7 B 9 A 7 B 9 A 7 C 9 D 7 A 9 D 7 B 9 A 7 D 10 C
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SENIOR REACTOR OPERATOR Page 2
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ANSWER SHEET Multiple Choice (Circle or X your choice). If you change your answer, write your selection in l the blan MULTIPLE CHOIC 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 abcd 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 l- cd 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 abcd 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
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l SENIOR REACTOR OPERATOR Page 3 ANSWER SHEET l
Multiple 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 l
048 a b c d 071 abcd 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 abcd 059 a b c d 082 a b c d 060 a b c d 083 a b c d M1 a b c d OM 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
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065 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 abcd
l SENIOR REACTOR OPERATOR Page 4 ANSWER SHEET i
Multiple Choice (Circle or X your choice). If you change your answer, write your selection in the blan l 092 a b c d i 093 a b c d
094 a b c d l
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095 a b c d !
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096 a b c d 097 a b c d i
098 a b c d l
099 a b c d 100 a b c d i
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Page 5 NRC t1ULES 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 penaltie ' 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 examinatio . Restroom trips are to be limited and only one applicant at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil OLa J to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the l
examination cover sheet and each ariswer shee . Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . The point value for each question is indicated in parentheses after the questio . If the intent of a question is unclear, ask questions of the examiner onl . When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap pape . Ensure allinformation you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80.00% or greate . There is a time limit of five (5) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoke ..~J w a5 Jr 4D w a m -.5i.4-' . - -J 4 - - - - = . -- -b4.a s'aA--.-a- - 4 M4-. , - -.+ - * J -. ;- - --
SENIOR REACTOR OPERATOR QUESTION: #1 (1.00)
Following a LOCA, the SPDS Cooling System injection Status display has Core Spray labeled as "INJ".
This indication should: be used by the operators as an indication that both core spray subsystems are injecting at their design flow rate, be used by the operators as an indication that at least one core spray subsystem is injecting at its design flow rat only be used with other indications because its based on system flow and the test flow valve being close only be used with other indication because its based on system flow and the test flow valve being ope ANSWER: Question Topic: SPDS for Injections system status REFERENCE: LP-107,Section IV.B.3.a.1, Page 28, fig 22 Learning Objective: R KA: G119 [3.0/3.O] Memory Level Material Required for Examination: No reference Question Source: 10/97 NRC exam
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SENIOR REACTOR OPERATOR QUESTION: #2 (1.0)
WHICH ONE of the following combinations of valve positions can damage a control rod drive if a scram were to o
REGION of the Power to Flow Map. Plant conditions prior to I l the event were as follows:
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Reactor power 90% of rated thermal power.
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APRMs indicated 90% +/- 3%.
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All LPRMs above downscale alarms and below upscale alarms.
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LPRMs near center of core indicate 95% +/- 3%.
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ALL SRMs are fully withdrawn i
WHICH ONE of the following neutron instrumentation responses POSITIVELY indicates reactor instability after core flow first reaches its lowest flow rate?
l APRMs indicate 60% with +/- 4% swings.
I LPRMs near center of the core indicate 70% +/- 4%. Period meter indicates strong negative and positive swings.
! LPRM downscale alarms occur at 5 seconds,30 seconds, and 90 second !
ANSWER: REFERENCE: HC.OP-AB.ZZ-0300, pg 3 and 4; LP-114 Learning Objective: ELO-3 i
KA: G 2.4.4 [4.0/4.3] Memory l
Material Required for Examination: NONE.
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SENIOR REACTOR OPERATOR QUESTION: #41 (1.00)
The plant was operating at 100% reactor power when the "A" reactor recirc pump tripped. The "B" recirc pump is now operating satisfactorily at 50% speed and 24,000 gpm drive flow.
HOW is core flow determined? Directly from Loop *B" flow indicatio By subtracting idle loop flow from the operating loop flow recorde By adding idle loop flow and the operating loop flow recorde Directly from the total core flow recorde l l
ANSWER: '
REFERENCE: HC.OP.SO.BB-0002, pg 33; LP-19, pg 69, Total flow recorde Learning Objective: ELO-R-16 and 2 KA: 295001AA1.01 [3.5/3.6] Higher Order Material Required for Examination: No reference Question Source: new
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SENIOR REACTOR OPERATOR QUESTION: # 42 (1.00)
i The plant is operating at 28% reactor power, all systems operable, when a loss of voltage occurs on the 4.16 KV buses
'(10A401, 10A402, 10A403,10A404) The reactor scram WHICH ONE of the following caused the reactor scram? ! High reactor pressure Turbine control valve fast closure Loss of power to RPS Low RPV water level l
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REFERENCE: LP-22, pgl Learning Objective: ELO-R1 KA: 295003AK305 [3.7/3.7) Higher Order Material Required for Examination: No reference Question Source: new
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SENIOR REACTOR OPERATOR QUESTION: #43 (1.00)
WHICH ONE of the following conditions requires emergency -
depressurization?
Assume a primary system is discharging into the areas liste HPCI (4111) - area temperature is 255 de RCIC (4110) - area temperature is 195 deg, TIPS room ARM reading 1,000 times norma HPCI room ARM reading 750 times norma . SACS *B&D" (4307) area temperature is 150 degrees .
RWCU pipe chase (4402) area temperature is 170 degrees i Core Spray pump rooms "B&D" (4104&4105) area temperatures J are 135 de RHR pump rooms - "B&D" (410 9 &4107 ) area temperatures are 130 de l
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l ANSWER: REFERENCE: LP-127, pg 15 & 29 ; EOP-103, table 1 and step RB/T-13. .
Learning Objective: ELO- KA: 295032EK3.01 [3.5/3.8] Higher Order Material Required for Examination: EOP-103 - white out entry condition Question Source: new i
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l SENIOR REACTOR OPERATOR l QUESTION: #44 (1.00)
A loss of all RPV level indication due to high drywell temperature has occurred. The reactor was successfully scrammed at 1020. The following conditions have existed since 112 SRVs manually opened
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RPV pressure - steady at 120 psig
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Supp chamber water level - 75"
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Supp chamber pressure - 10 psig and stable
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DW pressure - 11 psig and stable
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DW temperature - 195 degrees F. and stable
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Core spray pump "B" injecting
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RPV water level instrumentation is available WHICH ONE of the following actions should be taken? Continue to inject but at time 1240 terminate RPV injection for a maximum of 6 minutes or until RPV level indication is restore Continue to inject but at time 1240 terminate RPV injection for a maximum of 9 minutes or until RPV level indication is restore Continue to inject but at time 1207 terminate RPV injection for a maximum of 6 minutes or until RPV level indication is restore Continue to inject but at time 1207 terminate RPV injection for a maximum of 9 minutes or until RPV level indication is restore ANSWER: REFERENCE: EO-206, step RF-25; LP-134, pg 27, 2 Learning Objective: ELO-6 KA: 295028EK3.02 [3. 5 /3. 8) Higher Order Material Required for Examination: EO-20 Question Source: new t
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SENIOR REACTOR OPERATOR QUESTION: #45 (1.00)
Given the following plant conditions:
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A failure to scram occurred L
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Reactor power is 20%
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Supp pool water temperature is 112 degrees F.
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MSIVs were closed on main steam line High-High Rad l
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2 SRVs are cycling to control reactor pressure
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Drywell pressure is 2.2 PSIG
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RPV level is 50" and slowly lowering WHICH ONE of the following actions is required? Terminate and prevent all injection to the RPV except from CRD and boro Maintain RPV water level between -161" and +54".
c. Maintain RPV water level between -190" and -161".
d. Reopen the MSIVs to reestablish the condenser as a heat l sink.
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ANSWER: a.
l l REFERENCE: EOP-207, step LP-10; LP-135, pg 17.
l Learning Objective: ELO- KA: 295015AK1.03 [3. 8/3. 9) Higher Order Material Required for Examination: EOP-207 with entry conditions remove (..
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QUESTION: #46 (1.00)
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The plant is operating at 70% reactor power when the "A" outboard l l
MSIV fails close WHICH ONE of the following describes the response of the reactor power and reactor water level from steady state to steady state condition?
Assume no operator action is take Reactor power will decrease and stabilize at a lower powe RPV water level will decrease and then return to a normal leve Reactor power will decrease and stabilize at a lower powe RPV water level will increase and then return to a normal leve Reactor power will increase and stabilize at a higher powe RPV water level will increase and then return to a normal leve Reactor power will increase and stabilize at a higher j powe RPV water level will decrease and then return to a normal leve ANSWER: REFERENCE: AB202, sect 4.0; LP-11 Learning Objective: ELO-3 KA: 295020AA203 [3.7/3.7) Higher Order l
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SENIOR REACTOR OPERATOR QUESTION: #47 (1.00)
The following conditions exist:
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Suppression pool water level is 50 inche Suppression pool water temperature is 220 degrees Reactor pressure is 800 psi Under these conditions, Emergency Depressurization is: Not required since primary containment limits are not exceede Required to ensure that the suppression chamber design temperature is not exceeded, Prohibited since the downcomers are now exhausting to suppression pool free air spac Prohibited since the HPCI is now exhausting to the suppression chamber free air spac ANSWER: b Downcomers uncovered at 38.5" and HPCI exhaust uncovered at 25".
REFERENCE: EOP-102A, SP/T-9 l Learning Objective: LP-125A, ELO-9.
i KA: 295026K301 [3.8/4.1] Higher Order
Material Required for Examination: EOP Question Source: New l
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SENIOR REAC'f0R OPERATOR QUESTION: #48 (1.00)
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The plant is operating at 100% reactor power when a "CRD HYD UNIT TEMP HI" alarm is received.
l WHICH ONE of the following caused this condition?
! ! Eroded CRD cooling water orifice Leaking scram outlet valve CRD flow control valve fails open j Leaking scram inlet valve
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l REFERENCE: LP-6, table 2; ARP C6-C3, causes.
l l Learning Objective: ELO-18 & R2 KA: 295022AK302 [2.9/3.1] Higher Order Material Required for Examination: no reference.
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1 l QUESTION: #49 (1.00) l A fire protection header rupture has resu?.ted in 10 inches of 1
. water in the HPCI (4111) and RCIC (4110) rooms . All appropriate '
systems have been isolated and all sump pumps are operating but l level remains at 10 inche !
WHICH ONE of the following actions should be taken? A scram should be initiate j Recirc should be run back to minimu A normal shutdown should be commence No action is require .
ANSWER: REFERENCE: EOP-103, steps RB/L-10 & 11; LP-127, pg 27-2 Learning Objective: ELO- KA: 295036EK201 [3 .1/3. 2] Higher Order Material Required for Examination: EOP-103 with the entry conditions remove Question Source: new
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l SENIOR REACTOR OPERATOR QUESTION: #50 (1.00)
WHICH ONE of the following thermal limits are related to power oscillations at high power / low flow conditions? APLHGR LHGR CMFLPD l
i MCPR !
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ANSWER: d REFERENCE LP-106, pg 27-3 Learning Objective: ELO- KA:295014AK105 [ 3.7/4.2] Memory Material Required for Examination: nu referenc Question Source: new l
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SENIOR REACTOR OPERATOR QUESTION: 51 (1.00)
A LOCA has occurred and the following conditions exis Reactor pressure is 400 psig
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Reactor is shutdown
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Drywell pressure is 8.5 psig ,
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Drywell temperature SPDS point A2266 is 475 degrees )
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Drywell temperature is 270 degrees F. for all other i SPDS points
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Instrument reference leg temperatures are
<300 degrees F WHICH ONE of the following instruments would be the most accurate reactor vessel level indication under the listed conditions? LR-R623A-B21 Wide Range "A" LR-R623B-B21 Wide Range "B" LR-3683B Narrow Range "B" LR-3683A Narrow Range "A" ANSWER: b.
l REFERENCE: EOP caution 1; LP-126A.
l Learning Objective: LO-1c.
i KA:295012AK101 [3.3/3.5] Higher Order Material Required for Examination: EOP caution Question Source: new l
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l l SENIOR REACTOR OPERATOR QUESTION: 52 (1.00)
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WHICH ONE of the' following is the basis for the limit of 125" in the suppression pool? SRV tailpipes will be submerge Supp chamber vent path will be uncovere I Supp chamber to drywell vacuum breaker inlets are submerge Vent header drain lines will be submerge ANSWER: REFERENCE: EO-102B, step SP/L-24; LP-125B, pg 2 Learning Objective: ELO- KA: 295029EK101 [3.4/3.7) Memory Material Required for Examination: no referenc Question Source: new l
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SENIOR REACTOR OPERATOR QUESTION: #53(1.00)
Which of the following describes the effect of f ailing to restart the Turbine Building Ventilation System if it trips while operating in HC.OP-EO.ZZ-0104(Q)-FC, " Radioactive Release Control"? Assume a release in the turbine building is in progress, The Turbine Building releases will be monitored but not treate The total off-site calculated release could be higher, The Turbine Building will go to a slightly negative pressur The total off-site calculated release could be lowe ANSWER: REFERENCE. LP-128, Page 8, L.O. R-3 KA: 295038EA1.06[3.5/3.6) Higher order Material Required for Examination: No Referenc Question Source: HC requal bank
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SENIOR REACTOR OPERATOR QUESTION: 54 (1.00)
Due to a transient an offsite release is in progress. A sample analysis of the discharge as well as a projected offsite dose calculation has been done with the following result Noble gas at the site boundary will result in a dose rate of 2 rem / year total body
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The projected duration of the gaseous release at the site boundary will result in a TEDE of 150 mrem
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The projected duration of the gaseous release at the site boundary will result in a thyroid CDE of 400 mrem WHAT is the emer'ency plan classification of this event? General Emergency Site Area Emergency c. Alert i
d. Unusual Event ANSWER: REFERENCE: ECG, classification sect. 6.0; LP-0215-008.00B-00080 Learning Objective: LO KA: 295017AK2.06 [3.4/4.6] Higher Order
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Material Required for Examination: EALs only portion of ECGS.
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SENIOR REACTOR OPERATOR i l
1 QUESTION: 55 (1.00) '
Condenser backpressure is rapidly increasing with reactor power
.at 90%. !
WHICH ONE of the following is the required immediate operator action for this event per HC.OP-AB.ZZ-208(Q)? Reduce reactor power as necessary to maintain condenser pressure less than 6 inches HgA by inserting control rod Ensure proper operation of the steam jet air ejectors (SJAE) Ensure turbine steam seal pressure is norma Reduce reactor power as necessary to maintain condenser pressure less than 5 inches Hg .
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I ANSWER: REFERENCE: AB-208; LP-11 Learning Objective: ELO- KA: 295002AK3.09 [3.2/3.2] Memory 1 Material Required for Examination: no referenc ;
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j SENIOR REACTOR OPERATOR I
QUESTION: 56 (1.00)
l The-reactor is at 100% powe All systems are operable and in their normal alignmen I l WHICH ONE of the following would be the plant response to a loss l of the total steam flow signal to the feedwater level control system? Assume no operator action is take Vessel level control will automatically swap to single element control and control level in the normal band, Vessel >1evel will increase until the main turbine trip I Vessel level will decrease until the reactor scram ; Vessel level control will remain in 3 element control l and control level in the normal band.
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l REFERENCE: LP-59, pg 20; AB-200, sect 5.2; ARP B3-F Learning Objective: ELO-13.
! KA: 295009AA202 [3.6/3.7] Higher Order Material Required for Examination: no reference.
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QUESTION: 57 (1.00)
Given the following plant conditions: j
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Reactor pressure 500 psig l
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Reactor level 125 inches l
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Supp pool water temperature 182 degrees j WHICH ONE of the following bands contain the actual Heat Ccpacity Level Limit? " - 34" " - 37" " - 40" " - 43" ANSWER: REFERENCE: EOP-102A, step SP/L-5; LP-125B, pg 15-1 Learning Objective: ELO-R KA: 295030EK103 [3.8/4.1] Higher Order Material Required for Examination: provide EOP Question Source: new l
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SENIOR REACTOR OPERATOR QUESTION: 58(1.00)
A primary containment atmospheric sample indicates a hydrogen concentration of 2.2%.
WHICH ONE of the following actions are required per the EOPs?
Assume'all other reactor parameters are stable and are being controlled as required by the EOPs. All systems are operable and NO OFFSITE RELEASES ARE IN PROGRESS also offsite releases are expected to remain below the LCO limit during ventin Vent the supp chambe Vent the drywel Purge the supp chambe Purge the drywell.
ANSWER: a.
REFERENCE: EO-102B; LP-126C, pages 23-25.
Learning Objective: ELO-8.
KA: 500000EK3.08 [3.1/3. 6] Higher Order Material Required for Examination: EOPs Question Source: new o
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SENIOR REACTOR OP&'ATOR QUESTION: 59(1.00)
A loss of drywell cooling results in a drywell pressure reaching 2.6 psi WHICH ONE of the following describes EDG, RCIC, and RWCU expected response? Assume no operator action has been take EDGs - running and loaded RCIC - running and injecting RWCU - pumps tripped EDGs - running and NOT loaded RCIC - not affected RWCU - not affected EDGs - running and loaded RCIC - not affected RWCU - pumps tripped EDGs - running end NOT loaded RCIC - running and injecting RWCU - not affected ANSWER: REFERENCE:'AB-201; LP-11 Learning Objective: ELO- KA: 295010AK1.03 [3.2/3.4) Higher Order Material Required for Examination: no referenc Question Sou."ce: new w s ,
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SENIOR REACTOR OPERATOR *
I- QUESTION: 60 (1.00)
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! The MSIVs have closed and the reactor has scrammed due to a loss ( of condenser vacuum. Current plant conditions are as follows:
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All rods in
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Mode switch in Shutdown l
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Reactor level is 145 inches l
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Reactor pressure is 900 psig
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Drywell pressure is 1.4 psig
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Supp_ pool water temperature is 125 degrees WHICH ONE of the following is the Technical Specification action required at this time? Reduce the pool temperature to less or equal to 95 degrees F within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Reactor shall be depressurized to less than 200 psig within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The reactor shall be in a cold shutdown condition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Maintain Primary Containment integrity until pool temperature is reduced below 95 degrees !
ANSWER: REFERENCE: TS 3.6.2.1, action b.3.; LP-125A, pg 1 Learning Objective: ELO- KA: G2.1.12 [2.9/4.0] Higher Order l
Material Required for Examination: TS.
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l SENIOR REACTOR OPERATOR l QUESTION: 61 (1.00)
Given the following plant conditions:
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Reactor'is shutdown, all rods in
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RPV pressure is O psig
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Core Spray system "A" is injecting to the vessel
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Core Spray system *B" is injecting to the vessel
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CRD is injecting to the vessel
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All available normal injection subsystems are lined up and injecting
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RPV level is -161" and decreasing WHICH ONE of the following actions is required per EOPs? l i
a. Perform Primary containment flooding b. Perform emergency depressurization c. Perform Steam Cooling d. Line up alternate injection systems ANSWER: REFERENCE: EOP-201; LP-129, pg 2 Learning Objective: ELO- I KA: 295031EK302 [4 . 4 /4 . 7) sligher Order Material Required for Examination: EOPs Question Source: new l
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QUESTION: 62 (1.00) i l
All three primary and secondary condensate pumps are initially operating with reactor power at 40% when "A" & "B" secondary l condensate pumps both tri "A" & "B" feedwater pumps were l operating with "C" feedwater pump down for maintenanc l WHICH ONE of the statements below describes the response of the the feedwater pumps and the reactor recirc pumps assuming no operator action is taken? When the "A" and "B" secondary pumps trip both operating feedwater pumps trip on a signal from the secondary condensate pumps. Both reactor recirc pumps runback to 30% spee When the A" and "B" secondary pumps trip only one of the operating feedwater pumps trips on a signal from the secondary condensate pumps. Both reactor recirc ,
pumps runback to 45% spee When the "A" and "B" secondary pumps trip both l operating feedwater pumps trip on a signal from the (
secondary condensate pumps. The reactor recirc pumps will not runbac When the "A" and "B" secondary pumps trip only one of the operating feedwater pumps trips on a signal.from the secondary condensate pumps. The reactor recirc pumps will not runbac ANSWER: REFERENCE: LP-19, pg 59-60, 65; LP-58, pg 82, ELO-2 Learning Objective: ELO-1 .
KA: 256000K3.04 [3.6/3.2] Higher Order Material Required for Examination: no reference.
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SENIOR REACTOR OPERATOR l
QUESTION: #63 (1.00)
Given the following conditions:
- A plant startup is in progress with the Reactor Mode Switch in "Run"
- The Main Turbine is reset and is at 950 rpm accelerating to 1800 rpm
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- A loss of 125 VDC power from distribution panel 1CD318 to the EHC control logic l occurs Which of the following is the expected plant response?
I Main turbine trip Main turbine startup would continue at the selected acceleration rat j Main turbine speed will remain constant at 950 rp Main turbine control valves throttle closed due to a loss of the speed reference signa ANSWER: a Question Topic: Reactor Scram From Turbine Trip During Startup Due to Loss of 125 l VDC REFERENCE: LP-51, Page 33-34, L.O.12 KA: 295004K203 [3.3/3.31 Higher Order Material Required for Examination: No Reference Question Source: New l
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SENIOR REACTOR OPERATOR QUESTION: #64 Given the following conditions:
- The plant is operating at 35% power j * A main generator load reject has just occurred l
- The power / load unbalance circuit tripped unexpectedly during the load reject f WHICH ONE of the following is the expected response of the Turbine Control Valves and the Reactor Protection System (RPS) for the given conditions?
- The Turbine Control Valves throttle closed
-- RPS does not cause a SCRAM The Turbine Control Valves fast close
- RPS does cause a SCRAM The Turbine Control Valves throttle closed
- RPS does cause a SCRAM The Turbine Control Valves fast close
- RPS does not cause a SCRAM l
ANSWER: b ;
Question Topic: Power / Load Unbalance Trip - Expected Plant Response REFERENCE: LP-48, Page 9 Learning Objective: ELO-R10.b& 1 KA: 295005K201 [3.8/3.9) Higher Order Material Required for Examination: No Reference Question Source: New l
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QUESTION: #65 With the plant at 100% power, a severe overfeeding transient is occurring. Water level is
+ 50 inches and rising rapidl WHICH ONE of the following reactor water levels require termination of all feed to the reactor, closing the MSIVs and a reactor scram assuming none of these actions have occurred? + 54 inches + 65 inches + 90 inches + 118 inches 2 ANSWER: c Question Topic: Actions for Severe Reactor Overfeeding REFERENCE: AB-0200, Page Learning Objective: LP-114, L.O. ELO-1 & 3 KA: 295008AA2.01 [3.9/3.9] memory Material Required for Examination: No Reference Question Source: HC Bank Recual l
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l SENIOR REACTOR OPERATOR QUESTION: #66 Given the following conditions:
- SACS pumps "A" ano "C" are running, supplying TACS load * SACS pump "B" is run7in * SACS pump "D" is in s*andb * Due to improper testing, a low SACS pump "A" differential pressure signal is generated.
l' Following all automatic actions, identify the running SACS pumps and which SACS loop is supplying TACS loads.
l SACS pumps "B", and "D" are running and SACS loop "B"is supplying TACS loads.
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, SACS pumps "B", and "D" are running and both SACS loops are supplying TACS load SACS pumps "A", "B", "C", and "D" are running and SACS loop "B" is supplying TACS load SACS pumps "B", "C", and."D" are running and SACS loop "A" is supplying TACS load ANSWER: a l Question Topic: Status of SACS Following a Low Pump Differential Pressure Signal
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i REFERENCE: LP-80, Pages 18-24; L.O. ELO-16 KA: 295018K101 [3.5/3.6] Memory Material Required for Examination: No Reference
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SENIOR REACTOR OPERATOR QUESTION: #67 Given the following conditions:
- A loss of coolant accident has occurred and all systems responded normall * The Reactor Auxiliaries Cooling System (RACS) has been restored.
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WHICH ONE of the following describes the availability / response of the Emergency Instrument Air Compressor (EIAC) for these conditions should instrument air header
- pressure begin lowering?
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l The EIAC is not availa'le until the LOCA signalis reset or bypassed, and the l 1E breaker is closed.
I The EIAC is not available until the LOCA signal is reset or bypassed, and the non-1E breaker is close The EIAC will automatically start on instrument air header pressure less than
! 85 psig, The EIAC will not automatically start but may be started manually from the
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Control Room.
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ANSWER: a Question Topic: Emergency Air Compressor Responses ;
REFERENCE: LP-75, Page 43; ELO-7 l KA: 295019A101 (3.5/3.3] Higher Order Material Required for Examination: No Reference
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SENIOR REACTOR OPERATOR QUESTION: #68(1.00)
The plant has just reached COLD SHUTDOWN when an inadvertent shutdown cooling isolation occurs. HV-FOO8 (RHR shutdown cooling isolation valve) cannot be reopened. For these conditiens, decay heat should be removed by filling and venting the suction line, i starting a Reactor Recirculation pump, and:
! Opening the MSIVs, starting a mechanical vacuum pump and dumping steam to the main condenser, f Verifying the RWCU system is in service with maximum RACS flow to the non-regenerative heat exchange Raising RPV level to + 80 inches on the Shutdown Rang Placing alternate shutdown cooling in service using "C" to "A" RHR cross-tie.
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ANSWER: b JUSTIFICATION: Opening the MSIV's, starting a mechanical vacuum pump and dumping steam to the main condenser. - Alternate means are available and the plant is in cold shutdown -
no stea Verifying the RWCU System is in service with maximum RACS flow to the non-regenerative heat exchanger. - OK per step 4.7.a of AB-14 Raising RPV level to +80 inches on the Shutdown Range. - Not done if RR is starte Placing Suppression Pool Cooling in service, opening two SRV's and circulating the suppression pool through the reactor using and RHR pump. - Done only if all other possibilities are exhauste REFERENCE: AB-142, Pages 2 and Learning Objective: LP-28, ELO-R1 KA: 295021 AK3.05[3.6/3.8] Higher Order Material Required for Examination: No Referenc Question Source: HC Requa! Bank I
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SENIOR REACTOR OPERATOR l QUESTION: #69 (1.00)
l l Given the following conditions:
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- The plant has been operating at 100% power for several weeks.
l * Main Steam Line (MSL) radiation levels have been averaging 95 mR/hr but are slowly l trending upward * Before receipt of the MSL Hi Radiation alarm,it is noted that offgas activity and reactor coolant conductivity are both increasing.
l What are the immediate Operator Actions required for the given conditions? Place additional Condensate Demineralizers in service if possibl Scram the reactor and close the Main Steam isolation Valves when MSL levels are greater than 142 mR/h Direct Reactor Water Cleanup flow to the main condenser to reduce coolant activit Reduce reactor power to maintain MSL radiation levels less than 142 mR/h ANSWER: d Question Topic: Actions for MSL Hi-Hi Radiation REFERENCE: AB-100, Page 1; LP-221, pg 52-54.
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Learning Objective: LP-114, L.O. ELO- KA: 2950033EK1.03 [3.9/4.2] Memory Material Required for Examination: No Reference Question Source: HC requal bank l
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SENIOR REACTOR OPERATOR QUESTION: #70 (1.00)
Given the following conditions:
- The plant is or erating at 50% power
- All systems are operating normally
- One Reactor Building Ventilation System (RBVS) Exhaust Fan discharge damper has failed to the full "open" position with the fan running
- No other RBVS components have changed Which of the following describes how this will affect the initiation of the Emergency Core Cooling Systems (ECCS) and the reason for this? ECCS willinitiate after it is required because the failed damper raises Reactor Building pressure resultirg in a lower indicated drywell pressure, ECCS willinitiate before it is required because the failed damper lowers Reactor Building pressure resulting in a higher indicated drywell pressur ECCS will initiate after it is required because the failed damper lowers Reactor Building pressure resulting in a lower indicated drywell pressure, ECCS will initiate before it is required because the failed damper raises Reactor Building pressure resulting in a higher indicated drywell pressur ANSWER: b REFERENCE: AB-11 Learning Objective: LP-114, ELO-3 KA: 2950035EA1.01 (3.6/3.6] Higher Order Material Required for Examination: No Reference Question Source: NRC exam 2/98.
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SENIOR REACTOR OPERATOR QUESTION: #71 (1.00)
The reactor is in OPCON 5 for a short outage. During a containment inspection, the Operations Superintendent notices a radiation barricade rope around an area. A radiation sign on the rope reads " Caution - High Radiation Area" and indicates a maximum radiation level of 900 mrem /hr inside the roped are Which one of the following additional controls should be used for this area? The area should be kept locked and the keys kept under the administrative control of the Operations Superintendent, The area access should be controlled by issuance of a Radiation Work Permi The area should have a flashing light in tha immediate area as a warning devic The area should have a closed circuit TV monitor installed to provide radiation protection personnel with continuous monitoring capabilitie l
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ANSWER: b REFERENCE: NC.NA-AP.ZZ-0024, Pages 17-18; L.O. LP-113, ELO-52 KA: G.2.3.10 [2.9/3.3) Memory Material Required for Examination: No Reference Question Source: HC Bank t-
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SENIOR REACTOR OPERATOR QUESTION: #72 (1.00)
WHICH ONE of the following persons is authorized to provide selease authorization for a tag to be removed from a piece of l&C equipment during non-emergency conditions when the responsible individual can not be contacted? Only responsible individua Control Room Superviso Discipline superviso Operations Superintenden ANSWER: c Question Topic: Tag Removal REFERENCE: NC.NA.AP.ZZ-0015, Page 6, Section 5.1.1 Learning Objective: LP-113, ELO-4 KA: G.2.2.13 [3.6/3.8] Memory Material Required for Examination: No Reference Question Source: HC requal bank
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SENIOR REACTOR OPERATOR QUESTION: #73 0.00)
LAW NAP-13, Control of Temporary Modifications, WHICH ONE of the following modifications requires a 10 CFR 50.59 Safety Review prior to implementation? Installation of a temporary space heater to prevent freezing of a safety related system component during inclement weathe Connection of a sample tube to a sampling connection to obtain an RHR system sampl Installation of a pressure gauge on an instrument tap during the conduct of a
- system pressure test.
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! Hookup of an air supply hose to a station air manifold during maintenanc ANSWER: a.
i j Question Topic: Temporary Modification
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REFERENCE: NC.NA-AP.ZZ-OO13, Section 5. Learning Objective: LP-113, ELO-43.
l KA: G.2.2.11 [2.5/3.41 Memory l Material Required for Examination: No Reference
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QUESTION: #74 (1.00)
A RWCU Backwash Receiver Tank drain line has cracked during a transfer to Radwaste and is spilling into the Reactor Building. Reactor Building Area Radiation conditions are as follows:
Beginning of Shift Current l Reactor Buildina Area Area Radiation Area Radiation
- RWCU Pump Room 2 mr/hr 2400 mr/hr
- RWCU Heat Exchanger Room 3 mr/hr 1100 mr/hr
- RWCU Backwash Tank Room 3.5 mr/hr 4500 mr/hr
- General Area Outside of RWCU Backwash Tank Room 1.2 mr/hr 1000 mr/hr
- Other Reactor Building Areas 2 to 5 mr/hr Not in Alarm WHICH ONE of the following is the required action that must be directed by the NSS? Continue reactor operation and attempt to stop the tank leakag Commence a normal reactor shutdown to cold shutdow Scram the reactor and commence a normal cooldow Scram the reactor and commence an emergency depressurizatio ANSWER: b Question Topic:
REFERENCE: EOP 103, Step RB/R-9 Learning Objective: LP-127, LO- KA: G2.3.10 [2.9/3.3] Higher Order Material Required for Examination: EOP Question Source: NRC Exam Bank
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SENIOR REACTOR OPERATOR QUESTION: #75(1.00)
Plant conditions are as follows:
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All control rods have fully Inserte Reactor water level is -240 inches and decreasin Suppression pool level is 25 inche Containment radiation monitors indicate 20 R/hr and increasin Only one SRV can be opene WHICH ONE of the following states the alternate methods of rapidly depressurizing the reactor that would MINIMlZE the radioactive release to the environment? Turbine bypass valves, RCIC, HPCI, and RFPT Head spray, RFPTs, main steam line drains, and SJAE HPCI, RCIC, and head ven SJAEs, main steam line drains, and head ven l ANSWER: REFERENCE: EOP-202, Step ED-1 Learning Objective: LP-130, LO-3 KA: G2.3.11 [2.7/3.2] Higher Order Material Required for Examination: EOPs Question Source: NRC Exam Bank
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l QUESTION: #76 (1.0)
Emergency diesel generator "A" fails to start during a Loss of Offsite Power (LOP) and cannot be started manually. Emergency diesel generator "C" starts and loads and then trips. Emergency diesel generators "B" and "D" start and load as designe At this point, the emergency instrument air compressor:
i I Will start automatically on low air header pressure, Will not automatically start on low air header pressure, but can be manually l
started by the operator.
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! Will not start due to a loss of RAC l Will not start due to loss of power to the compresso ANSWER: d Question Topic: loss of electrical power for compresso PEFERENCE: LP-68, Pages 12 & 42, L.O. ELO-Sc
KA: 300000K2.02 [3.0/3.01 Higher Order Material Required for Examination: No Reference Question Source: New l
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QUESTION: #77(1.00)
The reactor is operating at 12% rated power with the Reactor Mode Switch in STARTUP/ HOT STANDBY. The APRMs have the following unbypassed LPRM input APRM 'A' APRM 'B' APRM 'C' APRM 'D' APRM 'E' APRM 'F'
LevelD 4 4 4 4 4 4 Level C 4 5 2 5 4 4 Level B 4 3 3 1 5 3 Level A 5 3 2 5 2 3 Average 16% 11 % 12% 11 % 13% 10% l Pwr 1 l No APRMs are bypassed for these conditions. What is the expected plant response? A full reactor scram will occur immediately due to APRM C and D INOP signals in RPS A and B respectivel j l A full reactor scram will occur immediately due to the APRM A UPSCALE and !
the APRM D INOP signals in RPS A and B respectivel i
! l l A full reactor scram will occur immediately due to the APRM A UPSCALE and l APRM C INOP signals in RPS A and the APRM D INOP signalin RPS B.
l A half scram due to the APRM A UPSCALE and APRM C INOP will be present l l in the RPS A logic.
l ANSWER: d REFERENCE: LP-16, Table 1, 2, Pages 36,17, 20; LP-22, Page 30, ELO-4 KA: 21200A3.01 [4.4/4.41 Higher Order
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Niaterial Required for Examination: No Reference
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Ouestion Source: HC requal bank
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SENIOR REACTOR OPERATOR
QUESTION: #78(1.00)
An earthquake was felt at the plant, and the Seismic Trigger SMA-3 Event Indicator (flag)
was confirmed " WHITE".
l The SNSS shall declare: No ECG level even An Unusual Even An Alert, A Site Area Emergenc ANSWER: b '
REFERENCE: LP-97, ECG Section 9. Learning Objective: R6 KA: G2.4.41 [2.3/4.1] Higher Order Material Required for Examination: ECG EALs onl Question Source: HC requal bank l
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' QUESTION: #79(1.00) i i
l Severe flooding is forecast and the Delaware River level is at 94' and is rising steadily. As , j CRS, WHICH ONE of the following actions are required?
' Monitor tide level at least once an hour and record level Place the plant in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if tide level is expected to reach 95'. Close all watertight perimeter flood doors within one hour if tide level is expected to reach 95'. Close all watertight perimeter flood doors within four hours if tide level is "
expected to reach 95'.
ANSWER: c Question Topic: severe flooding REFERENCE: AB-139, Page l i
Learning Objective: LP-114, ELO-3 KA: G.2.4.47 [3.4/3.7] memory Material Required for Examination: No Reference Question Source: New l
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SENIOR REACTOR OPERATOR I
l QUESTION: #80(1.00)
l l During a reactor startup, the following data was obtained with no control rod motion.
l l Time SRM A SRM B SRM C SRM D l
8:16:00 390 410 400 400 8:16:20 600 600 550 650 8:16:40 1200 1200 1100 1300 8:17:00 2400 2400 2300 2500 8:17:20 4500 4600 4500 4700 Determine the period and whether rod withdrawal may continu seconds; rod withdrawal may continu l .8 seconds; rod withdrawal may not continu .4 seconds; rod withdrawal may not continu seconds; rod withdrawal may continu ANSWER: REFERENCE: HC.OP-lO.ZZ-003, Section 3.1.8 and 5.2.1 Learning Objective: LP-13, LO-R10 KA: 2150004A1.02[3.6/3.7) Higher Order Material Required for Examination: No Reference Question Source: New
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l SENIOR REACTOR OPERATOR (
QUESTION: #81(1.00)
The reactor was operating at 55% power on the 90% rod pattern line with Reactor Recirculation Pump B idle and available. A controller failure causes Reactor Recirculation Pump A to runback, and core flow stabilizes at 42%. The operators shall: Place the Resctor Mode Switch in SHUTDOW Exit the region by inserting control rods or by increasing recirculation flow on the operating pum Exit the region by starting the idle recirculation pump, Within four hours, reduce the MCPR safety limit, MAPLHGR limit, and APRM scram and rod block setpoint ANSWER: b REFERENCE: AB-0300lmmediate Operator Action Learning Objective: LP-20, R6 KA: 202002K3.02[4.0/4.0] Memory Material Required for Examination: Power to Flow ma Question Source: HC requal bank.
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SENIOR REACTOR OPERATOR QUESTION: #82(1.00)
The plant is in OPCON 1. Core spray pumps AP-206 and CP-206 are operating in the full l flow test mod During the test, the position indication light switch for Core Spray Suction Valve BE-HV-F001C indication fails, causing a signal corresponding to intermediate position to be produce Choose the set of responses below, which describes the actual response (s) and/or status of )
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the system (s). CP-206 will continue running. Automatic start of CP-206 is inhibited but manual start is availabl CP-206 will trip. " CORE SPRAY LOGIC C OUT OF SVCE" annunciator will illuminate. Automatic and manual starts of CP-206 are inhibite CP-206 will continue running. Manual and automatic starts of CP-206 are not affecte CP-206 will trip. " CORE SPRAY LOGIC C OUT OF SVCE" annunciator will illuminate. CP-206 automatic start is inhibited, but manual start is availabl ANSWER: c JUSTIFICATION: There are no interlocks between the Core Spray Pumps and their suction valves. The pumps will run if the valve closes and will start on either a manual or automatic start signa REFERENCE: LP-27, Pages 16-17,26,30, L.O. R12 KA: 209002A3.01 [3.6/3.61 Higher Order Material Required for Examination: No Referenc Question Source: HC requal Bank l
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l QUESTION: #83(1.00)
The FRVS Vent System is in operation with RPV level at -40 inches. FRVS vent fan AV206 is in AUTO l..EAD (running) and vent fan BV206 is in AUTO (idle). The FRVS Unit BV206 substation normal / emergency transfer switch has been placed in the emergency positio WHICH ONE of the following set of conditions will stert the idle fan automatically? Low flow on AV206 for 30 second Low flow on AV206 for greater than 2 minute LOCA signal reset and low flow on AV206 for 30 second BV206 will not auto star ANSWER: d REFERENCE: LP-42, Pages 33-04, L.O. ELO-7, AB-115 KA: 261000K4.01 [3.7/3.8) Higher Order Material Required for Examination: No Referenc Question Source: New l
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SENIOR REACTOR OPERATOR QUESTION: 84 (1.00)
i The reactor building has isolated due to a valid 1.68 psig high
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drywell pressure conditio Subsequently drywell pressure drops to 0.7 psig. What actions must the operator take to reset the isolation, to reopen the reactor building isolation dampers and to prepare the RBVS fan units for restar Must reset PCIS isolation and then depress the open push button on panel 10C651E to reopen the dampers and reclose the non-1E breakers for the RBVS fans locally.
l Must reset PCIS isolation and then depress the open l push button on panel 10C651E to reopen the dampers and I
reclose the 1E breakers for the RBVS fans locall Must reset NSSSS isolation r,ad then depress the open push button on panel 10C651E to reopen the dampers and reclose the 1E breakers for the RBVS fans locall l Must reset NSSSS isolation and then depress the open push button on panel 10C651E to reopen the dampers and reclose the non-1E breakers for the RBVS fans locally, i l
ANSWER: REFERENCE: LP-42, pg 16-1 LO: R5, 1 KA: 290001A4.11[3.4/3.4] Higher Order MATERIAL REQUIRED FOR THE EXAM: none Question Source: New (modified HC bank)
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SENIOR REACTOR OPERATOR QUESTION: #85(1.00)
During accident conditions, which of the below describes why the post accident level indicators (PAMS) are the preferred instruments to use? ' They are electronically compensated for the higher accident temperatures in the drywell, The reference legs runs in the drywell are shorter, thus the effect of higher drywell temperatures are reduce The condensing pots are large, thus the effect of " chugging" is reduce They are electronically compensated to reduce the effect of rapid depressurization transient ANSWER: b REFERENCE: LP-OO2, Page 25, R15a KA: 259002K1.09[2.9/3 O] Memory Material Required for Examination: No Referenc Question Source: HC Requal Bank
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SENIOR REACTOR OPERATOR QUESTION: #86(1.00)
WHICH ONE of the following would be indicated by flashing of the Channel A, Amber, Containment isolation Manual Initiation, TRIPPED, light? Half trip of PCIS Channel A from Core Spray logi PCIS Channel A isolation from Level 2 logi Half trip of A inboard and outboard MSIV Isoletion of A inboard MSIV ANSWER: a JUSTlFICATION:- The lite will be on solid with a fullisolation signal also PCIS does not effect MSIV isolatio REFERENCE: LP-44, Pages 10-17, fig 5, L.O. ELO-2 KA: 223001 K1.01 [3.7/3.9) Memory Material Required for Examination: No Referenc Question Source: HC Requal Bank
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S,7NIOR REACTOR OPERATOR QUESTION: #87(1.00)
Given the following conditions:
- A LOCA has occurre * Drywell temperature is 175 degrees F.
- Suppression Chamber pressure is 9 psi Suppression chamber sprays are ra uired to be initiated at this pressure instead of drywell sprays to prevent: Exceeding the negative design pressure of the primary containmen Causing a stress failure of downcomer and vent header junction dua to cyclic condensatio Excessive accumulation of non-condensibles in the suppression chambe Drywell depressurization that exceeds the capacity of the suppression chamber to drywell vacuum breakers.
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ANSWER: b REFERENCE: LP-1268, Page 13, Step DW/P-3, L.O. ELO-R5 and R6 KA: Z2600113.03 [2.9/3.2] Memory Material Required for Examination: No Referenc Question Source: HC Requal Bank
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SENIOR REACTOR OPERATOR QUESTION: #88(1.00)
During a valid high reactor pressure condition, the Recirculation Pumps did NOT automatically trip as designe WHICH ONE t's following actions must be taken to open the Recirculation Pump Trip (RPT)
Breakers, Manually initiate both channels of the Redundant Reactivity Control System (RRCS).
b .' Verify the RPT Breakers trip when the Recirculation Pump MG Set Drive Motor Breakers are opene Direct the local tripping of the RPT Breaker Trip the RPT Breakers from the control roo I l
ANSWER: c REFERENCE: LP-19, Pages 35 and 65; TS Bases 3/4. Learning Objective: R :
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KA: 295007AA2.02[4.1/4.11 momory l i
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SENIOR REACTOR OPERATOR ,
QUESTION: #89(1.00)
Given the following conditions:
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- Reactor power is 30%.
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- The 1DG400 diesel is paralleled with the bus for surveillance testin * The 1DG400 load is currently 22OOK An operator in the control room inadvertently arms and depresses the "D" Core Spray manualinitiatioa pushbutton. WHICH ONE of the below describes the effect this error would have on the 1DG400 evolution in progress?
, The diesel generator remains in parallel operation, j l The normal bus supply breaker trips open and the diesel generator supplies i the 10A404 bus.
l The generator output breaker will trip and the diesel will shutdown and must be manually restarted.
l The generator output breaker will trip open and the diesel remains running.
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ANSWER: d REFERENCE: LP-68, Page 65, L.O. R7 l
KA: 26400K6.08 [3.6/3.7] Higher Order i
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Material Required for Examination: No Referenc Question Source: HC Requal Bank l
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QUESTION: #90 (1.00)
i The following conditions exist:
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The plant is operating at 75% powe At 0800, one Safety Relief Valve opene ,
At 0802, EOP-102 has been entered due to suppression pool water
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temperature reaching 95 degrees .
At what point should the reactor mode switch be placed in the shutdown position?
' Immediatel At 081 ! When suppression pool water temperature reaches 120 degrees When the " unsafe" region of the Heat Capacity Temperature Limit curve is entere ANSWER: a !
REFERENCE: AB-121, Step Learning Objective: LP-114, ELO-1 KA: 295013AA1.02[3.9/3.9) Memory
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SENIOR REACTOR OPERATOR QUESTION: #91 (1.00)
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WHICH ONE of the following is the reason for closing the MSIVs prior to evacuating the Control Room? To prevent uncontrolled reactor inventory loss and depressurization through the encrating turbine driven feedwater pumps, To provide positive reactor pressure and level control without reliance on automatic operation of the turbine bypass valve To quickly reduce the main turbine speed after it trips by reducing condenser vacuu To reduce radiation levels in areas that fire fighters and plant operators may require acces ANSWER: b REFERENCE: AB-130, Page Learning Objective: LP-114, ELO-3 KA: 295016K2.01 [4.4/4.51 Memory Material Required for Examination: No Referenc Question Source: NRC Bank i
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SENIOR REACTOR OPERATOR QUESTION: #92 (1.00)
The Unit has been shutdown for refueling following 100 days of power operation upon determining that several fuel assemblies have sustained cladding damage.
WHICH ONE of the following describes the effect of the damaged fuel assemblies on the p; ant during the refueling outage? Only gamma radiation streamin Only increased radiation levels in the fuel pool cooling system piping, Gamma radiation streaming and increased radiation levels in the shutdown cooling pipin Increased radiation levels in the fuel pool cooling system piping and the shutdown cooling pipin ANSWER: d REFERENCE: AB-101, Page '
Learning Objective: LP-114, ELO-3 KA: 295023AA1.02[2.9/3.11 higher order Material Required for Examination: No Referenc Question Source: NRC Bank
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EENIOR REACTOR OPERATOR QUESTION: #93 (1.00)
in EOP-101, RPV Control, if SRVs are cycling, the operator is directed to manually open SRVs until RPV pressure drops to 935 psi WHICH ONE of the following is the reason for stopping the reactor pressure reduction at 935 psig? To ensure the turbine bypass valves do not have the opportunity to stick closed, To prevent MSIVs from closing on low main steam line pressur .
. To minimize the amount of steam that is sent to the suppression poo N To prevent excessivo loss of reactor coolant inventory, i
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ANSWER: c REFERENCE: EOP-101, Step RC/P- Learning Objective: LP-124C, Page 19, L.O. ELO-8 KA: 29502 DEA 1.03 [4.4/4.4] Higher order Material Required for Examination: No Referenc Question Source: NRC Bank I
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SENIOR REACTOR OPERATOR QUESTION: 94 (1.00)
A complete loss of "A" feedwater heating train has occurred due to a system malfunctio Reector power le increasing. WHICH ONE of the following is the REQUIRED immediate operator actio Immediately reduce reactor power to its pretransient valu Immediately reduce reactor power to 25% below its pretransient valu Immediately reduce reactor power until power stabilize Immediately reduce reactor power to 20% below its pretransient valu l
ANSWER: d.
l l REFERENCE: AB-118, immediate operator action Learning Objective: LP-114, ELO- KA: 295014A1.02 [3.6/3.8) Memory Material Required for Examination: no referenc Question Source: new I
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SENIOR REACTOR OPERATOR QUESTION: #95 (1.00)
The following conditions exist:
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A failure to scram has occurre No boron has been injecte Reactor power is 30%.
- The Main Turbine is trippe The Main Condenser is availabl Suppression pool water level is norma Due to difficulty in establishing suppression pool cooling, the Heat Capacity Temperature Limit (HCTL) was exceede WHICH ONE of the following states the required operator action? Emergency depressurize using the SRV I Reduce reactor pressure at normal cooldown rate using the SRV l l
' Emergency deprcssurize using the main turbine bypass valve Reduce reactor pressure at normal :e,oldown rate using the main turbine bypass valves ANSWER: REFERENCE: EOP-102A, Steps SP/T-8 and 9 Learning Objective: LP-125A, Page 21, ELO-RS KA: 295013AK3.02[3.6/3.8) Higher Order Material Required for Examination: EOP Question Source: New
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SENIOR REACTOR OPERATOR QUESTION: #96(1.00)
WHICd ONE of the following actions allow the operator to disregard NPSH limits? After a successful reactor scram, Core Spray is being used to maintain level above -161 inche After a successful reactor screm, RCIC is being used to maintain level above
+ 12.5 inche During an ATWS, feedwater is being used to maintain level between + 1 to + 35 inche During an ATWS, following Fmergency Depressurization, condensate is being used for injectio ANSWER: a
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REFERENCE: EOP-201, Step ALC-2 l
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Learning Objective: LP-135, ELO-7.
l KA: 295031 A1.01 [4.4/4.4] Higher Order l
!- Material Required for Examination: EOP Question Source: New l
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SENIOR REACTOR OPERATOR QUESTION: #97(1.00)
The following conditions exist:
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A manual scram was inserted from 20% pow No other scram signals exis Reactor power is on intermediate range 5 and decreasin Three control rods are at position 06. All other rods are fully inserte EOP-100 has been entere WHICH ONE of the following is the required action? No other EOP entry is require Enter EOP-101, then exit EOP-101 and enter EOP-20 Exit EOP-100, and then enter EOP-101.
! Enter EOP-101, and execute concurrently with EOP-100.
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REFERENCE: EOP-100, Step S-1; EOP-101, Step RC/Q-16 Learning Objective: LP-1248, Pages 21-26, ELO-R8; LP-123, Page 8, ELO-3 KA: 295015 AK2.0113.8/3.9] Higher Order
Material Required for Examination: EOPs.
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QUESTION: #98(1.00)
While operating at 100% power, a recirculation pump sect failure causes EOP-102 entry on !
l high drywell pressure and high drywell temperature. Following initiation of suppression I
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chamber spray, drywell pressure stabilizes at 2.5 psig, suppression chamber pressure stabilizes at 2.0 psig, and drywell temperature stabilizes at 175 degrees WHICH ONE of the following actions is REQUIRED?
, Declare an Unusual Event and initiate drywell spray, Declare an Alert and initiate drywoll spray, Declare an Unusual Event. Do not initiate drywell spra i Declare an Alert. Do not initiate drywell spra i ANSWER: d i
REFERENCE: EOP-102A, Curve DWT-P-1; ECG Table 3, Section 3. Learning Objective: LP-0215-008.00B-000800,LO KA: 295010AK3.05 [3.5/3.4] higher order Material Required for Examination: EOPs and EPG's EPs.
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SENIOR REACTOR OPERATOR QUESTION: #99(1.00)
The following conditions exist:
- A successful automatic reactor scram occurred on high reactor pressur MSIVs are closed but the main condenser is availabl The operator is attempting to stabilize pressure between 900-1037 psig using SRV Re-establishing the main condenser as a heat sink: Is allowed only if no valid MSIV isolations exis Is not allowe Is required immediately after valid MSIV isolation signals are overridde l Is only allowed if the SRVs become unavailabl ANSWER: REFERENCE: EOP-101, Step / RC/P O cd 10 il Leaming Objective: LP-124C, Page 23, ELO R8 KA: 295025EA2.03[3.9/4.11 Higher Order
l Material Required for Examination: EOP Question Source: New l
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! QUESTION: #100(1.00)
The reactor was operating at 100% power when a reactor scram occurred. Only about one-third of the control rods fully inserted due to an undetected water level in the Scram Discharge Volume. Reactor power as indicated on the APRMs is 9%. Which one of the following actions, would you expect to be most effective in inserting control rods?
j Manually initiate AR De-energize scram solenoid Manually insert control rod Manually isolate and vent scram air heade ANSWER: c
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REFERENCE: EOP-101, Step RC/Q-16 Learning Objective: LP-1248, Pages 22-24, ELO R8 l KA: 295037EK2.05 [4.0/4.11 Higher Order
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l Attachment 2 SIMULATION FACILITY P.EPORT Facility Licensee: Hooe Creek Facility Docket No: 50-354 Operating Tests Administered from: Auoust 11-12,1998 l
l This form is used only to report simulator observations. These observations do not l
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constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that May be used in future evaluations. No licensee action is required in response to these observation A half core ATWS was planned for one scenario, but did not occur. After the exam it was determined that inserting the malfunction CD20 (SDV drain valves fail to close on scram)
canceled malfunction RP06 (half core ATWS). The licensee evaluated this modeling issue and found a computer code problem which was corrected.
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