IR 05000416/1998301
| ML20248A970 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/26/1998 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-416-98-301, NUDOCS 9806010119 | |
| Download: ML20248A970 (1) | |
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?f~p o AR 8.lNGTON, T E XAS 76011 8064 May 26,1998 NOTE TO: NRC Document Control Desk Mail Stop O-5-D-24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV SUBJECT: OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON MARCH 30 THROUGH APRIL 3,1998, AT GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET #50-416 On March 30 through April 3,1998, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
item #1 - a) Facility submitted outline and th'; initial exam submittal for distribution under RIDS Code A070.
b) As given operating examination, designated for distribution under RIDS l Code A070.
- Item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.
If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860-8253.
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, REGtON IV * / 611 RY AN PL AZA DRIVE. SulTE 400 L, .....J AR L INGTON, TE xAS 76011 8064 April 20, 1998 Joseph J. Hagan, Vice President Operations - Grand Gulf Nuclear Station Entergy Operations, Inc.
P.O. Box 756 Port Gibson, Mississippi 39150 SUBJECT: NRC INSPECTION REPORT 50-416/98-301
Dear Mr. Hagan:
From March 30 through April 3,1998, an NRC inspection was conducted at your Grand Gulf Nuclear Station reactor facility. The enclosed report presents the scope and results of that inspection.
The inspection included an evaluation of one applicant for an operator license and eight applicants for senior operator licenses. We determined that all applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room (PDR).
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, i John L. Pellet, Chief Operations Branch Division of Reactor Safety Docket No.: 50-416
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License No.: NPF-29
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Enclosure:
NRC Inspection Report 50-416/98-301 l ,
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Entergy Operations, Inc. -2-
REGION IV l l Docket No.: 50-416 License No.: NPF-29 l Report No.: 50-416/98-301 ; l Licensee: Entergy Operations, Inc. l Facility: Grand Gulf Nuclear Station ;
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Location: Waterloo Road Port Gibson, Mississippi
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Dates:- March 30 through April 3,1998 inspector (s): M. Murphy, Chief Examiner, Operations Branch H. Bundy, Senior Examiner, Operations Branch j t- T. Meadows, Senior Examiner, Operations Branch Approved By: John Pellet, Chief, Operations Branch Division of Reactor Safety l l i ATTACHMENTS: ,
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Attachment 1: SupplementalInformation Attachment 2: Simulation Facility Report Attachment 3: Final Written Examination and Answer Key i
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I EXECUTIVE SUMMARY Grand Gulf Nuclear Station NRC Inspection Report 50-416/98-301 NRC examiners evaluated the competency of one reactor operator and eight senior operator applicants for issuance of operating licenses at the Grand Gulf Nuclear Station. The licensee developed the initial license examinations using the guidance in NUREG-1021, interim Revision 8. January 1997. NRC examiners reviewed and approved the examinations. The initial written l examinations were administered to all nine applicants on March 27,1998, by facility proctors in accordance with the guidance in NUREG-1021, Interim Revision 8. The NRC examiners administered the operating tests March 30 through April 2,1998.
Ooerations I
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All nine (one reactor operator, eight senior operators) license applicants passed their examinations. Applicant communications and oversight of control room operations during the operating test were very strong. Peer and self checking were used extensively by the applicants. Senior operators effectively directed crew activities.
(Sections 04.1,04.2)
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The examination submitted was adequate for administration and required only limited enhancement and editorial corrections. The licensee staff was highly responsive to incorporating enhancement suggestions developed during the review process. 1 (Sections 05.1) ) l l
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-3-Reoort Details Summarv of Plant Status The plant operated at 98 percent power for the duration of this inspection.
l. Operations 04 Operator Knowledge and Performance 04.1 initial Written Examination a. Insoection Scoce On March 27,1998, the facility licensee proctored the administration of the written examination, approved by the chief examiner and NRC Region IV supervision, to one individual who had applied for an initial reactor operator license, six individuals who had applied for initial instant senior reactor operator licenses, and two individuals who had applied for initial upgrade senior operator licenses. The licensee graded the written examinations and evaluated the results for question validity and generic weaknesses.
The examiners reviewed the licensee's results.
b. Observations and Findinas The minimum passing score was 80 percent. The scores for the written examination ranged from 85 to 98 percent. The overall average score was 94.2 percent. The licensee's post-administration analysis identified that question 86 in the senior operator category was missed by six of the eight applicants. A post-examination evaluation determined that this was a combination of failing to properly read the stem information and an isolated training weakness in this area. No broad training or knowledge weaknesses were identified during review of applicant performance on the administered examinations.
c. Conclui!Om All nine license applicants passed the written examinations. No broad knowledge or training weaknesses were identified as a result of evaluation of the graded examinations.
04.2 Initial Ooeratina Test a. Insoection Scoce The examination team administered the various portions of the operating examination to the nine applicants on March 29 through April 2,1998. The reactor operator and two senior operator upgrades participated in two scenarios. The six senior operator instants' were administered three scenarios each. Eachef the instant senior operator applicants - _ - _ _ _ _ _ _ _ _
-4-and the reactor operator applicant also received a walkthrough test, which consisted of ten system tasks and four administrative areas. The two upgrade senior operator applicants were tested in five system tasks and four administrative areas.
b. Observations and Findinas All applicants passed all sections of the operating test. The examiners noted extensive use of peer and self-checking practices in all areas of the examinations. . The examiners also noted in the dynamic simulator scenarios, good oversight and effective communications that were routinely formal and three legged (i.e., request or direction, verbatim repeat back, acknowledgment of repeat back). The senior operators effectively directed crew activities. The applicants displayed good knowledge of technical specifications and facility abnormal and emergency procedures. While acting as the control board operators, the applicants displayed good knowledge of component controls and board awareness.
' The applicants performed well on the walkthrough and administrative sections of the examination.
c. Conclusions All nine applicants passed the operating tests, without significant individual weaknesses.
Communications and oversight of control room operations were very strong. Peer and self checking were used extensively. Senior operators effectively directed crew activities.
05 Operator Training and Qualification - 05.1 initial Licensina Examination Development
- The facility licensee developed the initial licensing examination in accordance with NUREG-1021, Interim Revision 8, " Operator Licensing Examination Standards for Power Reactors."
05.1.1 Examination Outlines a. Insoection Scone The facility licensee submitted the initial examination outline on January 30,1998. The chief examiner reviewed the submittal against the requirements of NUREG-1021, Interim Revision 8, _ _ _ _ _ _ _ - _ _ . _ _ _
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-5- 'b. Observations and Findinas i
The chief examiner determined that the initial examination outline met NRC requirements " and advised the licensee to proceed with enmination development.
c. Conclusions The licensee submitted an acceptable examination outline in a timely manner.
05.1.2 Examination Packaoe a. Inspection Scope The facility licensee submitted the completed draft examination package on March 2,1998. The chief examiner reviewed the submittal against the requirements of NUREG-1021, interim Revision 8. The examination material also received an independent review by another examiner against the same standards.
bl Observations and Findinas The draft written examination contained 125 questions, 75 of which were designated to' 1 be included in both reactor operator and senior reactor operator examinations, with 25 each to be used exclusively for each examination. Eighty-nine of the questions were new,16 were from the licensee's question bank, and 20 were from the bank,- but met the criteria as modified questions. The draft examination was considered technically valid, to discriminate at the proper level, and responsive to the sample plan submitted by the
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licensee on January 30,1998. Following two independent NRC examiner reviews, the chief examiner provided enhancement suggestions for about 16 percent of the questions and identified editorial corrections for an additional 8 percent. The suggested enhancements generally related to clarity of the question stem and distractor plausibility.
After discussion of the suggested enhancements, the licensee modified the examinations as agreed. The chief examiner concurred with the resolution of the comments and the final product.
The licensee submitted six scenarios, two of which were designated as backups. The six scenarios were reviewed and validated during the week of March 16,1998, with only minor enhancement and editorial comments to facilitate administration.
To support the system walkthrough section of the operating test, the facility licensee provided job performance measures developed to evaluate selecteri operator tasks that contained written task elements, performance standards, and comprehensive evaluator cues? Twenty-three job performance measures were submitted with two prescripted followup questions each. This provided two sets of ten job performance measures for the reactor operator and senior operator instants and one set of five for the senior operator upgrades with three backup tests. Personnel assignments and scheduling
. precluded any day-to-day repetition of operating tests. Two independent NRC examiner - ___-_ _
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reviews resulted in two enhancement and four editorial comments for the job l performance measures. The same reviews also produced three enhancement and three editorial comments for the associated prescripted questions. The licensee incorporated all comments. One job performance measure was replaced following validation when it was identified that it provided limited evaluation of the applicant. . The licensee submitted 13 administrative job performance measures and , 2 administrative topic questions. This provided two sets of 5 administrative job performance measures for the senior reactor operator applicants and one set of 4 administrative job performance measures with 2 administrative topic question for the reactor operator applicant. The two independent NRC reviews produced two enhancement and six editorial comments for the administrative topics section of the examination. The licensee incorporated all comments.
c. Conclusions The examination submitted was adequate for administration and required only limited enhancement and editorial corrections. The licensee staff was highly responsive to incorporating enhancement suggestions developed during the review process.
05.2 Simulation Facility Performance a. Insoection Scoce The examiners observed simulator performance with regard to fidelity during the examination validation and administration, b. Observations and Findings The simulation facility supported the validation and administration of the examination well, except, as described in Attachment 2, for one component failure that interrupted one scenario. The problem was identified and fixed by the simulator technical slipport personnel with minimal delay and no compromise in the examination evaluations. No fidelity problems were noted.
c. Conclusions The simulator and simulator staff supported the examinations well. No fidelity issues were identified.
05.3 Examination Security a. Eggp.g The examiners reviewed examination security both during on site preparation week and examination administration week for compliance with NUREG-1021 requirements.
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-7-b. Observations and Findinos Twenty-one members of the licensee's operations and training staff signed onto the NUREG-1021 approved examination security agreement acknowledging their responsibilities for examination security. The licensee maintained secure areas for examination review, validation, and reproduction. Signs were conspicuously posted to avoid inadvertent unauthorized access, and doors were maintained locked with good key control to ensure proper access to sensitive areas. The licensee installed deadbolts with no key access on all simulator access doors for positive traffic control. Applicants were maintained in one controlled area under constant supervision, and were always escorted to and from examination points. Simulator security was strictly complied with.
c. Conclusions Effective examination security was maintained.
V. Management Meetings X1 Exit Meeting Summary The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on April 3,1998. The licensee acknowledged the findings presented.
The licensee did not identify as proprietary any information or materials examined during the inspection.
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ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee B. Bryant, Operations Training Supervisor D. Cupstid, Technical Assistant, Operations L. Daughtery, Licensing Technical Coordinator
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W. Eaton, General Manager J. Hagan, Vice President D. Janacek, Director, Training M. McDowell, Technical Assistant, Training ; M. Rasch, Senior Operations Instructor S. Reeves, Senior Operations instructor J. Roberts, Director, Quality NB.C K. Weaver, Resident inspector INSPECTION PROCEDURES USED NUREG-1021 " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8
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ATTACHMENT 2 SIMULATION FACILITY REPORT Facility Licensee- Entergy Operations, Inc.
Facility Docket: 50-416 Operating Examinations Administered at: Grand Gulf Nuclear Station Operating Examinations Administered on: March 30 through April 2,1998
. These observations do not constitute audit or inspection findings and are not, without further
, verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations l do not affect NRC certification or approval of the simulation facility, other than to provide information, which may be used in future evaluations. No license action is required in response
- to these observations.
l Deficiencies identified Durina Examination Preparation ,
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None Deficiency identified Durina Examination Administration
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During the conduct of the first scenario on Thursday, April 2,1998, just as the crew had completed placing a reactor feed pump on line and were preparing for rod withdrawals to i raise power, a component failure in the simulator caused the simulated loss of two - buses, LCC 13801 and 13BD2, causing complete loss of instrument air and condenser vacuum. The crew responded to this event and since all other elements of the scenario appeared to be in order, the chief examiner elected to allow the new sequence to continue and evaluate the crew on their response to the new conditions. This worked until interference was detected in components of the RCIC system that would not allow continued valid simulation. The scenario was then terminated.
The simulator technical support personnel promptly responded, determined the cause i and replaced the failed component. After a short checkout run the simulator was , declared operable and those required missed portions of the scenario were run without compromise to the applicant evaluations.
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_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ . _ _ . _ ATTACHMENT 3 ' FINAL WRITTEN EXAMINATION AND ANSWER KEY _ _ _ - _ _ _ _ _ .___- -
, i FOR INSTRUCTOR USE ONLV FOR INSTRUCTOR USE OMLY I ~ 100 r90 5 -80 - 70' '60, pgg7 j .100 $0 - 80: 17 0 60" PART 1 E l
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_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ l ES-401 Site-Specific Written Examination Form ES-401-7 ! Cover Sheet I l U.S. Nuclear Regulatory Commission l Site-Specific l Written Examination Applicant Information l Name: Region: I / II / III /OV) Date: 27 March 1918 Facility / Unit: Grand Gulf 1 License Level: % / SRO Reactor Type: W/ CE / BW / GE Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.
Examination papers will be collected four hours after the examination starts. ! l i Applicant certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent I NUREG- 1021 i !
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 1 The following conditions are observed after a Loss of Coolant Accident: Reactor Pressure 90 psig 166' elev. temperature in the Dipvell 280*F Drywell Pressure 5.8 psig j 139' elev. temperature in the Containment 192 * F 119' elev. temperature in the Containment 181 *F Containment Pressure 2.0 psig i Shutdown Range LevelIndication + 20 inches J Upset Range LevelIndication + 50 inches Wide Rangei.evelIndication - 40 inches Which one of the following indicates actual level? A. Shutdown Range B. Upset Range C. Wide Range I D. Level Cannot be determined. I i QUESTION l NRC RECORD # WRI 1 ANSWER: C. SYSTEM # B21 K/A 295028 EK2.03: 3.6/3.8 EKl.01: 3.5/3.7 LP# GG-1-LP-RO-EP02.00 K/A 295027 EK1.02: 3.0/3.2 ; OBJ. 18 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-01-EP-2 Caution 1 NEW CLASS Modified Bank DIFF 3 annual exam ep-03 , DATE USED: RO SRO ROTH CFR 41.3/43.5 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 2 The plant conditions are as follows: Reactor Pressure 900 psig Reactor Water Level - 100 inches Drywell Pressure 1.10 psig LPCS Injection Line Pressure 450 psig Which of the following describes how the LPCS injection valve E21-F005 would respond ifits handswitch is taken to the OPEN position? A. The valve will not open.
B. The valve will not open until reactor level or drywell pressure has reached the LPCS System initiation setpoint.
C. The valve will open and remain open.
D. The valve will open, but will automatically close when it reaches its full cpen position.
QUESTION l NRC RECORD # WRI 2 ANSWER: C SYSTEM # E21 K/A 209001 K4.08: 3.8/4.0 LP# GG-1-LP-RO-E2100.00 K3.-: 3.5-3.9/3.5-3.9 OBJ. 8b,16 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-E21-1 sect. 3.11 NEW CLASS sect. 3.12 MODIFIED BANK l DIFF 3 annual exam l e21-05 l DATE USED: RO SRO BOTH CFR 41.7/41.8 3/17/98 t-----------------------_-----
U. S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR ,
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QUESTION 3 i The Plant is operating at 100 % power. l The Motor men Fire pump is out of service.
A fire in Transformer ESF 12 has initiated the Deluge system for the transformer.
The A Diesel Driven Fire Pump received a signal to start.
Which one of the following describes the starting limitations of the Diesel Driven Fire Pump? A. The diesel engine will attempt to start for 15 minutes. If the diesel does not start it l alarms in the Control Room, and it must be reset from the Control Room before it will attempt to start again.
B. The diesel engine will attempt to start for 15 seconds, then wait for 15 seconds. It will attempt this start sequence for 6 attempts. After that, it must manually be reset before any further start attempts occur.
C. The diesel engine will attempt to start for 15 seconds then wait for 15 minutes to
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allow the battery to recharge, then it will attempt this cycle again. After that it must manually be reset before any further start attempts occur.
D. The diesel engine will attempt to start as long as air pressure is > 60 psig. After that, the air bank must recharge before additional stm a uitempts can occur.
I QUESTION l NRC RECORD # WRI 3 ANSWER: B. SYSTEM # P64 K/A 286000 A2.08: 3.2/3.3 K5.05: 3.0/3.1 K4.07: 3.3/3.3 ; A3.01: 3.4/3.4 LP# OP-LO-SYS-LP-P64-05 A4.06: 3.4/3.4 OBJ. 5d SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 ; REFERENCE: ARI 04-S-02-SH13-P862 NEW CLASS ! 1A-B3; 1A-B5 MODIFIED BANK DIFF 3 sect.1.2; 2.1; & 4.5 ' DATE USED: RO SRO ROTH CFR 41.4 l l 3/17/98 l i __ _ _ - - _ _ - - . - .____.m _ _ . _ _ _ _ _ . . _ _ _
-_ _-_ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . __. _ _ _ _ _ _ _ _ _ _. ._ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR I QUESTION 4 . I High Drywell pressure of 2.2 psig has resulted in ECCS initiation. HPCS has injected to the reactor at rated pressure causing reactor level to increase to + 60 inches. The HPCS Injection Valve (E22-F004) has automatically closed. No operator action was taken.
Reactor water level has subsequently decreased to - 50 inches.
Which one of the following actions will be required to have HPCS inject to the reactor to restore waterlevel? A. Place the E22-F004 handswitch to CLOSE to reset the logic, and then to OPEN.
B. No further action is required. E22-F004 will auto reopen at - 41.6 inches.
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C. HPCS ManualInitiation pushbutton must be Armed and Depressed to reset l E22-F004 and allow it to open.
l D. Reset the HPCS Initiation Logic and allow a subsequent Initiation signal to occur. j l
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l QUESTION l NRC RECORD # WRI 4 ANSWER: B. SYSTEM # E22-1 K/A 209002 A3.01: 3.3/3.3 LP# GG-1-LP-RO-E2201.00 A2.10: 2.7/3.0 OBJ. 8b,16 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-E22-1 sect. 3.8 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7 l l l l 3/17/98 <
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U. S. NUCLEAR REGULATORY COMMISSION ) WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 5 The plant is in Mode 4 with Shutdown Cooling RHR A in service. RHR A was also aligned to blowdown reactor inventory to the Suppression Pool via the RHR A heat exchanger vent valves (E12-F073 A and E12-F074A) due to an outage on RWCU. After opening the heat exchanger vent valves (E12-F073A and E12-F074A) a significant amount, the operator assigned to monitor vessel level becomes distracted.
I Which one of the following best describes the response of the RHR A System to a lowering water level? , i (Assume no further operator actions) A. At - 41.6 inches, the RHR A Suction from the Reactor (E12-F008 & F009), and RHR A Shutdown Cooling R.eturn to Feedwater (E12-F053A) will isolate. This will cause the RHR A pump to trip.
B. At + 11.4 inches, RHR A Heat Exchanger Vents (E12-F073 A & F074A), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to go on minimum flow. l C. At - 41.6 inches, RHR A Heat Exchanger Vents (E12-F073A & F074A), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to go on minimum flow.
I D. At + 11.4 inches, the RHR Suction from the Reactor (E12-F008 & F009), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to trip.
QUESTION l NRC RECORD # WRI 5 ANSWER: D. SYSTEM # E12 K/A 205000 K4.03: 3.8/3.8
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LP# OP-LOR-ONEP-LP-001-04 A2.05: 3.5/3.7 OBJ. 31 A2.09: 3.6/3.8 ' LP# OP-LO-SYS-LP-E12-07 OBJ. 11,12 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 05-1-02-IH-5 Group 3 NEW CLASS ! 04-1-01-E12-1 MODIFIED BANK DIFF 2 sect. 4.2.2e(14) DATE USED: RO SRO ROT /[ CFR 41.7 i 3/17/98 _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - _. _ J
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 6 WIIICH ONE (1) of the following statements REOUIRE an immediate scram to be inserted by the operator-at-controls? l l A. The reactor is operating at 100% power when both mnning CCW Pumps trip, neither pump can be immediately restarted and the standby pump will not start.
B. The reactor is operating at 50% power, when a seismic event causes the "A" Reactor Recirculation Pump to trip. ; C. A reactor startup is in progress with reactor pressure at 400 psig when the mnning CkD Pump trips and one (1) scram accumulator is declared inop for a fully inserted control rod. ' D. The reactor is operating at 75% power when one (1) of the operating Reactor l Feed Pumps trips and level drops in the reactor to 25 inches QUESTION l NRC RECORD # WRI 6 ANSWER: A. SYSTEM # P42 K/A 295018 AA1.02: 3.3/3.4 LP# OP-LOR-ONEP-LP-001-
OBJ. 1 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-V-1 sect. NEW CLASS 01-S-06-2 sect. 6.3.6 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 I 3/17/98
I U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 7 Following a Recire line rupture, reactor level has decreased to - 80 inches.
Both trains of the Standby Gas Treatment System have initiated.
Which of the following best describes the operation of the Standby Gas Treatment System flow control dampers? A. When -0.2 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 90 seconds later the remaining flow control dampers throttle to their intermediate position.
B. When -0.25 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position.120 seconds later the remaining flow control dampers throttle to their intermediate position.
C. After 90 seconds, the flow control dampers will go to their intermediate positions to maintain - 0.75 inches water column in the Auxiliary Building and -0.25 inches I water column in the Enclosure Building.
D. After 90 seconds the flow control dampers throttle to maintain -0.25 inches water column. If the Enclosure Building pressure reaches -0.75 inches water column, the flow control dampers go to their intermediate positions.
QUESTION l NRC RECORD # WRI 7 ANSWER: A. SYSTEM tl T48 K/A 261000 A1.04: 3.0/3.3 LP# CG-1-LP-RO-T4801.00 OBJ. 7a, 8,15 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E- 1257- 08,11, 23 NEW CLASS MODIFIED BANK ; DIFF 4 annual exam wk 6 l DATE USED: RO SRO ROTH CFR 41.13 I i 3/17/98 l
U. S. NUCLEAR REGULATORY COMMISSION j WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR QUESTION 8 The plant is in a LOCA with ECCS systems injecting to the reactor.
Suppression Pool level has lowered to 13.5 feet.
Which one of the following is a condition that exists due to this level? ! A. The SRV tailpipe exhausts have been uncovered.
B. The RCIC Turbine Exhaust has been uncovered.
C. Suppression Pool temperature cannot be determined.
I D. Containment Pressure cannot be determined i l QUESTION l NRC RECORD # WRI 8 ANSWER: C. SYSTEM # E30 K/A 295030 EA2.02: 3.9/3.9 ! LP# OP-LO-EP-LP-005-03 OBJ. 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-EP-3 Caution 2 NEW CLASS MODIFIED BANK DIFF 2 annual exam SRO Rem 1 DATE USED: 110 SRO BOTH CFR 41.9
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 i REACTOR OPERATOR QUESTION 9 The plant is in a LOCA with ECCS systems injecting to the reactor.
Containment pressure has increased to 21.1 psig.
The Emergency Procedures have direction to vent the Containment irrespective of offsite release rates.
What is the bases for this action? A. The Primary Containment is not accessible to personnel to attempt mitigation of the event until pressure is decreased to less than 20 psig.
B. Venting the Primary Containment is done to prevent an uncontrolled release due to a breach of the Primary Containment.
C. If pressure in the Primary Containment exceeds 22 psig, the Suppression Pool Level instrumentation will fail causing an uncontrolled release of the Suppression Poolinto the Auxiliary Building.
D. Venting of Containment at this levelis postulated to not exceed the ALERT release limits of the Emergency Plan.
QUESTION l NRC RECORD # WRI 9 i ANSWER: B. SYSTEM # M41 K/A 295024 EK3.03: 3.6/4.1 LP# OP-LO-EPB-HN-003-02 LP# OP-LO-EP-LP-005-03 OBJ. 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-1-01-EP-3 NEW CLASS l MODIFIED BANK ' DIFF 3 DATE USED: RO SRO BOTH CFR 41.9 ! 3/17/98 L-----___--_-------------------------- - - - - - - -
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 10 The reactor was operating at the end of cyclejust prior to a refueling outage when a reactor scram occurred.
Which one of the following is a correct method of verifying the position of the control j rods? (The scram has NOT been reset.) j A. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw . Data and observe a blank display with only green LEDs for all control rods. I B. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe all control rods indicate 00 with a green LED for all control rods.
C. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe a blank display with only red LEDs for all control rods.
D. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe all control rods indicate 00 with a red ' LED for all control rods.
QUESTION l NRC RECORD # WRI 10 ANSWER: A. SYSTEM # C11-2; K/A 295006 AA2.02: 4.3/4.4 C11-1B 201005 A3.02: 3.5/3.5 LP# GG-1-LP-RO-C111B.00 A4.02: 3.7/3.7 OBJ. 3c LP# GG-1-LP-RO-C1102.02 OBJ. 10,11,22 l
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LP# OP-LO-ONEP-LP-001-04 OBJ. 1 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 ! REFERENCE: 04-1-01-C11-2 NEW CLASS sect. 4.7.2p & 4.8.2i MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR l 41.6/41.10/43.5 I 3/17/98 l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR l QUESTION 11 The Electrical line up is normal. A LOCA condition has caused Drywell Pressure to increase to 1.6 psig.
A switching error causes 500 KV voltage to decrease.
The voltage to ALL ESF busses DECREASES to 3290 volts.
\ The voltage transient duration is 10 seconds and then voltage returns to normal.
f Which ONE of the following statements is the condition of the ESF busses after this voltage transient? A. 15AAis being supplied from ESF 11 16AB is being supplied from ESF 21 17AC is being supplied from ESF 21 B. 15AAis being supplied from DivID/G 16AB is being supplied from DivIID/G 17AC is being supplied from Div III D/G C. 15AAis being supplied fromESF 11 16AB is being supplied from ESF 21 17AC is being supplied from Div III D/G D. ISAAis being supplied fromDivID/G 16AB is being supplied from DivIID/G 17ACis being supplied from ESF 21 QUESTION l NRC RECORD # WRI 11 ANSWER: B. SYSTEM # R21; K/A 262001 A2.11: 3.2/3.6 P75; P81 A3.01: 3.1/3.2 LP# GG-1-LP-RO-R2100.01 264000 K4.08: 3.8/3.7 OBJ. 12,14,20,22,28 LP# GG-1-LP-RO-P7500.00 OBJ. 8,15 LP# GG-1-LP-RO-P8100.01 OBJ. 8,15 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: 04-1-01-R21-1 sect. 5.1 NEW CLASS 04-1-01-P81-1 sect. 3.22 MODIFIED BANK DIFF 3 LOT 7/95 C 13 , DATE USED: RO SRO BOTH CFR 41.4
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U. S. NUCLEAR REGULATORY COMMISSION
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WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR j QUESTION 12
The LSS panels will perform which one of the following functions: A. Starting of all ESF Loads only when there has been a total loss of offsite power when i the Diesels are canying the buses.
B. Sequencing of ESF loads to ensure ESF Bus voltage and frequency are not degraded and to minimize stress on the diesel.
C. Starting of all loads on Division I, II, and III required following a DBA LOCA.
D. Sequences HPCS, LPCS and RHR to ensure the diesel engines are not damaged on a concurrent LOCA and loss of offsite power. i I QUESTION l NRC RECORD # WRI 12 ANSWER: B. SYSTEM # R21; K/A 264000 K5.06: 3.4/3.5 P75 LP# GG-1-LP-RO-R2100.01 264000 K4.08: 3.8/3.7 OBJ. 3a SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: Tech Spec Bases B3.8.1 NEW CLASS MODIFIED BANK DIFF 2 LOT 7/95 c13 DATE USED: RO SRO BOTH CFR 41.8
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_ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - U. S. NUCLEAR REGULATORY COMMISSION WRITTENEXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 13 The Auxiliary Building Isolation signal was received.
P53-F026.A (Instmment Air Aux. Bldg Isolation Valve) has a flashing red indication on the Isolation Status Panel.
Which one of the following best describes the reason for this indication? A. The valve has received an isolation signal and closed within the specified time B. The valve has received an isolation signal and has not repositioned within the specified time.
C. The valve may be reopened after the adjustable timer circuit has timed out.
D. The valve has been overridden open with an isolation signal present QUESTION _ _ l NRC RECORD # WRI 13 j ANSWER: B. SYSTEM # M72 K/A 223002 A4.04: 3.5/3.6
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A1.01: 3.5/3.5 LP# GG-1-LP-RO- A3.01: 3.4/3.4 M7200.00 OBJ. 5 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENC FSAR sect. 7.5 NEW CLASS l E: MODIFIED BANK , DIFF 2 LOT 7/95 c13 l DATE RO SRO SOTH CFR 41.8 l USED: 3/17/98 L ___ _ __ _ __ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 14 The plant is at 100 % power with I & C APRM A Surveillance in progress when the following indicators are illuminated on the H13-P680 panel.
Pushbutton HCU FAULT Pushbutton ROD DRIFT Pushbutton SCRAM VLVS Pushbutton ACKNHCUFAULT Annunciator"HCU TROUBLE" Annunciator"ControlRod Ddft" Which of the following could be a possible cause of these indications? ASSUME ALL OTHERINDICATIONS ARE NORMAL.
A. A control rod drifling out of the core.
B. A full reactor scram C. A single controlrod scram.
D. The use of the IN TIMER SKIP pushbutton rather than the INSERT pushbutton.
QUESTION l NRC RECORD # WRI 14 ANSWER: C. SYSTEM # C11-2 K/A 201005 A4.01: 3.7/3.7 LP# GG-1-LP-RO-C1102.02 OBJ. 10,11,22 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-C11-2 NEW CLASS sect. 4.7.2 & 4.8.2 MODIFIED BANK DIFF 3 04-1-02-H13-P680 LOT 2/98 rxinmu DATE USED: 4A2-D4 & 4A2-E4 RO SRO ROTII CFR 41.6 l 3/17/98 ! L- _
_ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ ___ _ . -- U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 , REACTOR OPERATOR QUESTION 15 GGNS is operating at 10% rated power with the mode switch in the STARTUP position, and total core flow at 53%. APRM E and H are bypassed due to failed power supplies.
The following is the present status of the APRMs vusus LPRM inputs and indicated power: APRM A B C D E F G H LPRM LVL D 5 5 2 2 3 2 4 5 LPRM LVL C 5 4 3 5 4 4 3 4 l LPRM LVL B 3 2 5 4 4 3 3 3 LPRM LVL A 2 4 4 4 4 4 5 3 INDICATED 10 % 13 % 12 % 14 % 0% 11 % 13 % 0% POWER byp byp , LPRM 26-43D has failed downscale and must be bypassed to allow troubleshooting.
With present conditions would this action be allowed? Attached is the LPRM vs APRM assignments table.
A. Yes, conditions are satisfactory.
B. Yes, however an LCO would have to be written on the associated APRM for Administrativeinputs.
C. No, this action would result in a half scram and administrative LCO requirements not to be met.
D. No, this action would result in a full reactor scram.
QUESTION l NRC RECORD # WRI 15 ANSWER: D. SYSTEM # C51-2 K/A 215000 K4.02: 4.1/4.2 LP# GG-1-LP-RO-C510400 A1.04: 3.6/3.6; A4.01: 3.7/3.7 OBJ. 4,9,los SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-C51-1 sect. 3.3 NEW CLASS 17-S-02-40 seet. 6.3 MODIFIED BANK l DIFF 3 Tech Spec Bases B3.3.1.1 l DATE USED: RO SRO BOTH CFR 41.6 l 3/17/98 - _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ _ _ _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 16 The plant is in a refueling outage with the refueling platform located over the Dryer Storage Area.
Which one of the following WILL PREVENT movement of the refueling platform over the reactor vessel core? A. The Main Hoist unloaded, control rod 28-37 is selected at position 00 on H13-P680, and the Reactor Mode Switch in REFUEL.
B. The Main Holst unloaded, one control rod at position 48, and the Reactor Mode Switch in REFUEL.
C. The Main Hoist unloaded, all control rods insened, and the Reactor Mode Switch in REF'UEL.
D. The Main Holst unloaded, control rod 28-37 selected in gang mode at H13-P680, and the Reactor Mode Switch in REFUEL.
QUESTION l NRC RECORD # WRI 16 ANSWER: D. SYSTEM # Fil K/A 234000 K6 03: 3.0/3.6 LP# GG-1-LP-RF-F1101.00 A3.02: 3.1/3.7 OB lla,c,28,36 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 3 J.
REFERENCE: 04-1-01-F11-1 Att. V NEW CLASS MODIFIED BANK DIFF 3 LOT 2/98 rxsys DATE USED: RO SRO ROTH CFR 41.4/43.7
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(_ __ _ _ _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 17 The plant has sorui id due to high reactor water level. The water level peaked at + 58 inches. The plant is now stable at 950 psig and +25 inches. All systems functioned properly , . following the scram.
Which one of the following is the correct status of Scram pilot solenoids. Backup scram solenoids, and ARI solenoids? ! SCRAM PILOT BACKUP SCRAM ARI SOLENOIDS SOLENOIDS SOLENOIDS ' A. De-energized Energized Energized L B. De-energized Energized De-erergized C. Energized Energized Energized l- D. Energized De-energized De-energized I
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QUESTION l NRC RECORD # WRI 17 ANSWER: B. SYSTEM # C71; K/A 212000 K1.06: 3.5/3.ti C11-1 A A2.19: 3.8/3.9 LP# GG-1-LP-RO-C7100.00 A2.20: 4.1/4.2 OBJ 13d,18 A4.12: 3.9/3.9 LP# GG-1-LP-RO-C111A.00 201001 K1.07: 3.4/3.4 OBJ 9f, 9g, 9h SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E- 1173 - 15-21 NEW CLASS E- 6066 - 03,06 MODIFIED BANK DIFF 3 LOT 2/98 rxsys.
DATE USED: RO SRO BOTH CFR 41.6 ! , ! i
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_ _ _ _ _ _ _ _ _____ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 18 The plant is operating at 60% rated power.
Both Recire Flow Control Valves are at 25% valve position.
A leak in the Drywell caused Drywell pressure to increase to approximately 1.5 psig.
Following the high Drywell pressure signal, the 'B' Reactor Feed Pump trips and level decreases to +14.2 inches before stabilizing at normallevel.
Which of the following statements best describes the response of the Recirc System? A. Flow Control Valves will runback to 15% valve position; Recirc Pumps in Fast Speed.
B. Flow Control Valves will remain at 25% valve position; Recirc Pumps in Slow Speed.
C. Flow Controls Valves will mnback to 15 % valve position; Recirc Pumps in Slow Speed.
D. Flow Control Valves will remain at 25% valve position; Recirc Pumps in Fast Speed.
QUESTION l NRC RECORD # WRI 18 ANSWER: B. SYSTEM # B33 K/A 202002 A2.06: 3.3/3.3 LP# GG-1-LP-RO-B3300.00 OBJ 19-22,51 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-B33-1 sect. 4.2 NEW CLASS ARI 04-1-02-II13-P680 MODIFIED BANK DIFF 4 3A-B7; 3A-B8 LOT 2/98 rxsys DATE USED: RO SRO ROTH CFR 41.6 l , 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 19 A LOCA has occurred.
Drywell pressure is 1.84 psig.
Reactor water level is +36" and stable.
High Pressure Core Spray Pump has been overridden to STOP.
Division III bus 17AC loses power and is subsequently reenergized by the Diesel Generator.
Which one of the following describes the condition of the HPCS7 A. Wd' l reset the overrides and HPCS will re-initiate.
B. Will reset the initiation logic and HPCS will remain secured.
C. Will NOT affect the initiation logic and HPCS will re-initiate.
D. Will NOT affect the initiation logic and HPCS will remain overridden.
QUESTION l NRC RECORD # WRI 19 ANSWER: D. SYSTEM # E22-1 K/A 209002 K2.03: 2.9/2.8 LP# GG 1-LP-RO-E2201.00 K2.01: 3.3/3.2 OBJ 9,11,13,16 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-1183- 023 NEW CLASS E-1188- 019 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.7/41.8 l l l l 3/17/98 a__________________-_____________. _
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i U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 20 f Which of the following situations would result in an automatic actuation of the Div. II SPMU valves? A. High drywellpressure of1.39 psig Suppression pool level of 17 feet 8 inches.
15 minutes since 1.39 psig drywell pressure ! B. Reactor waterlevel of-41.6 inches ! Suppression poollevel of18 feet.
! 26 minutes since -41.6 inches ! C. High drywellpressure of1.23 psig Suppression pool level of 17 feet 5 inches.
26 minutes since 1.23 psig drywell pressure D. Reactor waterlevel of-150.3" Suppression pool water level of 17 feet 4 inches.
15 minutes since -150.3 inches QUESTION l NRC RECORD # WRI 20 ANSWER: D. SYSTEM # E30 K/A 223001 A2.11: 3.6/3.8 LP# GG-1-LP-RO-E3000.00 A3.01: 3.4/3.5 OBJ 7 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARI 04-1-02-II13-P870 NEW CLASS 4A-A3; 4A-F1 MODIFIED BANK DIFF 3 LOT 2/98 nsys DATE USED: RO SRO BOTH CFR 41.7 , l ! 3/17/93 ___ ___ _______-__________ -
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- QUESTION 21 An ATWS is in progress with the MSIV's closed, and reactor water level is being controlled by . RCIC at the top of r.ctive fuel. Standby Liquid Control is injecting and all emergency procedure curves are in the SAFE region. Suppression Pool level is 17.8 feet and lowering.
Suppression Pool temperature is 125 'F and going up. RHR A and B are unavailable for SuppressionPoolCooling.
Which one of the following systems may be used to reduce the temperature of the Suppression Pool? A.- FuelPool Cooling and Cleanup B. ResidualHeat RemovalC C. Suppression PoolMake up D. Alternate DecayHeat Renoval QUESTION l NRC RECORD # WRI 21 ANSWER: C. SYSTEM # E12; K/A 295013 AA1.01: 3.9/3.9 E30; M41 LP# OP-LO-EP-LP-005-03 OBJ 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-EP-3 NEW CLASS STEPS 1 & 48 MODIFIED BANK
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DIFF 3 DATE USED: RO SRO BOTH CFR 41.9 l l l l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 22 A Loss of Coolant Accident is in progress. Reactor Level is being restored by HPCS and RCIC. Containment parameters are elevated.
Determine which of the following situations would result in Containment Spray initiating automatically.
A. 5 minutes since LOCA RHR A and B pumps were manually overridden to STOP 4 minutes aller LOCA Drywell pressure at 2 psig CTMT pressure at 9 psig B. 5 minutes since LOCA RHR A and B in LPCI mode on minimum flow.
Drywell pressureis at 2 psig CTMT pressure at 7 psig C. 15 minutes since LOCA RHR A and B pumps were manually overridden to STOP 4 minutes after LOCA Drywellpressureis 1.5 psig CTMT pressure at 8 psig D. 15 minutes since LOCA RHR A and B in suppression pool cooling Drywell pressure at 1.0 psig CTMT pressure at 2.2 psig QUESTION l NRC RECORD # WRI 22 ANSWER: C. SYSTEM # E12 K/A 226001 K4.09: 3.2/3.4 LP# OP-LO-SYS-LP-E12-07 OBJ 10 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS 17A-F3; 20A-B6 Af0DIFIED BANK ! DIFF 2 LOT 2/98 eccs , DATE USED: RO SRO BOTH CFR 41.9 l 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 23 A small break LOCA has occurred in the drywell. High Pressure Core Spray is not available due to maintenance on the Upper Motor Bearing. RCIC is injecting but level continues to decrease. Feedwater is the apparent source of the leak and has since been isolated. Levelis continuing to go down. LPCI A, B, & C and LPCS are operating on minimum flow.
Which one of the following choices best describes the operation ofADS? A. Once the 105 ac. timer has started, you MUST wait for the timer to time out and level to decrease to the top of active fuel and a high drywell pressure signal to be present to automaticallyinitiate ADS.
B. If drywell pressure increases to 1.39 psig a 9.2 minute timer will start and automatically initiate ADS with NO low level signal.
C. Completion of the ADS 9.2 minute timer will statt the 105 second timer iflevel remains below -150.3 inches with NO high drywell pressure signal, causing ADS to automaticallyinitiate.
D. Ifreactor level drops below +11.4 inches the ADS 9.2 minute timer starts. Iflevel drops to -150.3 inches, when the 105 second timer times out, ADS will automatically initiate.
QUESTION l NRC RECORD # WRI 23 ANSWER: C. SYSTEM # E22-2 K/A 218000 K5.01: 3.8/3.8 K4.02: 3.8/4.0 LP# GC-1-LP-RO-E2202-00 K4.03: 3.8/4.0 OBJ 10,21 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS 18A-A1; 18A-A2; 18A-C2 MODIFIED BANK DIFF 3 E - 1161-005 DATE USED: RO SRO ROTH CFR 41.8 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 24 The B21-F051B SRV handswitches at Division I and Division II Remote Shutdown panels are , in the "OFF" position.
Select the statement below that hgg describes the response of the Safety Relief Valve B21- I F051B while the switches are in this pcsition.
l > A. As reactor pressure increases foHowing a Group I Isolation, the valve will open when i ! Low-Low Setis initiated.
! B. All modes of operation of the SRV are inoperable except the Safety mode which will still open the valve.
f C. When the operator at the P601 panel takes the control switch for the valve to the open l position,it will open.
D. All modes of ope ation of the SRV are inoperable excyt the Safety and ADS modes I ofwhich either will still open the valve.
i QUESTION l NRC RECORD # WRI 24 ANSWER: B. SYSTEM # E22-2 K/A 239002 K4.05: 3.6/3.7 LP# GG-1-LP-RO-E2202-00 OBJ 9c SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 i REFERENCE: E-1161-11,14 NEW CLASS' MODIFIED BANK DIFF 3 LOT 2/98 eccs DATE USED: RO SRO BOTH CFR 41.7 ' . i l l 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 25 The plant has experienced a loss ofInstmment air due to a mpture in the common header piping between the Water Treatment Building and the Turbine Building. This has resulted in a complete loss of air. Maintenance estimates it will take 12 to 15 hours to repair. The plant is Shutdown following a 422 day mn. Air pressure to the ADS system is at 80 psig and lowedng.
Which one of the following is a method of restoring air pressure to the ADS Valves for Reactor Pressure Control? A. Cross tie the Instrument Air and Service Air Headers in the Auxiliary Building via hose fittings and chicago fittings.
B. Connect a diesel driven air compressor to the Instmment Air Header Drain Line in area 9, 139 fl. elevation of the Auxiliary Building.
C. Connect Nitrogen bottles in area 9, 139 fl. elevation to the Instmment Air Connection and isolate the Instrument Air Header fiom the ADS air header.
D. Enter Containment and connect Nitrogen bottles to the Instmment Air Header drain to the ADS air header.
QUESTION l NRC RECORD # WRI 25 ANSWER: C. SYSTEM # E22-2; K/A 295019 AA1.01: 3.5/3.3 PS3 LP# GG-1-LP-RO-E2202-00 OBJ 18d l LP# GG-1-LP-RO-EP07-00 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-V-9 sect. 3.9 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 26 The plant was operating at full power when an error while performing a surveillance resulted in a RecircFlow Controlmnback.
I Reactor Poweris presently 79 %. ! Total Core Flowis at 59 Mlbm/hr.
, Which one of the following describes the actions to be taken for the present situation? A. No actions required. Monitor for thermal hydraulic instability.
B. ControllM entry is allowed. Monitor for thermal hydraulic instability.
, C. Immediately scram the reactor.
D. Lrw.ediately take actions to exit the region. Monitor for thermal hydraulic instability.
QUESTION l NRC RECORD # WRI 26 ANSWER: D. SYSTEM # B33 K/A 295001 AA2.01: 3.5/3.8 LP# GG-1-LP-RO-B3300 40 AKl.02: 3.3/3.5 OBJ 46,47 LP# OF-LOR-ONEP-LP 401-04 OBJ 19 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-111-3 P/F MAP NEW CLASS l sect. 2.5 for Region II MODIFIED BANK DIFF 2 Fast Speed Recire.
DATE USED: RO SRO BOTH CFR 41.5/41.10/43.5 { l l l l l l l
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_ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _- t U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 27 The plant was operating at 45 % power when a transient on the Entergy Power Grid caused a fast closure on the Turbine Control Valves for the GGNS Main Turbine. The Reactor Water Level Control System maintained level within 10 inches ofnormallevel. Reactor Pressure increased slightly, but was handled by the Bypass valves.
Which one of the following describes the results of this transient? A. The plant remainvi et power with a reduced power due to a Recirculation Pump downshift to slow speed B. The plant remained at power with a reduced power due to a Recire Flow Control Valve Runback.
C. The plant has scrammed and the Recirculation Pumps have downshifted to slow speed.
D. The plant has scrror ued and the Recirculation Pumps have tripped to off.
QUESTION l NRC RECORD # WRI 27 ' ANSWER: C. SYSTEM # B33; K/A 202001 Kl.28 3.9/4.1 C71 LP# GG-1-LP-RO-B3300-00 OBJ 24, 25 l LP# GG-1-LP-RO-C7100.00 I OBJ 9 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-I-1 sect. 4.5 NEW CLASS i Tech Specs 3.3.4.1 & MODIFIED BANK l DIFF 3 3.3.1.1 > DATE USED: RO SRO BOTH CFR 41.5/41.6/43.6 , l-l l l
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_ _ _ _ _ - _ _ _ - _ - - _ _ _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 REACTOR OPERATOR QUESTION 28 The plant has experienced a small steam leak in the Drywell. Drywell pressure is currently 1.75 psig. Containment temperatures 98 'F. The Plant Supervisor has requested that Containment Cooling be maximized Which one of the following statements is correct conceming the ability to maximize cooling in the Containment with tie above conditions? A. Plant Chilled Water flow can be re-established after taking the Auxiliary Building Bypass Switches to BYPASS and restarting Plant ChillWater and stanting all Containment Coolers.
j B. Plant Chilled Water flow is unavailable such that the Containment Coolers are only able to recirculate the air in Containment.
- C. Drywell Chilled Water flow can be re-established by using the keylock switches to l open the Containment isolations and re-established water flow to the Containment Coolers which remain running.
l. D. Drywell Chilled Water is cross connected to Plant Chilled Water such that cooling is re-established to the Containment Coolers.
r , QUESTION l NRC RECORD # WRI 28 ANSWER: B. SYSTEM # P71; K/A 295011 AK2.01: 3.7/4.0 i M41 AA1.01: 3.6/3.9 LP# GG-1-LP-RO-M4100-00 OBJ 7d,e LP# GG-1-LP-RO-P7100.00 OBJ 6,11 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-HI-5 Grcup 6 NEW CLASS MODIFIED BANK l DIFF '2 l DATE USED: RO SRO BOTH CFR 41.9 i
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_ _ _ _ _ _ . ___ - __ - _ _ __ - __ _ _ __ ___ _ _ - _ - -_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l. REACTOR OPERATOR QUESTION 29 l The Control Room has been evacuated due to a freon leak indo the Control Room atmosphere, ! and plant control has been established at the Remote Shutdown Panels.
l l The plant was scranuned and level in the reactor is lowering. RCIC tripped on overspeed and i the MSIVs have closed. The Plant Supervisor has directed the use ofRHR A in the LPCI mode to maintain reactor water level.
l During the lineup ofRHR A in LPCI mode, you notice two handswitches for the LPCI A Injection Valve (E12-F042A). What is the reason for two handswitches? i ' . A. One handswitch is to swap to emergency, removing control from the control room, and l the other handswitch operates the valve OPEN or CLOSED.
B. One handswitch is to remove the auto features of the E12-F042A and allow the other I handswitch to have total control.
C. One handswitch enables the second handswitch to operate the valve in the open and closed positions.
D. One handswitch is used only when the Division I Lockouts have been transferred to 1 insent the pressure interlocks. The second handswitch operates the valve in the open ! and closed positions.
i QUESTION l NRC RECORD # WRI 29 1 ANSWER: C. SYSTEM # C61; K/A 295016 AK2.01: 4.4/4.5 E12 j LP# GG-1-LP-RO-C610040 OBJ 6c SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: E-1181- 037 NEW CLASS , MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.7 l
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 30 The plant is performing the In-Service Leak Test on the reactor following refueling , operations. A miscommunication results in a significant reactor pressure increase.
l Pressure as read on the Control Room Wide range Pressure indication on P680 is pegged upscale.
The Post Accident Pressure recorders indicate that pressure reached 1350 psig.
I Which one of the following is a correct statement with regard to the GGNS Safety Limit for ReactorPressure? A. Reactor Pressure was outside the Safety Limit of 1190 psig because this is referenced on the P680 Wide Range Instnnnent for Tech Specs.
B. Reactor Pressure was outside the Safety Limit of 1325 psig because the Post Accident indication comes from the Water Level instruments reference legs.
C. Reactor Pressure was within the Safety Limit of1375 psig because the Post Accident indication comes from the Bottom Head.. D. Reactor Pressure was within the Safety Limit of 1550 psig.
, QUESTION l NRC RECORD # WRI 30 l ANSWER: B. SYSTEM # K/A 295025 EK1.05: 4.6/4.7 Tech Specs EK1.02: 4.1/4.2 LP# OP-LO-PB-LP-001-02 Generic G2.2.22: 3.4/4.1 OBJ 8b & d G2.2.25: 2.5/3.7 LPN OP-LO-PB-LP-003-00 OBJ 4 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: Tech Specs 2.1.2 NEW CLASS Bases B2.1.2 MODIFIED BANK ' DIFF 3 DATE USED: RO SRO BOTH CFR 41.3/43.2 ! l l 3/17/98 i
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 31 The plant has scrammed due to a Reactor Feed Pump trip.
Reactor level decreased such that RCIC and HPCS auto started and restored level to the normal operating band.
Which one of the following best describes the condition of Ventilation Systems in the Auxiliary Building? A. All fan coil units operating, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are operating.
B. All the fan coil units shutdown, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are operating.
C. All fan coil units operating, the Secondary Containment Isolation Valves open, Standby Gas Treatment is shutdown, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are operating.
D. All the fan coil units shutdown, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are shutdown.
l l QUESTION l NRC RECORD # WRI 31 ANSWER: D. SYSTEM # T41; K/A 288000 K4.02: 3.7/3.8 T42; T43 K4.01: 3.7/3.9 LP# OP-LO-SYS-LP-T41-0? K4.03: 2.8/2.9 OBJ 4, 6, 7e,h A3.01: 3.8/3.8 LP# OP-LO-SYS-LP-T42-02 OBJ 4,6,8e,f LP# GG-1-LP-RO-T4801.00 i OBJ 7a, 9f,g,15 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3 i REFERENCE: ARI 04-1-02-H13-P870 NEW CLASS ! 2A-A3 MODIFIED BANK DIFF 3 05-1-02-III-5 AB VENT DATE USED: RO SRO BOTH CFR 41.3/43.2
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR
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, QUESTION 32 The plant underwent a transient which initiated a reactor scram. All control rods did not fully insert. Reactor water level was intentionally lowered.
One of three analyzed methods of adequate core cooling during emergency conditions is called l
" Steam Cooling with Injection".
What is the minimum reactor level at which this method can be said to be providing adequate core cooling? A. - 167 " B. - 192 "
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C. - 204 " ! F D. Level required to reduce Rx power to < 4%. j QUESTION l NRC RECORD # WRI 32 i ANSWER: B. SYSTEM # K/A 295031 EK1.01:4.6/4.7 EOP Bases LP# GG-1-LP-RO-EP02A.02 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-2A NEW CLASS ' MODIFIED BANK DIFF 3 LOT 7/95 ep & bases DATE USED: RO SRO BOTH CFR 41.10/43.5
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 33
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- The plant was operating at rated conditions when an ATWS occurred Several SRVs lifted and increased Suppression Pool temperature.
l The Plant Supervisor drected the initiation of Standby Liquid Control.
Level in the reactor was lowered to reduce power production.
The crew is now inserting rods by driving and sciruinii4 l Under which one of the following conditions would the reactor be considered shutdown and the termination of Standby Liquid Control be allowed? i
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l l A. All rods are at position 00 except for control rod 32-33 is at position 48.
B. The Reactor Engineer says that subcriticallity can be guaranteed to 200 'F. I
- C. All rods are at position 02 except three in different quadrants at position 04.
! D. Chemistry has confirmed that the Hot Shutdown Boron Weight of SLC has been
. injected l i
l QUESTION l NRC RECORD # WRI 33 ANSWER: A. SYSTEM # K/A 295015 AK1.01: 3.6/3.9 EOP Bases i LP# GG-1-LP-RO-EP02A.02 j OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 ' REFERENCE: 05-S-01-EP-2A NEW CLASS I MODIFIED BANK DIFF 3 LOT 7/95 ep & bases DATE USED: RO SRO BOTH CFR 41.8/43.6
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l U. S. NUCLEAR REGULATORY COMMISSION : WRITTEN EXAMINATION MARCH 1998 l
, REACTOR OPERATOR QUESTION 34 The plant is in a refueling outage with the reactor vessel head removed.
Core Alterations are in progress when the Refueling Bddge operator has a spent fuel bundle at the full up position traversing the Upper Containment Pool in the Reactor cavity area when the Fuel Grapple malfunctions and releases the fuel bundle. The bundle drops into the reactor vessel and falls against me north west wall of the reactor. The Refueling Bridge operator notices a large bubble start rising from the area of the fuel bundle.
Which one of the following describes the required actions for this situation? A. Move the Refueling Bridge to the Upper Containment Pool Fuel Storage area and wait there for the bubble to pass.
B. Pick the bundle up with the grapple and place it into the nearest fuel storage location, then move the Ramaling Bddge to the Cattle Chute.
C. Suspend core alterations and evacuate the Refueling Bddge area and suspend fuel handling until the cause can be determined j D. Keep the Refueling Bridge manned in its present position and contact the Refueling Floor Health Physicist to the Refueling Bddge and take radiation surveys.
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QUESTION l NRC RECORD # WRI 34 ANSWER: C. SYSTEM # ONEP K/A 295023 AA2.04: 3.4/4.1 AKl.01: 3.6/4.1 AK3.01: 3.6/4.3 LP# OP-LOR-ONEP-LP-001-04 Generic G2.4.11: 3.4/3.6 OBJ 1 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 3 REFERENCE: 05-1-01-II-8 sect. 2.1 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/41.12/ l 43.4/43.5/43.7 l 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 35 Given the following conditions: Reactor power 45% Reactorlevel-100 inches Reactor pressure 850 psig Suppression pooltemperature 125'F Suppression pool level 20 feet 5 inches 2 SRVs are open Which one of the following best describes the required actions to be taken given the above conditions? A. Immediately commence an Emergency Depressurization in accordance with EP-2A because limits in the Containment have been exceeded based on Suppression Pool Temperature.. B. Close the two SRVs and increase the Reactor Pressure band to a top end of 1000 psig, to reduce the amount of heat entering the Suppression Pool.
C. Lower Reactor Pressure using cooldown rates that may exceed 100 'F/Hr, to avoid l jeopardizing Containment by exceeding the heat capacity temperature limit of the Suppression Pool.
D. Conditions at present are acceptable, however all pumps taking a suction from the Suppression Pool should be secured.
QUESTION l NRC RECORD # WRI 35 ANSWER: C. SYSTEM # EOP K/A 295026 EKl.02: 3.4/3.8 HCTL Curve AK1.01: 3.6/4.1 AK3.01: 3.6/4.3 LP# GG-1-LP-RO-EP02A.02 Generic G2.4.11: 3.4/3.6 OBJ 2, 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 l REFERENCE: 05-1-01-EP-2 NEW CLASS l steps 38 & 40 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.9/41.10/ 43.5 3/17/98 i - - - - - _ _ _ _ . - _ - - - - -- j
- _ _ _ _ _ - - _ . _ _ _ _ _ _ - - - U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 36 The plantisin Mode 3,hci Shutdown. i A Station Blackout has occurred.
The Division III Diesel Generator is the only available source of4.16 Kv power.
High Pressure Core Spray has failed to start.
Reactor Levelis at - 140 inches and decreasing.
Preparations are being made to cross tie the Division III D/G to the Division II ESF Bus.
At a minimum, who must approve the cross tie ofDivision III D/G and WHY? A. Electrical Superintendent because this action could over load the Division III Diesel Generator.
B. General Manager, Plant Operations because this will require a change to a Safety Related Procedure.
C. Manager, Operations because he is required to approve all deviations from normal Operations Procedures.
D. Plant Shift Superintendent because the action is a deviation from requirements of 10 CFR 50 and the GGNS Operating License i QUESTION l NRC RECORD # WRI 36 ! ANSWER: D. SYSTEM # ONEP K/A 295003 AK1.06: 3.8/4.0 Generic G2.4.7: 3.1/3.8 G2.4.11: 3.4/3.6 LP# OP-LOR-ONEP-LP-001-04 G2.4.22: 3.0/4.0 OBJ 9 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 , REFERENCE: 05-1-01-I-4 sect. 3.2.8 NEW CLASS I 10 CFR 50.54x MODIFIED BANK DIFF 2 10 CFR 50 APP. A DATE USED: Criteria 17 RO SRO ROTH CFR 41.7 ' 3/17/98
_ _ _- ___ __ -__ _ . _ _ _ _ _ _ _ - _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 37 The plant is in a Refueling Outage with the reactor disassembled five (5) days after the plant was shutdownin Refueling Outage 08.
Reactor Coolant Temperature is 140 *F Reactor Water Level is at the Main Steam lines following Steam Line Plug "mstallation.
TheFuel shufBe has not begun.
i The inservice shutdown cooling pump hasjust tripped off.
l Assume no funher operator action.
Determine for this condition: I 1. The time to BOIL for the Reactor Vessel 2. The time for level to reach the Top of Active Fuel.
A. 1) 0.75 hours 2) 11 hours B. 1) 0.75 hours 2) 15 hours C. 1) 1.5 hours 2) 11 hours D. 1) 1.5 hours
- 2) 15 hours I
QUESTION l NRC RECORD # WRI 37 l ANSWER: A. SYSTEM # ONEP K/A 295021 AKl.01: 3.6/3.8
- LP# OP-LOR-ONEP-LP-001-04 AA2.01
- 3.5/3.6 OBJ 14
' SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: 05-1-01-11I-1 Att. I NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.5/43.5 i 3/17/98-t
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR i QUESTION 38 l Which one of the following describes the conditions that Cold Shutdown Boron Weight is designed to over come? , A. 70 F, xenon free, water level at steam lines, 50 % rod density.
B. 70 *F. xenon free, water level in normal band, all rods fully withdrawn. I
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C. 100 F, xenon free, water level in normal band, all rods fully withdrawn.
D, 100 'F, xenon free, water level at steam lines, 50 % rod density.
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QUESTION l NRC RECORD # WRI 38 ANSWER: B. SYSTEM # K/A 295037 EK3.05: 3.2/3.7 EOP-2A BASES LP# GG-1-LP-RO-EP02A.00 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-2A Bases NEW CLASS Step 21 MODIFIED BANK DIFF 2 . LOT 3/98 ep& bases DATE USED: RO SRO ROTH CFR 41.6/41.10/43.6 , , 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 39 The plant is operating at 100% power with the Offgas Post Treatment A Radiation Monitor tagged out for a power supply replacement.
A Non-Licensed Operator and trainee while conducting training take the handswtich for Offgas Post Treatment Radiation Monitor B to TEST.
Which one of the following describes the impact of this action?
(Assume no further operator action.)
A. The Control Room and back panels will receive an annunciator / alarm only.
B. The steam supply for the Steam Jet Air Ejector will isolate causing a loss of condenser vacuum.
C. The Offgas Charcoal Adsorber Bypass Valve, N64-F045 will close and open the Charcoal Adsorber Outlet Isolation Valve N64-F060 to place the system in treat mode.
D. The Offgas Charcoal Adsorber Outlet Isolation Valve N64-F060 will isolate causing a loss of condenser vacuum to occur.
I l QUESTION l NRC RECORD # WRI 39 ANSWER: D. SYSTEM # D17; K/A 272000 K3.05: 3.5/3.7 N64; N62 LP# GG-1-LP-RO-D1721.00 OBJ SR9 TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS 19A-C8 MODIFIED BANK DIFF 2 Tech Spec Loop Logics DATE USED: RO SRO BOTIf CFR 41.11/43.4
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' WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 40 The plant was operating at 100 % power when Suction Valve, N62-F003 A, for the in-service Steam Jet Air Ejector automatically closed. The Operator-at-the-Controls noticed Main Condenserlosing vacuum.
Which of the following best describes the automatic actions that will occur on a degrading condenser vacuum to 0 in Hg vacuum and which actions have a bypass available? A. 21" vac, Main turb trip' no bypass 16" vac, Main bypass valves close no bypass 12" vac, Rx feed pumps trip bypass available 9" vac, MSIV closure bypass available B. 21" vac, Main turb trip no bypass 16" vac, Rx feed pumps trip ' bypass available > 12" vac, Main bypass valves close no bypass 9" vac, MSIV closure bypass available i C. 21" vac, Main turb trip no bypass 16" vac, MSIV closure bypass available 12"vac, Main bypass valves close bypass available 9" vac, Rx feed pumps trip bypass available D. 21" vac, Main turb trip no bypass 16" vac, MSIV closure no bypass 12" vac,Rx feed pumps trip bypass available 9" vac, Main turb bypass valves close bypass available l ! QUESTION l NRC RECORD # WRI 40 ANSWER: B SYSTEM # N62 K/A 295002 AK1.03: 3.6/3.8 LP# CG-1-LF-RO-N6200.00 OBJ 7 LP# OP-LOR-ONEP-LF 001-04 OBJ 28 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-V-8 sect. 5.0 NEW CLASS MODIFIED BANK DIFF 2 LOT 3/98 stm cond DATE USED: RO SRO-BOTH- CFR 41.4 3/17/98 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ -
(_ _ _ __ _ ____ _-___ _ , U. S. NUCLEAR REGULATORY COMMISSION i WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 41 l l Plant conditions are as follows: Mode 1 l ReactorPower: 25 % , Which one of the following describes the response of the RCIS system if the Main Turbine were to trip with no reactor scram? RCIS will: l A. implement the constraints of the Rod Withdrawd Limiter allowing rod movements of up to 4 notches.
B. implement the constraints of the Rod Pattern Controller and depending on pattern j initiate Insen and/or Withdraw blocks.
C. be between the Rod Pattern Controller and the Rod Withdrawal Limiter indicating the l Low Power Alarm Point vith NO constraints on rod motion.
D. implement the constraints of the Rod Withdrawal Limiter allowing rod movements of up to 2 notches. .
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l QUESTION l NRC RECORD # WRI 41 ANSWER: B. SYSTEM # C11-2; K/A 295005 AA1.03: 2.7/2.8 N32-2 201005 K6.01: 3.2/3.2 l LP# CC-1-LP-RO-C1102.02 A1.01: 3.3/3.3 OBJ 6,16,22 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 1 REFERENCE: Tech Specs 3.1.6 NEW CLASS
& 3.3.2.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.7 l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR l QUESTION 42 The plant is operating at 20 % power when a Loss ofInstrument Air results in a reactor scram.
, The los ofinstrument air is a rupture of the instmment air header in the Water Treatment i Building. The Plant Supervisor has directed the Control Room Operator to maintain water levelin the reactor.
l Which of the following best describes the response of the Condensate and Feedwater System? l A. Feeding of the Reactor is not available with Condensate and Feedwater due to the isolation ofCondensate Cleanup System.
> B. Feeding of the Reactor is not available due to all of the Condensate and Feedwater Minimum Flow Valves failing open diverting all flow to the Condenser.
C. Feeding of the Reactor is available from the Feedwater system while steam is available to the RFPTs and afterwards at lower reactor pressures using Condensate and Booster pumps through the startup level control valve.
D. Feeding of the Reactor is available from the Feedwater system while steam is available to the RFPTs and afterwards at lower reactor pressures using Condensate and Booster pumps through the startup level control bypass valve.
QUESTION l NRC RECORD # WRI 42 ANSWER: D. SYSTEM # N19; K/A 259002 K6.01: 3.2/3.2 N21; N22; P53 LP# GG-1-LP-RO-N2100.00 ON 9e,12 LP# GC-1-LP-RO-N1900.00 ON 21,22g,25- SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: . 05-1-02-V-9 NEW CLASS sect. 5.21 - 5.24 MODIFIED BANK . DIFF 3 05-1-02-V-7 sect. 2.1.4 bank quest.
l DATE USED: RO SRO ROTN CFR 41.5
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' ! U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 3 REACTOR OPERATOR QUESTION 43 The plantis operating at 60% power.
Ti.e C011 A HPU oil pump is in operation for "A" RFPT with the C010A HPU oil pump tagged out for repairs.
HPU Oil Pump C011 A trips on motor overcurrent.
Which of the following describes the response of the Feedwater system and the plant? l (Assume no operator actions.)
A. The A RFPT will lockup at its present speed such that any changes in feed flow will have to be controlled by the B RFPT.
B. The A RFPT will trip on low governor oil pressure causing level to drop to level 3 causing a reactor scram on low level and a Recire Pump downshift.. ! C. The A RFPT will runback to minimum flow and the B RFPT will increase speed automatically to compensate.
D. The A RFPT will trip on low governor oil pressure and the B RFPT will increase speed automatically to compensate, a Recirc Flow Control Valve runback may occur.
QUESTION l NRC RECORD # WRI 43 ANSWER: D. SYSTEM # N21 K/A 259001 A1.05: 2.8/2.7 LP# GG-1-LP-RO-N2100.00 OBJ 7d,8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: 04-1-02-II13-P680 NEW CLASS 2A-A2 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 l l
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_ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ -_____ __ _ ____ _. _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 44 GGNS Main Generator has a limit to carry no more than i 250 MVARs.
What is the basis for this limitation? A. This is the Maximum reactive load allowed by the manufacturer due to the heat build up in the stator windings at full power.
B. GGNS is a base load station such that Entergy dispatchers are required to minimize the reactive load canied on the Main Generator.
C. GGNS Main Generator reverse power relays will not recognize a reverse power condition at high reactive load and will not provide the required trip.
D. The Generator V-Curves supplied by the manufacturer limit the power factor on the generator to reduce hysteresis losses.
QUESTION l NRC RECORD # WRI 44 ANSWER: C. SYSTEM # N41 K/A 245000 A4.14: 2.5/2.5 A3.10: 2.5/2.6 LP# GG-1-LP-RO-N1/51.00 K4.06: 2.7/2.8 OBJ 14 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 ; REFERENCE: 04-1-01-N40-1 seet. 3.8 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH I 3/17/98 l l l_ ._ . _ _ _ _ j
_ - _ __ --_ _ __ _ -_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR i- QUESTION 45 l 1' Inverter 1Y95 is operating on its alternate source following a transfer using its manual bypass switch. The operator has been requested to return the Inverter back to its normal power sourCC.
Which one of the following is NOT required to transfer the Inverter load back to the normal source? A. Alternate and Normal Source voltages to be matched.
B. Alternate and Normal Source output currents to be matched.
. C. Altemate and Normal Source frequencies to be matched.
D. Altemate and Nonnal Source phases to be in sync.
QUESTION l NRC RECORD # WRI 45 ANSWER: B. SYSTEM # L62 K/A 262002 K4.01: 3.1/3.4 LP# GG-1-LP-RO-L6200.00 OBJ 7, 11 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-L62-1 sect. 3.2 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 t REACTOR OPERATOR QUESTION 46 The plant is operating at full power. Health Physics personnel are setting up a controlled area at the Containment Steam Tunnel and cause the CCW to the RWCU Non-Regenerative Heat Exchanger valve (P42-F103) to go closed.
What is the affect of this valve going closed on the RWCU System?
(Assume no operator actions.)
A. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that RWCU Pump Suction CTMT OTBD Isol valve (G33-F004) will isolate, which willtrip the RWCU Pumps.
B.- The Non-Regenerative Heat Exchanger outlet temperature will increase to the po'mt that RWCU Pump Suction DRWL INBD Isol valve (G33-F001) and RWCU Supply to RWCU HXS valve (G33-F251) will isolate, which will trip the RWCU Pumps.
i C. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that RWCU Filter Demin Bypass valve (G33-F044) will open and lock the Filter Deminsin hold.
D. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that automatically reopens CCW to the RWCU Non-Regenerative Heat Exchanger valve (P42-F103). Iftemperatures continue to increase, at 150 *F the FilterDemins j will lock in hold, and RWCU Filter Demin Bypass valve (G33-F044) will open.
! QUESTION l NRC RECORD # WRI 46 ANSWER: A. SYSTEM # G33/36; K/A 204000 A3.04: 3.4/3.5 P42 A2.01: 3.2/3.4 LP# CG-1-LP-RO-C3336.01 A2.14: 3.2/3.2 OBJ 8, 9,10, 161, SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 17- l REFERENCE: 04-1-01-G33-1 sect. 3.1 NEW CLASS ARI 04-1-02-H13-P680 MODIFIED BANK lL DIFF 3 11A-C6 DATE USED: RO SRO BOTH CFR 41.4 i 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR' OPERATOR QUESTION 47 RHR A is operatmg in Suppression Pool Cooling Mode.
The plant experiences a LOCA.
What will be the response of the RHR A System and how can the system be returned to Suppression Pool Cooling Mode ofoperation with the LOCA? TheRHR A Systemwill: A. isolate RHR A Test Retum to the Suppression Pool valve (E12-F024A), and open RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for LPCI mode it is unable to be returned to Suppression Pool Cooling.
B.' - isolate RHR A Test Retum to the Suppression Pool valve (E12-F024A), and open RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for LPCI mode, E12-F024A and E12-F048A can be immediately manually overridden for Suppression PoolCooling.
( C. isolate RHR A Test Return to the Suppression Pool valve (E12-F024A), and open l RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for l - LPCI mode, E12-F024A can be humediately manually ovenidden open and E12- ' F048A closed after a time delay.
D. require manual realignment to the LPCI mode by closing RHR A Test Retum to the , Suppression Pool valve (E12-F024A), and opening RHR Heat Exchanger Bypass valve (E12-F048A). Once the system is realigned for LPCI mode, E12-F024A can be manually overridden open and E12-F048A closed after a thne delay, QUESTION l NRC RECORD # WRI 47 l ANSWER: C. SYSTEM # E12 K/A 219000 A1.08: 3.7/3.6 A2.14: 4.1/4.3 A3.01: 3.3/3.3 l LP# OP-LO-SYS-LP-E12-07 A4.06: 3.9/3.7 OBJ 7, 9 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-E12-1 sect. 3.4 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 3 20A-C6; 20A-B5 DATE USED: 05-1-02-HI-5 Group 5 RO SRO BOTH CFR 41.9 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 REACTOR OPERATOR QUESTION 48 RHR B Pump Room temperature increases to 170 * F.
Which one of the following identi6es the systems or components besides RHR B which will be affected by this temperature? A. RCIC B. 'RWCU C. RCIC and MSIVs D. HPCS QUESTION l NRC RECORD # WRI 48 ANSWER: A. SYSTEM # E12 K/A 219000 A1.08: 3.7/3.6 A2.14: 4.1/4.3 A3.01: 3.3/3.3 LP# GG-1-LP-RO-E5100.00 OBJ 14 LP# OP-LO-SYS-LP-E12-07 A4.06: 3.9/3.7 OBJ 8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 17.S-06-5 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 2 20A-B1 DATE USED: 05-1-02-III-5 Group 2,3, RO SRO BOTH CFR 41.4 1
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION hMRCH 1998 REACTOR OPERATOR QUESTION 49 RCIC is in a standby lineup aligned to the Condensate Storage Tank (CST) for a suction source.
Weeping SRVs cause Suppression Pool Level to increase such that High Suppression Pool Water Level alarms are received on H13-P601 and H13-P870 panels.
Which one of the following describes the response of the RCIC system to this condition? A. RCIC Saction from the Suppression Pool E51-F031 will open and E51-F010 Suction from the CST will close. This lineup can be manually overridden back until Suppression Pool Level is returned to normal at which time the transfer signal will clear.
B. RCIC Suction from the Suppression Pool E51-F031 will open and E51-F010 Suction from the CST will close. This lineup CANNOT be overridden back until the high Suppression PoolLevelis cleared.
C. RCIC Suetions will remain in standby configuration until a RCIC initiation signal is received at which time they will transfer with RCIC Suction from the Suppression Pool E51-F031 opening and E51-F010 Suction from the CST closing.
D. RCIC Suction from the Suppression Pool E51-F031 will open and E51-F010 Suction f from the CST will close. When Suppression Pool Level is returned to normal the lineup will automatically return to the original standby lineup.
QUESTION l NRC RECORD # WRI 49 ANSWER: B. SYSTEM # E51 K/A 295029 EA1.04: 3.4/3.5 LP# GG-1-LP-RO-E5100.00 OBJ 14 SROTIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 04-1-01-E51-1 sect.3.7 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 2 21A-C5 DATE USED: RO SRO BOTH CFR 41.4 I 3/17/98 _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR
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QUESTION 50 RCIC is operating in response to a LOCA signel.
Reactor water level increases to 55 inches.
l What is the expected response of RCIC?
(Assume no further operator actions.)
, A. The RCIC Steam Supply Isolation Valve, E51-F045, will close and the RCIC Trip / Throttle Valve will trip stopping the RCIC Turbine.
B. The RCIC TripfIhrottle Valve will trip and the RCIC Injection Valve E51-F013 will close.
l C. The RCIC Steam Supply Isolation Valves, E51-F063 and F064, will close and the ' RCIC Suppression Pool Suction Valve, E51-F031, if open will isolate.
D. The RCIC Steam Supply Isolation Valve, E51-F045, will close and the RCIC Injection j Valve E51-F013 will close. 1 l QUESTION l NRC RECORD # WRI 50 ANSWER: D. SYSTEM # E51 K/A 217000 A3.05: 3.9/3.9 LP# GG-1-LP-kO-E5100.00 OBJ Sc,j SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-1185- 1, 6, 34, 35, 42 NEW CLASS ARI 04-1-02-H13-P680 MODIFIED BANK DIFF 2 4A2-D1 DATE USED: RO SRO BOTH CFR 41.7 l l 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR GPERATOR.
QUESTION 51 A fire in Panel 11DK caused a complete loss of the K DC Bus.
Which one of the following is the effect this loss will have on the ATWS ARI/RPT System? A. The ATWS ARI portion of the system is INOP due to the loss of halfof the valves required to make the system operate, ATWS RPT will still trip the Recire Pumps.
B. The ATWS ARI portion of the system will operate because half of the valves will energize and depressurize the air header, ATWS RPT will still trip the Recirc Pumps.
C. The ATWS ARI portion of the system is INOP due to the loss of halfof the valves required to make the system operate, ATWS RPT is INOP because the Slow Speed breakers will NOT operate to trip the Recire Pumps.
D. The ATWS ARI portion of the system will operate because half of the valves will energize and depressurize the air header, ATWS RPT is INOP because the Slow Speed breakers will NOT operate to trip the Recire Pumps.
QUESTION l NRC RECORD # WRI 51 ANSWER: A. SYSTEM # C11; K/A 263000 K3.03: 3.4/3.8 B33 K3.02: 3.5/3.8 LP# GG-1sLP-RO-B3300.00 201001 K2.05: 4.5/4.5 OBJ 27 LP# GG-1-LP-RO-C111A.00 OBJ 9h, 18 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-6066 - 2, 5, 6 NEW CLASS MODIFIED BANK DIFF 4 DATE USED: RO SRO ROTH CFR 41.4 l 3/17/98 l L - - - - - - - - -
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 52 The plant is at 100 % power with the Electrical Distribution System in a preferred lineup.
Service Transformer 11 locked out on sudden pressure.
Which one of the following describes the status of the Main Steam Isolation Valves (MSIVs)? A. Inboard MSIVs have Isolated and the Outboard MSIVs are Open.
B. Inboard MSIVs are Open and the Outboard MSIVs have Isolated.
C. All MSIVs are Open with a loss of power to half of their pilot solenoids.
D. AllMSIVs areIsolated.
QUESTION l NRC RECORD # WRI 52 ANSWER: C. SYSTEM # B21; K/A 239001 K2.01: 3.2/3.3 R21; C71 K5.08: 2.6/2.7 LP# GG-1-LP-RO-C7100.00 K6.01: 3.1/3.3 OBJ 5,8d LP# GG-1-LP-RO-B1300.00 OBJ 4 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-C71-1 Att. III NEW CLASS 05-1-02-III-2 sect. 5.2 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 i . 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 53 The plant is at 100 % power.
The Main Steam Line Radiation Alarms for all four (4) divisions come in indicating rad levels greater than 3 times normal background.
Which one of the following describes the response expected from this signal? A. MSIVs and Reactor Sample Valves isolate, and the Reactor Scrams. ! i B. Reactor Sample Valves isolate and the Reactor Scrams.
C. MSIVs and Reactor Sample Valves isclate.
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D. Reactor Sample Valvesisolate.
QUESTION l NRC RECORD # WRI 53 ANSWER: D. SYSTEM # B21; K/A 295033 EK3.03: 3.8/3.9 D17 LP# GG-1-LP-RO-D1721.00 OBJ 13 LP# 0P-LOR-ONEP-LP-001-04 OBJ 31 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-III-5 Group 10 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 2 18A-C4; 19A-C4 DATE USED: RO SRO BOTH CFR 41.11/43.4 l l l l I 3/17/98
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_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 54 The plant is operating at 100 % power when the 11DA BUS is de-energized.
Electricians have found the cause of the loss and effected repairs and are ready to restore the bus to service.
Which one of the following describes how an inadvertent initiation of Division I ECCS Systems is prevented? A. The Initiation RESET Pushbutton for Division I is to be held depressed to prevent the initiation logic from picking up.
B. The Trips units for the ECCS initiation logics are powered from DC, however, the instrumentation which supply the inputs to the trip units are powered from Uninterruptable Power System (UPS) and have remained energized and in a normal state.
C. The Instrumentation which supply's the ECCS Initiation Logic Trip Units is arranged such that the instrument, on a loss of power, fails in a direction which will not cause an actuation of the ECCS Systems.
D. The Trip units for the ECCS initiation logics are arranged such that a time delay relay allows instrumentation and logics time to re-energize prior to sending an initiation signal.
QUESTION l NRC RECORD # WRI 54 ANSWER: D. SYSTEM # E12; K/A 295004 AKl.06: 3.3/3.6 E21; L11 LP#OP-LO-SYS-LP-E12-07 OBJ 9e,10e, 13 LP# GG-1-LP-RO-E2100.00 OBJ 6, 13, 16 SRO TIER 1 GROUP 2 / RO TIER I GROUP 2 REFERENCE: E-li82-23 & 26 NEW CLASS E-1181-68 & 82 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.7 3/17/98 _ - _ _ _ _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 55 , l The plant is in a startup following a 32 day outage. MSIV; are closed. Recire Loop Temperatures are at 180 * F. Control rods are being withdrawn to achieve criticality.
(minimal decay heat) The Operating CRD Pump tripped.
What will be the response of the plant? i (Assume no further operator actions) l A. The reactor wraer level will increase to the point that a reactor scram is received on High waterlevel.
B. The reactor water level will decrease to the point that a reactor scram is received on Low water level.
C. The plant will scram due to a loss of charging water pressure to the Hydraulic Control Units.
D. The reactor water level will remain stable at its present level.
QUESTION l NRC RECORD # WRI 55 ANSWER: B. SYSTEM # C11-1A; K/A 295022 AK2.04: 2.5/2.7 G33/36; IOI- 1 AK2.05: 2.4/2.5 LP# GG-1-LP-RO-G3336.01 AA1.04: 2.5/2.6 OBJ 2,12,21 LP# GG-1-LP-RO-C111A.00 OBJ 18 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 03-1-01-1 NEW CLASS sect. 2.2.5; 3.3.1d; 3.3.3a MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.5 j l l 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 56 Which one of the following is the basis for the automatic initiation of Standby Gas Treatment on High Ventilation Radiation Levels in Secondary Containment? l A. This provides for the recirculation of the Secondary Containment atmosphere without ! exhausting air outside of Containment.
l B. This provides for the filtration of the Secondary Containment atmosphere of l l radionuclides prior to their release into the environment, maintaining offsite releases to withinlimits.
! C. This provides for the cleeaup of the Secondary Containment atmosphere allowing personnel entry into the Secondary Containment during a DBA LOCA.
D. This provides for the maintenance of a positive pressure in the Secondasy Containment, therefore preventing any of the fission products released into the Containment from i being released into the environment.
QUESTION l NRC RECORD # WRI 56 ANSWER: B. SYSTEM # T48 K/A 295034 EK3.02: 4.1/4.1 LP# GG-1-LP-RO-T4801.00 261000 Kl.01: 3.4/3.6 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: Tech Specs B3.6.4.3 NEW CLASS MODIFTED BANK DIFF 2 i DATE USED: RO SRO ROTH CFR 41.7/41.13/43.4 i i ! 3/17/98
U. S. NUCLEAR REGULATORY COhtMISSION WRITTEN EXAMINATION MARCH 1998 ! REACTOR OPERATOR l QUESTION 57 The plant is operating at 100 % power normal operations.
l Which one of the following describes how the Secondary Containment Ventilation Systems prevent the release of radioactive contaminants? , , A. The Auxihary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation l Systems work together to maintain the Auxiliary Building at a negative pressure and ! monitor the exhaust to the atmosphere. Irradiation levels are excessive, signals are
sent to isolate the building and initiate an atmospheric treatment system.
l B. The Auxihary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a negative pressure and treet the exhaust of the Auxiliary Building to prevent any release of radioactive l materials.
l l C. The Auxdiary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a positive pressure and monitor the exhaust to the atmosphere. Irradiation levels are excessive, signals are ! sent to isolate the building.
D. The Auxiliary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a positive pressure and r monitor the exhaust to the atmosphere. Irradiation levels are excessive, signals are I sent to initiate an atmospheric treatment system.
, QUESTION l NRC RECORD # WRI 57 L ANSWER: A. SYSTEM # T42; K/A 295038 EK2.03: 3.6/3.8 I' T41; T48 EA1.06: 3.5/3.6 I LP# GG-1-LP-RO-T4801.00 l OBJ 7a,15
<
LP# OP-LO-SYS-LP-T42-02 OBJ 1, 2 LP# OP-LO-SYS-LP-T41-03 0"J 1,2r 4 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 04-1-01-T48-1 sect. 3.2 NEW CLASS 04-1-01-T42-1 sect. 3.1 MODIFIED BANK DIFF 2 05-1-02-IH-5 AB Vent DATE USED: ARI 04-1-02-H13-P870 RO SRO BOTH CFR 41.13/43.4 2A-A3 i 3/17/98 i
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR Q~UESTION 58 The plant is operating normally at 100 % power.
The Suppression Pool Hi/Lo Level and a LPCS Room Sump Level Hi-Hi annunciators have been received on the H13-P870 panel.
The Control Room Operator has noted that Suppression Pool Levelis at 18.4 feet. An operator dispatched to the room reports that water is spraying from the LPCS Suction piping, but he was unable to tell the exact location.
Which one of the following is appropriate actions for this event? A. Immediately scram the reactor, initiate Suppression Pool Makeup, and emergency depressurize the plant, and isolate the LPCS Suction from the Suppression Pool.
B. Ensure the LPCS Room sump pumps are operating, isolate LPCS Suction from the Suppression Pool and observe the status of the leak and makeup to the Suppression ' Pool via normal means, ifrequired open the LPCS Room Door. , C. Monitor and control LPCS Room sump levels, rack out the LPCS Pump Breaker and isolate LPCS Suction from the Suppression Pool, scram the reactor since the Max Safe l Level has been reached.
, D. Verify the LPCS Room sump pumps are operating, isolate LPCS Suction from the , Suppression Pool and rack out the LPCS Pump Breaker, and observe the status of the leak and makeup to the Suppression Pool via normal means.
, QUESTION l NRC RECORD # WRI 58 l ANSWER: D. SYSTEM # P45; K/A 295036 EA2.03: 3.4/3.8 i E12; EOP- 4 EK3.03: 3.5/3.6 I LP# OP-LO-EP-LP-005-03 EA2.02: 3.1/3.1 OBJ 2 LP# OP-LO-EP-LP-006-03 OBJ Sa SRO TIER 1 GROUP 2/ RO TIER 1 GkOUP 3 REFERENCE: 05-S-01-EP-3 step 48 NEW CLASS 05-S-01-EP-4 step 18 MODIFIED BANK DIFF 2 ARI 04-1-02-H13-P680 8Al-A4 DATE USED: ARI 04-1-02-H13-P870 RO SRO BOTH CFR 41.4 4A-A3; 2A-F1; 4A-C3 l 3/17/98 , ,
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 I REACTOR OPERATOR QUESTION 59 The reactor is shutdown and the plant is in a forced cooldown to achieve cold shutdown conditions.
Which one of the following best describes the method used to control CRD Flow and Drive pressure during the depressurization process? A. The Pressure Control Valve automatically throttles to maintain 250 psid Drive DP and the Flow Control Valve automatically throttles in response to a CRD flow setpoint of 60 GPM.
B. The Pressure Control Valve automatically throttles to maintain 250 psid Drive DP and the Flow Control Valve is manually throttled to maintain a CRD flow of 60 GPM.
C. The Pressure Control Valve is manually throttled to maintain 250 psid Drive DP and the Flow Control Valve automatically throttles in response to a CRD flow setpoint of 60 GPM.
D. The Pressure Control Valve is manually throttled to maintain 250 psid Drive DP and the Flow Control Valve is manually throttled to maintain a CRD flow of 60 GPM.
QUESTION l NRC RECORD # WRI 59 ANSWER: C. SYSTEM # C11-1A K/A 201001 K4.08: 3.1/3.0 LP# GG-1-LP-RO-C111A.00 OBJ 9d & e,13 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: M - 1081-B NEW CLASS E-1166- 003; 017 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6
3/17/98
_ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 60 l
l Which one of the following is the reason the LPCI Injection Valves, E12-F042A, B, and C, are designed to remain closed at normal reactor vessel pressure following a LOCA initiation signal? ) A. This allows the pump time to pressurize the header, thus minimizing the differential pressure across the injection valve.
l B. This ensures reactor pressure has decreased sufficiently to prevent the possibility I of overpressurizing low pressure piping.
C. This allows the pump to develop enough discharge head to overcome reactor pressure for injection preventing backflow of hot reactor water. I l D. This ensures reactor pressure has eqt.-lized with LPCI pressure to prevent the injection check valve E12-F041 A, B, C from slamming the injection piping causing damage.
l QUESTION l NRC RECORD # WRI 60 ANSWER: B. SYSTEM # E12 K/A K1.17: 4.0/4.0; K4.01: 4.2/4.2 203000 K4.02: 3.3/3.4; A3.01: 3.8/3.7 LP# OP-LO-SYS-LP-E12-07 A3.08: 4.1/4.1; A4.08: 4.3/4.3 OBJ 7h, 14 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-E12-1 seet. 3.4 NEW CLASS Tech Spec Bases B3.3.5.1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.8 3/17/98
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I U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 61 The plant isin an ATWS.
Standby Liquid Control is out ofsersice.
Which one of the following is a means of alternate Boron injection? i i A. Add sodium pentaborate to the RWCU Filter Demin Precoat tank and inject the boron l into the reactoi via RWCU.
i B. Add sodium pentaborate to the Suppression Pool and align RHR A or B m Suppression Pool Cooling mode to mix the solution then inject through any available ECCS pump taking a suction from the Suppression Pool.
I C. Add sodium pentaborate to the Condensate Storage Tank and mix with HPCS and
'
inject the boron into the reactor using RCIC with a suction on the CST.
D. Add sodium pentaborate to the Condensate Cleanup Precoat Tank and inject into the Condensate Cleanup system and use the Condensate / Feedwater Systems to provide a differential pressure to inject the boron into the reactor.
QUESTION l NRC RECORD # WRI 61 j ANSWER: C. SYSTEM # K/A K3.01: 4.3/4.4 EOP-2A 211000 LP# GG-1-LP-RO-EP02A.02 OBJ 3 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 05-S-01-EP-2 Att. 28 NEW CLASS MODIFIED BANK DIFF 2 i DATE USED: RO SRO ROTH CFR 41.6 l ! l . 3/17/98 l ;
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION. 62 The Reactor Wide Range Level Instrument on condensing pot D004A has its reference leg flash.
How will this affect the ability of an automatic Containment /Drywell Isolation? A. The Division I Containment /Drywell Isolation valves will isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will isolate on a low reactor waterlevel signal.
B. The Division I Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will-NOT isolate on a low reactor water level signal.
C. The Division I Containment /Drywell Isolation valves will isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal.
D. The Division I Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will . isolate on a low reactor water level signal. I i QUESTION l NRC RECORD # WRI 62 ANSWER: D. SYSTEM # B21; K/A 216000 K3.02: 4.0/4.3 M71 LP# GG-1-LP-RO-B2101.00 OBJ 5,7a, 8b,14 LP# GG-1-LP-RO-M7101.00 OBJ 7g, 9, 17 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 17-S-06-5 Att. II NEW CLASS Group 6A MODIFIED BANK DIFF 2 05-1-02-III-5 Group 6 DATE USED: RO SRO BOTH CFR 41.6 i l I
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 I L REACTOR OPERATOR QUESTION 63 , Under which of the following conditions would the control rods most likely insert l themselves WITHOUT assistance of the Control Rod Drive System and its Scram accumulators? l A. Reactor Pressure is 620 psig.
l Instrument Air to the Scram Air Header is pressurized.
B. Reactor Pressure is 620 psig.
Instrument Air to the Scram Air Header is depressurized.
! C. Reactor Pressure is 420 psig.
l Instrument Air to the Scram Air Header is pressurized.
D. Reactor Pressure is 420 psig.
Instrument Air to the Scram Air Header is depressurized.
QUESTION l NRC RECORD # WRI 63 ANSWER: B. SYSTEM # C11-1B K/A 201003 K3.02: 4.0/4.3 LP# CG-1-LP-RO-C111B.00 OBJ 13, 15 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 REFERENCE: Tech Spec Bases B3.1.5 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.2 f l 3/17/98
- __ - - _ _ _ _ -__ _ _ _ _ _ - _ _ _ ___ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 64 Much one of the following situations would require the Hydrogen Recombiners to be prevented from operation? A. Reactor Level is - 170 inches, Drywell Hydrogen Concentration is at 8.2 % and Containment Pressure is 10 psig.
B. Reactor Level is undetermined, Drywell Hydrogen Concentration is at 6.2 % and Containment Pressure is 1.50 psig.
C. Reactor Level is undetermined, Containment Hydrogen Concentration is at 8.2 % and Containment Pressure is 10 psig.
D. Reactor Level is - 170 inches, Containment Hydrogen Concentration is at 4.2 % and Containment Pressure is 1.50 psig.
QUESTION l NRC RECORD # WRI 64 ANSWER: C. SYSTEM # E61 K/A 500000 EK3.03: 3.0/3.5 LP# OP-LO-EP-LP-005-03 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-3 NEW CLASS Figure 5 & Step 66 MODIFIED BANK DIFF 2 D ATE USED: RO SRO ROTH CFR 41.10 i l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 65 The plant is involved in a LOCA.
Which one of the following situations would require an Emergency Depressurization of the reactor? A. Reactor Level is - 140 inches and rising Drywell Temperature is 230 * F Containment Temperature is 190 * F B. Reactor Level is - 120 inches and rising Drywell Temperature is 180 * F Containment Temperature is 110 * F C. Reactor Level is - 30 inches and lowering Drywell Temperature is 240 "F Containment Temperature is 180 * F D. Reactor Level is - 60 inches and lowering DrywellTemperatureis 180 *F Containment Temperature is 170 * F QUESTION l NRC RECORD # WRI 65 ANSWER: A. SYSTEM # K/A 295027 EA1.03: 3.5/3.8 EOP - 3 Pri Ctmt LP# OP-LO-EP-LP-005-03 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-S-01-EP-3 NEW CLASS ' Step 23 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.9/41.10/43.5 ,
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 66 The plant is operating at 75 % power. Plant Services personnel working inside Containment cause an inadvertent initiation of HPCS.
Which one of the following best describes the response of the Reactor Level Control System in Master Auto?
(Assume NO operator action.)
A. The Reactor Level Control System will remain static and lock the feed pumps at the current signal, and reactor water level will increase to the high level scram setpoint B. The Reactor Level Control System will increase feedwater flow in response to the lowering water level, and retum water level to close to the normal level.
C. The Reactor Level Control System will trip to single element, reduce feedwater flow in response to the rising water level, and lower water level to about the Low Level Alarm.
D. The Reactor Level Control System will reduce feedwater flow in response to the rising water level, and return water level close to the normal level.
QUESTION l NRC RECORD # WRI 66 ANSWER: D. SYSTEM # C34; K/A 295008 AK2.03: 3.6/3.7 N21 LP# OP-LOR-ONEP-LP-001-04 OBJ 1, 27 SRO TIER 1 GROUP 2/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-V-6 NEW CLASS ] sect. 2.1 & 4.1 MODIFIED BANK DIFF 3 DATE 1/ SED: RO SRO ROTH CFR 41.5 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 67 The plant is operating at 40 % power.
The "A" Circulating Water Pump develops a phase to phase short which trips the Circ Water Pump. The "B" Circulating Water Pump is tagged out for motor bearing replacement.
' Which one of the following best describes the response of the Main Condenser? ! (Assume NO operator action.)
A. Main Condenser vacuum will decrease and stabilize at approximately 15 inches Hg Vacuum.
' B. Main Condenser vacuum will decrease and stabilize above the turbine trip setpoint, as the Steam Jet Air Ejectors will control Main Condenser Vacuum.
, C. Main Condenser vacuum will decrease and approach 0 inches Hg Vacuum.
D. . Main Condenser vacuum will remain stable at its present value, as the Steam Jet Air Ejectors will control Main Condenser Vacuum.
! l QUESTION l NRC RECORD # WRI 67 j ANSWER: C. SYSTEM # N19; K/A 256000 K6.02: 3.1/3.1 N71 LP# GG-1-LP-RO-N1900.00 OBJ 22b,25 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 , REFERENCE: 05-1-02-V-8 sect. 4.1 & 3.3 NEW CLASS l MODIFIED BANK l DIFF 2
- . DATE USED
- RO SRO BOTH CFR 41.4
3/17/98 e __ _ _ _ _ _ - - - - _ - - - - - _ _ . - - - - _ - - - - - _ _ _ _ - - - - - - - - - - - - - - - - _ - - - - - - - - -
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 68 The plant is operating at 70 % power.
Which of the following best describes the response of the Reactor Water Level Control System on a failure of a single Feed Flow Transmitter UPSCALE 7 l A. The Digital Feed System will recognize the failure and de-select 3 - element l control and return level to the level setpoint.
l B. The Digital Feed System will decrease feed flow until reactor level decreases to 32 inches at which time it will become level dominant remaining in 3 - element control.
L C. The Digital Feed System will decrease feed flow and reactor level will stabilize out at a new low level below the low level alarm setpoint.
D. The Digital Feed System will lock up the controls and hold level at the normal level, remain in 3 - element control, and actuate the DFCS TROUBLE annunciator , on P680.
QUESTION l NRC RECORD # WRI 68 ANSWER: A. SYSTEM # C34 K/A 295009 AA2.02: 3.6/3.7 LP# GG-1-LP-RO-C3401.00 259002 K6.04: 3.1/3.1 OBJ 1.10 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: ARI 04-1-02-H13-P680 NEW CLASS 2A-C9 MODIFIEL BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7 i l l 3/17/98 l
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l U. S. NUCLEAR REGULATORY COMMISSION I WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 69 l Plant conditions are as follows: MODE: Mode 1 Rx power: 28 % T-G Load: 365 MWE Load Demand 390 MWE Bypass position: 0% All other parameters are per plant design.
The operator withdraws a control rod which increases Rx power to 29 %. How will the Turbine EHC Control System respond? A. The Bypass Control Valves will open by whatever amount is required to maintain RX pressure.
B. The Turbine Control Valves will open by whatever amount is required to maintain Rx pressure.
C. The Bypass Control Valves will close by whatever amount is required to maintain RX pressure.
D. The Tmbine Control Valves will close by whatever amount is required to maintain Rx pressure.
QUESTION l NRC RECORD # WRI 69 ANSWER: B. SYSTEM # N32-2 K/A 295007 AK2.01: 3.5/3.7 241000 A2.02: 3.7/3.7 LP# GG-1-LP-RO-N3202.00 K4.01: 3.8/3.8 OBJ 4 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 03-1-01-2 sect. 5.2 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH CFR 41.5 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 70 Plant Service Water is lost to the Auxiliary Building. This resulted in a trip of the operating Drywell Chillers on high condenser pressure.
Which one of the following best describes the affects on the Daywell Atmosphere?
(Assume NO operator action.)
A. Drywell temperature will increase, Drywell pressure will decrease to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the Drywell and Containment.
B. Drywell temperature will increase, Drywell pressure will increase to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the Drywell and Containment.
C. Drywell temperature will increase, Drywell pressure will remain constant due to the communication between the Containment and Drywell atmospheres. j D. Drywell temperature will increase, Drywell pressure will increase such that high Drywell pressure alarms will actuate and a reactor scram will occur due to a Idgh drywell pressure.
. QUESTION l NRC RECORD # WRI 70 ANSWER: D. SYSTEM # M51; K/A 295010 AK2.05: 3.7/3.8 P72 223001 K6.01: 3.6/3.8 LP# GG-1-LP-RO-M5100.00 A4.12: 3.3.5/3.6 OBJ 1,2, 21 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: ARI 04-1-02-H13-P870 NEW CLASS ' ' 3A- D4 MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH CFR 41.4 l
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3/17/98 m
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR l QUESTION 71 The plant is in a reactor startup just after reaching critical.
, The Operator-at-the-Controls is withdrawing SRMs.
l The following conditions exist: l AllIRMs are on Range 2.
! 4 3 SRM A reads 2 x 10 SRM D reads 6 x 10 3 4 SRM B reeds 8 x 10 SRM E reads 8 x 10 3 3 SRM C reads 2 x 10 SRM F reads 3 x 10 l Which one of the following best describes plant conditions? A. Half scram, halfrod block.
B. Full scram, rod block.
C. Rod block only.
D. No trips or blocks are present.
QUESTION l NRC RECORD # WRI 71 ANSWER: C. SYSTEM # C11-2; K/A 215004 A1.04: 3.5/3.5 C51; C71 A3.04: 3.6/3.6 LP# GG-1-LP-RO-C1102.02 OBJ 6 LP# GG-1-LP-RO-C51-1.05 , OBJ 7 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 ) REFERENCE: Tech Specs TR3.3.2.1 NEW CLASS MODIFIED BANK D1FF 2 LOT 3/98 rxinst DATE USED: RO SRO BOTH CFR 41.6 i i l I 3/17/98 J
_ _____ _ ___ _-___-__ _ _ ______-____ __ _ _ _ _ _ _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION , WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR QUESTION 72 l Plant conditions are as follows: i MODE: Mode 1 Rx power: 40 % T-G Load: 520 MWE Load Demand 510 MWE Bypass position: 0% ! All other parameters are per plant design.
l The Operator-at-the Controls continues withdrawing control rods to increase power. The other Control Room Operators are busy with surveillance and staning up BOP systems.
Power is increased 15 % by control rod movements and Load Demand on the Main Turbine hasNOT been adjusted.
Which one of the following best describes the response of the Turbine Control System? i A. The Turbine Control Valves will remain open at present positions and Reactor Pressure will increase to the point that the Low-Low Set SRVs open. Bypass Control Valves will remain closed due to the biasing of the control circuitry.. B. The Turbine Control Valves will open as power increases to control Reactor Pressure, increasing generator output. The Bypass Valves will remain closed.
C. The Reactor Pressure will increase corresponding to the 15 % increase, and the Tmbine Control Valves will remain at their limited load value, Bypass Valves will remain closed.
D. The Reactor Pressure will increase corresponding to the power increase to the point at which pressure overcomes the biasing, at this time the Bypass Valves will open to ; maintain Rx pressure. I
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QUESTION l NRC RECORD # WRI 72 ANSWER: D. SYSTEM # N32-2 K/A 241000 A1.14: 3.4/3.4 LP# GG-1-LP-RO-N3202.00 A1.12: 2.9/2.8 OBJ 1 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 03-1-01-2 sect. 5.2 caution NEW CLASS MODIFIED BANK _
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DIFF 3 DATE USED: RO SRO BOTH CFR 41.5 3/17/98 - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ - _
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 73
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' The plant is in a reactor startup with Reactor Power at 10 %. Which one of the following conditions would allow the withdrawal ofIntermediate Range ,
'
Neutron Instrumentation and the reason for withdrawing the detectors? A. The Mode Switch in Startup. Only those IRM detectors which have associated APRMs reading > 5 % power can be withdrawn. Detectors are removed to minimize detector burnout from high flux fields at high power operation.
~ B. The Mode Switch in Run. All IRM detectors can be withdrawn. Detectors are removed to minimize detector burnout from high flux fields at high power operation. j C. The Mode Switch in Startup. All IRM detectors can be withdrawn as long as IRM to APRM overlap is satisfactorily observed on at least two (2) IRMs. Detectors are removed to prevent localized overheating of the fuel due to the obstructed coolant flow.
D. The Mode Switch in Run. Only those detectors which have had IRM to APRM overlap satisfactorily observed may be withdrawn. Detectors are removed to prevent localized overheating of the fuel due to the obstructed coolant flow.
QUESTION l NRC RECORD # WRI 73 ANSWER: B. SYSTEM # C51 K/A 215003 K5.03: 3.0/3.1 LP# GG-1-LP-RO-C5102.00 K4.05: 2.9/3.0 OBJ 2a, 8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: 03-1-01-1 sect. 6.2.17 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.6 i 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR-
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QUESTION 74 A surveillance test was in progress operating HPCS in CST to CST mode, when an Auxiliary Building Isolation Signal was received.
An operator reports an alarm on the HPCS/RCIC Test Return Diaphragm in area 7, 119 Ft. elevation of the Auxiliary Building.
l Which one of the following best describes the impact of this alarm? i A. Secondary Containment should be INOP due to a failure of a Secondary Containment isolation.
B .' Primary and Secondary Containment are INOP due to a direct siphon path from the CST to the Suppression Pool via HPCS.
C. Primary and Secondary Containment are Operable since the Auxiliary Building Isolation Valves for the HPCS/RCIC CST Test Return are isolated.
D. Secondary Containment is Operable since HPCS is performing its ECCS Function for a LOCA condition.
QUESTION- l NRC RECORD # WRI 74 ANSWER: A. SYSTEM # E22; K/A 290001 A4.10: 3.4/3.3 T10 LP# GG-1-LP-RO-E2201.00 i OBJ 17 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 l REFERENCE: 06-OP-1T10-M-0001 NEW CLASS sect.1.1 & 5.3.3 MODIFIED BANK DIFF 2 Tech Specs 3.6.4.2 ! DATE USED: RO SRO BOTH CFR 41.7/43.4 ,
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7_ . U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 i REACTOR OPERATOR QUESTION 75 A Non-Licensed Operator is being sent out on ajob in a High Radiation Area.
The Dose rate in the area of thejob is 120 mrem /hr. Thejob is expected to take 45 minutes. The operator's exposure history to date for the year is 1800 mrem.
- Can the operator be utilized for this job and WHY?
l A. Yes, the operator will not exceed his administrative limits.
B. Yes, the operator will have to have an approved extension on dose limits before
- thejob.
C. No, the operator will exceed his federal dose limits.
D. No, the operator will exceed administrative dose limits which are not allowed to be extended.
QUESTION l NRC RECORD # WRI 75 ANSWER: A. SYSTEM # K/A G2.3.4: 2.5/3.1 ADMIN Rad Con Generic LP# EOI-S-LP-GET-RWT01.05 OBJ RWT31,32,33 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-08-2 sect. 6.3.2 NEF CLASS MODIFIED BANK DIFF 2 _ DATE USED: RO SRO ROTH CFR 41.12 I . 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR l
{
QUESTION 76 While in Mode 1, the "A" CRD Pump tripped.
Following an improper start of the "B" CRD Pump, the Operator-at-the-Controls noticed four (4) control rods drifting in with no drive command.
What is the proper action to take? A. Immediately scram the reactor.
B. Take no action until the control rod motion has stopped or reached the full-in position, then immediately return the control rods to their required position.
C. Select the control rod closest to the center of the core and apply a continuous insert signal to it until it reaches position 00.
D. Reduce core flow to 60%.
~l QUESTION RO 76 l NRC RECORD # WRI 101 ANSWER: A. SYSTEM # C11 K/A G2.4.1: 4.3 Generic LP# GG-1-LP-RO-C111A.00 OBJ, 19 LP# OP-LO-ONEP-LP-001-04 OBJ. I SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: 05-1-02-IV-1 sect. 2.2.3 NEW CLASS MODIFIED BANK DIFF 3 annual exam wk5 1 RO 4 DATE USED: RO SRO BOTH CFR 41.10/43.5
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 77 The plant was operating at 60% power when Division I UPS Power Panel 1Y89 trips its incoming circuit breaker.
Which one of the following is the response of the Reactor Protection System? A. RPS A system tripped resulting in a halfscram due to a loss ofpower to the RPS A logic relays.
B. RPS A system logic energized causing a half scram due to a loss of power to the RPS A Scram Pilot Valve solenoids. t
C. RPS A system tripped with no half scram because the RPS A solenoids still have power available.
l D. RPS A system logic energized causing alarms indicating the loss of power with no half scram due to RPS Bus A still being energized QUESTION RO 77 l NRC RECORD # WRI 102 ANSWER: A. SYSTEM # C71; K/A 2212000 K6.01: 3.6 1 L62 Kl.04: 3.4 )l LP# GG-1-LP-RO-C7100.00 OBJ. 13,18 LP# GG-1-LP-RO-L6200.00 OBJ. 5,13 SRO TIER GROUP / RO TIER 2 GROUP 1 REFERENCE: E-1026 NEW CLASS E- 1173 - 14,15, & 19 MODIFIED BANK j DATE USED: RO SRO BOTH CFR 41.2/41.6 i I I' 3/17/98 l
i U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR I QUESTION 78 l The plant was operating at 60% power when the "A" Recirculation Pump tripped. ONEP l actions have been taken and the Recirculation Pump Discharge valve (B33-FM7A) has been I reopened.
Select the statement that is correct concerning core flow indications: j
l A. When only one recirculation loop is in operation, total core flow is determined by
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i summing the flow from the flow elbow on the suction of the recirculation pump that is ! in operation and the flow elbow on the suction of the recirculation pump that has tripped.
B. When only one recirculation loop is in operation, total core flow is determined by , summing thejet pump flows for each loop, then subtracting the total flow from the I non-operating loop from the total flow of the operating loop.
C. Total core flow is normally determined by circuitry that takes the square root of the differential pressure across the core plate and displaying it on the P680 panel.
D. Total core flow is calculated by determining the " driving" flow through thejet pumps and adding the " driven" flow through thejet pumps.
QUESTION RO 78 l NRC RECORD # WRI 103 ANSWER: B. SYSTEM # B33 K/A 202002 A4.09: 3.2; A1.06: 3.4 LP# GG-1-LP-RO-B3300.00 A1.07: 3.1; A4.08: 3.3 OBJ. 35 LP# CG-1-LP-RO-B2104.00 OBJ. 5 SRO TIER GROUP / RO TIER 2 GROUP 1 REFERENCE: E- 1164 - 01 & 02 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.2/41.6 3/17/98 __
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 79 The plant is in a reactor startup at the point of adding heat when the control rod selected by the Operator-at-the-Controls stops control rod motion at position 12. The next control rod selected has an Insert Block indicated.
Which one of the following is the basis for the Insert Block? A. This is to prevent exceeding the Linear Heat Generation Rate (LHGR) Thermal Limit, thus causing plastic deformation of the fuel.
B. This is to mitigate the consequences of a Control Rod Drop Accident by limiting the maximum heat added to 280 cal /gm.
C. This is to prevent exceeding the Minimum Critical Power Ratio (MCPR) Thermal Limit, thus preventing the onset of transition boiling.
, D. This is to limit the amount of reactivity the operator can add in a single rod motion providing an observable response.
QUESTION RO 79 l NRC RECORD # WRI 104 ANSWER: B. SYSTEM # C11-2 K/A 201005 K4.02: 3.3 K5.04: 2.7 LP# GG-1-LP-RO-C1102.02 OBJ'. 2, 8 SRO TIER GROUP / RO TIER 2 GROUP 1 i REFERENCE: Tech Spec Bases NEW CLASS B3.1.6 & B3.3.2.1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.6/43.6
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 80 The plant is operating in the normal electrical lineup.
CCW pumps "A" and "C" are operating with "B" selected for STANDBY.
A Loss of Coolant Accident occurs resulting in a shedding ofloads. j The "C" CCW pump trips on overcurrent. l i i l
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Which of the following describes the resulting status of the CCW system? )
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(Assume NO operator action.)
A. Pump "A" operating; Pump "B" operating; Pump "C" tripped B. Pump "A" operating; Pump "B" not operating; Pump "C" tripped C. Pump "A" not operating; Pump "B" operating; Pump "C" tripped D. Pump "A" not operating; Pump "B" not operating; Pump "C" tripped MESTION RO 80 l NRC RECORD # WRI 105 ANSWER: B. SYSTEM # P42 K/A 400000 K4.01: 3.4 LP# GG-1-LP-RO-P4200.00 OBJ 6 SRO TIER GROUP / 'RO TIER 2 GROUP 2 REFERENCE: 04-1-01-R21-1 Table 1 NEW CLASS MODIFIED BANK DIFF 2 LOT 2/98 rxsys DATE USED: RO SRO BOTH CFR 41.4 i l 3/17/98 l
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 81 The reactor is operating at 55 % power when a pressure regulator failure causes the bypass valves to fully open.
RPV pressure decreases to 700 psig and the MSIVs fail to close.
Reactor power is 42 % when the operator manually closes the MSIVs.
The reactor scrams and RPV level is restored to normal with RCIC.
This is a violation ofwhich Safety Limit per GGNS Technical Specifications? A. Reactor at power above the Low Power Setpoint with Main Steam Bypass Valves open.
B. Reactor pressure and core flow at High Power being within the MCPR Limits for two loop operation.
C. Thermal power out of specification with Reactor pressure . D. Containment Isolation System failing to isolate at required setpoint.
QUESTION RO 81 l NRC RECORD # WRI 107 ANSWER: C. SYSTEM # K/A Generic G2.1.33: 3.4 Tech Specs LP# OP-LO-PB-LP-001-02 OBJ 8b & d LP# OP-LO-PB-LP-003-00 OBJ 4 SRO TIER GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs 2.1.1 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 43.2 3/17/98 _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 82 RFO9 is in progress.
The plant is aligning the Alternate Decay Her.t Removal System for operation.
The common suction for RHR A and B for Shutdown Cooling is being tested for LLRTs.
What would be the required alignment for ADHR to be utilized as the decay heat removal system? A. Vessel head removed with ADHR suction from the RHR C Suppression Pool Suction E12-F004C, returning to the reactor via the RHR C E12-F042C Injection Valve, through the reactor, and out the Safety Relief Valves to the Suppression Pool.
B, Vessel head removed with ADHR suction from the Spent Fuel Pool, returning to the reactor vessel via the RHR C .2-F042C Injection Valve with the gates between theUpperContaine ?ool and the Spent Fuel Pool removed.
C. Vessel head removed with ADHR suction from the Spent Fuel Pool, returning to the reactor vessel via RHR A or B Shutdown Cooling isolation Valve E12-F053A , or B with the gates between the Upper Containment Pool and the Spent Fuel Pool removed.
D. Vessel head removed with ADHR suction from the RHR A or B vi:: E12-F037 A or B from the Spent Fuel Pool, returning to the reactor vessel via the RHR C E12-F042C Injection Valve with the gates between the Upper Containment Pool and the Spent Fuel Pool removed.
QUESTION RO 82 l NRC RECORD # WRI 109 ANSWER: B. SYSTEM # E12-1 K/A 233000 K1.01: 2.6 LP# OP-LO-SYS-LP-E12-1- Kl.02: 2.9 OBJ 4 SRO TIER GROUP / RO TIER 2 GROUP 3 REFERENCE: 04-1-01-E12-1 sect. 5.13 NEW CLASS M-1085 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 !
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l t REACTOR OPERATOR QUESTION 83 Radwaste is discharging the Floor Drain Sample Tank to the River.
Which one of the following would result in an isolation of the G17-F355 Liquid Radwaste Discharge Isolation Valve? A. The Floor Drain Sample Pump discharge pressure is too low. j
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B. The blow down flow rate is too low.
C. The discharge flow rate is too low. l D. The effluent radiation monitor high radiation setpoint is reached.
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1 QUESTION RO 83 l NRC RECORD # WRI 110 ANSWER: B. SYSTEM # G17 K/A 268000 A1.02: 2.6 LP# GG-1-LP-RO-G17/18.00 OBJ 6a,11h,J SRO TIER GROUP / RO TIER 2 GROUP 3 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS l 19A-H7; 19A-H8 MODIFIED BANK DIFF 2 ARI 04-1-02-H13-P870 bank question DATE USED: 6A-F3 RO SRO BOTH CFR 41.13/43.4 I ( 3/17/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 84 Conceming the Fast Opening of One Recirculation Flow Control Valve transient, which one of the following CONDITIONS would result in the more severe transient on the . reactor? A Reactor Power is at 36 % with Recire in Fast Speed with Minimum Valve position.
B. Reactor Power is at 30 % with Recirc in Slow Speed with Maximum Valve position C. Reactor Power is at 60 % with Recirc in Fast Speed with Minimum Valve position.
l D. Reactor Power is at 100 % with Recirc in Fast Speed with 68% Valve position.
QUESTION RO 84 l NRC RECORD # WRI 116 ANSWER: C. SYSTEM # B33; K/A 295014 AA1.02: 3.6 FSAR CHPT 15 LP# OP-LO-DT-LP-011-02 202002 K1.02: 4.2; K3.02: 4.0 OBJ 2 SRO TIER GROUP / RO TIER 1 GROUP 1 REFERENCE: UFSAR IE4.5 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.6/43.6 I 3/17/98 _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR
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QUESTION 85 The Division III Diesel Generator Surveillance is being completed at this time. The normal Control Room STOP pushbutton was pressed five (5) minutes ago. and the diesel is cooling down. A LOCA signal and a loss of power to bus 17AC hasjust occurred.
Which of the following statements is correct concerning the Division III Diesel Generator? A. The condition of the diesel will not change until the cooldown cycle has completed at which time the diesel will restart and the output breaker will close.
B. The condition of the diesel will not change until the cooldown cycle has completed at which time the diesel must be manually reset by going into and out of MAINTENANCE.
C. The diesel will speed up to 900 rpm and the output breaker will close as a result of the LOP /BUV signal.
D. The diesel will speed up to 900 rpm, but the output breaker will not close until the local VOLTAGE SHUTDOWN RESET pushbutton is depressed.
l l QUESTION RO 85 l NRC RECORD # WRI 117 ANSWER: D. SYSTEM # P81 K/A 264000 A3.01: 3.0 LP# GG-1 LP-RO-P8100.01 OBJ. 15, 19 SRO TIER GROUP / RO TIER 2 GROUP 1 REFERENCE: 04-1-01-P81-1 sect. 3.21 NEW CLASS MODIFIED RANK DIFF 3 annual exam , P81-02 l DATE USED: RO SRO BOTH CFR 41.8 . , ! ! 3/17/98 ______ _ _ _ - _ _ _ -
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 86 The plant was operating at 20 % power when a Feedwater rupture in the Turbine Building cause Reactor water level to decrease. The Control Room Operator manually initiated HPCS and RCIC. Levelin the Reactor droppcd to - 32 inches before HPCS and RCIC turned level, and level is now increasing.
Which one of the following best describes the status of the Recirculation System? A. Recire Pumps are in Slow Speed with the Flow Control Valves Locked up (motion inhibit).
B. Recirc Pumps are tripped with the Flow Control Valves Locked up (motion inhibit).
C. Recirc Pumps are in Slow Speed with the Flow Control Valves in their pre-transient positions.
D. Recire Pumps are tripped with the Flow Control Valves in their pre-transient positions.
QUESTION RO 86 l NRC RECORD # WRI 121 ANSWER: C. SYSTEM # B33 K/A 295009 AKl.02: 3.0 AK2.03: 3.1 LP# GG-1-LP-RO-B3300.00 AA1.03: 3.0 OBJ. 22,$1 SRO TIER GROUP / RO TIER 1 GROUP 1 , REFERENCE: ARI 04-1-02-H13-P680 NEW CLASS i 3A-DJ; 3A-D10 MODIFIED BANK I DIFF 3 i DATE USED: RO SRO BOTH CFR 41.5/43.5 l I
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_ _ _ _ _ _ _ _ - _ - _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _____ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 87 Which one of the following conditions will allow a start of Recirculation Pump "A"? A. Steam Dome Pressure 950 psig Recirc Loop A Temperature 485 *F Recirc Loop B Temperature 529 'F Bottom Head Temperature 500 * F Reactor Power 30 % B. Steam Dome Pressure 1014 psig Recirc Loop A Temperature 495 *F Recirc Loop B Temperature 529 *F Bottom Head Temperature 435 *F Reactor Power 90 % C. Steam Dome Pressure 981 psig Recirc Loop A Temperature 499 *F Recire Loop B Temperature 529 *F Bottom Head Temperature 500 *F Reactor Power 60 % D. Steam Dome Pressure 960 psig Recire Loop A Temperature 481 *F l Recire Loop B Temperature 525 *F Bottom Head Temperature 495 *F Reactor Power 60 % i l I QUESTION RO 87 l NRC RECORD # WRI 122 ANSWER: C. SYSTEM # B33; K/A 216000 K1.23: 3.3 ! B21 202001 A4.01: 3.7 LP# GG-1-LP RO-B3300.00 OBJ. 28 SRO TIER GROUP / RO TIER 2 GROUP 1 l REFERENCE: 06-OP-1B33-V-0005 NEW CLASS l Data Sheet IV sect. 5.4 MODIFIED BANK l DIFF 3 I DATE USED: RO SRO BOTH CFR 41.5
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR i QUESTION 88 l l Which one of the following is utilized to reduce the possibility of a hydrogen buildup or incomplete recombination of the Ofigas system during startup or other low flow l conditions? , A. Nitrogen purge of the Ofigas System.
B. Oxygen addition to the Offgas Recombiner.
C. Venting of the Offgas System prior to the Ofigas Recombiner.
D. Instrument Air purge of the Ofigas System.
r QUESTION RO 88 l NRC RECORD # WRI 123 ANSWER: D. SYSTEM # N64 K/A 271000 K4.04: 3.3 LP# GG-1-LP-RO-N6465.00 OBJ. 14b,15 SRO TIER GROUP / RO TIER 2 GROUP 2 l REFERENCE: 04-1-01-N64-1 NEW CLASS l ' sect. 3.2, 3.7, 3.9, 5.5 MODIFIED BANK DIFF 2 DATE USED: RQ SRO BOTH CFR 41.4/41.13/43.4
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i U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 89 i Which one of the following describes the operation of the Control Room Standby Fresh Air units with control room intake air radiation levels indicating 6 mrem /hr? A. Control room A/C units trip and only the standby fresh air units are allowed to run.
B. The standby fresh air units start in the Fresh Air mode and are interlocked in this mode for 10 minutes.
C. The control room purge isolation dampers auto open to provide a positive pressure for l the control room. I D. The standby fresh air units start in the Recirc mode and are interlocked in this mode for 10 minutes.
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QUESTION RO 89 l NRC RECORD # WRI 124 ANSWER: D. SYSTEM # Z51 K/A 290003 K5.01: 3.2 LP# GG-1-LP-RO-Z5100.00 ! OBJ. 9 SRO TIER GROUP / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-Z51-1 sect. 5.4 NEW CLASS I ARI 04-1-02-H13-P855 MODIFIED RANK DIFF 2 1A-A5 LOT 7/95 vent DATE USED: ARI 04-1-02-H13-P601 RQ SRO BOTH CFR 41.4 19A-A10; 19A-A11 l l l l ,
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR QUESTION 90 The plant is operating at 100 % power.
The solenoid operated pilot valve for PS3-F001, Instrument Air Supply Header to Containment, burns out causing an isolation of the Instrument Air Header to Containment. ) i Which one of the following best describes the reaction of the Control Rod Drive System to this ' event? (Assume NO Operator Actions.)
A. The Scram Inlet and Outlet Valves will open, inserting Control Rods in a random ! pattern. The Scram Discharge Volume Vent and Drain Valves will fail closed, and the l CRD Flow Control Valve will drift open.
B. The Scram Inlet and Outlet Valves will open, inserting Control Rods in a random l pattern. The Scram Discharge Volume Vent and Drain Valves will fail closed, and the
- CRD Flow Control Valve will drift closed to minimum setting.
l C. The Scram Pilot Valve will vent on a Low Air Pressure signal causing the Scram Inlet and Outlet Valves to open, inserting Control Rods in a random pattern. The Scram i Discharge Volume Vent and Drain Valves will fail open, and the CRD Flow Control l Valve will drift closed to minimum setting.
D. The Scram Pilot Valve will vent on a Low Air Pressure signal causing the Scram Inlet - and Outlet Valves to open, inserting Control Rods The Scram Discharge Volume Vent and Drain Valves will fail open, and the CRD Flow Control Valve will fail open.
QUESTION RO 90 l NRC RECORD # WRI 125 j ANSWER: B. SYSTEM # P53; K/A 300000 K3.02: 3.3 C11-1A LP# OP-LOR-ONEP-LP-001-04 OBJ. 29 LP# GG-1-LP-RO-C111A.00 OBJ. 15b,16 SRO TIER GROUP / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-C11-1 sect. 3.1 NEW CLASS 05-1-02-V-9 sect. 5.1 MODIFIED BANK DIFF 3 l DATE USED: RO SRO BOTH CFR 41.6 3/17/98
l U. S. NUCLEAR REGULATORY COMMISSION I l WRITTEN EXAMINATION MARCH 1998 l REACTOR OPERATOR QUESTION 91 Which one of the following is an example of a method used to reduce personnel radiation exposure? '
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A. Assigning two (2) mechanics and a supervisor to replace the packing on a valve located l in a RWCU Pump Room"A".
I B. Sending an operator to verify a tagout in the Radwaste Building Tunnel on a RWCU i ' Backwash transferline vent valve.
l C. Stationing an operator on the Refuel Bridge over the Reactor Cavity to monitor l Hydrolazing in the Reactor Cavity area. 1 D. Notifying the Health Physics Department of a RCIC Surveillance that will operate in l full flow test to the Suppression Pool.
QUESTION RO 91 l NRC RECORD # WRI 126 ANSWER: D. SYSTEM # K/A Generic G2.3.1: 2.5 i ADMIN Rad Con LP# GG-1-LP RO-PROC.00 : OBJ. 29C & D SRO TIER GROUP / RO TIER 3 GROUP l REFERENCE: 01-S-08-2 sect. 6.1 NEW CLASS I MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.12 i i I i
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 92 l The Plant Supervisor has a temporary alteration which is to be installed on the relays that i initiate a Containment isolation for Drywell Chilled Water Valves for a LOCA signal. This temporary alteration will prevent the valves from isolating.
Which one of the following personnel may independently verify the installation of this
- temporary alteration?
i A. PM & C Engineer , B. a Nuclear Operator"B" C. Control Room Operator D. an I & C Technician l l QUESTION RO 92 l NRC RECORD # WRI 128 l ANSWER: C. SYSTEM # K/A Generic G2.211: 2.5 ADMIN G2.1.1: 3.7 Equip Control LP# GG-1-LP-RO-PROC.00 l OBJ. 13C & E SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-3 sect. 5.5 NEW CLASS 01-S-06-29 sect. 6.1.5 MODIFIED BANK DIFF 2 I DATE USED: RO SRO BOTH CFR 41.10 l l
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- _ _ _ - - _ _ _ _ - _ - _ _ _ _ _ _ __ _ - __ _ _______ __ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 REACTOR OPERATOR QUESTION 93 In accordance with Conduct of Operations, which of the following statements best desenhs the reason that personnel performing core alterations shall be in constant communications with the Operator-at-the-Controls? A. A core alteration is considered a change in reactivity which requires the knowledge and consent of the Operator-at-the-Controls using sound powered phones.
B. ' Core alterations are considered a special evolution requiring constant communication with the Control Room using portable radios. , l I
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C. To allow the Operator-at-the-Controls to monitor for inadvertent criticality and inform the refuel floor ofsuch event using sound powered phones.
l D. To allow the on-duty STA to perform a shutdown margin check required during Core Alterations using portable radios.
QUESTION RO 93 l NRC RECORD # WRI 132 ANSWER: C. SYSTEM # K/A Generic G2.1.16: 2.9 ADMIN G2.2.26: 2.5 Conduct of Ops G2.2.32: 3.5 LP# GG-1-LP-RO-PROC.00 OBJ. 12AA SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-2 act. 6.7.16 NEW CLASS 01-S-06-14 sect. 6.1.6 MODIFIED BANK DIFF 2 03-1-01-5 sect. 2.6 LOT 7/95 admin DATE USED: RO SRO BOTH CFR 41.10/43.7 3/17/98
i U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 94 The plant is operating at 100% power. You are the Operator-at-the-Controls. Annunciators have been received and indications indicate that reactor water level has risen to 54 inches on all ! three Narrow Range Level'mstmments.
] I The reactor did NOT automatically Scram.
NO other operators are present in the Control Room.
Which of the following best desentes your actions? , i A. Take control of reactor water level and decrease the level, then reset the scram annunciators. After level is normal inform the Plant Supervisor so that the appropriate LCOs can be completed.
B. Insert a manual reactor scram, then trip the main turbine, ifit did not trip. Take manual control of feedwater and retum level to nonnal, and carry out ONEP actions for a reactor scram.
C. Take control of reactor water level and decrease the level, then reset the scram annunciators. Once level is returned to normal, initiate a controlled plant shutdown per IOI- 2. j D. Manually trip the Main Turbine. Take manual control of feedwater, and return level to normal, and cany out ONEP actions for a reactor scram.
QUESTION RO 94 l NRC RECORD # WRI 133 ANSWER: B. SYSTEM # K/A Generic G2.1.1: 3.7 ADMIN Conduct of Ops LP# GG-1-LP-RO-PROC.00 OBJ. 12AA SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-2 sect. 6.2.1 NEW CLASS MODIFIED BANK DIFF 2 LOT 7/95 admin DATE USED: RO SRO BOTH CFR 41.10 3/17/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR
- QUESTION 95 The plant is at 135'F.
AllECCS systems arein standby.
The Reactor Mode Switch is in SHUTDOWN.
The Reactor Head is installed with all head closure bolts fully tensioned.. Primary and Secondary Containment is in effect.
With the above conditions, which one of the following is the Plant Operational Mode? A. Mode 2 - Startup B. Mode 3 -Hot Shutdown C. Mode 4 -Cold Shutdown D. Mode 5 -Refueling QUESTION RO 95 l NRC RECORD # WRI 142 ANSWER: C. SYSTEM # K/A Generic G2.1.22: 2.8 ADMIN Conduct of Ops.
LP# OP-LO-PB-LP-003-00 OBJ. 1 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs sect.1.1 NEW CLASS table 1.1-1 MODIFIED BANK DIFF 2 ' DATE USED: RO SRO BOTH l l 3/17/98
I U. S. NUCLEAR REGULNf0RY COMMISSION WRITTEN EXAMINATION MARCH 1998 . REACTOR OPERATOR QUESTION 96 A Control Room annunciator on the H13-P680 panel has a red "X" taped across the face of the annunciator window.
! l Which one of the following best describes the signi6cance of this indication? ! A. The annunciator window is operable with problem inputs bypassed B. The annunciator is a problem annunciator with its alarm card pulled.
C. The annunciator is a problem annunciator that is still functional out under investigation by I & C.. D. The annunciator is an EP-4 entry condition annunciator.
QUESTION RO 96 l NRC RECORD # WRI 145 ANSWER: B. SYSTEM # K/A Generic G2.4.31: 3.3 ADMIN Emergency Proc./ Plan I LP# GG-1-QC-RO-CR001.00 ) OBJ. Admin I LP# OP-LO-PD-LP-003-00 OBJ. 45B & E SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 02-S-01-25 sect. 6.2 & 6.3 NEW CLASS Att. IV MODIFIED BANK DIFF 2 02-S-01-4 sect. 6.4 DATE USED: RO SRO BOTH i l l 3/17/98
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_ _ _ - _ _ ________-__ - _____- - __-_-_ . _ _ . _ _ _ _ __ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 97 Given the following plant conditions: Reactor Fower 0% Reactor Level + 33 inches ReactorPressure 890 psig Containment Temperature 98 *F Containment Pressure 1.3 psig Suppression PoolTemperature 97 * F DrywellPressure 4.5 psig Drywell Temperature 208 * F Which one of the following describes the Emergency Procedures that are to be implemented? A. EP - 3 only B. EP - 2, and 4 C. EP - 2 and 3 D. EP - 2, 3, and 4 QUESTION RO 97 l NRC RECORD # WRI 146 ANSWER: C. SYSTEM # K/A Generic G2.4.2: 3.9 ADMIN Emergency Proc./ Plan LP# GG-1-LP-RO-EP07.00 OEJ. 1, 2, 3 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 05-1-01-EP-2 NEW CLASS 05-1-01-EP-3 MODIFIED BANK
' DIFF 2 05-1-01-EP-4 DATE USED: RO SRO BOTH CFR 41.10/43.5 ! l 3/17/98 i
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 REACTOR OPERATOR QUESTION 98 The plantisin an ATWS.
Which one of the following best describes the basis for injection of Standby Liquid Control (SLC) at 110 * F in the Suppression Pool? A. If Suppression Pool Temperature is allowed to rise further, SLC will have insufficient time to shutdown the reactor before Suppression Pool temperature reaches saturation temperature for Containment pressure at 1.23 psig. ;
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B. If Suppression Pool Temperature is allowed to increase higher, there is no assurance > that the Suppression Pool Heat Capacity will be sufficient to receive the energy from a future emergency depressurization and prevent jeopardizing Containment.
. C. If Suppression Pool Temperature is allowed to rise further, insufficient time is available to utdize attemate =thmis of SLC injection and maintain Suppression Pool ! Temperatures within the Heat Capacity Temperature Limit Curve.
D. This is the maximum Suppression Pool Temperature that water that may be used for injection into the reactor vessel will not dilute the effects of SLC with positive l reactivity thus resulting in further power increasesjeopardizing Containment i l i QUESTION RO 98 l NRC RECORD # WRI 147 ANSWER: B. SYSTEM # K/A Generic G2.2.25: 2.5 i ADMIN G2.1.27: 2.8 ! Equipment Cont.
LP# GC-1-LP-RO-EP2A.00 OBJ. 2 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: EP Bases NEW CLASS MODIFIED BANK DIFF 3 i DATE USED: RO SRO BOTH CFR 41.6/43.6 l
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[- U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 99 The Plant Supervisor- has ordered you to perform a RCIC operability test following l maintenance using the RCIC Quarterly Pump Operability Verification.
Which one of the following identifies the additional surveillance to be performed during the performance of this surveillance? A. Perform RCIC Monthly Valve Operability on the RCIC Exhaust Rupture Diaphragm.
B. Perform Chemistry analysis on the Suppression Pool water.
C. Perform Remote Shutdown Panel Post Accident Instrumentation Operability Checks.
D. Perform Suppression Pool Temperature Monitoring Checks.
QUESTION RO 99 l NRC RECORD # WRI 148 ANSWER: D. SYSTEM # K/A Generic G2.2.12: 3.0 ADMIN Equipment Cont.
LP# Simulator training OBJ. SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 04-1-01-E51-1 NEW CLASS sect. 5.2.le MODIFIED BANK 06-OP-1E51-Q-0003 f.ect. 2.3 DIFF 3 Tech Specs 3.6.2.1
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DATE USED: SR 3.6.2.1.1 RO SRO BOTH CFR 41.10/43.2
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U. S. NUCLEAR REGULATOR.Y COMMISSION WRITTEN EXAMINATION MARCH 1998 REACTOR OPERATOR QUESTION 100 In which one of the following situations would the MOV Test Switch be required to be in the TEST position? A. An operator is preparing to lineup the RWCU System to transfer from Pre-Pump mode to Post-Pump Mode ofoperation.
B. Aligning the RHR A system for LPCI operation with Reactor level at 0 inches.
C. Aligning SSW B to the Air Compressors during a loss of Turbine Building Coou g Water.
D. The HPCS system is being realigned to standby following an inadvenent initiation, and the E22-F004 HPCS Injection valve is stroking.
QUESTION RO 100 l NRC RECORD # WRI 149 3 ANSWER: A. SYSTEM # K/A Generic G2.1.28: 3.2 ) ADMIN Conduct of Ops.
LP# GG-1-LP-RO-G3336.00 OBJ. 20 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 04-1-01-G33-1 NEW CLASS sect. 3.18 MODIFIED BANK DIFF 2 01-S-06-2 sect. 6.9.3 DATE USED: RO SRO BOTH CFR 41.10
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ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet U.S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information Name: Region: I / II / III /(0/) Date: 27 March 1998 _ Facility / Unit: Grand Gulf 1 License Level: Ro / (SRdh Reactor Type: W / CE / BW / s- / GE Start Time: Finish Time: Instructions Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent.
Examination papers will be collected four hours after the examination starts.
Applicant certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent - _____ _
, U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 1 The following conditions are observed after a Loss of Coolant Accident:
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Reactor Pressure 90 psig 166' elev. temperature in the Drywell 280*F Drywell Pressure 5.8 psig 139' elev. temperature in the Containment 192 * F j 119' elev. temperature in the Containment 181 * F l ' Containment Pressure 2.0 psig Shutdown Range LevelIndication + 20 mehes ! Upset Range LevelIndication + 50 inches Wide Range LevelIndication - 40 inches Which one of the following indicates actual level? i A. Shutdown Range B. Upset Range C. Wide Range D. Level Cannot be determined.
QUESTION l NRC RECORD # WRI 1 ANSWER: C. SYSTEM # B21 K/A 295028 EK2.03: 3.6/3.8 EK1.01: 3.5/3.7 l LP# GG-1-LP-RO-EP02.00 K/A 295027 EK1.02: 3.0/3.2 OBJ. 18 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 ;
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REFERENCE: 05-1-01-EP-2 Caution 1 NEW CLASS Modified Bank DIFF 3 annual exam ep-03 DATE USED: RO SRO BOTH CFR 41.3/43.5 i l l I 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIORREACTOR OPERATOR QUESTION 2 The plant conditions are as follows: Reactor Pressure 900 psig Reactor Water Level - 100 inches DrywellPressure 1.10 psig LPCS Injection Line Pressure 450 psig Which of the following describes how the LPCS injection valve E21-F005 would respond ifits handswitch is taken to the OPEN position? A. The valve will not open.
B. The valve will not open until reactor level or drywell pressure has reached the LPCS System initiation setpoint.
C. The valve will open snd remain open.
D. The valve will open, but will automatically close when it reaches its full open ; position.
l ' QUESTION l NRC RECORD # WRI 2 ANSWER: C SYSTEM # E21 K/A 209001 K4.08: 3.8/4.0 , LP# GG-1-LP-RO-E2100.00 K3.-: 3,5-3.9/3.5-3.9 , OBJ. 8b,16 SRO TIER 2 GROUP 1 / RO TIER 2 GROUP 1 [ REFERENCE: 04-1-01-E21-1 sect. 3.11 NEW CLASS sect. 3.12 MODIFIED BANK j DIFF 3 annual exam I e21-05 DATE USED: RO SRO BOTH CFR 41.7/41.8
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 3 The Plant is operating at 100 % power.
The Motor Driven Fire pump is out of service.
A fire in Transfonner ESF 12 has initiated the Deluge system for the transformer.
The A Diesel Driven Fire Pump received a signal to start.
Which one of the following describes the starting limitations of the Diesel Driven Fire Pump? A. The diesel engine will attempt to start for 15 minutes. If the diesel does not start it alarms in the Control Room, and it must be reset from the Control Room before it will attempt to stan again.
B. The diesel engine will attempt to stan for 15 seconds, then wait for 15 seconds. It will attempt this start sequence for 6 attempts. After that, it must manually be reset before any further start attempts occur.
C. The diesel engine will attempt to start for 15 seconds then wait for 15 minutes to allow the battery to recharge, then it will attempt this cycle again. After that it must manually be reset before any further start attempts occur.
D. The diesel engine will attempt to start as long as air pressure is > 60 psig. After that, the air bank must recharge before additional start attempts can occur.
QUESTION l NRC RECORD # WRI 3 ANSWER: B. SYSTEM # P64 K/A 286000 A2.08: 3.2/3.3 K5.05: 3.0/3.1 j K4.07: 3.3/3.3 A3.01: 3.4/3.4 LP# OP-LO-SYS-LP-P64 05 A4.06: 3.4/3.4 OBJ. 5d SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 ! l REFERENCE: ARI 04-S-02-SH13-P862 NEW CLASS 1A-B3; 1A-B5 MODIFIED BANK l DIFF 3 sect.1.2; 2.1; & 4.5 DATE USED: RO SRO BOTH CFR 41.4 l 3/19/98 _.
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR
. QUESTION 4 . High Drywell pressure of 2.2 psig has resulted in ECCS initiation. HPCS has injected to the reactor at rated pressure causing reactor level to increase to + 60 inches. The HPCS Injection Valve (E22-F004) has autome' "y closed. No operator action was taken.-
Reactor water level has subuquently decreased to - 50 inches.
Which one of the following actions will be required to have HPCS inject to the reactor to restore waterlevel? A. Place the E22-F004 handswitch to CLOSE to reset the logic, and then to OPEN.
B. No further action is required. E22-F004 will auto reopen at - 41.6 inches.
C. HPCS Manual Initiation pushbutton must be Armed and Depressed to reset E22-F004 and allow it to open.
D. Reset the HPCS Initiation Logic and allow a subsequent Initiation signal to occur.
QUESTION l NRC RECORD # WRI 4 ANSWER: B. SYSTEM # E22-1 K/A 209002 A3.01: 3.3/3.3 LP# GG-1-LP-RO-E2201.co A2.10: 2.7/3.0 OBJ. 8b,16 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-E22-1 sect. 3.8 NEW CLASS MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.7 3/19/98 _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 5 The plant is in Mode 4 with Shutdown Cooling RHR A in service. RHR A was also aligned to blowdown reactor inventory to the Suppression Pool via the RHR A heat exchanger vent valves (E12-F073 A and E12-F074A) due to an outage on RWCU. After opening the heat exchanger vent valves (E12-F073 A and E12-F074A) a significant amount, the operator assigned to monitor vessel level becomes distracted.
Which one of the following best describes the response of the RHR A System to a lowering waterlevel?
(Assume no further operator actions)
A. At - 41.6 inches, the RHR A Suction from the Reactor (E12-F008 & F009), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to trip.
B. At + 11.4 inches, RHR A Heat Exchanger Vents (E12-F073A & F074A), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to go on minimum flow.
C. At - 41.6 inches, RHR A Heat Exchanger Vents (E12-F073A & F074A), and RHR A Shutdown Cooling Return to Feedwater (E12-F053 A) will isolate. This will cause the RHR A pump to go on minimum flow.
D. At + 11.4 inches, the RHR Suction from the Reactor (E12-F008 & F009), and RHR A Shutdown Cooling Return to Feedwater (E12-F053A) will isolate. This will cause the RHR A pump to trip.
QUESTION l NRC RECORD # WRI 5 ANSWER: D. SYSTEM # E12 K/A 205000 K4.03: 3.8/3.8 LP# OP-LOR-ONEP-LP-001-04 A2.05: 3.5/3.7 OBJ. 31 A2.09: 3.6/3.8 LP# OP-LO-SYS-LP-E12-07 OBJ. 11,12 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 05-1-02-III-5 Group 3 NEW CLASS 04-1-01-E12-1 MODIFIED BANK DIFF 2 sect. 4.2.2e(14) DATE USED: RO SRO ROTH CFR 41.7
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- - __-___ _ ---_- __ . . -_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 6 WHICH ONE (1) of the following statements REOUIRE an immediate scram to be inserted by the operator-at-controls? A. The reactor is operating at 100% power when both running CCW Pumps trip, neither pump can be immediately restarted and the standby pump will not start.
B. The reactor is operating at 50% power, when a seismic event causes the "A" Reactor Recirculation Pump to trip. 1 C. A reactor startup is in progress with reactor pressure at 400 psig when the mnning CRD Pump trips and one (1) scram accumulator is declared inop for a fully inserted control rod.
D. The reactor is operating at 75% power when one (1) of the operating Reactor Feed Pumps trips and level drops in the reactor to 25 inches QUESTION l NRC RECORD # WRI 6 ANSWER: A. SYSTEM # P42 K/A 295018 AA1.02: 3.3/3.4 LP# OP-LOR-ONEP-LP-001-
OBJ. 1 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-V-1 sect. NEW CLASS 01-S-06-2 sect. 6.3.6 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR 41.4 , i l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION l l WRITTEN EXAMINATION MARCH 1998 i SENIOR REACTOR OPERATOR ! r QUESTION 7 l
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Following a Recirc line rupture, reactor level has decreased to - 80 inches.
Both trains of the Standby Gas Treatment System have initiated.
Which of the following best describes the operation of the Standby Gas Treatment System flow control dampers? l A. When -0.2 inches water column is obtained in the Enclosure Building, the steam tunnel cooler dampers throttle to their intermediate position. 90 seconds later the remaining flow control dampers throttle to their intermediate position.
i B. When -0.25 inches water column is obtained in the Enclosure Building, the steam
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tunnel cooler dampers throttle to their intermediate position.120 seconds later the remaining flow control dampers throttle to their intermediate position.
I i C. After 90 seconds, the flow control dampers will go to their intermediate positions to maintain - 0.75 inches water column in the Auxiliary Building and -0.25 inches l water column in the Enclosure Building.
D. After 90 seconds the flow control dampers throttle to maintain -0.25 inches water column. If the Enclosure Building pressure reaches -0.75 inches water column, the flow control dampers go to their intermediate positions.
QUESTION l NRC RECORD # WRI 7 ANSWER: A. SYSTEM # T48 K/A 261000 A1.04: 3.0/3.3 LP# GG-1-LP-RO-T4801.00 OBJ. 7a, 8,15 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E- 1257- 08,11,23 NEW CLASS MODIFIED BANK DIFF 4 annual exam wk 6 DATE USED: RO SRO ROTH CFR 41.13 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 8 The plant is in a LOCA with ECCS systems injecting to the reactor.
Suppression Pool level has lowered to 13.5 feet.
Which one of the following is a condition that exists due to this level? A. The SRV tailpipe exhausts have been uncovered.
B. The RCIC Turbine Exhaust has been uncovered.
C. Suppression Pool temperature cannot be determined.
D. Containment Pressure cannot be determined 1
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QUESTION l NRC RECORD # WRI 8 ANSWER: C. SYSTEM # E30 K/A 295030 EA2.02: 3.9/3.9 j LP# OP-LO-EP-LP-005-03 ' OBJ. 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-EP-3 Caution 2 NEW CLASS MODIFIED BANK DIFF 2 annual exam SRO Rem 1 DATE USED: RO SRO BOTH CFR 41.9 l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 9 The plant is in a 1 OCA with ECCS systems injecting to the reactor.
Containment yessure has increased to 21.1 psig.
The Emergency Procedures have direction to vent the Containment irrespective of offsite release rates.
What is the bases for this action? A. The Primary Containment is not accessible to personnel to attempt mitigation cf the event until pressure is decreased to less than 20 psig.
B. Venting the Primary Containment is done to prevent an uncontrolled release due to a breach of the Primary Containment.
C. Ifpressure in the Primary Containment exceeds 22 psig, the Suppression Pool Level instrumentation will fail causing an uncontrolled release of the Suppression Poolinto the Auxiliary Building.
> D. Venting of Containment at this level is postulated to not exceed the ALERT release limits of the Emergency Plan.
QUESTION l NRC RECORD # WRI 9 ANSWER: B. SYSTEM # M41 K/A 295024 EK3.03: 3.6/4.1 LP# OP-LO-EPB-HN-003-02 LP# OP-LO-EP-LP-005-03 OBJ. 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-1-01-EP-3 NEW CLASS MODIFIED BANK DIFF 3 , DATE USED: RO SRO BOTIf CFR 41.9 , 3/19/98
, U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 10 The reactor was operating at the end of cycle just prior to a refueling outage when a reactor scram occurred.
Which one of the following is a correct method of verifying the position of the control rods? (The scram has NOT been reset.)
A. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw Data and observe a blank display with only green LEDs for all control rods.
B. Using the full core display on H13-P680, depress ALL RODS with RCIS in Raw i Data and observe all control rods indicate 00 with a green LED for all control i rods.
i C. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe a blank display with only red LEDs for all control rods.
D. Using the full core display on H13-P680, depress ALL RODS with RCIS out of Raw Data and observe all control rods indicate 00 with a red LED for all control rods.
QUESTION l NRC RECORD # WRI 10 ANSWER: A. SYSTEM # C11-2; K/A 295006 AA2.02: 4.3/4.4 C11-1B 201005 A3.02: 3.5/3.5 LP# CG-1-LP-RO-C1118.00 A4.02: 3.7/3.7 OBJ. 3c LP# GG-1-LP-RO-C1102.02 OBJ. 10,11,22 LP# OP-LO-ONEP-LP-001-04 OBJ. 1 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 04-1-01-C11-2 NEW CLASS ' sect. 4.7.2p & 4.8.2i MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTH CFR ! 41.6/41.10/43.5 3/19/98
- _ _ _ - _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTENEXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 11 The Electrical line up is normal. A LOCA condition has caused Drywell Pressure to increase to 1.6 psig.
A switching error causes 500 KV voltage to decrease.
The voltage to ALL ESF busses DECREASES to 3290 volts.
The voltage transient duration is 10 seconds and then voltage returns to normal.
Which ONE of the following statements is the condition of the ESF busses after this voltage transient?
' A. 15AAis being supplied from ESF 11 16AB is being supplied from ESF 21 17AC is being supplied from ESF 21 B. 15AAisbeing supplied from DivID/G 16AB isbeing supplied from DivIID/G 17AC is being supplied from Div III D/G C. 15AAis being supplied from ESF 11 16AB is being supplied fromESF 21 17AC is being supplied from Div III D/G D. 15AAis being supplied fromDivID/G 16ABis being supplied from DivIID/G 17ACis being supplied fromESF 21 QUESTION l NRC RECORD # WRI 11 ANSWER: B. SYSTEM # R21; K/A 262001 A2.11: 3.2/3.6
, P75; P81 A3.01: 3.1/3.2 LP# GC-1-LP-RO-R2100.01 264000 K4.08: 3.8/3.7 OBJ. 12,14,20,22,28 LP# GG-1-LP-RO-P7500.00 OBJ. 8,15 LP# GG-1-LP-RO-P8100.01 OBJ. 8,15 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: 04-1-01-R21-1 sect. 5.1 NEW CLASS 04-1-01-P81-1 sect. 3.22 MODIFIED BANK DIFF 3 LOT 7/95 C 13 , DATE USED: RO SRO BOTH CFR 41.4 l 3/19/98-
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 12 The LSS panels will perform which one of the following functions: A. Starting of all ESF Loads only when there has been a total loss of offsite power when the Diesels are carrying the buses.
B. Sequencing of ESF loads to ensure ESF Bus voltage and frequency are not degraded and to minimize stress on the diesel.
C. Starting of all loads on Division I, II, and III required following a DB A LOCA.
D. Sequences HPCS, LPCS and RHR to ensure the diesel engines aie not damaged on a concurrent LOCA and loss ofoffsite power.
QUESTION l NRC RECORD # WRI 12 I ANSWER: B. SYSTEM # R21; K/A 264000 K5.06: 3.4/3.5 P75 LP# GG 1-LP-RO-R2100.01 264000 K4.08: 3.8/3.7 OBJ. 3a SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: Tech Spec Bases B3.8.1 NEW CLASS MODIFIED RANK DIFF 2 LOT 7/95 c13 l DATE USED: RO SRO ROTH CFR 41.8 i I
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIORREACTOR OPERATOR QUESTION 13 The Auxiliary Building Isolation signal was received.
P53-F026A (Instrument Air Aux. Bldg Isolation Valve) has a flashing red indication on the Isolation Status Panel.
Which one of the following best describes the reason for this indication? A. The valve has received an isolation signal and closed within the specified time B, The valve has received an isolation signal and has not repositioned within the specified time.
C. The valve may be reopened after the adjustable timer circuit has timed out.
D. The valve has been overridden open with an isolation signal present QUESTION l NRC RECORD # WRI 13 ANSWER: B. SYSTEM # M72 K/A 223002 A4.04: 3.5/3.6 A1.01: 3.5/3.5 LP# GG-1-LP-RO- A3.01: 3.4/3.4 ' M7200.00 OBJ. 5 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENC FSAR sect. 7.5 NEW CLASS E: MODIFIED BANK DIFF 2 LOT 7/95 c13 DATE RO SRO ROTH CFR 41.8 USED: l 3/19/98 i
- - - - _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERc. TOR QUESTION 14 The plant is at 100 % power with I & C APRM A Surveillance in progress when the following indicators are illuminated on the H13-P680 panel.
Pushbutton HCUFAULT Pushbutton ROD DRIFT Pushbutton SCRAM VLVS Pushbutton ACKNHCUFAULT Annunciator"HCU TROUBLE" Annunciator"ControlRod Drif1" Which of the following could be a possible cause of these indications? ASSUME ALL OTHER INDICATIONS ARE NORMAL.
A. A control rod drifling out of the core.
B. A full reactor scram C. A single control rod scram D. The use of the IN TIhER SKIP pushbutton rather than the INSERT pushbutton.
QUESTION l NRC RECORD # WRI 14 ANSWER: C. SYSTEM # C11-2 K/A 201005 A4.01: 3.7/3.7 LP# GG-1-LP-RO-C1102.02 OBJ. 10,11,22 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-C11-2 NEW CLASS sect. 4.7.2 & 4.8.2 MODIFIED BANK DIFF 3 04-1-02-II13-P680 LOT 2/98 rxinmu DATE USED: 4A2-D4 & 4A2-E4 RO SRO ROTH CFR 41.6 I 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 15 , I GGNS is operating at 10% rated power with the mode switch in the STARTUP position, and ; total core flow at 53%. APRM E and H are bypassed due to failed power supplies. { The fdlowing is the present status of the APRMs versus LPRM inputs and indicated power: ) _ APRM A B C D E F G H
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LPRM LVL D 5 5 2 2 3 2 4 5 LPRM LVL C 5 4 3 5 4 4 3 4 , LPRM LVL B 3 2 5 4 4 3 3 3 LPRM LVL A 2 4 4 4 4 4 5 3 INDICATED 10 % 13 % 12 % 14 % 0% 11 % 13 % 0% l POWER byp byp LPRM 26-43D has failed downscale and must be bypassed to allow troubleshooting.
l With present conditions would this action be allowed? Attached is the LPRM vs APRM assignments table.
A. Yes, conditions are satisfactory.
B. Yes, however an LCO would have to be written on the associated APRM for Administrativeinputs.
C. No, this action would result in a half scram and administrative LCO requirements not to be met.
D No, this action would result in a full reactor scram.
QUESTION l NRC RECORD # WRI 15 , ANSWER: D. SYSTEM # C51-2 K/A 215000 K4.02: 4.1/4.2 LP# GG-1-LP-RO-C510400 A1.04: 3.6/3.6; A4.01: 3.7/3.7 OBJ. 4,9,10a SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-C51-1 sect. 3.3 NEW CLASS 17-S-02-40 sect. 6.3 MODIFIED BANK DIFF 3 Tech Spec Bases B3.3.1.1 DATE USED: RO SRO BOTH CFR 41.6 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 16 I i The plant is in a refueling outage with the refueling platform located over the Dryer Storage Area.
Which one of the following WILL PREVENT movement of the refueling platform over the reactor vessel core? A. The Main Hoist unloaded, control rod 28-37 is selected at position 00 on H13-P680, and the Reactor Mode Switch in REFUEL.
B. The Main Hoist unloaded, one control rod at position 48, and the Reactor Mode Switch in REFUEL.
C. The Main Holst unloaded, all control rods inserted, and the Reactor Mode Switch in REFUEL.
D. The Main Hoist unloaded, control rod 28-37 selected in gang mode at H13-P680, and the Reactor Mode Switch in REFUEL.
QUESTION l NRC RECORD # WR1 16 ANSWER: D. SYSTEM # Fil K/A 234000 K6.03: 3.0/3.6 LP# GG-1-LP-RF-F1101.00 A3.02: 3.1/3.7 OB 11a,c,28,36 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 3 J.
REFERENCE: 04-1-01-F11-1 Att. V NEW CLASS MODIFIED BANK DIFF 3 LOT 2/98 rxsys DATE USED: RO SRO BOTH CFR 41.4/43.7 l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 17 The plant has scrammed due to high reactor water level. The water level peaked at + 58 inches. The plant is now stable at 950 psig and +25 inches. All systems functioned properly following the scram.
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Which one of the following is the correct status of Scram pilot solencids, Backmp scram solenoids, and ARI solenoids? SCRAMPILOT BACKUP SCRAM ARI SOLENOIDS SOLENOIDS SOLENOIDS A. De-energized Energized Energized B. De-energized Energized De-energized C. Energized Energized Energized D. Energized De-energized De-energized QUESTION l NRC RECORD # WRI 17 ANSWER: B. SYSTEM # C71; K/A 212000 K1.06: 3.5/3.6 C11-1 A A2.19: 3.8/3.9 LP# GG-1-LP-RO-C7100.00 A2.20: 4.1/4.2 OBJ 13d,18 A4.12: 3.9/3.9 LP# GG-1-LP-RO-C111A.00 201001 K1.07: 3.4/3.4 OBJ 9f, 9g, 9h SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E- 1173 - 15-21 NEW CLASS E- 6066 - 03,06 MODIFIED BANK DIFF 3 LOT 2/98 rxsys DATE USED: RO SRO BOTH CFR 41.6 l i l f 3/19/98 L_____________.______.________
_ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 18 The plant is operating at 60% rated power.
Boti. Recire Flow Control Valves are at 25% valve position.
A leak in the Drywell caused Drywell pressure to increase to approximately 1.5 psig.
Following the high Drywell pressure signal, the 'B' Reactor Feed Pump trips and level decreases to +14.2 inches before stabilizing at normal level.
Which of the following statements best describes the response of the Recirc System? A. Flow Control Valves will runback to 15% valve position; Recirc Pumps in Fast Speed.
B. Flow Control Valves will remain at 25% valve position; Recire Pumps in Slow Speed.
C. Flow Controls Valves will runback to 15 % valve position; Recire Pumps in Slow Speed.
D. Flow Control Valves will remain at 25% valve position; Recirc Pumps in Fast ! Speed.
QUESTION l NRC RECORD # WRI 18 ANSWER: B. SYSTEM # B33 K/A 202002 A2.06: 3.3/3.3 LP# GG-1-LP-RO-B3300.00 OBJ 19-22,51 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-B33-1 sect. 4.2 NEW CLASS ARI 04-1-02-II13-P680 MODIFIED BANK DIFF 4 3A-B7; 3A-B8 LOT 2/98 rxsys DATE USED: RO SRO BOTH CFR 41.6 l ! l l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998
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SENIOR REACTOR OPERATOR QUESTION 19 A LOCA has occurred.
Drywell pressure is 1.84 psig.
Reactor water level is +36" and stable.
High Pressure Core Spray Pump has been overridden to STOP.
Division III bus 17AC loses power and is subsequently reenergized by the Diesel I Generator.
Which one of the following describes the condition of the HPCS7 l A. Will reset the overrides and HPCS will re-initiate.
B. Will reset the initiation logic and HPCS will remain secured.
C. Will NOT affect the initiation logic and HPCS will re-initiate.
D. Will NOT affect the initiation logic and HPCS will remain overridden.
, QUESTION l NRC RECORD # WRI 19 ANSWER: D. SYSTEM # E22-1 K/A 209002 K2.03: 2.9/2.8 LP# GG-1-LP-RO-E2201.00 K2.01: 3.3/3.2 OBJ 9,11,13,16 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-1183- 023 NEW CLASS E-1188- 019 MODIFIED BANK DIFF 3 DATE USED: RO SRO POTH CFR 41.7/41.8
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f U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR i QUESTION 20 Which of the following situations would result in an automatic actuation of the Div. II SPMU valves? A. High drywell pressure of1.39 psig ! Suppressic : poollevel of17 feet 8 inches.
15 minutes since 1.39 psig drywell pressure B. Reactor waterlevel of-41.6 inches ; Suppression poollevel of18 feet.
26 minutes since-41.6 inches C. High drywell pressure of1.23 psig Suppression pool level of 17 feet 5 inches.
26 minutes since 1.23 psig drywell pressure D. Reactor waterlevel of-150.3" Suppression pool water level of 17 feet 4 inches.
15 minutes since -150.3 inches QUESTION l NRC RECORD # WRI 20 ANSWER: D. SYSTEM # E30 K/A 223001 A2.11: 3.6/3.8 LP# GG-1-LP-RO-E3000.00 A3.01: 3.4/3.5 OBJ 7 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARI 04-1-02-U13-P870 NEW CLASS 4A-A3; 4A-F1 Af0DIFIED BANK DIFF 3 LOT 2/98 rxsys DATE USED: RO SRO BOTIf CFR 41.7
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3/19/98
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 21 An ATWS is in progress with the MSIV's closed, and reactor water level is being controlled by RCIC at the top of active fuel. Standby Liquid Control is injecting and all emergency procedure curves are in the SAFE region. Suppression Pool level is 17.8 feet and lowering.
Suppression Pool temperature is 125 *F and going up. RHR A and B are unavailable for Suppression Pool Cooling.
Which one of the following systems may be used to reduce the temperature of the Suppression , Pool? A. FuelPool Cooling and Cleanup B. ResidualHeat Removal C C. Suppression PoolMake up D. Alternate Decay Heat Removal QUESTION l NRC RECORD # WRI 21 l ANSWER: C. SYSTEM # E12; K/A 295013 AA1.01: 3.9/3.9 l E30; M41 l LP# OP-LO-EP-LP-005-03 I OBJ 3 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-EP-3 NEW CLASS STEPS 1 & 48 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR. 41.9 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUES 110N 22 A Loss of Coolant Accident is in progress. Reactor Levelis being restored by HPCS and RCIC. Containment parameters are elevated.
Determine which of the following situations would result in Containment Spray initiating automatically.
A. 5 minutes since LOCA RHR A and B pumps were manually overridden to STOP 4 minutes after LOCA Drywell pressure at 2 psig CTMT pressure at 9 psig B. 5 minutes since LOCA RHR A and B in LPCI mode on minimum flow.
Drywell pressureis at 2 psig CTMT pressure at 7 psig C. 15 minutes since LOCA RHR A and B pumps were manually overridden to STOP 4 minutes after LOCA Drywell pressureis 1.5 psig CTMT pressure at 8 psig D. 15 minutes since LOCA RHR A and B in suppression pool cooling Drywellpressure at 1.0 psig CTMT pressure at 2.2 psig ; i I l QUESTION l NRC RECORD # WRI 22 ANSWER: C. SYSTEM # E12 K/A 226001 K4.09: 3.2/3.4 4 I LP# OP-LO-SYS-LP-E12-07 OBJ 10 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS 17A-F3; 20A-B6 MODIFIED BANK DIFF 2 LOT 2/98 eccs DATE USED: RO SRO ROTII CFR 41.9
3/19/98
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'U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 23 A small break 'LOCA has occurred in the drywell. High Pressure Core Spray is not available due to maintenance on the Upper Motor Bearing. RCIC is injecting but level continues to decrease. Feedwater is the apparent source of the leak and has since been isolated. Level is continuing to go down. LPCI A, B, & C and LPCS are operating on minimum flow.
Which one of the following choices best describes the operation of ADS? A. Once the 105 sec. timer has started, you MUST wait for the timer to time out and level to decrease to the top of active fuel and a high drywell pressure signal to be present to automaticallyinitiate ADS.
B. If drywell pressure increases to 1.39 psig a 9.2 minute timer will start and automatically initiate ADS with NO low level signal.
C. Completion of the ADS 9.2 minute timer will start the 105 second timer iflevel remains below -150.3 inches with NO high drywell pressure signal, causing ADS to automaticallyinit(5tte.
.D. Ifreactor level drops below +11.4 inches the ADS 9.2 minute timer starts. Iflevel drops to -150.3 inches, when the 105 second timer times out, ADS will automatically initiate.
QUESTION l NRC RECORD # WRI 23 ANSWER: C. SYSTEM # E22-2 K/A 218000 K5.01: 3.8/3.8 l K4.02: 3.8/4.0 LP# GG-1-LP-RO-E2202-00 K4.03: 3.8/4.0 OBJ 10,21 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS 18A-A1; 18A-A2; 18A-C2 MODIFIED BANK DIFF 3 E - 1161-005 DATE USED: RO SRO BOTH CFR 41.8 3/19/98 t
_ _ - _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ - - - _ _ - . _- __ _ .- _ _ _ - _ _ _ _ _ _ _ l l l U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 24 l The B21-F051B SRV handswitches at Division I and Division II Remote Shutdown panels are in the "OFF" position.
l Select the statement below that beg describes the response of the Safety Relief Valve B21-F051B while the switches are in this position.
A. As reactor pressure increases following a Group I Isolation, the valve will open when Low-Low Set is initiated.
B. All modes of operation of the SRV are irioperable exce the Safety mode which will still open the valve.
C, When the operator at the P601 panel takes the control switch for the valve to the open position,it will open.
D. All modes of operation of the SRV are inoperable except the Safety and ADS modes ofwhich either will still open the valve.
QUESTION l NRC RECORD # WRI 24 ANSWER: B. SYSTEM # E22-2 K/A 239002 K4.05: 3.6/3.7 LP# GG-1-LP-RO-E2202-00 OBJ 9c SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-1161-11, 14 NEW CLASS MODIFIED BANK DIFF 3 LOT 2/98 cces DATE USED: RO SRO BOTH CFR 41.7 l l
f 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION
. WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 25 The plant has experienced a loss ofInstmment air due to a rupture in the common header piping between the Water Treatment Building and the Turbine Building. This has resulted in a complete loss of air. Maintenance estimates it will take 12 to 15 hours to repair. The plant is Shutdown following a 422 day mn. Air pressure to the ADS system is at 80 psig and lowering.
Which one of the following is a method of restoring air pressure to the ADS Valves for Reactor Pressure Control 7 A. Cross tie the Instrument Air and Senice Air Headers in the Auxiliary Building via hose fittings and chicago fittings.
B. Connect a diesel driven air compressor to the Instmment Air Header Drain Line in area 9, 139 ft. elevation of the Auxiliary Building.
C. Connect Nitrogen bottles in area 9, 139 fl. elevation to the Instmment Air Connection and isolate the Instrument Air Header from the ADS air header.
I D. Enter Containment and connect Nitrogen bottles to the Instmment Air Header drain to the ADS air header.
QUESTION l NRC RECORD # WRI 25 ANSWER: C. SYSTEM # E22-2; K/A 295019 AA1.01: 3.5/3.3 F53 LP# GG-1-LP-RO-E2202-00 OBJ 18d LP# GG-1-LP-RO-EP07-00 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-V-9 sect. 3.9 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 L } l l 3/19/98 i t
i-t U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR
' QUESTION 26 The plant was operating at full power when an error while perfomiing a surveillance resulted in aRecirc Flow Controlmnback.
ReactorPoweris presently 79 %. l TotalCore Flowis at 59 Mlbm/hr.
Which one of the following describes the actions to be taken for the present situation? A. No actions required. Monitor for thermal hydraulic instability.
B. Controlled entry is allowed. Monitor for thermal hydraulic instability.
C. Immediately scram the reactor.
D. Immediately take actions to exit the region. Monitor for thermal hydraulic instability.
QUESTION l NRC RECORD # WRI 26 ANSWER: D. SYSTEM # B33 K/A 295001 AA2.01: 3.5/3.8 LP# GG-1-LP-RO-B3300-00 AKl.02: 3.3/3.5 OBJ 46,47 LP# OP-LOR-ONEP-LP-001-04 OBJ 19 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-III-3 P/F MAP NEW- CLASS sect. 2.5 for Region II MODIFIED BANK DIFF 2~ Fast Speed Recire.
DATE USED: RO SRO BOTH CFR 41.5/41.10/43.5 l l 3/19/98 ,
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1993 SENIOR REACTOR OPERATOR QUESTION 27 The plant was operating at 45 % power when a transient on the Entergy Power Grid caused a fast closure on the Turbine Control Valves for the GGNS Main Turbine. The Reactor Water Level Control System maintained level within 10 inches ofnormal level. Reactor Pressure increased slightly, but was handled by the Bypass valves.
Which ene of the following describes the results of this transient? , A. The plant remained at power with a reduced power due to a Recirculation Pump downshift to slow speed.
B. The plant remained at power with a reduced power due to a Recirc Flow Control Valve Runback.
C. The plant has scrammed and the Recirculation Pumps have downshifted to slow speed.
D. The plant has scrammed and the Recirculation Pumps have tripped to off.
QUESTION l NRC RECORD # WRI 27 ANSWER: C. SYSTEM # B33; K/A 202001 K1.28: 3.9/4.1 C71 LP# GG-1-LP-RO-B3300-00 OBJ 24, 25 LP# GG-1-LP-RO-C7100.00 OBJ 9 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-I-1 sect. 4.5 NEW CLASS Tech Specs 3.3.4.1 & MODIFIED BANK DIFF 3 3.3.1.1 DATE USED: RO SRO BOTH- CFR 41.5/41.6/43.6 , I
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a ! 3/19/98 w_-___-__________-_________
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 28 The plant has experienced a small steam leak in the Drywell. Drywell pressure is currently 1.75 psig. Containment temperature is 98 * F. The Plant Supervisor has requested that Containment Cooling be maximized Which one of the following statements is correct concerning the ability to maximize cooling in the Containment with the above conditions? A. Plant Chilled Water flow can be re-established after taking the Auxiliary Building Bypass Switches to BYPASS and restarting Plant Chill Water and starting all Containment Coolers.
B. Plant Chilled Water flow is unavailable such that the Containment Coolers are only able to recirculate the air in Containment.
C. Drywell Chilled Water flow can be re-established by using the keylock switches to open the Containment isolations and re-established water flow to the Containment Coolers which remain mnning.
D. Drywell Chilled Water is cross connected to Plant Chilled Water such that cooling is re-established to the Containment Coolers.
QUESTION l NRC RECORD # WRI 28 ANSWER: B. SYSTEM # P71; K/A 295011 AK2.01: 3.7/4.0 M41 AA1.01: 3.6/3.9 LP# GG-1-LP-RO-M4100-00 OBJ 7d,e LP# GG-1-LP-RO-P7100.00 l OBJ 6, 11 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-III-5 Group 6 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.9 l l 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 29 The Control Room has been evacuated due to a freon leak into the Control Room atmosphere, and plant control has been established at the Remote Shutdown Panels.
The plant was scrammed and levelin the reactor is lowering. RCIC tripped on overspeed and the MSIVs have closed. The Plant Supervisor has directed the use ofRHR A in the LPCI mode to maintain reactor water level.
During the lineup ofRHR A in LPCI mode, you notice two handswitches for the LPCI A Injection Valve (E12-F042A). What is the reason for two handswitches? A. One handswitch is to swap to emergency, removing control from the control room, and the other handswitch operates the valve OPEN or CLOSED.
B. One handswitch is to remove the auto features of the E12-F042A and allow the other handswitch to have total control.
C. One handswitch enables the second handswitch to operate the valve in the open and closed positions.
D. One handswitch is used only when the Division I Lockouts have been transferred to l insert the pressure interlocks. The second handswitch operates the valve in the open ; and closed positions.
i QUESTION l NRC RECORD # WRI 29 ANSWER: C. SYSTEM # C61; K/A 295016 AK2.01: 4.4/4.5 l E12 LP# GG-1-LP-RO-C6100-00 OBJ 6c SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: E-1181- 037 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH CFR 41.7 ! 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION (' WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 30
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, The plant is performing the In-Service Leak Test on the reactor following refueling I operations. A miscommunication results in a significant reactor pressure increase.
Pressure as read on the Control Room Wide range Pressure indication on P680 is pegged l l upscale. l l The Post Accident Pressure recorders indicate that pressure reached 1350 psig.
i I i Which one of the following is a correct statement with regard to the GGNS Safety Limit for Reactor Pressure? A. Reactor Pressure was outside the Safety Limit of 1190 psig because this is referenced on the P680 Wide Range Instrument for Tech Specs.
B. Reactor Pressure was outside the Safety Limit of1325 psig because the Post Accident indication comes from the Water Level instmments reference legs.
C. Reactor Pressure was within the Safety Limit of 1375 psig because the Post Accident indication comes from the Bottom Head.. I D. Reactor Pressure was within the Safety Limit of 1550 psig.
QUESTION l NRC RECORD # WRI 30 ANSWER: .B. SYSTEM # K/A 295025 EK1.05: 4.6/4.7 Tech Specs EK1.02: 4.1/4.2 ;
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LP# OP-LO-PB-LP-001-02 Generic G2.2.22: 3.4/4.1 OBJ 8b & d G2.2.25: 2.5/3.7 LP# OP-LO-PB-LP-003-00 OBJ 4. SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 , REFERENCE: Tech Specs 2.1.2 NEW CLASS I Bases B2.1.2 MODIFIED BANK I DIFF 3 DATE USED: RO SRO BOTH CFR 41.3/43.2 i I l l I 3/19/98 - _ ________ - _ _______ _ ______ __ ____ _ _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 31 The plant has scrammed due to a Reactor Feed Pump trip.
Reactor level decreased such that RCIC and HPCS auto started and restored level to the normal operating band.
Which one of the following best describes the condition of Ventilation Systems in the Auxiliary Building? A. All fan coil units operating, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are operating.
B. All the fan coil units shutdown, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep ) Exhaust fans are operating. ) l C. All fan coil units operating, the Secondary Containment Isolation Valves open, Standby Gas Treatment is shutdown, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are operating.
D. All the fan coil units shutdown, the Secondary Containment Isolation Valves closed, Standby Gas Treatment is operating, Fuel Handling Area and Fuel Pool Sweep Exhaust fans are shutdown.
i l QUESTION l NRC RECORD # WRI 31 ' ANSWER: D. SYSTEM # T41; K/A 288000 K4.02: 3.7/3.8 T42; T48 K4.01: 3.7/3.9 1 K4.03: 2.8/2.9 ' LP# OP-LO-SYS-LP-T41-03 OBJ 4, 6, 7e,h A3.01: 3.8/3.8 LP# OP-LO-SYS-LP-T42-02 OBJ 4, 6, 8e,f LP# GG-1-LP-RO-T4801.00 OBJ 7a, 9f,g,15 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 3 REFERENCE: ARI 04-1-02-H13-P870 NEW CLASS 2A-A3 MODIFIED BANK DIFF 3 05-1-02-IH-5 AB VENT DATE USED: RO SRO BOTH CFR 41.3/43.2 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 32 The plant underwent a transient which initiated a reactor scram. All control rods did not fully insert. Reactor water level was intentionally lowered.
One of three analyzed methods of adequate core cooling during emergency conditions is called
" Steam Cooling with Injection".
What is the minimum reactor level at which this method can be said to be providing adequate core cooling? I A. - 167 " , B. - 192 " C. - 204 " D. Level required to reduce Rx power to < 4%. QUESTION l NRC RECORD # WRI 32 ANSWER: B. SYSTEM # K/A 295031 EKl.01:4.6/4.7 EOP Bases LP# GG-1-LP-RO-EP02A.02 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-2A NEW CLASS MODIFIED BANK DIFF 3 LOT 7/95 ep & bases DATE USED: RO SRO BOTH CFR 41.10/43.5 i i 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 33 The plant was operating at rated conditions when an ATWS occurred.
Several SRVs lifted and increased Suppression Pool temperature.
The Plant Supervisor directed the initiation of Standby Liquid Control.
Level in the reactor was lowered to reduce power production.
! The crew is now inserting rods by driving and scramming.
Under which one of the following conditions would the reactor be considered shutdown and the termination of Standby Liquid Control be allowed? f i A. All rods are at position 00 except for control rod 32-33 is at position 48.
{ B. The Reactor Engineer says that suberiticallity can be guaranteed to 200 'F.
C. All rods are at position 02 except three in different quadrants at position 04.
D. Chemistry has confumed that the Hot Shutdown Boron Weight of SLC has been injected.
QUESTION l NRC RECORD # WRI 33 ANSWER: A. SYSTEM # K/A 295015 AK1.01: 3.6/3.9 EOP Bases LP# GG-1-LP-RO-EP02A.02
- OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE
- 05-S-01-EP-2 A NEW CLASS
{ MODIFIED BANK DIFF 3 LOT 7/95 ep & i be.ses DATE USED: RO SRO BOTH CFR 41.8/43.6 i
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> 3/19/98
~ U. S.' NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR - QUESTION 34 The plant is in a refueling outage with the reactor vessel head removed.
Core Alterations are in progress when the Refueling Bridge operator has a spent fuel bundle at
~t he full up position traversing the Upper Containment Pool in the Reactor cavity area when the i' Fuel Grapple malfunctions and releases the fuel bundle. The bundle drops into the reactor vessel and falls against the north west wall of the reactor. The Refueling Bridge operator notices a large bubble start rising from the area of the fuel bundle.
Which one of the following describes the required actions for this situation? A. Move the Refueling B:idge to the Upper Containment Pool Fuel Storage area and wait there for the bubble to pass.
B. Pick the bundle up with the grapple and place it into the nearest fuel storage location, then move the Refueling Bridge to the Cattle Chute.
C. Suspend core alterations and evacuate the Refueling Bridge area and suspend fuel handling until the cause can be determined.
D. Keep the Refueling Bridge manned in its present position and contact the Refueling Floor Health Physicist to the Refueling Bridge and take radiation surveys. ,
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QUESTION l NRC RECORD # WRI 34 ANSWER: C. SYSTEM # ONEP K/A 295023 AA2.04: 3.4/4.1 AKl.01: 3.6/4.1 AK3.01: 3.6/4.3 LP# OP-LOR-ONEP-LP-00144 Generic G2.4.11: 3.4/3.6 i OBJ 1 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 3
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REFERENCE: 05-1-01-U-8 sect. 2.1 NEF CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/41.12/. 43.4/43.5/43.7 i 3/19/98 _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- - _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR
, QUESTION 35 Given the following conditions:
Reactor power 45% Reactorlevel-100 inches Reactor pressure 850 psig Suppression pooltemperature 125'F Suppression poollevel 20 feet 5 inches 2 SRVs are open Which one of the following best describes the required actions to be taken given the above conditions? A. Immediately commence an Emergency Depressurization in accordance with EP-2A because limits in the Containment have been exceeded based on Suppression Pool Temperature.. B. Close the two SRVs and increase the Reactor Pressure band to a top end of 1000 psig, to reduce the amount ofheat entering the Suppression Pool.
i C. Lower Reactor Pressure using cooldown rates that may exceed 100 *F/Hr, to avoid j jeopardizing Containment by exceeding the heat capacity temperature limit of the J Suppression Pool. ) l D. Conditions at present are acceptable, however all pumps taking a suction from the Suppression Pool should be secured.
QUESTION l NRC RECORD # WRI 35 ANSWER: C. SYSTEM # EOP K/A 295026 EKl.02: 3.4/3.8 HCTL Curve AK1.01: 3.6/4.1 AK3.01: 3.6/4.3 ;
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LP# GG-1-LP-RO-EP02A.02 Generic G2.4.11: 3.4/3.6 OBJ 2, 3 SRO TIER 1 GROUP 1/ RO TIER l' GROUP 2 REFERENCE: 05-1.01-EP-2 NEW CLASS steps 38 & 40 MODIFIED BANK DIFF 3 i DATE USED: RO SRO BOTH CFR 41.9/41.10/ 43.5 I l 3/19/98
l U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 36 l The plant isin Mode 3, Hot Shutdown.
A Station Blackout has occurred. q i The Division III Diesel Generator is the only available source of4.16 Kv power.
High Pressure Core Spray has failed to start. i i ReactorLevelis at - 140 inches and decreasing.
Preparations are being made to cross tie the Division III D/G to the Division II ESF Bus.
l At a minimum, who must approve the cross tie ofDivision III D/G and WHY? l
A. Electrical Superintendent because this action could over load the Didsion III Diesel Generator.
I B General Manager, Plant Operations because this will require a change to a Safety Related Procedure. ; C. Manager, Operations because he is required to approve all deviations from normal Operations Procedures.
D. Plant Shift Superintendent because the action is a deviation from requirements of 10 CFR 50 and the GGNS Operating Ijcense.
! QUESTION l NRC RECORD # WRI 36 ANSWER: D. SYSTEM # ONEP K/A 295003 AKl.06: 3.8/4.0 Generic G2.4.7: 3.1/3.8 ' G2.4.11: 3.4/3.6 LP# OP-LOR-ONEP-LP-001-04 G2.4.22: 3.0/4.0 1 OBJ 9 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-1-01-I-4 sect. 3.2.8 NEW CLASS 10 CFR 50.54x MODIFIED BANK DIFF 2 10 CFR 50 APP. A I DATE USED: Criteria 17 RO SRO BOTH CFR 41.7 I i 3/19/98 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
I i ' U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATIONMARCH 1998 SENIOR REACTOR OPERATOR , QUESTION 37 The plant is in a Refueling Outage with the reactor disassembled five (5) day:; after the plant was shutdownin Refueling Outage 08.
Reactor Coolant Temperature is 140 * F
' Reactor Water Level is at the Main Steam lines following Steam Line Plug installation.
The Fuel shufN has not begun.
The inservice sh .down cooling pump hasjust tripped off.
Assume no funher operator action.
Determine for this condition: 1. The time to BOIL for the Reactor Vessel 2. The time for level to reach the Top of Active Fuel l A. 1) 0.75 hours 2) 11 hours B. 1) 0.75 hours ! 2) 15 hours C. 1) 1.5 hours 2) 11 hours D. 1) 1.5 hours 2) 15 hours QUESTION l NRC RECORD # WRI 37 ANSWER: A. SYSTEM # ONEP K/A 295021 AKl.01: 3.6/3.8 LP# OP-LOR-ONEP-LP-001-04 AA2.01: 3.5/3.6 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3
.
REFERENCE: 05-1-01-III-1 Att. I NEW -CLASS MODIFIED bah % DIFF 3 l DATE USED: RO SRO BOTU CFR 41.5/43.5 f 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR t , QUESTION 38 Which one of the following describes the conditions that Cold Shutdown Boron Weight is designed to over come? A. 70 *F, xenon free, water level at steam lines, 50 % rod density.
B. 70 "F, xenon free, water level in normal band, all rods fully withdrawn.
I C. 100 F, xenon free, water level in normal band, all rods fully withdrawn.
D. 100 *F, xenon free, water level at steam lines, 50 % rod density.
QUESTION l NRC RECORD # WRI 38 ANSWER: B. SYSTEM # K/A 295037 EK3.05: 3.2/3.7 I EOP-2A BASES LP# GG-1-LP-RO-EP02A.00 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-2A Bases NEW CLASS Step 21 MODIFIED BANK DIFF 2 LOT 3/98 ep& bases DATE USED: RO SRO ROTII CFR 41.6/41.10/43.6
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3/19/98 u_________________________ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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- U. S. NUCLEAR REGULATORY COMhilSSION
! WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR j QUESTION 39 The plant is operating at 100% power with the Offgas Post Treatment A Radiation i Monitor tagged out for a power supply replacement. l I A Non-Licensed Operator and trainee while conducting training take the handswtich for l
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Offgas Post Treatment Radiation Monitor B to TEST.
Which one of the following describes the impact of this action?
(Assume no further operator action.) l ,
l
, A. The Control Room and back panels will receive an annunciator / alarm only.
B. The steam supply for the Steam Jet Air Ejector will isolate causing a loss of condenser vacuum.
C. The Offgas Charcoal Adsorber Bypass Valve, N64-F045 will close and open the Charcoal Adsorber Outlet Isolation Valve N64-F060 to place the system in treat mode.
D. The Offgas Charcoal Adsorber Outlet Isolation Valve N64-F060 will isolate causing a loss of condenser vacuum to occur.
, l QUESTION l NRC RECORD # WRI 39 ( ANSWER: D. SYSTEM # D17; K/A 272000 K3.05: 3.5/3.7 l N64; N62 l LP# GG-1-LP-RO-D1721.00 OBJ SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 l REFERENCE: ARI 04-1-02-H13-P601 NEW CLASS I , 19A-C8 MODIFIED BANK l l DIFF 2 Tech Spec Loop Logics )
, DATE USED: RO SRO ROTH CFR 41.11/43.4 I i
i
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l I l l \ l 3/19/98 l l
- U. S. NUCLEAR REGULATORY COMMISSION . WRITTEN EXAMINATION MARCH 1998 SENIORREACTOR OPERATOR QUESTION 40 The plant was operating at 100 % power when Suction Valve, N62-F003 A, for the in-service Steam Jet Air Ejector automatically closed. The Operator-at-the-Controls noticed Main Condenserlosing vacuum.
Which of the following best describes the automatic actions that will occur on a degrading condenser vacuum to 0 in Hg vacuum and which actions have a bypass available? A. 21" vac, Main turb trip no bypass 16" vac, Main bypass valves close no bypass 12" vac, Rx feed pumps trip bypass avadable 9" vac,MSIV closure bypass available B. 21" vac, Main turb trip no bypass 16" vac, Rx feed pumps trip bypass available 12" vac, Main bypass valves close no bypass 9" vac, MSIV closure bypass available C. 21" vac, Main tuit trip no bypass 16" vac,MSIV closure bypass available 12" vac, Main bypass valves close bypass available 9" vac, Rx feed pumps tdp bypass available
. D. 21" vac, Main turb trip no bypass 16" vac, MSIV closure no bypass 12" vac,Rx feed pumps trip bypass available 9" vac, Main tuit bypass valves close bypass available
)
i l QUESTION l NRC RECORD # WRI 40 ANSWER: B SYSTEM # N62 K/A 295002 AKl.03: 3.6/3.8 j LP# GG-1-LP-RO-N6200.00 l OBJ 7 LP# OP-LOR-ONEP-LP-001-04 i OBJ 28 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 l REFERENCE: 05-1-02-V-8 sect. 5.0 NEW CLASS MODIFIED BANK DIFF 2 LOT 3/98 stm cond l DATE USED: RO SRO BOTH CFR 41.4 j I 3/19/98 _ _________________________ _ __ -
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ . ___________ - ___ __ - ___-__ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 1 SENIOR REACTOR OPERATOR i QUESTION 41 Plant conditions are as follows: Mode 1 j ReactorPower: 25 % ! Which one of the following describes the response of the RCIS system if the Main Turbine were to trip with no reactor scram? RCIS will: A. implement the constraints of the Rod Withdrawal Limiter allowing rod movements of up to 4 notches.
B. implement the constraints of the Rod Pattern Controller and depending on pattern initiate Insert and/or Withdraw blocks.
C. be between the Rod Pattem Controller and the Rod Withdrawal Limiter indicating the low Power Alarm Point with NO constraints on rod motion.
D. implement +J.c constraints of the Rod Withdrawal Limiter allowing rod movements of up to 2 notches.
QUESTION l NRC RECORD # WRI 41 ANSWER: B. SYSTEM # C11-2; K/A 295005 AA1.03: 2.7/2.8 N32-2 201005 K6.01: 3.2/3.2 LP# GG-1-LP-RO-C1102.02 A1.01: 3.3/3.3 OBJ 6,16,22 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 1 REFERENCE: Tech Specs 3.1.6 NEW CLASS
& 3.3.2.1 MODIFIED BANK DIFF 3 l DATE USED: RO SRO BOTH CFR 41.7
, 3/19/98 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
U. S. NUCLEAR REGULATORY COhtMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR , QUESTION 42 The plant is operating at 20 % power when a Loss ofInstmment Air results in a reactor scram.
The loss ofinstmment air is a rupture of the instmment air header in the Water Treatment Building. The Plant Supervisor has directed the Control Room Operator to maintain water levelin the reactor.
Which of the following best describes the response of the Condensate and Feedwater System? A. Feeding of the Reactor is not available with Condensate and Feedwater due to the , isolation of Condensate Cleanup System.
B. Feeding of the Reactor is not available due to all of the Condensate and Feedwater Minimum Flow Valves failing open diverting all flow to the Condenser.
C. Feeding of the Reactor is available from the Feedwater system while steam is available to the RFPTs and afterwards at lower reactor pressures using Condensate and Booster pumps through the startup level control valve.
D. Feeding of the Reactor is available from the Feedwater system while steam is available to the RFPTs and afterwards at lower reactor pressures using Condensate and Booster l pumps through the startup level control bypass valve.
! I QUESTION l NRC RECORD # WRI 42 ANSWER: D. SYSTEM # N19; K/A 259002 K6.01: 3.2/3.2 N21; N22; P53 LP# GG-1-LP-RO-N2100.00 OBJ 9e,12 LP# GG-1-LP-RO-N1900.00 OBJ 21,22g,25 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 05-1-02-V-9 NEW CLASS sect. 5.21 - 5.24 MODIFIED BANK . DIFF 3 05-1-02-V-7 sect. 2.1.4 bank quest.
I DATE USED: RO SRO BOTH CFR 41.5 f 3/19/98
i U. S. NUCLEAR REGULATORY COMMISSION ! WRITTEN EXAMINATION MARCH 1998 i SENIOR REACTOR OPERATOR QUESTION 43 The plant is operating at 60% power.
The C011 A HPU oil pump is in operation for "A" RFPT with the C010A HPU oil pump tagged out for repairs.
HPU Oil Pump C011 A trips on motor overcunent.
Which of the following describes the response of the Feedwater system and the plant?
(Assume no operator actions.)
A. The A RFPT will lockup at its present speed such that any changes in feed flow will have to be controlled by the B RFPT.
B. The A RFPT will trip on low govemor oil pressure causing level to drop to level 3 causing a reactor scram on low level and a Recire Pump downshift.. C. The A RFPT will mnback to minimum flow and the B RFPT will increase speed ' automatically to compensate.
D. The A RFPT will trip on low governor oil pressure and the B RFPT will increase speed automatically to compensate, a Recirc Flow Control Valve mnback may occur.
QUESTION l NRC RECORD # WRI 43 ANSWER: D. SYSTEM # N21 K/A 259001 A1.05: 2.8/2.7 LP# GG-1-LP-RO-N2100.00 OBJ 7d,8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: 04-1-02-H13-P680 NEW CLASS 2A-A2 MODIFIED BANK DIFF 3 ' DATE USED: RO SRO BOTH CFR 41.4 i l l l l 3/19/98 '
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR QUESTION 44 GGNS Main Generator has a limit to carry no more than i 250 MVARs.
What is the basis for this limitation?
A. This is the Maximum reactive load allowed by the manufacturer due to the heat build up in the stator windings at full power.
l B. GGNS is a base load station such that Entergy dispatchers are required to minimize the l
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reactive load carried on the Main Generator.
C. GGNS Main Generator reverse power relays will not recognize a reverse power I condition at high reactive load and will not provide the required trip.
) D. The Generator V-Curves supplied by the manufacturer limit the power factor on the I generator to reduce hysteresis losses.
QUESTION l NRC RECORD # WRI 44 ANSWER: C. SYSTEM # N41 K/A 245000 A4.14: 2.5/2.5 A3.10: 2.5/2.6 LP# GG-1-LP-RO-NI/51.00 K4.06: 2.7/2.8 OBJ 14 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-N40-1 sect. 3.8 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH l ' 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 45 ! Invener 1Y95 is operating on its alternate source following a transfer using its manual bypass
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switch. The opera;or has been requested to return the Inverter back to its nonral power i
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source.
' ! Which one of the following is NOT required to transfer the Inverter load back to the normal source? ;
A. Alte:nate and Normal Source voltages to be matched. ; ' B. Alternate and Normal Source output currents to be matched. I l C. Alternate and Normal Source frequencies to be matched.
D. Alternate and Normal Source phases to be in sync.
! i QUESTION l NRC RECORD # WRI 45 ,
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ANSWER: B. SYSTEM # L62 K/A 262002 K4.01: 3.1/3.4 LP# GG-1-LP-RO-L6200.00 l OBJ 7,11 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 ! REFERENCE: 04-1-01-L62-1 sect. 3.2 NEW CLASS l MODIFIED BATE DIFF 2 j DATE USED: RO SRO BOTH CFR 41.4 ! 3/19/98
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,. .. . . .. . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 46 i The plant is operating at full power. Health Physics personnel are setting up a controlled area 1 at the Containment Steam Tunnel and cause the CCW to the RWCU Non-Regenerative Heat i Exchanger valve (P42-F103) to go closed.
What is the affect of this valve going closed on the RWCU System?
(Assume no operator actions.)
A. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that RWCU Pump Suction CTMT OTBD Isol valve (G33-F004) will isolate, which ; will trip the RWCU Pumps. i B. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that RWCU Pump Suction DRWL INBD Isol valve (G33-F001) and RWCU Supply to RWCU HXS valve (G33-F251) will isolate, which will trip the RWCU Pumps.
C. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that RWCU Filter Demin Bypass valve (G33-F044) will open and lock the Filter Deminsin hold.
D. The Non-Regenerative Heat Exchanger outlet temperature will increase to the point that automatically reopens CCW to the RWCU Non-Regenerative Heat Exchanger valve (P42-F103). If temperatures continue to increase, at 150 *F the FilterDemins will lock in hold, and RWCU Filter Demin Bypass valve (G33-F044) will open.
' QUESTION l NRC RECORD # WRI 46 ANSWER: A. SYSTEM # G33/36; K/A 204000 A3.04: 3.4/3.5 P42 A2.01: 3.2/3.4
.LP# GG-1-LP-RO-G3336.01 A2.14: 3.2/3.2 OBJ 8,9,10,16i, SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2
REFERENCE: 04-1-01-G33-1 sect. 3.1 NEW CLASS ARI 04-1-02-H13-P680 MODIFIED BANK DIFF 3 11A-C6 l DATE USED: RO SRO ROTH CFR 41.4 l l 3/19/98 l
I U. S. NUCLEAR REGULATORY COMMISSION .)
- WRITTEN EXAMINATION MARCH 1998 ,
SENIOR REACTOR OPERATOR l QUESTION 47 RHR A is operating in Suppression Pool Cooling Mode.
The plant experiences a LOCA. ' What will be the response of the RHR A System and how can the system be retumed to I Suppression Pool Cooling Mode of operation with the LOCA? The RHR A System will: j A. isolate RHR A Test Retum to the Suppression Pool valve (E12-F024A), and open l
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RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for i LPCI mode it is unable to be returned to Suppression Pool Cooling.
B. isolate RHR A Test Retum to the Suppression Pool valve (E12-F024A), and open RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for LPCI mode, E12-F024A and E12-F048A can be immediately manually overridden for Suppression Pool Cooling.
, C. isolatu RHR A Test Return to the Suppression Pool valve (E12-F024 A), and open RHR Heat Exchanger Bypass valve (E12-F048A). Once the system has realigned for LPCI mode, E12-F024A can be immediately manually overridden open and E12-F048A closed after a time delay.
D. require manual realignment to the LPCI mode by closing RHR A Test Retum to the j
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Suppression Pool valve (E12-F024A), and opening RHR Heat Exchanger Bypass valve (E12-F048A). Once the system is realigned for LPCI mode, E12-F024A can be l manually overridden open and E12-F048A closed after a time delay, i QUESTION l NRC RECORD # WRI 47 ANSWER: C. SY3 TEM # E12 K/A 219000 A1.08: 3.7/3.6 q
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A2.14: 4.1/4.3 A3.01: 3.3/3.3 LP# OP-LO-SYS-LP-E12-07 A4.06: 3.9/3.7 OBJ 7, 9 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2
"
REFERENCE: 04-1-01-E12-1 sect. 3.4 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 3 20A-C6; 20A-B5
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DATE USED: 05-1-02-HI-5 Group 5 RO SRO BOTH CFR 41.9 l l 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 48 RHR B Pump Room temperature increases to 170 ' F.
Which one of the following identifies the systems or components besides RHR B which will be affected by this temperature? A. RCIC B. RWCU C. RCIC and MSIVs I D. HPCS l
)
QUESTION l NRC RECORD # WRI 48 ANSWER: A. SYSTEM # E12 K/A 219000 A1.08: 3.7/3.6 A2.14: 4.1/4.3 A3.01: 3.3/3.3 LP# GG-1-LP-RO-E5100.00 OBJ 14 LP# OP-LO-SYS-LP-E12-07 A4.06: 1.9/3.7 OBJ 8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 2 REFERENCE: 17-S-06-5 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 2 20A-B1 DATE USED: 05-1-02-III-5 Group 2,3, RO SRO BOTH CFR 41.4
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l I l 3/19/98 - _
- _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ . - _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 49 RCIC is in a standby lineup aligned to the Condensate Storage Tank (CST) for a ruction source Weeping SRVs cause Suppression Pool Level to increase such that High Suppression Pool Water Level alarms are received on H13-P601 and H13-P870 panels.
Which one of the following describes the response of the RCIC system to this condition? A. RCIC Suction from the Suppression Pool E51-F031 will open and E51-F010 Suction from the CST will close. This lineup can be manually overddden back until Suppression Pool Level is returned to normal at which time the transfer signal will clear.
B. RCIC Suction from the Suppression Pool E51-F031 will open and E51-F010 Suction from the CST will close. This lineup CANNOT be overridden back until the high Suppression PoolLevelis cleared.
C. RCIC Suctions will remain in standby configuration until a RCIC initiation signal is received at which time they will transfer with RCIC Suction from the Suppression Pool E51-F031 opening and E51-F010 Suction from the CST closing.
D. RCIC Suction from the Suppression Pool E51-F031 will open and E51-F010 Suction from the CST will close. When Suppression Pool Level is retumed to normal the lineup will automatically return to the original standby lineup.
QUESTION l NRC RECORD # WRI 49 ANSWER: B. SYSTEM # E51 K/A 295029 EA1.04: 3.4/3.5 LP# GG-1-LP-RO-E5100.00 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 04-1-01-E51-1 sect.3.7 NEW CLASS ARI 04-1-02-H13-P601 MODIFIED BANK DIFF 2 21A-C5 l DATE USED: RO SRO BOTH CFR 41.4 3/19/98 l l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 50 ; RCIC is operating in response to a LOCA signal.
Reactor water level increases to 55 inches. j What is the expected response ofRCIC7 (Assume no further operator actions.)
A. The RCIC Steam Supply Isolation Valve, E51-F045, will close and the RCIC Trip / Throttle Valve will trip stopping the RCIC Turbine.
B. The RCIC Trip / Throttle Valve will trip and the RCIC Injection Valve E51-F013 will close.
C. The RCIC Steam Supply Isolation Valves, E51-F063 and F064, will close and the RCIC Suppression Pool Suction Valve, E51-F031, if open will isolate.
D. The RCIC Steam Supply Isolation Valve, E51-F045, will close and the RCIC Injection Valve E51-F013 will close.
QUESTION l NRC RECORD # WRI 50 ANSWER: D. SYSTEM # E51 K/A 217000 A3.05: 3.9/3.9 LP# GG-1-LP-RO-E5100.00 OBJ Sc, j SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-1185- 1, 6,34, 35, 42 NEW CLASS ARI 04-1-02-H13-P680 MODIFIED BANK DIFF 2 4A2-D1 DATE USED: RO SRO BOTH CFR 41.7
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3/19/98
- - - _ _ _ _ _ __ - __- _-_-_ - __ __ ________-_ _ _ _ - . _ _ _ _ _- _- - _ _ - _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 51 A fire in Panel 11DK caused a complete loss of the K DC Bus.
Which one of the following is the effect this loss will have on the ATWS ARI/RPT System? A. The ATWS ARI portion of the system is INOP due to the loss of halfof the valves required to make the system operate, ATWS RPT will still trip the Recire Pumps.
B. The ATWS ARI portion ofthe system will operate because half ofthe valves will energize and depressurize the air header, ATWS RPT will still tdp the Recirc Pumps.
C. The ATWS ARI portion of the system is INOP due to the loss of halfof the valves required to make the system operate, ATWS RPT is INOP because the Slow Speed breakers will NOT operate to trip the Recirc Pumps.
D. The ATWS ARI portion of the system will operate because half of the valves will energize and depressurize the air header, ATWS RPT is INOP because the Slow Speed breakers will NOT operate to trip the Recirc Pumps.
QUESTION l NRC RECORD # WRI 51 ANSWER: A. SYSTEM # C11; K/A 263000 K3.03: 3.4/3.8 B33 K3.02: 3.5/3.8 LP# GG-1-LP-RO-B3300.00 201001 K2.05: 4.5/4.5 OBJ 27 LP# CG-1-LP-RO-C111A.00 OBJ 9h,18 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: E-6066 - 2, 5, 6 _ NEW CLASS MODIFIED BANK DIFF 4 DATE USED: RO SRO BOTH CFR 41.4 i i 3/19/98
l U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR l QUESTION 52 l The plant is at 100 % power with the Electrical Distribution System in a preferred lineup.
Service Transformer 11 locked out on sudden pressure. < Which one of the following describes the status of the Main Steam Isolation Valves (MSIVs)? l A. Inboard MSIVs have Isolated and the Outboard MSIVs are Open.
B. Inboard MSIVs are Open and the Outboard MSIVs have Isolated.
C. All MSIVs are Open with a loss ofpower to halfof their pilot solenoids.
D. AllMSIVs areIsolated. ! QUESTION l NRC RECORD # WRI 52 ANSWER: C. SYSTEM # B21; K/A 239001 K2.01: 3.2/3.3 R21; C71 K5.08: 2.6/2.7 LP# GG-1-LP-RO-C7100.00 K6.01: 3.1/3.3 OBJ 5,8d LP# GG-1-LP-RO-B1300.00 OBJ 4 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 REFERENCE: 04-1-01-C71-1 Att. III NEW CLASS 05-1-02-III-2 sect. 5.2 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 l l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCII 1998 SENIOR REACTOR OPERATOR QUESTION 53 The plant is at 100 % power.
The Main Steam Line Radiation Alarms for all four (4) divisions come in indicating rad levels greater than 3 times normal background.
Which one of the following describes the response expected from this signal? A. MSIVs and Reactor Sample Valves isolate, and the Reactor Scrams.
B. Reactor Sample Valves isolate and the Reactor Scrams.
C. MSIVs and Reactor Sample Valves isolate.
D. Reactor Sample Valvesisolate.
t QUESTION l NRC RECORD # WRI 53 ANSWER: D. SYSTEM # B21; K/A 295033 EK3.03: 3.8/3.9 D17 LP# GG-1-LP-RO-D1721.00 OBJ 13 LP# OP-LOR ONEP-LP-001-04 OBJ 31 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 05-1-02-III-5 Group 10 NEW CLASS ARI 04-1-02-II13-P601 MODIFIED BANK DIFF 2 18A-C4; 19A-C4 DATE USED: RO SRO BOTU CFR 41.11/43.4 l l l l l l t i 3/19/98
_ _ _ _ _ _ _ - _ _ _ _ _ _ ______-_ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ - U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 54 The plant is operating at 100 % power when the 11DA BUS is de-energized.
Electricians have found the cause of the loss and effected repairs and are ready to restore the bus to service.
Which one of the following describes how an inadvertent initiation ofDivision I ECCS Systems is prevented?. A. The Initiation RESET Pushbutton for Division I is to be held depressed to prevent the initiation logic from picking up.
B. The Trips units for the ECCS initiation logics are powered from DC, however, the instrumentation which supply the inputs to the trip units are powered from Uninterruptable Power System (UPS) and have remained energized and in a normal state.
C. The Instmmentation which supply's the ECCS Initiation Logic Trip Units is arranged such that the instrument, on a loss of power, fails in a direction which will not cause an actuation of the ECCS Systems.
D. The Trip units for the ECCS initiation logics are arranged such that a time delay relay allows instrumentation and logics time to re-energize prior to sending an
- initiation signal.-
QUESTION NRC RECORD # WRI 54 ANSWER: D. SYSTEM # E12; K/A 295004 AKl.06: 3.3/3.6 E21; L11 LP#0P-LO-SYS-LP-E12-07 l. OBJ 9e,10e,13 I LP# GG-1-LP-RO-E2100.00 l OBJ 6, 13, 16 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: E-1182-23 & 26 NEW CLASS E-1181-68 & 82 MODIFIED BANK l DIFF 3 DATE USED: RO SRO BOTH CFR 41.7 3/19/98 L 1-__ _ _______
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 55 The plant is in a startup following a 32 day outage. MSIVs are closed. Recire Loop Te nperatures are at 180 'F. Control rods are being withdrawn to achieve criticality.
(minimal decay heat) The Operating CRD Pump tripped.
What will be the response of the plant?
(Assume no further operator actions)
A. The reactor water level will increase to the point that a reactor scram is received on High water level.
B. The reactor water level will decrease to the point that a reactor scram is received on Low water level.
C. The plant will scram due to a loss of charging water pressure to the Hydraulic Control Units.
D. The reactor water level will remain stable at its present level.
QUESTION l NRC RECORD # WRI 55 ANSWER: B. SYSTEM # C11-1A; K/A 295022 AK2.04: 2.5/2.7 G33/36; IOI- 1 AK2.05: 2.4/2.5 LP# GG-1-LP-RO-G3336.01 AA1.04: 2.5/2.6
- OBJ 2,12,21
, LP# GG-1-LP-RO-C111A.00 OBJ 18 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: 03-1-01-1 NEW CLASS sect. 2.2.5; 3.3.1d; 3.3.3a MODIFIED BANK , DIFF 3 i DATE USED: RO SRO BOTH CFR 41.5 l I 3/19/98 _ _ _ _ _ _ _ _ _ - _ _
, U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR l
- QUESTION 56
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i Which one of the following is the basis for the automatic initiation of Standby Gas Treatment on High Ventdation Radiation Levels in Secondary Containment? A. This provides for the recirculation of the Secondary Containment atmosphere without exhausting air outside ofContainment.
I B. This provides for the filtration of the Secondary Containment atmosphere of
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radionuclides prior to their release into the environment, maintaining offsite releases to withinlimits.
' C. This provides for the cleanup of the Secondary Containment atmosphere allowing personnel entry into the Secondary Containment during a DBA LOCA.
D. This provides for the maintenance of a positive pressure in the Secondary Containment, , therefore preventing any of the fission products released into the Containment from being released into the environment.
QUESTION l NRC RECORD # WRI 56 ANSWER: B. SYSTEM # T48 K/A 295034 EK3.02: 4.1/4.1 LP# GG-1-LP-RO-T4801.00 261000 K1.01: 3.4/3.6 OBJ 14 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 REFERENCE: Tech Specs B3.6.4.3 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.7/41.13/43.4 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR 1 l QUESTION 57
)
The plant is operating at 100 % power normal operations.
< Which one of the following describes how the Secondary Containment Ventilation Systems prevent the release ofradioactive contaminants? I A. The Auxdiary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation J Systems work together to maintain the Auxiliary Building at a negative pressure and ' monitor the extaust to the atmosphere. Irradiation levels are excessive, signals are sent to isolate the building and initiate an atmospheric treatment system.
B. The Auxiliary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a negative pressure and treat the exhaust of the Auxiliary Building to prevent any release of radioactive materials.
C. The Auxiliary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a positive pressure and monitor the exhaust to the atmosphere. IfRao ation levels are excessive, signals are sent toisolate the building.
D. The Auxiliary Building Ventilation and Fuel Handling Area / Pool Sweep Ventilation Systems work together to maintain the Auxiliary Building at a positive pressure and monitor the exhaust to the atmosphere. Irradiation levels are excessive, signals are sent to initiate an atmospheric treatment system.
QUESTION l NRC RECORD # WRI 57 i.
ANSWER: A. SYSTEM # T42; K/A 295038 EK2.03: 3.6/3.8 T41; T48 EA1.06: 3.5/3.6 LP# GG-1-LP-RO-T4801.00 OBJ 7a,15 LP# OP-LO-SYS-LP-T42-02 OBJ 1, 2 LP# OP-LO-SYS-LP-T41-03 OBJ 1, 2, 4 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2
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REFERENCE: 04-1-01-T48-1 sect. 3.2 NEW CLASS l 04-1-01-T42-1 sect. 3.1 MODIFIED BANK i
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DIFF 2 05-1-02-III-5 AB Vent DATE USED: ARI 04-1-02-H13-P870 RO SRO BOTU CFR 41.13/43.4 i 2A -A3 . l
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l ? ! l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR l QUESTION 58 The plant is operating normally at 100 % power.
The Suppression Pool Hi/Lo Level and a LPCS Room Sump Level Hi-Hi annunciators have been received on the H13-P870 panel.
The Control Room Operator has noted that Suppression Pool Level is at 18.4 feet. An operator dispatched to the room reports that water is spraying from the LPCS Suction piping, but he was unable to tell the exact location.
Which one of the following is appropriate actions for this event? A. Immediately scram the reactor, initiate Suppression Pool Makeup, and emergency depressurize the plant, and isolate the LPCS Suetion from the Suppression Pool.
B. Ensure the LPCS Room sump pumps are operating, isolate LPCS Suction from the Suppression Pool and observe the status of the leak and makeup to the Suppression Pool via normal means, ifrequired open the LPCS Room Door.
C. Monitor and control LPCS Room sump levels, rack out the LPCS Pump Breaker and isolate LPCS Suction from the Suppression Pool, scram the reactor since the Max Safe Level has been reached.
D. Verify the LPCS Room sump pumps are operating, isolate LPCS Suction from the Suppression Pool and rack out the LPCS Pump Breaker, and observe the status of the leak and makeup to the Suppression Pool via normal means.
QUESTION l NRC RECORD # WRI 58 ANSWER: D. SYSTEM # P45; K/A 295036 EA2.03: 3.4/3.8 E12; EOP-4 EK3.03: 3.5/3.6 LP# OP-LO-EP-LP-005-03 EA2.02: 3.1/3.1 OBJ 2 LP# OP-LO-EP-LP-006-03 OBJ Sa SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 3 REFERENCE: 05-S-01-EP-3 step 48 NEW CLASS 05-S-01-EP-4 step 18 MODIFIED BANK DIFF 2 ARI 04-1-02-H13-P680 8Al-A4 DATE USED: ARI 04-1-02-H13-P870 RO SRO ROTH CFR 41.4 4A-A3;2A-FI;4A-C3 3/19/98 _ _ _ _ .
- _ _ _ _ - _ _ _ _ _ _ _ - _-_-_- __ _ _ _________ _ _ _ _ - ._. - U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR l QUESTION 59 I l The reactor is shutdown and the plant is in a forced cooldown to achieve cold shutdown conditions.
Which one of the following best describes the method used to control CRD Flow and Drive pressure during the depressurization process? A. The Pressure Control Valve automatically throttles to maintain 250 psid Drive DP I and the Flow Control Valve automatically throttles in response to a CRD flow setpoint of 60 GPM.
B. The Pressure Control Valve automatically throttles to maintain 250 psid Drive DP and the Flow Control Valve is manually throttled to maintain a CRD flow of 60 GPM.
C. The Pressure Control Valve is manually throttled to maintain 250 psid Drive DP and the Flow Control Valve automatically throttles in response to a CRD flow setpoint of 60 GPM.
D. The Pressure Control Valve is manually throttled to maintain 250 psid Drive DP and the Flow Control Valve is manually throttled to maintain a CRD flow of 60 GPM.
QUESTION l NRC RECORD # WRI 59 ANSWER: C. SYSTEM # C11-1A K/A 201001 K4.08: 3.1/3.0 LP# GG-1-LP-RO-C111A.00 OBJ 9d & e,13 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: M - 1081-B NEW CLASS E-1166- 003; 017 MODIFIED BANK DIFF 3 DATE USED: RO SRO BOTN CFR 41.6
i ! l 3/19/98' J
- _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _-. ___ _ _ _ __ _- _ - _ - _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 60
' Which one of the following is the reason the LPCI Injection Valves, E12-F042A, B, and C, are designed to remain closed at normal reactor vessel pressure following a LOCA initiation signal?
A. This allows the pump time to pressurize the header, thus minimizing the differential pressure across the injection valve. 1 I B. This ensures reactor pressure has decreased sufficiently to prevent the possibility of overpressurizing low pressure piping.
C. This allows the pump to develop enough discharge head to overcome reactor pressure for injection preventing backflow of hot reactor water.
D. This ensures reactor pressure has equalized with LPCI pressure to prevent the injection check valve E12-F041 A, B, C from slamming the injection piping causing damage.
QUESTION- l NRC RECORD # WRI 60 ANSWER: B. SYSTEM # E12 K/A K1.17: 4.0/4.0; K4.01: 4.2/4.2 203000 K4.02: 3.3/3.4; A3.01: 3.8/3.7 LP# OP-LO-SYS-LP-E12-07 A3.08: 4.1/4.1; A4.08: 4.3/4.3 OBJ 7h, 14 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 04-1-01-E12-1 sect. 3.4 NEF CLASS Tech Spec Bases B3.3.5.1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.8 l l ! ! 3/19/98
,_-_ _ - _ - - - - - - - _ - _ - - _ _ - - - _ - . . _ _ _ _ _ - - - _ _ - - . _ _ . _ - _ _ _ _ - . _ _ _ - - i U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 ! SENIOR REACTOR OPERATOR
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QUESTION 61 The plant isin an ATWS.
Standby Liquid Control is out of service.
Which one of the following is a means of altemate Boron injection? l i A. Add sodium pentaborate to the RWCU Filter Demin Precoat tank and inject the boron into the reactorvia RWCU.
B. Add sodium pentaborate to the Suppression Pool and align RHR A or B in Suppression Pool Cooling mode to mix the solution then inject through any available l ECCS pump taking a suction from the Suppression Pool.
C. Add sodium pentaborate to the Condensate Storage Tank and mix with HPCS and inject the boron into the reactor using RCIC with a suction on the CST.
D. Add sodium pentaborate to the Condensate Cleanup Precoat Tank and inject into the Condensate Cleanup system and use the Condensate / Feedwater Systems to provide a , differential pressure to inject the boron into the reactor.
l QUESTION l NRC RECORD # WRI 61 ANSWER: C. SYSTEM # K/A K3.01: 4.3/4.4 EOP-2A 211000 LP# GG-1-LP-RO-EP02A.02 OBJ 3 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 05-S-01-EP-2 Att. 28 NEW CLASS MODIFIED BANK l DIFF 2 l DATE USED: RO SRO BOTH CFR 41.6 i i
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3/19/98 i
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 62 The Reactor Wide Range Level Instrument on condensing pot 9004A has its reference leg flash.
How will this affect the ability of an automatic Containment /Drywell Isolation? A. The Division I Containment /Drywell Isolation valves will isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will isolate on a low reactor waterlevel signal.
B. The Division I Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal.
C. The Division I Containment /Drywell Isolation valves will isolate on a low reactor water level signal; Division II Containment /Drywell Isolation valves will NOT isolate on a low reactor water level signal.
D. The Division I Containment /Drywell Isolation valves will NOT isolate on a low
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reactor water level signal; Division II Containment /Drywell Isolation valves will isolate on a low reactor water level signal.
QUESTION l NRC RECORD # WRI 62 ANSWER: D. SYSTEM # B21; K/A 216000 K3.02: 4.0/4.3 M71 LP# GG-1-LP-RO-B2101.00 ' OBJ 5,7a, 8b,14 LP# GG-1-LP-RO-M7101.00 OBJ 7g, 9, 17 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 17-S-06-5 Att. II NEW CLASS Group 6A MODIFIED BANK l t DIFF 2 05-1-02-III-5 Group 6 ) DATE USED: RO SRO BOTH CFR 41.6 I i
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 63 Under which of the following conditions would the control rods most likely insert i themselves WITHOUT assistance of the Control Rod Drive System and its Scram f accumulators? 1 A. Reactor Pressure is 620 psig.
Instrument Air to the Scram Air Header is pressurized.
B. Reactor Pressure is 620 psig.
Instrument Air to the Scram Air Header is depressurized.
C. Reactor Pressure is 420 psig.
Instrument Air to the Scram Air Header is pressurized.
D. Reactor Pressure is 420 psig.
Instrument Air to the Scram Air Header is depressurized.
~ QUESTION l NRC RECORD # WRI 63 ANSWER: B. SYSTEM # C11-1B K/A 201003 K3.02: 4.0/4.3 LP# GG-1-LP-RO-C111B.00 l1 OBJ 13, 15 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 REFERENCE: Tech Spec Bases B3.1.5 NEW CLASS i MODIFIED BANK ) DIFF 2 1 DATE USED: RO SRO BOTH CFR 41.2 I l I 3/19/98 l
. _ _ _ _ _ _ _ _ - _ _ - _ _ - _ - - _ _ - _ - - _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ - _ . . _ _ _ _ . _ _ - _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 64 Which one of the following situations would require the Hydrogen Recombiners to be prevented from operation? A. Reactor Level is - 170 inches, Drywell Hydrogen Concentration is at 8.2 % and Containment Pressure is 10 psig.
B. Reactor Levelis undetermined, Drywell Hydrogen Concentration is at 6.2 % and Containment Pressure is 1.50 psig. ) C. Reactor Levelis undetermined, Containment Hydrogen Concentration is at 8.2 % and Containment Pressure is 10 psig.
D. Reactor Levelis - 170 inches, Containment Hydrogen Concentration is at 4.2 % , and Containment Pressure is 1.50 psig.
QUESTION l NRC RECORD # WRI 64 ANSWER: C. SYSTEM # E61 K/A 500000 EK3.03: 3.0/3.5 I LP# OP-LO-EP-LP-005-03 l OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: 05-S-01-EP-3 NEW CLASS Figure 5 & Step 66 MODIFIED BANK ,
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DIFF .2 DATE USED: RO SRO BOTH CFR 41.10 l 3/19/98 l
_ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I l U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 65 The plant is involved in a LOCA.
Which one of the following situations would require an Emergency Depressurization of the reactor? A. Reactor Level is - 140 inches and rising Drywell Temperature is 230 * F Containment Temperature is 190 *F
B. Reactor Level is - 120 inches and rising Drywell Temperature is 180 * F Containment Temperature is 110 *F C. Reactor Level is - 30 inches and lowering Drywell Temperature is 240 *F Containment Temperature is 180 *F D. Reactor Level is - 60 inches and lowering Drywell Temperature is 180 * F Containment Temperature is 170 *F ; QUESTION l NRC RECORD # WRI 65 ANSWER: A. SYSTEM # K/A 295027 EA1.03: 3.5/3.8 EOP - 3 Pri Ctmt i LP# OP-LO-EP-LP-005-03 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 2 REFERENCE: 05-S-01-EP-3 NEW CLASS Step 23 MODIFIED BANK , DIFF 2 l DATE USED: RO SRO BOTH CFR l 41.9/41.10/43.5 l l 3/19/98 L _
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION M ARCH 1998 SENIOR REACTOR OPERATOR QUESTION 66 The plant is operating at 75 % power. Plant Services personnel working inside Containment cause an inadvertent initiation of HPCS.
Which one of the following best describes the response of the Reactor Level Control System in Master Auto?
(Assume NO operator action.)
A. The Reactor Level Control Systenc. will remain static and lock the feed pumps at the current signal, and reactor water level will increase to the high level scram setpoint B. The Reactor Level Control System will increase feedwater flow in response to the lowering water level, and return water level to close to the normal level.
C. The Reactor Level Control System will trip to single element, reduce feedwater flow in response to the rising water level, and lower water level to about the Low Level Alarm.
D. The Reactor Level Control System will reduce feedwater flow in response to the rising water level, and return water level close to the normal level. j QUESTION l NRC RECORD # WRI 66 ANSWER: D. SYSTEM # C34; K/A 295008 AK2.03: 3.6/3.7 N21 LP# OP-LOR-ONEP-LP-001-04 l OBJ ~ 1, 27 SRO TIER 1 GROUP 2 / RO TIER 1 GROUP 2 l REFERENCE: 05-1-02-V-6 NEW CLASS sect. 2.1 & 4.1 MODIFIED BANK DIFF 3 DATE USED: RO SRO ROTH CFR 41.5 1
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 67 The plant is operating at 40 % power.
The "A" Circulating Water Pump develops a phase to phase short which trips the Circ Water Pump. The "B" Circulating Water Pump is tagged out for motor bearing replacement.
Which one of the following best describes the response of the Main Condenser?
(Assume NO operator action.)
A. Main Condenser vacuum will decrease and stabilize at approximately 15 inches Hg Vacuum.
B. Main Condenser vacuum will decrease and stabilize above the turbine trip setpoint, as the Steam Jet Air Ejectors will control Main Condenser Vacuum.
' C. Main Condenser vacuum will decrease and approach 0 inches Hg Vacuum.
D. Main Condenser vacuum will remain stable at its present value, as the Steam Jet Air Ejectors will control Main Condenser Vacuum.
QUESTION l NRC RECORD # WRI 67 ANSWER: C. SYSTEM # N19; K/A 256000 K6.02: 3.1/3.1 N71 LP# GG-1-LP-RO-N1900.00 OBJ 22b,25 SRO TIER 2 GROUP 3 / RO TIER 2 GROUP 2 REFERENCE: 05-1-02-V-8 sect. 4.1 & 3.3 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.4 3/19/98 i
I U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR - l
. QUESTION 68 i
The plant is operating at 70 % power.
Which of the following best describes the response of the Reactor Water Level Cantrol System on a failure of a single Feed Flow Transmitter UPSCALE 7 i l A. The Digital Feed System will recognize the failure and de-select 3 - element control and return level to the level setpoint.
B. The Digital Feed System will decrease feed flow until reactor level decreases to 32 inches at which time it will become level dominant remaining in 3 - element control.
C. The Digital Feed System will decrease feed flow and reactor level will stabilize out at a new low level below the low level alarm setpoint.
D. The Digital Feed System will lock up the controls and hold level at the normal level, remain in 3 - element control, and actuate the DFCS TROUBLE annunciator on P680.
QUESTION l NRC RECORD # WRI 68 ANSWER: A. SYSTEM # C34 K/A 295009 AA2.02: 3.6/3.7 LP# GG-1-LP-RO-C3401.00 259002 K6.04: 3.1/3.1 OBJ 1.10 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: ARI 04-1-02-H13-P680 NEW CLASS 2A-C9 MODIFIED BANK DIFF 3 DATE USED: RO SRO'BOTH CFR 41.7 l l-l 3/19/98
)
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 69 Plant conditions are as follows: MODE: Mode 1 Rx power: 28 % T-G Load: 365 MWE Load Demand 390 MWE Bypass position: 0% All other parameters are per plant design.
The operator withdraws a control rod which increases Rx power to 29 %. How will the Turbine EHC Control System respond? A. The Bypass Control Valves will open by whatever amount is required to maintain RX pressure.
B. The Turbine Control Valves will open by whatever amount is required to maintain Rx pressure.
C. The Bypass Control Valves will close by whatever amount is required to maintain RX pressure.
D. The Turbine Control Valves will close by whatever amount is required to maintahi Rx pressure.
~ QUESTION l NRC RECORD # WRI 69 ANSWER: B. SYSTEM # N32-2 K/A 295007 AK2.01: 3.5/3.7 241000 A2.02: 3.7/3.7 LP# GG-1-LP-RO-N3202.00 K4.01: 3.8/3.8 OBJ 4 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 , REFERENCE: 03-1-01-2 sect. 5.2 NEW CLASS l MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH CFR 41.5 ! l 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 70
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Plant Service Water is lost to the Auxiliary Building. This resulted in a trip of the operating Drywell Chillers on high condenser pressure.
Which one of the following best describes the affects on the Drywell Atmosphere?
(Assume NO operator action.)
A. Drywell temperature will increase, Drywell pressure will decrease to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the Drywell and Containment.
B. Drywell temperature will increase, Drywell pressure will increase to the point that the Drywell Normal Vacuum relief valves will open equalizing pressure between the Drywell and Containment.
C. Drywell temperature will increase, Drywell pressure will remain constant due to the communication between the Containment and Drywell atmospheres.
D. Drywell temperature will increase, Drywell pressure will increase such that high Drywell pressure alarms will actuate and a reactor scram will occur due to a high drywell pressure.
. QUESTION l NRC RECORD # WRI 70 ANSWER: D. SYSTEM # M51; K/A 295010 AK2.05: 3.7/3.8 P72 223001 K6.01: 3.6/3.8 I l LP# GG-1 LP-RO-M5100.00 A4.12: 3.3.5/3.6 OBJ 1,2, 21 SRO TIER 1 GROUP 1/ RO TIER 1 GROUP 1 REFERENCE: ARI 04-1-02-1113-P870 NEW CLASS 3A- D4 MODIFIED BANK DIFF 2 DATE USED: RO SRO ROTH CFR 41.4 i l
3/19/98 - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
U. S. NUCLEAR REGULATORY COMMISSION . ' WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 71 The plant is in a reactor startup just afbr reaching critical.
The Operator-at-the-Controls is withdrawing SRMs.
The following conditions exist: AllIRMs are on Range 2.
SRM A reads 2 x 10 SRM D reads 6 x 10'
SRM B reads 8 x 10' SRM E reads 8 x 10
SRM C reads 2 x 10' SRM F reads 3 x 10 Which one of the following best describes plant conditions? A. Half scram, half rod block.
B. Full scram, rod block.
C. Rod block only.
D. No trips or blocks are present. I l QUESTION l NRC RECORD # WRI 71 ANSWER: C. SYSTEM # C11-2; K/A 215004 A1.04: 3.5/3.5 ! C51; C71 A3.04: 3.6/3.6 l LP# GG-1-LP-RO-C1102.02 ; OBJ 6 j LP# GG-1-LP-RO-C51-1.05 , OBJ 7 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 j REFERENCE: Tech Specs TR3.3.2.1 NEW CLASS MODIFIED BANK DIFF 2 LOT 3/98 rxinst DATE USED: RO SRO BOTII CFR 41.6 ' ! 3/19/98
r U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 72 Plant conditions are as fo!!ows: MODE: Mode 1 Rx power: 40 % T-G Load: 520 MWE Load Demand 510 MWE Bypass position: 0% All other parameters are per plant design.
The Operator-at-the Controls continues withdrawing control rods to increase power. The other Control Room Operators are busy with surveillance and starting up BOP systems.
Power is increased 15 % by control rod movements and Load Demand on the Main Turbine has NOT been adjusted.
Which one of the following best describes the response of the Turbine Control System? A. The Turbine Control Valves will remain open at present positions and Reactor Pressure will increase to the point that the Low-Low Set SRVs open. Bypass Control Valves will remain closed due to the biasing of the control circuitry.. B. The Turbine Control Valves will open as power increases to control Reactor Pressure, increasing generator output. The Bypass Valves will remain closed.
C. The Reactor Pressure will increase corresponding to the 15 % increase, and the Turbine Control Valves will remain at their limited load value, Bypass Valves will remain closed.
D. The Reactor Pressure will increase corresponding to the power increase to the point at which pressure overcomes the biasing, at this time the Bypass Valves will open to maintain Rx pressure.
QUESTION l NRC RECORD # WRI 72 ANSWER: D. SYSTEM # N32-2 K/A 241000 A1.14: 3.4/3.4 LP# GG-1-LP-RO-N3202.00 A1.12: 2.9/2.8 OBJ 1 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 1 REFERENCE: 03-1-01-2 sect. 5.2 caution NEW CLASS
- MODIFIED BANK l DIFF 3 l DATE USED
- RO SRO BOTH CFR 41.5 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR t QUESTION 73
The plant is in a reactor startup with Reactor Power at 10 %. Which one of the following conditions would allow the withdrawal ofIntermediate Range Neutron Instrumentation and the reason for withdrawing the detectors? i A. The Mode Switch in Startup. Only those IRM detectors which have associated i APRMs reading > 5 % power can be withdrawn. Detectors are removed to minimize detector burnout from high flux fields at high power operation.
B. The Mode Switch in Run. All IRM detectors can be withdrawn. Detectors are
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removed to minimize detector burnout from high flux fields at high power operation.
C. The Mode Switch in Startup. AllIRM detectors can be withdrawn as long as IRM to APRM overlap is satisfactorily observed on at least two (2) IRMs. Detectors are removed to prevent localized overheating of the fuel due to the obstructed coolant flow.
D. The Mode Switch in Run. Only those detectors which have had IRM to APRM overlap satisfactorily observed may be withdrawn. Detectors are removed to prevent localized overheating of the fuel due to the obstructed coolant flow.
QUESTION l NRC RECORD # WRI 73 , ANSWER: B. SYSTEM # C51 K/A 215003 K5.03: 3.0/3.1 LP# GG-1-LP-RO-C5102.00 K4.05: 2.9/3.0 OBJ 2a,8 SRO TIER 2 GROUP 2 / RO TIER 2 GROUP 1 REFERENCE: 03-1-01-1 sect. 6.2.17 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.6 l 3/19/98
-_ __ _ -_-__ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - - - _ _ _ _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACf0R OPERATOR QUESTION 74 A surveillance test was in progress operating HPCS in CST to CST mode, when an Auxiliary Building Isolation Signal was received.
An operator reports an alarm on the HPCS/RCIC Test Return Diaphragm in area 7, 119 Ft. elevation of the Auxiliary Building.
Which one of the following best describes the impact of this alarm? A. Secondary Containment should be INOP due to a failure of a Secondary Containment isolation.
B. Primary and Secondary Containment are INOP due to a direct siphon path from - the CST to the Suppression Pool via HPCS.
C. Primary and Secondary Containment are Operable since the Auxiliary Building Isolation Valves for the HPCS/RCIC CST Test Return are isolated.
D. Secondary Containment is Operable since HPCS is performing its ECCS Function for a LOCA condition.
QUESTION l NRC RECORD # WRI 74 ANSWER: A. SYSTEM # E22; K/A 290001 A4.10: 3.4/3.3 T10 LP# GG-1-LP-RO-E2201.00 OBJ 17 SRO TIER 2 GROUP 1/ RO TIER 2 GROUP 2 REFERENCE: 06-OP-1T10-M-0001 NEW CLASS sect.1.1 & 5.3.3 MODIFIED BANK DIFF 2 Tech Specs 3.6.4.2 DATE USED: RO SRO BOTH CFR 41.7/43.4 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACf0R OPERATOR QUESTION 75 A Non-Licensed Operator is being sent out on ajob in a High Radiation Area.
The Dose rate in the area of thejob is 120 mrem /hr. Thejob is expected to take 45 minutes. The operator's exposure history to date for the year is 1800 mrem.
Can the operator be utilized for thisjob and WHY? A. Yes, the operator will not exceed his administrative limits.
B. Yes, the operator will have to have an approved extension on dose limits before thejob.
C. No, the operator will exceed his federal dose limits.
D. No, the operator will exceed administrative dose limits which are not allowed to be extended.
QUESTION l NRC RECORD # WRI 75 ANSWER: A. SYSTEM # K/A G2.3.4: 2.5/3.1 ADMIN Rad Con Generic LP# EOI-S-LP-GET-RWT01.05 OBJ RWT31,32,33 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-08-2 sect. 6.3.2 NEW CLASS MODIFIED BAN'K DIFF 2 DATE USED: RO SRO BOTH CFR 41.12 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR
QUESTION 76 j The Containment Control Procedure, EP-3, Step 31, asks "IS CTMT PRESS ABOVE 2.2 PSIG". I l What is the Basis for this step? A. This is the lowest pressure that will result in the containment pressure remaining ! above atmospheric pressure when containment sprays are initiated.
B. This is the lowest pressure that will prevent evaporative cooling from taking place while the containment is being sprayed.
I C. This will assess whether the previous actions such as venting the containment j have been successful in reducing containment pressure.
l l D. This will ensure containment pressure can be maintained in the safe zone of the ! Pressure Suppression Pressure curve to assure the containment boundary is maintained while the RPV is at pressure. .
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QUESTION SRO 76 l NRC RECORD # WRI 100 ANSWER: A. SYSTEM # E12; K/A 295024 EK2.14: 3.9 ) M41 EK2.15: 3.9 ) LP# OP-LO-EP-LP-005-03 226001 A2.18: 3.5
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OBJ. 3 SRO TIER 1 GROUP 1 / RO TIER GROUP REFERENCE: 05-1-01-EP-3 BASES NEW CLASS 4 EPG BASES MODIFIED RANK DIFF 3 annual exam ep-02 DATE USED: RO SRO BOTH CFR 41.9 t 3/19/98 j
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 77 l The plant is increasing power from 80 % to 100 % power. The "A"Recirc Flow Control Valve hydraulic actuator sticks causing the valve to inadvertently slow open from 60 % valve position to 100 % valve position.
Reactor power stabilizes at 98 % power. j Total Core Flow 105 Mlbm/hr l Fraction ofCore Boiling Boundary 0.79 l MAPFAC(f) 0.91 MCPR 1.25 LHGRFAC(f) 0.89 Which one of the following identifies the thermal limits which have been violated? Power has NOT been changed since the transients.
Consider Flow Based Limits ONLY.
A. FCBB B. MCPR C. LHGR
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D. APLHGR QUESTION SRO 77 l NRC RECORD # WRI 106 , ANSWER: B. SYSTEM # Jil; K/A 295014 AA2.04: 4.4; AK1.05: 4.2 Tech Specs AK2.02: 4.2; AA2.05: 4.6 LP# OP-LO-PB-LP-003-00 OBJ SRO TIER 1 GROUP 1/ RO TIER GROUP REFERENCE: FSAR 15.4.5.3.3.3 NEW CLASS Tech Specs MODIFIED BANK , DIFF 3 3.2.1; 3.2.2; 3.2.3 DATE USED: RO SRO BOTH CFR 41.6/41.14/43.6
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Which of the following choices best fits the definition of Core Alteration? A. Control Rod movement from the control room to vent excess air.
B. With, drawing a control rod from a defueled quadrant of the core.
C. The removal ofan LPRM string.
D. The removal of a double blade guide from the core.
QUESTION SRO 78 l NRC RECORD # WRI 108 ANSWER: A. SYSTEM # K/A Generic G2.2.31: 3.8
. Tech Specs LP# 0P-LO-PB-LP-001-02 OBJ Sc LP# OP-LO-PB-LP-003-00 OBJ 1 SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: Tech Specs 1.1 NEW CLASS MODIFIED BANK DIFF 2-DATE USED: RO SRO BOTH CFR 43.7 l {
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 79 Given the following conditions: Reactor power 20% Reactorlevel-170 inches Reactor pressure 900 psig Suppression pooltemperature 126*F Suppression poollevel 17 feet 4 SRVs are open Which one of the following best describes the correct actions to be taken given the above conditions? A. Commence an emergency depressurization in accordance with EP-2.
B. Raise suppression pool level to avoid a depressurization.
C. Close the SRVs and allow pressure to increase to 1050 psig.
D. Reduce reactor pressure to avoid a depressurization.
' QUESTION SRO 79 l NRC RECORD # WRI 111 ANSWER: A. SYSTEM # Prim K/A 295030 EK1.03: 4.1 , CTMT EOP ' LP# OP-LO-EP-LP-005-03 , OBJ 3 SRO TIER 1 GROUP 1/ RO TIER GROUP REFERENCE: 05-S-01-EP-3 NEW CLASS Steps 49 & 50 MODIFIED BANK DIFF 3 Figures 2 & 7 LOT 7/95 ep DATE USED: RO SRO BOTH CFR 41.10/43.5 l 3/19/98 f L j
_- ___ - _ ____-_ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 80 The Radwaste contractor was attempting to load a High Intensity Cask (HIC) with spent Reactor Water Cleanup Resin when an equipment malfunction caused the filling equipment to spray approximately 2 cubic yards of dry spent resin out the railroad door of the Radwaste Building. The wind has dispersed the resin and its contaminants into the air. The Shift Superintendent has declared a General Emergency due to EAL 5.4.lb. Field monitoring teams and Chemistry have reported a 5450 mrem Thyroid CDE dose commitment at the Claiborne County Emergency Operations Center.
Which one of the following is the Protective Action Recommendation to be issued to the state? A. Evacuate 2 mile radius of the plant, and evacuate the 5 mile down wind sectors and shelter the remaining of the 10 mile Emergency Planning Zone.
B. Evacuate 2 mile radius of the plant, and evacuate the 10 mile down wind sectors and shelter the remaining of the 10 mile Emergency Planning Zone.
C. Evacuate 2 mile radius and the 5 mile radius of the plant and evacuate the 10 mile down wind sectors and shelter the remaining of the 10 mile Emergency Planning Zone.
D. Evacuate 2 mile radius, 5 mile radius, and 10 mile radius of the plant and shelter the 50 mile down wind sectors of the Emergency Planning Zone.
QUESTION SRO 80 l NRC RECORD # WRI 112 ANSWER: B. SYSTEM # K/A 295017 AK2.06: 4.6 EPP PARS LP# GG-1-LP-RO-EPTS6.00 OBJ 2 SRO TIER 1 GROUP 1/ RO TIER GROUP REFERENCE: 10-S-01-1 sect. 6.1.4 NEW CLASS EAL 5.4.lb MODIFIED BANK DIFF 3 5 mile EPZ Map DATE USED: RO SRO BOTH CFR 41.10/43.5 l l.
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 81 The Control Room has been abandoned due to a fire. Prior to leaving the Control Room, the reactor mode switch was placed in the SHUIDOWN position, all rods were verified fully inserted. When attempting to place RHR "A" in shutdown cooling, from the Remote Shutdown Panel RHR "A" shutdown cooling suction valve E12-F009 failed to open. The SRO has elected to use LPCI "A" as an alternate means of shutdown cooling and has directed opening of breakers 52-1C71105 (CB5A), 52-IC71107 (CB7A), 52-1C71205 (CB5B), and 52-1C71207 (CB7B).
Which one of the following describes the effect of opening the above breakers? l A. Opening all four breakers will cause a reactor scram.
B. Opening all four breakers will remove all interlocks associated with the LPCI "A" injection valve (E12-F042A).
! C. Opening all four breakers will close all MSIV's.
D. Opening all four breakers will cause an Auxiliary Building and Containment isolation.
QUESTION SRO 81 l NRC RECORD # WRI 151 ANSWER: C. SYSTEM # K/A 295016 AA1.04: 3.2 ADMIN Emergency Proc./ Plan S LP# GG-1-LP-RO-C7100.00 # OBJ. 18 SRO TIER 1 GROUP 1/ RO TIER 3 GROUP REFERENCE: E-1160 NEW CLASS 05-1-02-11-1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH l l 3/19/98 _ _ - - - _ - _ - _ - - - - - - - _ - _ - - - _ - - - - - - - - - - - - - - - - - - _ - - - - - _ - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - _ u
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 82 The plant is shutdown in Mode 4 preparing to enter Mode 5 to refuel the reactor.
Which one of the following best describes the method used to prevent an inadvertent l dump of the Upper Containment Pools while refueling evolutions are in progress? l A. Place the SPMU System Mode Select in OFF, SPMU Dump Test Switch in OFF, and red tag the SPMU Makeup valves in the closed position.
B. Place the SPMU System Mode Select in OFF, SPMU Dump Test Switch in TEST, and red tag the SPMU Makeup valves in the closed position. : C. Place the SPMU System Mode Select in AUTO and SPMU Dump Test Switch in ! TEST.
D. Place the SPMU System Mode Select in AUTO, SPMU Dump Test Switch in ! OFF, and red tag the SPMU Makeup valves in the closed position.
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QUESTION SRO 82 l NRC RECORD # WRI 114 ANSWER: A. SYSTEM # E30; K/A 295023 AKl.01: 4.1 ; IOI- Refueling AA2.02: 3.7 LP# GG-1-LP-RO-E3000.00 ' OBJ 10, 11 SRO TIER 1 GROUP 1/ RO TIER GROUP REFERENCE: 03-1-01-5 sect. 5.19 NEW CLASS 04-1-01-E30-1 MODIFIED BANK DIFF 3 sect. 3.2 & 5.2 DATE USED: RO SRO BOTH CFR 41.7/41.12/43.7
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 83 A fire has occurred at the Hydrogen Injection Skid on the Unit II side of the GGNS Site.
WLBT News covered the event and interviewed members of the Site Fire Brigade. The fire was extinguished in 8 minutes by the Site Fire Brigade.
Which one of the following describes the deportability of this event to the NRC7 A. This is not a reportable event since it did not occur inside the protected area.
B. Within one (1) hour.
C. Within four (4) hours.
D. Within twenty-four (24) hours.
QUESTION 83 l NRC RECORD # WRI 150 ANSWER: C. SYSTEM # K/A 600000 AK3.04: 3.4 Incident Reports LP# GG-1-LP-RO-PROC.00 OB 15 J SRO TIER 1 GROUP 2/ RO TIER GROUP J REFERENC 10-S-03-2 NEW CLASS E: sect. 6.2.3 MODIFIED BANK DIFF 4 01-S-06-5 DATE Att IIISectIIL7 RO SRO BOTH CFR 41.10/43.5 USED: i 3/19/98
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U. S. NUCLEAR REGULATORY COMMISSION
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WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 84 The reactor is operating at 100 % power and you have three (3) Senior Reactor Operators (Shift Superintendent, Shift Supervisor /STA, Plant Supervisor), and three (3) licensed Reactor Operators on shift in the Control Room. Four hours aRer your crew has relieved the shift, two (2) of the Reactor Operators and the Shift Superintendent fall ill with apparent food poisoning. The EMTs recommend sending the personnel to the hospital.
Which one of the following is the allowed actions for this situation? A. You cannot allow the personnel to leave the site because you will not meet Technical Specification manning requirements.
B. You may let the two (2) Reactor Operators leave, however, the Shift Superintendent cannot leave.
C. Allow the transport of all of the sick personnel to the hospital only after the unit has been placed in Mode 3 when stafling requirements are less.
D. Allow the transport of all of the sick personnel to the hospital and call in personnel to meet the minimum stafling requirements such that you are not less than minimum for more than two hours.
QUESTION SRO 84 l NRC RECORD # WRI 118 ANSWER: D. SYSTEM # K/A G2.1.4: 3.4 ADMIN Generic LP# GG-1-LP-RO-PROC.00 OBJ. 12U SRO TIER 3 GROUP / RO TIER GROUP REFERENCE: 01-S-06-2 sect. 6.5.la NEW CLASS Tech Specs 7.2.1 & 7.2.2 MODIFIED BANK DIFF 3 annual exam admin 04 DATE USED: RO SRO BOTH CFR 41.10/43.2 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 85 The plant is in a LOCA with the reactor water level at - 220 inches. Systems are being aligned to feed the reactor and emergency depressurization is in progress.
Drywell Pressure is at 15.8 psig.
Which of the following situations would require the initiation of Containment Sprays?
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A. Containment Pressure is 2.5 psig Containment Hydrogen Concentration is 6.2 % Containment Temperature is 165 *F B. Containment Pressure is 4.8 psig Containment Hydrogen Concentration is 9.2 % Containment Temperature is 175 *F C. Containment Pressure is 4.8 psig Containment Hydrogen Concentration is 6.2 % Containment Temperature is 175 *F D. Containment Pressure is 2.5 psig Containment Hydrogen Concentration is 4.2 % Containment Temperature is 185 *F QUESTION SRO 85 l NRC RECORD # WRI 119 ANSWER: B. SYSTEM # E61; K/A 500000 EA1.05: 3.3 EOP - 3 H2 Cont. EK2.06: 3.4 LP# OP-LO-EP-LP-005-03 A2.10: 2.7/3.0 OBJ. 2 SRO TIER 1 GROUP 1/ RO TIER GROUP REFERENCE: 05-1-01-EP-3 step 74 -76 NEW CLASS Figure 5 MODIFIED BANK DIFF 3
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DATE USED: RO SRO BOTH CFR 41.10/43.5 1 I
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 86 i A scram has occurred and both battery chargers supplying IIDA are inoperable. All control rods are inserted. RCIC is maintaining vessel level at approximately 30 inches following an j automatic initiation on low vessel level. The SRO has directed an RO to secure the RCIC gland i seal compressor from P601 to minimize DC loads.
No Operator actions have been taken with regard to RCIC operation.
Which one of the following best describes the method to secure the RCIC gland seal compressor from P60l? A. The RCIC Steam Supply Valve (E51-F045) must be closed, then place the handswitch for the RCIC gland seal compressor to STOP.
B. Place the handswitch for the RCIC gland seal compressor to STOP.
C. Depress the Div I LSS Reset pushbutton, then place the handswitch for the RCIC gland seal compressor to STOP.
D. Depress the RCIC Initiation Reset pushbutton, then place the handswitch for the RCIC gland seal compressor to STOP.
' i QUESTION SRO 86 l NRC RECORD # WRI 152 ANSWER: D. SYSTEM # K/A 295003 AK2.01: 3.2 ADMIN Emergency Proc / Plan )
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LP# GG-1-LP-RO-E5100.00 OBJ. 8n SRO TIER 1 GROUP 1 / RO TIER GROUP l REFERENCE: 04-1-01-E51-1 NEW CLASS 1 E-1185 MODIFIED BANK l DIFF 2 DATE USED: RO SRO BOTH l , a
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 87 The Control Room Operator has a tagout which requires verification.
Under which one of the following conditions can the Shift Superintendent waive Independent Verification? A. lineup on the Instrument Air Header Auxiliary Building Automatic Bleed off valve 8 foot off the floor in area 10, 166 ft elevation.
B. a Red Tag to be hung on a Main Steam Drain Valve on the HP Main Steam Stop Valve at 100 % Power.
C. a Temporary Alteration on the Division III Diesel Air Start Header.
D. a procedure step for lineup restoration following the Load Shedding and Sequencing Monthly surveillance.
QUESTION SRO 87 l NRC RECORD # WRI 127 ANSWER: B. SYSTEM # K/A Generic G2.3.2: 2.9 ADMIN Rad Con G2.2.13: 3.8 LP# GG-1-LP-RO-PROC.00 OBJ. 23E SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-1 sect. 6.1.13 NEW CLASS j 01-S-06-29 sect. 6.4.1 MODIFIED BANK DIFF 2 i DATE USED: RO SRO BOTH CFR 41.12/43.4 I I I i l f 3/19/98 i I t
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 88 The plant is in Mode 5 with the reactor head removed.
The Reactor Steam Separator is being prepared for removal from the reactor.
Which one of the following personnel must be notified prior to the lifting of the Steam Separator from the Reactor Vessel? A. ReactorEngineering Supervisor B. Refueling Floor SRO C. Shift Superintendent D. Refueling Outage Director QUESTION SRO 88 l NRC RECORD # WRI 129 ANSWER: C. SYSTEM # K/A Generic G2.2.26: 3.7 ADMIN G2.1.2: 4.0 Equip Control LP# GG-1-LP-RF-F1105.02 OBJ. Se SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-2 sect. 6.7.6 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 43.7 l
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 89 Select the individual who may be allowed to adjust the Reactor Recirculation Flow Control Valve position when increasing power to full power WITHOUT direct supervision.
A. The Reactor Engineer assigned to the shift to assist in bringing the unit to full power.
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B. The Shift Engineer, provided the knowledge and consent of the Operator-at-the- ] Controlsis obtained. j j C. An unlicensed individual in the RO Training Program.
D. The Control Room Operator assisting the Operator-at-the-Controls.
~ QUESTION SRO 89 l NRC RECORD # WRI 130 ANSWER: D. SYSTEM # K/A Generic G2.2.2: 3.5 ADMIN G2.2.1: 3.6 Equip Control LP# GG-1-LP-RO-PROC.00 OBJ. 12B.1& S SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-2 NEW CLASS sect. 5.1; 6.4.5; 6.4.6 MODIFIED BANK DIFF 2 LOT 7/95 admin i DATE USED: RO SRO BOTH CFR 43.2 l I I 3/19/98 L.
- _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ - - _ _ _ _ _ _ - - _ _ _ _ _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 90 Which of the following statements describes an acceptable means of" red tagging" a system or component per the Protective Tagging Procedure? A. Tagging a remote handswitch is adequate if a component is being tagged for personnel protection.
B. When a motor-operated valve is to be a fluid system boundary point, the power supply is tagged, but the local operator need NOT be tagged.
C. An air-operated valve that fails open will not be considered closed for clearance purposes unless it is jacked closed with an installed jacking desice.
D. An air-operated valve that fails open will be considered closed for clearance purposes if, once it is closed, its air supply is opened and appropriately tagged to maintain the air supply to the valve.
QUESTION SRO 90 l NRC RECORD # WRI 131 ANSWER: C. SYSTEM # K/A Generic G2.2.13: 3.8 ADMIN Equip Control LP# GG-1-LP-RO-PROC.00 OBJ. IIII SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-1 sect. 6.2.1h NEW CLASS MODIFIED RANK DIFF 2 LOT 7/95 admin DATE USED: RO SRO BOTH CFR 41.10
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U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 91 The plant is operating at 100% power. An electrician reports that the specific gravity of one of the battery cells for Battery 1 A3 is 1.189 and pilot cell specdic gravity is 1.188 (corrected for electrolyte temperature and level).
Select the response below which describes the required actions.
(Tech Specs are attached.)
A. Verify pilot cell's electrolyte level and float voltage is within the table category C limits within I hour and verify battery cell parameters meet the table category C limits within 24 hours and once per 7 days thereafter, and restore battery cell parameters to category A and B limits of the table within 31 days.
B. Verify pilot cell's electrolyte level and float voltage and battery cell parameters meet the table category C limits within 24 hours and once per 7 days thereafter, and restore battery cell parameters to category A and B limits of the table within 31 days.
C. Verify battery cell parameters are within the table category C limits within I hour and once per 8 hours thereafter, and restore battery cell parameters to category A and B limits of the table within 31 days.
D. Declare associated battery inoperable and be in Mode 3 in 12 hours and Mode 4 within 36 hours.
QUESTION SRO 91 l NRC RECORD # WRI 134 ANSWER: A. SYSTEM # K/A Generic G2.1.12: 4.0 , ADMIN G2.1.31: 3.9 Conduct of Ops LP# OP-LO-PB-LP-003-00 OBJ. 3 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs 3.8.6 NEW CLASS Condition A MODIFIED RANK DIFF 4 LOT 7/95 admin DATE USED: RO SRO BOTH CFR 43.2 l f 3/19/98
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i U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIORREACTOR OPERATOR QUESTION 92 . The plant is operating during an emergency. The shift determines that conditions are such that ! there is no appropriate action to be taken which would be in compliance with the station operatinglicense Whose permission at a MINIMUM is required to take the required actions to maintain the plant in a safe condition and when must the NRC be notified of such actions?
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A. The NRC Resident Inspector; notify the NRC within one (1) hour.
B. General Manager-Operations; notify the NRC within thhty (30) days in a written report.. C. Operations Shift Supervisor; notify the NRC within one (1) hour.
D. Licensed Reactor Operator; notify the NRC within thirty (30) days in a written report.
. QUESTION SRO 92 l NRC RECORD # WRI 135 ANSWER: C. SYSTEM # K/A Generic G2.1.1: 3.8 j ADMIN G2.1.2: 4.0 q' Conduct of Ops LP# GG-1-LP-RO-PROC.00 OBJ. 12M & 151 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-06-5 Att. III,1.4 NEW CLASS 01-S-06-2 sect. 6.2.le(4) MODIFIED RANK DIFF 3 LOT 7/95 admin DATE USED: RO SRO BOTH
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- _ - _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 93 Some routine repetitive tasks may be pedormed without a work authorization document provided certain conditions are met.
Which one of the following tasks would require a Work Order prior to its performance? A. Adding oil to the RWCU Pumps.
! B. Tightening the packing on the Instmment Air, Containment Isolation Valve, P53-F003.
C. Changing filters in air handling units for the Fuel Handling Area Fan Coil Unit.
l l D. Changing pens on the Post Accident Recorders in the Control Room.
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QUESTION SRO '93 l NRC RECORD # WRI 136 l ANSWER: B. SYSTEM # K/A Generic G2.2.19: 3.1 ADMIN I Conduct of Ops LP# GG-1-LP-RO-PROC.00 OBJ. 26G SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 01-S-07-1 sect. 6.1.4 NEW CLASS MODIFIED BANK DIFF 3 LOT 7/95 admin RO SRO BOTH
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U. S. NUCLEAR REGULATORY COMMISSION l WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR QUESTION 94
'Ihe plant is operating at 30 % power when chemistry reported to the control room the following results.-
Reactor pII 7.8 Feedwater conductivity 0.2 umho/cm Reactor water conductivity 0.9 umho/cm - Feedwater chlorides 6 ppb Reactor water chlorides 15 ppb Reactor water sulfates 33 ppb Condensate Pump discharge 0.03 umho/cm Which of the following best describes the required actions for these plant conditions?
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A. Restore to within limits within 48 hours or be in mode 3 in 12 hours and be in mode 4 in 36 hours.- B. Restore to within limits within 72 hours or be in mode 2 in 6 hours.
l C. If chemistry remains at these levels for 24 hours, begin a normal plant shutdown and j
' . proceed to cold shutdown as rapidly as operating conditions permit.
D. Inunediately begin plant shutdown and scram the reactor when IOI-2 permits and continue cooldown to Cold Shutdown. ,
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QUESTION SRO 94 l NRC RECORD # WRI 137 ANSWER: C. SYSTEM # K/A Generic G2.4.11: 3.6 ADMIN G2.1.34: 2.9 Emergency Proc / Plan , LP# OP-LO-SYS-LP-N22/P60412 l OBJ. 10 SRO TIER 3 GROUP / RO TIER 3 GROUP l REFERENCE: 05-1-02-V-12 NEW CLASS l Action Table Mode 1 MODIFIED BANK DIFF 3 Tech Specs TRM 6.4.1 LOT 7/95 onep DNrE USED: RO SRO BOTH CFR 41.10/43.5 i
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U. S. NUCLEAR REGULATORY COMMISSION
' WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 95 The plant is operating at 100 % power when the piping elbow on the common discharge of the Turbine Building Cooling Water (TBCW) Pumps blows out.
The standby TBCW Pump starts and Low TBCW Head Tank level and Discharge Pressure alarms are received in the Control Room.
Which one ofthe following describes the actions to be taken in the Control Room? A. Recover level in the TBCW Head Tank by manually opening the makeup bypass valve to the TBCW Head Tank. Reduce power to 60 % power using recire fk,w and control rods, maintaining within power to flow limitations. Reduce loads on TBCW as soon as possible. j B. Recover level in the TBCW Head Tank. Reduce power to 60 % porw using recirc flow and control rods, maintaining within power to flow limitations. When IOI-2 allows, manually scram the Reactor. Reduce loads on TBCW as soon as possible,
,
C. Manually- scram the Reactor and trip the' Main Turb'me. Initiate RCIC and trip . Condensate, Condensate Booster, Reactor Feed, and Heater Drain Pumps. j l D. Manually scram the Reactor and trip the Main Turbine.' Initiate RCIC and when l reactor level and pressure are stable close the MSIVs and shutdown the Condensate l and Feed Systems.
QUESTION SRO 95 l NRC RECORD # WRI 138 l ANSWER: C. SYSTEM # K/A Generic G2.4.24: 3.7 ADMIN G2.4.11: 3.6 Emergency Proc / Plan LP# OP-LO-ONEP-LP-401-04 OBJ. '1 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: 05-1-02-V-2 sect. 2.1 NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10/43.5 I I ! 3/19/98
__ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ U. S. NUCLEAR REGULATORY COMMIS.CION -
- WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 96 A LOCA condition exists and Containment Pressure has increased to 22.3 psig. The Emergency Procedures have directed the spraying of Containment irrespective of Adequate Core Cooling.
Which one of the following is the basis for this order? A. Above this point the Containment Pressure indicators are off-scale high.
B. This is the Maximum Containment Pressure at which the Containment Vent Valves, sized to reject all decay heat, can be opened and closed.
C. This is the Maximum Containment Design Pressure as defined by Tech Spec Plant Design Criteria.
D. At pressures above this the downstream ventilation piping will rupture when the Containment Vent Valves are opened.
QUESTION SRO 96 l NRC RECORD # WRI 139 ANSWER: B. SYSTEM # K/A Generic G2.4.18: 3.6
, ADMIN Emergency Proc /
Plan LP# OP-LOR-EP-LP-007-03 OBJ. 4 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: EP Bases NEW CLASS MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 41.10 l 3/19/98 -
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 97 The plant is shutdown at 150 * F and 0 psig. The Reactor is assembled following a Refueling Outage. The Fuel Handling Crew is shufIling spent fuel bundles from Containment to the Spent Fuel Pool. Startup is expected to occur in 3 days.
Which one of the following best describes the Primary and Secondary Containment requirements for plant conditions? A. Primary Containment is NOT required, and Secondary Containment is required.
B. Primary Containment is required, and Secondary Containment is NOT required.
C. Primary Containment is required, and Secondary Containment is required.
D. Primary Containment is NOT required, and Secondary Containment is NOT required.
QUESTION SRO 97 l NRC RECORD # WRI 140 ANSWER: A. SYSTEM # K/A Generic G2.2.22: 4.1 ADMIN G2.1.22: 3.3 Emergency Proc / G2.1.33: 4.0 Plan LP# OP-LO-PB-LP-003-00 OBJ. 3 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs 3.6.1.1 NEW CLASS 3.6.4.1 MODIFIED BANK DIFF 2 DATE USED: RO SRO BOTH CFR 43.2 l , 3/19/98
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 98 The plant is operating at 100 % power.
The LPCS monthly functional surveillance was found to be late by two (2) weeks. The surveillance has NOT been performed.
Which one of the following will be the actions to be taken by the Shift Supervisor'l A. Direct the Control Room Operator to perform the surveillance, and log the surveillance completion in the Surveillance Logbook. LPCS is operable until the surveillance is complete Sat or Unsat, provided the surveillance is performed within 24 hours of discovery. Initiate the Late Surveillance Notification Form.
B. Direct the Control Room Operator to perform the surveillance. Declare LPCS INOP and complete the appropriate LCO forms and take LCO actions as appropriate. Log the smveillance in the Surveillance Logbook. Initiate the Late Surveillance Notification Form.
C. Declare LPCS INOP and complete the appropriate LCO forms. Initiate the Late Surveillance Notification Form. Enter the applicable Condition for not meeting the LCO. Ifit is past the completion time allowed by the Condition, enter the applicable condition for not meeting the LCO.
D. Direct the Control Room Operator to perform the surveillance. LPCS may remain operable for up to a month past the discovery of the missed surveillance. Litiate the Late Surveillance Notification Form.
QUESTION SRO 98 l NRC RECORD # WRI 141 ANSWER: A. SYSTEM # K/A Generic G2.2.12: 3.4 ADMIN Equipment Cont.
LP# GG-1-LP-RO-PROC.00 OBJ. 17C,D,E LP# OP-LO-PB-L~P-003-00 OBJ. 3 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: Tech Specs SR3.0.3 NEW CLASS 01-S-06-12 sect. 6.4.11 MODIFIED BAhY DIFF 3 DATE USED: RO SRO BOTH CFR 41.10/43.2 3/19/98
- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ . U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 SENIOR REACTOR OPERATOR QUESTION 99
,In which one of the following situations would the Shift Superintendent be responsible to dispatch GONS personnel with regard to his responsibility for protection of the public?
A. The Claiborne County Fire Department calls the Control Room requesting the GGNS Fire Tmck be dispatched to the restaurant on Grand GulfRoad to put out a fire.
B. The Mississippi Department of Wildlife contacts the Control Room during high water conditions on the Mississippi River to enJ operators in the Operations boat to tie off a channel marker that has drifled south of Radial Well # 1 because it is a hazard to navigation.
C. A visitor to the Site Processing Facility has a heart attack, and the EMTs contact the Control Room to transport the visitor to the Claibome County Hospital in the Company Ambulance.
D. 'Ihere has been an ice storm and the Claibome County Highway Department has requested that a crew from GGNS spread sand on Grand Gulf Road.
QUESTION SRO 99 l NRC RECORD # WRI 143 ANSWER: C. SYSTEM # K/A Generic G2.1.2: 4.0 ADMIN G2.1.1: 3.8 Conduct of Ops.
LP# GG-1-LP-RO-PROC.00 OBJ. 12Q SRO TIER 3 GROUP / RO TIER 3 GROUP I REFERENCE: 01-S-06-2 sect. 6.1.2a NEW CLASS 10-S-01-19 MODIFIED BANK DIFF 3 sect. 6.1.2 & 6.3 DATE USED: RO SRO BOTH CFR 41.10
.
3/19/98 l
1
U. S. NUCLEAR REGULATORY COMMISSION WRITTEN EXAMINATION MARCH 1998 l SENIOR REACTOR OPERATOR i
:
i QUESTION 100 '
'Ihe plant is in mode 3. Cooldown is in progress for entry into mode 4. During a control room panel walkdown you discover annunciator MN CNDSR VAC LO BYP sealed in on panel P601.
Which one of the following best desenhs the significance of this alarm?
. A. The alarm indicates that the main turbine low vacuum trip is bypassed.
B. The alarm indicates that the MSIV isolation on low vacuum is bypassed C. The alarm indicates that the suction valve to the inservice steam jet air ejector has been closed.
D. The alarm indicates low vacuum in the main condenser and the need to place mechanicalvacuum pumpsin service.
. QUESTION SRO 100 l NRC RECORD # WRI 153 ANSWER: B. SYSTEM # K/A Generic G2.4.31: 3.4 ADMIN Emergency Proc 1 Plan LP# GG-1-LP-RO-M7100.00 OBJ. 17 SRO TIER 3 GROUP / RO TIER 3 GROUP REFERENCE: E-1160 NEW CLASS 04-1-02-1H13-P601-19A- MODIFIED BANK B1 DIFF 2 DATE USED: RO SRO BOTH , 3/19/98 _ _ _ - _ _ - _ - _ _ - _ - _ _ _ _ _ _ - _ - - _ - _ _ _ _ _ _ - _
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05-1-02-III-1 Revision: 20 Attechm:nt I Page 1 of 4 f
(Initial Water Temperature 140*F)
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17-S-02-40 R0vicion: 103 Attachment IV Page 1 of 1 RYDAM BYDACCTuf: Run TTunYDACCTM(1 f,DDMC LPRM/APRM f*ROSS REFERIMCI RArrTY Runstrn a 59D D G H 59C C B E F * 593 A H(41) C(42) D(43) G(44) 59A F A B E 51D F A B E F A Sic D G H C D G 513 B (35) E (3 6) F (37) SLA A (38) B (39) E (40) H C D G H C 43D G H C D 43C G H C E F A B E F A 438 C(28) D(29) G(30) 43A A H(31) C(32) D (33 ) G(34) B E F A B E 35D B E F A B E F 35C H C D G H C D 355 F (21) A(32) 35A B (23) E(24) F(25) A(26) B(27) D G H C D G H 27D C D G 27C H C D G A B E F 275 A B E G(14) H (15) C(16) D(17) 27A E F G (18) H (19 ) C(20) A B E F A 19D F A B E F A B 19C D G H C D G H 19B B (7) E(8) P (S) A(10) B (11) E (12) 19A H C D G F(13) H C D 11D G H C D G H 11C E F A B E F llB . C (1) D (2) 11A G(3) H(4) C(5) D(6) A B E F A B 10 18 26 34 42 50 58
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Condition Protective Action Recommendation EVACUATE: 2 Miles All Sectors General Emergency Declared and EVACUATE: 5 Miles in Downwind Sectors 10d SHELTER: Remainder of 10 Mile Emergency Planning Zone (EPZ) General Emergency EVACUATE: 2 Miles All Sectors Declared and ADA Dose Projection or EVACUATE: 10 Miles in Downwind Sectors Field Measurement , . and I at 2 5 miles corresponds to SHELTER: Remainder of 10 Mile Emergency 1 Rem TEDE Planning Zone (EPZ) . 0I 5 Rom Thyroid.CDE
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I SR Applicability 3.0
3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LC0 except as provided in SR 3.0.3.
Surveillance do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval < specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a
: "once per . . ." basis, the above frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LC0 not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.
if the Surveillance is not performed within the delay period, the LC0 must immediately be declared not met, and the applicable Condition (s) must be entered.
When the Surveillance is performed within the delay period and the Surveillance is not met, the LC0 must immediately be declared not met, and the applicable Condition (s) must be entered.
l (continued) l l l GRAND GULF 3.0-4 Amendment No. 120
_ _ _ _ - _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - SR Applicability 3.0 3.0 'SR APPLICABILITY (continued) SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be made unless the LC0's Surveillance have been met within their specified Frequency. This provision shall not prevent entry into MODES or other.specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the. unit.
SR 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.
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GRAND GULF ~ 3.0-5 Amendment No. 120
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_ - _ _ _ _ _ - _ _ - _ . ECCS-Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM j
3.5.1 ECCS-Operating , LC0 3.5.1 Each ECCS injection / spray subsystem and the Automatic Depressurization System (ADS) function of eight safety / l relief valves shall be OPERABLE.
i APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection / spray ECCS injection / spray subsystem inoperable. subsystem to OPERABLE status.
. B. High Pressure Core B.1 Verify by 1 hour Spray (HPCS) System administrative means inoperable. RCIC System is OPERABLE when RCIC is required to be OPERABLE.
AND B.2 Restore HPCS System 14 days to OPERABLE status.
I (continued) i GRAND GULF 3.5-1 Amendment No. 120 l
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ECCS-Operating 3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
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C. Two ECCS injection C .'1 Restore one ECCS 72 hours subsystems inoperable. injection / spray subsystem to OPERABLE E status.
One ECCS injection and one ECCS spray subsystem inoperable.
D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, AND B, or C not met.
D.2 Be in MODE 4. 36 hours E. One ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.
F. One ADS valve F.1 Restore ADS valve to 72 hours 1 inoperable. OPERABLE status.
AND @ One low pressure ECCS F.2 Restore low pressure 72 hours injection / spray ECCS injection / spray subsystem inoperable. subsystem to OPERABLE status.
(continued) i l GRAND GULF 3.5-2 Amendment No. 120
_ _ _ _ _ _ _ . _ _ _ - . _ _ _ _ _ _ _ _ _ - _ _ _ _ _- _ _ _ _ _ _ ___ _ _______ _ _ ____ ______ ECCS-Operatir.g 3.5.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME
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G. Two or more ADS valves G.1 Be in MODE 3. 12 hours inoperable.
AND M G.2 Reduce reactor steam 36 hours Required Action and dome pressure to associated Completion s 150 psig.
Time of Condition E or F not met.
H. HPCS and Low Pressure H.1 Enter LC0 3.0.3. Immediately Core Spray (LPCS) Systems inoperable.
M Three or more ECCS injection / spray subsystems inoperable.
M HPCS System and one or more ADS valves inoper_able.
M Two or more ECCS injection / spray subsystems and one or more ADS valves inoperable.
l ! GRAND GULF 3.5-3 Amendment No. 120 N ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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Titis: CondInfato liiyh No.: 05-1-02-V 12 R3 vision: 17 Pa90: 1 Conductivity
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1.0 PURPOSE /DISCUSSIOtt 1.1 Provide instructions concerning high conductivity in condensate, feedwater or reactor water while in Mode 1,2, and 3. For abnormal chemistry in other modes, or for chemistry parameters less than those in the Action Table, refer to 01 S-08 29, EPRI Water Chemistry Guidelinco.
1.2 The water chemistry limits of the reactor. coolant system are established to prevent damage to reactor materials in contact with coolant. The reactor, vessel acts as a large concentrator; therefore, low levels of conductivity or chlorides in Condensate System can result in high conductivity or chloride levels in the reactor. Chloride limits are specified to prevent stress corrosion cracking of stainless steel at elevated temperatures.
Chloride limits vary with power operation of the reactor. Conductivity measurements are required on a continuous basis, since changes in this parameter are an indication of abnormal conditions. When conductivity is within limits, pli, chlorides, and other impurities affecting conductivity must also be within their acceptable limits. The Condensate Cleanup System is designed to reduce conductivity and prevent chlorides from entering reactor vessel.
2.0 IMMEDIATE OPERATOR ACTIONS 2.1 None 3.0 SilDSEOUEffT ACTIO!!S 3.1 Contact Chemistry to take grab samples on condensate pump discharge, feedwater and reactor water as appropriate.
f 3.2 Based on above samples, enter appropriate section of the ACTION TABLE and take action as appropriate.
3.3 Observe Technical Specification Section 3/4.4.4, Reactor Coolant System Chemistry.
4.0 SYMPTOMS 4.1 One or more of the following alarms on Panel 11113-P600: 4.1.1 CNDS PMP DISCl! CNDCT !!I 4.1.2 CNDS DMIN EFL CNDCT llI.
4.1.3 RFP DISCH CNDCT HI 4.1.4 CNDSR TUBE SHEET LEAK !!I 4.1.5 CNDSR TUBE BREAK LEAK HI 4.1.6 RWCU FLTR DMIN INFL CNDCT HI/LO 4.2 One or more of the following alarms on Panel 1H22-P164: 4.2.1 CONDENSER TUDE SHEET LEAYAGE !!IGli 4.2.2 CONDENSER TUBE BREAK LEAKAGE HIGH l ! 4.3 Routine samples indicate increasing reactor coolant conductivity.
> I { . f M:\ TECH, PUB \ REVISION \0NEPV12.7
Titlo: Cond:nnta liigli 600 . 2 @3-8-@8 W-88 12w6eAem: 67/ vege: a
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Conductivity
4.4 Unexplained increase in hotwell level.
4.5 Rapid depletion of condensato domineralizero.
5.0 AtTrotiATIC ACTIOfiS 5.1' tione
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Titlei Cond ncate !!igh Ho.: 05 1-08-W48 moviuion: a mgs: e i Conductivity l l . J l ACTION TABLE FOR HODE 1 C11EMISTRY Chloride or Conductivity (pmho/cm) Sulfate Action (oob) - Cond. Pump Fecdwater Reactor Reactor 1. Determine remaining deep bed Dischg. (CPD) Water Water capacity and evaluato need for replacement.
>.10 p.07 >.3 >5' 2. Investigate RWCU precoat/ , but but but~ backwauh operation. ! cl0.0 <1.0 <20 3. Take' grab samples of reactor l water, feedwater, CST and DST.
Sampic for chloride, culfate and TOC.
4. For indications of.a circ water tubo break: a) Consider reducing power to concentrate chlorido source, b) Monitor circ water tube shoct/ break conductivity cells, c)' Determine which train has break and isolate the train per SOI 04-1-01-N71-1.
Plug broken tubes if possibic. . 5. Investigate changes in Makeup
* .
Water System or Radwaste System operation.
6. If. increasing reactor power, consult with Duty Manager for nunpencion of power increann. , H/A N/A >1.0 >20 1. Continue investigations as
*
but but .above.
<5.0 <100 2. If chemistry remains at these levcis for 24 hours,.bogin e normal plant Shutdown and proceed to Cold Shutdown as rapidly as operating l conditions ponnit .
>10.0 U/A >5.0 >100 1. When chemistry reaches.thesc l 1cvels, immediately begin plant Shutdown. Scram reactor when allowed by 10I-2 and continue to cool down to Cold Shutdown.
2. For condensor tube breaku, consider isolating CST llotwell reject. Isolate feedwater and cool down on CST water, if CST chemistry is satisfactory.
I l b
. H: \TECil_ PUB \REVISIDH\0NEPV12.T
_ _ --_ _ _______ ______ __ __ -.
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i Title: Condensate $Illgh No.: 05 1 03-V-18 movision: 37- Pages -4 Conductivity'
' ACTION TABLE FOR MODE 2/3 CHEMISTRY Chloride or ' conductivity (pmho/cm) Sulfate Action (ppb) . Cond. Pump .Feedwater Reactor Reactor 1. Determine remaining deep bed Dischg. (CPD) Water Water capacity and evaluate need , for replacement.
N/A- >.15 N/A' N/A 2. Investigate RWCU precoat/ backwash operation.
3. Take grab camples of reactor water, feedwater, CST and DST.
Sampic for chloride, sulfate and TOC.
4 For indications of a circ water tube break: a) Consider reducing power to concentrate chloride sourCO.
b) Monitor circ water tube shect/ break conductivity cells, c) Determine which train has break and isolate the train per SOI 04-1-01 N71 1.
Plug broken tubes if possible.
5. Investigate changes in Makeup j Water System or Radwaste System j I operation.
G. If increasing reactor power, j consult with Duty Manager for suspension of power increano. 1 N/A N/A >1.0 >100 1. Continue investigations as but but above, c5.0 <200 2. If chemistry remains at these levels for 24 hours, begin a normal plant Shutdown and proceed to Cold Shutdown as rapidly as operating conditions permit.
>10 U/A >5.0 >200 1. When chemistry reaches these levels, inunediately begin plant Shutdown. Scram reactor when allowed by IOI-2 and continue to cool down to Cold Shutdown.
2. For condenser tube breaks,
.
consider isolating CST llotwell reject. Isolate feedwater and
-
cool down on CST water, if CST l chemintry in natinfactory.
H:\ TECH _ PUB \REVISIOH\0HEPV12.T
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6.4 REACTOR COOLA!rr SYSTEM j 6.4.1 CHEMISTRY-LCO 6.4.1 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 6.4.1-1.
APPLICABILITY: At all times.
ACTIOllS
----~~-~~*****~~********* ........................................ MOTE ------=
LCO 3.0.3 is not applicable.
........................ ..................................................... CONDITIO!! REQUIRED ACTION COMPLETION TIME A. In MODE 1, with the A.1 - Restore to within 72 hours conductivity, chloride limits.
concentration or pH exceeding the-limit specified in-Table 6.4.1 1.
'l B. Required Action A.1 and B.1 Be in Mode 2. 6 hours associated Completion Time not met.
QB Conductivity or chloride concentration exceeds the limit.specified in Table * 6.4.1-1 while in MODE 1 for 2'336 hours in any 365 day period.' _ C. 'In MODE 2 and 3 with the C.1 Restore to within. 40 hours conductivity, chloride limits.
concentration or pit exceeding the. limit specified in Table 6.4.1 1.
(continued)
(
, i~ ' TRM 6.4-1 Rev. 6
_ _ _ _ _ _ _ _ - _ _ - _ - - _ _ _ _ _ _ - _ - _ _ . - _ _ _ _ _ _ - - _ _ _ _ _ _-_ ._ _- - - _ = . - - - - _ - _ _ _ _ _ _ ACTIOliS (continued) I REQUIRED ACTION COMPLETION TIME COffDITIOtt D ,' Required Action C.1 and D.1 Be in Mode 3. 12 hours ascociated Completion Time not met. 6Li2 D.2 De in Mode.4 - ---- -NOTE - -----
' QB Action should be taken ao rapidly as The identification practical within the while in MODE 1, that conductivity exceeds 10 cooldown rate limit if entry into Action pmho/cm at 25'C or in due to chloride concentration conductivity >10 exceedo 0.5 ppm pndio/cm or chloride concentration >0,5 ppm.
..................... 3G hours At all times other than E.1 Restore the *12 hours E.
MODE 1,2 or 3,with the conductivity and pH conductivity or pH to within the limit.
.I exceeding the limit specified in Table 6.4.1 1. . s F. At all times other than F.1 Restore chloride 24 houra MODE 1,2 or 3,with the concentration to chloride concentration within limit.
exceeding the limit specified in Table 6.4.1 1.
Prior to esitering
'G. Required Action F.1 and G.1 Perform an associated Completion engineering MODE 2 or 3.
Time not met. ' evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system.
l
!
i l F 1 i l N G.4-2 Rev. 6 l
- - _ _ - - _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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'
SURVEILLANCE REQUIREMENTS b
..................................... NOTES-------- --------------------------- )
The reactor coolant shall be determined to be within the specified chemistry limit by performance of the following:
..............................................................................
SURVEILLANCE FREQUENCY SR 6.4.1.1 Determine reactor coolant to be within 72 hours the specified chemistry limit by analyzing a sample of the reactor coolant AND for chlorideo. ........norg...... When conductivity in greater than the limit in Table 6.4.1-1.
.................. 8 hours SR 6.4.1.2 Determine reactor coolant to be within 72 houro.
, the specified chemistry limit by ( analyzing a sample of the reactor coolant for conductivity.
NOTE-..-......-...---- 72 houro Not required to be met in Modes 1, 2, 3, 4 or 5 when conductivity is 5 1pmhos/cm AND at 25'C.
....................................-.... .-..--.-NOTE------ When conductivity Determine reactor coolant to be within is greater than SR 6.4.1.3 the specified chemistry limit by the limit in analyzing a sample of the reactor coolant Table 6.4.1-1.
for pH.
B hours SR 6.4.1.4 .......---....-.... NOTE------------------ Not required to be met when obtaining in-i line conductivity measurements per SR j 6.4.1.5
.........................................
Record the conductivity of the reactor Continuously coolant.
TRM 6.4-3 Rev. 39 e-________-___-_____-_______________ _ _ _ _ . . _ . _
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,
I SURVEILLANCE FREQUENCY SR G.4.1.5 -- ----*- ----NOTE----- ------ - ----- ' flot required to be met when the continuoua recording conductivity monitor is operable.
............................................ -i Obtain an in line conductivity measurement. 4 hours in MODE 1, 2 or 3 AND 24 houru
----- ------ -----
SR 6.4.1.6 - ------.----------NOTE tiot required to be met when obtaining in-line conductivity measurements per SR 6 . 4 .1. 5
............................................
Perform a CHANNEL CllECK of the continuous *r days conductivity monitor with an in line flow cell. S At
...... NOTE- -----
When conductivity l is greater than the limit in Table 6.4.1-1..
.................
24 houro l 6.4-4 Rev. 6 _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - _ _ - _ _ _ - - _ _
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3.8 ' ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters Battery cell parameters for the Division 1, 2, and 3
.
. LCO. 3.8.6
. batteries shall be within the limits of Table 3.8.6-1.
When associated DC electrical power subsystems are- required APPLICABILITY: to be OPERABLE.
ACTIONS
-----
'
...............................-------NCTE--------------------------------
Separate Condition entry is allowed for each battery.
.............................................................................. COMPLETION TIME C0tlDITION REQUIRED ACTION A.1 Verify pilot cell's I hour A. One or'more batteries with one or more electrolyte level and
; float voltage meet battery cell parameters not within Table 3.8.6-1 Table 3.8.6-1 Category Category C limits.
A or 8 limits.
AND A.2 Verify battery cell 24 hours
; gameters meet Table 3.8.6 1 AND Category C limits.
thereafter AND Restore battery cell 31 days l A.3 parameters to Category A and B r ' limits of Table 3.8.6-1.
l (continued) i GRAND GULF 3.8-34 Amendment No. 120
=_____--______ _ -_ :
_-_ _ - _ - _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ . I ACTIONS (continued) COMPLETION TIME CONDITION REQUIRED ACTION Declare associated Innediately B. Required Action and B.1 associated Completion battery inoperable.
Time of Condition A not met.
E One or more batteries with average electrolyte temperature of the representative cells
< 60'F.
@ One or more batteries with one or more battery cell parameters not within Category C limits.
, SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY l Verify battery cell parameters meet 7 days !- SR 3.8.6.1 Table 3.8.6-1 Category A limits.
J (continued) , i i GRAND GULF 3.0-35 Amendment No. 120 l f i L_-________-___________.
a.v.v I SURVEILLANCE REQUIREMENTS (continued)- FREQUENCY SURVEILLANCE 92 days SR 3.8.6.2 Verify battery cell parameters meet Table 3.8.6-1 Category 8 limits.
AND Once within 72 hours after battery overcharge
> 150 V 92 days SR 3.8.6.3 Verify average electrolyte temperature of representative cells is 2: 60*F.
i i
. I'
ND GULF' 3.8-35 Amendment No. 120
- _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _
J.u.o
. ' Table 3.8.6-1 (page 1 of 1)
Battery Cell Parameter Requirements CATEGORY A: CATEGORY B: CATEGORY C: LIMITS FOR EACH LIMITS FOR EACH LIMITS FOR EACll DESIGNATED PILOT CONNECTED CELL CONNECTED CELL PARAMETER CELL
> Minimum icyc1 > Minimum level Above top of Electrolyte level plates, and not indication mark, indication mark, and s 1/4 inch overflowing ar.d s 1/4 inch above maximum above maximum levei : ndication leve; : ndication mark;ai mark,ap a 2.13 V d 2.13 V > 2.07 V Float Voltage a 1.195 2 1.190 Not more than Specifi 0.020 below Gravity (b)(c) ' AND average of all connected cells Average of all connected cells AND .
d 1.200 Average of all
)
connected cells a 1.190 I
{
j
'(a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum level during equalizing charges provided it is not overflowing.
(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float
)
charge. l
) (c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits following a battery
, recharge, for. a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each j connected cell shall be measured prior to expiration of the 7 day l l 3}IUWdilQ. 1 l f Amendment no. no l GRAND Gulf 3.8-37
i
Battery Cell Parameters B 3.8.6
,
B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.6 Battery Cell Parameters BASES BACKGROUND This LC0 delineates the limits on electrolyte temperature, level, float voltage, and specific gravity for the DC power source batteries. A discussion of these batteries and their OPERABILITY requirements is provided in the Bases for LC0 3.8.4, "DC Sources-Operating," and LC0 3.8.5, "DC Sources-Shutdown."
APPLICABLE The initial conditions of Design Basis Accident (DBA) and SAFETY ANALYSES transient analyses in UFSAR, Chapter 6 (Ref. 1) and Chapter 15 (Ref. 2), assume Engineered Safety Feature systems are OPERABLE. The DC electrical power subsystems provide normal and emergency DC electrical power for the diesel generators, emergency auxiliaries, and control :.nd switching during all MODES of operation.
I The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and is based
' upon meeting the design basis of the unit. This includes I'
maintaining at least one division of DC sources OPERABLE during accident conditions, in the event of: ,
,
a. An assumed loss of all offsite AC power or all onsite AC power; and b. A worst case single failure.
Since battery cell parameters support the operation of the DC power sources, they satisfy Criterion 3 of the NRC Policy Statement.
i'
.
LCO Battery cell parameters must remain within acceptable limits j to ensure availability of the required DC power to shut down i
'
the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA.
Electrolyte limits are conservatively established, allowing !
continued DC electrical system function even with limits not i !- met. l
~
(corTinued)
i o GRAND GULF B 3.8-65 Revision No. O ! i
i Battery Cell Parameters B 3.8.6 ! ! BASES (continued) l l APPLICABILITY The battery cell parameters are required solely for the support of the associated DC electrical power subsystem.
' Therefore, battery electrolyte is only required when the DC power source is required to be OPERABLE. Refer to the Applicability discussion in Bases for LC0 3.8.4 and LC0 3.8.5.
ACTIONS A.1, A.2. and A.3 With parameters of one or more cells in one or more l batteries not within limits (i.e., Category A limits not l' met, Category B limits not met, or Category A and B limits not met) but within the Category C limits specified in Table 3.8.6-1, the battery is degraded but there is still sufficient capacity to perform the intended function.
' Therefore, the affected battery is not required to be considered inoperable solely as a result of Category A or B limits not met, and continued operation is permitted for a l limited period.
The pilot cell electrolyte level and float voltage are required to be verified to meet Category C limits within Ihour(RequiredActionA.1). This check provides a quick
indication of the status of the remainder of the battery ! cells. One hour provides time to inspect the electrolyte level and to confirm the float vditage of the pilot cell.
One hour is considered a reasonable amount of time to perform the required verification.
Verification that the Category C limits are met (Required Action A.2) provides assurance that, during the time needed to restore the parameters to the Category A and B limits, the battery is still capable of performing its intended function. A period of 24 hours is allowed to complete the initial verification because specific gravity measurements must be obtained for each connected cell. Taking into consideration both the time required to perform the required verification and the assurance that the battery cell parameters are not severely degraded, this time is considered reasonable. The verification is repeated at 7 day intervals until the parameters are restored to Category A and B limits. This periodic verification is consistent with the normal Frequency of pilot cell
. Surveillance. .
(continued)
,
' GRAND GULF B 3.8-66 Revision No. O L
!
b .i.6.0 BASES l
' '
ACTIONS A.1, A.2, and A.3 (continued) Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits. I this time is acceptable for operation prior to declaring the DC batteries inoperable. j
. ' ' ' B .1 When any battery parameter is outside the Category C limit for any connected cell, sufficient capacity to supply the .
maximum expected load requirement is not assured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 60'F, also are cause for imediately declaring the associated DC electrical power subsystem inoperable.
, SURVEILLANCE SR 3.8.6.1 REQUIREMENTS The SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), which recommends regular battery inspections including float voltage, specific gravity, and electrolyte level of pilot cells. The 7 day Frequency ensures that these inspections are performed within that frequency, recomended by IEEE-450 (Ref. 3).
SR 3.8.6.2 < . The quarterly inspection of specific gravity and. voltage is ' consistent with IEEE-450 (Ref. 3). In addition, within 72 hours of a battery overcharge > 150 V, the battery must ' l
.be demonstrated to meet. Category B limits. This inspection ' is also consistent with IEEE-450 (Ref. 3), which recomends i special inspections following a severe discharge or
! overcharge, to ensure that no significant degradation of the l battery occurs as a consequence of such overcharge.
(continued) GRAND GULF B 3.8-67 Revision No. 1
.
.
.
.
/03 r
.. . . . .
B 3.8.6 i
.
BASES SURVEILLANCE SR 3.8.6.3
- 1 REQUIREMENTS (continued) This Surveillance verification that the average temperature of representative cells (every sixth connected cell) is I 2 60*F is consistent with a recommendation of IEEE-450 ' (Ref. 3), which states that the temperature of electrolytes ,
in representative cells should be determined on a quarterly l basis. ! I Lower than normal temperatures act to inhibit or reduce I battery capacity. This SR ensures that the operating I temperatures remain within an acceptable operating range. 1 This limit is' based on manufacturer's recommendations. I Table 3.8.6-1 This table ' delineates the limits on electrolyte level, float voltage, and specific gravity for three different categories. The meaning of each category is discussed below. . Category A defines the normal parameter limit for each designated pilot cell in each battery. The cells selected as pilot cells are those whose level, float voltage, and level, float specific gravity approximate the, state of charge of the entire battery.
The Category A limits specified for electrolyte level are based on manufacturer's recommendations and are consister,t with the guidance in IEEE-450 (Ref. 3), with the extra
'
1/4 inch allowance above the high water level indication for operating margin to account for temperature and charge effects. In addition to this allowance, footnote a to Table 3.8.6-1 permits the e'iectrolyte level to be above the specified maximum level during equalizing charge, provided
-
it is not overflowing. These limits ensure that the plates suffer no physical damage, and that. adequate electron transfer capability is maintained in the event of transient conditions. IEEE-450 (Ref. 3) recomends that electrolyte
, level readings should be made only after the battery has been at float charge for _at least 72 hours.
, ' The Category A limit specified for float' voltage is { a 2.13 V per cell. This value is based on the l ' recommendation of IEEE-450 (Ref. 3), which states that {
{
prolonged operation of cells. below 2.13 V can reduce the j life expectancy of cells.
.
{\
(continued)
GRAND GULF B 3.8-68- ' Revision No. 1 J L .
Battery Cell Parameters
, B 3.8.6 BASES SURVEILLANCE Table 3.8.6-1 (continued)
REQUIREMENTS The Category A limit specified for specific gravity for each
-
pilot cell is 21.195(0.015 below the manufacturer's fully charged nominal specific gravity). This value is characteristic of a charged cell with adequate capacity.
According to IEEE-450 (Ref. 3), the specific gravity readings are based on a temperature of 77'F (25'C).
The specific gravity readings are corrected for actual electrolyte temperature and level. For each 3*F (1.67'C) above 77'F (25'C), 1 point (0.001) is added to the reading; 1 point is subtracted for each 3*F below 77'F. The specific gravity of the electrolyte in a cell increases with a loss of water due to electrolysis or evaporation.- Level correction will be in accordance with manufacturer's recommendations.
Category B defines the normal parameter limits for each connected cell. The term " connected cell" excludes any battery cell that may be jumpered out.
The Category B limits specified for electrolyte level and float voltage are the same as those specified for Category A and have been discussed above. The Category B limit spacified for specific gravity for each connected cell is a 1.190 (0.020 below the manufacturer's fully charged, nominal specific gravity) with the average. of all connected cells a 1.200 (0.010 below the manufacturer's fully charged, nominal specific gravity). These values are based on manufacturer's recommendations. The minimum specific gravity value required for each cell ensures that the effects of a highly charged or newly installed cell do not mask overall degradation of the battery. l Category C defines the limit for each connected cell. These values, although reduced, provide assurance that sufficient capacity exists to perform the intended function and maintain a margin of safety. When any battery parameter is outside the Category C limit, the assurance of sufficient , capacity described above no longer exists, and the battery I must be declared inoperable. { l f
(continued)
GRAND GULF B 3.8-69 Revision No. O
Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE Table 3.8.6-1 (continued) REQUIREMENTS The Category C limit specified for electrolyte level (above the top of the plates and not overflowing) ensures that the plates suffer no physical damage and maintain adequate electron transfer capability. The Category C limit for float voltage is based on IEEE-450 (Ref. 3), which states that a cell voltage of 2.07 V or below, under float conditions and not caused by elevated temperature of the cell, indicates internal cell problems and may require cell replacement.
The Category C limit of average specific gravity (2: 1.190), is based on manufacturer's recommendations (0.020 below the manufacturer's recommended fully charged, nominal specific gravity). In addition to that limit, it is required that
,
the specific gravity for each connected cell must be no less than 0.020 below the average of all connected cells. This limit ensures that the effect of a highly charged or new-cell does not mask overall degradation of the battery.
The footnotes to Table 3.8.6-1 that apply to specific gravity are applicable to Category A, B, and C specific ,*, gravity. . Footnote b in Table 3.8.6-1 requires the above mentioned, correction for electrolyte level and temperature, with the exception that level correction is not required when battery charging current is < 2 amps on float charge. This current provides, in general, an indication of overall battery condition.
Because of specific gravity gradients that are produced during the recharging process, delays of several days may ' occur while waiting for the specific gravity to stabilize.
A stabilized charger current is an acceptable alternative to specific gravity measurement for determining the state of
, .
charge. This phenomenon is discussed in IEEE-450 (Ref. 3).
Footnote c to Table 3.8.6-1 allows the float charge current to be used as an alternate to specific gravity for up to
7 days following a battery recharge. Within 7 days each I connected cell's specific gravity must be measured to confirm the state of charge. Following a minor battery recharge (such as equalizing charge that does not follow a (continued) l GRAND GULF B 3.8-70 Revision No. 0
_ _ - _ - _ _ _ _ _ - - _ _ _ _ . - _ - _ - _ _ - - _ - _ - _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - ____ Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE. Table 3.8.6-1 (continued) REQUIREMENTS deep discharge) specific gravity gradients are not
.significant, and confirming measurements may be made in less than 7 days.
REFERENCES 1. . UFSAR, Chapter 6.
2. UFSAR, Chapter 15.
3. IEEE Standard 450, 1987.
< i l.
l l' l' GRAND GULF B 3.8-71 Revision No. O __.__--______________j
l>rimary Luntainment 3.6.1.1
'- _ '3.6' CONTAINMENT SYSTEMS 1 3.6.1.l. Primary Containment - LC0 '3;6.l'.1 Primary containment-shall be 0PERABLE.
. APPLICABILITY: MODES 1, 2, and 3.
ACTIONS REQUIRED ACTION' COMPLETION TIME CONDITION A. Primary containmer,t A.1 Restore primary 1 hour inoperable. containment to OPERABLE status.
B .- Required Action and' B.1 Be in MODE 3. 12 hours associated Completion _g Time not met. AND B.2 Be in MODE 4. 36 hours <-
} J. , . --
3.6-1 Amendment No. 120 GRAND' GULF l-l t .
-
- - ___-_ - _ _ _ _ _ _ _ _ _ _ _ _ .-_ _ _ _ _ . -_ - _
_ 3.6.4.1 i 3.6' CONTAINMENT SYSTEMS-3.6.4.1' Secondary Containment.
-
'
LCO' 3.6.4.1 The- secondary containment shall. be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the primary or secondary containment, During CORE ALTERATIONS, . During operations with..a potential . for draining the reactor-vessel (0PDRVs).
ACTIONS COMPLETION TIME CONDITION REQUIRED ACTION Secondary containment A.1 Restore secondary 4 hours
< A.
inoperable in MODE 1, -containment to 2, or 3.. OPERABLE status, i Required Action and B.1 Be in MODE 3.- 12-hours B.
associated Completion Time of Condition A AND not met.
B.2 Be in MODE 4. 36 hours I
(continued)
- i L
. f GRAND GULF- 3.6-42 Amendment No. 120 2
l e
'
bcLUlludi'J LUllL d illhiUll L 3.6.k.1 ACTIONS- (continued) l REQUIRED ACTION COMPLETION TIME CONDITION C. Secondary C.1 --------NOTE--------- containment LC0 3.0.3 is not inoperable during applicable.
l
---------------------
' movement of irradiated fuel assemblies in the Innediately primary or secondary Suspend movement of containment, during irradiated fuel CORE ALTERATIONS, or. assemblies in the during OPDRVs. primary and secondary containment.
AND C.2 Suspend CORE Immediately ' ALTERATIONS.
Aji0, C.3 Initiate action to Immediately suspend OPDRVs.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify all auxiliary building'and 31 days SR 3.6.4.1.1 enclosure building equipment hatches and blowout panels are closed and sealed.
3.6.4.1.2 Verify each auxiliary building and 31 days SR enclosure building access door is closed, except when the access opening is being used for entry and exit.
l (continued) ! I , GRAND GULF 3.6-43 Amendment No. 120 - _ _ - _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _
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