IR 05000413/1992302
| ML20128E389 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/02/1992 |
| From: | Aiello R, Ernstes M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20128E375 | List: |
| References | |
| 50-413-92-302, NUDOCS 9212080073 | |
| Download: ML20128E389 (408) | |
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UNITED ST ATES - h!UCLEAR nEGULATOnY COMMISSION
[. ' 0,#N REGION 11 $ $ 101 MARIETTA STRE ET, N.W - ?, f ATL ANTA, GEORGI A 30323 :
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ENCLOSURE 1 EXAMINATION REPORT - 50-413/92-302 facility Licensee: Duke Power Company Facility Name: Catawba Nuclear Station Facility Docket Nos.: 50-413 and 50-414 Examinations were administered at Catawba Nuclear Station near York, South Carolin Chief Examiner: %[/>da 1 /h 2/b RppTdF.Aiello/ Date Signed E
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Approved By: @ 4r I, / ~ /7/12 Michael E. Ernstes, Chief Date Signed Operator Licensing Section 2 Operat>ons Branch Division of Reactor Safety SUMMARY Scope: Written examinations and operating tests were administered to five Reactor Operator (RO) and seven Senior Reactor Operator (SRO) applicants during the: week of October 19, 199 Results: Six SR0s and one R0 passed these examinations. All others faile A weakness was found in many knowledge areas tested on the written examination. Half the applicants failed the written examination (para 4.a).
.9212080073 921202 PDR ADOCK 05000413 V PDR
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_ REPORT DETAILS 1. Persons Contacted W. Barron, Director, Operator Training S. Bradshaw, Shift Operations Manager M. Brady, Shift Supervisor M. Geddie, Operations Superinterdent R. Katalinich, Nuclear Instructor W. McCollum, Catawba Station Manager M. Tuckman, Catawba Site Vice President 2. NRC Examiners _ R. Aiello, Chief Examiner, Region 11 T. Guilfoil, Sonalysts W. McGonegal, Sonalysts K. Parkinson, Sonalysts 3. Other NRC Personnel Attending Exit Interview R. Watkins, Region 11 J. Zeiler, Resident Inspector 4. Discussion Written Examination The results of the written examination were surprisingly poo Four of the five R0 candidates failed. The R0 candidates had an average score of 75.1 percent. The SR0 upgrades, who currently hold an R0 license, had an average score of 86.1 percent. Since approximately 75 percent of the questions were similar between the - two examinations, this suggests that the training department may be ineffective in preparing the candidates for the examinatio One of the seven SR0 candidates also failed the examinatio The written exam identified many weak knowledge areas. There were several questions which were missed by the majority of the candidates. The following lists some of the operator knowledge areas tested by these questions. The question numbers are in []. R0 Written Exam Planned emergency exposure [5] Nonlicensed operator manipulation of controls (13] Reactor Coolant Pump (NCP) operation [19],[66] Feedwater system (CF) operation [32),[48] ECCS valve interlocks [41] Pressurizer heater power supplies [42] Reactor coolant flow detector design [47]
Report Details 2 Spent-Fuel Pool f rstem (KF) response to a blackout [50] Loss of condenser vacuum actions (72] Loss of All AC immediate action basis [73] Digital Rod Position Indication (DRPI) failure [84] Preventing Fuel Water Sterage Tank (FWST) depletion [89] SR0 Written Exam Planned Emergency Exposure [6] Precautions for placing a battery charger in service [35] Reactor Coolant Flow Detector design [45] Spent Fuel Pool System (KF) response to a blackout (48] Control Rod Bank selector switch use (57' Reactor Coolant Pump (NLP) operation [653 Pressurizer PORV leak indication [82] A strength was noted in the quality of the exam prereview. The facility members assigned to prereview the examination took a proactive approach to ensure the questions were valid, technically accurate, and worded using terms familiar to the candidate Operating Examination 4 The candidates used reference material excessively when addressing JPM follow-up questions. This indicated a lack of familiarity with some important topics. Specifically, the candidates were weak in conduct of operations and radiological requirements. Many of the candidates labored through the administration procedure The simulator examinations showed improved communication between the operators, especially during plant transient The candidates' utilization of the Annunciator Response and Emergency Procedures was noteworth Inspection The inspector conducted a review of the licensee's 1992 Licensed Operators Requalification Program Sample Plan -according to NUREG-1021, ES-601, Attachment 2. No weaknesses or deficiencies were identifie The inspector reviewed the licensee's process for tracking licensed operator control evolutions identified in 10 LFR 55.59 (C)(3)(i). This review included a random sampling of individual records to verify compliance. No deficiencies in the tracking process were identified. However, written feedback from the instructors was not consistently provided to the individual operators and their respective Shift Superviso ,
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Repur. Details 3
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The inspector reviewed the facility's program for E0P rewrite during the preparation week. The communications between the procedure writers, the training department and the operations department have been effective in incorporating the concerns of the operator A strength was identified in the development of the " Simulator Management Observation Book." This book catalogs the written comments provided by the Site Vice President, Station Manager, Superintendent of Operations, Director of Operations, and Oaerations line management, who are all required to routinely oaserve the simulator trainin Material Condition of the Plant The label plate for bus lockout relay 86S was missing from the breaker pane " Magic marker" had been used to write the identifying information on the label plate fac The examiner identified a potential safety hazard in the Unit 1 Auxiliary Feedwater (CA) pump room. A wooden scaffolding-pad was improperly positioned. Specifically, the- post was resting on the side of the wooden block, compromising stabilit The kick-plate next to the hatch in the ETA switchgear room was loose at the corners resulting in a potential trip hazar The inspector found one procedure in the simulator to be out of date. The training staff was using Revision 8 of Procedure OP/1/A/6200/01 (Shift and Vent NV Seal Water Injection Filters) in the simulator. The control room copy was Revistan 9. The simulator copy was replaced with the proper revisio No problems were encountered with either security or health physics in-processin . Exit Meeting At the conclusion of the site visit, the examiners ret with representatives of the plant staff, indicated in paragraph 1 above, to , discuss the results of the examinations and inspection findings. The licensee did not identify as proprietary any material- provided to or reviewed by the examiners. The examiners further described trie areas inspected and discussed in detail the inspection findings.
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ENCLOSURE 2 SIMULATOR FIDEllTY REPORT Facility Licensee: Duke Powar Company Facility Docket No.: 50-413 Operating Tests Administered On: October 20 - 24, 1992 This form is to be used only to report observations. These observations, in and of themselves, do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the- simulation facility other than to provide information which-may be used in future evaluations. No licensee action is required solely in responst to these observation No '.imulator fidelity items were identifie . - -- - , w -,n-
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October 20, 1992 i TO: Mr. Mike Ernstes, Chief Examinor SUBJECT: NRC Exam Comments Catawba Nuclear Station The following attached comments and references have been prepared as a result of our post exam review of the CNS NRC administered SRO and RO exams on 10/16/92.. Please take into account these comments in your grading of the examinee's exams to provide an accurate evaluation of their individual responses. We are very pleased at the high quality exam that was administered, Ne enmmend you in your effort If you have any questions, please contact Reggie Kin. ray at (803) 831-311 ' tt/1 4 % H. Barron Director Operations Training Catawba Nuclear Station REK/kkg Attachments cc: D. Wylie E. M. Geddie M. J. Brady
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..- 1 Question: RO #57 and SRO #53
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Which ONE [1] of the following reasons explains.why Component l Cooling (KC) valves.56A and:81B [ND Heat Exchanger Inlets) open , or. Lo FWST level-following Ss (Safety Injection]? (Refer To KC ! drawing below). Prevent ND pump runout in the event of loss of on trai Prevent KC pump runout in the event-of-loss of one trai < Prevent water hammer in-the ND syste Prevent water hammer'in the KC syste . Answer: A Response: KC is opened to ND to provide cooling when pumping hot containment sump water. See attached. No correct answe Recommendation: Delete from exa . _ _ , --- ,
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OP-CN-PSS-KC
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2) Close at 75% l 3) Recirc and Sample Assured Source - RN Chemicals Added 1) Sodium Nitrite, to pick up Oxygen i 2) Sodium Tetraborate (Borax), to maintain a high pH ,, for carbon steel corresion prot.ectio I 3) Sodium Bicarbonate, Buffer p ) Benzo Tri Azule (BET), to protect coppe ) Precautions: a) Wash off any KC water thoroughly, may cause an irritation to the ski b) Immediately wash eyes, will burn the ) Notify chemistry after the makeup to Surge tank and recire is complet Level Indication in C ) Low level in Surge tank (computer point) 50% 2) Low low level a) 34% b) Closes train related non-essential header isolation valve B. KC Pumps Two per train - one train normally runnin . Power Supply - ETA /B Normal Parameters (ISS OBJ #6) Pressure - 100 peig (CR indication) Flow (RO/SRO OBJ #4, #5) 1) CR indication 2) Aux. S/D Panel indication 3) Flow will depend on the components in service a) ND Hx-5000 gpm b) KF Hx-3000 gpm c) other components normally in service-supply 3500 gpm f Runout Flow 1) 49'+0 gpm per pump 2) ALARH AT 4800 gpm - ont, pump 3) High Flow Annunciator for 2 Pumpe - 9400 gpm 4) Since KC trains. are currently aligned together, runout of one train of KC pumps following failure Rev. 13/03-03-92/GFW l '
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OP-CN-P88-KC
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I of the other train could occur. .The logic on_the following KC vive. prevent this possible runouts a) KCs, vive. 1A.-25,: SOA & 538 (AB:hdr) close~on s Sp or Lo FWST lvl. following an 8s.-
b) KC' vive. 3A, ISB, 228B & 230Al-(RB Isol.)- Close on Sp or_Lo FW8T;1vl. following S _
, , , c) -KC vive._56A & 81B(ND Rx inlets) open on Sp- " !
or Lo-FWST Iv1. following an S : This logic applies to both Unit 1 & Unit 2.- Monitor lights for 1 & 2 KC 56-A & 81B on monitor panel Group Minimum Flow 1) 1000 gpm/ pump 2) KC37/40 Auto open at 3150 gpm decreasing and close at 5800 gpm increasing Cool their ovn motor via essential heade . Auto start Train related blackout b.- Ss Controls Control Room Auxiliary Shutdown Panel Location (Unit 1 - Aux. Bldg. 560', Unit.2 -_ Aux. Bldg. 577) KC Hx's (ISS/RO/SRO #2) . Tube Side RN - Easier to clean Shell side KC KC pressure > RN pressure- Tube leak to RN b. - Keeps excessive' Chlorides out of KC , RN flow temp, controlled l 1) KC out 90 F 2) Hi- High Temp. Alarm 120 0 F , l_ Location - Aux. Bldg. 577' The "KC HX A(B) RN Outlet flow-lo" annunciator wa . constantly in alarm-during normal operations in the "KC
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, ! TEMP" mode due to relatively-low heat load. An-NSM was t installed to eliminate this nuisance alarm-by interlocking.
! each annunciator with a Ss signal such that the: alarm will-
-only be'anabled while * Ss signal is present. It also 1 L provides.a 72-second time de?.ay after the1Ss signal'to allow the RN valve to stroke full open.-
Rev. 13/03-03-92/GFW
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, Spe CNS 1573.KC 00-0001 -
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Date: December 2,1991
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signalis generated. Flow will commence only after KC$6A is open, however. The Ss signal overrides aD non safety modulation control. In the non safety configuration, this valve modulates to maintain a desired flow rate. Modulation control is activated from the Control Room by depnssing the ambar ' RESET' push-button (KC 57A ND Hx 'A' Flow Control Ss Reset). . KCS7A fails to the open position when ASP train 'A'is activated. This allows Dow tnrough the ND heat , exchanger if desired by the operators to achieve a controlled cooldows of the Unit. There are no controls for
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this valve on the AS Valve: KC81B Description: KC to Residual Heat Removal Heat Exchanger 'B' Supply Isolation Valve Active: Yes IE Power: Yes ESF: Yes IWV: Category B
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ESF lLition: Open Stroke Time: 60 seconds ESF Response Time: 76 seconds Cont. Iso. Tune: N/A
. - fd KC81B is a motor operated valve which serves to isolate KC Dow to the ND heat exchenger during normal 4N o,erations; divating Dow to the non-essential headers. Since the ND system is required to mitigate the consequences of a large break LOCA, this valve automaticaDy opens to admit cooling flow to the ND heat
,' < ya exchanger 'B' when the following signals have been generated: , s
- - 853 i. to FwST ievei fotiowing a safety injection
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2. Sp (Phase 'B' isolation signal)
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Comrols for KC81B are provided on ASP 'D' and MC11 in the control room which comists of
, 'OPF.N CLOSE' pushbuttons and ' red green'indicatim. lights and give the operator the ability to control this valve as desired to achieve a controlled cooldown c,ithe Unit. When ASP 'B'is activated, this valve receives no auto actuation signals but transfers 'as is'. ~
Valve: KCS2B Description: Residual Heat Removal Heat Exchanger 'B' Control Valve Active: Yes IE Power: Yes ! ESF: Yes JWV: Category B ESF Position: Open Stroke Time: 40 seconds ESF Response Time: 76 seconds Cont. Iso. Time: N/A KC82B is an air operated valve which serves to regulate KC flow to the ND heat exchanger to support Unit shutdown operations. Since the ND system is also required to mitigate the consequences of a krge break LOCA, this valve automatically opens to unisolate the outlet side of the ND heat exchanger 'B' when an Ss signal is generated. Flow will commence only after KC81B is open, however. The Ss signal overrides all non safety modulation control. In the non safety configuration, this valve modulates to maintain a desired flow rate. Modulation control is activated from the Control Room by depressing the amber 'RIESET' push-button (KC 82B ND Hx 'B' Flow Control Ss Reset).
KCS2B fails to the open position when ASP train 'B' is activated. This aBows flow through the ND heat exchanger if desired by the operators to achieve a controlled cooldown of the Unit. There are no controls for this valve on the AS . 20. DI. SIGN BASIS AND CRill:RIA 20-26
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. Spe CNS.1571KC 00 0001
* Date: December 2,1991 e
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Yalve: KCMB Description: Train 'B' Recirculation line isolation -
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Active: Yes IE Power: Yes ESF: No IWV: N/A ESF Position: N/A Stroke Time: N/A , ESF Response Time: N/A Cont. Iso. Time: N/A KCMB is a motor operated valve which serves to isolate the recirculation line to the KC 'B' train surge tank. This valve is normally opened only to mix the contents of the surge tank after chemical manipulation This valve shall have the ability to close during a design basis event to prevent the possible degradation of
'B' train KC due to excessive casential header bypass flow. This valve is remotely controlled only from the T control room by its control switch on MCll which consists of 'OPEN CLOSE' pushbuttons and ' red. green' , indicating lights and receives no signals to auto actuat Valv KC56A
_ Description: KC to Residual Heat Removal Heat Exchanger 'A' Supply Isolation Valve Active: Yes IE Power: Yes ESF: Yes IWV: Category B
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ESF Position: Open Stroke Time: 60 seconds
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ESF Response Time: 76 seconds Cont. Iso. Time: I N/A
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KC56A is a motor operated valve which serves to isolate KC flow to the ND heat exchanger during normal fd . operations; diverting flow to the non-essential headers. Since the ND system is required to mitigate the ff57 consequences of a large break LOCA, this v21ve automatically opens to admit cooling flow to the ND heat
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6N exchanger 'A' when the following signals have been generated: IN 1. Lo FWST level following a safety injection 2. Sp (Phase 'B' isolation signal) Controls for KC56A are provided on ASP 'A' and MCll in the control room which consists of
'OPEN CLOSE' pushbuttons and ' red. green' indicating lights and give the operator the ability to control this valve u desired to achieve a controlled cooldown of the Unit. When ASP 'A'is activated, this valve ( receives no auto actuation signals but transfers 'as is'.
Valve: KC57A l Description: Residual Heat Removal Heat Exchanger 'A' Control Valve ! Active: Yes IE Power: Yes l l , ESF: Yes IWV: Category B ESF Position: Open Stroke Time: 35 seconds ESF Response Time: 76 seconds Cont. Iso. Time: N/A l l KC57A is an air operated valve which serves to regulate KC flow to the ND heat exchanger to suppon Unit shutdown operations. Since the ND system is also required to mitigate the consequences of a large break 1.OCA, this valve automatically opens to unisolate the outlet side of the ND heat exchanger 'A' when an Ss 20. DI. SIGN BASIS AND CRITI:RI A 20-25
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,- Question: SRO #39
.See attachect exam questio Response:-
Answer key is incorrect. The key states that the_ correct-answer is 1, 8, 1, The question Cross-reference document states that the correct answer is 4, 8, 2, , Recommendation: Change answer to 4, 8, 2, _ _ - _ _ _ _ - _ _ - - _ _ - _ - _ . _ _ _ _ - _ _ _ _ _ - - _ .
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QUESTION: 039 (2.00) ,$ 8d de For the ECCS drawing below, match the interlock conditions from Column B that are required to open the valve in Column (NOTE: Items in Column B can only be used once, and only a single answer may occupy one cnswer space.] .
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Column A Column B (VALVE) (INTERLOCK CONDITION)
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..................... NI 147 NC pressure less than 385 psig FW 27 closed FW 55 closed NI 185A ND 28 closed ND 1 or 2 closed ND 36 or 37 closed NI 115 closed NI 136 closed n NORMAL CHARGING LINE Ab NI19 NV 294;' NV A COLD eb -
44 NV 252 LEOS ~ ' O C6 ,
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Og NI10 N CON NS 43 2E3 SPRAY ND 28 Ny a HDR A ' ;' 00 _ lN.'E38 COLD LEOS I' NC N C p NI173 ND26 ND HX A ND26 ND A ND ND HOT I - ,, I Ob
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2 1 $ N1183 h 4 27 COW N1185 FWST
" 32 NV TO FW 27 '
HOT B&C 2-
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=65 135 HX g '"; 4'
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N1178 , ND ,, pw gg A 4~1 . . , ,58 o q,, o SUMP y;gg4 u B b" ND'IA) M N NX ND B
- ND ND
'-44 % -+ NC HOT COLD LEGS ND $9 37 36 CONT. NS 638 ' LEO C SPRAY .u 0; -
' ' ' - Ni A HDR B A ' '; J N1103 B
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N1136 ,_ ;> i N{]16 4 HOT LEGS TO a
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N1100 COLni N1162u 3Ni113 NI B N1 C4- LEGS' 0> N1162 u l'NI150 n ~ N1147 u
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ENCLOSUJE 4 i
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NRC RESOLUTION OF FACIllTi COMMENTS
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SR0 #39 NRC Resolution: Comment accepted. Answer key changed to:
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a) 4, 8 b) 2, 5 SR0 #53 and RO #57 NRC Resolution: Comment accepte Question deleted from both examination :
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! Facility: O Catawba 1 & 2 O Exam Date: 1992/10/16 Knowledge and Ability Record Form 0c -/a o l> r COUNT MATRIX OMU AI 4 J !" " 9c2- J o Z Summarizing Counts by K/A Group for PWR - Senior Reactor Operator Total Plant Wide Generics .................................. 17 K1 K2 K3 K4 K5 K6 Al A2 A3 A4 SG Plant Systems I 2 1 1 1 1 1 1 2 1 2 6 19 Plant Systems II 1 1 2 4 1 1 1 1 2 1 1 16 Plant Systems III 1 0 0 1 0 0 0 0 1 0 1 4 Emergency /Abn I i 1 3 ........ ^6 2 ..... 11 24 Emergency /Abn II 1 0 1 ........ 5 4 ..... 5 16 Emergency /Abn III O O 1 ........ 0 1 .... 1 3 Totals 6 3 8 6 2 2 13 10 4 3 25 ===== Model Total .....................,............ 99 cc
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Kno ledge and Ability Record 1 rm PLANT-WIDE GENERIC RESPONS1BILITIES PWR - Senior Reactor Operator Target: 17 % Actual: 17.2 % K/A Rep Topic Rating _ _ R/S_ 194001A104 Ability to operate the plant phone, paging system, and 3.0/ two-way radio 194001A106 Ability to maintain accurate, clear and concise logs, 3.4/ records, status boards and reports 194001A107 Ability to obtain and interpret statica electrical and 3.5/ mechanical drawings _94001A109 2 Ability to coordinate personnel activities inside the 2.7/ control room 194001A111 Ability to direct personnel activities inside the 2.8/ Cohirol room 194001A116 2 Ability to take actions called for in the Facility 3.1/ Emergency Plan, including (if required) supporting or acting as the Emergency Coordinator 194001K101 2 Knowledge of how to conduct and verify valve lineups 3.6/ K102 Knowledge of tagging and clearance procedures 3.~// K103 Knowledge of 10 CFR 20 and related facility radjation 2.8/ control requirements 194001K104 3 Knowledge of facility ALARA program 3.3/ K105 Knowledge of facility requirements for controlling 4.1/ access to vital / control areas 194001K109 Knowledge of safety procedures related to high 3.4/ pretsure I
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O O Knowledge and Ability Record Form PLANT SYSTEMS PWR - Senior Reactor Operator - 40 %
-Group I-Plant Systems -Target: 19 %' Actual: 19.2 %
001-Control Rod Drive 017 In-Core Temperature 061 Aux./Emer. Feedwater 003 Reactor Coolant Pump 022 Containment Cooling 063 DC Electrical Dis ~004 Chemical & Volume 025 Ice Condense Liqui.c Radwasto 013 E. Saftey Actuation 026 Containment Spray 071 Wast Gas Disposal 014 Rod Position Indi Condensate System 072 Area Radiation Mo Nuclear Instrumen Main Feedwater System K/A Rep Topic Rating __R/S__- 001000G006 Knowledge of bases in technical specifications for 2.9/ limiting conditions for operations and safety limits 001000K103 CRDM . 3.4/ K558 Reason for overlap of control banks 2.7/ G006 Knowledge of bases in technical specifications'for 2.7/ limiting conditions for operations and safety limits 003000K201 RCPS . 3.1/ A401 Boron and control rod reactivity effects 3.8/ A205 CIAS, SIAS 4.1/ A105 Main steam pressure 3.4/ G008 Knowledge of tne annunciator alarms and indications, 2.9/ and use of the response instructions 015000A?03 Verification of proper functioning / operability 3.9/ K406 Reactor trip bypasses 3.9/ K604 Bistables and logic circuits 3.1/ CK101 Plant computer (COLSS) 3.2/ A404 Valves in the CCS 3.1/ K301 Containment 3.8/ G005 Knowledge of limiting corditions for operations and 3.'3/ safety limits 061000A202 Loss of air to steam supply valve 3.2/ G010 Ability to explain and apply all system limits and 3.1/ precautions 0720 BOG 007 Knowledge of purpose and function of major system 2.6/ components and controls I I s
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O Knowledge and Ability Record Form 0: PLANT SYSTEMS PWR - Seni_or Reactor Operator - 40 %
. Group II Plant Systems Target: _ 17 % Actual: 16.. %
002-RCS' 012 RPS 029 CPS 039 MRSS 073 PRM 006 ECCS 016 NNIS 033 SFPCS 055 CARS 075 CIRC 010 PZRPRS 027 CIRS 034 FHES 062 AC 079-SAS 011 PZRLCS 028 HRPS 035 S/GS 064 ED/G 086-FP Containment K/A Rep Topic Rating __R/S__ 002000K403 Venting the RCS 2.9/ K603 Reactor vessel level indication _ 3.1/ G006 Knowledge of bases in technical specifications for 2.9/ limiting conditions for operations and safety limits 00L(30K409 Safety injection valve interlocks 3.8/ A403 PORV and block valves 4.0/ K201 PZR heaters 3.0/ A211 Failure of PZR level instrument-low 3.4/ K501 DNB 3.3/ A301 Automatic selection of NNIS inputs to control systems' 2.9/ K101 RCS 3.4/ A101 Hydrogen concentration 3.4/ K302 Containment entry 2.9/ K303 Spent fuel' temperature 3.0/ A302 Isolation of the MRSS 3.1/3.5-
-064000K411 Automatic load sequencer: safeguaLJs 3.5/ K406 CO} 3.0/ pn. .n '
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_. . - . _ .=_ - _ - _ _ _._ _ ---__..___. _ ___.._- _ _ _ _ .. _ O Knowledge and Ability Record Form O PLANT SYSTEMS PWR - Senior Reactor Operator - 40 % Group III Plant Systems Target: 4 % Actual!. 4.0 % 005 Residual Heat Removal System 045 Main Turbine Generator 007 PZR Relief Tank / Quench 076 Service Water System 008 Component Cooling Water System 078 Instrument Air System 041 Steam Dump System Bypass Control K/A Rep Topic Ratin .__R/ S_ 005000G007 Knowledge of purpose and function of major system 3.3/ components and controls 008000K102 Loads cooled by CCWS 3.3/3.4 008030A304 Automatic actions associated with the CCWS that occur 3.6/ as a result of a safety injection signal 041020K417 Reactor trip 3.7/ ..
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O Knowledge and Ability Record Form O . EMERGENCY PLANT. EVOLUTIONS PWR - Senior Reactor. Operator - 43 % Group I Emergency and Abnormal Plant Evolutions Target: 24 % Actual: 24.2 % 000001 Continuous' Rod Wit Loss of CCW 000059 LRW Release 000003 Dropped Control Rod 000029 ATWS 000067 Plant Fire Onsite 000005 Inoperable / Stuck Rod 000040 Steam Line Rupture 000068- CR Evacuation 000011 Large Break LOCA 000051 Loss of Vacuum 000069 Loss Containment 000015 RCP Motor Malfunction 000055 Blackout 000074 Inadeq. Core Cool 000024 Emergency Boration 000057 Loss of AC Ele High RCS Activity Instrument Bus K/A Rep Topic Rating __R/S__ 000001G007 Ability to explain and apply all system limits and 3.1/ ' precautions 000001G008 Ability to recognize indications for system operating 3.2/ parameters which are entry-level conditions for technica) sp; "ications 000003G005 Knowledge of . annunciator alarms and indications, 3.6/ and use of t w response instructions
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000003G010 Ability to perform without reference to procedures 3.9/ those actions that require immediate operation of system components or controls 000005G005 Knowledge of the annunciator alarms and indications, 3.1/ and use of the response instructions 000011A103 Securing of RCPs 4.0/ A104 ESF actuation system in manual 4.4/ A111 Long-term cooling of core 4.2/ G010 Ability to perform without reference to procedures 4.5/4.5-l those actions that require immediate operation of system components or controls 000015G010 Ability to perform without reference to procedures 3.4/3.4 l those actio s that require immediate operation of ' system components or controls 000015K207 RCP seals 2.9/ A105 The CCWS surge tank, including level control and level 3.1/ , alarms, and radiation alarm 000029G011 Ability to recognize abnormal indications for system 4.4/4.6-operating parameters which are entry-level conditions-for emergency and abnormal operating procedures l 000029K312 Actions contained in EOP for ATWS . 4.4/ G012 Ability to utilize symptom based procedures 3.8/ A202 Conditions requiring reactor and/or turbine trip 3.9/ A101 Manual inverter swapping 3.7/ A102 ARM system 3.3/3.4 ,
'000067A203 Fire alarm 3.3/3.5 L 000067G012 Ability to utilize symptom based procedures 3.4/3.4 L 000074K103 . Processes for~ removing decay heat.from the core 4.5/4.9-
- 000074K304 Tripping RCPs 3.9/4.2-l 000074K311 Guidance contained in EOP for Inadequate Core Cooling 4.0/ G012 Ability to utilize symptom based procedures 2.9/ p. s r
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O O Knowledge and Ability Record Form
~ : EMERGENCY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43 %
Group _II. Emergency and Abnormal Plant Evolutions Target: 16 % Actual: '16.2 %- 000007- Reactor Trip 000027 PZR PCS Malfunction 000054 Loss of MFW 000008 Stuck Relief Value 000032 Loss of SRNI 000058 Loss of DC 000009 Small Break LOCA 000033 Loss of_IRNI 000060 GRW Release 000022 Lcss of RCS Makeup 000037 SG Tube Leak 000061 ARMS Alarm 000025 Loss of Residual Heat 000038 SG Tube Rupture 000065 Loss of IAS K/A Rep Topic Rating- __R/S__ 000007A103 RCS pressure and temperature 4.2/ G010 Ability to perform without reference to procedures 4.2/ those actions that require immediate operation of sy, stem components or contrels 000008K101 Thermodynamics and flow characteristics of open or -3.2/ leaking valves 000009A237 Existence of adequate natural circulation 4.2/ A102 CVCS charging low flow alarm, sensor, and indicator 3.0/ A109 RCP seal iluws, temperatures, prescuras, and 3.2/ vibrations 000025G011 Ability to recognize abnormal indications for system 3.6/ operating parameters which are entry-level conditions for emergency and abnormal operating procedures 000027A218 Operable control channel 3.4/ G008 Ability to recognize indications for system operating 2.8/ parameters which are entry-level conditions for technical specifications 000033A202 Indications of unreliable intermediate-range channel 3.3/ operation 000038K302 Prevention of secondary PORV cycling 4.4/ G012 Ability to utilize symptom iased procedures 3.2/ A101 Crosstic of the affected de bus with the alternate 3.4/ supply 000065A103 Restoration of systems served by instrument air when 2.9/ pressure-is regained
- 000065A201 Cause and effect of low-pressure instrument air alarm 2.9/3.2 l
000065G010 Ability to perform wit? it reference to procedures 3.2/3.3 j those actions that requt immediate operation.of ' system components or controls l - ,
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O O Knowledge and Ability Reward Form EMERGE!1CY PLANT EVOLUTIONS PWR - Senior Reactor Operator - 43 % Group III Emergency and Abnormal Plant Evolutions Target: 3 % Actual: $. 000028 Pressure Level Malfunction 000056 Loss of OffSite Power 000036 Fuel Handling Accident K/A Rep Topic Rating _ R/S_ 000028A202 PZR level as a function of power level or T-ave 3.4/ including interpretation of malfunction 7010 Ability to perform without reference to procedures 3.7/ those actions that require immediate operatic;) of system components or controls 000056K302 Actions contained in EOP for loss of offsite power 4.4/ _ s
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:D, O O U. S. NUCLEAR REGULATORY COMMISSION-SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION -2 CANDIDATE'S NAME:
_ FACILITY: Catawba 1 & 2 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 92/10/16 INSTRUCTIONS TO CANDIDATE: -Use the answer sheets provided to document your answer Staple this cover-cheet on top of the answer sheets. . Points for each question are indicated-in parentheses after the question. The passing grade requires a final grado of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE % __ _ 100.00 % TOTALS FINAL GRADE All work done on'this examination i.s my ow I have neither given nor received aid, i-Candidate's Signature Vl} Q D ~ ] . '
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- SENIO".; REACTOR OPERATO Page 2 ANSWER SHEET ,
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Multiple Choice -(Circle or X your choice) If you char.ge your answer, writo your selection in the blan 'I MULTIPLL CHOICE 023- a b c 6 001 a b c d 024 a b c d i 002 a b c d 025 a b c d
- 003 a b c d 02e a b c d
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004 a b c d 027 a b c d
003 a b c d 028 a b c d 006 a b c d a o c d _ 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d
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010 a 5- c d 033 a b c d 011 a b c d 034 a b c d
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012 a b c d 035 a b c d .
- 013: a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c- d 038 a b c d - 016' a b c d 039 MK2CHING 017 a b c d a .
Rf!Il 9 " [ d 1 018 a b c' d 2 h ; -[l:- d . . sI
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gO GC he , 020 a b c d 2-021 a b c d _ MULTIPLE CHOICE U22 a b c d _,_ 040 a .b- d
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_ _ _ _ _ . O O ; SENIOR REACTOR OPERATOR Pago 3 ; ANSWER S 11 E E T , !
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Multiple choice (circio or X your enoice) ; , If you chango your answer, writo your selection in the b)an a b c d 064 a b c d i 042 a b c d ___ 065 a b c d ,
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043 a b c d 066 a b c d 044 a b c d 067 a b c d
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045 a b n d 068 a b c d 046 a b c d 069 a b c d , 047 a b c d 070 a b c d E 048 a b c d 071 a b- c d 049 a b c d 072 a b c d r
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050 a b c d 073 a b c d 051 a b c d 074 a b c d 05'. a b c d 075 a b c d ' 053 a b c d 076 a b c d _; 054 a b c d ,7 a b c d '
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055 a b c d 078 a b c d 056 a b c d 079 a b c d ___ 057 a b c d 000 a b c d r 058 a 1, c d 081 a b c -d n ., n 9~~p, 059 a b c d ___ 082 a b c d {l , j ,jh
'060- a b c d 083 a b c d / c ,3 p Wi. 2 - t -: e U* Ip &)0-061 --b c d 084 a b- c -d __
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Multiple choice (circle or X your choice) If you change your answer, writo your selection in the blank.
007 a b c d 008 a b c d 089 a b c d 090 a b c d 091 a b c d 092 a b c d 093 a b c d 094 a b c d __ 095 a b c d 096 a b c d 097 a b c d 098 a b c d 099 a b c d _ _ _ , o ?
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_ . _ .. . _ . _ _ _ _ _ _ _ _ _ . ~ _ _ _ _ _ . _ _ _ O O Page 5 ! NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS-During tho administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie you must sign the statement on 2. After the examin the cover tion has been sheet indicating thatcompleted,is the work your own and you have not received or given assistance in completing the examination. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination-room to avoid even the appearance or possibility of cheating.
, 4. Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provided. USE ONLY Tile PAPER PROVIDED AND DO NOT WRITE ON Tile DACK SIDE OF Tile PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing * jour answers on the exumination question pag . Use abbreviations o;;1y if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answe Write it ou . The point value for each question is indicated in parentheses after the question.
' ' 10. Show all calculations, methods, or assumptions used to obtain an answer to any_short answer question . Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that is provided may count against yo For example, if a question is-worth one point and asks for four responsen, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth-0.20 point If one of your five responses is incorrect, 0.20 will be deducted _and your total credit for that question will be 0.80 instead of 1.00 oven though you got the four correct answer . If the intent of a question is' unclear, ask questions of the examiner onl ,
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Page 6 l 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheet In addition, turn in all scrap paper, 15. Ensure all information you wish to have evaluated as part of your arasver is on your answer sheet. Scrap paper will be disposed of immediately following the examination.
16. To_ pass the examination, you must achieve a grade of 80% or greater.
17. There is a time limit of four (4) hours f or completion of the examinatio , 18. When you are done and have turned in your examination, leave the examination area (EXAMIllER WILL DEFIllE Tile AREA) . If you are found in this area while the examination is still in progress, your license may be denied or_ revoke i l V?_(? (; n l * } k . vLh
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. SD110R REACTOR OPERATOR Page 7 ,
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l QUESTIOll: 001 (1.00) If a valve is found out of position during independent verification of 'f its position, which OllE (1) of the following actions must be taken or i parformed?
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a. Immediately notify the operator at the controls (OATC) and reposition the valve with the OATC's permissio & b. Immediately notify the shift supervisor (SS) and reposition the valve with the SS's permission, c. Immediately reposition the valve, then notify the OAT i d. Immediately reposition the valve, then notify S ,
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- OstSTION: 002 (1.00)
Which ONE (1) of the following statements describes the PURPOSE (as defined in OMP 2-18 "Tagout Removal and Restoration (R&R) Procedure") of the Venting Restoration Sheet? a. Explain how to vent equipment prior to maintenanc b. Describe how to vent infrequently operated equipment prior to operation of the equipmen c. Record equipment placed in an "OUT OF NORMAL" condition and to - insure the equipment is returned to its normal conditio d. Document independent verification of vont valve e v-i 'f1
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_ _ O O SENIOR REACTOR OPERATOR Page 9 QUESTION: 003 (1.00) If a safety tag violation has occurred, which ONE (1) of the following statements describes how the safety tag violation is reported? a. A Tag Occurrence Report Form is sent to the Station Manage b. A Tagout/ Removal & Rostoration Record sheet is annotated with a description of the violation and retained with the tagout record c. In accordanca with Technical Specification Section Administrative Controls, In accorc.ance wit; OMP 2-11, " Operating Department Reporting".
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O O : SENIOR REACTOR OPERATOR Page 10 f l
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i QUESTION: 004 (1.00) l Which ONE (1) of the following radiation exposures describes a Duke Power company administrative radiation exposure limit? a. 3000 mRems/ quarter gamma exposure to the whole bod b. 6000 mRems/ quarter gamma exposure to the ski c. 52 MPC-hours internal radiation exposure in any seven day d. 1500 MPC-hours internal radiation exposure por quarte ; e
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O O SENIOR REACTOR OPERATOR Page 11 QUESTION: 005 (1.00) Which ONE [1] of the following statements describes how operators are identified by the use of color coded identification tape while in the bower Containment during outages? a. Yellow tape worn on-the upper portion of the ar b. Red tape worn on the shoulde c. Magenta and yellow tape worn on the ches d. Green tape worn on the upper part of the bod t
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1 QUESTIOlit 006 (1.00) j If all factors are equal, which ONE (1) of the following volunteers chould be selected for a Planned 2mergency Exposure? a. 20 year old man, b. 30 year old woman, c. 40 year old ma d. 50 year old woma ., t i
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QUESTION: 007 (1.00) The following conditions exists
- You are the Shift Supervisor (SS).
- 1t is 0135 in the mornin A 25 year old man has volunteered to remedy or correct a - situation for. life saving purpose i Which ONE [1] of the following is the maximum whole body doso you may authorize the man to roccive while remedying the situation? a. 5 rem b. 25 rom remo, rem . f t _' 4 > 3 M
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l O O ) SENIOR REACTOR OPERATOR Page 14 l l l l QUESTION: 000 (1.00) Which ONE (1) of the following statements describes the control of high .;
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radiation master keys?
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a. Master keys retained in the Radiation Protection Office are issued to security personnel during a Site Assembl b. Master keys are retained in a scaled box in the Shift Supervisor's office and may only be used in an emergency.- c. Shift Supervisor with concurrence of Radiation Protection must approve issue of a master key and only in an emerganc d. Radiation Protection must approve issue of a master key prior to issuanc . h t b b
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O O . SE!1IOR REACN R OPERATOR Page 15 QUESTION: 009 (1.00) , The following plant conditions exists
- The plant is in MOCE Personnel must work in the main steam doghous ,
Which ONE (1) of the follcaing positions must be notified of personnel working in the main steam doghouse? a. Operator at the controls and Shift Superviso b. Operator at the controls and Balance of plant operato c. Individual's work supervisor and Protected Area First Aid Staf d. Individuals's work supervisor and Security Medical Emergency Response Team [MERT).
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QUESTION: 010 (1.00) : During a fire or medical omorgency, which ONE (1) of the following individuals is responsible for controlling Motorola !!T 440 two-way radio communications? < Shift Technical Adviso i
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b. Operator at the Control c. Unit cuperviso d. Shift Superviso .
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O O SENIOR REACTOR OPERATOR Page 17 QUESTION: 011 (1.00) Which ONE (1) of'the following conditions or items should have an entry in the Open Item Summary of the control Room Logbook? a. Any control Room equipment placed in an "Out of Normal" position and not R&R' b. Any OPERABLE automatic valve in an abnormal position and R&R' c. Initiation of Shutdown Request d. Temporary Modifications (Brown Tago).
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QUESTION: 012 (1.00) i l Which ONE (1) of the following statements describes how revisions are donignated on the Control Room copy of electrical olomentary diagrams (CNEE)? j
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a. The diagram will be red-marked to show the details of the modification and stamped " OPS INTERIM".
b. The diagram will be red-marked to show the details of the modification and stamped "AS-BUILT".
c. The revised area will bo outlined in rod ink and the diagram : will be stamped " sed INTERIM AS-BUILT". 4 d. The revised arou will be blacked out and the diagram stamped ,
"SEE OPS INTERIM IN Tile DRAWING SUPPLEMENT FILE." ,
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QUESTION: 013 (1.00) If the Charlotte Dispatcher notifies Catawba Nuclear Station that the Unit Interface controller (UIC) operating limits require resetting, which ONE (1) of the following positions is responsible for resetting the UIC operating limito? a. Nuclear control Operator or Shift Support Technicia b. Operator at the Controls or shift Technical Advisor, c. Shift Operations Manager or Unit operations Manage d. Maintenance Engineering Services or Protective Relay Engineering Grou , f
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O O SENIOR REACTOR OPERATOR Page 20 QUESTION: 014 (1.00) Which ONE [1] of the following individual (s) is authorized to adjust the sound level of annunciators in the control Room? a. Operator at the Control b. Shift Support Technician, c. Nuclear Control Operato d. I&E personnel.
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QUESTION: 015 (1.00) A non-licensed operator (NLO) Who is NOT in an approved license traininy ; class in doing Control Room observation. Which ONE (1) of the following
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control Room equipment or operations may tho.NLO operate or perform .
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-under the direct observation of an actively licensed RO or.SRO? a. Boration of Reactor Coolant Syste I b. Startup of Residual Heat Removal Syste c. Reactor Coolant Pum , d. Turbine Generato ,
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O O Sell 1GR PEACTOR OPERATOR Page 22 QUECT10t's 016 (1.0G) Ch;c- Jr E (1) ni the following statemento describou when follow-up nr. ; ! l cations to the county and State Warning Pointo are requirod? a. Within b minutes of the implementation of any emergencj procedure, b. Within 15 minuten of any change in Protective Action Pocommendnt ions: [ PAlts ) . ' Every 30 mvnutos if an omorgency han boon clasuified an an Alert or higho Every hour if an emergency has been classified as a Sito Area Emergency or higher.
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O O SENIOR REACTOR OPERATOR Pago 23 QUESTION 017 (1.00) . -Which ONE (1) of thu following conditions warrants a Si'ce Assembly? a. Radiation levels in unrestricted arnas of the Auxiliary Building are equal to or grea*.e* than 1 mr/h b. Radiation levels in restricted arcas of the Auxiliary Building are greater than 2 mr/h EMF-41 indicates Auxiliary Building Airborne Radiation Level is g estor than 1,000,000 cp d. Radiation invels in unrestricted aroes of the Auxiliary. Building are equal to oc greater than 1 mr/hr with EMF-41 Indicating Auxiliary Du11 ding Airborns Radiation Level greater than 10,000 cp { fEuD f*'? { '!) im 4a. 1
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O O SENIOR REACTOR OPERATOR Pago 24 QUESTION: 018 (1.00) Which ONE (1) of the following descriptions explains why 125 Vdc and 70 Vdc power supplies are used in the DC Hold Cabinot? a. The 70 Vdc is used to latch the grippers and the 120 Vdc is used to hold the gripper b. The grippers are latchoO by adding the 125 Vdc and 70 Vdc then the 125 Vdc is used to hold the grippors, c. 125 Vdc is used to latch the grippers and 70 Vdc is used to hold the grippors, d. 125 Vdc is used to latch the grippors and as an alternate hold voltage; and 70 Vdc is used to hold the grippers and as an alternato latch voltag , e' I
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I QUESTION: 019 (1.00) i Which ONE (1) of the following reasons explajos why the control banks are overlapped? a. Permits the use of a lower rod insertion limi b. Reduces the number of required control Rod Drive Mechanisms, f c. Reduces the consequences of a continuous rod-withdrawal . accident, t d. Gives a more uniform reactivity addition por rod step,
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SEllIOR REACTOR OPERATOR Page 26 QUESTIOll: 020 (1.00) The following plant conditions exicts
- Unit 2 is in Mode Rod movement tests are in progress in accordance with Technical Specification 4.1.3. All the rods in control bank A, group 1 fail to move due to a control malfunctio Which O!1E (1) of the following statements describes the action required to be taken? [ Technical Specifications 3.1.3.1 and 4.1.3.1.2 are attached)
a. De in llOT STAllDBY within 6 hour Determine that the SilUTDOW!i MARGIll is satisfied within 1 hour and be in llOT STAliDBY within 6 hour Enter Technical Specification 3. d. Continue plant operation while TS 3.1.3.1 action d. la take F ' "' i i ? t- 1 f I,
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O O " ,R SENIOR REACTOR OPER Pago 27 QUESTION: 021 (1.00) Which ONE (1) of the following statements explains Wily cach reactor coolant pump (NCP) has a safety breaker installed in serios with its 6900 V supply breakor? a. Provent damage to the high voltage suppl b. To prevent a mechanical failure of the electrical penetration assembly due to fault current if one protectivo devico fails, c. To provent da.aago to the high voltago (6900 V) supply breaker in , the event of an NCP sheared shaf ; d. To provent damage to the high voltage (6900 V) supply breaker.in the event of an NCP locked roto r
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_ _.__.._____ _._ ___._. . O O ! ' SENIOR REAC7JH OPERATOR Pago 28 j l l-QUFSTION: 022 (1.00)
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t If NC [ Reactor Coolant) System cold log temperature is 2GE degroos_F, which ONE [1' of the following statomonts explains why secondary water temperature must be loss than 315 degrees F prior to starting the NC Pumps? i a. Protect against NC System overpressure caused by onorgy i additions from the secondar b. Protect agair4st a loss of shutdown margin caused by nnergy loss- 1 to the secondar r c. Provent excessive NC System cooldown rato d. Provent loss of pressurizer loyal due to excessivo NC System cooldow , g N .j n
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QUESTIOll: 023 (1 00) The following plant conditions exist
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All cyotems are operabl Control Bank "D" starts stopping in slowly, but at a noticeable rat Which UNE (1) of the following events will cause this response? , n. A lonk has developed in the Regenerative llent Exchange . b. A leak has developed in the Lotdown llont Exchange c. A leak has developed in the tubo bundle of the Seal Water licat , Exchange d. A leak in the Excess Letdown licat Exchange i
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O O SENIOR REACTOR OPERATOR Page 30 t t QUESTION: 024 (1.00) f The following plant conditions exist: [
- Boric Acid Transfer Pump 1A running following an auto start -,
domand from the Reactor Makeup system i
- A spurious So signal wi.s roccived Which ONE [1] of the following operations must be performed to shutdown Bor10 Acid Transfer Pt.mp 1A following roset of the Ss (Safety Injection)
coquencor? Select "OFF" on the pump control switc b. Start the other Borj a Acid Transfer Pum c. Vcrify that the VCT level is above the auto makeup sotpoint and i select "OFF" on the pump control switc . d. Roset the Boric Acid Transfer Pulap and select "OFF" on the pump control switch, r i
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. . - . . . . - . - _ . ~ .O O JENIdR ' P.EACTOR OPERATOR Page 31 QUESTION: 025 (1.00)
If a steam leak on Unit 2 is causing steamline pressure to decrease at_a_ constant rate, which ONE !1) of the following statoacnts explains the plant's response to the steam leak? a. A steam 31ne low pressure satety injection (SI) will occur at an indicated pressure less than the setpoint, b. A steamline low pressure SI will occur at an indicated pressure-greater than the setpoint, c. Since the steamline low pressure SI circuit is rats sensitive, the rate of decrease is multiplied by 10 and an SI will occur when the rate of decrease reaches 100 psi /se d. Since the steamline low pressure SI circuit is rate sensitive, the constant rate of change is multiplied by 10 and an SI will occur at an indicated pressure less than the setpoint or the rate of decrease is greater than 100 psi /sec.
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O O . SENIOR REACTOR OPERATOR Page 32 QUESTION: 026 (1.00) The following plant conditions exist:
-
Unit 2 at 43% powe The Digital Rod Display System "NON-URGENT" Annunciator is in alar No other alarms have bsen receive Which ONE [1] of the following conditions caused the alarm? a. Data B failure, b. All three control units faile c Data A rod position is 180 steps and Data B rod position is 192 steps, d. A rod is stuck.
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O o iSENIOR REACTOR OPERATOR -Page 33 QUESTION: 027' (1.00) Tho'following readings were noted on the Power Range and Intermediate
-
Range Channels: N-35 0 5 x 10E-5 N-36 0 8 x 10E-6 N-41 0 8.5% N-42 0 9% N-43.0 8.5% N-44 0 9% Which ONE [1] of the following-describes the problem indicated by these readings? [See figure below) a. N-35 reading high for current conditions, b. N-36 reading low for current-conditions, c. N-41 and N-43 adjusted improperly during last calorimetri d. N-42 and N-44 adjusted-improperly during last calorimetri DaTTRMEtaAft PowtR
. RANGE RANCE
, 10 . , 200%
' 100 %
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0 O SENIOR REACTOR OPERATOR Page 34 QUESTION: 028 (1.00) The reactor is operating at 13% power when Intermediate Range Nuc)aar Instrument N-35 fails LO Which ONE [1] of the following responses will occur, Reactor Trips on Source Range liigh Flux Level Tri Intermediate Range [IR) High Flux Red Stop (C-1) comes in. . c. Power Above Permission P-10 status light for N-35 goes out, d. Power Above Permissive P-6 status light'for N-35 goes out, s l l l
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SENIOR REACTOR OPERATOR QUESTION: 029 (1.00)- If a Containment entry is made during reactor startup with the - Intermediate Range reading less than 10 E-10 amps, which ONE (1) of the following reasons explains why the Containment Evacuation Alarm cannot be used to warn personnel to evacuate containment? a. High Flux at Shutdown Alarm is blocked, f b. High Flux at Shutdown Alarm only sounds the annunciator in the Control Roo c. Source Range High Level Reactor Trip is blocke d. Alarm actuation may initiate magnetic interference that could cause a spurious reactor trip.
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O O ' SENIOR REACTOR OPERATOR- Page.36 QUESTION: 030 (1.00) Which ONE (1) of the following statements explains how Control Room operators can recognize that an incore thermocouple is being monitored at the Incore Thermocouple Panel? a. The plant computer thermocouple temperature will be indicated by a yellow readin b. The plant computer will show an alarm because the signal to 'hec computer will be disconnecte c. The local control annunciator will alar d. The thermocouple temperature reading will flash on the plant computer scree 'f i tt ; .- % A
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O O SENIOR ~ REACTOR OPERATOR' Page 37 QUESTION: 031 (1.00) If the Lower-Containment Ventilation units are shifted to HIGH speed, which ONE (1) of the following statements describes how the Containment Ventilation (VV) System responds? a. Both the thermostat valve and the cooling water bypass fail close b. The thermostat valve fails closed and the cooling water bypass is throttled to control temperature, c. Both the cooling water bypass valve and thermostatically controlled valve fail ope d. The cooling water bypass fails open to allow the thermostat valve to control on temperatur DFG'7(']} ta < s1
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O O SENIOR-REACTOR OPERATOR Page 38 QUESTION: 032- (1.00) According to Technical Specification Bases,'which ONE-[1] of the . following statements describes the effect of operation with-the Ice Condenser Doors open? In the event of a LOCA, the containment peak pressure transient = may exceed 14.7 psi b. In the event of a LOCA, the containment peak pressure transient will be less than 14.7 psi c. During a LOCA the released Reactor Coolant System Fluid may be diverted away from the ice condenser bay Inadequate sublimation of the ice bed may not occur because of warm air intrusion.
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-QUESTION:R033- (1.00)'
Theffollowing plant conditions exist:
- > Unit 1 t.t 63% powe . .
. .
'NS Pump 1A recirc valve to the FWST is open while performing-a
.
-
test-on NS Pump 1 NS53B;was-shut to perform maintenance on NS Pump 1 Other NS-valves are aligned as indicated in the-NS drawing provided belo Which ONE (1) of the following actions is required? -(Technical Specification 3.6.2.is attached]. a. Perform surveillance:4.6.2.a. on completion of the test on the NS Pump 1 b. Provide an alternate flow path by opening NS 38B and NI 184 c. Enter technical specification 3. d. Verify that NS pump 1A develops a differential pressure head greater than or equal to 185 psi " A n un; w ga om g_ =g
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O O a i SENIOR REACTOR OPERATOR Page 40 ' QUESTION: 034 (1.00)
.Which'ONE (1) of the following statements concerning the Auxiliary Fosdwater System describes the operation of '.he Turbine Driven Pump (TDCA]?
a. If instrument air pressure is lost, the' steam isolation valves-fall open to start the TDC b. If either of the Motor Driven Pump's [MDCA) flow exceeds 780 gpm after a valid start signal, the TDCA will auto start.- c. The TDCA pump will auto start upon loss of both Main Feedwater pump d. The TDCA will automatically start upon an undervoltage signal on one (1) NCP bu Q? ?\ ' 5fb h;;t stur (D U aNMOi/
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O O : SENIOR-REACTOR OPERATOR Page 41
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QUESTION: 035 (1.00)
.Which ONE [1] of the following reasons explains why the.AC input breaker >;
must'be. closed first when placing a-battery charger in service?- a. To prevent exceeding the auctioneering Diode Assembly peak' voltage limit when closing the output breata b. To ensure charging rectifiers are not exposed to an overvoltage condition when closing circuit breakers, c. To protect the output breakers from heavy surge current d. To limit charging current to minimize battery explosion potential,
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O O SENIOR REACTOR OPERATOR Page 42 QUESTION: 036 (1.00)
'
LWhich ONE [1] of the following statements describe the'use of installed sources in Area Radiation Monitors (EMF)? a. NC filters area EMF uses an installed source to verify alarm b. Steam line EMF use an installed source to prevent spurious alarm c. Reactor Building Refueling Bridge EMF uses an installed source
'
to prevent spurious alarm d. Check sources have been removed from Area Radiation Monitors for ALARA concern f('lG
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II i \ O ;O JSENIOR REACTOR OPERATOR Page 43
: QUESTION: 037 (1. 00)
If:all Reactor Coolant (NC) Pumps are off, which ONE (1) of=the-following Reactor Vessel Level Indicating System (RVLIS) indications-is/are valid? a. Train A upper range and Train B D/P Rang b. Train A upper range and Train B lower rang c. Train A D/P range and Train B upper range, d. Train A D/P range and Train A lower rang pp,,3 EE ' , ;d
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SENIOR; REACTOR OPERATOR- Page 44 QUESTION: 038 (1.00)
'Tha following plant conditions exist: - The reactor vessel head is in plac The reactor coolant system has been drained to mid loo The S/G primary side manways have been remove Which ONE (1) of the following statements explains why at least one-hotleg is maintained with no nozzle dam installed?.
a. Allows the hotlegs to be vented to containmen b. Prevents syphoning coolant from the core in the event of a loss-of core coolin c. Vents non-condensible gases that could inhibit natural circulation, Eliminates need for low temperature overpressure protection.
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QUESTION: 039 (2.00) l For the ECCS drawing below, match the interlock conditions from Column B l that are required to open the valve in Column (NOTE: Items in ; Column B can only be used once, and only a single answer may occupy one i answer space.)
Column A Column B
[ VALVE) [ INTERLOCK CONDITION)
_____ . _ __ _________________ NI 147 NC pressure less than 385 psig FW 27 closed FW 55 closed NI 185A ND 28 closed ND 1 or 2 closed ND 36 cr 37 closed NI 115 closed NI 136 closed a NORMAL CHARGING LINE A (-- N!19 NV 294 ': NV A COLD 8b-~ 4(- M NV 252 O LEGS -* Ch o 4 - Dp NI10 NV CONT,
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NS 43 253 SPRAY ND ,28 NV B gi 334 HDR A (); " ) ; --- COLD LEGS -)4 NC N N ND26 ND HX A ND 25 ND A ND ND HOT 1 -,, t C p NI173 .,
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LEO 3, < >3 D (-- "NDi r h d ND - 2 1 B 3[j'3 NI183 'W 24 " " 27 COM NI185 FW 27 ' l FWST
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A p N1178 U ND u ,, NI184 FW 55 ' p", , ,58 n n SUMP n/ ND ND B p ND 60
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NO3 ' '; j NC HOT COLD LEGS ND HX B 37 36 LEG C CONT. gg $33 ND 59 HDR W NI A M N1103 A i
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9 9 R SENIOR REACTOR OPERATOR- Page 46 y e i. - r-. QUESTION: 040 (1.00) Which ONE (1) of the-following statements describes the technical Specification BASES for removal of power to the accumulator power- ) operated isolation valves? a. The valve motor operators have a history of overheatin b. Ensures that the' safety analysis assumptions used for accumulator pressure and volume are me c. The-valves fail'to meet single failure _ criteri d. Valve stroking time may exceed the accident analysis value p;i 7! p r' p p .
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' SENIOR REACTOR OPERATOR Page 47-t-QUESTION: 041 (1.00)
Which ONE (1) of the following-identifies the power supplies for Pressurizer lleater Group A? a. LXD and CD b. LXH and CD , LXH and CD d. LXI and CD , i f ,- f* Ji "r -g J
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O: O-SENIOR REACTOR OPERATOR Page 48- , _ QUESTION: 042 (1.00) The;following plant conditions exist: .
- Pressurizer PORV isolation valve NC-31B is-shut to isolate leaking PORV NC-32B - PORV isolation valve NC-33A is open and PORV NC-34A is shu PORV isolation valve NC-35B remained open when its control switch was placed in close to isolate a leaking PORV NC-36B Which ONE [1] of the following actions should be taken to SHUT PORV isolation valve NC-35B?
a.. Manually close NC-35B circuit breaker on CD b. Manually close NC-35B circuit breaker on EMX c. Position NC-35B Control Switch to " OVERRIDE" positio Position NC-35B Control Switch at the ASP to "CLOSE".
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SENIOR REACTOR OPERATOR .Page 49 QUESTION: 043 (1.00) The following plant conditions exist:
- Unit 1 is at 100% powe All control systems are in automati Pressurizer Level Contral Switch is in position 1- Pressurizer Level Channel II-fails ID Without operator intervention, which ONE (1) of the following plant responses will occur?
a. Letdown isolates, Pressurizer Heaters shut of b. No effect on charging flow, no Letdown _ Isolation, Pressurizer Heaters shut of c. Charging flow to maximum, 'o Letdown Isolation, all Pressurizer Heaters shut of d. Charging flow to maximum, Letdown Isolates, Pressurizer Heaters shut of E O
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SENIOR REACTOR OPERATOR Page 50 QUESTION: 044 (1.00) The following plant conditions exist
- Delta T is 100% ~ Over Temperature [OT) Delta T setpoint is 138%
,
- Pressurizer pressure is decreasing due to a small Reactor Coolant System leak Which ONE (1) of the following responses describes the OT Delta T setpoint relationship to Actual Delta T? Setpoint increases causing the difference between the setpoint and actual Delta T to increas Setpoint decreaues causing the difference between the setpoint and actual Delta T to decreas , Actual Delta T increases and the setpoint decrease causing the difference between setpoint and actual Delta T to increas Actual Delta T decreases and the setpoint increases causing the difference between setpoint and actual Delta T to decreas ;
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n O oO SENIOR PEACTOR OPEkATOR Page 51 QUESTION: 045 (1.00) Which ONE [1] of the following statements explLins why the outer Reactor Coolant Flew Detector is offset by 30 degreec along the pipe bend? (See-the Reactor Coolant Flow Detector belov) a. To minimize detector output for reverse flow condition b. To maximize detector output for normal flow condition c. To minimize detector damage by flow erosio d. To maximize detector output during natural circulation, s 15 i 15- /
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SENIOR REACTOR OPERATOR Page 52 QUESTION: 046 (1.00) Which ONE (1) of the following conditions will cause all Steam Generators (S/Gs) on the respective Unit to swap to manual control?' a. Unit 1 S/G "A" Narrow Range Level instrument fai's low and S/G
"A" Wide Range instrument fails high, b. Unit 1 Power Range Instrument N41 fails low and S/r 'A" Feedwater F'ow Median fails low, -
c. Unit 2 Power Range Instruments N41 and N42 fail lo d. Unit 2 Feedwater Temperature Median fails hig MASTEit 30PY
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Q-SENIOR REACTOR OPERATOR-. Page 53 QUESTION: 047 (1.00) If-the Hydrogen Recor.biners are not avai]able following a LOCA,_which ONE (1) of the followin~ containment Hydrogen concentrations will be exceeded? a. 3.5% by weight, b. 4% by volum cc/k cc/k *s ';;m m e?) he
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O O SENIOR REACTOR OPERATOR- Page 54
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QUESTION: 048 (1.00) <Which ONE'[1] of t!/. following statements describes the response of the Spent Fuel Pool Cooling System (KF) following a Blackout? [ NOTE: Assume No operator action is taken) a. Spent Fuel Pool level and temperature will increas Spent Fuel Cooling Pumps restart automatically but the KF' Skimmer Pump must be manually restarte c. Makeup to the Spent Fuel Pool from the--Fueling Water Storage-Tank (FWST) will be isolate d. Spent Fuel Pool level will increase because the KF makeup valves (KP-101B and KF-103A) automatically open,
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SENIOR REACTOR OPERATOR Page 55.
' QUESTION: 049
- (1.00)
If.the Main Steam Isolation Valves ~ [MSIVs] on Unit 2 are opened before pressure is completely equalized, which ONE (1) of the following responses may occur? a. Reactor Trip due to High Pressurizer Leve Safety Injection due to Steam Generator swel c. Main Steam Line Isolation due to high steam line pressure rat Steam Generator PORVs lifting due to moisture in the Main Steam Lines flashing to stea MASTER CO?Y
o a q SENIOR REACTOR OPERATOR Page_56-QUELTION: 050 (1.00) The following plant conditions exist:
-
A Blackout has occurra A LOCA occurs 13 seconds after the Blackou Which ONE [1] of the following m atements describes the Emergency Diesel ' Generator Sequencer operation? a. Sequencer resets; sheds all loads; and then initiates the LOCA sequenc b. Sheds all loads; starts the accelerated sequence followed by the committed sequenc c. Sheds non-LOCA loads and continues starting-LOCA loads in the LOCA sequenc d. Sequencer resets; sheds non LOCA loads; starts the accelerated sequence and the committed sequence.
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O O SENIOR REACTOR OPERATOR -Page 57 QUESTION: 051 (1. 00) Which ONE (1) of the following statements explains how CO2 pressure is maintained in the Diesel Generator Low Pressure CO2 storage tank? a. The refrigeration package is manually started at 295 psig tank pressure to maintain the CO2 in a liquid stat b. The refrigeration package is manually started at 295 psig tank pressure to maintain the CO2 in a gaseous state, c. The refrigeration package is automatically started at 305 psig tank pressure to maintain the CO2 in a liquid stat d. The refrigeration package is automatically started at 305 psig tank pressure to maintain the CO2 in a gaseous state.
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, L) 'L) SENIOR REACTOR OPERATOR Page 58 QUESTION: 052 (1.00) Containment Sump Isolations NI-184B and NI-18SA automatically OPEN following a Safety Injection [Ss) with 2/4 Low FWST [Pueling Water Storage Tank) level [37%). IlOW and WHY is the auto swap over blocked?
(Refer to ND drawing below)
a. Pressing the " Defeat" pushbutton to prevent the ND pumps from losing suction during a LOCA outside containmen Placing the valve control switch in Pull To Lock to prevent loss of ND Pump suction if the swap over occurs prematurel c. Go to the CLOSE position at the ASP to prevent inadvertent draining of the FWST to the Containment Sum Go to the CLOSE position at the SSF if the valves have spuriously opened due to a fire 3n the Control Roo LOOP B HOT LEG
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'SEH10R REACTOR OPERATOR Page 59 QUESTION: 053 (1.00)
Which ONE (1) of the following reasona explains why.Corponent Cooling
[KC) valves SGA and 81B (HD Heat Exchanger Inlets) open on Lo FWST level following So (Safety Inject; ion)? (Refer To XC drawing below) j l
a. Provent HD pump runout in the event of loss of one trai l b. Provent KC pump runout in the event of loss of uno trai ,
!
I c. Provent water hammer in the HD Syste d. Provent water hammer in the Kc syste KC SURGE TANK A@4 yKC SURGE TANK B - KC SUMP PUMP DISCHO RX BLDG ' SUPPLY HDR - 61A T T 540_
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SUPPLY ES PUMP MOTOR EMF [1 I OA 22b} s m HDR ES PUMP ( I' MOTOR COOLERS 9 9 n, EMF COOLERS
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'
KC SUMP PUMP DISCiG RX DLDG - ~g SUPPLY HDR > SfA T T 540 . .. AUX DLOG
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f O J X SUPPLY
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I OA 22 D, HDP! ES PUMP hb EUU EMF [] .
(JOTOR f MOTOR COOLERS EMF COOLERS * ^ ' , :: : :- , = ] ,, 56A 0-X f SOA 530 f X-0 e
810 (ND) MECH MECH IND) f
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... . . .. O O SENIOR REACTOR OPERATOR Page 61
QUESTION: 055 (1.00) Which ONE (1) of the following statomonts describos a purpose of the Containment Purgo (VP) System? a. Reduce llydrogen concentration in Containment following a 1.nC b. Docrease Containment humliity to within acceptable limits for proper operation of refueling instrument c. Provido additional cooling to upper Containment during _ refuelin d. Roduce fission product concentrations in conte.inmont atmosphere to acceptable limits for personnel access.
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- . .
_ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ . _ - _ _ _ - - _ -
. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . . O O SEli10F, REACTOR OPERATolt Page 62 QUESTION: 056 (1.00)
The following plant conditions exist:
-
A reactor trip has occurred from 80% power on Unit Reactor Trip Breaker "B" failed to ope Which OllE (1) of the following statements describes how the Steam Dump control System will respond?
?lant Trip Load Rejection Condenser Atmospheric Controller Controller Dumps Dumps a. Enabled Enabled Armed Armed b. Enabled Hot Enabled liot Armed Armed c. Ilot Enabled Enabled Armed Not Armed d. Not Enabled flot Enabled flot Armed 11ot Armed
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. _ . . .__ _ . - _ ___ _ _ ._ . . . _. , . . . . __. . . _ . . _ _ _ _ .
l SEllIOR REACTOR OPERATOR Page 63 l
QUESTIO!1 057 (1.00) Which OllE (1) of the following describes when the Individual control Bank Positions on the Bank Selector Switch may be used? a. During an emergency and test operation b. When less than 15% power during reactor startup or shutdown, c. During normal operatien greater than 15% powe d. When adjusting rod height to compensate'for a dropped rod and manually shutting down the reacto , MASTER 90PY
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_ . . _ _ _ . _ _ .-_ _.___._ _._ - -___ _ O O . l SENIOR REACTOR OPERATOR Page 64 QUESTIONt 058 (1.00) Control Rod F8 has been datormined to be trippable but inoperable due to an open in the lift coil circuit but it is only two stops lower than the remaining rods in the grou Which ONE (2) of the following statomonts ; describes the action which must be taken? (TS 3.1.1.1 and 3.1.3.6 are i attached) ; a. Determine that the Shutdown Margin requiremont of TS "t.1.1.1_is ! untisfied within 1 hour and be in llot Standby within 6 hour I b. Reduce turbino power to equal to or less than 75% of Rated Thermal Power within 1 hour and reduce the III Neutron Flux Trip Sotpoint to 85% within the following 4 hour c. Datormine the Shutdown Margin of TS 3.1.1.1 at least once overy 12 hours and that FN Delta 11 is within limit : i d. Maintain the remaining rods in the group within plus or minus 12 steps of F8 while maintaining the rod sequence and insertion-limits of TS 3.1.3.6 during subsequent operatio , e i MASTER CDPY
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SENIOR REACTOR OPERATOR Page 65 i t l
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QUESTIOll: 059 (1.00) l i The following plant conditions exist on Unit is -
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Reactor pcwor is 40%.
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Control Bank rod B-4 has droppe l t
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Which ONE (1) of the following actions will cause a " Rod Control Urgent Failure" alarm actuation during the dropped rod recovery? , n. Placing Rod Control in manua ' b. Rosetting the P/A convertor to zer :
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c. Withdrawing Rod B-4 to its bank positio d. Placing the lift coil disconnect switches for the remaining rods i in the control group to of !
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SENIOR REACTOR OPERATOR Page 66
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QUESTION 0G0 (1.00) The following plant conditions exist on Unit 1
- Reactor power is 43%. !
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"RDI AT BOTTOM ROD DROP" in alar "RPI TWO OR MORE RODS AT BOTTOM" in alar Bank D control rods stopping out in automati Tavo 3 degrees F below Tro "COMPARATOR P/R CllANNEL DEVIATION" in alar Which ONE (1) of the following statements describes the IMMEDIATE ACTION to be taken?
a. Manually trip the reactor and enter EP/1/A/5000/01, " Reactor Trip or Safety Injection".
b. Place rod control in manual and withdraw control rods as noconaary to match Tavo and Tro c. PILeo rod control in manual and increase turbino load to maintain Tavo and Trof matched, d. Place rod control in manual and stop any turbino load changes in progress.
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O O SENIOR REACTOR OPERATOR Pago 67 QUESTIOlis 061 (1.00) Which ONE (1) of the following responses would be an indication of a control bank failure to novo during a power increase from 50% power to 100% power with rod control in automatic? " ROD CONTROL SYS URGENT FAILURE" in alar b. Increasing Tavg and possible "NC SYS !!I/LO TAVG" in alar c. liigh pressurizer levol indication and possible alar d. " ROD CONTROL SYS lioll-URGENT FAILURE" in alar l MkSTER CDPy
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O O ' SENIOR. REACTOR OPERATOR Page 68 I-QUESTION: 062 (1.00) Ths-following plant condition exist on Unit 11
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- Recovery from a LOCA is in progres i
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So has been terminate ,
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Imminent PTS Condition (EP201) exist Which ONE [1] of the following conditjons requires manual reinitiation of Safety Injection by the operator? a. Pressurizer level less than 5%. b. NC subcooling greater than 50 degrees F.
, c. Core Exit thermocouple readings greater than 700 degrees d. All non-faulted steam generator narrow range icvels greater than . 82%. .
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SE!1IOR REACTOR OPERATOR Page 69
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QUESTIO!1: 063 (1.00)
Which OllE (1) of the following times describos the time interval after a ! LOCA that 110 and til pumps are aligned for liot Leg Recirculation? a. 4 hour ' b. 9 hour c. 12 hour d. 15 hour MASTER ggpy
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SENIOR REACTOR OPERATOR Page 70
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QUESTION: 064 (1.00) The following plant conditions exist on Unit 1:
' - Reactor Trip has just occurre Operators are performing the actions of EP-01, " Reactor Trip or Safety Injection".
- " Ice condenser Dooru Open" is in alar ; Which ONE (1) of the following procedures should the operators be directed to enter? a. EP-1B, "SI Termination Following Spurious Safety Injection".
b, EP-1C, "High Energy Line Break Inside Containment". . c. EP-ID, "Steamline Break Outside Containment".
d. EP-18, " Steam Generator Tube Rupture".
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O O SENIOR REACTOR OPERATOR Pago 71 i
l l QUESTION: 065 (1.00) l l AP/1/A/5500/08 "Halfunction of Reactor Coolant Pump" Iminodiato Action directs operators to trip an lic Pump if the Pumn Lower ' earing . Temperature is 225 degrees F or greate Which OliE (1) of the following reasons explains why the lic pump is tripped if the Pump Bearing Temperature Limit is exceeded?
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a. Minimize the risk of molting bearing babitted surfaces and impair llCP coastelow ' b. Reduce the possibility of soal leakoff flashing to steam and damaging the liCP seal Increased bearing friction may result in uncontrolled riso in ' bearing temperattire d. Minimize the risk of an lic Pump Lube Oil fir , i i l C
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_ O O SENIOR REACTOR OPERATOR Page 72 QUESTION: 066 (1.00) The following Unit 1 plant conditions exist:
- Operating in accordance with OP/1/A/6100/03 " Controlling Procedure For Unit Operation".
- Reactor power is 18% proceeding to 85%.
- NC (Reactor Coolant) Pump 1A was tripped when the 1D Pump Lower Bearing Temperature exceeded the alarm setpoin The operator recognised that the wrong NC pump was tripped and tripped NC Pump 1 Which ONE (1) of the following actions is required to be performed? '
a. Start NC Pump 1B in accordance with OP-6150/02A "NC Pump Operation". Enter OP/1/A/6100/02 " Controlling Procedure For Unit Shutdown".
c. Ensure the reactor tripped and go to EP/1/A/5000/1 " Reactor Trip or Safety Injection".
d. Enter Technical Specification 3.0.3.
, MASTER
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QUESTION: 067 (1,00) The following plant conditions _ exists
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A normal plant cooldown is in progres The computer alarm for the KC Surgo Tank indicates a need for makeup from the domineralized water syste ,
- The water level in Surge Tank 1A is slowly DECREASIN Locally, there is no apparent cause for the decreasing. surge tank leve The associated makeup valve to surge tank 1A is fully ope Which ONE (1) of the following explanations is the cause of decreasing ourge tank level?
a. There is a leak in the KC heat exchange b. There is a leak in the Letdown heat-exchange c. There is a leak in the Reactor Coolant Pump Thermal barrier-heat exchanger, d. The plant cooldown rate is excessiv MASTER COPY
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O O SEllIOR REACTOR OPERATOR -Page 74 QUESTION: 068 (1.00)
'When responding to an ATWS condition in accordance with EP/1/A/5000/2A1 " Nuclear Power Generation /ATWS", which ONE (1) of the following reasons explains why Pressurizer pressure is verified to be less than the PORV cotpvint?
a. Ensures that a LOCA will not be created by a faulty PZR POR b. Ensures that the PZR PORVs have shut after opening to reduce NC pressure following tripping the turbin c. Allows sufficient boration flow rate to the NC syste d. Prevent reactor pressure from exceeding safety limits during an overpressure transien MASTER COPY _
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' SENIOR REACTOR OPERATOR Page 75 QUESTION: 069 (1.00) .The following plant conditions exist on Unit 2 -
The plant is operating at 83% powe A control rod drops into the cor One Negative Rate Trip Bistable light is li About one minute later a second rod drops into the core in another quadran A second Negative Rate Trip Distable light is actuate Power and Tave have decreased slightly but otherwise operations are undisturbe Which ONE (1) of the following actions should be taken? a. Trip the reactor and enter EP/2/A/5000/01 " Reactor Trip or Safety Injection".
b. Reduce power in accordance with OP/2/A/6100/03 " Controlling Procedure for Unit Operation"'in preparation for entering OP/2/A/6100/02 " Controlling Procedure For Unit Shutdown".
c. Recover the dropped rods in accordance with AP/2/A/5500/14
" Control Rod Misalignment".
d. Reduce power and enter AP/2/A/5500/16 " Case IV, Power Range Malfunction".
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SENIOR REACTOR OPERATOR Page 76 l
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QUESTION! 070 (1.00) i
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The following plant conditions exist on Unit 1
- The plant has exporlenced a main steamline ruptur EP/1/A/5000/2D1 "Imminont Pressurized Thormal Shock" has boon !
ontore ' Which ONE (1) of the following statomonts describes the major actions to be taken? I a. Maintain NC System cooldown rato; maintain NC System pressur j b. Maintain NC System cooldown rato; depressurize NC Syste c. Stop NC System cooldown; maintain NC System pressur d. Stop NC System cooldown; depressurize NC Syste > l
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'SEllIOR REACTOR OPERATOR Pago 77 ;
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. QUESTION: 071 (1.00)
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The following plant cor.ditions exist:
- Unit 1 is at 35% power increasing to 100% power in accordance with OP/1/A/6100/03 " Controlling Proceduro For Unit Operation". ' -
AP/1/A/5500/23 " Loss of Condensor Vacuum" was entered due to . Condonsor vacuum indication DECREASIll Condonsor vacuum is currently stable at 24.5 inches of li Exhaust flood Temperature is 228 degroon P and decreasing slowl ! Which OllE (1) of thu following actions should be taken? i a. Trip the turbine; trip the reactor; enter EP/1/A/5000/01 -
" Reactor Trip or Safety Injection".
b. Trip the turbino; enter AP/1/A/5500/02 "Turbino Gonorator Trip". i c. Stop the power increase and reduce power in accordance with OP/1/A/6100/03 " Controlling Proccdure For Unit Operation" Enclosure 4.2 " Power Decrease". , d. Increaso load por OP/1/D/6'300/01 "Turbino Generator".
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O O SEllIOR REACTOR OPERATOR Page 78 QUESTIOll 072 (1.00) Which ONE [1] of the fellowing statomonts describes the result of taking Inverter 1EID wanual Dypass Switch to the "DYPASS" position with the "IN SYNC" light not lit? a. There is no power on the alternato source; thereforo power would be lost to IERP b. Since the "Ill SYNC" light muut be lit before tSo switch can be moved, tho switch would not operat c. Since 1ERPD and IVRD would momentarily be connected out of phase, the switch would probably be damage d. The switch would operato; but the connection would not be transferred because the Kirk-key interlock has not boon mad !
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ -_____________ _ _ _ O O set 110R REACTOR OPEP.ATOR Pago 79 t QUESTIOll: 073 (1.00) ; During an authorized release of liquid vante, EMF-49 alarm Which ONE ;
(1) of the following statements describos the action to be taken by the ;
oporator? i
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a. Recot the alarm and valt to soo if the alarm return b. Writo a work request to have the alarm checked by IA c. Notify Operations Duty Engincar of the alarm conditio d. Ensure automatic actions occurre !
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O O SENIOR REACTOR OPERATOR Page 80
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QUESTION: 074 (1.00) Which ONE [1] of the following descriptions identifies the location of a fire which would automatically actuato a Halon Piro Protection Systom? a. Operator's Storage Rooms, Document Control Storage Are c. Cable Room Corridors on Units 1 & Steam Production Office Building.
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. O O SEllIOR REACTOR OPERATOR Pago 81 QUESTI0ll: 075 (1.00) , Which OllE (1) of the following individuals becomes the Fire Captain when a firo is reported on-sito? a. Superintendent of Operation b. Safety Associato/ Specialist, Fire Protectio c. Shift Supervisor,
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d. Assistant Shift Superviso .
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p i V V SENIOR RP. ACTOR OPERATOR Page 82 QUESTION: 076 (1.00) The following plant conditions exist:
- EP/1/A/5000/2C1 " Loss of Secondary lleat Sink" has been entere Feedwater flow cannot be establishe Which ONE (1) of the following feed and bleed paths will be used?
a. Feeding to the Steam Generators (S/Gs) and using steam dump to bleed stea _ reading to NC (Reactor Coolant) System and using Pressurizer PORVs to bleed stea c. Feeding to NC System and using Normal Letdown to blood the NC Syste Feeding to the S/Go and using S/G PORVs to bleed steam.
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I O O i SE!4IOR REACTOR OPERATOR 'Page 83
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i QUEST 10!is 077 (1.00) ,
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If EP/1/A/5000/2C1 " Loss of Secondary Heat Sink" has boon entered, which ONE (1) of the following reasons explains why NC (Reactor Coolant) Pumps , are stopped in loops with Steam Generator wido rango level less than 5%? ; a. Stops adding unnecessary hea b. Prevent lic Pump seal damago due to steam formation in the Number 1 Sca . c. Allows rostoration of food to the steam Generato l P d. Allows cooling the associated loop by natural circulatio ,
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O O : SENIOR REACTOR OPERATOR Page 84
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t i QUESTION: 078 (1.00) When operating in accordance with EP/1/A/5000/201 " Loss of Secondary Heat Sink", which ONE (1) of the following reasons explain why cubcooling and CETs (Core Exit Thermocouples) are monitored throughout the ("ocedure? a. They provide ' 'te feed and bleed initiation criteria, b. They provide ; le criteria for exiting from EP2C , c. They provide the criteria for securing the reactor coolant pump d. They provide the criteria for feeding the steam generators from the RN Syste ,
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r QUESTIOllt 079 (1.00) NC (Roactor Coolant) Systen activity has ranged from 1.1 microcurion por I gram doso equiemlent I-131 to 1.5 microcuries por gram dono equivalent .
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I-131 for the laut 48 continuous hours. Which ONE (1) of the following , actions should be taken? (Technical Specit'ication 3.4.8 is attached) ! a. Shutdown the reactor to llot Standby and notify the NR b. Trip the reactor; entor EP/1/A/5000/1 " Reactor Trip or Safety Injection"; and notify the NR c. Reduco reactor power to 1000 than 75% and notify the NR d. Reduce reactor power to 50% and notify the NRC if activity increase PT3
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F O O l SENIOR REACTOR OPERATOR Page 86 [
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QUESTION: 000 (1.00) [ f
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If EP/1/A/5000/1A1 "Naeurrl Circulation Cooldown" has boon entered from EP/1/A/5000/1A "Roactor Trip Response", which ONE (1) of the following indications is used to datormino when to stop NC (Roactor Coolant) System depressurization and allow further upper head cooling? . a. All T-110To greater than 200 degroos b. All NC wide range temperatures greator than 200 degroes c. Reactor Vessel UR ILvol loss than 97 , d. NC subcooling greator than 65 degroos F with all CRD vont fans ! o ! ,
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_ - _ _ _ _ - _ _ _ _ - - _ _ _ _ _ - _ . _ ._ O @ SENIOR REACTOR OPERATOR Page 87 QUESTION: 081 (1.00) EP/1/A/5000/01 " Reactor Trip or Safety Injection" Step 3 Action / Expected - Response reads: Verify either 1 ETA or 1ETB is energize Which ONE (1) of the following star,ments describes how 1 ETA or 1ETB are verified energized? ETA or ETB Line Voltage equal to or greats an 4160 volt b. Diesel Ger.erator A and B Blackout Load Scquencer Actuated Statas Lights are dark, ETA and ETB Normal or Standby incoming breaker amps are greater than ETA and ETB X, Y, and Z Phase UV Status Lights are dar '.- hwiejsk CDPY
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- ' SENIOR REACTOR OPERATOR Page 88 -
- QUESTION
- 082 (1.00)
The following plant conditions exist on Unit 1:
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Reactor is shutdown in Mode Pressurizer level is 63%.
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NC (Reactor Coolant) System pressure is 1385 psi .T (Pressurizer Relief Tank) pressure is 6 psi s If one pressurizer PORV is leaking slightly, which ONE (1) of the following temperatures will be indicated on the Relief Valve discharge - RTD (INCRD 5940]? a. 247 degrees F.
b. 263 degrees c. 275 degrees d. 282 degrees K '
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O-SENIOR REACTOR OPERATOR _ Page 89-QUESTION: 083 (1.00) EP/1/A/5000/1A1." Natural Circulation Cooldown" Step 11 directs the operators to: Monitor NC System cooldow Which ONE [1] of_the following sets of indications demonstrate existence of adequate natural circulation? a. Core exit thermocoup'es decreasing; T-HOT decreasing; N(1 subcooling increasin b. Core exit thermocouples decreasing; T-HOT decleasing; NC subcooling decreasin c. T-IiOT steady _or increasing slowly; T-COLD decreasing; NC subcooling increasin d. Core exit thermocouples steady or increasing slowly; Steam Generator pressure decreasing; NC subcooling decreasing.
i: MASTER COPY O O SENIOR REAC1'OR OPERATOR Page 90
. QUESTION: 084 (1.00)
The following Unit _1 conditions exist following a Ss (Safety Injection):
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NCS pressure is 1050 psi Tave is 557 degrees All ESF Monitor Panel lights are properly LI Containment Pressure is 2.8 psi EMF 53A (B] indicates 4 R/h NV/SI flcw is 500 gp NCS boron is 500 pp Core Burnup is 200 EFP Which ONE (1) of the following actions should be performed by the ! operator? a. Borate the NCS to 850 ppm Boron for proper SDM prior to cooldown to less than 200 degrees F NC Wide Range Temperatur Ensure NV-202 and NV-203 are ope Secura NS Pumps that are in servic Secure all NC pump j < MASTER COPY
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OL O-SENIOR REACTOR OPERATOR Page 91-
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QUESTION: 085 : (1. 00) _
'Which ONE [1] of: the following contains indications that are listed in AP/1/A/5500/12 " Loss of Charging or Letdown" as' symptoms of a~ Loss of Charging?
a. "NCP SEAL WATER LO FLOW" alarm LIT and 'VCT LO LEVEL" alarm LI ' b. "CilARGING- LINE HI/LO FLOW" alarm LIT and "PZR 11I LEVEL DEV CONTROL"' alarm LI " CHARGING LINE HI/LO FLOW" alarI;. LIT and " REGEN HX LETDN HI TEMP" alarm LI "PZR HI LEVEL DEV CONTROL" alarm LIT and "VCT LO LEVEL" alarm LI O ? f"
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- O SENIOR RFACTOR OPERATOR Page 92
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-QUESTION: 086 (1.00)
The following plant conditions exist:
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Unit 1 is in Mode NC (Reactor Coolant) pressure is 950 psi NC temperature is 455 degrees'F and rising slowl All NC Pumps are operatin #1 Seal leakoff is 0.9 gpm per pump.
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Seal injection flow is 7.9 gpm per pum #1 Seal leakoff temperature on NC Pump A is 205 degrees F and rising slowl Which ONE (1) of the following actions should be taken? a. Open the #1 Seal bypass valv b. Stop NC Pump Increase Seal injection flo d. Lower the temperature of the KC (Component Cooling) Wate b 1 tt -
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O OL SENIOR-REACTOR OPERATOR Page 93-QUESTION: 087 _ (1.00)
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The following Unit 1 plant conditions exist:
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Shutdown in Mode 1 ,
- - Pressurizer NC level is decreasing (Reactor Coolant] pressure isslowly. _ing decreas slowl ' -
ND-(Residual Heat Removal) flow is hig "ND & NS ROOMS SUMP LEVEL EMERG HI" is in alar Which ONE (1) of the following IMMEDIATE ACTIONS must be performed?- a. Secure operating ND pump (s) and SHUT ND System suction and discharge valves to the NC Syste b. Secure operating ND and NC pump (s).
c. Secure operating ND pump [s] and manually initiate-Phase A isolatio Secure operating ND and NC pump [s] and manually initiate-Phase B-isolation.
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' SENIOR REACTOR OPERATOR- Page 94-QUESTION: 088 (1.00) During Mode 1 operation, with the Pressure Control Selector Switch selected to the 1-2 position, which ONE (1) of the following statements describes the effects of Pressuriser Pressure Control Channel I failing ; LOW? (NOTE: Assume no operator action is taken) All heaters on; PORV NC34 blocked; NC pressure rises; PORVs 32 & 36 open; NC pressure modulates between 2235 and 2315 psig, Sprays full on; PORV 34 opens; PORV 34 closes at 2185 psig; Low Pressure Reactor Trip at 1945 psig; Low Pressure SI at 1F45 psi c. All heaters on; PORV 32 & 36 open; PORV 32 & 36-close at 2185 psig; spray valves modulate NC pressure between 2260 end 2310 psi Sprays full on; PORVs 32 & 36 blocked; all heaters on at 2210 psig; heaters modulate to maintain NC pressure between 2210 and 2218 psi , 3,c
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SENIOR REACTOR OPERATOR- ;Page 95 q
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QUESTION: 089 (1.00) The following plant conditions exist:
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Unit 1 is in Mode Source Range N31 indicates 1.2 x 10E4 cp Intermediate Range N35 indicates 1.1 x 10E-10 amp Intermediate Range N36 indicates 9 x 10E-11 amp Control Rod Bank C is being withdrawn to criticalit Source Range N32 fails off-scale low Which ONE (1) of the following actions should be performed? [See figure below) (Technical Specification 3.3.1 is attached) a. Suspend all operations involving positive reactivity change b. Restore Fource Range Channel N32 to OPERABLE within 48 hours or
- open the Reactor Trip Breakers within the next hou c. Block both Channels of Source Range High Level-Reactor-Trip and continue the startup, d. Place Source Range Channel N31 in the tripped _ position and continue the startu '"L"" ZE
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SENIOR REACTOR OPERATOR Page 96 QUESTION: 09 (1.00) Tha following plant conditions exist:
- Reactor startup is in progres Source Range Channel.N31-indicates 7E3 cp Source Range Channel N32 indicates 8E3 cp Intermediate Range Channel N35 indicates 3E-11 amp Intermediate Range Channel N36 indicates 1.5E-10 amps.
' Which ONE (1) of the following statements describes the status of the nuclear instruments? (Note: Figure CN-IC-ENB-3 below) a. N35 is over compensate b. N35 is under compensated, c. N36 is.over compensated, N36 is under compensate S? O 4
. ,0 .,
100 %
#
10 % 11 * 1% PWR 14 * to MMME MhM 4
. to' #
10
. se '* ' 10 ,
to AMM to
- 1Q .. MASER .. .
CDPY
. . . . _ . _ _ . . - . - . .. .- - . . -. .O O .
SEMIOR REACTOR OPERATOR Page 97 QUESTION: 091 (1.00) Which ONE (1) of the following reasons describes the basis for maintaining the pressure in a ruptured S/G (Steam Generator) less-than' q 1125 psig?
' Provide a margin to lifting the steam line safety valve b. Minimize secondary system contamination from the NC (Reacter Coolant) System, c. Prevent excessive cooldown of the NC System from overfeeding the S/ d. Allow NC System depressurization and rapid equalizing with S/G pressur MASTER COPY .
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OL y(-)- SENIOR REACTOR OPERATOR Page 98 QUESTION: 092 (1.00) Th'e foll'ouing plant conditions exist:
-
AP/1/A/5500/06 " Loss of S/G Feedwater" was entered due to a-loss of feedwater supply to the S/G Case II actions of AP06 have been completed through Step 1 Hotwell levels are 0.22 feet CA flow is 590 gp Which ONE (1) of the following actions should be performed? '(NOTE: AP/1/A/5500/06 is attached) a. Shift CA suction to the CACST and US b. Isolate RN from the CA Pump suctio c. Run only one CA Pump, Stop all CA Pump np G-g! g J.) f a v i" {l L 4 u s s I ,y m _w0 a
'q
,xg p> if k3[? k B
__
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O O SENIOR REACTOR OPERATOR Page 99 QUESTION: 093 (1.00) Which ONE [1] of the following conditions must-be' satisfied before blackout switchgear control power may be swapped back from the alternate cupply to the normal supply? a. The undervoltage relay on CDA must be rese VDC Battery 1DPB cr 2DPB breakers are close Supply breakers to 4160 & 600V control power and CAPT control power are close d. Dattery charger ECA has its AC input and DC output breakers closed.
F n . > s es a po n ;,q O O
' SENIOR REACTOR OPERATOR- Page100 QUESTION: 094 (1.00)
Which ONE [1] of the following reasons describes the MOST LIKELY cause of a reactor trip during a sustained loss of Instrument Ai a. NC [ Reactor Coolant] Low Pressure due to excessive cooldown-from S/G [ Steam Generator] PORVs failing ope b. S/G Lo Lo Level due to CF (Main Feed] control valves failing close c. NC High Pressure due to failure of pressurizer spra d. Pressurizer High Level due to failure of Steam Dumps to operat MASTER l COPY
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n Or U
- SENIOR REACTOR OPERATOR- Pag 0101 t
QUESTION: 095 (1.00) Which ONE (1) of the following indications is used to determine the position of VS-78 [VS to VI) in AP/0/A/5500/22 " Loss of VI"? a. Valve ste b. Local indicating ligh c. MC13 in CR OPEN/CLOSE indicatio d. OAC poin d Mz.sai COPY
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O O SENIOR REACTOR OPERATOR page102 QUESTION: 096 (1.00) In addition to ensuring the Unit 2 Control Operator is notified, which ONE [1] of the following actions must be pert'ormed immediately if Instrument Air pressure falls to the low pressure alarm setpoint on Unit
a. Dispatch operators to VI [ Instrument Air] compressor panels to verify proper compressor operation and to start the backup temporary VI compressor, b. Dispatch operators to VI compressor panels.to verify proper compressor operation and dispatch operators to locate and isolate the leak on the VI syste Dispatch operators to locate and isolate the leak on the VI system and start the backup temporary VI compresso Dispatch operators to isolate the leak on the VI system and to open IVS-78 locally, i , MASTER COPY
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~ SENIOR REACTOR' OPERATOR Page103- . QUESTION: 097 (1.00) The following plant conditions exist:
-
Reactor power is 90%. .
. -
Pressurizer Level Control Switch is in Position 1-2.-
-
Letdown isolate Pressurizer heaters are off.-
-
Pressurizer level is increasin Charging flow has increased to maximu Which ONE [1] of the following reasons explain the cause of the indication listed above? a. Level II has failed hig . b. Level I has failed hig c. Level I has failed low, d. Level II has failed lo Et f4 ** ?" h}w{h ; L CDPY
.. - . .-
0 0 SENIOR REACTOR OPERATOR Page104
.
-QUESTION: 098 (1.00) Which ONE (1) of the following statements describes the entry conditions to EP3B " Loss of All AC Power Recovery With S/I Required" from EP03
" Loss of All AC Power"?
a. Power restored to either 1 ETA or 1ETB and graater than 0 degree F subcooling, greater than 5% Pressurizer level and no auto S b. Power restored to either 1 ETA or 1ETB or greater than 0 degree F subcooling, greater than 5% Pressurizer level and no auto S c. Power restored to either 1 ETA or 1ETB and no subcooling or Pressurizer level greater than 5% or auto SI presen d. Power restored to either 1 ETA or 1ETB or no subcooling or Pressurizer level greater than 51, or auto SI-presen MASTER COPY
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SENIOR REACTOR OPERATOR Page105 t QUESTION: 099 (1.00) Two of the Immediate Actions for a Loss of All AC Power are:-
- Manua'E Trip Reactor and Verify Reactor Tri Verify Turbino Tri Which ONE (1) of the following statements lists the remaining IMMEDINTE ACTIONS?
a. Verify D/G [ Diesel Generator) status, b. Verify D/G status, dispatch 2 operators to establish NC (Reactor Coolant) Pump Seal injectio c. Verify D/G status, dispatch 2 operators to establish NC Pump Seal injection, verify NC System isolate d. Verify D/G status, dispatch 2 operators to establish NC Pump-Seal injection, verify NC System isolated, Ensure RN flow to
.
D/ MASTER COPY (********** END OF EXAMINATION **********)
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SENIOR REACTOR OPERATOR- Page106 ANSWER: 001 (1.00) b.
REFERENCE: OP-CN-ADM-SD, Rev. 09/06-09-92/ELR, Page 30,-Para, G.1.& KA 194001K101 [3.6/3.7) Both-RO and SRO 194001K101 ..(KA's) ANSWER: 002 (1.00) d.
REFERENCE: OMP 2-18, Rev. 31, Page 2, Para. 3. KA 194001K101 [3.6/3.7) Both RO and SRO 194001K101 ..(KA's) ANSWER: 003 (1.00) MASTER cdOPY
.
,( - . -- . . - - . .. --
~.
O' Or.. -SENIOR REACTOR OPERATOR Page107-
-REFERENCE:
'OP-CN-ADM-SD, Rev. 09/06-09-92/ELR,_Page 13, Para. 4. arc KA 194001K102 [3.7/4.1)
SRO only 194001K102 ..(KA's) ANSWER: 004 (1.00) REFERENCE: CNS Directive 3.8.6 [TS), Rev. 22, Page 5, Para. 4.1, Page 4, Para. KA 194001K103 (2.8/3.4) Both RO and SRO 194001K103 ..(KA's)
>
ANSWER: 005 (1.00) a.
-REFERENCE: i
CNS Directive 3.8.8, Rev.-27, page 16, Para. 5.17.2 & 5.1 ! KA 194001K104 [3.3/3.5) Both RO and SR K104 ..(KA's) Iit k~S k[,f
.
n-O O SENIOR REACTOR OPERATOR Page108-
. . ANSWER:- 006 (1.00) REFERENCE:
CNS Directive 3.8.5 (TS],~ Revision 18, Page 3, Para. 6. KA 194001K104 (3.3/3.5] Both RO and SRO
,
194001K104 ..(KA's)
.
ANSWER: 007 (1.00) REFERENCE: Response Procedures RP-04 and RP-0 KA 194001K104 [3.3/3.5] SRO only 194001K104 ..(KA's) ANSWER: 008 (1.00) h8 /3 <
[?
LM4e{t-*=}?
.
L O O SENIOR REACTOR OPERATOR Page109 REFERENCE: OMP 2-9, Rev. 13, Page 16, Para. 1 KA 194001K105 [3.1/3.4] SRO only 194001K105 ..(KA's) ANSWER: 009 (1.00)
- %
REFERENCE: SD 2.11.12, Rev. O KA 194001K109 (3.4/3.4) Both RO and SRO 194001K109 ..(KA's) ANSWER: 010 (1.00) ' REFERENCE: OMP 2-21, Rev. 11, Page 7, Para. 11.1 KA 194001A104 (3.0/3.2] SRO only 194001A104 ..(KA's) fM/ L ..'; n i P
, . j ju
~s
[ ,,'<a (~2 n)gU_
._ -_ - - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ - _ _ _ - _ - _ _ - - _ _ _ - _ _ - - - _ _ _ __
- -- - __
SENIOR REACTOR OPERATOR Page110 ,. ANSWER: 011 (1.00) REFERENCE: OMP 2-17, Rev. 23, Page 2, Para. 6. KA 194001A106 [3.4/3.4] Both RO and SRO -- 194001A106 ..(KA's)
, ,
ANSWER: 012 (1.00) . REFERENCE: OMP 2-10, Re , Page 6, Para. 8.5.A& KA 194001A107 [2.5/3.2) Both RO and SRO _ 194001A107 ..(KA's) ANSWER: 013 (1.00) k m/,.- p
.,
-
- -v,f jf ?Y
*
- _ _ _ - - _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ . _ ____ _ _-
OL O SENIOR REACTOR OPERATOR = Page111-
.
REFERENCE: L --l 'OMP 2-11, Rev. 16, Page 2, Para. 4.4 and Page 3, Para. KAL194001A109 [2.7/3.9) Both RO and SRO 194001A109 ..(KA's) ANSWER: 014 (1.00) d.
REFERENCE: OMP 2-16, Rev. 12, Page 4, Para. 6. KA 194001A109 [2.7/3.9) Both RO and SRO 194001A109 ..(KA's) ANSWER: 015 (1.00) c.
REFERENCE: OMP 1-8, Rev._18, Page 11, Para. 1 CN-OP-ADM-OMP, Rev. 09/03-23-92/SWH, Page 23, Para. 'lU\ -194 001A111 [2.8/4.1) Both RO and SRO 194001A111 ..(KA's) h rl, 'a . e
. .
i Y I'd
-/
,
i:n . I'*
-
__
- .-. . . .- -.
o o : - SENIOR REACTOR OPERATOR Page112 ,
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' ANSWER: 016 (1.00)
- REFERENCE:
-RP/0/A/5000/06, Change 5 to Retype #6,, Page 1, Para. KA.194001A116 (3.1/3.4)
-SRO only 194001A116 ..(KA's)
:
ANSWER: 017 (1.00) , REFERENCE: RP/0/A/5000/10, Page 1, Para. KA 194001A116 (3.1/4.4) SRO only 194001A116 ..(KA's) ANSWER: 018 -(1.00)
-c.
f _ 0 I n o r?e"g-0013l(
. +
.
V m G' V-SENIOR REACTOR OPERATOR Page113 5
.
REFERENCE:
'
- OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 19, Para. 8.d.
.KA 001000K103 (3.4/3.6] JBoth RO and SRO 001000K103 ..(KA's) ANSWER: 019 (1.00) d.
REFERENCE: OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 30, Para. KA 001000K558 (2.7/3.2] Both RO and SRO 001000K558 ..(KA's)
>
ANSWER: 020 (1.00) d.
REFERENCE:
"'echnical Specification, page B 3/4 1- .
-KA 001000G006 [2.9/3.8)
-SRO only 001000G006 ..(KA's)
l.a?,Y,P' Copy ~
.- _ _ __
.
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' SENIOR REACTOR OPERATOR Page114 ANSWER: -021 (1.00) REFERENCE:
OP-CN-CNT-CNT, Rev. 10/06-02-92/JWH, Page 28, Para. 2.4. Catawba Exam Bank Question NCP-13 KA 003000K201 [3.1/3.1] Both RO and SRO 003000K201 ..(KA's) ANSWER: 022 (1.00) REFERENCE: Technical Specification Bases, Page B 3/4 4-1 KA 003000G006 (2.7/3.8) Both RO and SRO 003000G006 ..(KA's) ANSWER: 023 (1.00) AMS.Tgg cc LO.Py .
-
. . - - . . . . - - . . . ..- -, . - - . - . . . . . . . . - .
.
SENIOR REACTOR OPERATOR Pago115 !
-
.
REFERENCE: >
)
OP-CH-PSS-KC, Rev. 1.*:03-03-92/GFW, Page 9, Par B.3.a.- OP-CH-PS-11V, Rev. 04/04-10-92/PEV, Page 25, Para.- B.14. . KA 004000A401 (3.8/3 9) Beth RO and SRO l
,
004000A401 ..(KA's)
-ANSWER: 024 (1.00) l
,
REFERENCE: OP-CH-PS-llV, Rev. 04/04-10-92/PEV, Pago 26, Para.- Catawba Examination Bank Question llV-89 KA 004010A205 (4.1/4.3) Both RO and SRO 004010A205 ..(KA's)
>
-ANSWER: 025 (1. 00) REFERENCE:
'
OP-CH-ECCS-ISE, Rev. 1C/03-11-92/CTK,-Page 16, Para. KA 013000A105 (3.4/3.6) Both RO and SRO
,
,
013000A105 ..(KA's) i e. '
".
.N1
. ,,j
.. _ . _ _ . _ . . . - . . _ _ _ _ _ . . _ .
O O l SENIOR REACTOR OPERATOR Pago116 ) t
; ,
ANSWERt 026 (1.00) ,
,
a4 , i REFEREllCE: Catawbu Examination Bank Question EDA-15 , _)U4 014000G008 (2.9/3.1) , Both Ro and SRO .
!
014000G000 ..(KA'u)
!
f ANSWER: 027 (1.00) r REFERE!1CE: ,
.
OP-CN-IC-END, DWG Cll-IC-EllD-3 KA 015000A303 [3.9/3.9) Both no and SRo 015000A303 ..(KA's) t ANSWER 028 (1.00)
' MSTER COPY
_____ __ _ ___ __________- .- _ -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - _ _ _ _ _ _ - _ _ _- - ._ _ _ _ _- _ _ _ _ - _ _ _ _ _ - _ _ - _ _ O O i
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SENIOR REACTOR OP2RATOR Page117 REFERENC l
=OP-CN-IC-EllB, Rov. 09/03-19-92/JLY, Page 11, Para. fl . KA 015000KG04 (3.1/3.2) !
l Both RO and SRO l t
.
015000KG04 . . (KA's) A11SWER: 020 (1.00)
: REFERENCE:
OP-CH-IC-END, Rov. '9/03-19-92/JLY, Pago 7, Para. C. KA 015000K406 [3.9/9.2) Both RO and SRO
:
015000K406 . . (KA's)
;
ANSWER: 030 (1.00) f ! REFERENCE: L OP-CN-IC-ENA, Rev. 05/08-05-90/GFW, Page 10, Para. 1 KA 017000K101'[3.2/3.2) EJth RO and SRO 017020K101 . . (KA's) g MRSTER nmf, p f y ..
%i %h 0
.
---tv*g ,,w- g y-T'7wTw-wyt w w v wi Tr'r =rrwseSvqar --
4 t new *- p y ef- - * - = s er- +a- * y abb M - - ' - -
. . - . . _ - . .._ ~ . .. . . . _ -. .- . . - . , - - -- - . .. - -. . . . _ - --- . .
'
O O SENIOR REACTOR OPERATOR Page118 i i
!
1 ANSWER: 031 (1.00) l i ;
;
REFERENCE:
!
Catawba Examination Bank Question CNT-VV-9 ! t KA 022000A404 (3.1/3.2) ;
!
Both RO and SRO 022000A404 . . (KA's) . ANSWER: 032 (1.00)
. 't REFERENCE:
Catawba Technical specifications, Pago B 3/4 6- ' KA 025000K301 [3.8/3.8) Both RO and SRO 025000K301 . . (KA's) q, ANSWER: 033 (1.00) ,
,
i hMSTER : my J-
.
r ew -e-y-+w,-e.~-e -,_v - e w w --m--w -- - 3-i e rw------ e t----. w -,e w -- < - - - + ~. - -em+r--~r- -ar---e,m--- ---t~r --,+--rw--s +r = +-+yeve1-4 erw-+- e - =rdr
\
O O SENIOR REACTOR OPERATOR Page119 REFERENCE: ' Technical Specification 3.6.2 and 3. KA 026000G005 (3.3/3.9) SRO only 026000G005 ..(KA's) ANSWER: 034 (1.00) , !
REFERENCE:
,
Catawaba Examination Bank Question CA-31 KA 061000A202 (3.2/3.6) Both RO and SRO i
'
061000A202 ..(KA'c)
:
ANSWER: 035 (1.00) , > REFERENCE: OP/0/B/6350/07, Page 1,-Para. KA 06?000G010- (3.1/3.2) Both RO and SRO
'
063000G010 . (KA's) MASTER .
CDPY 1
-~ ..,_._,.,_, . _ , . . ..._ _ . . . . . - - . , . _ . , . , _ . _,, .,..., --,_ . _. ._-.__ . =r, , _ -
'
O O : GENIOR REACTOR OPERATOR Pago120 , l
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ANSWER: 036 (1.00) !
!
REFERENCE: r
!
OP-CH-WE-EMP, Rev. 09/12-11-91/CTX, Page 12, Par . ! l KA 072000G007 [2.6/2.9) i t Both RO and SRO
:
072000G007 ..(KA's)
!
I ANSWER: 037 (1.00) :
-, !
,
REFERENCE: OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 27, Para. 5.b.3.c. and 5.b. !
'KA 002000K603 [3.1/3.6)
^
Both Ro and SRO t 002000K603 ..(KA's) -!
,
ANSWER: 038 (1.00)
. O f8 L[J]a(A'}f,'
CUPY _ ___ _ _ _ _ _ _ _
. / T / ', 'q , x .j/
SE!110R REACTOR OPI:RATOR Page121 REl'OR EllCE : OP-C11-PS-11C, Rev. 08/02-25-92/GlH, Page 30, Par . KA 002000K403 (2.9/3.2) SRO only 002000K403 ..(KA'n) AllSWER: 039 (2.00) . . . . [0.5 each) rder not specific) REl'EREllCE: Drawing C!l-ECCS-111- KA 006000K409 (3.8/4.1) SRO only 006000K409 ..(KA's) A!!SWER: 040 (1.00) tIh 9-* j.e : ,, s;,u c'p*[(*_VWj.,g{ n,,- 3()
. O O SENIOR REACTOR OPERATOR Pago122 REFERENCE: Technical Specification Danes, Pago D3/4 5- KA 006000G006 (2.9/4.0) SRO only 006000G006 ..(KA's) ANSWER: 041 (1.00) REFERENCE: OP-CN-PS-IPE, Rev. 04/10-03-90/GFW, Page 17, Para. 2. KA 010000K201 (3.0/3.4) Both RO and SRO 010000K201 ..(KA's) ANS;iER: 042 (1.00) PEFERENCE: OP-CN-PS-IPE, Rev. 04/10-03-90/GFW, Page 13, P. ira. 8. KA 010000A403 (4.0/3.8) Both RO and SRO 010000A403 ..(KA's) 00 [\ Whir t"n B: -1';) ) f;[{
.:
& d ,L(9 r
. .. .. . ......_._ J
O O i SENIOR REACTOR OPERATOR Page123
-)
j I ANSWER: 043 (1.00)
, j
!
'
REFERENCE:
'
Catawba Examination Bank Question ILE-26
!
KA 011000A211 (3.4/3.G) Both RO and SRO
'
011000A211 ..(KA's) ' ANSWER: 044 (1.00) ' REFERENCE: Catawba Examination Bank Question IPX-30 KA 012000K501 (3.3/3.8) + Both RO and SRO 012000K501 ..(KA's) P ANSWER: 045 (1.00) - r MSTER COPY
. = .- - . _ _-- - -
_ . . _ _ _ _ _ _ _ _ _ _ - - _ - _ _ - - _ - - . - - - - _ - - - - _ - _ . - - - - - -
,
O O i
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SENIOR REACTOR OPERATOR Page124 l
,
i REFERENCE: DrcWing CH-PS-NC-5 KA 016000K101 (3.4/3.4) l Both RO and SRO l l 016000K101 ..(KA'a) l l ANSWER: 046 (1.00) t REFERENCE: j OP-CH-CF-IFE, Rev. 06/03-09-92/GFW, Drawing CF-IFE-3
't
.
KA 016000A301 (2.9/2.9) l SRO only ,
,
016000A301 ..(KA's) ANSWER: 047 (1.00) b.- REFERENCE:
'O-CH-CNT-VX, Rev. 06/03-10-92/PEV, Page 8, Para. 2.2. KA 028000A101 (3.4/3.8) F n'AO only 028000A101 ..(KA's)-
ANSWER:- -048 (1.00) Ad COPY _ _ _ _ _ -
' I O O SENIOR REACTOR OPERATOR Pago125 REFERENCE:
'
OP-CN-Fil-KF, Rov. 06/11-28-90/GPW, Pago 18, Para D. KA 033000K303 (3.0/3.3) Both RO and SRO 033000K303 ..(KA'n) ANSWER: 049 (1.00) REFERENCE: OP-CH-STH-SM, Rev. 09/02-06-92/RJP, Pago 15, Para 2.a. In KA 039000A302 (3.1/3.5) Both RO and SRO 039000A302 ..(KA's) ANSWER: 050 (1.00) REFERENCE: OP-CN-DG-EQG, Drawing CN-DG-EQG-12 KA 064000K411 (3.5/4.0) Both RO and SRO 064000K411 ..(KA's) ANSWER: 051 (1.00) U* 2 -
.
ew D Ga'n, y
. . . . . . . . . . . . . . . . . . .
. _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
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O O SENIOR REACTOR OPERATOR Pago126 REFERENCE: OP-CN-SS-RFY, Rev. 02/12-11-89/SMD, Page 22, Para.1 KA 086000K406 (3.0/3.3) Both RO and SRO ; t 006000K406 ..(KA's) - t ANSWER: 052 (1.00) REFERENCE: OP-CN-PS-ND, Rev. 09/02-27-92/GFW, Page 13, Para D. KA 005000G007 (3.3/3.4) SRO only 0050000007 ..(KA's)
.
ANSWER: 053 (1.00) b.
^ REFERENCE: ,! OP-CN-PSS-KC, Rev. 13/03-03-92/GFW, pages 10 & 11,_ Para B.3. 'KA 000000K102 (3.3/3.4] 7 Both RO and SRO ; 008000K102 ..(KA's)
.
ANSWER: 054 (1.00) tJ flopy '
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._- ____-___ __-______ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _
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O O SENIOR REACTOR OPERATOR Pago127 REFERENCEi OP-CN-PSS-KC, Rev. 13/03-03-92/GFW. Page 10, Para B.3.c. & KA 008030A304 [3.6/3.7) ! Both no and SRo !
,
008030A304 ..(KA's) ANSWER: 055 (1.00) . REFERENCE: Catawba Examination Batth Question VP-6
'
KA 029000K302 [2.9/3.5)
-Both no and SRo 029000K302 ..(KA's)
,.
ANSWER: 056 (1.00) REFERENCE: OP-CN-STM-IDE, Rov 05/02-25-92/DPM, Page 16, Para 5. KA 041020K417 [3.7/3.9) Both RO and SRO 041020K417 ..(KA's)
.
f
' ANSWER: 057 (1.00) kf 4 g
'
. _ _ _ _
. _ _ . . . _ . _ ___._ _ __ __ ._. _ . ._ __ _ __ O O SENIOR REACTOR OPERATOR Page128 REFERENCE:
OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Pago 26, Para KA 000001G007 (3.1/3.3) SRO only
000001G007 ..(KA's) I I I
ANSWER: 058 (1.00) ) d.
REFERENCE: OP-CH-IC-IRE, Rev. 07/03-16-92/GFW, Page 28, Pera 2. ' KA 0000010008 (3.2/3.8) Both RO and SRO 000001G008 ..(KA's) kNSWER: 059 (1.00) c.
REFERENCE: Catabwa Examination Bank Question IRE-20 KA 0000'03G005 (3.6/3.6) SRO only 000003G005 ..(KA's) t ANSWER: 060 (1.00) , .. _ wWdt >> c w h a 1, qm
_ . . . _ _ . _ .
O- O SENIOR REACTOR OPERATOR -Pago129 l
, *
REFERENCE: < Catabwa Examination bank Question IRE-19 KA 000003G010 (3.9/3.8)
'
SRO only
'
000003G010 ..(KA's) ANSWER: 061 (1.00) , n.
REFERENCE: Catabwa Examination Bank Question IRE-75 KA 000005G005 [3.1/3.3) Both RO and SRO
>
00000SG005 ..(KA's) ANSWER: 062 (1.00) a.
REFERENCE: Catabwa Examination Bank Question CSF-42' KA 000011A104 (4.4/4.4) SRO only_
.000011A104 ..(KA's) ,
ANSWER: 063 (1.00)
'
b.-
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O O SENIOR REACTOR OPERATOR Page130 REFERENCE: Catabwa Examination Bank Question HI-8 KA 000011A111 (4.1/4.2) SRO only 000011A111 ..(KA's) ANSWER: 064 (1.00) REFERENCE: Catawba Examination Bank Question EP1-45 KA 000011G010 [4.5/4.5) SRO only 000011G010 ..(KA's) ANSWER: 065 (1.00) REFERENCE: OP-CH-PS-NCP, Rev. 04/09-18-90/GFW, Page 39, Para KA 000015K207 [2.9/2.9) Both RO and SRO , t 000015K207 ..(KA's)
.
ANSWER: 066 ( 1. 0 0)- ,
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._ . __ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _
_ .m. _._.m__.__ . _ .m..__. s O O SENIOR REACTOR OPERATOR Pa90131
.
REFERENCE: r AP/1/A/5500/04, Rotype #7, Page 2, Para KA 000010G010 (3.4/3.4) Both RO and SRO 000015G010 ..(KA's) ANSWER: 067 (1.00) REFERENCEt Catawba Examination Bank Quontion KC-3 KA 000026A105 [3.1/3.1) Both-RO and SRO 000026A105 ..(KA's) ANSWER: 068 (1.00) REFERENCE: OP-CN-EP-CSP, Rev. 11/05-27-92/DPM, Pago 11, Para 2. KA'000029K312 [4.4/4.7) Doth RO and SRO 000029K312 ..(KA's)
< . ANSWER: _069 (1.00) (LO 3 p 7* *
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- _ - _ - _ _ ___ - _ - _ - _ _ _ _ _ _ _ _ - _ _ _ - _ - - - _ _ _ - - _ _ _ _ - _ _ _ _
O O SENIOR REACTOR OPERATOR Page132 l
!
REFERENCE:
,
EP/2/A/5000/01, Rotype /13, page 1, Para 3 j
!
KA 000029G011 (4.4/4.6) Both Ro and Sao i
:
-!
000029G011 ..(KA's)
,
-I ANSWER: 070 (1.00) REFERENCE:
Catawba Examination Bank Question CSF-60 KA 000040G012 (3.8/4.1) , Both RO and SRO ,. 000040G012 ..(KA'a) ANSWER: 071 (1.00) r REFERENCE: AP/1/A/5500/23, Rotype #6, Pago u, Para 8.b.- I
'OP-CN-MT-HT-1, Rev. 04/12-11-90/GFW, Page 28, Para 5. t KA 000051A202 (3.9/4.1) ,
Both Ro and SRO 000051A202 ..(KA's)- -
,
ANSWER: 072 (1.00) pyp **b0AO p(g ue
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O O SENIOR REACTOR OPERATOR Pago133 ;
!
REFERENCE OP-CN-EL-EPL, Pov. 04/01-27-92/CWN, Page 16, Para 2.1. j
,
KA 000057A101.[3.7/3.7) I Both RO and SRO
!
000057A101 ..(KA's)
,
ANSWER: 073 (1.00) REFERENCE Catawba Examination Bank Question EMF-20 KA 000059A102 (3.3/3.4) Both RO and Sho 000059A102 ..(KA's) ANSWER 074 (1.00) b.- REFERENCE: OP-CN-SS-RFY, Rev._ 02/12-11-89/SMD, Page 20, Para 2.2. KA 000067A203 (3.3/3.5) Both RO and SRO-000067A203 ..(KA's)
-ANSWER: 075 (1.00) i "'
l DDPY . - - _ - - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ i O OL i SENIOR REACTOR OPERATOR .Page134 ) i
'
REFERENCE: OP-CH-SS-RTY, Rov. 02/12-11-09/SMD, Pago 26, Para 2.3. ' KA 000067G012 (3.4/3.4) Both RO and SRO
!
! 000067G012 ..(KA's) > i ANSWER: 076 (1.00) REFERENCE:
'
OP-CN-EP-CSP, Rev. 11/OS/27/92/DPM, Page 18, Para k.-& KA 000074K103 (4.5/4.9) < Doth RO and SRO i 000074K103 . . ( }G ' F ) ANSWER: 077 (1.00) ; ; REFERENCE: OP-CN-EP-CSF, Rev. 11/05-27-92/DPM, Page 17, Para i KA 000074K304 (3.9/4.2) Both-no and-SRO 000074K304- . .' (lm 's) ANSWER: 078 .(1.00) a.- 30Il QTi? J
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O .O SENIOR REACTOR OPERATOR - Page135
.
REFERENCEt OP-CN-EP-CSF, Rev. 11 5-27-92/DPM, Page 17, Para KA 000074K311 (4.0/4.4) Both RO and SRO 000074K311 . . (KA's) ANSWER: 079 (1.00) REFERENCE:
,
OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 33, Para 2.3. ; KA 000076G012 (2.9/3.1) i Both RO and SRO 000076G012 . . (KA's) ANSWER: 080 (1.00) REFERENCE: OP-CN-EP-EP1, Rev. 17/03-30-92/RJK, Page 38, Para e EP/1/A/5000/1A1, Retype #9, Page 1e, Para 0. KA 000007A103-(4.2/4.1) , SRO only 000007A103 . . (KA's) . M}f D p ~' "' ~') LAa i wh
' ANSWER: 08 (1.00)
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O O SENIOR REACTOR OPERATOR Page136 , REFERENCE: EP/1/A/5000/01, Rotypo #13, Pago 5,_ Step 3 KA 000007G010 (4.2/4.1)
<
Both Ro and SRO 000007G010 . . (KA's) ANSWER: 082 (1.00) REFERENCE OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 22, Para 2.2.A.1 KA 000008K101 (3.2/3.7) Both RO and SRO 000008K101 . . (KA's) ANSWER: 083 (1.00)
: REFERENCE:
EP/1/A/5000/1A1, Rotypo #9, Pago 6, Stop 11 KA 000009A237 [4.2/4.5) Both-RO and SRO 000009A237 . . (CA's) ANSWER:. 084 (1.00) 5 d .- 10 Q 6 7 P djpgdj{g() P bu(fD? g. .
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O O SENIOR REACTOR OPERATOR Page137 REFERE!1CE: Catawba Examination Bank Question EP1-53 Ya 000011A103 (4.0/4.0) Both RO and SRO 000011A103 ..(KA's) ANSWER: 085 (1.00) REFERE! ICE AP/1/A/5500/12, Rotype #5, Page 1 Para Ya 000022A102 (3.0/2.9) SRo-_only 000022A102 ..(KA's) ANSWER: 086 (1.00)
. REFERENCE:
OP-CN-PC-NCP, Rev.-04/09-18-90, Page 26, Para 2. KA 000022A109 (3.2/3.3) SRO only-
000022A109- ..(KA'a) L-ANSWERt 007 (1.00) -k{QQ
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L _= - . -. . .~. __ . _ _ . ._ -....-._..-.; . - . . . . - - . - . - - - . - - - . - - . _ . - _ - . O O i SENIOR. REACTOR OPERATOR Pago138
.
REFERENCE: , AP/1/A/5500/19, Rotype /19, Page 17, Para C.1 & OP-CN-PS-ND, Rev. 09/02-27-92/GFW, Page 24, Para 2.3. KA 000025G011 (3.6/3.9) Both RO and SRO
!
, 000025G011 ..(KA's) ANSWER: 088 (1.00) REFERENCE: OP-CN-PS-IPE, Rev. 04/10-03-90/GFW, Page 23, Para 2.6. KA 000027A218 (3.4/3.5) Both RO and SRO 1 000027A218 ..(KA's) ANSWER: 089 (1.00) REFERENCE: OP-CN "^-ENB,-Rev. 09/03-19-92/JLY, Page 7, Para 2.1.C. ka-0000320008 (2.8/3.3) SRO only-
,
000032G008 ..(KA's) -
,
MR 7
+
090 MM'"Qn I*i( 7-ANSWER: (1.00) . . - - _ . . . . _ ~ . . .. _ . .,_ ._ . .,.... . , . - ,
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ - - - - . _ - _ _ _
O O ; l SENIOR REACTOR OPERATOR Page139
!
,
REFERENCE: OP-CN-IC-END, Rev. 09/03-19-92/JLY, Pago 9, Para 2.1. ! KA 000033A202 (3.3/3.6) !
>
Both RO and SRO i 000033A202 ..(KA's) , ANSWER: 091 (1.00)
' REFERENCE:
OP-CH-EP-EP4, Rev. 11/12-12-91/PEV, Page 12, Para 2.1. f KA 000038K302 (4.4/4.5) Both no and SRo , A 000038K302 ..(KA's) i
,
ANSWER: 092 (1.00) REFERENCE: AP/1/A/5500/06, Retype #10, Page 12, Para 'KA 000054G012 (3.2/3.2) s Both RO and SRO 000054G012 ..(KA*s) ANSWER: 093 (1.00) fd!q[O'
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SENIOR REACTOR OPERATOR Page140 REFERENCE: Catawba Examination Bank Question EPF-1 . KA 000058A101 (3.3/3.5) Both RO and SRO 000058A101 ..(KA's) ANSWER: 094 (1.00) b.
REFERENCE: OP-CN-SS-VI, Rev. 10/05-21-92/GFW, Page 22, Para 2.1.M.3. KA 000065A201 (2.9/3.2) Both RO and SRO 00006SA201 .(KA's) ANSWER: 095 (1.00) d.
REFERENCE: Catabwa Examination Bank Question VI-55 KA 000065A103 [2.9/3.1) sRo only 000065A103 ..(KA's)
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i:
ANSWER: 096 (1.00) ; ;,s g h
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O O SENIOR REACTOR OPERATOR Page141 REFERENCE OP-CH-SS-VI, Rev. 10/05-21-92/GFW, Page 21, Para 2.1.M. KA 000065G010 (3.2/3.3) Both no and SRO 000065G010 ..(KA'a) ANSWER: 097 (1.00)
- REFERENCE:
'
OP-CN-PS-ILE, Rev. 05/03-23-92/RJK, Page 15, para 2.6.D. KA 000028A202_(3.4/3.8) SRO only 000028A202 ..(KA's) ANSWER: 098 (1.00) REFERENCE: } l OP-CN-EF-EP5, Rev. 08/07-22-91/CWN, page 19, para 2. KA 000056K302 [4.4/4.7) Sao only 000056K302 ..(KA's) 0 T* T' *r,, ly% c?b *
"i . ANSWER: 099 - (1.00)
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.
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O O SENIOR REACTOR OPERATOR Page142 i
!
i.
REFERENCE: , s OP-CN-EP-EPS, Rov. 08/07-22-91/CWN, Page 10, Para 2.1.D.3 & * KA 0000SGG010 (3.7/3.9) l sRo only : 000056G010 ..(KA's)
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(**********-END OF EXAMINATION **********)L
0- . SENIOR REACTOR OPERATOR Page l' ANSWER KEY MULTIPLE CHOICE 023 c 001 b 024 d 002 d 025 b 003 a 026 a , 004 b 027 b 005 a 028 d 096 d 029 a 007 d 030 b 008 b 031 d 009 a 032 a 010 c 033 c 011 a 012 c 035 c
^13 -a 036 b-014 d 037 b '015 c 038 a 016 b 039 MATCHING 017 c a 4
- ~018 c 2 8
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021 b MULTIPLE CHOICE 022 a 040 c i
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-. . , . - . . . . . - . :;; 'O O SENIOR REACTOR OPERATOR .Pag '
A N:S W E R K.E Y 4
,
b
041- d 064 b 042 c 065 b , 043 a 066 c ,
- 044 b 067 -a-045 a 068 c 046 d 069 a 047 'b 070 d 048 a 071 b - 039- c 072 c ,
050 d 073 d
- 051 c 074 b
,
052 a 075 d'
- 053 b 076 b 054- d 077 a 055 d - 078- a : 056 c 079 a 057 a 080 c 058-- d 081 d
-
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059 c 082- b-060 d 083 a z 061 a' 084 d
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062 a 085 c P It " ' " a !#
\f,.[;l;3;; d s 1A - . 063 b 086 a
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- _ _ - - - _ _ - - - - - _ _ - - - _ .
O OL-SENIOR REACTOR OPERATOR Page -3 + ANSWER 'K-E-Y~ _ I 08'F -b 088 a 089 c * 090 d 091 a 092 d
.093 c ,
,
094 b
'095 d 096 a 097 c ,
098 c 099 b
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Wh i' t . 30PY ( * * * * * * * * * * END OF EXVIINATION * * * * * * * * * * )
.- -, -. - - . _ . -
O TEST CROSS REFERENCE, O Paga- 1 --
.
,
. SRO Exam PWR R e a_c:t_o r orqanizod b y- Question Numbor i QUESTION VALUE REFERENCE 001 1.00 9000251 002 1.00 _9000252 ,
003 1.00 9000253 004 1.00 9000254 005 1.00 9000255 006 1.00 9000256 007 1 00 9000257 008 1.00 9000258 009 1.00 9000259 010 1.00 9000261 011 1.00 9000262- , 012 1.00 9000264 013 1.00 9000265 014 1.00 9000266 015 1.00 9000267 016 1.00 9000268 5 017 1.00 9000269 018 1.00 9000270 019 1.00 9000272 020 1.00 9000274 021 1.00 9000275 022 1.00 9000277 023 1.00 9000278 024 1.00 9000279 025 1.00 9000280 026 1.00 .9000282 027- 1.00 9000283 028 1.00 9000284 029 1.90 9000285 030 1.00 9000286 031 1.00 9000287 032 1.0 ' 033 1.00 9000289 034 1.00 900029 .00 -9000294 036 1.00 9000296 037 1.00 9000298 038 1.00 9000299 039 2.00 -9000301 040 1.00 9000302 . _ - . 041 1.00 9000303 p Il D Fa IP P f)- 042 1.00 9000304 g (3[j) { 043 1.00 9000305 044 -1.00 ~9000306 045 1.00 9000308 _-fac -Q ] .f 046 1.00 9000310 - 047 1.00 9000311 048 1.00 9000313 049 1.00 9000314
- . _ -
.
--
.
TEST CROSS REFERENCE Pago 2 S R'O- Exam PWR R e a-c t o r
~
i Organized by Q u e-s t i o n N u m b o QUESTION VALUE REFERENCE 050 1.00 9000316 051 1.00 9000317 052 1.00 9000319 053 1.00 9000321 054 1.00 9000322 055 1.00 9000324 056 1.00 9000326 057 1.00 9000328 058 1.00 9000329 059 1.00 9000330 060 1.00 9000331 061 1.00 9000331 062 ' 1. 0 0 - 9000333' 063- 1.00 9000334 064 1.00 9000335 065 1.00 9000336 > 066 1.00 9000337 067 1.00 9000338 068 1.00 9000339 069 1.00 9000340 070 1.00 9000341-071 1.00 9000342 072 1.00 9000345 073 1.00 9000346 074 1.00= 9000347 075 1.00 9000348 076 1.00 9000349 077 1.00 9000350 078 1.00 9000351 079 1.00 9000352-080 1.00 9000355 081 1.00- 9000356 082 -1.00 9000357 083 1.00- 9000358 084 1.00 9000359 085 1.00 9000361 086 1.00 9000362 087 1.00 9000363 088 1.00 -9000364-089 1.00 9000366 , 090 1.00 9000367 gg'{y" 091- 1.00 9000368- xx g 092 1.00 9000369- ' nd * f"( 093' 1.00 9000370 mg g gy 094 1.00 9000371 - f 095 1.00 9000372 096 1.00 9000373 097 1.00 9000374 098 1.00 9000375
, . . . .. ....- - _ - - -- -- ._ .
CF TEST-CROSS REFERENCE
.
O Page- 3' SRO~ - E x 'a m P! W R R e~a c t'o r O r g a'n 1 z'a d 'b y Question N u m b-e r' QUESTION- VALUE REFERENCE 099 1.00 9000377 i
.. ..
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. . . . . .. . .---_-- - - - . .. . -. .. .
TEST CROSS REFERENCE :Pago S- R 0 Exam PWR R e-a c t Organized by KA G r.o u p
PLANT WIDE GENERICS QUESTION VALUE KA f 010 1.00 194001A104 011 1.00 194001A106 012 1.00 194001A107 033 1.00 194001A109 014 1.00 194001A109 015 1.00 194001A111 016 1.00 194001A116 017 1.00 194001A116 002 1.00 194001K101 001 1.00 194001K101 003 1.00 194003K102 004 1.00 194001K103 005 1.00 194001K104 006 1.00 194001K104 007 1.00 194001K104 008 1.00 194001K105 009 1.00 194001K109 ______ PWG Total 17.00 PLANT SYSTEMS Group I QUESTION VALUE KA 020 1.00 0010000006 018 1.00 001000K103 019 1.00 001000K558-022 -1.00 003000G006 021 1.00 003000K201 023 1.00 004000A401 024 1.00 004010A205 025 1.00 013000A105 026 1.00 014000G008 027 .1.00 015000A303 029 1.00 015000K406 028 1.00 015000K604 - . 030 1.00 017020K101 I , 031 1.00 022000A404 EJLu wtssl*=W 032 1.00 025000K301 033 1.00 026000G005 .), [g,g fg yf 034 1.00 061000A202 4J kJ d _ R _ 035 1.00 063000G010 036 1.00 072000G007 ______ q > - - - -
- . .
- -
TEST-CROSS REFERENCE Pago. 5 SRO E x a-m P:W R: R e a'c t o-r-Organized b y- KA G r: 0 u9 PLANT SYSTEMS Group-I QUESTION VALUE KA PS-I Total 19.00 Group II _ QUESTION VALUE KA 038 1.00 002000K403 037 1.00 002000K603 040 1.00 006000G006 039 2.00 006000K409 042 1.00 010000A403 ~ 041 1.00 010000K201 043 1.00 011000A211-044 1.00 012000K501 046 1.00 016000A301 045 1.00 016000K101 047 1.00 028000A101 055 1.00 029000K302 048 1.00 033000K303 049 1.00 039000A302 050 1.00 064000K411 051 1.00 086000K406 _______ _- PS-II Total 17.00 Group III QUESTION VALUE KA 052 1.00 005000G007 053 1.00 008000K102 054 1.00 008030A304 056- 1.00 041020K417 ______ PS-III Total 4.00 ______ ______ PS Total 40.00 hh b kh EMERGENCY PLANT EVOLUTIONS $ibI * U $ = R Group I
. {j{j{3}f QUESTION VALUE KA
, . . . .
,n ,
\_x
\
TEST CROSS REFERENCE Paga 6 SRO Exam PWR Reactar Organi zed by KA Group EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 057 1.00 000001G007 058 1.00 000001G008 059 1.00 000003G005 060 1.00 000003G010 ' 061 1.00 000005G005 084 1.00 000011A103 062 1.00 000011A104 063 1.00 000011A111 064 1.00 000011G010 066 1.00 000015G010 065 1.00 000015K207 067 1.00 000026A105 069 1.00 000029G011 068 1.00 000029K312 070 1.00 000040G012 071 1.00 000051A202 072 1.00 000057A101 073 1.00 000059A102 074 1.00 000067A203 a 075 1.00 000067G012
"5 076 1.00 000074K103 077 1,00 000074K304 078 1.00 000074K311 079 1.00 000076G012 ,
______ EPE-I Total 24.00 Group II QUESTION VALUE KA 080 1.00 000007A103 081 1.00 000007G010 082 1.00 000008K101 083 1.00 000009A237 085 1.00 000022A102 086 1.00 000022A109 087 1.00 000025G011 .~ 088 1.00 000027A218 p(pqm
' $t
,
-
089 1.00 000032G008 aw; < l' 090 1.00 000033A202 '.* 091 1.00 000038K302 I t idI- .Y! 092 1.00 000054G012 k5ld d 093 1.00 000058A101 095 1.00 000065A103
. . . . . . . . ;
- . . - . . . _ . -. - .
f _ TEST CROSS REFERENCE Paga 7 S4 RO Exam PWR . Reactor -i Organizod by K A- G r-o u p EMERGENCY PLANT EVOLUTIONS Group II
,
QUESTION VALUE KA 094 1.00 000065A201 096 1.00 000065G010 ______ EPE-II Total 16.00 Group III QUESTION VALUE KA
,
097 1.00 000028A202 099 1.00 000056G010 098 1.00 000056K302 ______ EPE-III Total 3.00 ______ ______ EPE Total 43.00 ______ ______ ______ Test Total 100.00
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n ,o V () REACTOR COOLANT SYSTEM ( 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION .
't,3.4.8 The specific activity of the reactor coolant shall be limited to: Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and Less than or equal to 100/E microcuries per gram of gross specific activit APPLICABILITY: MODES 1, 2, 3, 4, and ACTION:
MODES 1, 2 and 3*: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500'F within 6 hours; With_the gross specific activity of the reactor coolant greater than 100/E microcuries per gram of gross radioactivity, be in at least HOT ST ANDBY with T,yg less than 500 F within 6 hours; and The provi:. ions of Specification 3.0.4 are not applicabl > B!! h R O' { }d 4, . . A.,
(f ; e b d.mMJii'l greater than or equal to 500 I ( *With I, 3/4 4-27 Amendment No.25 (Unit 1) I CATAWBA - UNITS 1 & 2 '
Amendment No.15 (Unit 2)
---___________________________________-______________-_________________-________________-____-______________a
. r,. ,m \ b U REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION T - ACTION (Continued) NODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-cu ies per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the spf.cific activity of the reactor coolant is restored to within its limit SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolans shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4- n ;ra t ,
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CATAWBA - UNITS 1 & 2 3/4 4-28 Amendment No.25 (Unit 1) Amendment No.15 (Unit 2)
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y .
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ACCEPTABLE
--
2 . L') , u OPERATION
< 50
m O I w f ' w m l- o l , o t
.
t
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-
-- _ - - - _
.
--
i 0 -- --
-
30 40 50 60 70 80 90 100
I
PERCENT OF' RATED THERMAL POWER FIGURE 3.4-1 DUSE tQuiMLuu i-m .,:iAc,oa waLAH f bet'CIFIC AC71VITY LIMIT VEF2 ' PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC'
{s- . ACTIVITY > 1 pCi/ gram DOSE EQUIVALENT I-131 n ov gPs;;'h my B:,i,.g i t ;,
t
; 5 j ? >{ --
a j; g g ; >;- t j (w 3
:
,,
.,m .p
- - -- 9- ;s.*.
_ CATAWBA - UNITS 1.AND 2 3/4 4-29 $J N E
.
.
TABLE 4.4-4 9 R[ACIOR C09t ANT SPECIflC ACTIVilY SAMPLE N AND ANALYSIS PROGRAM Ea " H0 DES IN WilICil SAMPLE TYPE Of HEASUR[HENT SAMPLE AN3 ANALYSIS At:1) ANAt YSIS FREQUENCY AND ANALYSIS REQUIRED { m Gross Radioactivity Determination ** At least once per 72 hours 1, 2, 3, 4 Isotopic Analysis for DOSE EQUIVA- 1 per 14 days 1 [ LENI l-131 Concentration Radiochemical for 5 Determination *** 1 per 6 months * 1
# #
1,2,3,4,$ # # # g Isotopic Analysis for Iodine a) Once per 4 hours, lecintiing I-131,1-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE EQUIVALENT I-131 E or 100/l pCi/ gram of j,^ gross radioactivity, and E b) One sample between 2 1,2,3 and 6 hours following a TilERMAL POVER change exceeding 15% * of the RATED 111ERHAL POWER within a 1-hour perio O m '
, * - =
g- S
. t% .
ET1 { t, M if R~ m
- , . . ..
~
~
O O TABLE 4.4-4 (Cbntinued)'
.
- r TABLE NOTATIONS
>
#Until the specific activity of the Reactcr Coolant System is restored:
*
within its limits- .
, * Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION =
have elapsed since reactor was last subcritical for 48 hours or longe **A gross radioactivity analysis shall consist of the quantitative measurement of the total-specific activity of the reactor coolant except for radionuclides with half-livss less than 10 minutes and all radioiodines. !The total specific-activity shall be the sum of the degassed beta gamma activity and the tota of. a11' identified gaseous activities _ in the sample within 2 hours af ter the
- . sample is taken and extrapolated back to when-the sample was take Dete r-mination of the contributors to thi gross specific activity shall be based - .upon those energy peaks identifiable with .a 95% confidence-level. -The latest -
available data may be used for pure beta-emitting radionuclide ***A radiochemical analysis for 5 shall consist of -the quantitative measurement of the specific activity for each 'radionuclide,- except_ for radionuclides with - half-lives 1ess than 10 minutes and all radiciodines which is identified in e the reactor coolant. The specifjc activities fgr these individual, radio- _f nuclides shall be used in the determination of_E for the reactor coolant ( sampl Determination of the contributors to E shall be based upon those-energy peaks identifiable with a 95% confidence leve . ji j !1 a yy
! s-L ta l h , -j g }
c .. fat by
u bs ykk u E
,
tATAWRA - IINTTR I A I- 7. /a - 4- 31 -
. ,
, Form 34912 (8-83) PAGE NO, i , CNS AP/1/A/5500/06 LOSS OF S/G FEEDWATER Retype #1r A
-
.
.
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TABLE OF CONTENTS
'
E!!1n A. Purpose ,
B. Symptoms 1 Case Loss Of CF Supply To S/Gs Case II." Loss Of Normal CA Supply Case Loss Of CF Supply To S/Gs Operator Actions 2 Immediate Actions 2 Subsequent Actions 2 Case I Loss Of Normal CA Supply Operator Actions 5 Immediate Actions 5 i Subsequent Actions 5 Enclosure 1_- Continuous Monitoring of CA Pump Parameters 12 Enclosure 2 - CA Pump Suction Pressure Based on Hotwell 17 Level and Flowrate Vd? 0(-] ?l D ta:w a 2 i COPY
. .
, Forrn 34912 (8 83) g
*
PAGE NO, CNS - AP/1/A/3500/06 LOSS OF S/G FEEDWATER 1 of-17 Retype #1C ua-
.
A. RyRPOSE o To verify proper response to a loss of feedwater supply to the S/Gs o To verify proper response to a loss of normal supply of auxiliary feedwate B. E_YMPTOMS
.
Case I . -Loss Of CF Supply To S/Gs o CFPT A and B - TRIPPED o S/G A (B,C,D) LO LEVEL ALERT alarm (s) (IAD-4) - LIT o S/r A (B,C,D) FLOW MISMATCH L0 CF FLOW alarm (s) (1AD-4) - LIT, CASE II. Loss Of_ Normal CA Supply o "CACST Lu LEVEL" alarm (IAD-5, H-4) - LI Unp
'
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'
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'"
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.
y , g;)lhi; kt r Page 1 of 17
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. Form 34913 (8-82) : V - J PAGE N .CNS~
CASE I. . LOSS OF CF SUPPLY TO S/Gs -2 of 17: a
'
AP/l/A/5500/06 Rotype #1r- [. n
.
>
RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE C. OPERATOR AGI!ONS 15ESJAIK_A_qT10M
.
None , o HQJI MEU_ ACTIONS Verify Reactor power - LESS Perform the following l _ * THAN 5%. _ Manually trip Reacto _., b . GQ Ig EP/1/A/5000/01, Reactor Trip or Safety injectio . Verify Teedwater Isolation Manually-initiate-Feedwater .; an indicated by following Isolation , status lights (1SI-5) - LIT: __. o TRN A
._. o "S/G A CF CONT ISOL _o TRN VLVS CLSD"
_o "S/G B CF CONT ISOL VLVS CLSD" e _o "S/G C CF CONT ISOL ' VLVS CLSD" _o "S/G D CF CONT ISOL VI.VS CLSD."
Verify total CA flow -- Perform the following: _ GREATER TilAN 8504 CP _ Manually start CA purnp , II'CA flow cannot be ,
-established, THEN 00 TO EP/1/A/5000/2C1, L'oss of Seconda ryl Ilea t. Sink .
"
d nyp n
" * 9 : yjy
-
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.f Page 2 of 17
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Form 34913 (8 82) h 1 PAGE N CNS AP/1/A/5500/06 CASE LOSS OF CF SUPPLY TO S/Gs 3 3f 37 Rotype'#10
"~
,
.
sa RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE Control S/0 levels: __ Verify S/G N/R levels - __ Maintain total feed flow AT 38%. greater than or equal to 450 gp EllE.ti N/R level increasing in at least one S/G, THEN depress CA System
. valve control " RESET" pushbuttons:
__ o TRN A __ o TRN __ Throttic feed flow to maintain S/G N/R levels at 38%. Verify CA supply - ADEQUATE:
, Ensure the following valves - OPEN:
__ o ICA-4 (CA Pmps Suct From UST) __ o ICA-6 (CA Pmps Suct '-
~
From CA CST).
__ Verify "CACST LO LEVEL" Perform the following: alarm ( I AD- 5 , 11-4 ) - DAR __ 1) Ensure ICA-6 (CA Pmps Suct From CA CST) - CLOSE __ 2) REFER TO Case II, Loss of Normal CA Supply and perform applicable steps concurrently with
~
romaindor of this case.
, y .h S
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,
Page 3 of 17 bOS i
- ., . .. . -
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,
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Fe'rm 34913 (8-82)
PAGE NO.- CNS 4 oi ;7' AP/1/A/5500/06 -CASE I, LOSS OF CF_ SUPPLY.TO S/Gs Rotype #10
'3 It SPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE
: Determine required notifications:
.
.__ o . REIIB LO RP/0/A/5000/01, C' asification of Emergency .
-
,,_., o REFER LO RP/0/B/5000/13, NRC Notification Requirement _ Determine and correct cause of -
losi of CT suppl _ Determine long term plant statu BETURN 10 procedure in effect, E1D
, >
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~
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.
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Nyk$k Page 4 of 17 f
- . , , - . . _ . ,, -. - , . , . -
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L' Form 34913 (8 82) PAGE N CNS CASE II. LOSS OF NORMAL CA SUPPLY- 5 of 17 AP/1/A/5500/06 Retype #10 C
.
RESPONSE NOT OBTAINED - ACTION / EXPECTED RESPONSE _ C. 9PERATOR ACTIONS LMBQLALC A91 ion 5
.
None
,
SUBSEQUENT ACTIONS H_0TE - Closing ICS-69 will isolate the CACST f rom the Unit 1 and Unit 2 CA Pump _ Ensure 1CA-6 (CA Pmps Suet _ Dispatch operator to unlock and Trom CACST) - CLOSE close 1CS-69 (CACST To Unit 1 and Unit 2 CA Supplies) (SD 619, T-26). Initiate makeup to UST: __ o Pump CST to UST with CST Pumps 1A and 1 _o Open lYM-100 (UST M/V CTRL).
_ Verify at least one hotwell CO 7,9 Step 1 pump is available for maGup to the US .
'y* O A " "' ]
:) -
$ d
'[L,;_ , .; 0 buiP'll
. Page 5 of 17
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__
- ~.
Foim 34913 (8-87) g PAGE NO, CNS CASE II. LOSS OF NORMAL CA SUPPLY 6 of 17 AP/1/A/5500/06 Rotype #1r
.
' RESPONSE NCT OBTAINED ACTION / EXPECTED RESPONSE 4 Makeup to UST from hotwells
' Dispatch operator to:
'
__ 1) Open ICM-35 (110twell Hi Lvl Ctrl Byp),
(TB-575, IL-25).
2) Ensure the following
,
valves - CLOSED: __ o ICS-24 (Normal Hotwell Makeup Control Inlet)
(TB-595, 1G-30)
o ICS-60 (Normal Hotwell Makeup Control Bypass)
(TB-596, 1G-30)
__ o ICS-32 (Hotwell Recirc Makeup Valve Inlet)
(TB-573, IG-29)
__ o ICS-35 (Hotwell Recirc Makeup Valves Bypass)
(TB-578, IG-29). IE all hotwell pumps are-of f , THEN:
__ 1) Notify Chemistry to ensure all CM polishing domineralizers isolate . 2) Ensure the following valves - CLOSED: __ o Polish Demin Byp Ct rl
._ o ICM-83 (Gen Load , . , y r '
Reject Bypass). F n-gy3?" !;* 7(4 D
, n la e t
__ Start and stop hotwell pump (s) .
. :g: <t
--
as required to makeup from 5 1 y id )$/
'
hotwel (I%d 5 N Page 6 of 17
-1 '
Form 34913 (B 82) -' - PAGE N t g g ._ AP/1/A/5500/06- CASE II. LOSS OF NORMAL-CA SUPPLY- _7..~ o f 17 ' Retype #10
.
.
ACTIOf XPECTED RESPONSE RESPONSE NOTOBTAINED jion o If the CA pumps are taking a suction on the hotwell with the UST depleted and the CA auto start circuitry is actuated, then the CA pump suction will automatica11y_ align to' the - RN_ assured makeup sourc , o If CA has been reset and CA pump suction is aligned to the hotwell wit'a the UST-depleted, then the CA pumps will trip on low suction pressur . Verify UST level -' ADEQUATE:
,_._ o "CACST/UST LO LEVELS" o Break condenser vacuum, alarm (IAD-5, H-6) -
DAR _ H hotwell level greater than 0.5 ft, THEN RETURN TO Step ___ 7 . H]iKH hotwell level less than 0.5 ft AND "CACST/UST LO LEVELS" alarm (1AD-5, H-6) lit, T]iEN close 1CA-4 (CA Pmps Suet From UST).
~ l: l
)h
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e
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L Page 7 of 17 l t
,- , , _ _ _ _ _ _ _ _ - -: _
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O Q~
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Form 34913 (B 82) PAGE N CNS CASE _II, LOSS OF NORMAL CA SUPPLY a of 17-AP/1/A/5500/06 Retype /11'
.
RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE _ Ensure CA pumps suction aligned to RN: o TRN A valvas - OPEN: .
.__ o 1RN-250A (RN Hdr A To CA Pmp Suct Isol)
__ o ICA-116A (CA Pump //1
- Suct Frm RN Hdr A)
_o ICA-ISA (CA Pump 1A Suct Frm RN Isol), o TRN B valves - OPEN: _o ICA-85B (CA Pump (l1 Suct Frm RN Udr B)
,__ o ICA-18B (CA Pump 1B Suct From RN Isol)
_o 1RN-310B (RN Hdr B To i CA Pmp Suct Isol), [ " SET CA XTER TO RC" alarm (1AD-5, G-4) lit, ! THEN ensure CA Pump #1 cuetion aligned to RC: l
.__ o ICA-178 (RC Supply To CA Pump- Isol) - OPEN
_o ICA-174 (RC To CA Suct Isol) .OPEN ___ o ICA-175 (RC To CA Suct Isol) - OPE . Determine long term plant' o Continue monit- ing CA Idbs # ' pump statu , Di}if
: S
_. o RETURN To procedure in effect.
I Page 8 of'17
.m * -
Form 34913 (8 83)
-w PAGE N CASE 11. -LOSS OF NORMAL CA SUPPLY 9 of 17 AP/1/A/5500/06 Retype #10 j .
RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE NOTE The CA pumps will trip on low suction pressure with CA rese and CA pump suction aligned to the hotwell when the UST is deplete . Establish operator control of CA System __ [[ CA Pump #1 is needed,
' TtiE_ff place CA Pump #1 switch !n "0N" Depress CA System valve control " RESET" pushbuttons:
_o TRN A _o TRN B.
.,
-
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.
Page 9 of 17
..,. = -. . .- . . -. .- - ,
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; F'orm 34913 (8 82) - ,
PAGE N CNS AP/1/A/5500/06 CASE II. LOSS OF NORMAL CA SUPPL ' 10 -_ o r .17 ,, Retype #1'- _
:- - ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED -
1 Maintain all S/G N/R levels at 38X by using one CA pump at a times
'
o IE one motor driven CA pump available, TlLEE dispatch operator to open the following valves: o ICA-111 (CA Pump 1A & 1B Disch Xover To S/G Isol) (AB-552, BB-50, Rm 250)
,_ o ICA-112 (CA Pump 1A & IB Disch Xover To S/G Isol) (AB-552, BB-50, Rm 250).
OR o LE CA Pump #1 available, THEN open the following valves: _, o ICA-66B (CA Pump 1 Disch To S/G 1A Isol) __ o ICA-38A (CA Pump 1-Disch To S/G 1D Isol).
NOTE The CA pump trip on low suction pressure is being defeated in the following step so that the hotwell can be used as a CA suction source, . I ___ 1 Initiate SWR 10403 to have [$p q IAE. defeat the-CA pump trip f f[*f.Ui,- 4, [* Pt
.
on low suction pressure on all:3 CA pumps, j?b g ,. Y b'h Y . Page 10 of 17 _ _
..
Form 34913 (8 82);
._
PAGE N CASE II. LOSS OF NORMAI,CA SUPPL 1 11 of.17 AP/1/A/5500/06 Retype #10 t' l ._:, '
.
ACTION / EXPECTED RESPONSE - RESPON5E NOT OBTAINED
.1 Defeat the automatic swap of CA pump suction to the RC assured makeup source by dispatching operator to '
open the following breakers at the SST (Elev. 611): __ a) SDSP1 Bkr #4 __ b) SDSD-F01D (RC Supply To Aux Feedwater Isol Valve ICA-174) __ c) SDSD-F02C (RC Supply To Aux Feedwater Isol Valve ICA-175).
__ 1 Continuor:1y monitor the parameters given on Enclosure 1 and perform any required action __ 1 Determine long term plant statu RETURN 70 procedure in effec E!LD _ l
.
Jsne} se3 e 1.si h(J 24!hhfh N Page 11 of 17
_ .
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Form 34913 (8-82) () PAGE N CNS LOSS OF S/G FEED'dA7t.R ' AP/1/A/5500/06 ENCLOSURE 1 - Page 1 of 5 12 of 17 Retype #h n
-
ITESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE CONTINUOUS MONITORING OF CA PUMP' PARAMETERS NOTE If the CA pumps are taking a suct'on on the hotwell with the UST depleted and the CA auto start circuitry is actuated, then the CA pump suction will automatically align to the RN assured makeup sourc NOTE The preferred method of determining CA pump suction- pressure is to compare hotwell level and CA pump flowrate (Enclosure 2). Continuously monitor CA pump suction pressure using any of the following methods: i _o Hotwell level and CA pump flowrate , B_EJJ3. TO Enclosure 2
,
3
._ o Pressure gauges on IMC-4 l _o Local pressure gauges in the CA pump roo [M[[[{
h Q!T[@*!
~ "[T'J"'
f[1 y I_I'CA pump suction pressure drops to less than or equal PU O3 P* F "
-
to 8.2 psig, IJLEN perform the following: _, a . Stop all CA pump _ b.- Close ICA-4 (CA Pmps __ Dispatch c>perator to unlock Suct From UST). and close ICS-19'(CAl Pumps Supply From UST) (TB-620, ID-30). i e Page 12 of 17
:
. .. . -.. - . - . . . . . . _- .- _ - . . . __ -~ -. . . - . . . _ . . . . -. . '
F$rm 34013 (8 87) l
^ "U'
CNS LOSS OF S/G TEEDWATER ! AP/1/A/5500/06 r.:: CLOSURE 1 Page 2 of $ 13 of'17 i Retype #10 n
.
ACTION / EXPECTED RESPONSE RESPONSE NOT ODTAINED
, (Continued) .
, , LE RN availabic, IljK!f align -
'
CA pump suction to the RN - assured makeup sources: 1) Alig: A TRN RN supply ' by opening the '_ following valves:
,
_o IRN-250A (RN lidr A To CA Pep Suct Isol) , _o ICA-116A (CA Pump #1 Suct Frm RN lldr A)
.__ o ICA-ISA (CA Pump 1A Suct Frm RN 18o1),
2) Align B TRN RN supply by opening the following ' valves: t
,_ o ICA-85B (CA Pump #1 Suct Frm RN Hdr B) ,
_o ICA . s (CA Pump IB
., Suct Frm RN-1801)
_o IRN-310D (RN lidt B To CA Pmp Suct I r.o l ) . h.ht ,bh paw a L
>T (b d}fj\$ $I Pa r,e 13 of 17
- - .-_ . . _ - __ . _ _ .. . _ _ . . ._. - . _ . . - _ _ _ _ . _ . -
. _ ,
4' otrn 34913 (8 82) ,
!
PAGE N CNS I,0SS OF S/C TEEDWATER AP/1/A/5500/06 ENCLOSURE 1 - Page 3 of 5 14 of 17 ! Retype I/) l R t i RESPONSE NOT *)DTAINED ACTibrd/ EXPECTED RESPONSE (Continued)
! Restore the automatic swap of CA purnp suction -
[
to the RC assured makeup ;
'
source by dispatching operator to close the following breakers at
, the SSF (Elev. 611): .__ 1) SDSP1 Bkt (14
_ 2) SDSD*F01D (RC Supply To Aux Teedwater Isol Valve ICA-174) _ _ 3) SDSD F02C (Rt Supply , To Aux Feedwater Isol Valve ICA-17 Start CA pump (s) as necessary to maintain S/G N/R levels at 38%. _ 3, MilEE the "CACST/UST LO . LEVELS" alarm (1AD 5, H 6) is lit, 11153 break Corutenser vacuu Of 0
-
Ly p ;)@'.g . f,= g(. 00PY Page 14 of 17 -
,
.r4e cc, . . , = e i- ,- , .-_,r-- r-- ,. -
----,,-----,,.-,w--m, - . ,
_ _ _ - - - _ _ - - _ _ _ _ _ _ - - - - - - _ _ _ _ _ _ - - _ - - - . - - - -.---- ._ - - - .
,
i
' '
' FCrm 34913 (8 82)- !
,
PAGE N LOSS OF S/G FEEDWATER { CNS AP/1/A/5500/06 ENCLOSURE 1 - Page 4 cf 5 15'of 17 : Retype fl10 ,
,
'
. .
?
RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE
,
I Fj H hotwell level is less i 84 . than or equal to 0.5 ft, EM perform the following: *
.
1 Stop all CA pump ;
._, b . Close 1CA-4 (CA Pmps-Suct Dispatch operator to unlock l From UST). and close 105-19 (CA Pumps i Supply From UST) (TD-620, l
*
10 30)'.
~
i II RN available, E M - !gn CA pump suction to the RN . assured makeup sources: 1) Align A TRN RN supply Fy opening the : following valves: l -
;
,_ o 1RN 250A (RN lldr A 3 To CA Pmp Suct Isol)
_o ICA-116A (CA Pump //l i Suct Fim RN lldr A) > _o- ICA-15A (C/ Pump 1A Suct Frm RN 1 sol).
~ 2) Align B TRN RN. supply by l
-
-
opening the tallowing r valves I &
,_, o - 1CA-85B (CA Pump //1- ,
Suct_Frm RN lidr B) . _o 1CA 18B (CA Pump ID Suct Frm RN lsol) _o 1RN-310B-(RN lldr D g
- To CA Pmp-Suct-Isol). w -
P .
.
)
-- fEA <a {
>
fi!Qy y . y ly ' _-
.
Pago.13 of 17 ,
,
F
-. . - ; , _ _ - . - . - . _ . _ . . . . - . . , . , . . . . . . - - , - ._-,_.,,,,..,_.__,;,n,-
l Form 34913 (0 82) h
#U " U'
CNS 1,055 0F S/G FEEDWATER ENCLOSURE 1 - Page 5 of 5 16 of 17 AP/1/A/5500/06 Retype (1)
-
I
~
-
!
RE$PONSE NOT OBTAINED l ACitt *: MCTED RESPON$E
. . (Continued) Restore the automatic swap of CA pump suction -
to the RC assured makeup . source by dispatching operator to close the following breakers at
.
the SSF (Elev. 611): , _1) SDSP1 Ilkr (14 _. 2 ) SDSD-F01D (RC .- Supply To Aux Feedwater Isol Valve ICA-174) ;
._ 3 ) SDSD-F02C (RC
,
Supply To Aux Feedwater Isol Valve 1CA-17 ,
*
,_ Start CA pump (s) as necessary to nalntain ,
>
S/G N/R leveln at 38*. O':0!E& l ff_&
,
.
.-
Page 16 of 17 i l
+
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- - ,. , . . .
.- . . .... - . - - . .- . . ~ - .-. . -._ - - _ . .. . .. _
- .
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' Form 3W2 (0 631
, . _ _ w a-
: t r -- - - PAGE N i CNS LOSS OP S/G TEED %lATER ,
ENCLOSURE 2 Page 1 of 1 17 of 17 l AP/1/A/5500/06 Retype #10 ,
. .
t 9A._f3R SUCTION _ PRISSyBI BA5KD ON H0TWKl.L_ LKIKL_.AN.R ff,JERME I
.
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!!91% o This graph assumes that the CA pumps are aligned to the hotwell - and that the CACST and UST are depleted, i
o This graph is valid for either one or two CA pumps taking n suction.on the hotwell with the UST deplete Q A p vp
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.
O O t %, 3/4.3 INSTRUMENTATION L 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION
;
LIMITING CONDITION FOR OPERATION i
3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3- APPLICABILITY: As shown in Table 3.3- ACTION: As shown in Table 3.3- s SURVEILLANCE REQUIREMENTS
4.3.1.1 Each Reactor Trip System instrumentation channel and interlock 'and the automatic trip logic shall be demonstrated OPERABLE by the performance of ] the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1, 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months."
Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels .,
*
are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the
" Total No. of Channels" column of Table 3.3-1. The response time of RTDs associated with the Reactor Trip System shall be demonstrated to be within their limits at least once per 18 month [ b i? Y l'
W g0 J Cor7
*This surveillance need not be performed for the primary RTO response time ' '.
testing portion of items 7 and 8 from Table'3.3-2 for Unit 2 until prior to entering STARTUP following the Unit 2 first refuelin CATAWBA - UNITS 1 & 2 3/4 3-1 Amendment No. 40 (Unit 1) Amendment No. 33 (Unit L)
-- - - . . ____ __----
,. __ g TABLE 3.3-1
> ..
.
..
75 ,. .c -
REACTOR TRIP ~ SYSTEM INSTRUMENTATION
' h T? : , t --
r MINIMUM r?E- TOIAL N CHANNELS CHANNELS APPLICABLE -
-U FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NODES ACTION KV , =8 Manual. Reactor Trip 2 1 2 1, 2 1 y -
2 1 2 3* , 4 * , 5* 10 g$'7[ t
, .>
r Power Range, Neutron Flux
- Hig). Setpoint Lov Setpoint
4
2
3 1, 2 17##, 2
2 g Power Range, Neutron Flux 4 2 3 1, 2 2 i High Positive Rate g Power Range, Neutron Flux, 4 2 3 1, 2 2 l
* High Negative Rate E Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 Source Range, Neutron Flux Startup 2 1 2 2ft 4 Shutdowm 2 1 2 3* , 4 * , $* 10 gg Overtemperature AT =g Four Loop Operation 4 2 3 1, 2 6 l @@ l e= - Overpower AT 55 Four Loop Operation 4 2 3 1, 2 6 l
-- Pressurizer Pressure-Low 4 2 3 1 6** l NE -
22 ra
==_> ** (~ := .--. , A' .$ + ) .
w my 2 E e t
*
.
r ! -
'
~ ~ s ,
.
TABLE 3.3-1 (Continued) .
,
' y REACTOR' TRIP SYSTEM INSTRUMENTATION
#
MINIMUM E TOTAL N CHANNELS CHANNELS APPLICABLE
] . FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
[, 1 Pressurizer Pressure-High - 4 2 3 1, 2 6** l
-[ 11. . Pressurizer Water Level-High_ 3 2~ 2 1 6 l
'- 1 Reactor Coolant Flow-Low . . Single Loop (Above P-8) 3/ loop 2/ loop in 2/1 cop in 1 6 ! ,
'
any oper- each oper-ating loop ating loop Two Loops (Above P-7 and' 3/ loop 2/ loop in 2/ loom 1 6 l w below P-8). two oper- eac) per- "
'l , ating loops ating loop
-
, . . .
.O 1 Steam Generator Water 4/sta 2/str gen 3/sta gen 1. 2 6** l Level--Low-Low
'
gen in any each
.
.
*
operating operating 1 - sta gen sta gen
,- 14. .Undervoltage-Reactor Coolant 4-1/ bus 2 3 1 6 l gy Pumps (Above P-7) 'o 5 ' t , 'y'
e 1 Underfre Pumps (Above -7) Reactor Coolant 4-1/ bus' =2 3 1 6 l h
?y , -
xi 1 Turbine Trip '
? P4 t R a.t.-' 'Stop Valve EH Pressure- 4 2 3 1MM 6 l O$ "
-: Low . .. _
. Turbine Stop Valve Closure
~ INN 11'
.EE l
~
' l { 1 Safety Injection Input from ESF g pg
. .
1 2 1, 2 9 vv yl . sg N * k ,h/- es ff
'
q;
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% p g,7
'
i
-- --
, _ _ - _
r TABLE 3.3-1 (Continued)
,O REACTOR TRIP SYSTEM INSTRUMENTATION .~ p ; - ,7 7Y' , ,. -
MI;4IPUM
;. - - -
TOTAL N CHANNELS CHANNELS APPLICABLE
~-l e FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERASLE MODES ACTION i E3 gog 1 Reactor Trip System Interlocks ,. Intermediate Range - Neutron Flux, P-6 2 1 2 2D 8 h;N 2 $ Low Power Reactor [[ ' Trips Block, P-7 .',- .
P-10 Input or 4 2 3 1 8
~
g' P-13 Input 2 1 2 1 8 c.'lPowerRangeNeutron 51ux, P-8 4 2 3 1 8 '
) - ,
w Fower Range Neutron 4 2 3 1 8 i Flux, P-9 -
- Power Range Neutron-Flux, P-10 4 2 I 8
. Power Range Neutron Flux, Not P-10 4 3 '4 1, 2 , 8
, Turbine Impulse Chamber 3
-
Pressure, P-13 2 1 2 1 8 W 1 Reactor Trip Breakers 2 1 2 1, 2 9 2 1 2 3*, 4*, 5* 10 l 20. Automatic Trip and Interlock 2 1 2 1, 2 9 ! Logic 2 1 2 3* , 4* , 5* 10
-
)
Nm i 3 ' Jg l
%I m ,
r (- ( .
~ ,
f h
'
_ . _ - - _ - _ - _ _ - . _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ O O . J
, TABLE 3.3-1 (Continued)
TABLE NOTATIONS
*0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive system is capable of rod withdrawa ** Comply with the provisions of Specification 3.3.2, for any portion of the channel required to be OPERABLE by Specification 3. #GBelow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint ###Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoin ####Above the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoin .
ACTION STATEMENTS r ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the .iext 6 hour ACTION 2 - With the number of OPERABLE channels one less than the Total Numbe.* of Channels, STARTUP and/or POWER (PERATION may proceed 4 provided the following conditions are satisfied: The inoperable channel is placed in the tripped condition '
.
within 6 hours, ,' The Minimum Chsnnels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Spec;ticati .3.1.1, and Either, THERMAL POWER is restricted to less thari or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux trip setpoint is reduced tu less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at '.aast once per - 12 hours per Specification 4.2. ACTION 3 - With the number of channels OPERnBLE one less than the Minimum Channels OPERABLE requirement and wi',h t..e THERMAL POWER level:- Below the P-6 (Intermediate Range Neutron Flux interlock) Setpoint, restore the inoperable channe. to OPERABLE status prior to increasing THERMAL POWE' above +he P-6 Setpoint; or (?,7,4":
.
4-
-
at" ,I '. 1* u Above the P-6 (Intermediate Range NeuM q f ua intenlock) W2 c5m Setpoint but below 10% of RATED THERMAI. POWER. restorc the ( ,, ffE3 5d inoperable channel to OPERABLE status prior to increasing y)j,3% THFRMAL FGWER above 10% of FATED THERMA POWER.
,
,
CATAWBA - UNITS 1 & 2 3/4 3-5 Amendment No.48 (Unit 1) Amendment No.41 (Unit 2) _ . - _
' O O TABLE 3.3 1 (Continued) ACTION STATEMENTS (Continued) ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity change ACTION 5 - Delet<! ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: The inoperable channel is placed in the tripped condition within 6 hours, aid i The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3. ACTION 7 - Delete ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive status light (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3. ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, - a proviced the other channel is OPERABL g 7,,
/C110N 10 - With the number of OPERABLE channels one less than the Minimum C- O Channels OPERABLE requirement, restore the inoperable channel % ' **
to OPERABLE status within 48 hours or open the Peactor trip breakers within the next hou ad (* ra p
AC110N 11 - With the number of OPERABLE channels less than the Total Number 3 of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hour r ACTION 12- With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION , The bitaker shall not be bypassed while one of th', diverse trip i features is inopeiable except for the time required for perform- j
-
ing maintenance tu restore the breaker to OPERABLE statu Witn I the breaker bypassed, apply AtTION l
rCTION 13- With any reactor trip bypads breakd inoperable, restore the bypass breaker to OPERABLE status prior to placing it in servir CATAVBA - UN!TS 1 6 r 3/ n
. * A. endment No.63 (Unit 1)
Amendment No.57 (Unit 2)
?
u n ; i TABLE 3.3-2 k REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
$ . FUNCTIONAL UNIT RESPONSE TIME C
5 Manual Reactor Trip v1
~ Power Range, Neutron Flux < 0.5 second*
e-m Power Range, Neutron Flux, High Positive Rate . Power Range, Neutron Flux,
. High Negative Rate < 0.5 second* .. Intermediate Range, Neutron Flux N. #M Source Range, Heutron Flux '" Overtemperature AT ' # .
5 4(0 ) seconds * l Overpower AT # 5 4(8 ) seconds l Pressurizer Pressure-Low '
<
_ 2 seconds 1 Pressurizer Pressure-High ~
-< 2 seconds gg (% yD @@ 1 Pressurizer Water Level-High p+ ~ ^' .. a # yg g gg >
_ ee .~ _s z z s? ?
.
w :
, .
w%.
*U $ -
22 * Neutron detectors are exempt from response time testin Response time of the neut6on flux signal portion
,11 of the channel shall be measured from detector output or input of first electronic component in channe ' " "
# Applicable upon deletion of RID Bypass Syste l
.-
.
.- TABLE 3.3-2 (Continued) REACTOR TRIP SYSTEM INSTRtMENTATION RESPONSE TIMES si e FUNCTIONAL UNIT RESPONSE TIME C 5 1 Low Reactor Coolant Flow
$
w Singic Loop (Above P-8) $ 1 second e- Two Loops (Above P-7 and below P-8) $ 1 second u 1 Stear Generator Water Level-Low-Low g Unit 1 < 3.5 seconds Unit 2 {2.0 seconds 1 Undervoltage-Reactor Coolant Pumps 5 1.5 seconds y 15. Underfrequency-Reactor Coolant Pumps $ 0.6 second
[ 1 Turbine Trip Stop Valve EH Pressure-Low l Turbine Stop Valve Closure . Safety Injection Input from ESF . Reactor Trip System Interlocks [[ ==
1 Reactor Trip Breakers- g
[[
sa
[[ 2 Automatic Trip and laterlock logic ??
W mm 3 * 33 C7; 2 05 C N% ms v v kr./ r . ,. .,j ( %$ l 'it 'y -, w mv w - t _ _ . . . .
._
.
TABLE 4.3-1
"
>
$ REACTOR IRIP SYSTEM INSTRUMENTATION SURVEILLA.NCE REQUIREMENIS
$
# TRIP
ANALOG ACTUATING MODES FOR Cl!ANNEL DEVICE milch E ACTUATICI* SURVEILLANCC O CHANNEL CHAN.. s+
g 4 (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWE O Y provisions of Specification 4.0.4 are not appilcable for entry into MODE 2 or The
;$*g (7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASI (8) With power greater than or equal to the interlock setroint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive status ligh (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifi-cation that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive status ligh (10) Setpoint verification is not applicabl (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifi-
. cation of the operability of the Undervoltage and Shunt trip (12) Deleted (13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Steam Generator Water Level Low-law is adjusted to a value less than or equal to 1.5 seconds.
(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual l Reactor Trip Function. The test shall also verify the OPERABILITY of the t
Bypass Breaker trip circuit (s).
t CATAWBA - UNITS 1 & 2 3/4 1-12 Amendment No. 63 (Unit 1) t *
. - i .
Amendment No. 5/ (Unit 2)
o O O
'
TABLE 4.3-1 (Continued) ,
\ ,* TABLE NOTATIONS l (15) A local manual shunt trip on the bypass breakers shall be performed prior ,
to placing breaker in rervic ,
(16) The automatic undervoltage trip capability shall be verified OPERABL ,
.,
l ,' ,
,
i i I e
,
t i
.
E CATAWBA --UNITS 1 & 2 - 3/4 3-12a Amendment No. 6* (Unit- ? ) Amen,dment N s. 57 (" iit 1) :
,
e..-- ,.- . , _ , , . - , - . . , , . , , - . . ,,, _ - _ - _ _ _ _ - _ _ _ _
-
O O CONTAINHENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND C00 LINO SYSTEMS 'r CONTAINHENT SPRAY SYSTEM
.
LIMITING CONDITION FOR OPERATION 3.6.2 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the refueling water storage tank and transferring sunion to the containment sum APPLICABILITY: H0 DES 1, 2, 3, and AC, TION: With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status withir 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hour SURVEILLANCE REQUIREMENTS
~
l
-
4.6.2 Each Containment Spray System shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked,
'
sealed, or otherwise secured in position, is in its correct position;
/M ,,
i By verifying, that on recirculation flow, each pump develops a differential pressure of greater than or equal to 185 psid when
/'/ ' / j Q 3
. , , Ogg tested pursuant to Specification 4.0.5; e ' ' ig
' At least once per 18 months during shutdown,"* by: f f 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Phase S" Isolation test signal, and
'
2) Verifying that each spray pump starts automatically on a Phase "B" Isolation test signal, 3) Verifying that each spray pump is prevented from starting oy the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid, and is allowed to start at greater than or equal to 0.45 psid relative to the outside atmosphere,
,
a*This surveillance need not be perfomed until prior to entering HOT SHUT 00VN following the Unit i first refuelin CATAVBA - UNITS 1 & 2 3/4 6-18 j
,
- _ _ _ _ _
__. _ _ _ _ _ _ _ _ - - _ - _ _ - _ - - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ - _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ o i O :
-
-
.
CONTAINMENT SYSTEMS 1
.
SURVE!LLANCE RE0VIREMENTS (Continued)
,
4) Verifying that esch spray pump discharge valve closes or is prevented from cpening by the Containment Pressure Control , System when the containment atmosphere pressure is less than or equal to 0.25 ,nsid and is allowed to open at greater than or equal to 0.45 psid relative to the outside atmosphere, and
. 5) Verifying that each spray pump 1s automatically deenergized -
by the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid relative to the outside atmosphere, At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed, ;
.
,
CATAWBA - UNITS 1 & 2 3/4 6-19 ,
l .
..
o o ; REACITV!YY CONTROL SYSTEMS ! N P* 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 42fy&] *b '
- '
,
LIMITING CONDITION FOR OPERATION l
!
3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and : positioned within 112 steps (indicated position) of their group step countt:r + demand positio APPLICABILITY: MODES 1* and 2*. ACTION: i With one or more full-length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-- ment of Specification 3.1.1.1 is satisfied within I hour and be in HOT STANDBY within 6 hour Vith more than one full-length rod misaligned from the group' step counter demand pnsition by more than 112 steps-(indicated position), be in HOT STANDBY'within G hour With one full-length rod-trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than 112 steps (indicated 3 J-position) POWER OPERATION may continue provided that within 1 hour: -j The rod is restored to OPERABLE status-within the above alignment requirements, or The rod is declared inoperable and the remainder of the rods in-the group with the inoperable rod are aligned to within-i 12 steps of the inoperable rod while maintaining the Jod sequence and -- insertion limits of Specification -3.1.3.6.. The THERMAL- POWER l: level shall .be restricted pursuant to Specification 3.L 3.6 during subsequent operation, or The rod is declared inoperable and the SHUTDOWN MARGIN require-ment of Specification 3.1.1;1 is satisfied.- POWER OPERATION _~ may then continue provided that: a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days;.this reevaluation shall confirm-that the previously analyzed results of these accidents-remain valid for the duration of operation under these conditions;- - b) The SHUTDOWN MARGIN: requirement of Specification 3.1.1.1 is - determined at least once per 12 b .rs;
*See Special Test Exceptions Specifications 3.10.2 and 3.1 ].
J CATAWBA - UNITS 1 & 2' 3/4 1-14- Amendment No. 74 Amendment.( _ N Unit 2) 68 (Unit 1) Q
4 m < mg m h
' '
, . % REACTIVITY CONTROL SYSTEM I ( LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable N incore detectors and F0(Z) and F are verified to be LH within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERHAL POWER within the next hour and within the following 4 hourt the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERWsl POWE With more than one full-length rod trippable but inoperable due to causes other than addressed by ACTION a above, POWER OPERATION may continue provided that: Within 1 hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within 112 steps of the inoperable rods while maintaining the rod sequence and inser-tion limits of Specification 3.1. The THERMAL POWER level l shall be restricted pursuant to Specification 3.1.3.6 during (
<
f subsequent operation, and The inoperable rods are restored to OPERABLE status within 72 hour SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hour .1.3. Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 day (
-
l CATAWBA - UNITS 1 & 2 3/4 1-15 Amendment No. 74 (Unit 1) i Amendment No. 68(Unit 2) _ - - _ _ _ _ - _ _ _ _ _ _____ _ _ _ _ - _ - _ _ _ _ _ _ _ - _ - _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ s
,
- -
O O s TABLE 3.1-1 . i ACCIDENT ANALYSES REQUIRING REEVALUATION , IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Hisalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at full Power Major Reactor Coolant System Pipe Ruptu'res (Loss of Coolant Accident) hajor Secondary Coolant System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
%)
L fl O d : ,
].(<ek, ,
Gdyy
'
)'i CATAWBA - UNITS 1 & 2 3/4 1-16
;
-
.
O
'
O
'
3/4.1 REACTIVITY CONTROL SYSTEMS ( 3/4.1.1 BORATION CONfROL SHUTDOWN MARGIN - T,yg >200*' LIMITING CONDITION FOR OPERATION ,, I3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% ak/k for four loop operatio APPLICABILITY: MODES 1, 2*, 3, and ACTION: With the SHUTDOWN MARGIN less than 1.3% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN HARGIN is restore SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shal' be determined to be greater than or equal to 1.3% ak/k: Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperabl (* If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased-allowance for the withdrawn worth of the immovable or untrippable control rod (s);
- When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Soecification 3.1.3.6; When in MODE 2 with K,ff less than 1, within 4 hours prior to achieving reactor criticality by verifying that the predicted '
critical control rod position is within the limits of Specification
*
3.1.3.6; Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le. below, with the control banks at the maximum inservice limit of Specification 3.1.3.6; and kk0 Q% e j "See Special Test Exceptions Specification 3.1 Ddj[
,u ~
CATAWBA - UNITS 1 & 2 3/4 1-1 _ _a
r O O
'
REACTIVITY CONTROL SYSTEMS
'I S,U,RVE!LLANCE Rfcu!REMENTS (Continued) ) When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
1) Reactor Coolant System boron concentration, 2) Control rod position, 3) Reactor Coolant System average temperature, 4) Fuel burnup based on gross thermal energy generation, 5) Xenon concentration, and 6) Samarium concentratio .1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 1% ak/k at least once per 31 Ef fective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le., abov The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions * prior to exceeding a fuel burnup of 60 EFPD af ter each fuel loadin nn t . ' ' O O %, ll' ?,,a;
!. , , , ,
* NQ
;elijf Qr#{M}';f
~
.
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CATAWBA - UNITS 1&2 3/4 1-2 Amendment No. 39 (Unit 1) Amendment No. 31 (Unit 2) J
r O O 4EACTIVITY CONTROL SYSTEMS k( CONTROL BANK INSERTION LIMITS LIMITING CONDITION FOR OPERATION
~{ 3.1. 3. 6 The control banks shall be limited in physical insertion as shown in Figure 3.1- APPLICABILITY: MODES la and 2*#.
ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2: Restore the control banks to within the limits within 2 hours, or Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figure, or Be in at least HOT STANDBY within 6 hour SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hour ?
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Pccilityt Catawba 1f'-2 Exau-Dato 1992/10/16
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Ca /a u b a-Knowledge and Ability Hocord Form i COUllT HATRIX M O Alas /e s
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Summarizing Counts by K/A Group l for * PWR ~ Reactor Operator
*otal Plant Wido Genorics .................................. 13 K1 K2 KJ K4 K5 K6 Al A2 A3 A4 SG Plant Systems 1 2 1 1 2 1 1 1 3 1 5 5 23 Plant Systems II 3 1 2 3 1 1 -0 1 1 1 5 19 Plant Systems III 1 0 0 3 0 0 0 1 1 0 1 7 Emergency /Abn I 1_ 1 3 ........ 2 4 ..... 5 16; '
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Emergency /Abn II 1 0 3 ........ 4 3 ..... G 17 Emergency /Abn III O O O ........ 0 2 ..... 1 3 Totals 8 3 9 8 2 2 7 14 3 6 23 ===== Model Total .................................. 98 . t b
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K wiedge and Ability Record rm PLANT-WIDE GENERIC RESPONSIBILITIES FWR - Reactor Operator Target: 13 % Actual: 13.3 % K/A Rep Topic Rating __R/S__ 194001A106 2 Ability to maintain accurate, clear and concise logs, 3.4/ records, status boards and reports 194001A107 Ability to obtain and interpret station electrical and 2.5/ mechanical drawings 194001A109 2 Ability to coordinate personnel activities inside the 2.7/ contr^1 room 194001A111 Abilff t r- direct oorsonnel activities inside the 2.8/ contico , ,o m 194001K101 2 Knowledge of how to condtet and verify valve lineups 3.6/3.7 194001K103 Knowledge of 10 CFR 20 and related facility radiation 2.8/ control requirements 194001K104 2 Knowledge of facility ALARA program 3.3/3.5 194001K109 Knowledge of safety procedures related to high 3.4/ pressure 194001K114 Knowledge of safety procedures related to confined 3.3/ spaces e e e s -w ;m 7)
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Kr wiedge and Ability Record orm PLANT SYSTEMS PWR - Reactor Operator - 51 % Group I Plant Systep4 Target: 23 % Actual: 23.5 % C01 Cont,'l Rod Drive 017 In-Core Temperature 061 Aux./Emer. Feedwater 003 Reactor Coolant Pump 022 Containment Cooling 068 Liquid Radwaste 004 Chemical & Volume 025 Ice condenser 071 Wast Cas Disposal 013 E. Saftey Actuation 056 Condensate System 072 Area Radiation Mo Huclear Instrumen Main Feedwater System K/A Rep Topic Rating __R/S__ 001000G010 Ability to explain and apply all system limits and 3.3/ precautions 001000K103 CRDM 3.4/ K407 Rod stops 3.7/ K558 Reason for overlap of control banks 2.7/ A401 Seal injection 3.3/ G006 Knowledge of bases in technical specifications for 2.7/ limiting conditions for operations and safety limits 003000K201 RCPS 3.1/ A401 Boron and control rod reactivity effects 3.8/ A205 CIAS, SIAS 4.1/ A105 Main steam pressure 3.4/ A403 ESPAS initiation 4.5/ A303 Verification of proper functioning / operability 3.9/ K406 Reactor trip bypasses 3.9/ K604 Bistables and logic circuits 3.1/ K101 Plant computer (COLSS) 3.2/ A404 Valves in the CCS 3.1/ K301 Containment 3.8/ A411 Recovery from automatic feedwater isolation 3.1/ G010 Ability to explain and apply all system limits and 2.9/ precautions 061000A202 Loss of air to steam supply valve 3.2/ G002 Knowledge of system statis criteria which require the 3.0/ notification of plant personnel 072000A202 Detector failure 2.8/ G007 Knowledge of purpose and function of major system 2.6/ components and controls If : b
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e Kn wiedge and Ability Record rm PLANT SYSTEMS PWR - Reactor Operator - 51 % Group II Plant Systems Target: 20 % Actual: 19.4 % 002 RCS 011 PZRLCS 016 NNIS 033 SFPCS 055 CARS 064 ED/G 079 SAS 006 ECCS 012 RPS 026 CSS 035 S/GS 062 AC 073 PRM 006 WDS 010 PZRPRS 014 RPIS 029 CPS 039 MRSS 063 DC 075 CIRC K/A Rep Topic Rating __R/S__ 002000K109 PZR 4.1/ K603 Reactor vessel level indication 3.1/ K409 Safety injection val'9 interlocks 3.8/ )onn0A4n3 PORV and block valves 4.0/ PZR heaters 3.0/ t Failure of PZR level instrument-low 3.4/ C' DNB 3.3/ .ar 2 Knowledge of the annunciator alarms and indications, 2.9/ and use of the response instructions RCS 3.4/ AFW system 3.6/ ( , Ability to locate and operate components, including 3.6/ local controls 029000K302 Containment entry 2.9/ K303 Spent fuel temperature 3.0/ A302 Isolation of the MRSS 3.1/ G009 Ability to locate and operate components, including 3.2/ local controls 063000G010 Ability to explain and apply all system limits and 3.1/ precautions 064000K411 Automatic load sequencer: safeguards 3.5/ K406 CO) 3.0/ if} * n n , v, ~9
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Kno'wledge and Ability Record rm PLANT SYSTEMS PWR - Reactor Operator - 51 % Group III Plant Systems Target: 8 % Actual: 7.1 % 605 RHRS 008 CCWSS 028 HRPS 041 SDS 076 SWS 103 Containment 007 PRTS 027 CIRS 034 FHES 045 MT/G 078 IAS K/A Rep Topic Rating _ _R/S_ 005000K407 System protection logics, including high-pressure 3.2/ interlock, reset controls, and valve interlocks 007000K401 Quench tank cooling 2.6/ K102 Loads cooled by CCWS 3.3/ A304 Automatic actions associated with the CCWS that occur 3.6/ as a result of a safety injection signal 034000A201 Dropped fuel element 3.6/ K417 Reactor trip 3.7/ G010 Ability to explain and apply all system limits and 2.7/ precautions
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s-O KnowledgeandAbilityRecord9orm EMERGENCY PLANT EVOLUTIONS PWR - Reactor Operator - 36 % Group I Emergency and Abnormal Plant Evolutions Target: 16 % Actual: 16.3 % 000005 Inoperable / Stuck Rod 000040 Steam Line Rupture 000067 Plant Fire Onsite 000015 RCP Motor Malfunction 000051 Loss of Vacuum 000068 CR Evacuation 000024 Emergency Boration 000055 Blackout 000069 Loss Containment 000026 Loss of CCW 000057 Loss of AC Elec 000074 Inadeq. Core Cool 000027 PZR PCS Malfunction Instrument Bus 000076 High RCS Activity K/A Rep Topic Rating __R/S__ 000005G005 Knowledge of the annunciator alarms and indications, 3.1/ and use of the response instructions 000015G010 Ability to perform without reference to procedures 3.4/ those actions that require immediate operation of system components or controls 000015K207 RCP seals 2.9/ A105 The CCWS surge tank, including level control and level 3.1/ alarms, and radiation alarm 000027A218 Operable control channel 3.4/ G012 Ability to utilize symptom based procedures 3.8/ A202 Conditions requiring reactor and/or turbine trip 3.9/ A203 Actions necessary to restore power 3.9/ K302 Actions contained in EOP for loss of offsite and 4.3/ onsite power 000057A101 Manual inverter swapping 3.7/ A203 Fire alarm 3.3/ G012 Ability to utilize symptom based procedures 3.4/ K103 Processes for removing decay heat from the core 4.5/ K304 Tripping RCPs 3.9/ K311 Guidance contained in EOP for Inadequate Core Cooling 4.0/ G012 Ability to utilize symptom based procedures 2 9/ l 5) LL _ gk
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r Kn wiedge and Ability Record orm EMERGENCY PLANT EVOLUTIONS PWR - Reactor Operator - 36 % Group II Emergency and Abnormal Plant Evolutions Target: 17 % Actual: 17.4 % 000001 CRW 000022 Loss of RCS Makeup OG0038 SG Tube Rupture 000003 Dropped Control Rod 000025 Loss of RHR 000054 Loss of MFW 000007 Reactor Trip 000029 ATWS 000058 Loss of DC 000008 Stuck Relief Valve 000032 Loss of SRNI 000059 Loss Release 000009 Small Break LOCA 000033 Loss of IRNI 000060 GRW Release 000011 Large Break LOCA 000037 SG Tube Leak 000061 ARMS Alarm K/A Rep Topic Rating __R/S__ 000001G008 Ability to recognize indications for system operating 3.2/ parameters which are entry-level conditions for technical specifications 000001G010 Ability to perform without reference to procedures 3.9/ those actions that require immediate operation of system components or controls 000003A201 Rod position indication to actual rod position 3.7/ G010 Ability to perform without reference to procedures 4.2/ those actions that require immediate operation of system components or controls 000008K101 Thermodynamics and flow characteristics of open or 3.2/ leaking valves 000009A237 Existence of adequate natural circulation 4.2/ A103 Securing of RCPs 4.0/ A305 Manual and/or automatic transfer of suction of 4.3/ charging pumps to borated source 000025G011 Ability to recognize abnormal indications for system 3.6/ operating parameters which are entry-level conditions for emergency and abnormal operating procedures 000029G011 Ability to recognize abnormal indications for system 4.4/ operating parameters which are entry-level conditions for emergency and abnormal operating procedures 000029K301 Verifying a reactor trip; methods 4.2/ K312 Actions contained in EOP for ATWS 4.4/ A202 Indications of unreliable intermediate .ange channel 3.3/ operation 000038K302 Prevention of secondary PORV cycling 4.4/ G012 Ability to utilize symptom based procedures 3.2/ A101 Crosstie of the affected dc bus with the alternate 3.4/ supply 000059A102 ARM system 3.3/ r- o-Lwi n b Y 1992/09/25 16: 30:42 page 7
n K wiedge and_ Ability Record . rm EMERGENCY--PLANT EVOLUTIONS i PWR.- Reactor Operator'- 36 % Group III--Emergency and' Abnormal Plc.nt Evolutions Target: 3 % Actual: 3.1 % 000028-Pressure Level Malfunction 000056 Loss of OffSite Power 000036 Fuel Handling Accident 000065 Loss of Inscrument Air K/A_ Re Topic Rating __R/S__ 000056A214 Operational status of ED/Gs (A nd B) 4.4/ A201 Cause and effect of low-pressu- instrument air alarm 2.9/3.2-000065G010 Ability to perform without ref ance to procedures 3.2/ those actions that require im-sdiate operation of system components or controls i
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r; O O U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME: FACILITY: Catawba 1 & 2 9 REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 92/10/16 INSTRUCTIONS TO CANDIDATE:
'Uso the answer sheets provided to document your answers.- Staple this cover --
chcot on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE % _ 100.00 % TOTALS FINAL GRADE All work'done on this examination is my ow I have neither given nor received ai Candidate's Signature
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O O REACTOR OPERATOR Page 2 A N-S W E S H E E-T Multiple Choice -(Circle or X your choice) If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d
- -003 a b c d 026 a b c -d 004 a b c d 027 a b c d'
005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d' 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011- a b c d ___ 034 a b c d 012 a b c d 035 a b c -d 013 a- b c d 036 a b .c d 014 a b c d 037 -a b c d 015 a b c d 038 a b d 016 a b c d 039 MATCHING 017 a b c d a- EO $ $ () k($['$1 Ois L Jl 018 a b c d b 019- a b c d c 020 a b c d MULTIPLE CHOICE 021 a b d 040 a b c d-022- a b c d
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ANSWER SHEET-
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Multiple Choice .(Circle or X your choicc)
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If you change your answer, write your selection in the blan MATCHING 060 a b c d-a 061 a b c d b 062 a b- c- d c 063 a b c d MULTIPLE CHOICE 064 s b c d
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042 a b c d 065 a b c d 043 a b c d 066 a b c d 044 a b c d 067 a b c d 045 a b c d t 48 a -b c d ,. 046 a b c d 06; a b c d __
'047, a b c d 070 a- b c d 048 a b c d 071 a b c d 049 a b c d 072 a b c d v
050 a b c d 073 a b c d 051- a b c d _ _ _ 074- a b c d-052 a b c d -075 a b c d 053 a b c d 076- a b c a b c d 077 a b' c d 055 a b c d 078 a b c d y'; n gga gp &p 056 a b c d 079 a b c d b U uN"N a; Ik$
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r i O O REACTOR OPERATOR' Page- 4 A-N S W E R- S 11 E E T .; Multiple choice- (circle or X your choice) If you-change your answer, write your selection in the blan _ 083- a b c d 084 a b c d 085 a b c d 086 a b c d 087 a b c d 088 a b c d __ i 089 a b- c d OSC a b c d _ _ _ 091 a b c d _ 092 a b c d 093 a b c 'd 094 a b c d 095 a b c d 096 a b c d , '097 -a- b c d 098- a b c d-099 a b c d
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'A f h . b!!k COPY (********** END OF EXAMINATION **********) , - -- ..
O O Page S NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your applicatio'- and could result in more severe penaltie . After the examination has been completed, you must sign the ctatement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination-room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consc ~utively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it out, 9. The point value for each question is indicated in parentheses after the question.
10. Show all calculations, methods, or assurotions used to obtain an answer to any short answer questions.
11. Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE JEY ANSWER BLANK.
12. Proportional grading will be applied. Any additional wrong information that.is provided may count against yo For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you give.five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect,10.20 will be-deducted and your total credit for that question will be 0.80 instead of-1.00 even though you got the four correct answers.
13. If the intent of a question is unclear, ask questions of the examiner onl pg g pg -
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=14. When turning in your examination, assemble the completed examination with-examination questions, examination aids and answer sheet In addition, ;
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turn in all scrap pape ' 15. Ensure all information you wish to have evaluated as part of your answer is . on your answer shoot. Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have tw .sd in your examination, leave the examinatica area (EXAMINER WILL DEFII,J hE AREA) . If you are found in this area while the examination is still in progreer, your license may be denied or revoked.
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O O RREACPOR OPERATOR Page 7 QUESTION: 001 (1.00) If a valve is found out of position during independent verification of its position, which ONE (1) of tPm following actions must be taken or-performed?. a. Immediately notify the operator at the controls (OATC) and reposition the valve with the OATC's permission, b. Immediately notify the shift supervisor (SS) and reposition the valve with the SS's permisalo c. Immediately reposition the valve, then notify the OAT d. Immediately reposition the valve, then notify SS.
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-QUESTION: 002 (1.00)
Which ONE (1) of the'following statements describes the PURPOSE (as defined in OMP 2-18 "Tagout Removal and Restoration (R&R) Procedure") of-the Venting Restoration Sheet? a. Explain how to vent equipment prior to maintenance, b. Describe how to vent infrequently operated equipment prior to operation of the equipment, c. Record equipment placed in an "OUT OF NORMAL" condition and to-insure the equipment is_ returned to its normal conditio d. Document independent verification of vent valve l
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O O REACTOR OPERATOR 'Page 9 QUESTION: 003 (1.00) Which ONE [1] of the following radiation exposures describes a Duke Power Company administrative radiation exposure limit? a. 3000 mRems/ quarter gamma exposure to the whole bod b. 6000 mRems/ quarter gamma exposure to the ski c. 52 MPC-hours internal radiation exposure in any seven day d. 1500 MPC-hours internal radiation exposure per quarte ; I l r I i filASTER I
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O O REACTOR' OPERATOR Page 10-QUESTION: 004 (1.00)
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Which ONE (1) of the:-following statements' describes'how operators--are identified by ;the use of color coded identification tape while in. the: Lower Containment during outages? a. Yellow tape worn on the upper portion of the ar b. Red tape worn on the shoulde , c. Magenta and yellow tape worn on the ches d. Green tape worn on the upper part of the bod i a[ Q, Ti*I% .
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~ QUESTION: 005 ( 1. 0 0 ) _ If;all factors are equal, which ONE [1] of the following volunteers-should be selected ~for a Planned-Emergency Exposure? a. 20 year old man, year old wonan, c. 40 year old man, year old woma .
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. QUESTION: 006 (1.00)
The following plant conditions exist:
- The plant is in MODE Personnel must work in the main steam doghous Which ONE [1] of the following positions must be notified of personnel working in the main steam doghouse?
a. Operator at the controls and Shift Supervisor, b. Operator at the controls and Balance of Plant Operato c. Individual's work supervisor and Protected Area First Aid Staf d. Individuals's work supervisor and Security Medical Emergency Response Team [MERT).. _
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O O REACTOR OPERATOR Page:13 QUESTION: 007 (1.00) Which ONE (1) of the following must grant approval prior to entry into areas posted as confined Spaces? a. Radiation Protectio b. Control Roo c. Safety Sectio d. Securit , f
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_ _ O O-REACTOR'OPERATok Page 14
:QUESTIOll: 008 - (1.00)
Which ONE [1] of the following conditions or items-should have_an entry in tho Open Item' Summary of the Control Room Logbook?- a. Any Control Room equipment placed in an "Out of Normal" position and not R&R' b. Any OPERABLE automatic valve in an abnormal position and R&R' Initiation of Shutdown Requests, d. Temporary Modifications (Brown Tags].
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QUESTION: 009 (1.00) Which_ONE (1) of_the following'p.sitions is responsible for maintaining-the Control Room Annunciator Status Log? a. Balance of Plant Operato b. Operator at the Centrols, c. Shift Supervisor, d. Unit Superviso Qni p w n .p, :', ~na ;x
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t iQUESTIONs C10 (1.00) Which ONE (1) of the following statements describes how revisions are designated on the C_ntrol Room copy of. electrical elementary diagrams
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a. The diagram will be red-marked to show the details of-the modification and stamped " OPS INTERIM". b. The diagram will be red-marked to show the details of the modification and stamped "AS-BUILT".
c. The revised area will be outlined in red ink and-the diagram will Pa stamped "SEE INTERIM AS-BUILT".
d. The revised area will be blacked out and the diagram stamped
"SEE OPS INTERIM IN THE DRAWING SUPPLEMENT FILE."
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u_ _ . _.... _ . . . ._ . O O REACTOR OPk.4 TOR Page 17 ? QUESTION: 011 (1.00) If the Charlotte Dispatcher notifies Catawba Nuclear Station-that the Unit Interface controller [UIC) operating limits require. resetting, which ONE (1).of the following positions is responsible-for resetting the UIC operating limits? a. Nuclear Control Operator or Shift Support Technicia b. Operator at the Controls or Shift Technical Advisor, c. Shift operations Manager or Unit operations Manager, d. Maintenance Engineering Services or Protective Relay Engineering Grou }, 's 1-
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O G V V REACTOR: OPERATOR Page 18' QUESTION: 012 (1.00) 4-Which ONE (1] of the following individual [s] is authorized to adjust the sound level of annunciators in the control Room? a. Operator at the Control L
- Shift Support Technicia c. Nuclear Control Operato I I&E personne ,.
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Jy: .i O O-REACTOR OPERATOR Page 19-QUESTION: 013 (1.00) A non-licensed operator (NLO) who is NOT in an approved license training class is doing Control Room observation. Which ONE [1] of.the following Control Room equipment or operations may the NLO operate or. perform under the direct observation of en actively licensed RO or SRO? a. Boration of Reactor Coolant Syste b. Startup of Recidual Heat Removal System, c. Reactor Coolant Pum d. Turbine Generato , s, 4
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QUESTION: 014 (1.00) Which ONE (1) of the following descriptions ~ explains why 125 Vdc and
- 70-Vdc power supplies are used in the DC lloid Cabinet?
a. The 70 Vdc is used to latch the grippers and the 125 Vdc is used to hold tho' grippers, b. The grippers are latched by adding the 125 Vdc and 70 Vdc then the 125 Vdc is used to hold the gripper c. 12f "-ic is used to latch the grippers and 70 Vdc is used to hold' the < ipper d. 125 Vdc is used to latch the grippers and as an alternate hold voltage; and 70 Vdc is used to hold the grippers and as an alternate latch voltage.
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REACTOR _ OPERATOR- _ Page 21-QUESTION: 015 (1.00) Which ONE-[1] of the following statements describes a-feature of the-
" Intermediate Range (IR] High Flux Rod Stop"?'
a. Blocked by IR H1 flux trip block pushbutto b. Affects only automatic withdrawa Setpoint is the current equivalent to 20% power on 2/2 channel Prevents automatic rod withdrawal when less than 15% powe nlT ;" b c c reo r,.w
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, . . . . . O O REACTOR OPERATOR Page 23 QUESTIOll: 017 (1.00) Which OllE (1) of the following reasons explains Wily the Urgent Failure Alarm Reset pushbutt'-n should not be depressed for greater than ono {1) second? a. CRD'4 coils could overheat and f ai Supervisory Memory Duffer Circuit could satutate and cause bank overlap to occur prematurel c. Multiplexiag thyristors and sampling resistors may be damaged from excessive cyclin Urgent Failure Alarm may llOT actuate due to overheating of pushbutton contact ,
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QUESTION 018 (1.00) Which ONE [1] of the followirig statomonts explains WlIY cach reactor ! coolant pump (NCP) has a safety breaker installed in oorica with its ! 6900 V aupply breakor? a. Provent damage to the high voltage cupply, b. To provent a mechanical iallure of the electrical penetration i ascombly due to fault current if one procectivo device fails, c. To provent damage to the high voltago (6900 V) oupply brot':e: in the event of an HCP t>hearoc: anaft, d. To provent damage to the high voltage (6900 V) supply bret.kot in the event of an HCP locked roto p p n r% ',' T* ; i.:. < ;.
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O O REACTOR OPERATOR page 25
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QUESTION: 019 (1.00) Which ONE (1) of the following statemento explains how an operator , determinos that the water supply to a Reactor Coolant Pump [HCP) #2 seal ; is adequato? ;
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a. Observing #1 ocal leakoff of 3 gp j b. Oboorving NCP coal injection flow of 8 gp c. Observing NCDT preocure of 15 psi , d. Observing VCT proosure of 20 psi L t-L
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y O O HEMJI'on OPERATOR Pago 26 QUESTIoll: 020 (1.00) If lic (Roactor Coolant) Systein cold leg teinporaturo is 265 degrees T, which OllE [1] of the following statomonts explains why secondary water temperature must be loss than 315 degroos F prior to starting the 11C Pumps? a. Protect against 11C System overpressure caused by energy additions from the secondar b. Protect against a loss of shutdown margin caused by energy loss to the secondar c. Provent excessivo lic System cooldown rato d. Provent loss of pressurizer level duo to excessivo 11C System cooldow __ _f . 'I Q) g
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QUESTION: 021 (1.00) The following plant conditions exist: ; P
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The plant is operating at 100% power at BO All systems are operabl Control Bank "D" starts stopping in slowly, but at a noticeablo [ rat , Which ONE (1) of the following ovents will cause this responso? f
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a. A leak has developed in the Rogonorativo lleat Exchange b. A le ik has developed in the Lotdown licat Exchange g
c. A leak has developed in the tubo bundio of the seal Water float , Exchange d. A leak in the Excess Letdown llent Exchange . - r ( dhk COPY ! r y, ,.,-.v -w ,,..-p ,,-u, ., .r..myn,,.,,s, , - . . , , . , ..,9.u,. .,,e..e . . , ,,,, -- y, ,,. ,.p. mp %% , - 14c .- w e-,,+yy.-r +.r-73--. gym.,--yn,,,.g--4, r-T **FT PP Mr *9NN P f t'
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O O l REACTOR OPERATOR Page 28 l QUESTIOll: 022 (1.00) The following plant conditions exist: l
- Boric Acid Transfer Pump 1A running following an auto start i domand from the Ronctor Makeup system - A spurious So signal was roccived Which OllE [1] of the following operations must be performed to shutdown ;
Boric Acid Transfer Pump 1A following roset of the Ss (Safoty Injection) ; sequencor? a. Select "0FF" on the pump control switc b. Start the other Boric Acid Transfer Pum c. Vorify that the VCT level is above the auto makeup sotpoint and select "0FF" on the pump control swltc Roset the Boric Acid Transfer Pump and select "0FF" on the pump control switc pll p n V i" Li; 4 Li
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REACTOR- OPERATOR Page 29 ; QUESTION: 023 (1.00) , If a steam leak on Unit 2 is causing steamline pressure to decrease at a i constant rato, which ONE (1) of the following statomonts explains the plant's response to the steam leak? ,
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a. A steamlino low pressure safety injection [SI) will occur at an indicated pressure loss than the sotpoin ! b. A steamline low pressure SI will occur at an indicated pressure greater than the sotpoin , c.-Since the steamlino low pressure SI circuit is rato snnsitivo, the rate of decrease is multiplied by 10-and an SI will-occur when the rate of decrease reachos 100 psi /sec.- d. Since the steamline low pressuro SI circuit is rate sensitive, the constant rate of change is moltiplied by 10 and an SI will occur at an indicated pressure less than the sotpoint or the rato of decrease is greater than 100 psi / soc.
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QUESTioll: 024 -(1.00) _ l If an operator initiates a manual food water isolation by pushing the Train A RESET pushbutton in error (rather than- the Train A IllITIATE , pushbutton) and then pushes the Train B--IllITIATE pushbutton, which ol1E (1) of the following statements describes the plant response? a. Both Trains will isolat b. lleither Train will isolat c. Only Train A will isolat d. Only Train B will isolat i
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___ O O REACTOR OPERATOR Page 31 QUESTIO!1: 025 (1.00) The following plant conditions exist:
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Unit 2 at 43% powe The Digital Rod Dicplay System "llOli-URGEllT" Annunciator is in alar other alarms have been received.
> Which OllE (1) of the following conditions caused the alarm? Data B failur b. All three control units faile Data A rod position is 180 steps and Data D rod position is 102 steps, A rod in stuc , 7, .- - ; . ,, 1 La h
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REACTOR OPERATOR Page-32 r i QUESTIOll: 026 (1.00) The following readings wore noted on the Power Range and Intermediate Rango Channelus ., 11-3 5 0 5 x 10E-5 11-36 0 8 x 10E-6 14 - 4 1 0 8.5% i 11-4209% 11 - 4 3 0 8.5% ; li-44 0 9% , i Which ONE (1) of the following describes the problem indicated by these readings? (Sco figure below) , a. 11-35 reading high for current condition , b. 14-36 reading low Ior current condition c.11-41 and 11-43 adjusted improperly during last calorimetri d. 11-4 2 and 11-4 4 adjusted improporly during last calorimetri ! DFtfeMffAAff Pows A suWGt RJWot
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" ' QUESTIoll: 027 (1.00)
The reactor is operating at 13% power when Intermediate Rango Nuclear Instrument H-35 fails LOW. Which OllE (1) of the following responson-will occu ! e Reactor Trips on Source Range High Flux.Lovel Tri Intermediato Rango (IR) High Flux Rod Stop [C-1) comes i t c. Power Above Permission P-10 status light for N-35 goes ou * d. Power Abovo Permissivo P-6 status light for N-35 goes ou > c V s
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o O , REACTOR OPERATOR Pago 34 i i
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QUESTION: 028 (1.00)
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If'a Containment entry is mado during reactor startup with the Intermediate Range reading loss than 10 E "0 . amps, which ONE (1) of the
:following ransons explains why the Containr.ent Evacuation klarm ca:Inot .i be used to warn personnel to ovocuato containment?
a. liigh Flux at Shutdown Alarm is blocke b..lligh Flux at Shutdown Alarm only sounds the annunciator in the control Roo ;
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c. Source Range liigh Level Reactor Trip is blocke d. Alarm actuation may initiato magnetic interference that could cause a spurious reactor tri :
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QUESTIoti: 029 (1.00) Which ONE [1] of the following statements explains how control Room operators can recognize that an incore thermocouple is being monitored : at the Incore Thermocouplo Panel? , a. The plant computer thermocouple temperature will be indicated by a yollow reading, b. The plant computer will show an alarm becauno the signal to the computer will be disconnecte , 0. The local control annunciator will alar . d. The thermocouple temperature reading will flash on the plant "i computer scree r t
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QU'ESTIOll: 030 (1.00) ,
'if the Lcwor Containment Ventilation units are shif ted to !!IGil spood, l'
which OllE (1) of the following statomonts describes how the containment Vontilation (VV) System responds? . a. Both the thermostat valvo and the cooling water bypass fail close ' b. The thermostat v61ve fails closed and the cooling water bypass' is throttled to control temperatur c. Bo'hc the cooling water bypass valvo and thermostatically : controlled valve fail ope , d. The cooling water bypacs fails open to allow tho thermostat , valve to control on temperatur i
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QUESTIolis 031 (1.00) , According to Technical Specification Unses, which OllE (1) of the following statomonts describes the offect of operation with the Ice condonsor Doors opon? ; a. In the event of a LOCA, the containment peak pressure transient t may exceed 14.7 psi ; b. In the event of a LOCA, the containment peak pressure transient f will be loss than 14.7 psig, c. During a LOCA the roloased Reactor Coolant System Fluid may=be .
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diverted away from the ice condensor bay d. Inadequato sublimction of the ice bed may not occur.bocause of I warm air intrusio i o
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QUESTIOll: 037 (1.00) The following plant conditions exists
! - Unit 2 at 54% power ' - P-14 (111 111 S/G Loyol) came in momentarily on the Unit 2 "C" S/G I
and then cleare i Which OllE (1) of the following statements describe how the Feodwater !
' (CF) System on Unit 2 responds to these conditionsY a. Feedwater isolation occurs because a turbine trip and reactor trip will be generate i b. The CF Bypass to CA flozzle valves start to close and then ,
reope c. The CF Pump Dischargo valves go closed and remain close ;
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d. The Containment Inolation Bypass valvos go open and then clos i P
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.t REACTOR OPERATOR Page 39 ;
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i QUESTIO!1 033 (1.00) Which OllE (1) of the following statomonts explain why the Feodwater (CF) i Pump dischargo valvo is opened last and shut first when starting and stopping a CF pump? I
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a. Provent pump oversnee l b. Provent nuction line overpressur c. Provent pump runou d. Provent pump cavitatio !
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'QUESTIO!12 034 (1.00)
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Which OllE (1) of the following statements concerning the Auxiliary Foodwater System describes the operation of the Turbine Driven Pump (TDCA)? a. If instrument air pressure is lost, the steam isolation valves fail open to start the TDC b. If either of the Motor Driven Pump's (MDCA) flow exceeds 780 gpm after a valid ctart signal, the TDCA will auto star c. The TDCA pump will auto start upon loss of both Hain Feedwater pumps, d. The TDCA will automatically start upon an undervoltage signal on one (1) 11CP bus, i
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Which ONE f 1) of the following staterbento describes the control placed on CA-67 [Sco Drawing CN-CF-CA-2 belov) a. CA-6 muut remain open to allow overflow to the UST (Upper Surgo Tank).
b. CA-6 must remain open to provent overflow of the Auxiliary Feedwater Condensato Storage tan c. CA-6 cannot be closed unless the UST is less than 90%. d. CA-6 cannot be closed without management approva _ ta 6<r[u$$)
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> 0 0 REACTOR OPERATOR Page'42 QUESTIOll: 036 (1.00) Which OllE [1] of the following reasons explains why the AC input breaker must be closed first when placing a battery charger in service? a. To provent excooding the auctionooring Diodo Assembly peak
*toltage limit when closing the output breake :
l b. To ensure charging rectifiers are r,at exposed to an overvoltage condition when closing circuit breakers, i-l c. To protect the output breakers-from heavy surgo current l d. To limit charging current to minimizo battery explosion potentia ; i
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_ O O REACTOR OPERATOR Page 43 QUESTION: 037 (1.00) Refer to the RP 30A Module shown below:
- The TRIP 2 light is off - The TRIP 1 light is off - The OPERATE light is off Which ONE [1] of the following reasons explains why the OPERATE light is off?
a. The INT position is selected so all pulses less than the preset threshold are ignore b. The signal has been lost due to detector failur c. The THIP 1 signal has c1 cared but has liot been rese d. The TRIP 2 signal has cJeared but has not been rese ~
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_ _ _ - - - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ O O REACTOP OPERATOR Page 44 I QUESTIOll: 038 (1.00) Which OllE (1) of the following statements describe the use of installed sources in Area Radiation Monitors (EMF)? a lic filters area EMF uses an installed cource to verify alarms, b. Steam lino EMF use an installed source to prevent spurious alarm c. Reactor Building Refueling Bridge EMP uses an installed source to prevent spurioua alarm d. Check sources have been removed from Area Radiation Monitors for ALARA concern ~ I l'" p ,o i ,, V
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O O REACTOR OPERATOR Page 45 QUESTIO!!: 039 (1.50) For the drawing below, match the Reactor Coolant (14C] loop or system from Column B with the connections in Column (110TE: Items in Column B can only be used once, and only a singic answer may occupy one answer space.)
Column A Column B
[ CotillECTIoff ) (LOOP / SYSTEM)
____________ _________ ___
----- Pressurizer (PCR) Loop A
----- Loop B liot Leg Connection "A" Loop B
----- Loop C Cold Leg Connection "B" Loop C Loop L' D pump "A" Suction ormal Charging
'/ . liI Discharge (Injection Phase) Pressurizer Spray n
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QUESTION: 040 (1.00) ; I If all Reactor Coolant [NC) Pumps are off, which CNE (1J of the
.following Reactor Vessel Level Indicating System (RVLIS) indications ,
is/are valid? a. Train A upper range and Train B D/P Rang ,
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b. Train A upper range and Train B lower rang c. Train A D/P range and Train B upper range, d. Train A D/P range and Train A lower range.
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REACTOR OPERATOR Page 47 l i QUESTION: 041 (1.50) For the ECCS drawing below match the interlock condition from Column B that is required to open the valves in Column [ NOTE: Items in Column B can only be used once, and only a single answer may occupy one answer space.)
Column A Column B
[ VALVES) [ INTERLOCK CONnITION] --- ...-- -------------------- NI 136 and NI 184 NC pressure less than 385 psig FW 27 closed ND 28 and N1 185 FW 55 closed ND 28 closed NI 147 and FW 55 ND 1 or 2 closed ND 36 or 37 closed NI 336 closed NI 147 open NI 185 open NORMAL CHARGING LINE
A (-- NI19 NV D42 NV A m NV 282 COLD H (- -t(- .
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]N CON NS 43 263 SPRAY ND, ,28 NV B N1334
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o N1100 HOT LEGS " N1118 Ni a Ni C 4--- COLD ( >f NI162 ~ 135 LEOS NI147 NI 152 ' NI150 -
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O O REACTOR OPERATOR Page 48 QUESTIOll: 042 (1.00) Which OllE (1) of the following identifies the ,awer supplies for Pressurizer lleater Group A? LXD and CD LXil and CD c. LXil and CD d. LXI and CD %hhk i D an-a21 ~ n 3,bi VLa $g*af _ - _ _ -_ _ _ - - _ _- _______-______-_ _______ -
_ _ _ _ _ _ _ _ _ .__ . _ _ _ _ . __ . _ . _ _ O O ! REACTOR OPERATOR Page 49 j QUESTION: 043 (1.00) The following plant conditions exist: ,
- Pressurizer PORV isolation valve NC-318 is shut to isolate leaking PORV NC-32B - PORV isolation valve NC-33A is open and PORV NC-34A is shu PORV isolation valve NC-35B remained open when its control switch - l was placed in close to isolate a leaking PORV NC-36B Which ONE (1) of the following actions should be taken to HUT PORV isolation valve NC-35B?
P a. Manually close NC-35B circuit breaker on CD b. Manually close NC-35B circuit breaker on EMX Position NC-35B Control Switch to " OVERRIDE" positio Position NC-35B Control Switch at the ASP to "CLOSE".
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..; :I .OUESTION: 044 (1.00)
The following plant. conditions exist:
- Unit 1.is at-100% powe All control systems are in automatit Pressurizer Level Control Switch is in position 1- Preusurizer Level Channel II fails LO Without operator intervention, which ONE {1) of the following plant responsus will occur?
a. Letdown isolates, Pressurizer Heaters shut of b. No effcct on charging' flow, no Letdown Isolation, Pressurizer Peaters stuit of c. Charging flow to maximum, no Letdown Isolation, all Pressurizer-lleaters shut of . d. Charging flow to maximum, Letdown Isolates, Pressurizer Heaters shut of . Is IVilst5 Iilii CDPV
l O O REACTOR OPERATOR Page 51 QUESTION: 045 (1.00) The following plant conditions exist:
- Delta T is 100% - Over Temperature [OT) Delta T setpoint is 138% - Pressurizer pressure is decreasing due to a small Reactor Coolant System leak Which ONE [1] of the following responses describes the OT Delta T setpoint relationship to Actual Delta T?
a. Setpoint increases causing the di.fc .. ice between the setpoint and actual Delta . t- nerease Setpoint decreases causing the difference between the setpoint and actual Delta T to decreas .. c. Actual Delta T increases and the setpoie; decrease causing the difference between setpoint and actual Delta T to iscrease, Actual Delta T decreases and the setpoint increases causing the difference between setpoint and actual Delta T to decreas _ M
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- REACTOR OPERATOR Page-52 .. ' QUESTION: 046 (1.00)
If an Urgent Alarm has been eceived because DRPI has been unable to calculate the position of a particular control rod, how will the position indication of that rod be affected? a. It will be unaffectei as long as the rod is not move b. All position LEDs and the GW LED will be flashin c. All LEDs will be out except the GW [ flashing) and the ROD BOTTOM LE d. All LEDs will be out except the GW-LED [ flashing).
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O O REACTOR OPERATOR -Page 53 QUESTION: 047 (1.00) Which ONE (1) of the following statements explains why the outer Reactor Coolant Plow Detector is offset by 30 degrees along the pipe bend? (See the Reactor Coolant Flow Detector below) I a. To minimize detector output for reverse flow conditions, b. To maximize detector output for normal flesw conditions, c. To minimize detector damage by flow erosion, d. To maximize detector output during natural circulatio s 15 , 15 /
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O O REACTOR OPERATOR Page 54 QUESTION: 048 (1.00) During a power decrease on Unit 2, which ONE {1] of the following signals must be present to receive the White Swap enable light for the CA Nozzle? CF to CA Bypass Valve is ope CM-839 [ Reverse Purge Condenser 1A Isolation) closed.
, c. CF-100 (Tempering Flow Isolation] Ope d. The associated Steam Generator's CF Containment Isolation Valve is close _
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n -s REACTOR OPERATOR Page 55 QUESTION: 049 (1.00) Which ONE [1] of the following sets of conditions MUST be satisfied to cool the FWST using the 1A Containment Spray (NS) Pump and cooler? [NS Drawing provided below) a. Close NS53B and open NSIB and NI184 b. Close NS20A and open NA18A and NI185 c. Place CPCS [ Containment Pressure Control Systetr) in TEST and adjust the TEST SIGNAL to obtain the CPCS Permissive Signa d. Open the 1c ked closed recirc valve and close the NS1A pump breaker at the ETB switchges % helN NS4M
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O O REACTOR _ OPERATOR Page 56 QUESTION: 050 (1.00)
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Which ONE [1] of the following statements describes the response of the Spent Fuel Pool Cooling System [KF) following a-Blackout?
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Assume No operator action is taken) a. Spent Fuel Pool level and temperature will increas b. Spent Fuel Cooling Pumps restart automatically but the KF Skimmer Pump must be manually restarte c. Makeup to the Spent Fuel Pool from thre Fueling Water Storage Tank [FWST) will be isolate Spent Fuel Pool level will increase because the KF makeup valves
[KP-101B and KF-103A) automatically ope ,
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O O " REACTOR OPERATOR -Page 57 4 ; -QUESTION: 051- (1.00) LIf the Main Steam Isolation Valves [MSIVs) on Unit 2 are opened before pressure is completely equalized, which ONE [1] of the:following responses may occur? a. Reactor Trip due to High Fressurizer Level, b. Safety Injection due to Steam Generator swel c. Main Steam Line Isolation due to high steam line pressure rat d. Steam Generator PORVs lifting due to moisture in the Main Steam Lines flashing to stea , nR c%4v' Wl iilt
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REACTOR OPERATO .Page 58-
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QUESTION: 052- (1.00)
.Which ONE (1) of:the following statements describes how to remove Pneumatic Circuit Breaker PCB 17 from service? [See figure below);
a. Place the control Room _ control switch inl" TRIP" and pull-to~- lockout; closo disconnect ,
,
b. Place the Relay llouse control switch in " TRIP" and pull to lockouc; capon disconnect , c. Plsco-the. local control switch in " TRIP" and pull to lockout;- open disconnect ' d. Install grounding device; open disconnects; rack out.- PCD 17 : L PCD a i-230 KV-
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_ _ _ _ _ _ _ . _ REACTOR OPERATOR Page 59 QUESTION: 053 (1.00) The following plant conditions exist:
-
A Blackout has occurre A LOCA occurs 13 seconds after the Blackout.
s Which ONE (1) of the following statements describes the Emergency Djesel ) Generator Sequencer operation? Sequencer resets; sheds all loads; and then initiates the LOCA sequenc _ a b. Sheds all loads; starts the accelerated sequence followed by the committed sequence, c. Sheds non-LOCA loads and continues starting LOCA loads in the , LOCA sequence, e d. Sequencer resets; sheds non LOCA loads; starte the accelerated ' sequence and the committed sequenc _
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, - . . O O REACTOR OPERATOR _Page 60 QUESTION: 054 (1.00) Which'ONE [1] of the following statements explains how CO2 pressure is maintained in~the Diecel Generator Low Pressure CO2-storage-tank? a. The refrigeration package is manually started at 295 psig= tank pressure tc. maintain the CO2 in a liquid-state.-
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b. The refrigeration package is manually started at 295 psig tank pressure to maintain the 002 in a gaseous _ stat c. The refrigeration package is automatically started at 305 psig tank pressure to maintain the CO2 in a liquid stat d. The refrigeration package is automatically started at-305 psig tank pressure to maintain the CO2 in a gaseous stat { li " 4% n ; LOL . ,' u md4-
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QUESTION: 055 (1.00) Containaient Sump Isolation valves NI-3 84B and NI-185A autoulatically -OPEN following a Sa-[ Safety Injection) with 2/4 FWST [ Low Fueling _ Water-Storage Tank) level (37%). Which ONE [1] m the following conditions will require operator manual action to open the valves? -
(See ND' system; drawing below) - ECCS is rese ECCS has not been reset.
. c. C-Leg Recirc FWSr To CONT SUMP FNAP ENABLE light not li d. NS-43A (388), NS-18A(1B), NI-185A (184B] and ND-28A (NI-1369] close LOOP B HOT LEO ._
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.' REACTOR OPERATOR: PageL62 ' QUESTION: 056 - (1. 0 0) .
I Which.ONE [1] of the following methods is used to cool the PRT (Pressurizer Rclief _ Tank]?: -(Refer to PRT drawing-below) a. NCDT Recirc and Spray from the RMWS b. NCDT Recire and-ND Recir c. Spray from the RMWST and ND Recir d. Spray from the RMWST and venting to Waste. Gas Decay Tank.1 _ PZR : 'o
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v/ b REAC'rOR OPERATOR Page 63 QUESTION: 057 (1.00) Which ONE [1] of the following reasons explains why Component Cooling (KC] valves 56A and 81B [ND Heat Exchanger Inlets) open on Lo FWST level following Ss (Safety Injection)? (Refer To XC drawing below) Preve7t ND pump runout in the event of loss of one train, b. Prevent KC pump runout in the event of loss of one trai Prevent water hammet in the ND Syste d. Prevent water hammer in the KC syste KC SURGE TANK A KC SURGE TANK B
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QUESTION:'-058_ ( 1.' 0 0 ) .
:Which;ONE (1) of the-following: statements describes how Component Cooling-(KC] system valves align-in response to=an Sp'[ Phase-:B) signal- 'following an Ss.[ Safety' Injection) signal? [ Refer to KC drawing.below]'
a. KC valves 1A, 3A,'50A and 230A ope b. KC valves 2B, 53B open and 18B and 228B clos .; c. KC valves 2B, 53B close and 3A and 230A ope >
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Page'65
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QUESTION: 059 (1.00) If only one Hydrogen Recombiner is availz.ble.following a-LOCA, which'ONE
[1] of the following Containment Hydrogen concentrations will NOT be exceeded?
a. 31, by weigh b. 4% by volum !
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o o REACTOR OPERATOR Page 66 QUESTION: 060 (1.00) khich ONE (1) of the following statementr describes a purposo of the Containment Purge (VP] System? a. Reduce Hydrogen concentration in Containment following a LOC b. Decrease Containment humidity to within acceptable limits-for proper operation of refue)ing instrument c. Provide additional cooling to upper Containment during refuelin d. Reduce fission pioduct concentrations in Containment atmosphere to acceptable 3 :. mits for personnel acces ,
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, .QUESTTON: 061. -(1.00) -The following plant conditions exist: -
Refueling in progress on Unit A spent fuel element is being moved from the reactor-to the upende The spent fuel element in dropped to the bottom of the cana Which ONE [1] of the following products released from the spent fuel-clement will reesent the most immediate hazard?- a. Hydrogen ga b. Alpha radiction from fission product c. Gamma radiation from fission and corrosion products, d. Icdine and Krypton gase _.
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; QUESTION:-062 (l.00) -The following plant conditions exist: ; -
A reactor trip has occurred from 80% power on Unit _ Reactor Trip Breaker "B" failed-to ope Which ONE (1) of the following statements describes how the Steam Dump- s control System will respond? Plant Trip Load Re.jection Condenser Atmospheric Controller Controller Dumps Dumps a. Enabled Enabled Armed Armed ' b. Enabled Not Enabled Not Armed Armed c. Not Enabled Enabled Armed Not Armed d. Not Enabled Not Enabled Not Armed Not Armed
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REACTOR OPERATOR Page 69 QUESTION: 063 (1.00) f Which-ONE (1) of the following reasons explain why the RN [ Nuclear Service Water; System is aligned to the SNSWP [ Standby Nuclear Service Water Pond) in cold weather? , a. Prevent severe ice accumulation on the surface of the pond, b. Normally closed Pump House isolation valves (1RN3A and 1RN4B) from the SNSWP may fail to automatically reposition on a Pit Emergency Low Level due to ice accumulation, c. Prevent a loss of RN due to ice formation in the source and inta):e section structures in Lake Wyli d. Prevent a loss of RN due to ice formation in discharge etructures in Lake Wylie, ik ,. , ri h
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C) L U REACTOR OPERATOR Page 70 QUESTION: 064 (1.00) Control Rod F8 has been determined to be trippable but inoperable due to an open in the lift coil circuit but it is only two steps lower than the remaining rods in the grou Which ONE [1] of the following statements describes the action which must be taken? (TS 3.1.1.1 and 3.1.3.6 are attached) Determine that the Shutdown Margin requirement of TS 3.1.1.1 is satisfied within 1 hour and be in Hot Standby within 6 hours, b. Reduce turbine power to equal to or less than 75% of Rated .- Thermal Power within 1 hour and reduce the HI Neutron Flux Trip Setpoint to 85% within the following 4 hours, c. Determine the Shutdown Margin of TS 3.1.1.1 at least once every 12 hours and that FN Delta H is within limits.
d. Maintain the remaining rods in the group within plus or minus 12 steps of F8 while maintaining the rod sequence and insertion limits of TS 3.1.3,6 during subsequent operatio .
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QUESTION: 065 (1.00) Which10NE (1) of.che fo'. lowing responses would be'an indication of a control bank failure to .e during a power increase from 50%. power to 100% power with rod control in automatic? a. " ROD CONTROL SYS URGENT FAILURE" in alar b. Increasing Tavg and possible "NC SYS HI/LO TAVG" in alar c. High' pressurizer level indication and possible alar d. " ROD CONTROL SYS NON-URGENT FAILURE" in alar .. i i ry t. 5 4
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_ _ _ - _ - - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - - - - _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ , _ _ _ - O O REACTOR OPERATOR _- Page 72 QUESTICH:-066 (1.00)_ AP/1/A/5500/08 " Malfunction of Reactor Coolant Pump" Immediate Action directs operators to trip an NC Pump if the Pump Lower Bearing Temperature is 225 degrees F or greate Which'ONE [1].of the following-reasons explains why the NC pump is tripped if the Pump Bearing Temperature Limit is exceeded?- a. Minimize the risk of molting bearing babitted surfaces-and impair NCP coastdow b. Reduce the possibility of seal leakoff flashing to steam and - damaging the NCP seal c. Increased beariTg friction may result in uncontrolled rise in bearing temperature d. Minimize the risk of an NC Pump Lube Oil fir . MASTER
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V)- REACTOR OPERATOR Page 73 QUESTION: 067 (1.00) The following Unit 1 plant conditions exist:
' -
Operating in accordance with OP/1/A/6100/03 " Controlling Procedure For Unit Operation".
- Reactor power is 18% proceeding to 85%.
-
NC { Reactor Coolant] Pump 1A was tripped when the 1B Pump-Lower Bearing Temperature exceeded the alarm setpofn The operator recognized that the wrong NC pump was tripped and tripped NC Pump 1 Which ONE (1) of the.following actions is required to be performed?- a. Start NC Pump 1B in accordance with OP-6150/02A "NC Pump operation".
b. Enter OP/1/A/6100/02 " Controlling Procedure For Unit Shutdown", c. Ensure the reactor tripped and go to EP/1/A/5000/1 " Reactor Trip or Safety Injection".
d. Enter Technical Specification 3. h alariza1a, g, l i
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O O REACTOR OPERATOR LPage-74 < QUESTION: 068 (1.00)~ The following' plant condition; exist:
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A normal plant cooldown is in progres The computer alarm for-the KC Surge Tank indicates a need for makeup from the domineralized water syste The water level i n Surge Tank 1A is slowly DECREASIN Locally,-there is no apparent cause for the. decreasing surge tank leve .
- The associated makeup valve to surge tank 1A is fully ope Which ONE'[1] of the following explanations is the.cause of decreas'ing . surge tank level?
a. There is a leak in the KC heat exchanger, b. There is a leak in the Letdown heat exchanger, c. There is a leak in the Reactor Coolant Pump Thermal barrier heat exchanger, d. The plant cooldown rate is excessiv ,lD ' _ If
O b Page'75 z REACTOR OPERA'/OR QUESTION: 069 (1.00) When responding to an ATWS condition in accordance with EP/1/A/5000/2A1
" Nuclear Power Generation /ATWS", which ONE [1] of the following reasons explains why Pressuriger precsure is verified to be less than the PORV setpoint?
' a. Ensures that a LOCA will not be created by a faulty PZR POR Ensures that the PZR PORVs have shut after opening to Iaduce NC pressure following tripping the turbin c. Allows sufficient boration flow rate to the NC system, Prevent reactor pressure from exceeding safety limits during an overpressure transient.
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s O 0: REA' "OR OPERATOR- Page 7 QUESTION: 070 (1.00) The following-plant-conditions exist on Unit 2:
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The plant is operating at 83% power, s
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A control rod drops into the cor One Negative Rate Trip Bistable light is li About one minute later a second rod drops into the core in another quadrant.,
-
A second Negative Rate Trip Bistable light is actuate Power and Tave have decreased slightly but otherwise operations are undisturbe _ Which ONE [1] of the following actions should be taken? , Trip the reactor and enter EP/2/A/b000/01 " Reactor Trip or Safety Injection", Reduce power in accordance with OP/2/A/6100/03 " Controlling Procedure-for Unit Operation" in preparation for entering OP/2/A/5100/02 " Controlling Procedure For Unit Shutdown".
c. Recover the dropped rods in accordance with AP/2/A/5500/14
" Control Rod Misalignment", Reduce power and enter AP f 7/A/5500/16 " Case IV, Power Range Malfunction".
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0 0 REACTOR OPERATOR Page 77 QUESTION: 071 (1.00)- The following plant conditions exist on Unit 1:
- The plant has-experienced a main steamline ruptur EP/1/A/5000/2D1 " Imminent Pressurized Thermal Shock" has been '
entere Which ONE (1) of the following statements describes the major-actions to be taken? a. Maintain NC System cooldown rate; maintain NC System pressur b. Maintain NC System cooldown rate; depressurize NC Syste c. Stop NC System cooldown; maintain NC System pressure, d. Stop NC System cooldown; depressurize NC System.
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__ _ . . . . . . . . _ _ _ 7 .- O O REAlf0R OPERATOR Page 78-1 _- _ _ - QUl!GTIOll: 073 (1.00) . The following plant conditions exist:
- Unit 1 is at 351, power increasing to 1000 power in accordance .E with GP/1/A/6100/03 " Controlling Procedure For Unit Optiratior ".
_
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AP/1/A/5500/23 " Loss of Condenser Vacuum" was entered due to Condenser vacuum Indication DECREASIll Condencer vacuum is currently stable at 24.5 inches of li Exhaust Hood Temperature is 228 degrees F and decreasing slowl Which Ol4E (1) of the following actions should be taken? ~ r a. Trip the turbine; trip the reactor; enter EP/1/A/5000/01 = " Reactor Trip or Safety Injection".
' __ b. Trin the turbine; enter AP/1/A/Sa00/u2 " Turbine Generator Trip".
c. Stop the powar increase and reduw powet in accordance with ,
-
OP/1/A/6100/03 " Controlling Procedure For Unit Operation" -
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Enclosure 4.2 " Power Decrease".
-
- Increase load per OP/1/B/6300/01 " Turbine Generator".
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QUESTIO!!! 073 (1.00) EP/1/A/5000/03 "Lotas of All AC ' -%" includes the Immediate Actions ! Verify Turbino Tri i Which OllE (1) of the following rossons is the basis for " Verify!ng Turbino Trip"? l a. Limit CA Pump thrust due to operation near the runout conditio b. Prevent an urcontrolled cooldow c. Limit depletion of steam generator inventor d. Provent an unnecessary Safety Injection.
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l QUESTION: 074 (1.00) The following plant conditions oxists -!
- A loss of all AC power has occurre EP/1/A/5000/03 " Loss of All AC Power has been entere Doth D/Gs (Diesel Generators) are RUNNING but neither has !
automatically supplied its essential bu Which ONE (1) of the following actior.s is required to be performed ! immediately?
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a. Relet the train related D/G load sequencers and auto load the D/G : b. Manually load both D/G ,
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c. Stop and then man.lally start both D/Gs from the Control Room, Stop both D/G t , e i i t i
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QUESTInit: ~ 07 5 (1.J0) ' Which OllE [1] of the tv)1ov',cy statomonts describes the result of taking Inverter 1EIB Nanual Byptiva Switch to the " BYPASS" position with the "Ill SYllC" light not lit?
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a. There is no power on the a1tornato source; therefore power would be lost to IERP b. Si nco the "Ill S) ', ," light must be lit before the switch can be moved, the switen would not operat c. Since 1ERPB and IVRD would momentarily bo connected or.t of phase, the switch would probably be damage ' d. The switch would operato; but the connection would not be transferred because the Kirk-key interlock has not boon inad ! h
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'O O , . REACTOR OPERATOR Page 82 f i
I i QUEST 10!it 076 (1.00) During an authorized releano of liquid wanto, EMF-49 alarm Which ONE :
(1) of the following statements describes the action to be taken by the _,
operator? ,
!
a. React the alarm and wait to see if the alarm return b. Write a work request to have the alarm chocked by IA c. Ilotify Operations Duty Enginater of the alarm conditio [ Ensure automatic actions occurre ,
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O !" O REACTOR OPERATOR Page 83 ) QUESTIoll: 077 (1.no) Which OllE (1) of the following descriptions identifies the location of a fire which would automatically actuate a llalon Fire Protection System? Operator's Storage Rooms, b. Document Cintrol Storage Are c. Cable lloom Corridors on Unito 1 & Steam Production Office Buildin e i
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O o Pago 8 REACTOR OPERATon > i QUESTI014: 078 (1.00)
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Which OliE (1) of the following-individuals becoraes the Fire Captain when-a fire is reported on-site? !
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a. Superintendent of Operation b. Safoty Associato/ Specialist, Fire Protection, c. Shift Superviso d. Assistant Shift Superviso , e l f
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' .O O-HEACTOR Oli tATOR Page 85
i QUESTION: 079 (1.00) f-The following plant conditions exist i
- EP/1/A/5000/2C1 " Loss of Secondary lleat Sink" has been entere ! -
Feedwater flow cannot be establishe Which ONE (1) of the following feed and bleed paths will be usnd? a. Feeding to the stea'.a Generators [S/Gs) and using steam durap to blood stea {
'
b. Feeding to NC (Roactor Coolant] System and using Pressurizer-PORVs to bleed stea c. Feeding to NC System and using Normal Letdown to blood the NC-
.
Syste d. Feeding to the S/Gs and using S/G PORVs to bleed steam. .
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QUESTION: 080 (1.00) If EP/1/A/5000/2C1 " Loss of Secondary licat Sink" has boon entered, which ONE (1) of the following reasons explains why NC (Roactor' Coolant) Pumps are stopped in loops with Steam Generator wide rango level less than 5%?._ a. Stops adding unnecessary hea ' b. Provant NC Pump seal damago due to steam formation in the Number 1 Sca c. Allows restoration of food to the Steam Generato d. Allows cooling the associated loop by natural circulatio <
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.m REACTOR OP);RATOR O c Page 87 QUESTIOll: 083 (1.00)
hhen operating in accordance with EP/1/A/F000/2C1 " Loss Of Secondary liont Sink", which OllE (1) of the following reasons explain why subcooling and CETs (Core Exit Thermocouples) are monitored throughout the procedure? a. They provide the feed and bleed initiation criteri b. They provide the criteria for exiting from EP2C c. They provide the criteria for securing the reactor coolant pump d. They provide the criteria for feeding the steam generators from the R11 System.
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_ _ _ _ - _ _ _ _ _ - _ _ . . _ . . . O O REACTOR OPERATOR Page 68 QUESTIoll: 082 (1.00) llc (Reactor Coolant) System activity has ranged from 1.1 microturies per gram done equivalent I-131 to 1.5 microCuries por gram dose equivalent , I-131 for the lact 48 continuous hours. Which OllE (1) of the following s actions should be taken? (Technical Specification 3.4.8 is attached) a. Shutdown the reactor to liot Standby and notify the llR b. Trip the reactor; enter "P/1/A/5000/1 " Reactor Trip or Safety Injection"; and notify the liRC.
I c. Reduce reactor poh-r to lens than 75% and notify the 11R d. Reduce reactor power to 501, and notify the 11RC if activity increase , w' ; a f
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:
QUESTIOll: 083 (1.00) While in the Automatic mode of operation, which OllE (1) of the following - IMMEDIATE ACTIOlis should be taken in the event of continuous rod movement' a. Select " Manual" on Rod Control and verify Manual Control by ; driving roda in then ou i
,
b. Select " Manual" on Rod Cor. trol and verify all motion stops.
' c. Selnct " Manual" on Rod Control; terminate boron concentration-change; verify all snotion stops, d. Select "lianual" cn Rod Control; terminate turbine load chango and boron concentration changes; verify all motion stop . I
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REACTOR 'PERATOR . Page 90
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i I: I QUESTION: 084 (1.00) control Rod Group D is at 30 steps when the DRP1 (Digital Rod Position ' Indication) suf fers a Data "A" Failuro llalf Accurat:y casualty. Which , ONE (1) of the following describes the group indication following the casualty? a. 18 step * b. 24 stop ,
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c. 30 step d. 36 step ,
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O O : REACTOR OPERATOR Pa9e 91 i i QUESTION: 085 (1.00) , EP/1/A/5000/01 " Reactor Trip or Safety Injection" Step 3 Action / Expected norponse roads: Verify olther 1 ETA or 1ETB is energize . Which ONE [1] of the following statemento describes hov 1 ETA or 1ETE are i verified energized? a. ETA or ETB Line Voltage equal to or greater than 4160 volt b. Dlecol Generator A and B Blackout Load Sequencer Actuated Status Lights are dar c. ETA and ETB Hormal or Standby incoming breaker amps are greater than d. ETA and ETB X, Y, and Z Phace UV Status Lights-are dar , i b h
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. REACTOR OPERAT')R Page 92 i i
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QUESTIoll: 036 (1.00) The following plant conditions exist on Unit l' i
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Reactor is shutdown in Mode 3.- - ' ~ Pressurizer level is 63 NC (Roactor Coolant) System pressure is 1305 psi PRT [ Pressurizer Relief Tank) pressure is 6 psi l t Ii one pressurizer PORV is leaking slightly, which ONE [1] of the follow.4.ng + .iperatures will be ine'.icated on the Relief Valve discharge RTD [1NCRD 5940)? a. 247 degre7.s b. 363 degrees degrees degrees .
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O O REACTOR O'PEIUtTOR Page 93
:
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' - QUEST 10!!: 087 (1.00)
i EP/1/A/S000/1A1 "liatural Circulation Cooldown" Step 11 directs the operators to: j Monitor HC Syctem cooldown: Which ONE [1] of the following sets of indications demonstrate existence : of adequate netural circulation? a. Core exit thermoccuples decreasing; T-HOT decreasing; NC - uubcooling increasing.
' b. Core exit thermocouples decreasing; T-110T decreasing; NC subcooling decreasin ; c. T-i!OT eteady or increasing slowly; T-COLD decreasing; HC subcooling increasin d. Core exit thermocouples steady or increasing slowly; Steam Generator pressure decreasing; NC subcooling decreasin ,
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G O R'EACTOR OPERATOR QUESTIO!1: 088 (1.00) The io)1owing Unit I conditions exist following a 80 (Safety injection):
- 110S pronsure in 1050 psi Tave la b57 Cogreco All ESP Monitor Panel lights are properly LI Containment Pressure is 2.8 poi :!{F53A [B) indicates 4 R/h NV/SI flow is 500 gp ItCS boron 10 500 pp Ccre ilurnup is 200 EFP Which O!3E (1) of the following actions should be performed by the operatof?
a. Borate the 11CS to 850 ppm Boron for proper SDM prior to cooldown a to loss than 200 degrees F lic Wide Range Temperaturo.
' b. Ensure 14V-202 and 11V-203 are ope c. Secure NS ' imps that are in servic Secure all llc pumps.
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O O REACTOR OPI:H.% TOR Page 95 QUESTION: 089 (1.00) i
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The following plant conditions exist on Unit 18
- A LOCA has occurred and the plant is being recovere ! - NC prmouro is 950 psi i -
Pressurizor levol is 4%. 1
' - All lic Pumps have been r,ocure Containment sump level is 11 foo FWST level is 30%. ' - Cold Leg Recirculation has boon initiate FWST [ Fueling Water Storage Tankj suction valvo (FWS5B) is open and will not clos ;
Which ollE [1) of the following statemet.to describes the action to be taken?
,
a. Dispatch an operator to close FW55B manuall i b. Let the pumps run until FWST level is 11% than shift to-natural- ' circulation cooldow c. Closo sump valvo NI1848 and stop 11D Pump B to provent excessive FHST depletio , d. Shif t to llot Log Rocirculation to minimize the depletion of the FWST until FW55D can be close fi/]n';j+pID-
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QUESTION: 090 (1.00) l The following Unit 1 plant conditions exist I
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Shutdown in Hodta !
- Proscurizer loval is daereasing slowl ! -
NC [ Reactor Coolant) pressure is decreasing slowl ND (Residual llcat Removal) flow is hig ,
- "ND & NS ROOMS SUMP LEVEL EMERG llI" is in alar Which ONE (1) of the following IMMEDIATE ACTIotM must be performed?
a. Secure <>porating ND pump (s) and SIIUT ND System suction and dischargo v'ilves to the NC Syste o. Sect, e operating ND and NC pump (s). Secure operating HD pump (s) and manually initiate Phase _ n
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isolatio d. Eocuro operating ND and NC pttmp(s) and manually initiate Phase B isolatio >
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O O 4 REACTOR OPERATOR Page 97-QUESTIOll: 091 (1.00) . During Mode 1 operation, with the Prosauro control Solsetor Switch selected to the 1-2 position, which OllE (1) of the following statomonts describes the offacts of Pressurizer Prosauro control Channel I failing t LOW? (NOTE Assumo no operator action is taken] ! a. All heators on; PORV NC34 blocked; NC pressure rises;.PORVs 32 & 36 open* NC prosauro modulates betwoon 2235 and 2315 psi Sprays full on; PORV 34 opens; PORV 34 closos at 2185 psigl Low Pressuro Rocctor Trip at 1945 psig; Low Pressure SI at 1845 , psi ' c. All heators on; PORV 32 & 36 open; PORV 32 & 36 close at 2165 psig; spray valvos modulate 11C pronsure betweer 2260 and 231n poi _; r d. Sprays full on; PORVs 32 & 36 blocked; all heators on at 221 poig; heators modulate to maintain NC prassure betwoon 2210 and 2218 psi !
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_ - O O REACTOR OPERATOR Page 98 QUESTION: 092 (1.00) . EP/1/A/5000/2A1 4 Nuclear Power Generaticn/WrWS" has been entere Which ONE (1) of the fi'. lowing sets of indications is used to verify Reactor Trip?' a. Rapidly decret t.J ng power, zero start up rate, Tave steady.
i b. Rod bottom lights LIT, Intermediate Range amps decreasing, Control Rod Drive M-G Set breakers ope c. Rod bottom lights out, positive 7 tart up rate, SI Actttated-status light LI d. P.od bottom-lightr. LIT, Intermediate Range amps decreasing, Reactor Trip breaker s ope !
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r i O O .I i REACTOR OPERATOR Page 99 '] i QUESTIOll: 093 (1.00) i The following plant conditions exists
- Reactor startup is in progres Source Range Channel 1431 indicatos 7E3 cp ! - Source Rango Channel 1132 indicates BE3 cp Intermediate Range Channel 1135 indicates- 3E-11 amp Intermediate Rango Channel 1136 indicates 1.5E-10 amp ,
Which ONE (1) of the following-statomonts describes the status of the nuclear instruments?' [liotot Figure CH-IC-Elill-3 below)
:
a . 143 5 is over compensate b. 113 5 is under compensate ' c. 143 6 is over compensated, d. 113 6 is under compensate "O"" E '
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i QUESTIOll: 094 (1.00) !
!
Which OllE (1) of the following reasons describes the basis for _
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'
maintaining the pressure in a ruptured S/G (Steam Generator] less than 1125 psig? ,
,
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a. Provide a margin to lifting the steam line safety valve b. Minimize secondary system contamination from the 11C '[ Reactor ! Coolant) Syste c. Prevent excessive cooldown of the lic System from overfeeding the S/ ' d. Allow lic System depresa,urization and rapid equalizing with 5/G pressur i f i-r L t t COPY .
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REACTOR OPERATOR Page101 - r l QUESTION: 095 (1.00) . The following plant conditions exist:
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AP/1/A/5500/06 " Loss of S/G Feedwater" was entered due to a insa ;
*
of feedV '.or s"pply to the S/G Case II uctions of APOS have been completed through Step 1 Hotwell levels are 0.22 feet CA flow is 590 gp i Which ONE (1) of the following actions should be performed?
'
[ NOTE:
'AP/1/A/5500/06 is attached)
i a. Shift CA suction to the CACST and US ' b. Isolate RN from the CA Pump suctio c. Run only one CA Pump.
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d. Stop all CA Pump . E h CDPy :
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O O REACTOR OPERATOR Page102 QUESTION: 096 (1.00) Which ONE (1) of the following conditions must be satisfied before blackout switchgear control power may be swapped back from the alternate supply to the normal supply? a. The undervoltage relay on CDA must be rese b. 250 VDC Battery 1DPD or 2DPB breakers are closed, c. Supply breakers to 4160 & 600V control power'and CAPT control power are closed, d. Battery charger ECA has its AC input and DC output breakers close i
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' REACTOR OPERATOR Page103
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-Q'ESTION: 097 (1.00) 1 Which ONE (1) of the following reasons describes the MOST LIKELY cause of a reactor trip during a sustained loss of Instrument Ai a. NC (Reactor Coolant) Low Pressure due to excessive cooldown from S/G [ Steam Generator) PORVs failing open, b. S/G Lo Lo Level dite to CP (Main Feed) control valves f ailing close ,
'l c. NC High Pressure due to failure of pressurizer spra ]
d. Pressurizer High Level _due to failure of Steam Dumps to operat I
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O O-REACTOR OPERATOR Page104 QUESTION: 098 (1.00) In addition to ensuring the Unit 2 control operator is notified, which ONE [1] of the following actions must be performed immediately if Instrument Air pressure falls to the low pressure alarm setpoint on-Unit
a. Dispatch operators to VI (Instrument Air) compressor panels to verify proper compressor operation and to start the backup temporary VI compressor, b. Dispatch operators to VI compressor panels to. verify proper-compressor operation and dispatch operators to locate and isolate the leak on the VI syste c. Dispatch operators to locate and isolate the leak on the VI system and start the backup temporary VI comprensor, d. Dispatch operators to isolate the leak on the VI system and to open IVS-78 locall hiASTER COPY
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QUESTION: 099 (1.00) If a D/G [ Diesel Generator) is supplying its associated bus as a result of a Loss of Offcito Power, which ONE (1) of the following reasons explains the basis for NOT exceeding 5750 KW on the D/G7 a. To limit generator winding temperature ris b. To ensure RH can provide adequate D/G coolin c. To limit engine crankshaft and piston stresse d. Tc encure an underfrequency condition does not occur oil the bu ,
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(********** END.-OF EXAMINATION **********)
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1K101 (3.6/3.?) b ^. n and SRO 194 (;a1K101 ..(KA'G) ANSWER: 002 (1.00) REFERENCE: OMP * t 8, Re , Page 2, Para. 3. KA 13 001K101 [3.6/3.7) Both RO and SRO 194001K101 ..(KA'C;
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~ REACTOR OPERATOR 'Pago107
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REFERENCE:-
'CNS Directive 3.8.6 (TS), Rev. 22, Page-3, Para.=4.1,. Page-4,. Para. 4.4_
KA'194001K103 (2.8/3.4) ' Both RO and SRO
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.
ANSWER: 004 (1.00) REFERENCE: CNS Directive 3.8.8, Rev. 27, page 16, Para. 5.17.2 &-5.1 ;4 KA'194001K104 (3.3/3.5) Both RO .'nd SRO 19%001K104 . . (KA's)
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,- ANSWEP: 005 (1.00) . REFERENCE: Ct!S Directive 3.8.5 (TS), Revision 18, Page 3, Para. 6.1.4.
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Both RO and SRO 194004K104 . . (KA's) l
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Both RO.trd SRO- , 194001K109- . . (KA's) Af .i'ER: 007 (1.00) REFERENCE: iD 2.11.4 (SS), Rev. 16, Page 1, Para. 2. KA 194001K114 (3.3/3.6] *
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-REFERENCE:
OMP 2-17, Rev. 23, Page 2, Para. 6. A106 [3.4/3.4) , Both RO and SRO 194001A106 ..(KA's) . ANSWER: 009 (1.00) 'REFEkENCE: OMP 2-31, Rev. 17, Page 1, Para. KA 194001A106 (3.4/3.4) RO only
194001A100 ..(KA's)
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ANSWEP: 010 (1.00) REFERENCE: OMP 2-10, Re , Page 6, Para. 8.5.A& KA 394001A107'(2.5/3.2) Both RO and SRO 194001A107 ..(KA's) P h;fId
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- REFERENCE:
, OMP 2-11, Rev -16, Page 2, Para. 4.4 and Page 3, Para. KA'194^01A109 (2.7/3.9] Both RO and SRO 194001A109 ..(KA's) ' ANSWER: 012 (1.00) , REFERENCE: OMP 2-16, Rev. 12, Pace 4, Para. 6 . 0 . 11 . KA 194001A109 [2.7/3.9) Both RO and SRO
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194001A109 ..(KA's) ANSWER: 013 -( MASTEg m3 NN
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,
194001A111 ..(KA's) ANSWER:- 014 (1.00) REFERENCE: OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 19, Para. 8.6.- KA 001000K103 [3.4/3.6] Loth RO and SRO 001000K103 ..(KA's) ANSWER: 015 (1.00) u, REFERENCE:
.OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 25, Para. 2.4.A.1.a.
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un
-
0: O REACTOR OPERATOR Page112
? ANSWER: 010 (1.00)
Ed.- REFERENCE: OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 10, Para. ; KA 001000K558 [2.7/3.2] Both RO and SRO , 001000K558 ..(KA's) .
.ANSW'. .: 01'l- (1.00) REFERENCE:
' OP-CN-IC-IRE, Rev.'07/03-16-19/GFW, page 27, Para. 1 KA 001000G010 [3.3/3.5] 1Ro only 001000G010 ..(KA's)
.
ANSWER: 018 (1.00) b.
.
.
CE!Jit j 11 . eng p{ , :> 0 r,y & _ W d ..,
<4,, , - -
x-.- + , -
,
O O - -
-
REACTOR' OPERATOR Page113
= REFERENCE:
OP-CN-CNT-CNT, Rev.-10/06-02-92/JWH,- Page 28, Para. 2.4. ^ Catawba Exam Bank Question NCP-13-KA 0030u0K201 (3.1/3.1) , Both RO and SRO 003000K201 ..(KA's) ANSWER: 019 (1.00) REFERENCE: OP-CN-PS-NCP, Rev. 04/09-18-90/GFW, Page 17, Para. 4.b 9(a)
.
KA 003000A401 [3.3/3.2) RO only 003000A401 ..(KA's) ANSWER: 020 (1.00) REFERENCE: Technical Specification Bases, Page B 3/4-4-1 KA 003000G006 (2.7/3.8) Both RO and SRO
.n 003000G006 ..(KA's)
L IM A.n e IY%j $
ine7 bdfk,,,
- ~
_ . _ ,.... . _._,._ _-... _. -
.._ . _ . . _ . ' '
OL O
.' REACTOR OPERATO Pageli4-021 (1.00)- ' -ANSWER:
c.-
,
REFERENCE: OP-CN-PSS-KC, Rev. 13/03-03-92/GFW, Page 9, Para. B. OP-CN-PS-NV, Rev. 04/04-10-92/PEV, Page 25,' Para.-B.14. KA.004000A401 (3.8/3.9)
-Both RO and SRO 00400CA401 ..(KA's)
ANSWER: 022 (1.00) REFERENCE:
.OP-CN-PS-NV, Rev. 04/04-10-92/PEV, Page 26, Para. Catawba Examination Bank Question NV-89-KA 004010A205 (4.1/4.3)
Both RO and SRO 004010A205 ..(KA's)
-ANSWER:-- 023 (1.00)
b.
' MASTER CDPY
. -- ..
.
.
...
'
. 10 0 REACTOR OPERATCR lPage115 REFERENCE:
OP-CN-ECCS-ISE, Rev.'10/03-11-92/CTK, Page 16, Para. KA 013000A105 [3.4/3.6] Both RO and SRO
.;
013000A1.05 ..(KA's)
,
ANSWER: 024 (1.00) 't REFERENCE:
-OP-CN-ECCS-ISE, Rev. 10/03-11-92/CTK, Page 19, Para. F. KA 013000A403 (4.5/4.7) ~
3 only
.13000A403 ..(KA's)
-ANSWER: 025 (1.00) ,
'
~ REFERENCE:
-
Catawba Examination Bank Question EDA-15' KA 014000G008 [2.9/3.1] ' Both RO and SRO 014000G008 ..(KA's) =-
'
. CaPy 9 9 + , y e - - - - , - -
--- - - _ _ - _ _ _ ____ _ _ __ .
l O O REACTOR OPERATOR Page116 ANSWER: 026 (1.00) REFERENCE: OP-CN-IC-ENB, DWG CN-IC-ENB-3 KA 015000A303 [3.9/3.9] Both RO and SRO 015000A303 ..(KA's) o
ANSWER: 027 (1.00) REFERENCE: OP-CN-IC-END, Rev. 09/03-19-92/JLY, Page 11, Para. KA 015000K604 [3.1/3.2] Both RO and SRO
-
015000K604 ..(KA's)
.
ANSWER: 028 (1.00) MASTER m
,
b i' uY >4
- - - _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ - - - _ - _ - ____ _ __ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ - _ _ _ _ - _ _
;
;O 0; REACTOP OPERATOR ': Pago117
*
-REFERENCE:
OP-CN-IC-ENB, Dev. 09/03-19-92/JLY, Page 7, Pcca. C. KA 015000K406 [3.9/4.2) Both RO:and SRO
,
015000K406 ..(KA's) ' ANSWER: 029 (1.00)
'
b.
' REFERENCE:
-OP-CH-IC-ENA, Rev. 05/08 05-90/GFW, Page 20, Para. 1 t KA'017000:~101 [3.2/3.2) .Both RO end SR K101 ..(KA's)
. NSWER: 030 (1.00) d.
. ,
. REFERENCE: : Catawba' Examination Bank Question CNT-VV-9
' KA 022000A404- (3.1/3.2) _
<
Both RO and SR '.'-404 ...(KA's) OY g prm ovt .
. ,
. _ _ . .. . . .. . . _ _ . _ - . .
*
LO O REACTOR OPERATOR Page118' fANSWER: 031 (1.00) . REFERENCE: Catawba Technical Specifications, Page B 3/4 6 - 5 ~.
-KA 025000K301 [3.8/3.8) .
Both RO and SRO 025000K301 ..(KA's) ' ANSWER: 032 (1.00)
- REFERENCE:
OP-CN-CF-CF, Ru / . 09/02-25-92/JLY, Page 29', Para. KA 059000A411 [3.1/3.9)
--RO only 059000A411 ..(KA's)
ANSWER: 033 (1 00) .
-
.- -
[Eurica y Lgq Pf
%[jDi# m )( y a-y -
$
,p_ .-. -- ,, . . . . ..
.
- . -
O O * 4 REACTOR OPERATOR Pago119 ,
.,.
- RE'"ERENCE :
.OP-CN-CF-CF, Rev. 09/02-25-92/JLY,-Page 11, Para. 4. .
}
KA'059000G010 (2.9/2.9)
.RO.only 059000G010 ..(KA's)
,
. ANSWER: 034 (1.00)
-- a ..
REFERENCF.: Catawaba Examina on Sank Question CA-31 KA 061000A202 [3.2/3.5) Both RO and SRO
-061000A202 ..(KA's) ~ ANSWER: 035 (1.00) - ~
',. REFERENCE: OP-CN-CF-CA, Rev. 10/O'a-19-22/ELR KU 061000C002 . ( 3. 0/ 3. 8 ) - RO only 061000G002 ..(KA's) ng
. a j p w g ,g , !. b:- .
_ ~ . . . . .. .- - . - -- - , . - hi
.A -~ U-REACTOR OPERATOR Page120~
= ANSWER: 036 -(1. 00) i REFERENCE:
bP/0/B/6350/07, Page 1, Para. KA 063000G010 ( 3 . .',/ 3 . 2 ) Both RO and SRO 063000G010 ..(KA's) ANSWER: 037 (1.00) , REFERENCEt
'
OP-CN-WE-EMF, Rev. 09/le-11-91/CTK, Page 13, Para. 3.a.4 < KA 072000A202 (2.8/2.9) RO.only 072000A202 ..(KA's)
. ANSWER: '038 (1.00)
b.- b$ 0 O b *)Gfy ?.e p ), COPY ..
& . - r , p,' f-
_ - _ - _ . _ . . _ _ _ _ . _ . _ O O : REACTr/R OPERATOR Page121 REFERENCE:
*
OP-CH-WE-EMF, Rev. 09/12-11-91/CTK, Page 12, Par .c.
't KA 072000G007 [2.6/2.9) Both RO and 3RO 072000G007 ..(KA's)' ANSWER: 039 (1.50) , . . . [ 0.5 each) _, REFERENCE: Drawing CN-PS-NC-09 KA 002000K109 [4.1/4.4] RO~only a
002000K109 ..(KA's) l l
'
L ANSWER: 040 (1.00) l- . b. REFERENCE:
- OP-CN-PS-NC,' Rev. 08/02-25-92/GFW, Page 27, Par S.b.3.c. and.5.b, KA 002000K60J (3.1/3.6)
(; Both E3 and SRO l 002000K603~ ..(KA's) f [$ 7 :M ny hkfJ$j 73 )-j w[YA[g de COP!
- .
- ,
t [D (/ REACTOR OPERATOR Page122 ANSWER: 041 (1.50) [0.5 each] REFERENCE: Drawing CH-ECCS-JI-7 KA 006000K409 [3.8/4.1] RO only 006000K409 ..(KA's) ANSWER: 042 (1.00) REFERENCE: OP-CN-PS-IPE, Rev. 04/10-03-90/GFW, Page 17, Para. 2. KA 010000K201 [3.0/3.4]- Both RO and SRO 010000K201 ..(KA s)
.
ANSWER: 043 (1.00) e O! "
, w np hg) p l n ;; . ::, y
,
,
?
O O REACTOR. OPERATOR- Pagit123 -
' /
REFERENCE: OP-CN-PS-IPE, Rev. 04/10-03-90/GFW, Page 13, Para. 8. KA 010000A403 (4.0/3.8) , Both RO and SRO 010000A403 ..(KA's)
.
ANSWER: 044 (1.00) REFERENCE:
!atawba Examination Bank Question ILE-26 KA 011000A211 (3.4/ .6]
Both RO and SRO 011000A211 ..(KA's) * ANSWER: 045 (1.00) b.
, REFERENCE: ' Catawba Examination Bank Question IFX-30 ,
.KA'012000K501 [3.3/3.8] -
i Both RO and SRO I
- . 012000K501 ..(KA's)
'
l
- O li f ? T' .
l Ebb %M [, . l COPY
+
_ _ . . , . . ._. _ __ . . . _ . .., o g
- REACTOR; OPERATOR- . Pa ge12_4 ~ ANSWiR: L046 (1.00)' ,
d.-
.
ItEFERENCE: Catawba Examination Bank Question EDA-33 KA 014000G008 (2.9/3.1] , RO Only
,
014000G008 . . (KA's) ANSWER; 047 (1.00) ,
, REFERENCE:
.
D.rawing CN-PS-NC-5 KA 016000K101 (3.4/3.4]
~Both Ro and SRo
,
016000K101 . . (KA's)
-- ANSWER: 048 (1.00)
a;
t i
- -~
-
O *j; (ud ?! %
$5'dd 3 bh n
C~uHurw . l
O O REACTOR' OPERATOR Page125 REFERENCE:
.OP-CH-CF-IFE, Rev. 06/03-09-92/GFW, Page 21,_ para. L.3.a. 2 .KA 016000K106 [3.6/3.5]
RO;only 016000K106 ..(KA's) ANSWER: 049 (1.00) REFERENCE: OP-CN-ECCA-NS, Rev. 09/01-08-92/RJK, Page 11, Para. 2. ! KA 026000Guo9 [3.6/3.6]
'l 'O only R
026000G009 ..(KA's)- ' R 1 ANSWER: 050 (1.00) REFERENCE: OF'CN FH-KF, Rev. 06/11-28-90/GFW, Page 18, Para B. KA-033000K30J [3.0/3.3] Both RO and SRO i
!
033000K303 ..(KA's) ANSWER:- 051 (1.00) - I COPY _
~ . .- . .- .
. , . ..
O O REACTOR OPERATOR _ Page126'
-REFERENCE:
OP-CN-STM-St i, Rev. 09/02-06-92/RJF, Page 15, Para 2.a. &. KA 039000A302 [3.1/3.5)
'Both RO-and SRO -
039000A302 ..(KA's) ANSWER: 052 (1.00) REFERENCE: OP-CN-EL-EP, Rev. 06/03-30-92/ELR, Page 14, Para 7.a. & KA 062000G009 [3.2/3.3)
-RO only 062000G00e ..(KA's)
, ANSWER: 053 (1.00)
' *
REFERENCE: OP-CN-DG-EQG, Drawing CN-DG-EQG-12
.KA 064000K411 (3.5/4.0)
Both RO and SRO 0640003411 ..(KA*s) l'
*
*
I Cupy
..
. . . . . ~- .- . , .,.
, O O ~ REACTOR OPERATOR Page127 EREFERENCE: . OP-CN-SS-RFY, Rev. 02/12-11-89/SMD, l' age 22, Para 1 KA 086000K406 [3.0/3.3) Both RO'and SRO
,
086000K406 ..(KA's) ANSPER: 055 (1.00) REFERENCE: OP-CN-PS-ND, Rev. 09/02-27-92/GFW, Page 12, Para D. KA 005000K407 [3.2/3.5) RO only 005000K407 ..(KA's) ANSWER: 056 (1.00) REFERENCE * OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 22, Para 1 KA 007000K401 (2.6/2.9) RO only 007000K401 ..(KA's)
.
ANSWER: 057 (1.00) c g, );,Qll};f
;I'xl ; gg . p u ,,
' $ sib /(,,
.
&
O O? REACTOR OPERATOR Page128 REFERENCE: OP-CN-PSS-KC, Rev.-13/03-03-92/GFW, pages 10 & 11, Para B.3. K102 [3.3/3.4]
.Doth RO and SRO 008000K102 ..(KA's)
ANSWER: 058 (1.00) , d.
REFERENCE: -OP-CN-PSS-KC, Rev. 13/03-03-92/GFW, Page 10, Para B.3.c 4.a-. & KA 008030A304 (3.6/3.7] Both RO and SRO
,
0080' 0A304
. ..(KA's)
ANSWER: 059 (1.00) b.
REFERENCE: OP-CN-CNT-VX, Rev. 06/03-10-92/PEV, Page-10, Para KA 028000K601-[2.6/3.1] -RO only
.
028000K601 ..(KA's) ANSWER: 060 (1.00) 3 3 4 f4'Y"p 0hk e o f]k COPY
. . ._ .
'l
-
\.)
l REACTOR OPERATOR Page129 EREFERENCE: Catawba Examination Bank Question VP-6
,KA-029000K302 (2.9/3.5)
Both RO and SRO 029000K302 ..(KA's) ANSW?R: 061 (1.00) REFERENCE: Catawba Examination Bank Question FL-19 KA 034000A201 [3.6/4.4] RO only 034000A201 ..(KA's) ANSWER: 062 (3.00)
~ REFERENCE:
i OP-CN-STM-IDE, Rev. 05/02-25-92/DPM, Page 16, Para 5. KA1041020K417 -(3 7/3.9) Both-RO and SRO ' 041020K417 . (KA's) L ANSWER: 063 (1.00) f . copy
-
a _ _ ,
. . - .. -
)
.
o o-REACTOR OPERATOR Page130
- REFERENCE:
OP-CN-PSS-RN, Rev. 13/02-05-92/ELW, Pages 14 & 15, Para KA 076000G010 [2.7/2.9) RO only 076000G010 ..(KA's) ANSWER: '064 (1.00) REFERENCE: OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 28, Para 2. KA 000001G008 (3.2/3.8) Both RO and SRO 000001G008 ..(KA's) ANSWER: 065 (1.00) REFERENCE: Catabwa Examination Bank Question IRE-75 KA 000005G005-[3.1/3.3) Both RO and SRO 000005G005 ..(KA's) ANSWER: 066 (1.00) h0 .fd lyp*% r: ?,,pa a y u wy y 1 -- , a COPY
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, -REACTOR I: OPERATOR-- Pagc1311 . # ' ;f , , , : REFERENCE:
OP-CN-PS-NCP, Rev.-J04[09-18-90/GFW,.P3ge 39, Para , D LKA,000015K207'(2.9/2.9)
. --
i< Both RO:and-SRO -
--
-
L
-:000015K207 ..(KA's)
:
ANSWER: 067 (1.005
: REFERENCE:
.
'
AP/1/A/5500/04, Retype #7, Page 2, Para .t . ..
!
'KA-.000015G010 (3.4/3.4)-
.Bsth ROJand SRO 000015G010 ..(KA's)
.
' ANSWER: -068 (1.00)
' a' , .
-
-
' REFERENCE:
~
W C2tawba Examination Bank-Question.KC-3
-.KA~00002D '05 [3.1/3.1)
.
.Both RC c.d SRO .
000026A105 ..(KA's)' .
"
ANSWER: 069L (1.00)- , N ^ . L g '
@py~? , r, , ., ._, - . . 2 ,
. . .
'
'l A A tj U l REACTOR OPERATOR Page132 REFERENCE:
OP-CN-EP-CSF, Rev. 11/05-27-92/DPM, Page 11, Para 2. KA 000029K312 [4.4/4.7) Both RO and SRO 000029K312 ..(KA's)
-ANSWER: 070 (1.00) RZFERENCE:
EP/2/A/5000/01, Retype #13, page 1, Para 3 KA 000029G011 [4.4/4.6] Both RO and SRO 000029G011 ..(KA's) ANSWER: 071 (1.00) REFERENCE: Catawba Examination Bank Question CSF-60 KA 000040G012 [3.8/4.1) Both RO and SRO 000040G012 ..(KA's) ! ANSWER: 072 (1.00) l f n?l?O":n t, g y 3 7 - m
,s
/1r ,
.
b[ j;;< l.,
.- _ . . - . . . - - . - - , . . . . - - . - . , - . - . . _ . _ . .
O- O
: 1EEACTOR OPERATOR .Page133 - REFERENCE:=
AP/1/A/5500/23, Retype /6, Page 4, Para 8. OP-CN-MT-MT-1, Rev. 04/12-11-90/GFW, Page 28, Para 5. 'KA-000051A202 (3.9/4.1) Both RO and SRO 000051A202 ..(KA's) ANSWER: 073 (1.00) REFERENCE: OP-CN-EP-EPS, Rev. 08/07-22-91/CWN, Page 9, Para KA 000055K302 (4.3/4.6] RO only 000055K302 ..(KA's) f ANSWER: 074 (1.00) REFERENCE: EP/1/A/5000/03, Retype #10, Page 3, Para KA 000055A203 [3.9/4.7] RO only 000055A203 ..(KA's) 075 (1.00) : hpv//.3/J pg 7.,
. ANSWER:
-
,- o ies, ; ,:
N ,? : f! fh N! ) +- pp
.Vpyf f I
__ _
. -. . - .- ._ _ -. . - _ .O 0: ~ REACTOR-OPERATOR Page134 REFFRENCE:
OP-CN-EL-EPL, Rev. 04/01-27-92/CWN, Page 16, Para 2,1. KA 000057A101 (3.7/3.7)
'Both RO and SRO 000057A101 ..(KA's) -ANSWER: 076 (1.00) REFERENCE:
' Catawba Examinaticn Bank Question EMF-20 KA 000059A102 [3.3/3.4] Both RO and SRO 000059A102 ..(KA's) ANSWER: 077 (1.00) REFERENCE: OP-CN-SS-RFY, Rev. 02 /12 -11-89 / SMD , Page 20, Para 2.2. v.A 000067A203 [3.3/3.5) Both RO and SRO 000067A203 ..(KA's) ANSWER: -078 (1.00) .pp p c d p *n, f NN!;/ f pn4 vegay;"
- - . . . _ .- .
-,
n-,--
.
O O REACTOR OPERATOR- Page135 REFERENCE: OP-CN-SS-RFY, Rev. 02/12-11-89/SMD,.Page 26, Para 2.3. KA'000067G012 (3.4/3.4)
--Both RO and SRO 000067G012 ..(KA's)
^
ANSWER: 079 (1.00) REFERENCE: OP-CN-EP-CSF, Rev. 11/05/27/92/DPM, Page 38, Para & KA 000074K103-(4.5/4.9) Both RO and SRO 000074K103 ..(KA's) ANSWER: 080 (1.00) a.- REFERENCE: OP-CN-EP-CSF, Rev. 11/05-27-92/DPM, Page 17, Para KA 000074K304 [3.9/4.2) l l- 'Both RO and SRO
..(KA's) ~
l- 000074K304 l l-ANSWER: 081 (1.00)
*- u
' b Y ?Q i;
mm (7; ; , ~ bf
.
_ _ _ . . O O REACTOR OPERATOR Page136 REFERENCE: OP-CN-EP-CSF, Rev. 11/05-27-92/DPM, Page 17, Para KA 000074K311 [4.0/4.4) Both RO and SRO , 000074K311 ..(KA's) ANSWER: 082 (1.00) - REFERENCE: OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 33, Para 2.3. KA 000076G012 (2.9/3.1) Both RO and SRO 000076G012 ..(KA's) ANSWER: 083 (1.00)
- REFERENCE:
OP-CN-IC-IRE, Rev. 07/03-16-92/GFW, Page 28, Para 2.5.C. KA 000001G010 (3.9/4.0] RO only 000001G010 ..(KA's) ANSWER: 084 (1.00) n
/
h /'/ : c',_. n,,P
, .g ci
%
'
. s y'
, _ - - - - _ - - - - - - - - _ - - - - - - -- -
. , - . . , . .
O- O i I REACTOR OPERATOR- Page137 l REFERENCE: , l
--OP-CN-IC-EDA, Rev. 05/07 13-91/GFW, Page 12, Para 2.3.H.2. & KA 000003A201 (3.7/3.9)
RO only 000003A201 ..(KA's) ANSWER: 085 (1.00)
. REFERENCE:
EP/1/A/5000/01, Retype #13, Page 5, Step 3 KA 000007G010 f4.2/4.1) Both RO and SRO 000007G010 ..(KA's) ANSWER: 086 (1.00) REFERENCE: OP-CN-PS-NC, Rev. 08/02-25-92/GFW, Page 22, Para 2.2.A.1 _KA 000008K101 (3.2/3.7] ' Both RO and SRO 000008K101 ..(KA's) l-I
- ANSWER
- -087 (1.00)
l' 10 [j (a va:;6
: .er
-
: Q
" #Y 61 1 5, i[
k L I i- _ - .
- -
-
.
_ ._ ' O LO
' REACTOR OPERATOR Page138 REFERENCE:
EP/1/A/5000/1A1,-Retype #9, Page 6, Step 11 KA 000009A237 (4.2/4.5) Both-RO and SRO 000009A237 ..(KA's) ANSWER: 088 (1.00) - REFERENCE: Catawba-Examination Bank Question EP1-53 KA'000011A103 (4.0/4.0] Both RO and SRO 000011A103 ..(KA's) ~ ANSWER: 089 (1.00) REFERENCE: OP-CN-EP-EP2, Rev. 11/03-03-92/GFW, Page 28, Para 2.4. KA 000011A105 (4.3/3.9] RO only 000011A105 ..(KA's)
-ANSWER: 090 .(1.00) -- ! j ha) (y*@*7 u. imD a .Inf _
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, REACTOR OPERATOR Page139-
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REFERENCE: AP/1/A/5500/19, Retype /19, Page 17,. Para C.1-6 OP-CN-PS-ND, Rev. 09/02-27-92/GFW, Page 24, Para 2.3.J.~2 KA 000025G011 (3.6/3.9) Both RO and SRO 000025G011 ..(KA's)
' ANSWER: 091 (1.00) I REFERENC OP-CP-PS-IPE, Rev. 04/10-03-90/GFW, Page 23, Para 2.6. KA 000027A218 [3.4/3.5]
Both RO and SRO 000027A218 ..(KA's) l t ANSWER: 092 (1.00) REFERENCE: EP/1/A/5000/2A1, Retype #5, Page 1,- Para 1.
- KA 000029K301 [4.2/4.5)
. RO only l OLOO29K301 ..(KA's) l- ' ANSWER: 093 (1.00) 6-y;;p p o ~ ;l' Q -1
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Pago140
:
1 REFERENCE:
.OP-CN"IC-ENB, Rev. 09/03-19-92/JLY, Page 9, Para 2.1. KA~000033A202 [3.3/3.6)
Both RO and SRO 000033A202 ..(KA's) *
'
ANSWER: 094 (1.00) REFERENCE: OP-CN-EP-EP4, Rev. 11/12-12-91/PEV, Page 12, Para 2.1. KA 000038K302 (4.4/4.5] Both'RO and SRO 000038K302 ..(KA's) ANSWER: 095 (1.00)
- REFERENCE
AP/1/A/5500/06, Retype #10, Page 12, Para '
:KA 000054G012 [3.2/3.2]
Both..RO and SRO .. 000054G012 ..(KA's) ANSWER: 096- (1.00)
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O O REACTOR OPERATOR Pagel41 REFERENCE: Catawba Examination Bank Question EPF-1 KA 000058A101 [3.3/3.5] Both RO and SRO 000058A101 ..(KA's) ANSWER: 097 (1.00) - REFERENCE: OP-CN-SS-VI, Rev. 10/05-21-92/GFW, Page 22, Para 2.1.M.3. KA 000065A201 [2.9/3.2] Ecth RO and SRO 000065A201 ..(KA's) ANSWER: 098 (1.00) REFERENCE: OP-CN-SS-VI, Rev. 10/05-21-92/GFW, Page 21, Para 2.1.M. KA 000065G010 [3.2/3.3] Both RO and SRO 000065G010 ..(KA's) ANSWER: 099 (1.00) /' /: ,,
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I OP-Cli-EP- EPS , Rev.:08/07-22-91/CWN, page 20, para'2.3. KA.000056A214 ( 4 '. 4 / 4 . 6 ) l RO only I 000056A214 .(KA's)- l i i , i l l-l'
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REACTOR OPERATOR .Page :1 ANSW-ER KEY -
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MULTIPLE CHOICE 023 b 001 b 024 a
'002 d 025 a 003 b 026 b 004 a 027 d-005 d 028 a 006 a 029 b 007 c 030 d 008 a 031 a 009 b 032 b 010 c 033 b 011 a 034 a 012 d 035 d 013 c 036 c 914 c 037 b 015 a 038 b
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i 016 - d 039 MATCHING 017 a a 2 018 - b b 5- !. _ c- -7 020- a- MULTIPLE CHOICE ' 021- c- 040 b- , fy9 - 0 f% r* m , d
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_ L O O l ' REAUTOR OPERATOR Page 2 A N S'W E R KEY 041 MATCHING 060 d a 6 061 d b 5 062 c c 7 063 a MULTIPLE CHOICE 064 d 042 d 065 a 043 c 066 b 044 a 067 c 045 b 068 a 046 c 069 c 047 a 070 a 048 a 071 d 049 e 072 b _ 050 a 073 b 051 c 074 d 052 c 075 c 053 d 076 d 054 c 077 b 055 c 078 d 056 a 079 b 057 b 080 a
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.. O O REACTOR OPERATOR Page 3 ANSWER KEY 083 b 084 b 085 d 086 b - 087 a 088 d 089 c 090 b 091 a 092 d 093 d 094 a 095 d _ 096 c 097 b 098 a-099 c
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. TEST' CROSS REFERENCE Pags ll
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R 0- Exam -P W R React _or organized by Queation N-u n b e r QUESTION VALUE REFER':.NCE 001 1.00 9000251' 002 1.00 9000252 003 1.00 -9000254 004 1.00 9000255 005 1.00 9000256 006 1.00 9000259 007 1.00 9000260 008 1.00 9000262 009 1.00 90002F3-010 1.00 9000264 011 1.00 9000265 012 1.00 9000266 013 1.00 9000267-014 1.00 9000270 015 1.00 9000271 016 1.0 .00 9000273 018 1.00 9000275 019 1.00 -9000276 020 1.00 9000277 021 1.00 9000278 022 1.00 9000279 023 1.00 9000280 024 1.00 9000281 025 1.00 9000282 026 1.00- 9000283 027 1.00 9000284 028 1.00 9000285 029 1.00 9000286 030 1.00 9000287 031 1.00 9000288 032 1.00 9000290-033 1.00 9000291 034 1.00 9000292' 035 1.00 9000293 036 1.00 9000294 037 1.00 9000295 038- 1.00 9000296 039 1.50 9000297 04 .00- 9000298-041 1.50 900030 .00- 9000303-043 1.00 9000304
-
044 1.00- 9000305 045 046 1.00 1.00 9000306 9000307 (OIJf-{? M +
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i .00 9000308 048 1.00 9000309
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049 1.00 9000312
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O TEST CROSS REFERENCE O - Page -2 RO_ -E.x a m- P W-R Reactor O r g a n i z e'd- by Q u1e s t l'o n' N u-m b o r-
;
QUESTION' VALE REFERENCE 050 1.00 9000313 051- 1.00 9000314 052 1.00 9000315 053 1.00 9000316 054 1.00 9000317-055 1.00 9000318-056 1.00 9000320 057 1.00 9000321 058 1.00 9000322 ; 059 1.00 9000323 060 1.00 9000324 061 1.00 9000325 062 1.00 9000326-063 1.00 9000327 064 1.00 9000329 065 1.00 9000332 066 1.00 9000336 067 1.00 9000337 068 1.00 9000338 069 1.00 9000339 070 1.00' 9000340 071 1.00 9000341 072 1.00 9000342 073 1.00 9000343 074 1.00 9000344-075 1.00 9000345 076 1.00 9000346 077 1.00 9000347 078 1.00 9000348 079 1.00 9000349- -! 080 1.00 9000350 081 1.00 9000351 082 1.00 9000352 083 1.00 9000353 ~I 084 1.00 9000354 085 1.00 9000356 086 1.00 9000357
.087 1.00 9000358 088 1.00 9000359 089 1.00 9000360 !
090 1.00 9000363 3 p 7 ', .
*
. 091 1.00 9000364 g- _ _ 092 1.00 9000365 dl [o 3 L 093- 1.0 .0 .00 9000368 9000369 h 3 'E
-!
096 1.00 9000370 - 097 1.00 9000371
, -098 1.00- 9000373 i i ..
TEST CROSS REFERENCE Page 3 RO Exam PWR- Reactor Organi?, ed by Quootion Numbor QUESTION VALUE REFERENCE 099 1.00 9000376 __ __ 100.00 ______ em WB WD _ _ _ 100.00 _
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80PY
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TEST CROSS REFERENCE Page- 4 RO Exam PWR R e a c-t o r Organized by KA G r'o u p PLANT WIDE GENERICS QUESTION VALU3 KA 008 1.00 194001A106 009 1.00 194001A106 010 1.00 194001A107 011 1.00 194001A109 012 1.00 194001A109 013 1.00 194001A111 002 1.00 194001K101 001 1.00 194001K101 003 1.00 194001K103 004 1.00 194001K104 005 1.00 194001K104 006 1.00 194001K109 007 1.00 194001K114 ______ PWG Total 13.00 PLANT SYSTEMS Group I QUESTION VALUE KA 017 1.00 001000G010 014 1.00 001000K103 015 1.00 001000K407 016- 1.00 001000K558 019 1.00 003000A401 020 1.00 003000G006 018 1.00 003000K201 021 1.00 004000A401 022 1.00 004010A205 023 1.00 013000A105 024 1.00 013000A403 026 1.00 015000A303 ! 028 1.00 015000K406
027 1.00 015000K604 , 029 1.00 017020K101 030 1.00 022000A404 031 1.00 025000K301 .fjf; p ,3 l 032 1.00 '059000A411 t q j j, .
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_ ' 033 1.00 059000G010 034 1.00 ~061000A202 gg .
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035 1.00 061000GOG2 037 1.00 072000A202 i8 . , 038 1.00 .072000G007 ______ L i
__ ___ _ _. _ _, , . O TEST CROSS REFERENCE
-
0: Pag I R O= E x a' s PWR - R.o a c t'o r-O r_g-a n i 2 e d by KA G-r:o'u-p PLANT SYSTEMS Group I QUESTION VALUE KA PS-I Total- 23.00 Group Il QUESTION VALUE KA 039 1.50 002000K109 040 1.00 002000K603-041 -1.50 006000K409 043 1.00 010000A403 042 1.00 010000K201 044 1.00 011000A211 045 1.00 012000K501 025 1.00 014000G008 046 1.00 014000G008 047 1.00 016000K101 048 1.00 016000K106 049 1.00 026000G009 060 1.00 029000K302 050 1.00 033000K303 051- 1.00 039000A302-052 1.00 062000G009-036 1.00 063000G010 053 1.00 '064000K411 054 1.00 086000K406 ______-
-PS-II Total 20.00 Group III QUESTION -VALUE KA
<- 055- 1.00 005000K407 056 1.~ 0 0 - 007000K401-057 1.00 008000K102 i: 058- 1.00 008030A304 (- 059 1.00 028000K601 i 061- 1.00 034000A201 062 1.00 -041020K417 hh fSj Q'j" * 3 .
- -063 1.00
____._- -076000G010 LFff9 fy j ~jwilf/ - PS-III~ Total B.00 L?d 'Id "^ 3 < ______ ______ h , TC. Tota .00
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Q TEST' CROSS REFERENCE-h .Pago 6
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L R- O E'x a u PWR Rcactor_ + Organized by KA G r o~u p-EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 065 1.00 000005G005-067 1.00 OC0015G010 066 1.00 000015K207 068 1.00 000026A105 091 1.00 000027A218 " 071 1.00 000040G012 072 1.00 000051A202 074 1.00 000055A203 073 1.00 000055K302 075 1.00 000057A101 077 1.00 000067A203 078 1.00 000067G012 079 1.00 000074K103 080 1.00 000074K304 081 1.00 000074K311 082 1.00 0000766012 ______ EPE-I Total 16.00 Group II QUESTIOW VALUE KA 064 1.00 000001G008 083 1.00 000001G010 084 1.0 A201 - 085 1.00 000007G010 086 1.00 000008K101 087 1.00 - 000009A237 088 1.00 000031A103 089 1.00- - 000011A105-090 1.00 000025G01 ; 070 1.00 000029G011 092 1.00 000029K301 069 1.00 000029K312 093 1.00 000033A202 094 1.00 - 000038K302 095' 1.00 000054G012 096 1.00 000058A101 f 1 076 1.00 000059A102- [j[Q pp {p gjg.p) .yh ca s d
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l ______ . l EPE-II Total 17.00 g - Group III
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i OL TEST CROSS' REFERENCE 01 Pago-. 7 R0 Exam PWR R_e a.c t o O r g a n i--z e d by K- A Grou EMERGENCY PLANT EVOLUTIONS Group III QUESTION VALUE KA 099 1.00 000056A214 097 1.00 000065A201 098 1.00 000065G010 ______ EPE-III Total 3.00 ______ ___ __ EPE Total 36.00 _ ____ ______ _ _ _ _ . , _ Test-Total 100.00
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O O REACTOR COOLANT SYSTEM ( 3/4.4.8 SPECIFIC dCTIVITY LIMITING CONDITION FOR OPERATION _
'l 3.4.8 The specific activity of the reactor coolant shall be limited to: Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and Less than or equal to 100/E microCuries per gram of gross specific )
!
activit APPLICABILITY: H0 DES 1, 2, 3, 4, and ACTION: MODES 1, 2 and 3*: With the specific activity of the -eactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours ' during one continuous time interval or exceeding the limit line shown on Figure 3,4-1, be in-at least HOT STANDBY with T avg less than 500*F within 6 hours; With.the gross specific activity of the reactor coolant greater than ' 100/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T avg less than 500 F within 6 hours; and The prov', fans of Specification 3.0.4 arn not applicable, n
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l:$ , ,[ ( ^With T grea Gr than or equal to 500 avg CATAWdA - UNITS 1 & 2 3/4 4-27 Amendment No.25 (Unit 1) Amendment No.15 (Unit 2)
REACTOR COOLANT SYSTEM O O
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QMITINGCONDITIONFOROPERATION ACTION (Continued) MODES 1, 2, 3, 4, and 5: With the snecific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro- 4
'
curies per gram, perform the sampling and analysis requirements of Item 4,a) of Table 4.4-4 until the specific activity of the rea: tor coolant is restored to within its limit SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity s e reactor coolant shall be determined to be within the limits by performa'a of the sampling and analysis program of
.able 4.4- .
r ( khlIhS{ ' ;+ Copj' C _ CATAWBA - UNITS 1 & 2 3/4 4-28 Amendment No. 25 (Unit 1) Amendment No.15 (Unit 2)
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20 30 40 50 60 70- 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE hQUlvALLin ' i-m . iAL.ua t. JLAh r $PECIFIC ACT1VITY LlHIT VERT"; y PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT. SPECIFIC
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-- ACTIVITY > 1 pCi/grm 005E EQUIVALENT I-131 3 n , ,, 3 -.yn3c-
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$dh$d a b CATAWBA - UNITS 1 AND'2- 3/4 4-29
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TABir 4.4-4 4 9 RfACTOR C001Atli SPICIFIC ACilVITY SAMPlf jj ASD AilALY515 PRUf5AM Gi
" MODES IN WillCil SAMPLE 1YPE Of MLA5tlRfMENT SAMPLE Af10 ANALYSIS AND AftAlYSIS FREQlIENCY AND ANALYSIS REQt11 RED
[
{
m Grost " ioactivity Determination ** At least once per 72 hours 1, 2, 3, 4 Isotopic Analysis for DOSE EQUIVA- 1 per 14 days 1
[ LENI I-131 Concev^. ration Radiochemical for l Determination *** 1 per 6 months * 1
# 1 i 9 R g
4 Isotopic Analysis for Iodine a) Once per 4 hours, 1,2,3,4,5 Including 1-131, 1-133, and 1-135 whenever the specific activity exceeds 1 pCi/ gram DOSE EQUIVALENT I-131
% or 100/E pCi/ gram of j,' gross radioactivity, and a b) One sarpple between 2 1, 2, 3 and 6 hours following a filERMAL POWER change exceeding 15% *
of the RAIED Til[RMAL POWER within a 1-hour perio O
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IC R R
l - _ - _ _ _ _ _ _ _
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O O TABLE 4.4-4 (C'ontinued) TABLE NOTATIONS
#U ntil,the specific activity of the Reactor Coolant System is restored within its limit * Sample to be taken after a minimum of 2 EF00 and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longe "A gros. radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less the.n 10 minutes and all radiciodines. The total specific activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours af ter the sample is taken and extrapolated back to when the sample was take De te r-mination of the contributors to thrs gross specific ictivity shall be based
.upon those energy peaks identifiable with ,a 95% confidence level. The latast available data may be used for pure t' eta-emitting radionuclide *"A radiochemical analysis for 5 shall cnnsist of the quantitative measurement of the specific activi'.y for each radionuclide, except for radionuclides with half-lives 1.ess than 10 minutes and all racioiodines, which is identified in the reactor coolant. The specific activities fgr these individual radio-( nuclides shall be used in the cetermination of.E for the reactor coolant sampl Determination of toe contributors to E shall be based upon tnose energy peaks identifiable with a 95% confidence leve O R ?] n fl j ip '~
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,3/4.1. 'i HOVABLE CONTROL ASSEMBLIEb
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GROUP HEIGHT -
.. u LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be C/ ERA'BLE and positioned within 112 steps (indicated position) of their group step counter demand positio APPLICABILITY: MODES 1* and 2*.
ACTION: Vith one or more full-len;,th roc ' inoperable due to being immovable as a result of exc1ssive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN reo" ire-ment of Specification 3.1.1.1 is satisfied within 1 ho w and be in HOT STANDBY within 6 hours, With more than one full-lengt. -o,1 misaligned from the group step counter demand pnsition by more than 212 steps (indicated position), be in HOT STANDBY within 6 hours, With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by (nora than 112 steps (indicated i-position), POWER OPERATION may continue provided that within 1 hour: ) The rod is restored to OPERABLE status within the above alignment requirements, or The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod ani aligned to within i 12 steps of the inoperable rod while maintainiag the rod stquence and insertion limits of Specifi ntion 3.1. The THERMAL POWER _ l level shall.be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or The rod is declared inoperable and the LHUTDOWN MARGIN require- , ment of Specification 3.1.1.1 is satisfied POWER OPERATION
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may then continue provided that: a) A reevrluation of each accident analysis of Table 3.I n is pet ?NMd within 5 days; this reevaluation shall confirm thar ~ previously analyzed results of these accidents rema'n 2 aid for the duration of operation under these cont. . u t ons ; b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours;
*See Special Test Exceptions Specifications 3.10.2 and 3.1 CATAWBA - UNITS 1 & 2 3/4 1-14 Amendment No. N (Unit 1~)
Amendment No. 68 (Ur.it 2)
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PEACTIVITY CONTROL SYSTEMS ( L1HITING CONDITION FOR OPERATION ACTIOM (Continued) c) A power distribution map is obtained from the movable N incore detectors and F (Z) and F are verified to be A bH within their limits within 72 hours; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours the High Neutron-Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWE With more than one full-length rod trippable but inoperable due to , causes other than addressed by ACTION a above, POWER OPERATION may continue provided that: Within I hour, the remainder of the rods in the bank (s) with the inopeiable rods are aligned to within 112 steps of the inoperable rods while maintaining the rod sequence and inser-tion limits of Specification 3.1. The THERMAL POWER level l
g shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and ( The inoperable rods are restored to OPERABLE status within 72 hour SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be vithin the group damand limi' * y verifying the individual rod positions at least once per 12 hours ex.ept during time interval: when the Rod Position Deviation Monitor inoperable, then verify the group positions ht least once per 4 hour .1.3.1.2 Esch full-1crath rod not fully inserted in the core shall be determined to be OPERABU by movement of at least 10 steps in arf one direction at least once per 01 day m p n O I il d) hhh .N
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, CATAWBA - UNITS 1 & 2 3/4 1-15 Amendment No. 74 (Unit 1) l Amendment No, 68(Unit 2)
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TABLE 3.1-1 ACCID"IT ANALYSES RE001 RING REF. VALUATION # IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD , Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Hisaliep.cnt loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident) Major Secondary Coolant System Pipe Rupture Rupture of a Control Pod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
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sATAWBA - UNITS 1 & 2 3/4 1-16
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3/4.1 REACTIVITY CONTROL SYSTEMS ( 3/4.1.1 BORATION CONTROL
,5jiUT00WN HARGIN - T avg >200'F LIMITING CONDITION FOR OPERATION D.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% ok/k for four loop operatio APPLICABILITY: MODES 1, 2*, 3, and ACTION: ,
With the SHUTDOWN MARGIN less than 1.3% ok/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7900 ppm baron or equivalent until the required SHUTDOWN MARGIN is restore SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% ak/k: . Within 1 hour after detection of an inoperable control red (s) and ( at least once per 12 hours thereafter while the rod (s) is inoperabl If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);
- When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with K,ff less than 1, within 4 hours prior to ,
achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; Prior to initial operation above 5% RATED THERMAL POWiR 4.f ter each fuel loading, by consideration of the factors of Specification 4.1.1.1.le. below, with the control banks at the maximum inservice limit of Specification 3.1.3.6; and mpe j}
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"See Special Test Exceptions Specification 3.1 .W e . ..n
, _ , , - .g CATAWBA - UNITS 1 & 2 3/4 1-1
_ _ _ _ _ _ _ _ _ _ - _ O O REACTIVITY CONTROL SYSTEMS
'I SURVEILLANCE REQUIREMENTS (Continued) ) When in MODE 3 or 4, at least once per 24 hours by consideration of the following factors:
1) Reactor Coolant System boron concentration, 2) Control rod position, 3) Reactor Coolant System average temperature, 4) Feel burnup based on gross thermal energy generation, 5) Xenon concentration, and 6) Samarium concentratio , 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% ok/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those f actors stated in Specification 4.1.1.1.le., abov The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD af ter each fuel loadin , e 7, i
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CATAWBA - UNITS 1&2 3/4 1-2 Amendment No. 39 (Unit 1) Amendment No. 31 (Unit 2)
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CONTAINMENT SYSTEMS 9 5 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS , CONTAINMENT SPRAY SYSTEM ,
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LIMITING CONDITION FOR OPERATION J.6.2 'Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the refueling water storage tank and transferring suction to the containment sum APPLICABILITY: H0 DES 1, 2, 3, and ACTION: With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within
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the next 6. hours; restore the inoperable Spray System to OPERABLE status trithin the next 48 hours or be in COLD SHUTDOWN within the following 30 hour . SURVEILLANCE REQUIREMENTS l
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4.6.2 Each Containment Spray System shall be demonstrated OPERABLE: At least once per power-operated, or 31 days by) automatic verifying in the thatthat flow path each valve is not (manual, locked ' sealed.. or otherwise secured in position, is in its correct position; L By verifying, that un recirculation flow, each pump develops a differential pressure of greater than or equal to 185 psid when tested pursuant to Specification 4.0.5; At least once per 18 sonths during shutdown,"* by: 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Phase "A" Isolation test signal, and 2) Verifying that each spray pump starts automatically on a Phase "B" Isolation test signa pq, 'c 3) Verifying that each spray pump is prevented from starting by Lu ; , ~ ; I y] the Containment Pressure Control System when the cor.tainment atmosphere pressure is less than or equal to 0.25 psid, and is allowed to start at greater than or equal to 0.45 psid f%:nf-hgfI relative to the outside atmosphere, e CCThis surveillance need not be perfortred until prior to entering HOT SHUTDOWN following the Unit 1 first refuelin CATAWBA - UNITS 1 & 2 3/4 6-18 l
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I' , CONTAINMENT $YSTEMS
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B SURVEILLANCE REQUIREMENTS (Continued)
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4) Verifying that each spray pump discharge valve closes or is prevented from opening by the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid and is allowed to open at greater than or equal to 0.45 psid relative to the outside atmosphere, and 1 . 5) Verifying that each spray pump is automatically deenergized by the Containment Pressure Control System when the containment atmosphere pressure is less than or equal to 0.25 psid relative .. to the outside atmospher d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray no:21e is unobstructe .
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CATAWBA - INITS 1 & 2 3/4 6-19
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Form 34012 (8-83)
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PAGE N CNS i AP/1/A/5500/06 LOSS OF S/G FEEDWATER Retype #1r
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TABLE OF CONTENTS l'#10 A. Purpose ,
B. Symptoma 1 Case Loss of CF Supply To S/Gs
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Case II.'Los. Of Mormal CA Supply Case Loss of CF Supply To S/Os Operator Actions 2 g Immediato Actions 2 Subsequent Actions 2 Cas, 1 Losn Of Normal CA Supply Operator Actions 5 Immediate Actions 5
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Subsequent Actions 5 Enclosurn l_- Continuous Monitoring of CA Pump Parameters 12 Encionuro 2 - CA Pump Suction Pressure Based on llotwell 17 Level and Flowrate
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', Form 34912 (8-82) - -
PAGE N CNS
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AP/1/A/5500/06 LOSS OF S/G TEEDWATER 1 of 17 Retype #1C
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A. PURPOEX o To verify proper response to a '.oss of feedwater supply to the S/Gs o To verify proper response to a loss of normal supply of auxiliary feedwate B . pl M. E LO M
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Case I.-Loss of CP Supply To S/Os o CFPT A and B - TRIPPED o S/G A (B,C,D) LO LEVEL ALERT alarm (s) (IAD-4) - LIT o S/G A (B,C,D) TLOW MISMATCH LO CF FLOW alarm (s) (1 AD-4) - LI CASE II. Loss Of Normal CA Supply o "CACST LO LEVEL" alarm (1AD 5, H-4) - LI L
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-, , 4-k Page 1 of 17
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Ferm 34913 (8 83) g g PAGE N , CNS ' - AP/1/A/$$00/06 CASE LOSS OF CF SUPPLY TO S/Gs 2 of 17 Retype #1f n RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE C. QEERATOR ACTIOKE '
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15fLEDL&IE_ ACTIONE
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None
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SHSJQVENT ACT!_QNE Verify Reactor power - LESS Perform the following: _ THAN 5%. _ Manually trip Reacto _ GO TO EP/1/A/5000/01, Reactor Trip or Safety injectio . Verify 1eedwater Isolation Manually initiate Teedwater as indicated by following Isolation statuu lights (1S1-5) - LIT: _o TRN A _o "S/G A CF CONT ISOL _o TRN VLVS CLSD" _o "S/G B CF CONT 150L VLYS CLSD" s _o "S/G C CF CONT ISOL VLVS CLSD" _o "S/G D CF CONT ISOL VLVS CLSD." Verify total CA flow - Perform the following: GREATER THAN 450 GP _ Manually start CA purnp _ II CA flow cannot be established,-IREN G0 10 EP/1/A/5000/2C1, Loss of Secondary IIent Sin g n, 6 a ' ,"" 9e a 1 + th 4
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Form 34913 (8 83) g PAGE N CNS AP/1/A/5500/06 CASE LOSS OF CF SUPPLY TO S/Gs 3 of 17 l Retype //10 l m
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_ RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE I Control S/G 1evels:
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_ Verify S/G N/R levels - _ Maintain total feed flow AT 38%. greater than or equal to' , 450 gp l Initti N/R level increasing in at least one S/G, IILEJ depress CA System
. valyc control " RESET" pushbuttons: ._. o TRN A
_ o TRN , _ Throttle feed flow to maintain S/G N/R levels at 38%. Verify CA supply - ADEQUATE: Ensure the following valves - OPEN: ,
-t
_ o ICA 4 (CA Pmps Suct From UST) _ o ICA-6 (CA Pmps Suct -
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From CA CST).
_ Verify "CACST LO LEVEL" Perform the foilowing: alarm (IAD-5, H-4).- DAR __ 1 ) Ensure 1CA-6 (CA Pmps Suct from CA CST) - CLOSE _ 2) RETER 10 Case II, Loss-of Normal CA Supply and perform applicable steps concurrently with remainder of this cas , 13;di h".u93
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Fe'rm 34913 (8-82) PAGE NO, CNS CASE 1. LOSS OF CF SUPPLY TO S/Gs 4 of 17 AP/1/A/5500/06 Retype //10
'n ACTION / EXPECTED RESPONSg RESPONSE NOT OBTAINED Determine required notifications ._. o BITIB H RP/0/A/5000/01, Classification of Emergency .
_o BLEEB H RP/0/B/5000/13. NRC Notification Requirement __ 7 . Determine and correct cause of loss of CT supply, i __ 0 Determine long term plant statu RETURN E procedure in effec END _
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hi l' b' ;b ud\VY J ni Page 4 of 17 .;
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n -- i Fcrm 34913 (0 82)
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g PAGE .NO,
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CNS
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CASE II. LOSS OF NORMAL CA SUPPLY -$ of 17 I AP/1/A/5500/06 Retype #30 l n'
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RESPONSE NOT OBTAINED l ACTION / EXPECTED RESPONSE
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C. Q RRATOR ACTIONS LM.LEDI ATE ACT!RfiS
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None
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SUBS [qUNT ACTIONS N_01C-Closing 105-69 will isolate the CACST from,the Unit 1 and Unit 2 CA Pump __, 1 . Ensure ICA-6 (CA Peps Suct ,,_, Dispatch operator to unlock and Trom CACST) - CLOSE close 1CS 69 (CACST To Unit 1 and Unit 2 CA Supplies) (SB 619, T-26). Initiate makeup to UST: ___ o Purnp CST to UST with CST Pumps 1A and 1 .,_. o Open 1YH-100 (UST H/U CTRL),
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___ Verify at least one hotwell 99 LO ftep 1 pump is available for mak~eup to the US fA P O ' 'y u[i'i
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;i u['-u$ f .sa-('ti_) A 6 ** #
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h . Page-5 of 17
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*. . Fcrm 34913 (8 B2) g =
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PAGE N CN* CASE !!. LOSS OF NORHAL CA SUPPLY 6 of 17 AP/1/c '5500/06 Rotype #1r
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RESPONSE NOT OBTAINED ACTION / EXPECTED RE$PON$t 4 Makeup to UST from hotwells Dispatch op,.itor to:
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_1) Open ICM-35 (flotwell Ili Lvl Ctrl Byp),
(TB-575, IL-25).
2) Ensure the following
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valves CLOSED: _o ICS-24 (Normal llotwell Makeup Control inlet)
(TB-595, 1G 30) ._, o ICS-6u (Normal llo+.well Hakeup Control Bypass) (TB-596, 1G-30)-
_, o ICS-32 (llotwell Recirc Makeup Valve inlet)
(TB-573, IG-29)
_o 105-35 (flotwell Recirc Makeup Valves Bypass) J (TB-578, IG-29). LE all hotwell pumps are-off. THEE: - _1) Notify Chemistry to ensure all CM polishing domineralizers isolate . 2) Ensure the following valves - CLOSED: _o Polish Domin Byp Ctrl l _o ICH-83 (Gen Load fl g l
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7 7 ;ra - 3
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Reject Bypass). {' _' $.H!
. . . . ah
_ Start and stop hotwell pump (s) .& c g, as required to makeup'from ies'
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' hotwel .. page 6 of 17 l L
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . Form M913 (8 82) h PAGE N CNS AP/1/A/5500/06 CASE I LOSS OF NORMAL CA SUPPLY 7 of 17 Retype #10 m
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ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED , HOE o If the CA pumps are taking a suction on the hotwell with the UST depleted and the CA auto start circuitry is actuated, ' then the CA pump suction will automatically align to the RN assured makeup sourc , o If CA has been reset and CA pump suction is aligned to the hotwell with the U3T depleted, then the CA pumps will trip on low suction pressur ;
. Verify UST level - ADEQUATE:
_o "CACST/UST LO LEVELS" _ o Break coridenser vacuu alarm (I AD-5,11-6) - DAR _, II hotwell level greater than 0.5 f t, ELN REI.U_M LO Step 8 .__. EliEE hotwell level less than 0.5 ft 63D "CACST/UST LO LEVELS" alarm (1AD-5, H 6) lit, IliEE close 1CA-84 (CA Peps Suct Trom UST).
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Lt.C+ +. !? f.tQQ\ VJil k Page 7 of 17 _ w n -w-, -w-w--- --v,-,-r-w,w-y ---wr~p- w -~m- vv-* -em-- g--+n---m_ w-m-- ,-e w
r Form 34913 (8 821 O e PAGE N CNS CASE II. LOSS OP NORMAL CA SUPPLY .8 of 17 AP/1/A/5500/06 Retype #1r
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RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE Ensure CA pumps suction aligned to RN: . o TRN A valves - OPEN: . __ o 1RN-250A (RN lidt A To CA Pmp Suct isol) o 1CA-116A (CA Pump #1
- Suct Frm RN ildr A)
_._, o ICA-15A (CA Pump 1A Suct Frm RN Isol). ! o TRN D valves - OPEN: __., o ICA-85B (CA Pump #1 Suct Frm RN lldr B) ___ o ICA-16B (CA Pump 13 Suct Frca RN Isol) _o 1RN-310B (RN Hdr B To CA Pmp Suct Isol). IT "SST CA XTER TO RC" alarm (1AD-5, G-4) lit, ljiEE ensure CA Pump #1 suetion aligned to RC: _ o ICA-178 (RC Supply To CA Pumps Isol) - OPEN _ o ICA-174 (RC Tr> CA suct Isol) - OPEN o ICA-175 (RC To CA Suct Isol) - OPE t' " TS)
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Determine long term plant , b> > - 1 jw status ,; .,p .g
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o Continue monitoring CA b[/hO pump status, _A RQ1LRE 10 procedure in ef f c Page 8 of'17 t
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Form 34913 (8 82) - g
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PAGE N CASE I LOSS OF NOR!iAL CA SUPPLY 9 of 17 AP/1/A/5500/06 Retype (110 m
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" RESPONSE NOT ODTAINED ACTION / EXPECTED RESPONSE H_Olt; The CA pumps will trip on low suction pressure with CA reset and CA pump suction aligned to the hotwell when the UST is deplete . Establish operator control of CA Systemt
_ LE CA Pump #1 is needed,
' TREN place CA Pump (11 switch in "0N". Depress CA System valve
' control " RESET" pushbuttons: _o TRN A _o TEN _
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Form 34913 (B 82) g g PAGE N CNS CASE 1 LOSS OF NORMAL CA SUPPLY 10 og 17 AP/1/A/5500/06 R e t ype (11' _ _ _
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RESPONSE NOT OBTAINED ACTION / EXPECTED RESPONSE _ 1 Maintain all S/G H/R levels at 38Y. by using one CA pump at a times o LE one motor driven CA pump available, I}{EN dispatch operator to open the following valves:
[o ICA-111 (CA Pump 1A 6 IB Disch Xover To S/G 1 sol) (AB-552, BB-50, Rm 250)
_o ICA-112 (CA P ;mp 1A & IB Disch Xover To S/6 Isol) (AB-552, BD-50, Rm 250).
OR o LE CA Pump (11 available, IHEN open the following valves: _o ICA-66B (CA Pump 1 Disch To S/G 1A Isol)
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_o ICA-38A (CA Pump 1 Disch To S/G ID isol).
MQlE The CA pump trip on low suction pressure is being defeated in the following step so that the hotwnll can be used as a CA suction source 13 Initiate SWR-10403 to have
{7}--} p n (W'
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. *'nf' 71)
g : $3 IAE defeat the CA pump trip j ,U ' 6>
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on low suction pressure on all 3 CA pump $f M7 i ( [w a[ib Pan. 10 of 17 i . . - _ _ _ _ _ _ _ _ _ _ _ - - _ _ -_ _ _ _ _ _ __ __- _-
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Fo rn 34911(8 82> 0 e A N J CNS AP/1/A/5500/06 CASE I LOSS OF NORMAL CA SUPPLY 11 of 17 Retype #10 t- _ ACTION / EXPECTED RESPONSE HESPONSE NOT OBTAINED 1 44 . Defeat ther automatic seap of CA pump suction to the RC assured makeup source by dispatching operator to ' open the following breakers at the SSF (Elev. 611):
,_, a ) SDSP1 Bkr #4
___ b ) SDSD-F01D (RC Supply To Aux Feodwater Isol Valve ICA-174) ___ c ) SDSD-r02C (RC Supply To Aux Feedwater Isol Valvo ICA-175).
_, 15 . Continuously monitor the parameters given on Enclosure 1 and perform any required action _ 1 Determine long term plant
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statu RETUR!{ L0 procedure in effec E.!!D _
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j_ rage 11 of 17 {
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Form 34913 (8 92)
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PAGE N LOSS Of S/G FEEDWATER CNS ENCLOSURE 1 - Page 1 of 5 12 of 17
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AP/1/A/5500/06 Retype #. 1
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- FIESPONSE NOT OBTAINED >
ACTION / EXPECTED RESPONSE _, EDX7]E1[0US MQN} IPA)14LQLQA PVMP PARAMETEEH HQTU If the CA pumps are taking a suction on the hotwell with the UST depleted and the CA auto start circuitry is actuated, then the CA pump suction will automatically align to the RN assured makeup sourc HRIK The preferred method of determining CA pump suction pressure is to compare hotwell level and CA pump flowrote (Enclosure 2), Continuously monitor CA pump suction pressure using any of the following methods: __ o llotwell levni and CA pump flowrate BETJB IQ Enclosure 2 0B y
.__ o Pressure gauges on lHC-4 - 9B
_,,. o Local pressure gnuges in the CA pump roo . II CA pump suction pressure drops to less than or equal to D.2 psig, THIM perform the following ___ a . Sto;) all CA pump b.- Closo ICA-4 (CA Pmps _ b. Dispatch operator to unlock and close
._.,
Suct From UST). Supply From UST) -((f!Pfi20,ICS-19(CAP 10-30). ,
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Fo64 M*13 (8 82) g ^
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cI PAGE NO' i CNS LOSS OF S/G TEEDWATER [ ENCLOSURE 1 - Page 2 of 5 13 of 17 AP/1/A/5500/06 Retype #10 t a : q - RESPONSE NOT OBTAINED
' - * * ACTION / EXPECTED RESPONSE - -
t I te (Continued) ,
! [ Ed: available, Ilgy align -
CA pump suttlon to the RN Osured makeup course =. O Align A TRN RN supply by opening the following valves:
.
__ o 1RN-250A (RN lidt A To CA Pep Suct 1 sol) _o ICA-116A (CA Pump #1 Suct Frm RN Hdr A)
._ o ICA-15A (CA Pump 1A Suct Frm RN 1 sol).
2) Align D TRN RN supply by opening the following valves:
,o ICA-85B (CA Pump #1 Suct Frm RN lidr 10 ,_ o ICA-18B (CA Pump 1B ,,,
Suct Frm RN 1so1' _o 1RN-310B (RN lidr B To CA Pmp Suct 18o1).
, qqrjp pq' 1: n
,, W e 3 n s.. jOQ
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Npis USI' . Yg Pagn 13 of 17
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F tm 34913 (8 BM PAGE N ll LOSS OF S/G FEEDWATER CNS - ENct.oSURE 1 - Pagn 3 of 5 14 of_17 ll AP/1/A/5500/06 Retype (1) ' n 1 ,
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RESPONSF NOT OBTAINED ACTION /2XPECTED RESPONS5 (Continued) Restorn the automatic swap of CA pump suction to the RC assured makeup j
, source by dispatching operator to close the fo'21cwing breakers at , the SL<' (Elwv. 611): ._ 1) SDSP1 Bkr //4
_ 2) SDSD-F01D (RC Supply To Aux Feedwater isol Valve ICA-174) _ 3) SDSD-F02C (RC Supply To Aux Feedwater Isol Valve ICA-17 __ Start CA pump (s) as necessary to maintain S/G N/R levels at 3 8* . __ ?. ill{EN tive "CACST/UST Lo 3 O ; , ,. y LEVELS" alarm (iAD 5, H 6) , is lit. IljEN break -?,f,3 .
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's C 5 C Corulenser vacuut idi i i page 14 of 17 (1
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Fcem 34913 (S 82) AGE N CNS LOSS OF S/G TEEDWATER ENCLOSURE 1 - Psgo 4 of 5 15 of 17 AP/1/A/5500/06 Retype #10
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RESPONSE NOT OBTAINED ACTION /E XPECTED RESPONSE I EHEM hotwell level is less than or equal to 0.5 ft, IHEN perform the following: . __ Stop all CA pump __ Close ICA-4 (CA Pmps Suct __ Dispatch operator to unlock From UST). and close 1C5-19 (CA Pumps-Supply From UST) (TD-620,
- 1D- 30 ) '. II RN available TitEE align CA pump suction to the RN assured makeup sources:
1) Align A TRN RN supply by opening the following valves: __ o 1RN-7.50A (RN lidr A To CA Pmp Suct Isol) __ ICA-116A (CA " ump di Suct Frm RN lldr A) __ o ICA-15A (CA Pump 1A Suct Frm RN ! sol). r
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2) Align B TRN RN supply by opening the following valves: __ o ICA-85D (fA Pump #1 Suct Frm RN lidr D) P l? l1 ,, . ,
'l A, -J _A-
__ o ICA-18D (CA Pump 1B 'f : s ,. Suct Frm RN lsol) * L'J 0 2[d , b. #j bn it J ,, __ o 1RN-310D (RN lidr B To CA Pmp Suct isol). U O80'i* _ l Page 15 of 17 t
r-- e s , Form 34913 (8-021 g -
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LOSS OF S/0 FEEDWATER CNS EtC.0SURE 1 - Par;o 5 of 5 16 of 17 AP/1/A/5500/06 Rotype //)
,
h , RESPONSE NOT ODTAINED ACTION / EXPECTED RESPONSE - 84 . (Continued) Restore the automatic - swap of CA pump suctfon to the RC assured makeup source by dir. patching operator to close the following breakers at
.
the SSF (Elev. 611): __ 1) SDSP1 fikr //4
._. 2 ) SDSD-Fold (RC Supply To Aux Feedwater Isol Valve ICA-174)
_ 3) SDSD-F02C (RC Supply To Aux Feedwater Isol Valvo ICA-17 .__ Start CA pump (s) as necessary to mnintain S/G N/R levels at 38%. 3g nn
d4 .;: 2 ..$1 COPY
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PAGE N CliS LOSS Or S/G TEEDWATER . ENCLOSURE 2 Fage 1 of' 1 11 of 17 AP/1/A/5500/06 Rutype //10
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G.A. PUl'IP_S.lLCIL93 fRESSURE .BM@_9N HOTWELL LE. lek ANIQ fkOWRATE -
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12.00- - - w cr!- .. . .,
; + i -+i+ + ; 4 4; v i+ -
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+ + -i , -
11.75- +-r+- - - -
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+- r r- - * +
^o 11 50-
-
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Page 17 of 17 h y --
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