W3F1-2006-0065, Initial Exam - 11/2006 - Licensee Post Exam Comments
| ML063530149 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 11/22/2006 |
| From: | Fletcher R Entergy Corp, Entergy Nuclear South |
| To: | Nease R NRC Region 4 |
| References | |
| W3F1-2006-0065 | |
| Download: ML063530149 (42) | |
Text
Entergy Nuclear South Entergy Operations, Inc.
17265 River Road Killona, LA 70057 W3F1-2006-0065 November 22, 2006 Rebecca L. Nease U.S. Nuclear Regulatory Commission Region IV 61 1 Ryan Plaza Drive, Suite 400 Arlington, TX 7601 1-8064
Subject:
Post Operator Licensing Examination Documentation and Comments Submittal Waterford Steam Electric Station, Unit 3 waterford 3)
Docket No. 50-382 License No. NPF-38
Dear Ms. Nease:
Per guidance provided in NUREG-1021 Revision 9, ES-501.C.1.a, Entergy is hereby forwarding the specified post operator licensing examination documentation. The subject operator licensing examination was conducted at Waterford 3 SES from November 10 through 16, 2006.
Also, per guidance provided in NUREG-I 021 Revision 9, ES-402, Section E, Entergy is hereby submitting formal comments associated with the completed Examination.
There are no new commitments contained in this submittal. If you have any questions concerning this submittal, please contact Robert Fletcher at (504) 739-6001.
R.W. Fletcher Manager, Training &
RWF/OPP/cbh Attachments: 1. Listing of Enclosed Operator Licensing Examination Documentation
- 2. Formal Examination Comments
Enclosure:
Operator Licensing Examination Documentation
W3F1-2006-0065 Page 2 (w/o Attachments) cc: U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. Me1 B. Fields Mail Stop 0-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004 U.S. Nuclear Regulatory Commission Region IV ATTN: Thomas F. Stetka 611 Ryan Plaza Drive, Suite 400 Arlington, TX 7601 1-8064
Attachment 1 to W3F1-2006-0065 Listing of Enclosed Operator Licensing Examination Documentation to W3F1-2006-0065 Page 1 of 1 Post Operator Licensing Exam Documentation Listing The following examination documentation is enclosed as applicable as specified in NUREG 1021, Revision 9, ES-501 Section C . l .a:
- 1. The graded written examinations (Le., each applicants original answer and examination cover sheets) plus a clean copy of each applicants answer sheet (ES-403, grading initial site-specific written examinations).
- 2. The master examination(s) and answer key(s), annotated to indicate any changes made while administering and grading the examination(s) (ES-402, administering initial written examinations, and ES-403).
- 3. Any questions asked by and answers given to the applicants during the written examination (ES-402).
- 4. The written examination seating chart (ES-402).
- 5. A completed form ES-403-1, written examination grading quality checklist (ES-403 and section d.1).
- 6. The results of any written examination performance analysis that was performed, with recommended substantive changes (ES-403).
- 7. Original form(s) ES-201-3, examination security agreement, with a pre- and post-examination signature by every individual who had detailed knowledge of any part of the operating tests or written examination before they were administered.
The following are additional items enclosed as requested by the Chief Examiner:
- 1. CD containing as given written and operating exam.
- 2. New ES-401-6 sheet with numbers incorporating prep week comment resolution.
- 3. Expanded ES -301-5 with individual count for each combination.
- 4. New clean ES -301-6 competencies checklist.
- 5. CD containing Simulator scenario trend data.
Attachment 2 to W3F1-2006-0065 Formal Examination Comments
Waterford 3 is requesting the NRC to review 5 questions that were included in the 2006 Reactor Operator examination.
I Q I uestion I Recommended Action 5 Accept C&D 2 Delete from exam 3 Delete from exam 1 17 64 Delete from exam Delete from exam The attachments contain the specificjustification for each question.
The number of questions recommended for deletion and changed answers is greater than the 4 questions specified in ES-501 C.2.c., second paragraph. Condition report WF3-2006-3510 has been generated to evaluate this condition.
1
Waterford 3 2006 RO Exam Question 5 0 Mode 1, 100% power 0 RCP 2B THRUST BRNG TEMPERATURE HI annunciator is in alarm RCP 2 8 Motor Lube Oil Reservoir temperature is 185°F and stable 0 RCP 28 Upper Thrust Rearing Temperature indicates 206°F and stab!^
What ONE action is required by OP-901-130, Reactor Coolant Pump Malfunction, as a result of the PRESENT indications?
1A Trip the reactor and secure RCP 28.
1B I Commence a plant shutdown.
C Start ACCW Pumps and WCT Fans.
D Start one RCP 2B lift oil DurnD.
Waterford 3 recommends accepting two answers on this question - Answer C, which is the original correct answer per the key, and also Answer D for the following reason: A procedure revision to Off-Normai Procedure, OP-90 i - i 30, Reactor Coolant Pump Malfunction, completed prior to the exam, added additional guidance that allows starting a Reactor Coolant Pump lift oil pump as well. This choice was listed in answer D.
Answer C and D are both correct. 5 of 5 Reactor Operator Candidates missed this question.
1
Waterford 3 2006 RO Exam Question 5 I
Question Information I
1 I
Examination Outline Cross-Reference I
Tech References OP-901-130 KIA 4.2-Aq 517 7-AA2.02
/I Ref Supplied , NIP. Imp. Rating I 2.8 I I Learning Objective WLP-OPS-PPO-10 04 I 1 Proposed Question The following plant conditions exist:
- Mode 1, ?OO% pcwer RCP 2B THRUST BRNG TEMPERATURE HI annunciator is in alarm 0 RCP 2B Motor Lube Oil Reservoir temperature is 185°F and stable RCP 2 8 Upper Thrust Bearing Temperature indicates 206°F and stable What ONE action is required by OP-901-130, Reactor Coolant Pump Malfunction, as a result of the PRESENT indications?
A Trip the reactor and secure RCP 2B.
B Commence a plant shutdown.
C Start ACCW Pumps and WCT Fans.
D Start one RCP 28 lift oil w m p .
Answer 1C Explanation C is correct; when bearing temperatures exceed 205"F, Duty Plant Manager and System Engineer are notified and attempt to lower bearing temperature(s) by EITHER of the following:
0 Start Dry Cooling Tower Fans.
0 Start Auxiliary Component Cooling Water Pump(sj AND associated Wet Cooling Tower Fans.
A is incorrect; Reactor and RCP are required to be tripped when bearing temperatures exceed 225°F.
B is incorrect; if any bearing temperature exceeds 212"F, then commence Rapid Plant power reduction D is incorrect: Oil Lift pump is required to be started only if the ARD has a high temperature condition.
Old question matched WA AA2.08 for RCP trip conditions on high bearing temperature.
1 RCP 2 8 THRUST BRNG TEMPERATURE HI annunciator F-9 setpoint is 203°F for upper bearing, per
)I 1 ARP OP-500-008 Rev 17.
I 2
' SAFETY RELATED !
REQUESTiAPPROVAL PAGE Required Review Level (check one)
! OSRC PROCEDURE NUMBER: OP-901-130 REVISION: 4 CHANGE: 0 DEVIATION: NIA I
CROSS NIA NIA DISCIPLINE REV1EWS NIA NJA (List Groups)
NIA NJA NIA NIA GM, PLANT OPERATIONS Review VICE PRESIDENT, OPERATIONS Approval (sign) / NIA Effective Date / Milestone (if applicable): N/A
Off Normal Procedure OP-90 1-130 Reactor Coolant Pump Malfunction Revision 4 E3 BEARING TEMPERATURE HIGH PLACEKEEPER START DONE N/A
- 1. --
IF ANY of the following occurs, THEN notify Duty Plant Manager and System Engineer:
0 -
ANY bearing temperature exceeds 205°F e Reactor Coolant Pump Lube Oil Coder digerentlal -
L!
temperature reaches 30°F A change in ANY bearing temperature of >IO"F over i hour CAUTION (I)CCW TEMPERATURES OF ~ 7 5 F" COULD LEAD TO ESSENTIAL CHILLER TRIPS ,
ON EVAPORATOR LOW REFRIGERANT PRESSURE.
(2) CCW TEMPERATURE SHOULD BE CHANGED AT A RATE OF ~ 1 0 ° F IN ONE HOUR TO PREVENT DEGREDATION OF THE REACTOR COOLANT PUMP SEALS.
- 2. Attempt to lower bearing temperature(s) by tol,v&ing:
c I+.>
ANY of the 0 Start Dry Cooling Tower Fans.
0 c l n Start Auxiliary Component Cooling Water Pump(s) 0 AND associated Wet Cooling Tower Fans.
0 0 Start Auxiliary Component Cooling Water Pump(s)
AND lower ACC-l26A(B) setpoint.
n u e Start an oil lift pump cn affected RCP fer 5-70 minutes.
ilci 14
Waterford 3 2006 RO Exam Question 2 I I B SDM calculation is NOT; because adequate boration is assumed due to the SIAS.
1I C I depressurization is; to provide more Safety Injection fiow to the core.
depressurization is NOT; because break fiow is assumed to reduce pressure enough to permit SDC entrv.
Vvaterforb 3 reconimeiicls deleting this question for the following reason: The required knowledge is considered too detailed and infrequently used to be performed from memory. 4 of 5 Reactor Operator Candidates missed this question.
The referenced objective states: Given plant andor equipment conditions, INTERPRET content, location, and sequencing of procedural steps applicable to the LOCA procedure, OP-902-002. [18]
The correct answer per the exam key was B. The question author was referring to step 19 of OP-902-002, which directs performing a controlled plant cooldown to less than 350F. Step 19 is not preceded by a step to perform a Shutdown Margin Calculation. The Technical guide for step 19 states the following:
A plant cooldown will change the reactivity conditions in the reactor core. Therefore, whenever u cooldown is pegormed, the operator must consider the effects of the cooldown on reactivity control and take appropriate uctions to muintuin the shutdown margin. This may include boruting and sampling the RCS, both-prior to and during the cooldown. This step assumes einergency boration is in progress due to an SIAS and therefore no direction is given to commence boration.
Step 68 of OP-902-002 also directs performing a controlled plant cooldown to less than 35OF. The preceding step. 67, DOES procedurally direct performing a Shutdown Margin Calculation. The question should have had OP-902-002 supplied as a reference to allow the students to interpret the procedural steps, as required by the objective. It should be noted that all students missing this question chose answer A.
Waterford 3 2006 RO Exam Question 2 ProDosed Question Plant conditions are as follows:
A pressurizer safety valve has failed partially open.
RCS pressure is 1150 psia and slowly dropping.
All systems have operated per design.
0 The crew has entered OP-902-002, Loss of Coolant Accident.
The procedure directs performing a controlled cooldown to < 350°F using steam bypass or atmospheric dumps valves.
In this situation, a procedurally directed . . ,
Explanation B is correct; SDM calculation is NOT required by OP-902-002. The Tech Guide states that the effects of the cooldown on core reactivity should be considered, but adequate boration is assumed due to the SIAS.
A is incorrect; S D M calculation is NOT required by OP-902-002.
C and D are incorrect; the Tech Guide states that depressurization may be necessary to permit SDC entry.
Comments 2
- Number: WLP-OPS-PPE02 ENTERGY NUCLEAR *Revision: O6 Page: 5 of 9 Rkype: T1.03 LESSON PLAN Given plant and/or equipment conditions, EVALUATE Safety Function Status and DETERM!NE actior! tc! !E taker! if Safety Function Status Check!is? Criteria are not being met as applicable to the LOCA Recovery Procedure, OP-902-002. [I61 Given plant and/or equipment conditions, IDENTIFY and APPLY procedural steps, cautions, and notes applicable to the LOCA procedure or to plant andlor equipment conditions which may mtry into the LOCA procedure, OP-902-002. [ 171 Given plant and/or equipment conditions, INTERPRET content, location, and sequencing of procedural steps applicable to the LOCA procedure, OP-902-002.
S81 STATE the criteria required AND the basis for each of the following operations in the LOCA procedure, OP-902-002: [ 191 Verify SlAS Actuation RCP Trip Strategy Verify RCP Operating Limits Verify Containment Isolation and Cooling Verify Containment Spray Actuation HPSI Throttle Criteria LPSl Pump Stop Criteria Check Single Phase Natural Circulation Verify Heat Removal Under Two Phase Natural Circulation RCP Restart Criteria RAS Initiation Criteria Containment Spray Termination Hot and Cold Leg Injection SDC Entry Conditions
WATERFORD 3 SES OP-902-002 Revision 11 Page 17 of 65 LOSS OF COOLANT ACCIDENT RECOVERY I
INSTRUCTIONS CONTINGENCY ACTIONS Restore Instrument Air
- 16. Verlfv instrument air is available:
- a. Check BOTH of the following are a. 1 REFER TO Appendix 18, operating: "Aligning Potable Water to Instrument Air Compressors" and TCWpump aiian potable water to the CWpump instrument air compressors.
- b. Check instrument air pressure is b. 1 Dispatch an operator to start ALL greater than 95 psig. available air compressors.
- c. Check IA 909, CNTMT
- c. 1 IF instrument air pressure is ISOLATION INSTRUMENT AIR greater than 95 psig, THEN open valve is open.
IA 909, CNTMT ISOLATION INSTRUMENT AIR valve.
Restore Power to DCT Sump Pumps
- 17. IF offsite power has been lost AND can NOT be restored within 30 minutes, THEN REFER TO Appendix 20, "Energize DCT Sump Pumps" and energize at least one DCT sump pump in each sump.
LOCA Condition Isolated
- 18. IF the LOCA is isolated, THEN GO TO step 53.
WATERFORD 3 SES OP-902-002 Revision 11 LCSS CF COOLANT ACCiDEI\!T RECOVERY I Page 78 of 65 I
INSTRUCTIONS CONTINGENCY ACTIONS The following forms may be required during the cooldown and depressurization:
Appendix 3-8, ;'Pressurizer/RCS Cooldown Log" Appendix 3-C, "PZR Spray Transient Logsheet" OP-010-004 Attachment 9.9, "Design Cycle Transient Logsheet" Perform Controiied Cooldown
- 19, Cooldown the RCS to less than 35OOF 19. Cooldown the RCS to less than 35OOF TH or CET temperature using the TH or CET temperature using the steam bypass control valves. atmospheric dump valves.
I I
I NSTR UCTIO NS CONTINGENCY ACTIONS Cooldown NOT Desired
- 66. IF a plant cooldown is NOT desired.
THEN:
- a. Maintain the plant in a stabilized condition.
- b. WHEN BOTH of the following conditions are met, At least one RCP is qerating MFW is available to the steam generators THEN GO TO the appropriate General Operating Procedure.
Maintain Shutdown Margin During the Cooldown
- 67. REFER TO OP-903-090, "Shutdown Margin" and perform the following:
- a. Calculate Shutdown Margin for desired RCS temperature.
- b. Borate the RCS to maintain shutdown margin throughout the cooldown.
WATERFORD 3 SES I OP-902-002 Revision 11 INSTRUCTIONS CONTINGENCY ACTIONS The following forms may be required during the cooldown and depressurization:
Appendix 3-A, "Pressurizer/RCS Cooldown Log" Appendix 3-C, "PZR Spray Transient Logsheet" OP-010-004Attachment 9.9, "Design Cycle Transient Logsheet" Perform Cont roIIed CooId own
- 68. Cooldown the RCS to less than 350°F 68.1 Cooldown the RCS to less than 350°F TH or CET temperature using the TH or CET temperature using the steam bypass control valves. atmospheric dump valves.
f T e c h n i i G i i d e for Loss of TG-OP-902-002 I Cooiant Accident Recovery Procedure Revision I G Step Number I 9 Perform Controlled Cooldown Obiective The intent of this step is to initiate a controlled cooldown of the plant. Tnis action will reduce the pressure and temperature of the plant. In the case of a LOCA, SGTR, or ESD this limits the associated break flow by reducing the stored energy in the plant.
A plant cooldown will change the reactivity conditions in the reactor core. Therefore, whenever a cooldown is performed the operators must consider the effects of the cooldown on reactivity control and take appropriate actions to maintain the shutdown margin. This may include borating and sampling the RCS, both prior to and during the cooldown. This step assumes that emergency boration is in progress due to an SlAS and therefore no direction is given to commence boration.
Inst ructio ns The preferred method to perform the plant cooldown if the main condenser is available, is to use the steam bypass control system. Use of the steam bypass valves conserves secondary inventory and provides for greater control over possible releases from the steam generators to the atmosphere.
Conti nqencv Actions If systems are not available to support using the steam bypass valves, such as a loss of offsite power (LOOP) atmospheric dump valves (ADVs) are used as an alternate method to cooldown the plant.
Justification for Deviations There are no deviations.
References None 42
I Technical Guide for Loss of TG-OP-902-002 i
Step Number 67 Maintain Shutdown Margin During the Cooldown Obiective The intent of this step is to maintain the reactor shutdown throughout the cooldown Instructions The operator should perfarm a shutdown margin calculatior; and borate the RCS to maintain shutdown margin to shutdown cooling entry conditions.
Conti nqencv Actions None Justification for Deviations There are no deviations.
References None 115
Coolant Accident Recovery Procedure 1 Fievision IO Step Number 68 Perform Controlled Cooldown 0b iective The intent of this step is to initiate a controlled cooldown of the plant. This action will reduce the pressure and temperature of the plant. In the case of a LOCA, SGTR, or ESD this limits the associated break flow by reducing the stored energy in the piant.
k piant cooldown will change the reactivity conditions in the reactor core. Therefore, whenever a cooldown is performed the operators must consider the effects of the cooldown on reactivity control and take appropriate actions to maintain the shutdown margin. This may include borating and sampling the RCS, both prior to and during the cooldown.
Instructions The preferred method to perform the plant cooldown if the main condenser is available, is tc use the steam bypass control system. Use of the steam bypass valves canserdes secondary inventory and provides for greater control over possible releases from the steam generators to the atmosphere.
Cont inqency Actions If systems are not available to support using the steam bypass valves, such as a loss of offsite power (LOOP) atmospheric dump valves (P,DVs) are used as an alternate methcd to cooldown the plant.
Justification for Deviations There are no deviations.
References None 117
Waterford 3 2006 RO Exam Question 3 7- ~~
~ The 90°F temperature limit for the Containment applies
.A. in Modes 1-4; Clad B in Modes 1-4; Containment C above 70% power; clad 1 D above 70% power; Containment Waterford 3 recommends deleting this question for the following reason: The question tested at a meinorization level that was not required by the referenced objective. 4 of 5 Reactor Operator Candidates missed this question.
The referenced objective states:
Given the Technical Specifications: Technical Requirements Manual. and plant status, IDENTIFY and EVALUATE the operability, actions, and bases for the following: (6) 3.3 2 ESFAS Instrumentation 3.6.1.1 Containment Integrity 3.6.1.2 Containment Leakage 3.6.1.3 Containment Air Locks 7614 Internal Presslire 3.6.1.5 Containment Air Temperature (high) 3.6.1.5 (TRM) Containment Air Temperature (low) 3.6.1.6 Containment Vessel Structural Integrity 3.6.3 Containment Isolation Valves 3.6.5 Vacuum Relief Valves 3.8.4 Electrical System Protective Devices 3.9.4 Containment Building Penetrations ( Refueling)
No reference was supplied for this question thereby requiring the students to have memorized Technical Specification 3.6.1.5 and it basis.
1
Waterford 3 2006 RO Exam Question 3 I
QID New Modified Direct from Bank 3 R06003 X I Question Information Examination Outline Cross-Reference Previous Bank QID N/A Level RO 1
NRC Exam History 1 N/A TierlGroup 1/1 Tech References TS and Basis 3.6.1.5 KIA 4.1-E9-EK3.16 iI I I II Ref Supplied N/A Imp. Rating 3.8 I
Cognitive Level 2 I O CFR 55.41(b) 5 Learning Objective WLP-OPS-CBOO 06 I
, ProDosed Question The 90°F temperature limit for the Containment applies 1 This limit ensures LOCA.
temperature is maintained within design specifications during a
~ A in Modes 1-4; clad B in Modes 1-4; Containment C above 70% power; clad 1 1D I above 70% power: Containment Answer C Explanation C is correct; TS 3.6.1.5 limits the low temperature to 90°F only above 70% power, and the basis states this is to ensure peak clad temperature remains 5 2200°F.
A and B are incorrect; TS 3.6.1.5 limits the low temperature to 90°F only above 70% power.
5 is incorrect; TS 3.6.1.5 basis states this is to ensure peak clad temperature remains 5 2200°F.
Comments This limit applies to a SB LOCA because it is based on the difference in the rate ECCS is injecting versus the rate at which the RCS is emptying into the Containment. At lower Containment pressures, this
WLP-OPS-CBOO Given the Technical Specifications, Technical Requirements Manual, and plant status, IDENTIFY and EVALUATE the operability, actions, and bases for the following: (6) a 3.3.2 ESFAS Instrumenta ti on a 3.6.1.1 Containment Integrity a 3.6.1.2 Containment Leakage a 3.6.1.3 Containment Air Locks a 3.6.1.4 Internal Pressure a 3.6.1.5 Containment Air Temperature (high) a 3.6.1.5 (TRM) Containment Air Temperature (low) a 3.6.1.6 Containment Vessel Structural Integrity c 3.6.3 Containment Isofatior; Valves a 3.6.5 Vacuum Relief Valves a 3.8.4 Electrical System Protective Devices a 3.9.4 Containment Building Penetrations ( Refueling)
From memory, IDENTIFY and EVALUATE the operability, actions, and bases for the following Technical specifications that require actions in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or less: (7) 3.3.2 ESFAS Instrumentation a 3.6.1 .I Containment Integrity 3.6.1.2 Containment Leakage 3.6.1.4 Internal Pressure a 3.9.4 Containment Building Penetrations ( Refueling) p 1.
J w e n piant status, EXPLAIN the bases for procedc!raI steps and requirements, including notes and cautions for the following procedures: (8)
Of-008-005 Revision 8 Page 6 of 25
CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall 120°F
" I I &
, I APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With the containment average air temperature the average air temperature to within the limit- within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetlcai average of the tefiperaitires at any three of the foiiowing iocaiions and shall be determined at least once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
Location
- a. Containment Fan Cooler No. 1A Air Intake
- b. Containment Fan Cooler No. 1B Air Intake
- c. Containment Fan Cooler No. 1C ,Air !ntake
- d. Containment Fan Cooler No. I D Air Intake WATERFORD - UNIT 3 3/4 6-13 - . . .
CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the a n ~ u l u satmosphere of 0.65 psid, (2) the containment peak pressure does not exceed the design pressure of 44 psig during either LOCA or steam line break conditions, and (3)the minimum pressure of the ECCS performance analysis (BTP CSB 61) is satisfied.
The limit of +27 inches water (approximately 1.O psig) for initial positive containment pressure is consistent with the limiting containment pressure and temperature response analyses inputs arid assumptions.
The limit of 14.275 psia for initial negative containment pressure ensures that the minimum containment pressure is consistent with the ECCS performance analysis ensuring core reflood under LOCA conditions, thus ensuring peak cladding temperature and cladding oxidation remain within limits. The 14.275 psia limit also ensures the containment pressure will not exceed the containment design negative pressure differential with respect to the annulus atmosphere in the event of an inadvertent actuation of the containment spray system.
The limit of 120.F on high average containment temperature is consistent with the limiting containment pressure and temperature response analyses inputs and assumptions.
The limits currently adopted by Waterford 3 are 269.30 F during LOCA conditions and 41 3.5- F during MSLB conditions.
120. F maximum value, specified in the TS 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintainedcomparabie io the originai aesign standards for ine iife of ihe faciiity. Structurai integrity is required to ensure that the containment vessel will withstand the maximum pressure resulting from the design basis LOCA and main steam line break accident. A visual inspection in conjunction with Type A leakage test is sufficient to demonstrate this capability.
AivlENDMENT NO. s, WATERFORD - UNIT 3 B 314 6-2 CHANGE NO. w,
Waterford 3 2006 RO Exam Question 7
! I 1 Which of the following components may be operated from CP-1 following a loss of the TGB-DC Bus? 1 A Steam Bypass Control Valves 1
B 2A(B) to 3A( B) Tie Breakers C Condensate Pumps 1D Main Feedwater Pumps Waterford 3 reccmrnesds delering this question for the fol!=wing reason: The required knowledge is considered too detailed and infrequently used to be expected to be performed from memory. 4 of 5 Reactor Operator Candidates missed this question.
The referenced objective requires interpreting procedural steps, not memorizing them.
To successfully answer the question, the student would have to memorize the automatic actions of Off-Normal Procedure, OP-901-3 13, Loss of 125 VDC Bus, and the fourth item in a Caution on page 30 of Off-Nornial Procedure, OP-90 1-313, Loss of 125 VDC Bus.
The objectives from WLP-OPS-FWPOO, Feedwater Pumps, and WLP-OPS-DCOO, 125V DC Distribution were reviewed. No objective was found that would require a student to recall from memory the effects of a loss of DC power on the Main Feedwater Pumps.
1
Waterford 3 2006 RO Exam Question 17 jm 1 QiB I 1 New I
Modified I
Direct from Bank
/I 17 R06017 X Question Information Examination Outline Cross-Reference c Level I 29 I I Tech References OP-901-313 KIA 4.2-A58-AAI .03 Ref Supplied , Imp. Rating I 31 I
Cognitive Level 2 10 CFR 55.41(b) 1 7
~ Learning Objective WLP-OPS-PPO30 04 I
I 1 Proposed Question 1 Which of the following components may be operated from CP-1 following a loss of the TGB-DC Bus?
- ~____
iA Steam Bypass Control Valves B 2A(B) to 3A(B) Tie Breakers C Condensate Pumps 1D Main Feedwater Pumps ~~-
1 1
~ ~~
inswer D Explanation D is correct; Stated in OP-901-313 p 3lstep 9 and in the preceding Caution.
A is incorrect; Stated in OP-901-313 p 30 Caution 4.
E B and C are incorrect; OP-901-313 p 16 step C.4 states that all on the 7KV and 4KV TGB switchgear have no control Dower and are disabled.
Comments
OBJECTIVES: Permanent objective numbers listed in brackets I].
Objectives [I] through 161 applicable to LO and STA.
Objective 161 applicable to NAO.
DIAGNOSE the event and IDENTIFY the appropriate Off-Normal Operating Procedure to use to combat the event. [I]
IDENTIFY and APPLY the Immediate Operator Actions (if any) applicable to the diagnosed event, from memory. 121 IDENTIFY and APPLY procedural steps, cautions, and notes applicable to the diagnosed event. [3]
INTERPRET content, location, and sequencing of procedural steps applicable to the diagnosed event. [4j IDENTIFY and APPLY Technical Specifications applicable to the diagnosed event. [5]
EXPLAIN the tasks performed outside the Control Room regarding applicable Off-Normal Operating Procedures. 161
Loss of a 125 Volt DC Bus OP-901-313 Revisbn 3 C. AUTOMATICACTIONS (CONT'D)
- 4. LOSSOf 125 Volt DC BUSTGB-DC 0 Station Air AND Instrument Air compressors will fail to load.
0 Loss of Air Side Seal Oil Backup Pump.
0 Loss of Main Turbine DC Emergency Bearing Oil Pumps.
0 Loss of Emergency Lube Oil Pumps for the Main Feed Pump Turbines.
0 LQSSof Voltage Regulator control.
0 The following 4KV AND 7KV switchgear are left without control power.
remote manual control AND automatic protection of the switchgear AND associated connected components are disabled:
0 7KVBusIA 0 7KVBusIB 4KVBus2A 0 4KVBus2B 0 4KVBus4A 4KVBus4B Ii NOTE 480V breakers overcurrent protection is available from Electronic Current Sensing Device (ECS).
0 The following 480V Switchgear are left without control power. remote manual control of the switchgear AND associated connected components are disabled:
0 480V BUS21A 480VBus21B 480VBus22A 480VBus22B 0 480V BUS23A 480V BUS23B.
16
Loss of a 125 Volt DC Bus OP-901-313 Revision 3 Eq LOSS OF 125 VOLTDC BUS TGB-DC CAUTION
- 1. THE MAIN TURBINE AND MAIN FEEDWATER PUMP TURBINE (FWPTS) DO NOT HAVE DC LUBE O!L PUMPS AVAILABLE DUE TO LOSS OF BL'S TGB-DC.
EXTREME CARE MUST BE TAKEN TO ENSURE AC LUBE OIL PUMPS DO NOT LOSE POWER UNTIL TURBINE SHAFTS STOP. POWER SUPPLIES FOR AC LUBE OIL PUMPS ARE:
MAIN TURBINE AC BEARING OIL PUMP LOG-E3KR3?3AS-2M 0 FWPTAMOPI LOF-EBKR211A-5D FWPTBMOPI LOF-EBKR212A-5D 0 FWPTAMOP2 LOF-EBKR212B-5D 0 FWPTBMOP2 LOF-EBKR211B-5D.
- 2. ALL NON-SAFETY 7KV, 4KV SWITCHGEAR ARE LEFT WITHOUT CONTROL POWER. ALL REMOTE MANUAL CONTROL AND AUTOMATIC PROTECTION OF THESE LOAD CENTERS AND CONNECTED COMPONENTS ARE DISABLED.
- 3. 480V BREAKERS WILL HAVE OVERCURRENT PROTECTION FROM ECS, BUT ARE LEFT WITHOUT CONTROL POWER.
- 4. IF TGB-DC POWER CANNOT BE RESTORED, THEN ALL MAIN STEAM BYPASS VALVES CANNOT BE OPENED FROM CP-I AND WILL NOT OPEN ON A REACTOR TRIP CUTBACK.
PLACEKEEPER START DONE
- 1. Verify Automatic Actions (Section C>take place as designed.
0 0
- 2. Start Emergency Diesel Generators A AND B on CP-1.
c l n
- 3. Manually close Condensate Pump A, 6 C recirculation valves CD 137A, CD 137B, AND CD 137C.
NOTE Buses 2A and 2B will not transfer to Startup Transformers in Steps 4 and 5 due to the loss of control power to their supply breakers.
- 4. Transfer bus 1A supply from Unit Auxiliary Transformer (UAT)
A to Startup Transformer (SUT) A by positioning Bus A Transfer selector switch (CP-1j to SUT.
30
Loss of a 125 Volt DC Bus OP-901-313 Revision 3 E4 Loss OF 125 VOLTDC Bus TGB-DC <CzliiiD)
PLACEKEEPER START DONE
- 5. Transfer bus 1B supply from UAT B to SUT B by positioning Bus B Transfer selector switch (CP-1) to SUT.
- 6. Verify Main Generator Breaker A AND Main Generator Breaker 0 B (CP-I) indicate closed.
7.1 Close Startup Transformer A 4KV ISOLATION breaker (4KV-EBKR2A-4) by depressing the 0
MANUAL CLOSE pushbutton.
7.2 Open Feeder From Unit Auxiliary Transformer A breaker 4KV-EBKRZA-1 by depressing MANUAL TRIP pushbutton.
n u 8.1 Close Startup Transformer B 4KV Isolation breaker 4KV-EBKR2B-4 by depressing MANUAL CLOSE pushbutton 8.2 Open Feeder From Unit Auxiliary Transformer B breaker 4KV-EBKR2B-1 by depressing MANUAL 0
TRIP pushbutton.
MAIN FEEDWATER PUMP AUTOMATIC TRIPS, WITH THE EXCEPTION OF MECHANICAL OVERSPEED, ARE LOST. MAIN FEEDWATER PUMP GOVERNOR CONTROL VALVE ON CP-1 WILL INDICATE GREEN (LESS); HOWEVER, MAIN FEEDWATER PUMP WILL NOT BE IN MANUAL.
- 9. Maintain discharge pressure of BOTH Main Feedwater Pumps cont.
approximately 100 psia greater than Steam Generator pressure by adjusting SGFP Speed Controller FW-IHIC-1107, F\N-IHIC-1108 in Manual using indications on CP-1.
31
~ATEP,FORD3 Number:WbP-QPS -FWPOO Revision: 03 NUCLEAR PLANT 5 50 APPROXIMATE TIME: 4 Hours LESSON PLAN R~pe:
OBJECTIVES:
Upon completion of this lesson and given applicable reference materials, the student should be able to:
0 SUMMARIZE the operation and function of: [2]
0 Steam Generator Feedwater Pump (SGFP) Turbine 0 Steam Generator Feed Pump 0 SGFP Governor/Stop Valves 0 SGFP Turbine HP Steam Supply Drain Valves 0 SGFP Turbine Thrust Bearing 0 SGFP Turbine Rupture Discs 0 SGFP Gland Seal System 0 SGFP Lube Oil System 0 SGFP Turning Gear 0 System Instrumentation, Controls and Alarms 0 SGFP Trips SUMMARIZE the design features which provide for: [3]
0 SGFPWarmup 0 SGFP Turbine Manual Trip and Reset 0 SGFP Lube Oil Sump High and Low Level Test SGFP Gland Seal Cooling Prevention of water intrusion on high Hotwell level 0 IDENTIFY the interlocks which provide for: [4]
Automatic Turbine/Reactor Trip Runback Start permissives for SGFP 0 Automatic SGFP Recirculation Flow Discharge Valve Closure
IDENTIFY the physical connections and STATE the interrelations with: [I]
0 Condensate 0 Main Steam 0 Turbine Closed Cooling Water Reheat Steam 0 Main Condenser 0 Lube Oil Storage, Transfer and Purification System 0 EXPLAIN the following theoretical concepts as they apply to the Steam Generator Feedwater Pump and Turbine: [5]
0 Reason for balancing Steam Generator Feed Pump loads 0 Characteristics of level, flow and pressure indications 0 Given plant status, EXPLAIN the bases for procedural steps and requirements, including notes and cautions for the following procedures: [6]
OP-003-033, Main Feedwater (As it pertains to Feedwater Pumps) 0 INTRODUCTION TP-1 0 Review Objectives 0 System Purpose TP-2 The Steam Generator Feed Pumps, in conjunction with the Feedwater and Feedwater Control Systems, are designed to supply feedwater to the steam generators at the required flow rates and pressures under all conditions.
o The Steam Generator Feed Pump Turbines are used to convert therms! energy tz mechanica! energy tz drive the SGFPs.
Design Bases 0 Designed to supply full load feedwater plus steam generator blowdown flow at maximum anticipated operating pressures,
OBJECTIVES:
STATE the purpose of the DC Distribution system. [ I ]
STATE the design basis of the system. [2]
STATE the purpose of the major components: [3]
Battery Chargers Battery Distribution Panels EXPLAIN how a battery charger operates. [4]
Given a one line diagram of the system, LABEL the components, buses, and PDPs. [5]
IDENTIFY the causes of alarms associated with the system. [6]
Given the Technical Specifications and plant status, IDENTIFY and EVALUATE the operability, actions and bases for the following: [7] [SI 3.8.2.1 3.8.2.2 3.8.3.1 0 3.8.3.2 TRM 3.8.3.Icompliance Given OP-901-313, SUMMARIZE the effects of a loss of the different DC distribution buses: [8]
125DC BUSA-DC 0 125DC BUSB-DC 125DC BUSAB-BC 125DC BUSTGB-DC
Waterford 3 2006 RO Exam Question 64 1
I 64 Chemistry has informed the Control Room that the in service Gas Decay Tank Oxygen concentration is reading 4.7% and the Hydrogen concentration is reading 57.7%.
Per Technical Specifications, we must suspend all additions of waste gas, and reduce the concentration of Oxygen to 5 4%.
B Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> / Within 1 Hour Waterford 3 recommends deletion of this question for the following reason: The question tested at a memorization level that was not required by the referenced objective. 5 of 5 Reactor Operator Candidates missed this question.
The referenced objective states: Given the Technical Specijications andplant status, tdc;ztij& m d cvaiuatc the qwaSili+l, actions, and Sascsfor !he fsilawt3g:
3.3.3.11 3.11.2.5 a 3.11.2.6 Additionally, WLP-OPS-TS02, Non System Technical Specifications, objective 10 states: STATE the LCO, ApplicabiliQ and Action Requirement @f 1 houyjfor the foliowing Tech Specs: (see attached reference for the entire list). Technical Specification 3.1 1.2.5 is not listed among these required to state.
This question deals with Technical Specification 3.1 1.3.5, Radioactive Effluents Explosive Gas Mixture. No reference was supplied for this question thereby requiring the students to recall from memory Technical Specification 3.1 1.2.5 Limiting Condition for Operation and the associated actions.
If the condition given in the stem of the question were to exist in the plant, the Control Room would have already received Annunciator G090.5, WASTE GAS ANAL SYS H2/02 CONCENTRATION HIGH. The Annunciator Response Procedure, OP-500-007, gives guidance to refer to Technical Specification 3.1 1.2.5 and to coordinate with the Shift Chemist to verify Technical Specification 3.1 1.2.5 is met.
1
Waterford 3 2006 RO Exam Question 64
/j 64 1 R06064 I I
~
I i
I Ix Learning Objective WLP-OPS-GWMOO 1 07 Chemistry has informed the Control Room that the in service Gas Decay Tank Oxygen concentration is reading 4.7% and the Hydrogen concentration is reading 57.7%.
Per Technical Specifications, we must reduce the concentration of Oxygen to 5 4%.
A 1 Immediately / Immediately suspend all additions of waste gas, and i
B Within 1 hour/ Within 1 Hour 1 C l*/ithin 1 hgUT! iinmediately I D lmmediatelv /Within IHour Answer A Explanation A is correct; TS Action b -with Oxygen concentration >4% and Hydrogen >4%, immediately suspend all additions of waste gases to the system and immediately reduce Oxygen level <4%.
B,C, and D are incorrect; but plausible combinations.
1 Comments /j 1 Old only addressed time of TS, old values were 8, 24, and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Q 2
Explain the relationship of hydrogen/oxygen concentrations to flammability.
Given the Technical Specifications and plant status, identify and evaluate the operability, actions, and bases for the following:
3.3.3.1 I 3.1 1.2.5 3.1 I.2.6 EXPLAIN the relationship of hydrogedoxygen concentrations to flammability. [6]
Given the Technical Specifications and plant status, IDENTIFY and EVALUATE the operability, actions, and bases for the following: [?I (Not ayplicable to NAOs)
? I ?
3.32.1 I 3.11.2.5 3.11 2.6 12
RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen i n the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by vol urne.
APPLICABILITY: A t all times.
ACTION:
- a. With the concentration o f oxygen in the WASTE-GAS HOLDUP SYSTEM greater than 2% by volume but less than or equal to 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- b. With the concentration o f oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by vol ume and the hydrogen concentration greater than 4% by voiume, immeaiateiy suspend a l l additions o f waste gases to the system and immediately reduce the concentration o f oxygen to less than or equal to 4% by volume and then take the ACTION in a.
above.
C. The provisions o f Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations o f hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 a f Specification 3.3.3.11.
WATERFORD - UNIT 3 3/4 11-16
K-5 GO905 CWD K903 WASTE GAS ANAL SYS H2/02 CONCENTRATION HIGH INITIATING DEVICE SETPOINT High 0 2 or H2 alarm on Waste Gas NA Analyzer Local Panel POSSIBLE EFFECTS
- 1. Refer to Technical Specification 3.1 1.2.5 CONTROL ROOM INDICATIONS LOCAL INDICATIONS NONE Waste Gas Analyzer Local Panel POSSIBLE CAUSES RECOMMENDED ACTIONS
- 1. See initiating device. 1.IInform duty shift chemist of alarm.
1.2 Coordinate with shift chemist to verify Technical Specification 3.1 1.2.5 is met.
OP-500-007 Revision 11 Attachment 4.85(1 of I) 126
INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The explosive gas monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE wi th'their alarmitrip setpoints set to ensure t h a t the lfmits o f Specification 3.11.2.5 are not exceeded.
APPLICABILITY: As shown in Table 3.3-13.
ACTION:
- a. With an explosive gas monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown i n Table 3.3-13.
- b. With less than the minimum number of explosive gas monitoring instrumentation channel s OPERABLE, take the ACTION shown i n Table 3.3-13. R e s t m e the tnaporable instrumentation t c OPERABLE status within 30 days and, if unsuccessful , prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.
WATERFORD - UNIT 3 3 i 4 3-60 AMENDMENT NO. 27,68