ML032450535
| ML032450535 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/21/2003 |
| From: | Berry P Entergy Nuclear Northeast |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-333/03-301 | |
| Download: ML032450535 (16) | |
Text
LOI-03-0 1 POST EXAMINATION REVIEW ANSWER KEY CHANGE QUESTION NUMBER 0 12/13 ISSUE 0 2 Correct answers, A and B RECOMMENDATION 0 Accept both answers as correct RESULT 0 No impact on sample plan. All candidates demonstrated K/A knowledge IMPACT 0 Exam balance of coverage and content validity are preserved BASIS A11 incorrect respondents selected distracter B 0 No proctor notes, candidate notes or debrief comments 0 Additional review conducted based upon high error rate CONCLUSION Although not specifically stated, the intent of the question was to recognize that the directhmmediate impact of the high pressure signal is to trip the recirculation pump drive motor breakers and energize the alternate rod insertion solenoids thus making A the correct response.
The examination development process failed to recognize that the generator field breakers would also trip as a result of the respective drive motor breaker trip.
The high-pressure signal referenced in the stem is a direct trip of the recirculation pump drive motor breakers (71-10110/10210) (Reference attached ARP-09-4-2-16). The generator field breakers will open approximately 17 seconds later (ReferenceNOTE preceeding step F. 1.5 of the attached OP-27, Section F. 1 excerpt).
The question (stem) conditions did not specify a sequence or direct result condition therefore B is also a correct response.
I I
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 High Reactor Pressure 1 3 Group # 1 1 Ability to operate andlor monitor the following as K/A # 295025 EA1.07 EA1.07 they apply to HIGH REACTOR PRESSURE:
(CFR: 41.7 145.6)
ARIIRPTIATWS: Plant-Specific Importance Rating 4.1 4.1 Proposed Question: Which QNJ of the following describes the effect a reactor vessel pressure signal of 1170 psig will have on the reactor recirculation pumps and alternate rod insertion (ARI) system?
The Recirculation motorIgenerator...
a) drive motor breakers will trip and the ARI solenoid valves will energize ROlSRO b) generator field breakers will trip and the ARI solenoid valves will energize 12/13 c) drive motor breakers will trip and the ARI solenoid valves will de-energize d) generator field breakers will trip and the ARI solenoid valves will de-energize.
Proposed Answer: a) drive motor breakers will trip and the ARI solenoid valves will energize.
Explanation (Optional):
Technical Reference(s): ITS-3.3.4.1ISR-3.3.4.1.4 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: SDLP-02H EO 1.05.C.2, SDLP-O3C EO1.05.C,2 (As available)
Question Source: Bank # Quad Cities 1 INPO Bank # 16832 (Modified for JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 311611998 (Optional - Questions validated at the facility since 10195 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 2 Comments:
RECIRCULATION SYSTEM CAUTION F.l.l Verify the following:
All control rods are full in.
Each running RWR MG set is at minimum speed ( 3 0 8 ) .
NOTE 1: Annunciator 09-4-3-2 RInIR LOOP A OUT OF SERVICE will alarm when only XWR Pump B is running.
NOTE 2 : RhT P u p A will trip when 02MOV-53A is 12% open.
F.1.2 Close RWR PMP A DISCH 02MOV-53A.
F.1.3 Verify RWR PMP 02-2P-1A is tripped.
F.1.4 Place RWR PMP 02-2P-1A control switch in PULL TO LOCK.
NOTE : RWR MG Set A generator field breaker will open approximately 17 seconds after RWR Pump A drive motor breaker trips.
F.1.5 Verify open RWR MG A GEN FlELD BKR.
Rev. No. 59 Page 43 of 176
ANNUNCIATOR RWR MG A ARP 09-4-2-16 LEGEND ATWS TRIP DEVICE 0 02-3LT-72A or C and 02-3LT-72B or D 0 02-3PT-102A or C and 02-3PT-102B or D SETPOINT Reactor pressure
- CCTSI LESS THAN OR EQUAL TO 1 1 5 5 psig (when,either zero or one SRV is out of service and MODE switch in Run)
CITSI 11153 psig with 210 SRVs OPERABLE
- I:CTSI LESS THAN OR EQUAL TO 1120 psig (when two or more SRVs are out of service and MODE switch in Run)
[ITS1 11118 psig with 10 SRVs OPERABLE 0 Reactor water level
- GREATER THAN OR EQUAL TO 105.4 inches above TAF (when MODE switch in Run)
Reference:
1.62-150, 1.62-151, ESK-7FH, FE-1E CAUSES 0 Reactor water level - Low Low 0 Reactor pressure - High AUTOMATIC ACTIONS 71-10110 (feed to RWR MG set A motor) TRIP PROCEDURE Verify RWR PMP 02-2P-1A is tripped.
Rev. No. 7 I (F Page 1of 2
Interoffice Correspondence August 4,2003 JENG-03-0215 MEMO TO:
FROM: STEVE BONO
SUBJECT:
TRIPS ASSOCIATED WITH THE RWR-MG SET GENERATOR FIELD BREAKER Generator Trip Circuit Elements The generator field breaker's trip circuit is comprised of three discreet contacts in the trip logic as shown on attached print 1.62-151. Two of the contacts are associated with the Generator Lockout circuit as shown on print 1.62-150. An actuation of the lockout circuit is designed to energize the trip coils for both the drive motor breaker and the generator field breaker. The last contact in the generator field breaker trip circuit is the Generator Loss of Field Aux relay. The Generator Loss of Field Aux relay is actuated by two contacts in series. The series contacts are Field Application and Under Voltage relay and Generator Loss of Field relay. The Field Application and Under Voltage relay (print 1.62-152) is energized during the pump start sequence to block the trip of the generator field breaker for the Generator Loss of Field relay until the start sequence is complete. The Generator Loss of Field relay (print 1.62-153) is the field undervoltage relay and during a start of the system the field is in an undervoltage condition so this trip would require a by-pass until the field voltage is established.
System response to a manual trip of the Drive Motor Source Breaker or any other trip that is not common to the Generator Lockout circuit The manual trip of the Drive Motor Source Breaker will de-energize the drive motor. The inertia of the MG set will maintain the generator output for some finite period of time.
The output of the generator will decrease in voltage as the generator slows down. The generator field is powered from the exciter, a smaller generator driven directly by the Drive Motor. Exciter output voltage also decreases as motor speed decays. When the voltage from the exciter decreases to the set point of Generator Loss of Field relay the Generator Field Breaker will trip.
a
MEMO TO: PAT BERRY August 4,2003 FROM: STEVE BONO JENG-03-0215
SUBJECT:
TRIPS ASSOCIATED WITH THE RWR-MG SET GENERATOR FIELD BREAKER Page 2 of 2 System response to a Generator Lockout circuit actuation The actuation of the Generator Lockout circuit will energize the trip coils to both the drive motor breaker and the generator field breaker de-energizing the drive motor and the generator field with no coast down output of the generator.
The above information was supplied at the request of Rick Devercelly and if you have any further questions feel free to contact Keith Brazeau at x6014.
KEITH BRAZEAU PEER REVIEW SYSTEM ENGINEER LARRY LEITER SYSTEM ENGINEER Attachments KB/S B/dj c CC: JENG File
LOI-03-0 1 POST EXAMINATION REVIEW ANSWER KEY CHANGE QUESTION NUMBER 23/26 ISSUE Two correct answers, B and D RECOMMENDATION Accept both answers as correct RESULT No impact on sample plan. All candidates demonstrated K/A knowledge IMPACT Exam balance of coverage and content validity are preserved BASIS All incorrect responses selected distractor By.
No proctor notes, candidate notes or debrief comments Additional review conducted based upon exam analysis CONCLUSION Isolation of extraction steam to a Feedwater heater will result in a reduction of Feedwater heating and a new stable higher power level thus making D a correct response as originally intended.
Distracter B indicated that a rise in bus frequency supplying the recirculation motor generator set had occurred. Applying standard AC motor theory, this rise in fiequency will result in a new higher generator speed resulting in a new higher recirculation pump speed and ultimately a new higher reactor power.
Distracter Bywas written with the understanding that the speed feedback loop would return the generator to the previously selected speed demand thus returning reactor power to the previous value.
Post examination analysis revealed a plant modification (F 1-87-043) that removed the speed feedback loop to address a speed oscillation issue. System Engineering provided information on the modification during a series of communications evaluating this question.
Considering the removal of the speed feedback loop, a rise in bus frequency will result in a new higher stable power level thus making B a correct response.
L Examination 0utline Cross-reference: Level RO SRO Tier # 1 1 Inadvertent Reactivity Addition / 1 Group # 2 2 Knowledge of the interrelations between KIA # 295014 AK2.07 AK2.07 INADVERTENT REACTIVITY ADDITION and the following: (CFR: 41.7 / 45.8)
Reactor power Importance Rating 3.9 3.9 Proposed Question: From normal full power operation, which of the following will result in a stable higher power level?
a) Inadvertently isolating the Reactor Water Cleanup System ROlSRO b) Raising 10100 Bus frequency.
23/26 c) Main Condenser Circulating Pump Trip.
d) Closing the manual extraction steam valve for Feed Heater 6B Proposed Answer: d) Closing the manual extraction steam valve for Feed Heater 6B Explanation (Optional): Explanation:
a) Inadvertently isolating the Reactor Water Cleanup System results in higher feedwater temperature- therefore a lower power level.
b)Raising 10100 bus frequency will momentarily raise Recirc MG speed. Speed vs Speed demand will reduce it back down.
c)A Main Condenser Circulating Pump Trip will result in higher condensate and therefore feedwater temperature resulting in a lower power level.
d)The manual extraction steam valve for Feed Heater 66 closing will prevent the heating of the feedwater in the 6B heater, thereby, causing colder feedwater to enter the vessel and drive reactor power up.
Technical Reference(s): AOP-62, AOP-32, OP-3A (Attach if not previously provided)
Proposed references to be provided to applicants during examination: None Learning Objective: LP-AOP EO 1.02 (As available)
.Question Source: Bank # Clinton INPO # 20412 (Modified to JAF)
Modified Bank # (Note changes or attach parent)
New Question History: Last NRC Exam 7/23/200 1 (Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Comments:
ASTER Course/Program: BWR Technology ModuWLP ID: SDLP-021
Title:
Recirculation Flow Control Course Code:
Preparer: E. Riley p
/& Revision/No: 9 (PrintlSignature)
~
Prerequisites: Date:
(Handwritten)
FiMD Est. Teach Time: 4 Hours Lead Accreditation Specialist Review (Rev 0 required) Date:
PrinUSignature flh Technical Review:
PrintlSignature Training Supervision Approval Date:
I Y/dr PtWSignature:
OBJECTIVES The following shall be accomplished from memory and without error except where otherwise noted:
Objective # Licensed OperaforIShift Technical Advisor Objective Description 1.01 State the purpose(s) of the Recirculation Flow Control System.
1.02 State the Recirulation Flow Control System design bases as referenced in the FSAR.
1.03 NIA 1.04 State the ELECTRICAL DISTRIBUTION SYSTEM which powers the components listed below:
- a. M G Set Speed Control Scoop Tube Actuator
- b. MG Set Speed Limiter Signal Generator
- c. MG Set Speed Demand Limiter
- d. #2 Speed Limiters
- e. Jet Pump Flow Square Root Extractors G:UPLANS\sdlplnew fonnat\sdlp021R9 new formatdoc Page 1 of 58
ContentlSkills ActivitieslNoteslObjectives II. Presentation A. Purpose
- 1. The 'Recirculation Flow Control System provides: 1.01
- a. A high degree of stability for the recirculation system
- b. A reasonably fast control of reactor power
- c. A means of automatically reducing reactor power due to abnormal plant conditions
- d. A means of preventing plant transients caused by a loss of the control system signal B. Design Bases
- 1. Power Generation Design Basis: The Recirculation Flow 1.02 Control System is designed to allow variation of the recirculation flow rate.
- 2. Safety Design Basis: The Recirculation Flow Control System functions so that no abnormal operational transient resulting from a malfunction in the system can result in damaging the fuel or exceeding the Reactor Coolant System pressure limits.
C. System Description Figure 1
- 1. Reactor recirculation flow rate is changed by adjusting the speed of the two reactor recirculating pumps.
The master controller is located on panel 09-5 and can NOTE: Modification F1-87-043 be used to control the pump speed. In manual, the removed the speed control feedback signals controller will hold the recirculation pumps at the set to the M/A transfer speed. The master controller is pinned in the manual station. It is now used for position and automatic operation is prohibited. The speed indication only.
speed demand limiter provides the upper and lower speed limits which the master controller can send to the MIA transfer stations. The high limit is set to give 102.5% speed and the low limit is set at approximately 44% pump speed.
new format.doc G:\Training~Master~Files\Operations\lnitial\SDLP-O2l~sdlp021R9 Page 18 of 58
Conte nt/SkiI Is Activities/Notes/Objectives H. Technical Specifications
- 1. Discuss Section 3/4.2.6, Recirc Pump Trip Instrumentation (CTS
- 2. Discuss Section 3.3.4.1, ATWS-RPT Instrumentation (ITS)
I. Industry/Plant Operating Experience
- 1. Modification F1-87-150 During several years in the 1980's both scoop tube positioners were locked during normal plant operation due to noise problems in the recirculation flow control system. Modification F1-87-150 added "Auto - Unlock" switches which would automatically unlock the scoop tube brake and allow the scoop tubes to "Runback" to the #2 speed limiter (44% position) when the "Auto-Unlock" switch is "ON" and a low vessel level with < 2 RFPs exists.
- 2. Modification F1-87-043 Recirculation Flow Control Improvement Both recirculation speed control loops were modified from closed to open loop type control by this modification. The Speed demand indicator was moved in the circuitry so as to provide "instant" speed indication to the operator. A new rate limiter was added to limit rate of change in recirc pump speed.
- 3. RlCSlL #066 4/1/94 One of two recirculation pumps inadvertently increased in speed from minimum to maximum speed and remained at maximum speed for 30 minutes. The plant was in a refueling outage with vessel level near the vessel flange. As a result of this speed increase, excessive crud was released negatively impacting visibility and radiation levels. Per GE Nuclear Energy's recommendation and our operating experience evaluation, OP-27 requires the scoop tubes be locked if the recirculation pump will be at minimum speed for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
new formatdoc G:\Training~Masfe~FileslOperafions\lnifiai\SDLP-O2i\sdlpO~lR9 Page 45 of 58
LOI-03-01 POST EXAMINATION REVIEW ANSWER KEY CHANGE QUESTION NUMBER 0 35/47 ISSUE 0 No Correct answer RECOMMENDATION 0 Discharge from examination RESULT 0 Reduces Tier 1 Group 1 by one A1 item leaving 2 IMPACT 0 Exam balance of coverage and content validity are preserved BASIS All incorrect responses selected the D distracter.
0 During exam administration, 2 candidates questioned if UPS was still out of service. Both were told YES. Both responded that this resulted in no correct answer.
One candidate documented an assumption on the exam (attached) indicating after UPS returned and then selected his response.
0 Two candidates documented a dilemma with Blue SCRAM lamps and SCRAM Rod Positions Five candidates debriefed a dilemma with using the blue scram lamps to verify scram rod positions 0 An additional instructor review also indicated that there was no correct answer.
CONCLUSION A loss of the UPS power supply will de-energize RPIS resulting in a loss of all rod position indication.
In the event of a reactor scram, the loss of rod position indication will require EOP-3 entry.
The question (stem) conditions ask how the operator can verify rod position. This is not possible with UPS de-energized.
None of the responses offers that no rod position information is available.
Per AOP-21, excerpt attached, the operator can only verify that rods scrammed by observing a downward power trend and individual rod blue scram lamps and yellow accumulator trouble lamps.
Those candidates who selected the intended response apparently made an unstated reasonable assumption that scram rod positions translated to confirm reactor scram as stated in AOP-2 1.
One candidate has a stated reasonable assumption that U P S was restored and selected an incorrect response that was consistent with the assumption. Page 47 of this candidates examination package is attached.
Misleading stem information and the lack of proctor guidance forced assumptions on the part of the candidates to support an answer selection when no correct answer existed.
Examination Outline Cross-reference: Level RO SRO Tier # 1 1 SCRAM Condltion Present and Power Above APRM Group # 1 1 Downscale or Unknown / 1 Ability to determine andlor interpret the following WA # 295037 EA2.05 EA2.05 as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
(CFR. 41 10f43.5/45 13)
Control rod posltion Importance Rating 42 4.3 Proposed Question: While operating at full power, the plant has experienced a complete loss of UPS Operator actions failed to prevent a Reactor SCRAM.
Plant control is directed by (1 1 and SCRAM Rod Positions are verified by (2) a) (1) EOP-2 (2) Green Full In Lamps on the Full Core Display b) (1) EOP-2 RO/SRO (2) Blue SCRAM Lamps on the Full Core Display 35/47 C) (1) EOP-3 (2) Blue SCRAM Lamps on the Full Core Display d) (1) EOP-3 (2) Green Full In Lamps on the Full Core Display Proposed Answer: C)(1 ) EOP-3 (2) Blue SCRAM Lamps on the Full Core Display Explanation (Optional): Question was rewritten as SRO Only swapped original RO(SR0 question 35147 to make it a SRO Only- S35,SRO original question S35 was made ROlSRO question 35/47.
RPlS is inoperable with a loss of the UPS. Per AOP-21, the SCRAM is verified by confirming the Blue 8 Yellow lamps lit on the Full Core Display. This indication only confirms that the SCRAM Inlet 8 Outlet Valves opened and the accumulator discharged. With no immediate way of confirming Rod Digital Position Indication, the Operators are forced to conclude that Rod Position is Unknown, thus entry into EOP-3 is required.
Technical Reference(s): EOP-3 (Attach if not previously provided)
Proposed references to be provided to applicants during examination: EOPs Learning Objective: LP-AOP, EO-1.03, EOP3LP, EO-I. 07 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New NEW Question History: Last NRC Exam
(Optional - Questions validated at the facility since 10/95 will generally undergo less rigorous review by the NRC; failure to provide the information Will necessitate a detailed review of every question.)
Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Comments:
I .
2003 Initial NRC Exam SRO
- 47. While operating at full power, the plant has experienced a complete loss of UPS.Operator actions failed to prevent a ReactoeRAM.,
Plant control is directed by (1) and SCRAM Rod Positions are verified by (2) k (1) EOP-2
./
(2) Green Full In Lamps on Core Display (1) EOP-2 (2) Blue SCRAM Lamps on the Full Core Display C. (1) EOP-3 (2) Blue SCRAM Lamps on the Full Core Display
@(1) EOP-3 (2) Green Full In Lamps on the Full Core Display
LOSS OF UPS* AOP-21 (J) c.1 Complete Loss of UPS - P r o m p t A c t i o n s (cont)
C.1.4 IF a reactor scram occurs, THEN perform AOP-1 using the following guidance in any order or concurrently.: I NOTE 1: SRM and IRM detectors cannot be driven into core.
NOTE 2: APRM/IRM and SRM recorders are not operable.
- a. Verify reactor power is lowering using the following indications:
SRM meters and IRM downscale indicators at panel 09-5 (-1 EPIC APRM indication (-1 NOTE: RPIS is not operable.
- b. Confirm reactor scram by verifying the following lights are on at full core display:
Yellow ACCUM Blue SCRAM NOTE: SRV acoustic monitors are not operable.
Rev. No. 18 Page 10 of 23