ML102020502
| ML102020502 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/20/2010 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-259/10-301, 50-260/10-301, 50-296/10-301, ES-401, ES-401-5 50-259/10-301, 50-260/10-301, 50-296/10-301 | |
| Download: ML102020502 (21) | |
Text
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RD SRO G2.1.8 (10CFR 55.41.10) Tier#
Ability to coordinate personnel activities outside the control room. Group#
KIA# G2.1.8
- sf Importance Rating 3.4 L Question
- # 68 Unit 2 is operating at 90% Reactor Power, when a power reduction is required to be performed locally at the VFD.
Communications and coordination have been established between the operator at the VFD and the Unit Operator in the Control Room.
In accordance with 2-01-68, Reactor Recirculation System, which ONE of the following is the MINIMUM personnel requirement at the VFD to perform Speed Control manipulations?
A. Reactor Operator ONLY B. Reactor Operator AND a second Reactor Operator for peer checking C. Reactor Operator AND a Senior Reactor Operator for oversight D. Assistant Unit Operator AND a Reactor Operator directly supervising L Answer: A Post Exam Applicant Accept Both Answers A and C Comment DOCKET NUMBER 055-22940
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Cha11ene BASIS:
There are two procedures in which it is stated that a Unit Supervisor SHALL be present or provide direct oversight when reactivity manipulations are performed. As a result, C is also a correct answer to this question.
OPDP-1 Conduct of Operations Rev 16 page 13 step 3.6.B.2, states that the Unit Supervisor is responsible for all manipulations that affect reactivity and is charged to personally oversee all reactivity changes or assign another SRO to oversee the reactivity change if unable to give his/her undivided attention. During periods of frequent reactivity manipulations or significant plant evolutions (such as startup or shutdown), another SRO may be assigned to perform the reactivity management oversight function.
SPP-1O.4 Reactivity Management Program Rev 8 page 12, step 3.2.6.J; states that the Unit Supervisor shall provide direct oversight (line of sight within normal conversation level distance) for all reactivity manipulations (may be performed by a dedicated SRO for major reactivity evolutions such as reactor startup).
OPDP 1 B. The Unit Supervisor is responsible for all manipulations that affect reactivity and is charged to:
- 1. Promulgate and enforce the standards and expectations associated with reactivity management.
- 2. Give permission to Unit Operators to make reactivity changes. Personally oversee all reactivity changes or assign another SRO to oversee the reactivity change if unable to give his/her undivided attention. During periods of frequent reactivity manipulations or significant plant evolutions (such as startup or shutdown), another SRO may be assigned to perform the reactivity management oversight function
- 3. Utilize approved reactivity plans or notify the shift Reactor Engineer of unplanned reactivity changes greater than 1% thermal power.
- 4. Review and approve all planned reactivity changes or core alterations in accordance with approved procedures or instructions developed by Reactor Engineering. Ensure Reactor Engineering is available in the control room during planned significant reactivity evolutions.
- 5. Ensures that pre-job briefs for work activities address potential reactivity effects.
Personnel involved in reactivity manipulations or working on reactivity control equipment must be properly trained, must understand their roles and responsibilities, and must be briefed on management expectations. (Refer to Attachment 5, Pre-Job Briefing Guidelines.)
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet SPP 10.4 3.2.6 Unit Supervisors A. Are sensitive to the reactivity effects that may result from normal and infrequent evolutions.
B. Ensure that planned work activities have received appropriate reactivity management reviews and the necessary controls have been implemented into the work packages and/or procedures, including contingency or compensatory actions as needed.
C. Place emphasis during turnover and control board walk downs on items important to reactivity management.
D. Have the authority to terminate any activity in which the effects on reactivity control are unknown or non-conservative.
E. Have the responsibility and authority to trip the unit if there is uncertainty as to the units status with respect to the control of reactivity and control of the plant.
F. Ensure that the specific details of events or equipment problems related to the control of reactivity are recorded and initiates corrective actions.
G. Maintains a cautious approach to the adjustment or interpretation of power indication by questioning the reasons behind discrepancies that may exist between power measurements.
H. Evaluate the recommendations provided by the Reactivity Control Plan or verbally by the on-call Reactor Engineer. However, the election to take actions more conservative than the recommendations is within the Reactivity Management Philosophy.
I. Ensure all control rod movements are made in a deliberate, carefully controlled manner while constantly monitoring nuclear instrumentation and redundant indications of reactor power and neutron flux.
J. Provide direct oversight (line of sight within normal conversation level distance) for all reactivity manipulations (may be performed by a dedicated SRO for major reactivity evolutions such as reactor startup).
K. Attend Reactivity Management Review Board when requested.
FACILITY Supports the applicants comment; in accordance with SPP 2.0 Procedures and Document Control, the SPP and QPDP are both a higher tier document.
Below is and excerpt form SPP 2.0 and the procedure hierarchy 3.0 POLICIES AND MANAGEMENT EXPECTATIONS C. All applicable procedures must be followed when conducting activities encompassed by a written procedure. The performer is responsible for performance of the procedure as written, even though the procedure may not be required to be in hand or is being performed from memory.
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Figure 1 (Page 1 ofl)
NPG Procedure Hierachy EXTERNAL REQ UI REM ENTS WA POLICY Codes Reguiwians Siandards
- I 1
V WA PRINCIPLES, MPG PROGRAM PRAC11CES, & QA LICENSING REQ U IREM ENTS MANUALS DOCUMENTS PROCEDURES 4 NQAP so Tech Specs/S4R (As/WlKI NOM J J 1
Standard Programsl Business Processes arid Practices Standard Site Physical Site Radiological Department Procedures Security Plans Emergency Plans lon-QLaüty efated Qsaly Retate Department Ir Site Technical 1 1 Technical Instructions Site Physical Documents Security Corporate EPIPs Gthdes/Spcciftc&ons OPS4Wtht7tenence Instructions SIs/TJsIOther StandardsJOTher I
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Explanation A CORRECT: A Licensed Reactor Operator is the minimum qualification in (Optional): accordance with 2-01-68 B INCORRECT: Plausible in that peer checking would more than likely be used but not a requirement of 2-01-68 C INCORRECT: Plausible in that SRO oversight may be provided, but once again it is not a requirement D INCORRECT: Plausible in that it would be true if the AUO were enrolled in a License Training Program.
Justification: Requires knowledge of MINIMUM qualification AND coordination requirements to perform LOCAL VFD Speed Control manipulations for Reactor Recirc Pumps.
Technical Reference(s): 2-01-68 Rev. 134 (Attach if not previously provided)
OPDP-1 Rev. 15 Proposed references to be provided to applicants during examination: NONE Learning Objective: OPL171 071 V.B.6 (As available)
Question Source: Duane Arnold 2
Bank#
Modified Bank # (Note changes or attach parent)
New Question History: Duane Arnold Last NRC Exam 2007 (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 eve,y question.)
Question Cognitive Level: Memory or Fundamental x Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 X 55.43 Comments:
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO 295016 Control Room Abandonment /
Tier # 1 AA2.05 (IOCFR 55.43.5 SRO Only)
Ability to determine and/or interpret the following as they apply to Group# 1 CONTROL ROOM ABANDONMENT: KIA# 295016AA2.05
. Drywell pressure Importance Rating . 3.9 Question: # 77 Unit 2 was operating at 100% Reactor Power when the following series of events occurred:
- At 0200 an AIR LINE rupture in the Drywell results in a High Drywell Pressure Scram
- At 0205 Unit 2 Control Room evacuation is initiated due to a fire in the Control Bay
- At 0230 the Backup Control Panel, 2-25-32, is manned Which ONE of the following completes the statements?
In accordance with EPIP-1, Emergency Plan Implementing Procedure, the HIGHEST emergency action level classification that is required for these conditions is a (an) _(1)_.
In implementing 2-AOl-i 00-2, Control Room Abandonment, HPCI will cycle, upon demand, between the initiation AND high level trip setpoint until _(2)_.
[REFERENCE PROVIDED]
A. (1)Alert (2) it is secured in accordance with the Subsequent Actions B. (1)Alert (2) HPCI flow control is established at the Backup Control Panel C. (1) Site Area Emergency (2) it is secured in accordance with the Subsequent Actions D. (1) Site Area Emergency (2) HPCI flow control is established at the Backup Control Panel Answer: C POST EXAM APPLICANT COMMENT ACCEPT BOTH ANSWER A and C DOCKET NUMBER 055-23054 Explanation A INCORRECT: Part 1 incorrect Plausible in that per EPIP-1 2.1-A, Drywell (Optional): pressure at or above 2.45 psig AND Indication of Primary System leakage into Primary Containment requires declaration of an Alert. However, this is not the highest level of EAL Classification required. Part 2 correct as detailed in C below.
ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet B INCORRECT: Part 1 incorrect as detailed in A above. Part 1 is incorrect Plausible in that RCIC flow control is available at the Backup Control Panel, but HPCI is not. The similarity between these systems often causes confusion.
c CORRECT: Part 1 correct (see attached excerpts) Per EPIP-1 6.2-S, Control Room Abandonment from entry into AOl-i 00-2 or SSI-1 6 for ANY Unit Control Room AND Control of reactor water level, reactor pressure, and reactor power (for Modes 1, or 2, or 3) or decay heat removal (for Modes 4, or 5) per AOl-i 00-2 or SSI-1 6 as applicable, can NOT be established within 20 minutes after evacuation is initiated requires declaration of a Site Area Emergency. Part 2 correct (see attached excerpts) In the subsequent actions of AOl-i 00-2, local breaker operation results in the shutdown and inoperability of H PCI.
D INCORRECT: Part 1 correct as detailed in C above. Part 2 incorrect as detailed in B above.
Justification: To successfully answer the question, candidate must recognize that HPCI is required to be secured after initiating on a High Drywell Pressure signal. Additionally, candidate must correctly classify the event (relative to the Drywell Pressure and Control Room Abandonment) which is a function unique to the SRO position. This question is rated as COMP due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome.
Technical Reference(s): EPIP-1 Matrix Rev. 45 (Attach if not previously provided) 2-AOl-i 00-2, Rev. 52 Proposed references to be provided to applicants during examination: EPIP-1 Matrix Rev. 45 Section 2 Section 6 Learning Objective: OPL171.208V.B.4 (As available)
Question Source: Bank #
Modified Bank # (Note changes or attach parent)
New x 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 eveiy question.)
Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 X Comments:
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Challenge BASIS: The question asks for the highest level of EAL classification, and the basis document of EPIP-1 6.2-S gives a possible justification for NOT classifying the event as a Site Area Emergency based on certain key parameters being controlled by automatic functions. HPCI initiation and automatic operation is specifically given as one of these instances. With no other evidence of uncontrolled key parameters in the question stem, the event could be classified as an Alert.
SITE AREA EMERGENCY EAL: Control Room Abandonment from entiy into 1, 2, or 3-AOI-100-2 or 0-SSI-16 for ANY Unit Control Room AND Control of reactor water level, reactor pressure, and reactor power (for Modes 1, or 2, or 3) or decay heat removal (for Modes 4, or 5) per 1, 2, or 3-AOl-100-2 or 0-SSI-16 as applicable, can NOT be established within 20 minutes after evacuation is initiated.
OPERATING CONDITION: ALL BASIS: This event classification is intended to recognize loss of control of critical parameters either by failure of equipment designed to automatically initiate for control of the parameter or failure to expeditiously transfer safety system control to the backup controls.
Fission product barrier damage may not yet be indicated but should be considered by assessing available parameters versus the status of safety systems and the ability to control critical parameters. In Mode 4 and Mode 5 operator concern should be directed towards maintaining core cooling using decay heat removal systems. In power operation, hot standby, and hot shutdown operator concern is primarily directed toward maintaining critical parameters, (i.e., level, pressure, power, and heat sink) and thereby assuring fission product barrier integrity. The 20 minute time period is based on time required for personnel to leave the control room, arrive at the appropriate backup control station, and take control of critical parameters before core uncovety or core damage has occurred. This timeframe has been projected within the Tennessee Valley Authority, Browns Ferty Nuclear Plant, Fire Protection Report. During execution of procedures and transfer of equipment control, the listed critical parameters may be considered as being controlled if the parameters can be verified as being maintained within safe value ranges by appropriate equipment and automatic initiation functions designed to control the parameter (example: HPCI auto initiated and raised RPV level to a value above the initiation setpoint.).
Escalation to General Emergency is by fission product barrier degradation radioactivity release, or Emergency Director Judgment event classification.
Based on these references, the correct answer would be A.
The Facility does NOT SUpport the applicants comment that there are two correct answers.
The correct Emergency plan classification is Site Area Emergency, can not verify parameters are being maintained within safe value ranges within 20 minutes if the Backup Control Panel is not manned until 25 minutes after evacuation.
ES-401 Sample Written Examination Form ES-401-5 Question Worksheet BROWNS FERRY EMERGENCY CLASSIFICATION PROCEDURE EVENT CLASSIFICATION MATRIX EPIP-1 I I I I 6.3-UI I I I Turbine failure resulting in casing penetration OR C Significant damage to turbine or generator seals during operation.
m OPERATING CONDITION:
Model,or2 6.2-Al 6.3-Al Control Room Abandonment from entry into Turbine failure resulting in visible structural 1 2, or 3-AOl- 100-2 or O-SSI- 16 for ANY Unit damage to or visible penetration of ANY of the Control Room. following structures from missles:
- Reactor Building *Diesel Generator Building .
- Intake Structure *Control Bay m OPERATING CONDITION:
OPERATING CONDITION: Mode 1 or 2 ALL 6.2-SI I I I I Control Room Abandonment from entry into 1 2, or 3-Aol-I 00-2 or 0-SSI- IS for ANY Unit Control Room AND m Control of reactor water level, reactor pressure, m and reactor power (for Modes 1, or 2. or 3) or decay heat removal (for Modes 4. or 5) per 1, 2, or 3-AOl-i 00-2 or 0-SSI-16 as applicable, can C)
NOT be established within 20 minutes after evacuation is initiated, C)
OPERATING CONDITION:
ALL I I I I 0
m z
rTi I
PAGE 55 OF 206 REVISION 45
JPM NO. 202 REV. NO. 0 PAGE 1 OF 11 BROWNS FERRY NUCLEAR PLANT OBPERFOCEMEAS OPERATOR:_______________________
RO___ SRO___ DATE:________
JPM NUMBER: 202 TASK NUMBER: U-003-NO-04 TASK TITLE: Place a Second/Third RFPT in Service KIA NUMBER: 259001 A4.02 K/A RATING: R019 SRO: 3.7 TASK STANDARD: Places the Third RFPT in Service.
LOCATION OF PERFORMANCE: SIMULATOR REFERENCES/PROCEDURES NEEDED: 3-01-3 VALIDATION TIME: 20 minutes MAX. TIME ALLOWED: (Completed for Time Critical JPMs only)
PERFORMANCE TIME:
COMMENTS:
Additional comment sheets attached? YES NO RESULTS: SATISFACTORY____ UNSATISFACTORY____
SIGNATURE: DATE:
EXAMINER
BROWNS FERRY NUCLEAR PLANT INITIAL CONDITIONS: You are the Unit Operator at the controls. RFPT 3A is warmed and ready to be placed in service. Precautions and limitations have been reviewed. Radiation Protection has been notified that an RPHP is in effect for the impending action to place RFPT 3A in service.
Time of notification has been recorded in the NOMS Narrative Log. Appropriate data and signatures have been recorded on Appendix A.
INITIATING CUES: The Unit Supervisor directs you to place RFPT 3A in service and in automatic level control per 3-01-3 Reactor Feedwater System section 5.7.
JPM NO. 202 REV. NO. 0 PAGE 3 OF 12 IN-SIMULATOR: I will explain the initial conditions and state the task to be performed. I will provide initiating cues and reports on other actions when directed by you. When you complete the task successfully, the objective for this job performance measure will be satisfied. When your task is given, you will repeat the task and I will acknowledge Thats Correct. (OR Thats Incorrect, if applicable). When you have completed your assigned task, you will say, my task is complete and I will acknowledge that your task is complete.
INITIAL CONDITIONS: You are the Unit Operator at the controls. RFPT 3A is warmed and ready to be placed in service. Precautions and limitations have been reviewed. Radiation Protection has been notified that an RPHP is in effect for the impending action to place RFPT 3A in service.
Time of notification has been recorded in the NOMS Narrative Log. Appropriate data and signatures have been recorded on Appendix A.
INITIATING CUES: The Unit Supervisor directs you to place RFPT 3A in service and in automatic level control per 3-01-3 Reactor Feedwater System section 5.7.
JPM NO. 202 REV. NO. 0 PAGE 4 OF 12 START TIME____
Performance Step 1: Critical_ Not Critical _X 5.7 Placing the Second and Third RFP/RFPT In Service CAUTIONS
- 1) FAILURE to monitor SJAE/OG CNDR CNDS FLOW, 3-FI-2-42, on Panel 3-9-6 for proper flow (between 2 x 106 and 3 x 106 lbmlhr) may result in SJAE isolation.
- 2) Changes in Condensate System flow may require adjustment to SPE CNDS BYPASS, 3-FCV-002-0 190.
NOTE Placing RFP 3A(3B)(3C) MIN FLOW VALVE, 3-HS-3-20(13)(6), in OPEN position will lock it open, preventing minimum flow valve oscillations at low flow.
[1] BEFORE placing a RFPT in service:
[1 .1] NOTIFY Radiation Protection that an RPHP is in effect for the impending action to place RFPT 3A in service. RECORD time Radiation Protection notified in NOMS Narrative Log
[1.2] VERIFY appropriate data and signatures recorded on Appendix A per Appendix A instructions
[2] IF RFP/RFPT is NOT warmed, reset and rolling, THEN PERFORM the following: (Otherwise N/A)
Standard:
Given in initial conditions that step 1 and 2 are complete.
SAT_ UNSAT N/A COMMENTS:___________
JPM NO. 202 REV. NO. 0 PAGE 5 OF 12 Performance Step 2: CriticaiX Not Critical
[3] VERIFY RFP 3A MIN FLOW VALVE, 3-HS-3-20, in OPEN position.
CHECK OPEN MN FLOW VALVE, 3-FCV-3-20.
Standard:
Places RFP 3A MN FLOW VALVE, 3-HS-3-20, in OPEN SAT_ UNSAT N/A COMMENTS:_____________________________
FACILITY COMMENT:
In accordance with NUREG 1021, Appendix C, Job Performance Measures Guideline, every procedural step that the examinee must perform correctly (i.e., accurately, in the proper sequence, and at the proper time) in order to accomplish the task standard shall be identified as a critical step and shall have an associated performance standard.
For this JPM, Unit 3 was less than 70% Reactor Power with the B and C Reactor Feed Pumps in service. Placing RFP 3A MIN FLOW VALVE, 3-HS-3-20, in OPEN position will lock it open, preventing minimum flow valve oscillations at low flow. Although it is desired to minimize valve oscillations, such oscillations at given conditions are well within the capability of the FWLC System to respond to and will not result in a significant plant transient or significant diminished margin to safety.
The task standard for this JPM is to place RFPT 3A in service and in automatic level control per 3-01-3 Reactor Feedwater System section 5.7. Failure to perform this step will not prevent successful completion of the task standard. Therefore, we recommend this performance step to be changed to Not Critical.
JPM NO. 202 REV. NO. 0 PAGE 6 OF 12 Performance Step 3: Critical_ Not Critical X
[4] SLOWLY RAISE speed of RFPT using RFPT 3A SPEED CONT RAISE/LOWER, 3-HS-46-8A, to establish flow and maintain level in vessel.
[5] WHEN RFPT discharge pressure is within 250 psig of reactor pressure, THEN VERIFY OPEN REP 3A DISCHARGE VALVE, 3-FCV-3-19.
Standard:
Raises RFPT speed and verifies discharge valve open SAT_ UNSAT N/A COMMENTS:_________
JPM NO. 202 REV. NO. 0 PAGE 7 OF 12 Performance Step 4: CriticaiX Not Critical_
[6] SLOWLY RAISE RFPT speed using RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A, to slowly raise RFP discharge pressure and flow on the following indications (Panel 3-9-6):
Raises RFPT speed and commences injection to the Reactor SAT UNSAT N/A _COMMENTS:____________________________
Performance Step 5: Critical_ Not Critical X
[7] WHEN sufficient flow is established to maintain RFP 3A MIN FLOW VALVE, 3-FCV-3-20, in CLOSED position (approximately 2 x 106 lbmlhr),
THEN PLACE RFP 3A MIN FLOW VALVE, 3-HS-3-20, in AUTO.
Standard:
Places RFP 3A MIN FLOW VALVE in AUTO SAT_ UNSAT N/A _COMMENTS:_______
JPM NO. 202 REV. NO. 0 PAGE 8 OF 12 Performance Step 6: Critical_ Not Critical _X
[8] OBSERVE lowering in speed and discharge flows of other operating RFPs.
Standard:
Raises injection flow of RFPT 3A and monitors feedwater flow of other operating feed pumps.
SAT_ UNSAT N/A COMMENTS:____________________________
Performance Step 7: *CnticalX Not Critical_
[9] IF transferring RFPT from MANUAL GOVERNOR to individual RFPT Speed Control PDS, THEN PERFORM the following: (Otherwise N/A)
[9.2] VERIFY amber light at switch extinguished above RFPT 3A SPEED CONT RAISE/LOWER switch, 3-HS-46-8A.
3-SIC-46-8 (Panel 3-9-5):
[9.3.1] SELECT Column 3.
[9.3.2] VERIFY PDS in MANUAL.
Standard:
Pulls up on 3-HS-46-8A, verifies amber light extinguishes, selects column 3 and verifies in manual.
SAT_ UNSAT N/A _COMMENTS:______________________________
JPM NO. 202 REV. NO. 0 PAGE 9 OF 12 Performance Step 8:
- Critical X Not Critical_
[10] IF transferring control of RFPT from individual RFPT Speed Control PDS to AUTO control using REACTOR WATER LEVEL CONTROL PDS, 3-LIC-46-5, THEN PERFORM the following: (Otherwise N/A)
[10.1] VERIFY REACTOR WATER LEVEL CONTROL (PDS),
3-LIC- 46-5 is functioning properly and ready to control second or third RFP.
- [lo.3] WHEN RFP speed is approximately equal to operating REP(s) speed, THEN PERFORM the following on RFPT 3A SPEED CONTROL (PDS), 3-SIC-46-8:
[10.3.1] PLACE PDS in AUTO.
[10.3.2] VERIFY Column 3 selected.
Standard:
Raises REP discharge flow and pressure, places PDS in auto and verifies column 3 selected.
SAT_ UNSAT N/A COMMENTS:_________________________
JPM NO. 202 REV. NO. 0 PAGE 10 OF 12 Performance Step 9: Critical_ Not CriticaiX
[11] WHEN RFP in automatic mode on REACTOR WATER LEVEL CONTROL, (PDS) 3-LIC-46-5, THEN CLOSE the following valves:
Standard:
Closes the above listed valves and contacts AUO to verify the last two valves.
SAT_ UNSAT N/A _COMMENTS:_______________________
CUE: RFPT A HP STEAM SHUTOFF ABOVE SEAT DRAIN is closed RFPT A LP STEAM SHUTOFF ABOVE SEAT DRAIN is closed
JPM NO. 202 REV. NO. 0 PAGE 11 OF 12 Performance Step 10: Critical_ Not CriticaiX
[12] VERIFY CLOSED the following valves on first RFP started in Section 5.5:
Standard:
Verifies closed the above listed valves and contacts AUO to verify the last valves.
SAT_ UNSAT N/A _COMMENTS:_________________________
CUE: RFPT B, C LP STEAM SHUTOFF ABOVE SEAT DRAIN are closed
JPM NO. 202 REV. NO. 0 PAGE 12 OF 12 Performance Step 11: Critical_ Not CriticaiX
[13] VERIFY both RFPT Main Oil Pumps running.
[14] IF desired to stop Turning Gear for in service RFPT, THEN PLACE appropriate handswitch in STOP and RETURN to AUTO:
- RFPT 3A TURMNG GEAR MOTOR, 3-HS-3-1O1A
[15] REFER TO Section 6.0.
- CONTROL and MONITOR RFW system operation.
Standard:
Verifies both RFPT Main Lube Oil Pumps running, step 14 is NA and verifies proper operation of RFW system.
SAT UNSAT N/A _COMMENTS:_______
END OF TASK STOP TIME