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| =Text= | | =Text= |
| {{#Wiki_filter:. . - .. .. | | {{#Wiki_filter:}} |
| p# t "'%c UNITED STATES 3 '4 NUCLEAR REGULATORY COMMISSION 4 ,
| |
| S E j REGION IV 8
| |
| j 611 RYAN PLAZA DRIVE, SUITE 400 i
| |
| , ARLINGTON, TEXAS 760118064 March 25,1999 NOTE TO: NRC Document Control Desk Mai: Stop O-5-D-24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV
| |
| | |
| ==SUBJECT:==
| |
| OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON JANUARY 25-28,199's, AT DIABLO CANYON NUCLEAR POWER PLANT.
| |
| DOCKETS #50-275; 50-323 On January 25-28,1999, Operator Licensing Examinations were administered at the referenced facility. Attached you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
| |
| Item #1 - a) Facility submitted outline and the initial exam submittal for distribution under RIDS Code A070.
| |
| b) As given operating examination, designated for distribution under RIDS l Code A070. ;
| |
| n i
| |
| ~$
| |
| item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.
| |
| If you have any questions, please contact Laura Hurley, Licensing Assistant, Operations Branch, Region IV at (817) 860-8253.
| |
| On D A K 05 75 V PDR
| |
| | |
| m
| |
| . I ES-401 PWR SRO :xamination Outline Form ES 401-3 1
| |
| Facility: DCPP Units 1/2 Date of Exam: January 25,1999 Exam Level: SRO Tier Group K/A Category Points Po n 1 K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G a
| |
| : 1. 1 3 3 5 E$ ((( @ 5 5 $$$ $ 3 24 Emergency & 2 3 3 3 M @ I% 3 3 $$ 1 16 Abnormal Plant 3 0 0 1 % M $3 0 1 M fh$ 1 3 Evolutions TierTotals 6 5 9 %N$ 9 9 $5 $$ 5 43 l
| |
| : 2. 1 2 1 2 2 1 1 2 2 3 1 2 19 Plant 2 1 1 1 2 2 1 2 2 1 1 3 17 Systems 3 0 0 1 1 0 0 0 1 0 1 0 4 TierTotals 3 2 4 5 3 2 4 5 4 3 5 40
| |
| : 3. Generic Knowledge and Abilities Cat 1 Cat 2 Cat 3 Cat 4 4 5 3 5 17 Note:
| |
| * Attempt to distribute topics among all KIA categories; select at least one topic from every K/A category within each tier.
| |
| Actual point totals must match those specified in the table.
| |
| Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
| |
| Systems / evolutions within each group are identified on the associated outline. !
| |
| The shaded areas are not applicable to the category / tier.
| |
| /
| |
| (
| |
| NUREG-1021 1 of 8 Interim Rev. 8. January 1997 hD70
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| 7 0
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| 8 0
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| 1 5 6 7
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| 8 7 A U E 0 4
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| 0 4
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| 0 l
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| K 0 0 0 0 N
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| ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-5 4
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| Facility: DCPP Units 1/2 Date of Exam: January 25,1999 Exam Level: SRO Category K/A # Topic Imp I Points Conductof 2.1.2 Responsibilities of the Work Control Shift Fore; nan 4.0 1.0 Operations 2.1.7 Determine actions to recover RCS temperature following a transient 4.4 1.0 2.1.14 Changes in plant status that require notification of C&RP 3.3 1.0 2.1.33 Conditions requiring establishment of containment integrity 4.0 1.0 Total 4.0 Equipment 2.2.3 Reason for procedure difference based on unit design difference 3.3 1.0 Control 2.2.8 Formal commurications LBIE screen requirements 3.3 1.0 2.2.18 Determine allowed maintenance activity during an outage 3.6 1.0 2.2.25 Bases for 2 RCPs prior to exceeding 350*F 3.7 1.0 2.2.31 Responsibilities of the fuel handling SRO 3.8 1.0 Total 5C Radiation 2.3.4 Determining when contaminated (count rate) when using a frisker 3.1 Control 2.3.9 Effects of failed radiation monitor on containment purge 3.4 1.0 2.3.10 Actions required to minimize personnel exposure on VCT rupture 3.3 1.0 Total 3.0 Emergency 2.4.2 Actions required for Hi CTMT pressure while in FR-H.1 4.1 1.0 Procedures 2.4.6 Mitigation strategy for loss of all AC 4.0 1.0 and E-Plan 2.4.9 Mitigation strategy for inadequate decay heat removal 3.9 1.0 2.4.21 Determine CSFST monitoring frequency from SPDS parameters 4.3 1.0 2.4.40 E-plan responsibilities for the Interim Site Emergency Coordinator (ISEC) 4.0 1.0 Total 5.0 Tiar3 TargetPointTotal(SRO) 17 NUREG-1021 8 of 8 Interim Rev. 8, January 1997
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| ES-301 Administrative Topics Outline Form ES-301-1 i
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| Facility:- DCPP Units 1&2 Date of Examination: _ _ 01/25/99 Examination Level: __ SRO Operating Test Number: 1 Administrative Describe method of evaluation:
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| Topic / Subject . 1. ONE Administrative JPM, OR l Description 2. TWO Administrative Questions A.1 Conductof G2.1.23 (4.0) Ability to perform specific system integrated plant procedures Operations during allmodes of plantoperation.
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| (JPM) Reviewan(faulted)ECP G2.1.20 (4.2) Ability to execute procedure steps (Ques) Required steps when SFM suspects current procedure is inappropriate for plant conditions.
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| G2.1.12 (4.0) Ability to apply technical specifications for a system (Ques) Determination of equipmentinoperability A.2 Equipment Control G2.2.13 (3.8) Knowledge of tagging and clearance procedures (JPM) Review completed (faulted) clearance for FCV-95 A.3 Radiation Control G2.3.1 (3.0) Knowledge of 10CFR20 and related facility radiation control ,
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| requirements.
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| (Ques) Authorization for exceeding emergency exposure limits G2.3.4 (3.1) Knowledge of radiation exposure limits and contamination control including permissable levels in excess of those authorized.
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| (Ques) Interpret DCPP administrative radiation control requirements A.4 Emergency Plan G2.4.41 (4.1) Knowledge of the emergency action level thresholds and classifications (JPM) Classify a loss of decay heat removal NUREG-1021 21 OF 26 Interim Rev. 8, January 1997
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| \
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| , I ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: DCPP Units 1&2 Date of Examination: 01/25/99 Exam Level: _SR0(l)ISR0(U) Operating Test No.: 1 System /JPM Title / Type Codes
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| * Safety Planned Follow-up Questions:
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| Function K/A/G -Importance - Description
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| : 1. ECCS / Transfer to CL recirculation IV a. 005000K4.11 (3.9) Time to reach CL recirculation (D, L, S) Shutdown b. 005000A2.02 (3.7) RHR solid plant resp to PCV failure
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| : 2. Cont Spray / Manually initiate Containment V a. G2.1.33 (4.0) Determine system misalignment i Spray conditions fr entryinto TI 303 (D, A, L, S) Shutdown b. 026000K3.01 (4.1) Loss of CSP 1-1 effects on CCS
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| : 3. Instrument Respond to SCM LO/LO alarm Vil a. 017000K4.01 (3.7) Determine SCM with failed PT (M, S) 100% b. G2.1.12 (4.0) SCMM LCO requirements w/ failed PT l
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| : 4. RPS/ Manuallyisolate phase A 11 a. 013000K4.13 (3.9) Reseting MFW isolation after MSLI components (D, A, S) Shutdown b. 013000A3.01 (3.9) Effect of RPS testing on plant ops
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| : 5. Elect./ Crosstie vital bus G and H VI a. 062000A2.01 (3.9) Mode 5 elect bus operation l
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| (0, S) Shutdown b. 062000A1.01(3.8) D/Gloadlimit
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| : 6. Rod Controll Verify misaligned rod is not I a. 001000K5.01 (3.7) Determine bank overlap stuck (D, S) 100 % b. 001000A2.15 (4.2) Determine QPTR
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| : 7. PZR / Respond to PZR HI pressure ala m til a. 010000A1.09 (3.7) Determine tailpipe temperature (D, S) 100% b. 010000K1.02 (4.1) Effect of Sl on BU heater operation
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| : 8. AFW / Align a!!emate AFW from the FWST IV a. 061000K2.01 (3.3) Determine pwr supply to FCV-95 (M, R) Emergency b. 061000A1.04 (3.9) Determine time in Hot Standby
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| : 9. CCW/ Crosstie the CCW system Vill a. 008000A3.05 (3.1) Component isolation by phase B (N, R) Emergency b. G2.1.32 (3.8) CCWS LOCA response on loss of N2
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| : 10. EDG / Perform local start of a Diesel Gen. VI a. 064000K1.04 (3.9) D/G operability (D, A, P) Abnormal b. G2.4.30(3.6) Missed surveillance reportability
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| * Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)ltemate path, (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant,(R)CA NUREG-1021 22 of 26 Interirn Rev. 8, January 1997
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| Appendix D Scenario Outline F:at ES-D-1 l Facility: DCPP Units 1 & 2 Scenario No.: 1 Op-Test No.: Backup Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to an Eagle 21 system failure Evaluate the crew's ability to diagnose and respond to a letdown pressure controller failure Evaluate the crew's ability to diagnose and respond to a S/G Level Channel failure Evaluate the crew's ability to diagnose and respond to a RCP # 1 seal failure Evaluate the evaluate the crew in using E0Ps during a RCP seal failure LOCA Evaluate the crew's ability to respond to a partial P-4 actuation Evaluate the crew 's ability to diagnose and respond to a undercompensated IR channel Initial Conditions: 100% power, equilibrium xenon, Beginning of cycle (IC-1)
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| Tumover: 10 gpd leak on S/G 1-3, D/G 1-1 is 00S, CCW pump 1-3 is 00S __
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| MFW pp 1-2 Increased Vibrations - ramp unit to 50% for repair ASAP Time Event Malf. Event Event min No. No. Type
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| * Description 0 1 n/a N, RO Ramp power to 50% for repair of MFW pp 1-2 10 2 xmt I, BOP S/G Level Channel LT-549 fails HIGH mfw46 15 3 cnh cyc8 C, BOP Letdown pressure control valve PCV-135 fails CLOSED in auto, operatesin manual 20 4 ppl6a I, RO LCP Halt Protection Set 1, Rack 1 30 5 rep 1b C, RO RCP 12 # 1 sealfailure 35 6 ser 1244 RCP 1-2 # 1 High Vibration cond on 7 rep 2b, M ALL RCP 1-2 seal package failure.
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| trip rcp3b ppl5a C,SFM Rx Trip Breaker"RTA' fails to OPEN res3b Loop 2 RCS SB LOCA cond 8 nis 3b I, RO IR NI 36 is undercompensated (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
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| SCENARIO 01 OVERVIEW ,
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| The crew is directed to ramp to 50% power to remove MFP 1-2 from service (Event 1).
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| The RO will have to borate to ramp turbine and reactor power to 50%.
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| S/G Level transmitter LT-456 fails High (Event 2). The BOPCO and CO should identify the failed channel and report to the SFM. The SFM should refer to annunciator response procedure PK04-07 for a Protection Channel Activated and enter OP AP-5 for a Malfunction of Protection or Control channel. The SFM should infomi the RO and BOPCO of the affects of the failure on AFWlevel controlif needed.
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| Letdown pressure controller PCV-135 fails closed while in the auto position (Event 3).
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| BOPCO should identify the failure of PCV-135 to controlletdown pressure and report to the SFM. SFM should refer to PK04-21 for letdown press / flow. SFM should diract that the failed controller be placed in manual to regain letdown pressure control and ensure that the letdown relief valve re-seats.
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| Eagle 21 (LCP Halt) failure comes in (Event 4). The CO and BOPCO should identify the tadure and report to the SFM. The SFM should refer to annunciator response procedure PK06-01 ar.d enter OP AP-5. Crew should identify all instrumentation that is affected by the failure. LT-459 and PT-455 are failed as is and LT-459 and PT-455 should be deselected for control.
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| RCP 1-2 seal leakoff flow increases to 60 gpm (over 20 minutes) giving indications of a RCP 1-2 # 1 seal problem (Event 5). SFM should refer to annunciator response procedure PK05-02. RCP parameters should be monitored and Operations Management should be informed.
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| RCP 1-2 High Vibration (Event 6). SFM should refer to annunciator response procedure PK05-05. SFM should direct a Reactor Trip, followed by the tripping of RCP 1-2 and the subsequent isolation of the RCP 1-2 # 1 seal leakoff valve.
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| Following the transition to EOP E-0.1, RCP 1-2 seal package will fail, causing a 60gpm LOCA (Event 7), additionally Reactor Trip Breaker "A" fails to open. The LOCA results in decreasing pressurizer level accompanied by decreasing RCS pressure. The SFM should direct a manual SI prior to an automatic SI at 1850 psig and re-entry into EOP E-0 followed by a transition to EOP E-1. The SFM should direct the local action to open RTA.
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| Approximately 10 to 15 minutes following the reactor trip, the intermediate range NI should be low enough to energize the SR instruments except that NI-36 is undercompensated (Event 8) which results in only 1 of 2 intermediate Nis below P-6.
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| The RO/ BOP should recognize the failure of SR instruments to energize and the SFM should direct the manual re-energizing of the SR detectors.
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| The scenario will be terminated following diagnosis of the RCS not intact and a transition out of EOP E-1 to EOP E-1.2.
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| Appendix D Scenario Outline Forrn ES-D-1 1
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| Facility: DCPP Units 1 & 2 Scenario No.: 2 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to a RCS makeup control failure Evaluate the crew's ability to diagnose and respond to a Radiation monitor failure Evaluate the crew's ability to diagnose and respond to a Thermal barrier leak Evaluate the crew's ability to diagnose and respond to a NI42 Failure Evaluate the crew's ability to respond to a seismic event Evaluate the crew's ability to diagnose an ATWS and take timely actions Evaluate the crew's ability to diagnose and respond to a stuck open PORV Evaluate the crew in using the E0Ps during recovery from SB LOCA initial Conditions: 75% power, equilibrium xenon, end of cycle (IC-36)
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| Tumover: D/G 1-1 is 00S, for lube oil heater replacement Ramp to 100% power following repairs to #2 Htr Drip pump.
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| Time Event Malf. Event Event min No. No. Type
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| * Description 0 1 n/a N, SFM Ramp power to 100% following repairs to #2 Htr Drip pump R, RO 5 2 v!v cyc14 C, Ro FCV-111B VCT inlet from RCS makeup fails closed 10 3 xmtrms34 I, BOP instrument failure of RM44 results in High rad alarm and CVI 15 4 nis6b 1,ALL Instrument failure of NIS42 high 20 5 ccw3c C,ALL RCP 1-3 Thermal Barrier Leak (30 gpm) to CCW system 30 6 ppl5a M,ALL 0.35g Earthquake causes loss of startup power, Auto and manual ppl5b C, BOP Reactor trip blocked, PORV PCV456 fails open with no power to block valve.
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| (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
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| SCENARIO 02 OVERVIEW .
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| The crew is directed to ramp to 100 % power (Event 1), #2 Htr. Drip pump repairs are complete with the pump warmed up and running. The RO will have to dilute to ramp turbine and reactor power to 100%.
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| During the ramp to 100% power, FCi' 1118 will fail to open in auto or manual (Event 2) ;
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| causing RCS normal dilution flowpb to become inoperable. The SFM should refer to l l PK05-11 and enter AP-19 for malfunction of reactor control makeup system. SFM should direct local operators to investigate FCV-111B failure and direct the CO to use
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| ! Alternate Dilution for RCS makeup.
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| Containment Exhaust radiation monitor RM-44A will fail high, over-ranged (Event 3),
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| causing an inadvertent containment vent isolation (CVI). SFM should refer to the annunciator response procedure PK011-21, High Radiation and BOPCO should be sent to control room radiation monitors to locate the source of the high alarm condition.
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| SFM should refer to annunciator response procedure PK01-17 and determine that RM-12, Containment Gaseous Radiation Monitor is out of service and direct the BOPCO to place the CFCU drain collection system in service.
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| Nuclear Instrument NI-42 will fail high (Event 4), CO should report the failure to the SFM. BOPCO should be sent to Ni panels in the control room to investigate the failure.
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| SFM should enter OP AP-5 for the Nl failurr. sad determine that NI-42 can be defeated.
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| SFM should direct the BOPCO to defeat NI-42 using OP AP-5 Attachment 4.1. BOPCO should defeat NI-42 at the NI Racks.
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| Reactor coolant pump 1-3 thermal barrier leakage (Event 5), ramps from 0 to 30 gpm causing increase in CCW head tank level, Radiation monitors RM-17A & B alarm and close of the CCW head tank vent RCV-16. CO and BOPCO should report to the SFM.
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| The SFM should enter AP-1 for Excessive Reactor Coolant Leakage and directs CO and BOPCO in the determination of leak location. SFM enters AP-1; Malfunction of CCW System and directs control room and local actions to isolate thermal barrier for RCP-13.
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| Earthquake with a magnitude of 0.35g will occur (Event 5) causing the loss of startup power. An ATWS will exist due to the magnitude of the earthquake and the failure of the reactor to trip. SFM should direct the CO and BOPCO to perform the immediate actions for the ATWS and direct the opening of MG supply load center breakers to cause the l reactor to trip. During the event PORV 456 will open and stay open, power to the block l valve is not available, resulting in a SB LOCA. The steam space break results in
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| ]
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| increasing pressurizer level accompanied by decreasing RCS pressure. The SFM i should direct a m inual Si prior to an automatic Si at 1850 psig and re-entry into EOP j E-0 followed by a tsnsition to EOP E-1. l The scenario will be terminated following diagnosis of the RCS not intact and a transition out of EOP E-1 to EOP E-1.2. 1 1
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| 1
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| Appendix D Scenario Outline Forni ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 3 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respcad to a 10% dump valve failure Evaluate the crew's ability to diagnose and respond to a SJAE 7,ajiation inonitor failure Evaluate the crew's ability to diagnose and respond to a loss of DRPi indication Evaluate the crew's ability to diagnose and respond to a Loop 4 Delta T failure Evaluate the crew in using the AOP and EOP to respond to SG tube leak l Evaluate the crew's ability to diagnose and respond to a SG tube rupture i Evaluate the crew's ability to respond to a seismic event Evaluate the crew 's ability to diagnose and respond to a failure of the MSIVs to close l
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| InitialConditicns: 2% power, equilibrium xenon, end of cycle (10-40) l Turnover Continue startup at step 6.5 of L-3 l
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| l Timo Event Malf. Event Event min No. No. Type
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| * Description 0 1 n/a N, BOP Transfer House loads to startup power 5 2 xmt I,RO Loop 4 Delta T Fails high res142 10 3 ovr I, BOP SJAE RM-15R failure due to loss of power xdmiO33p 15 4,5 seit I,SFM 0.1g Earthquake causes loss of DRPI Indication and PY-15 dscrodi dsc eps29 20 6 mss 3 C, BOP PCV-20 fails full open, will close with backup air system 25 7 res4b M,ALL SG tubeleak 0-200 gpm over 20 mindes Cond 8 mss 7-10 C, BOP MSIVs fail to close from control room (N)ormal (R)eactivity (1)nstrument (C)omponent (M)ajor
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| l .
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| SCENARIO 03 OVERVIEW The crew is directed to continue the plant startup using OP L-3. First action for the crew will be for the BOPCO to transfer house loads to startup (Event 1) per step 6.5 of OP L-3.
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| During the transfer to startup, Loop 4 DeltaT fails High (Event 2) causing alarms associated with the failure. CO and BOPCO should report the failure to the SFM. The SFM should enter OP AP-5 for the failure and direct the CO to defeat the inputs to Delta T for the failed channel.
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| Steam Jet Air Ejector (SJAE) radiation monitor RM-15R will fail due to a loss of power (Event 3). CO and BOPCO should report to SFM the failure indications of the radiation monitor and the SFM should reference AR PK11-07 for the SJAE radiation monitor failure. The SFM should contact TM to determine the cause of the alarm and an operator to determine the status of the RM-15R input to SPDS.
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| Earthquake with a magnitude of 0.01g will occur which will result in normal supply breaker for DRPl tripping open, and the loss of all DRPI indications (Event 4) additionally PY-15 output bkr will fail open (Event 5). CO and BOPCO should report the loss of DRPI and the PY-15 to the SFM. The SFM should refer to annunciator response procedure PK03-21 and determine the Tech Spec 3.0.3 applies and directs local operations to place DRPI on backup power supply. SFM should refer to OP AP-4 for loss of PY-15 and direct actions to restore PY-15. SFM should enter CP M-4 in response to the earthquake.
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| PCV-20,10% steam dump valve will fail open (Event 6) causing Tavg to decrease and the 40% dump valves to modulate closed. The SFM should direct the BOPCO to manually close PCV-20 at the controller. when it is determined that the controller is failed, the SFM should direct the BOPCO to use the backup air source to close the PCV. The SFM should direct the CO to control TAVG using the 40% dumps once PCV-20 is closed.
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| Steam Generator 1-2 will develop a tube leak (Event 7). CO and BOPCO should identify the indications of the tube leak and report them to the SFM. SFM should enter and take actions per AP-3 for a Steam Generator Tube Failure. CO should determine when leakage is greater than 50 gpm and report to the SFM. SFM should direct a Manual Safety injection and the performance of E-0 immediate actions. SFM may direct the BOPCO to perform an early isolation of SG 1-2 using E-3, Steam Generator Tube Rupture.
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| While performing E-3 the BOPCO should recognize the failure of the MSIVs to close (Event 8). The failure should be reported to the SFM. The SFM should direct the actions of E-0 and transition to E-3. When it is determined the MSIVs will not close remotely, local operators should be dispatched to close the MSIVs locally. With isolation from intact SGs not possible the SFM should transition to ECA-3.1, SGTR with Loss of Coolant - Subcooled recovery desired.
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| Once the SFM transitions to ECA-3.1 the MSIVs will be closed and the scenario terminated.
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| Appendix D Scenario Outline Form ES-D-1 F
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| Facility: DCPP Units 1 & 2 Scenario No.: 4 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l _
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| Objectives: Evaluate the crew's ability to diagnose and respond to a Condensate system disturbance Evaluate the crew's ability to diagnose and respond to a Pressurizer master controller failure Evaluate the crew's ability to diagnose and respond to a ASW pump 0. C. Trip Evaluate the crew in using the AOP and E0P to respond to fullload rejection 1 Evaluate the crew's ability to diagnose and respond to a Loss of startup power Evaluate the crew's ability to diagnose and respond to a Loss of all AC event Evaluate the crew 's ability to diagnose and respond to a safety valve failure Initial Conditions: 100% power, equilibrium xenon, Beginning of cycle (IC-1)
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| Turnover: D/G 1-1 is 00S, for Lube oil heater replacement, CFCU 1-4 failed 10 minutes ago, TS have not been addressed Time Event Malf. Event Event min No. No. Type
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| * Description 0 1 none N, CO STP R-1 A Exercising Full Length Control Rods 12 2 cnd4 R, RO FCV-55 fails open 20 3 pn4 1, RO Pressurizer Master pressure controller HC-455K fails High - manual available 28 4 asw1 C, BOP ASW pp 1-10.C. trip also trips 4Kv F Bus on 0.C., no auto start on asw2 standby ASW pp1-2, can be started manually.
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| eps4c,1 33 5 syd1 M, ALL 100% full load rejection due to 500 Kv system disturbance 34 6 syd1 M,ALL Loss of startup power due to system disturbance degib C, BOP Diesel generator 1-2 fails to start, can be started manually 34 7 mss 6d C, BOP Main steam safety valve lead four RV-58 stuck open (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
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| t SCENARIO 04 OVERVIEW Control operator we begin performance of STP R-1 A (Event 1), Exercising full length control rods.
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| . Condensate system bypasa valve FCV-55 will fail open (Event 2). The RO and BOPCO should recognize the symptoms and identify the failure and inform the SFM of the transients affects on the plant. SFM should direct the BOPCO to attempt to close l FCV-55 from the control room. SFM should direct the CO to reduce turbine load to reduce reactor power and direct local isolation of FCV-55.
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| Pressurizer master controller HC-455K fails high (Event 3), causing pressurizer spray valves to open and pressurizer pressure to drop. CO should recognize the failure and inform the SFM. The SFM should enter. Annunciator Response procedure PK05-17 and transition to OP AP-13 Malfunction of Reactor Pressure control system. Pressurizer spray valves should be manually closed prior to a low pressure Reactor Trip. Pzr Master controller should be controlled in manual and spray valves retumed to auto.
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| The running Aux. Salt Water pump 1-1 trips on Overcurrent'(Event 4). The Overcuurent trip also causes an Overcurrent Trip on 4Kv bus F causing the loss of the 4Kv bus. The CO and BOPCO should recognize losses and report them to the SFM. The SFM should enter Annunciator Response procedure PK01-01 and OP AP-10 to restore the ASW system. The SFM should direct the CO/BOPCO to start the standby ASW pump.
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| The SFM should reference Annunciator Response procedure PK18-17 and PK18-22 and direct the actions for the failure of the 4Kv bus F loss including the start of CCW Pp.1-3 A 500 Kv system disturbance will cause a Full load rejection (Event 5). The CO and BOPCO should recognize loss and report to the SFM. SFM should reference Abnormal procedure OP AP-2 and direct the actions of the CO/BOPCO per the procedure. Plant will not survive the Load Rejection and Emergency procedure E-0 should be entered.
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| During performance of initial steps of E-0 Startup Power will be lost along with the a failure of DSL Gen 1-2 to start (Event 6). SFM should recognize the entry conditions for ECA 0.0 and direct the CO/BOPCO in actions for ! s of All AC Fower. SFM should
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| , direct the Manual starting of DSL Gen 1-2 and the restoration of AC power to Bus G.
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| SFM should determine that with one bus operable a transition to E-0 should be done.
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| During the initial Load rejection a Main Steam line safety valve will stick open (Event 7).
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| The BOPCO should determine that the 1-4 Steam generator is not intact and inform the SFM. The SFM should direct the BOPCO to perform an early isolation of SG 1-4 while the CO and the SFM perform the steps in E-0. SFM should determine the need to transition to EOP E-2 for the faulted SG. SFM should determine that a transition to EOP E-1 is required at the completion of EOP E-2.
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| 1 The scenario will be terminated following transition out of EOP E-1 to EOP E-1.1 SI Termination.
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| l i
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| l 1
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| 1 i
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| I
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| 'V i
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| 1 4
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| I j s REVISION 1 l
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| ES-301 Administrative Topics Outline Form ES-301-1 1
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| l i
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| Facility: DCPP Units 1&2 Date of Examination: 01/25/99 Examination Level: _ SRO Operating Test Number. 1 l Administrative Describe method of evaluation:
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| Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions l l
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| A.1 Conduct of G2.1.23 (4.0) Ability to perform specific system irtegrated plant procedures Operations during allmodes of plant operation. i I
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| (JPM ADMNRC-1) Review an (faulted) ECP l
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| G2.1.20 (4.2) Ability to execute procedure steps (Ques) Required steps when SFM suspects current procedure is inappropriate for plant conditions.
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| G2.1.12 (4.0) Ability to apply technical specifications for a system l
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| (Ques) Application of mode transition times and allowed out of service time (A0T) l A.2 Equipment Control G2.2.13 (3.8) Knowledge of tagging and clearance procedures (JPM ADMNRC-2) Review completed (faulted) clearance for FCV-95 l
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| l l
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| l A.3 Radiation Control G2.3.1 (3.0) Knowledge of 10CFR20 and related facility radiation control I
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| requirements.
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| (Ques) Authorizationforreceiving emergencyexposurelimits G2.3.4 (3.1) Knowledge of radiation exposure limits and contamination control including permissible levels in excess of those authorized.
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| (Ques) Interpret DCPP administrative radiation control requirements A.4 Emergency Plan G2.4.41 (4.1) Knowledge of the emergency action level thresholds and classifications (JPM ADMNRC-3) Classify a loss of decay heat removal NUREG-1021 21 OF 26 Interim Rev. 8, January 1997 h
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| NUCLEAR POWER GENERATION DtABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: ADMNRC-1
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| ==Title:==
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| REVIEW AN ESTIMATED CRITICAL POSITION Examinee:
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| Evaluator:
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| Print Signature Date Results: Sat Unsat Total Time: minutes Comments: Provide the " Student Copy" of the completed STP R-17 Attachment 8.2 for review.
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| ==References:==
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| STP R-17, Estimated Critical Position, Rev.12A Vol. 9B: Fig. R17-lF-3, Rev. I1; Fig. R17-lT-2, Rev.13; Fig. R17-IT-3, Rev. 8; Fig. R4-1F-4, Rev. 8; R17-1F-2, Rev. 8; R17-1T-4, Rev. 9 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 25 minutes I
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| Critical Steps: 2 - 9, and 11 - 18 !
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| Job Designation: SRO K/A Number G2.1.23 (4.0) i AUTHOR: JOHN BECERRA DATE: 11/17/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR i
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| APPROVED BY: NA DATE: NA TRAINING LEADER REv.O l
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| I JPM TITLE: REVIEW AN ESTNATED CRITICAL PoSmoe JPM NUMBER: ADMNRC-1 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the I
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| performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin. l Required Materials: Calculator and a complete STP R-17, Attachment 8.2 (ECP).
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| Initial Conditions: Unit 1 tripped from 100% power 28 hours ago and is in HOT STANDBY with shutdown banks withdrawn. Core burnup is 6000 l MWD /MTU Initiating Cue: Reactor Engineering delivered an ECP to the control room. A reactor i startup is scheduled for two hours from now. You are to review and approve the ECP.
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| Task Standard: The ECP is reviewed and approved after finding and correcting the three ;
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| (3) technical errors.
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| NOTE: Recalculations due to errors are not considered " technical" errors. l ADMNRC-1 PAGE 2 OF 8 REV.0 1
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| JPM TITLE: REVIEW AN ESTIMATED CRmCAL PosmON JPM MUMBER:- ADMNRC-1 INSTRUCTOR WORKSHEET Start Time:
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| Step Expected Operator Actions
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| )
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| : 1. Obtain the correct procedure. 1.1 Operator references STP R-17 and Volume 9.
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| Step was: Sat: Unsat *
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| : 2. ** Check reference power defect. 2.1 Operator verifies the reference power defect to be +1165 pcm from Table 1 R17-lF 3. !
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| [ \
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| Step was: Sat: Unsat !
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| : 3. *
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| * Check refere e xenon worth. 3.1 Operator verifies the reference xenon worth to be -2901 pcm from Table R17-lT-2.
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| Step was: Sat: Unsat
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| : 4. ** Check estimated xenon worth at 4.1 Operator verifies the estimated xenon criticality, worth to be -1885 pcm frem Table R17-lT-2.
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| Step was: Sat: Unsat
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| : 5. *
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| * Calculate net change in xenon worth. 5.1 Operator calculates the net change in xenon worth to be +1016 pcm.
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| 5.2 Determines Step C.4 in error due to incorrect sign.
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| Step was: Sat: Unsat
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| : 6. *
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| * Check reference samaritun worth. 6.1 Operator determines the reference samarium worth to be -943 from Table R17-1T-3 by interpolation.
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| Step was: Sat: Unsat
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| * Denotes an entry required on the JPM cover sheet.
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| ** Denotes a Critical Step.
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| ADMNRC-1 PAGE 3 OF 8 REV.O
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| JPM TITLE: REVIEW AN ESTIMATED CRmCAL Pof., mon JPM NUMBER:- ADMNRC-1 INSTRUCTOR WORKSHEET
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| : 7. *
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| * Check estimated samarium worth at 7.1 Operator determines the estimated criticality. samarium worth to be 1069 pcm ,
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| from Table RI?-lT-3 by l interpolation.
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| I 7.2 Determines Step C.6 in error due to transposition error (1096 vice 1069).
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| l:
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| Step was: Sat: Unsat i l
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| : 8. ** Calculate net change in samarium 8.1 Operator recalculates the net change I worth. in samarium worth to be -126 pcm. I I
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| Step was: Sat: Unsat i
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| : 9. *
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| * Check reference integral rod worth. 9.1 Operator verifies the reference !
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| integral rod worth to be 0 to -5 pcm from Figure R4-lF-4.
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| Note: Operator niust use Figure R4-1F-4 for hot full power reference conditions vs. Table R17-1T-4, which is for hot zero power.
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| Step was: Sat: Unsat ,
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| * I i
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| : 10. Check critical rod position. 10.1 Operator agrees with a critical rod !
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| position of control bank D at 150 steps.
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| Step was: Sat: Unsat i j
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| 1 l
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| ADMNRC-1 PAGE4OF8 REV.O
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| JPM T.TLE- REVIEW AN ESTIMATED CRITICAL POSITION JPM NUMBER:- ADMNRC-1 INSTRUCTOR WORKSHEET
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| : 11. " Check integral rod worth. I1.1 Operator verifies the integral rod worth for control bank D at 150 steps to be -202 pcm from Table R17-lT-4.
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| Note: Operator must use Ta' ole ,
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| R17-1T-4 for this step and not Figure R4-1F-4 which was used in Step 9.
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| I 1.2 Determines Step C.10 in error due to misread table.
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| Step was: Sat: Unsat
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| : 12. ** Calculate net change in integral rod 12.1 Opemtor recalculates the net change worth. in integral rod worth to be -197 pcm. l Step was: Sat: Unr.at
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| : 13. ** Calculate the net reactivity changes. 13.1 Operator recalculates the sum of the ,
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| net reactivity changes to be +1858. I Step was: Sat: _
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| Unsat
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| : 14. " Calculate preliminary change in 14.1 Operator recalculates preliminary boron concentration. change in boron concentration to be
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| +206 ppm.
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| Step was: Sat: Unsat
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| : 15. " Calculate preliminary boron 15.1 Operator recalculates the preliminary concentration. boron concentration to be 1506 ppm.
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| Step was: Sat: Unat *
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| ADMNRC-1 PAGE 5 OF 8 REV.0
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| JPM TITLE: REVIEW AN ESTIMATED CRmCAL PosmON JPM NUMBER:- ADMNRC-1
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| < INSTRUCTOR WORKSHEET
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| : 16. ** Check actual differential boron 16.1 Operator determines the actual i worth, differential boron worth to be -7.7 I pcm/ ppm from Figure R17-IF-2.
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| Note: -7.7 pcm/ ppm is based on 6,000 MWD /MTU. Thus the value is l i
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| taken 1/4 of the way between the HOL and EOL curves.
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| Step was: Sat: Unsat 1
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| : 17. ** Calculate the boron concentration 17.1 Operator recalculates the estimated change to offset the net reactivity change in boron concentration to be change. +241 ppm.
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| Unsat
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| * Step was: Sat:
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| : 18. *
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| * Calculate the estimated critical boron 18.1 Operator recalculates the estimated concentration, critical boron concentration to be 1541 ppm.
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| 18.2 Agrees with the estimated critical boron concentration calculated by Reactor Engineering.
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| Step was: Sat: Unsat
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| : 19. Analyze acceptance criteria. 19.1 Operator checks the YES box for rod insertion limit criteria.
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| 19.2 Checks the YES box for withdrawal limits.
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| Step was: Sat: Unsat
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| : 20. Complete Shift Foreman review. 20.1 Operator completes Item A.6.
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| 20.2 Signs name and enters the date and time.
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| Step was: Sat: Unsat Stop Time:
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| Total Time: (Enter total time on the cover page)
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| ADMNRC-1 PAGE 6 OF 8 REV.O
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| 7-l JPM TITLE: JPM NUMBER: ADMNRC-1 i- EXAMINES CUE SHEET i
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| i l Initial Conditions: Unit 1 tripped from 100% power 28 hours ago and is in HOT j STANDBY with shutdown banks withdrawn. Core burnup is 6000
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| ! MWD /MTU Initiating Cue: Reactor Engineering delivered an ECP to the control room. A reactor l startup is scheduled for two hours from now. You are to review and approve the ECP.
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| l Task Standard: The ECP is reviewed and approved after finding and co Tecting the three (3) technical errors.
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| NOTE: Recalculations due to errors are not considered " technical" errors.
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| 1 l
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| l l
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| 1 l
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| ADMNPC-1 PAGE7OF8 REV.O
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| i JPM TITLE: REVIEW AN ESTIMATED CRITICAL POSmON JPM NUMBER: ADMNRC-1 1 ATTACHMENT 1, SIMULATOR SETUP O The simulator is rot needed for the performance of this JPM.
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| i l
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| l l
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| l l
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| l 1
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| REv.0 ADMNRC-1 PAGE 8 OF 8 I
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| - Page1of1 DIABLO CANYON POWER PLANT STP R-17 ATTACHMENT 8.2 AND TITLE: MANUAL CALCULATION OF ESTIMATED CRITICAL POSITION I A. GENERAllNFORMATION 4 EXPECTE0 0ATEITIME CRITICAL 01f27199 I 1235
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| : 1. UNIT 1 CYCLE 9 STARTUP NO._p9 01 5. LENGTH 0F SHUTDOWN 30 HRS
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| : 2. BURNUP (MWDlMTV) 6,000 6. DATEITIME THIS ECP EXPIRES I (ADO 4 HRS TO STEP A.4 OR SFM REVIEW, WHICHEVER EARLIER)
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| : 3. SHUT 00WN DATEITIME 01126199___l___0635 B. REFERENCE CRITICAL CONDITIONS 0. CALCULATION
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| : 1. DATE 01126199___ TIME 0635 1. SUM 0F NET REACTIVITY CHANGES
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| : 2. R00 POSITION: BANK D AT 222 STEPS (C.1 + C.4 + C.7 + C.11) (-) 239 PCM
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| : 3. BORON CONCENTRATION 1300 PPM PREllMINARY ECP
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| : 4. POWER LEVEL 100 PERCENT 2. BORON CONCENTRATION CHANGE TO OFFSET NET REACTIVITY CHANGE (-(0.1)+ 9 PCMlPPM) (+) 27 PPM NOTE: PD Xe & Sm blocks are N/A for initial criticality. 3. PRELIMINARY CRITICAL BORON C. NET REACTIVITY CHANGES CONCENTRATION (B.3 + D.2) 1327 PPM POWER DEFECT N/A [ l
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| : 1. NET CHANGE - O . REFERENCE PWR DEF FROM R171F 3(R17 2F 3) (+) 1165 PCM XENON N/A [ l FINAL ECP
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| : 2. RGERENCE XENON WORTH FROM 4. ACTUAL DIFFERENTIAL BORON WORTH L171T 2(R17 2T 2s (T-0) (-) 190,LPCM FOR CONCENTRATMN D.3. FROM
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| : 3. ESTIMATE 0 XENEN WORTH AT R171F 2(R17 2F 2) (-)_Q PCM! PPM CRIY!CALITY FFiOM R171T 2
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| : 5. BORON CONCENTRATION CHANGE (R17 2T 2) (T- A.5) (-) j8,05_PCM TO OFFSET h... REACTIVITY
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| : 4. NET CHANGE IN XENON WORTH CHANGE ( + 0.4) (+)_LO_ PPM (C.3 C.2) (-) 1016 PCM
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| : 6. ESTIMATED t tilTICAL BORON SAMARIUM N/A [ l CONCENTRATION (B.3 + 0.5) 1330 PPM
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| : 5. REFERENCE SAMARIUM PWMy She (PPE)
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| REACTIVITY FROM R171T 3 (-) 943 PCM (R17 2T 3) (T-0) Reviewed by Shnstre (SPPE)
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| : 6. ESTIMATED SAMARIUM WORTH AT CRITICALITY FROM R171T 3 (- LLO96,,PCM E. ACCEPTANCE CRITERIA YES N0 IWA (R17 2T 3) (T- A.5) 1. ECP MEETS R00 INSERTION LIMIT
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| : 7. NET CHANGE IN SAMARIUM CRITiRIA 0F T.S 3.1.3.6 WITHIN WORTH (C.6 C.5) (-1 153 PCM 4 HRS OF CRITICAlfTY. (x! !l []
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| : 2. ECP WITHIN WITH0RAWAL LIMITS OF CONTROL RODS N/A l l OP L 2 (T.S. 3.1.1.3) WITHIN
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| : 8. REFERENCE INTEGRAL P00 4 HRS OF CRITICALITY. (1 [] []
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| WORTH FROM R41F4 or (-) 5 PCM SFM REVIEW:
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| R171T4 (R4-2F4 or R17 2T4) (NOTE:SFM COMPLETE A.6 00 NOT G0 CRITICAL AFTER THIS TIME)
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| : 9. CRITICAL CONTROL R00 POSITION D 9150 STEPS SFM Shnstre DATEITIME dite I time
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| : 10. INTEGRAL R00 WORTH FOR POSITION SPECIFIED IN C.9. (-) 240 PCM F. ACTUAL CRITICAL POSITION (DATA TAKEN AT 10' AMPS)
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| FROM R171T4 (R17 2T4) DATE TIME Tm
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| : 11. NET CHANGE IN R00 WORTH BORON CONCENTRATION PPM (C.10 . C.8) (-) 235 PCM CONTROL R00 POSITION STEPS DEVIATl0N: Ap rods - Ap baron - Ap Xe/Sm - SUM - pcm.
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| REMARKS:
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| STUDENT COPY
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| I Page1of1 I DIABLO CANYON POWE1: *NT STP R-17 l ATTACHMENT 8.2 AND l TITLE: MANUAL CALCULATION OF ESTIMATED CRITICAL POSITION A. GENERAL INFORMATION 4 EXPECTED OATEITIME CRITICAL ___01l27199 l_ 1235 1 UMT 1 CYCLE 9 STARTUP NO. 99 01 5. LENGTH 0F SHUTDOWN 30 HRS l
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| : 2. BURNUP (MWDIMTU) 6.000 6. DATEITIME THIS ECP EXPlRES _,0jl27199_ _l__ _1635 (ADD 4 HRS TO STEP A.4 OR SFM REVIEW, WHICHEVER EARI.lER)
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| : 3. SHUTOOWN DATElTIME 01f26!99 I 0635 )
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| B. REFERENCE CRITICAL CONDITIONS D. CALCULATION j i
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| : 1. DATE 01126I99 TIME 0635 1. SUM OF NET REACTIVITY CHANG
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| : 2. RDO POS!T! Oft BANK D AT 222 STEPS (C.1 + C.4 + C.7 + C.11) + ) 1858 PCM
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| : 3. BORON CONCENTRATION 1300 PPM PRFLIMINARY ECP
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| : 4. POWER LEVEL _100 PERCENT 2. BORON CONCENTRA H TO OFFSET NET R CHANGE (- (0.1) ) (+) 206 PPM 4 NOTE: PD, Xe & Sm blocks are N/A for initial criticality. S. PRELIMINA l C. NET REACTMTY CHANGES CONCENTR N (B. 1506 PPM I POWER DEFECT N/A ! ]
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| : 1. NET CHANGE - 0. REFERENCE PWR DEF FROM R171F 3(R17 2F-3) (+) 1165 PCM l XENON N/A l l M CP
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| : 2. REFERENCE XENON WORTH FROM . L DIFFERENTIAL BORON WORTH R171T 2(R17 2T 2) (T-0) (-) 2901 PCM NTRATl0N D.3. FROM
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| : 3. ESTIMATED XENON WORTH AT 171F 2(R17 2F 2) (-), 7.7,_ PCMIPPM CRITICALITY FROM R171T 2 RON CONCENTRATION CHANGE gg pug (R17 2T 2) (T- A.5) (-)J8_8,5_P TO OFFSET NET REACTIVITY l I ce4tREg4. NU CHANGE IN XENON WORTH p
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| - 6. ESTIMATED CRITICAL BORON SAMARIUM i- CONCENTRATION (B.3 + D.5) 1541 PPM
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| : 5. REFERENCE SAMARIUM Performed by Sbnstre (PPE)
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| REACTMTY FROM R171T 3 (R17 2T 3) (T-9) Reviewed by Sbnsttre (SPPE) gggcg puE 6. ESTIMATE 0 SAMARIUM 79%/Sfb51TD( AT CRITICALITY FROM R1 9 PCM E. ACCEPTANCE CRITERIA HS NO N/A (R17 2T 3) (T* A. 1. FrP MEETS ROD INSERTION LIMIT
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| : 7. NET CHANGE IN S g;ITERIA 0F T.S. 3.1.3.6 WITHIN WORTH (Cfar.51 (-)_,1.[6_PCM 4 HRS OF CRITICALITY. Ixl t1 (l
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| . 2. ECP WITHIN WITHDRAWAlLIMITS OF 1 CONTROL RODS N/Al l OP L 2 (T.S. 3.1.1.3) WITHIN
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| : 8. REFERENCEI R3 4 HRS OF CRITICALITY. l:] ll l1 WORTH FROM F4 or (-) 5_ PCM SFM REYH-R171T4 (R4-2F r R17 2T4) (NOTE:SFM C0lWrtETE A.6 DO NOT G0 CRITICAL AFTER THIS TIME)
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| : 5. CRITICAL CONTROL ROD POSITION _D 9150 STEPS SFM Sbnstre DATEITIME - - - -/ste - -i- -time FER0R M 10. INTEGRALRODWORTH FOR p A48+ M WY POSITl0N SPECIFIED IN C.9. (-! 202 PCM F. ACTUAL CRITICAL POSITION (DATA TAKEN AT 10* AMPS)
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| TMAS FROM R17 IT4 (R17 2T4) DATE TIME Tm
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| : 11. NET CHANGE IN R00 WORTH BORON CONCENTRATION PPM (C.10 C.8) (-) 197 PCM CONTROL ROD, POSITION STEPS DEVIATION: Ap rods - Ap boron - Ap XefSm - SUM. pcm.
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| REMARKS:
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| R17 key. doc
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| Sheet 1 of 3 *-
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| l DIABLO CANYON POWER PLANT OPERATION DATA i FIGURE R17-1F-3 '
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| Total Power Defect as a Function of Power Level at BOL Cycle 9 for Burnup 0 - 8000 MWD /MTU l
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| 1400 l l l l l j 1000 ppm i
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| 1200 #
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| / , 1300 ppm
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| / /
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| / / , 1600 ppm 1000 4 p f / / 1900 ppm f / / // /
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| s / / /
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| 800 >
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| G / // /
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| 600 # ' #
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| f 0/$
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| 0 10 20 30 40 50 60 70 80 90 100 POWER LEVEL (%)
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| SOURCE: WCAP-14855, Rev 0, Figure D.4-1 Page ID-7 Revision 11 Date 05/22/97
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| Sheet 'l of 2 8
| |
| DIABLO CANYON POWER PLANT OPERATION DATA FIGURE R41F-4 Diberential and integral Rod Worth vs. Steps Withdrawn, Banks D, C, B and A Moving with 100 Step Overlap, at BOL, HFP, Equilibrium Xenon Cycle 9 for Burnup < 12200 MWD /MTU 17.5 3500 s %
| |
| 3000 \ %
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| 15.0,g a3 4
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| h 2.5 N i N i 0 M 0.0 0 100 200 ROD BANK POSITION (Steps Withdrawn) .
| |
| BANK D 10 1100 1228 l
| |
| BANK C lo 1100 1228 !
| |
| BANKB lo n 00 1228 !
| |
| BANK Alo 1100 1228 Note: Boron Concentration Held Constant at the HFP ARo Value. ARo Shown as 228 (Nominal) but can be 1225 and 5231 Steps Withdrawn SOURCE: WCAP.14855, Rev 0, Figure 61 ,
| |
| 1 I
| |
| l l
| |
| Page 18-4 Revision 8 Date 05/22/97
| |
| | |
| I 1
| |
| DIABLO CANYON POWER PLANT OPERATION DATA TABLE R17-1T-4 BOL AND EOL, HZP INTEGRAL WORTH AS A FUNCTION OF STEPS WITHDRAWN UN FOR BANKS D, C, AND B WITH 100 STEP OVERLAP CYCLE 9
| |
| - (Data Plotted on Figures A-4 and A-5) 0- 12200 MWD /MTU 12000-24490 MWD /MTU BOL, HZP EOL, HZP Steps Withdrawn Figure A-4 Figure A 5 Bank Bank Bank int. Rod Worth int. Rod Worth B C D (pcm) (pCm) 228 228 228 0.0 0.0 228 228 220 -3.8 -25.0 228 228 210 -17.2 -115.6 228 228 200 -39.3 -253.1 228 228 190 -67.0 -393.9 228 228 180 -98.0 -516.3 228 228 170 -131.0 -617.9 228 228 160 -165.7 -702.0 228 228 150 -202.0 -771.8 228 228 140 -240.0 -829.5 228 228 130 -279.8 -876.2 228 228 120 -321.5 -912.8 228 228 110 -364.9 -940.3 228 228 100 -409.8 -961.0 228 218 90 -457.7 -1001.7 228 208 80 510.3 -1088.1 228 198 70 -567.2 -1203.1 228 188 60 -627.4 -1316.7 228 178 50 -689.2 -1416.2 228 168 40 -751.3 -1502.3 228 158 30 -812.3 -1576.4 228 148 20 -871.3 -1638.3 228 138 10 -927.1 -1687.1 228 128 0 -979.4 -1723.7 228 120 0 1017.5 -1744.9 228 110 0 -1063.2 -1763.7 228 100 0 -1110.9 -1777.5 218 90 0 -1161.8 -1809.4 208 80 0 -1216.7 1880.8 198 70 0 -1275.4 -1977.6 188 60 0 -1336.7 -2074.0 178 50 0 -1398.9 -2157.2 Note: Boron Concentration Changed to Maintain Critical Conditions as Rods are Moved. ARO shown as 228 (nominal) but can be 3225 and s231 steps withdrawn.
| |
| SOURCE: WCAP 14855. Rev 0, Figure A-4. Figure A-5 Page 18-5 Revision 9 Date 05/22/97
| |
| | |
| T l DIABLO CANYON POWER PLANT OPERATION DATA l FIGURE R17-1F-2 l
| |
| Differential Boron Worth vs. Boron Concentration at HZP, BOL, No Xenon, l Peak Samarium, and EOL with HFP Equilibrium Xenon, Equilibrium Samarium Cycle 9
| |
| -6.5 I
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| 0 200 400 600 800 1000 1200 1400 1600 1800 2000 BORON CONCENTRATION (PPM) 1 SOURCE: WCAP-14855, Rev 0, Figure A-7 Page IA-9 Revision 8 Date 05/22/97
| |
| | |
| NUCLEAR POWER GENERATION ,
| |
| DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: ADMNRC-2
| |
| | |
| ==Title:==
| |
| REVIEW A CLEARANCE FOR TECHNICAL ACCURACY Examinee: ,
| |
| l Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
| |
| l l
| |
| 1
| |
| | |
| ==References:==
| |
| Operator Valve Identification Diagram,106704, Sheet 4 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2 1
| |
| Job Designation: SRO l K/A Numher G2.2.13 (3.8) l AUTHOR: JOHN BECERRA DATE: 11/11/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REv.0
| |
| | |
| [ ]
| |
| JP.'A TITLE: REVIEW A CLEARANCE FOR TECHNICAL ACCURACY JPM NUMBER: ADMNRC-2 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are l needed. The examinee may be given the procedure and told the step l
| |
| with which to begin. ,
| |
| l Required Materials: Clearance NBR: 00034803, " Install and Remove Test Equipment on FCV-95 Initial Conditions: Unit 1 is 100% power, steady state.
| |
| Initiating Cue: Perform a review of all clearance points on Clearance NBR: 00034803, l Install and Remove Test Equipment on FCV-95, for technical errors. l Task Standard: Find and correct three (3) technical clearance errors within the clearance points.
| |
| 1 l
| |
| i l
| |
| ADMNRC-2 PAGE 2 OF 5 REV.O
| |
| | |
| p t
| |
| JPM TITLE: REVIEW A CLEARANCE FOR TECHNICAL ACCURACY JPM NUMBER: ADMNRC-2 INSTRUCTOR WORKSHEET Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. Obrain the correct reference material. Cue: Provide operator with clearance sheet.
| |
| 1.1 Operator selects Operator Valve Identification Diagram (OVID).
| |
| 1.2 Selects Section 106704, Sheet 4.
| |
| Note: Operator may obtain valve and breaker number using optional .
| |
| l reference material.
| |
| Note: Clearance Legend TAGS:
| |
| MAN-ON-LINE (M) i CAUTION (C)
| |
| CONTROL BOARD (CBC)
| |
| POSITIONS:
| |
| RACKED OUT (RO)
| |
| RACKED IN (RI)
| |
| CLOSED (CL)
| |
| OPEN (OP)
| |
| OPEN & UNCAPPED (OPU)
| |
| Step was: Sat: Unsat *
| |
| ' Denotes an entry required on the JPM cover sheet.
| |
| " Denotes a Critical Step.
| |
| ADMNRC-2 PAGE 3 OF 5 REV.O
| |
| | |
| JPM TITLE: REVIEW A CLEARANCE Fon TECHNICAL ACCURACY JPM NUMBER: ADMNRC-2 INSTRUCTOR WORKSHEET I
| |
| ! 2. **1dentify MS-1-FCV-95 clearance 2.1 Determines point #4 tag should be errors. MAN-ON-LINE (M) tag.
| |
| 2.2 Determines point #5 breaker should be 52-1F-30.
| |
| 2.3 Determines point #7 valve (MS FCV-37) position should be closed.
| |
| Step was: Sat: Unsat
| |
| * Stop Time: i 1
| |
| Total Time: (Enter total time on the cover page) l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| ADMNRC-2 PAGE 4 OF 5 REV.O
| |
| | |
| 1 JPM TITLE: JPM NUMBER: ADMNRC-2 EXAMINEE CUE SHEET t Initial Conditions: Unit 1 is 100% power, steady state.
| |
| )
| |
| Initiating Cue: Perform a review of all clearance points on Clearance NBR: 00034803, Install and Remove Test Equipment on FCV-95, for technical errors.
| |
| l Task Standard: Find and correct three (3) technical clearance errors within the clearance i points. !
| |
| l I
| |
| i i 1
| |
| i
| |
| ,i i
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| I ADMNRC-2 PAGE 5 OF 5 REV.O
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| |
| | |
| : NUCLEAR POWER GENERATION .r DIABLO CANYON POWER PLANT JOs PERFORMANCE MEASURE Number: ADMNRC-3
| |
| | |
| ==Title:==
| |
| CLASSIFY A LOSS OF DECAY HEAT REMOVAL Examinee:
| |
| Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
| |
| | |
| ==References:==
| |
| OP AP SD-0, Rev. 7A EP G-1, Accident Classification and Emergency Plan Activation, Rev.28 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 5 minutes Critical Steps: 2,3,4,5,6 Job Designation: SRO K/A Number: G2.4.41 (4.1)
| |
| AUTHOR: JOHN BECERRA DATE: 11/11/98 l
| |
| REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| I l JPM TITLE: CMSSIFY A LOSS OF DECAY HEAT REMOVAL JPM NUMBER: ADMNRC-3 l INSTRUCTOR WORKSHEET i
| |
| Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by l the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are ,
| |
| needed. The examinee may be given the procedure and told the step I l with which to begin.
| |
| Required Materials: Calculator Initial Conditions: Unit 1 is in Mode 5,27 days into a refueling outcge. Core reload has been completed, replacing 1/3 of the core with new fuel. Completed tensioning the RX vessel head while filling the RCS per OP A-2:lli with level @ 114". A common mode failure on RHR pp breakers caused both RHR pps to trip 15 minutes ago. Pumps were shutdown 15 minutes ago. Core exit thermocouple readings are presently 110 F.
| |
| Initiating Cue: The Shift Supervisor has directed you to classify the event and to report the event classification to him.
| |
| Task Standard: Determine the proper event classification and report the classification to the Shift Supervisor, i
| |
| l I
| |
| i l
| |
| ADMNRC-3 PAGE2OF6 REV.O
| |
| | |
| l l
| |
| . JPM TITLE: CLASSIFY A LOSS OF DECAY HEAT REMOVAL JPM NUMBER: ADMNRC-3 lNSTRUCTOR WORKSHEET l
| |
| I l
| |
| Start Time.
| |
| Expected Operator Actions Step
| |
| : 1. Obtain correct procedure. Note: Operator may reference EP G-1 prior to performing Appendix B.
| |
| 1.1 Operator references OP AP SD series procedures, Appendix B.
| |
| Note: All six of the OP AP SD series procedures contain Appendix B.
| |
| Step was: Sat: Unsat
| |
| : 2. * *Detennine the predicted heat load. 2.1 Operator estimates heat load from Appendix B curve for 27 days after shutdown to be 6 MW (10.5 MW).
| |
| Step was: Sat: Unsat
| |
| : 3. *
| |
| * Determine the reduction factor for 3.1 Operator uses predicted heat load refueled cores. from previous step.
| |
| 3.2 Uses 0.67 as the fraction of previously used assemblies installed in the core (0.66 to 0.67).
| |
| 3.3 Calculates estimated decay heat load to be 4MW (3.6 - 4.4).
| |
| Step was: Sat: Unsat
| |
| : 4. *
| |
| * Predict the heat up rate. 4.1 Operator uses estimated decay heat load from previous step.
| |
| 4.2 Determines the inventory factor to be 0.33.
| |
| 4.3 Calculates predicted heat up rate to be 1.32 F/ min (1.22 to 1.42 F/ minute)
| |
| Step was: Sat: Unsat l
| |
| * Denotes an entry required on the JPM cover sheet. ,
| |
| *
| |
| * Denotes a Critical Step. l ADMNRC-3 PAGE 3 OF 6 REV. 0 l 1
| |
| | |
| 1
| |
| : ' JPM TITLE: CLASSIFY A 1.OSS OF DECAY HEnT REMOVAL JPM NUMBER: ADMNRC-3 lNSTRUCTOR WORKSHEET l
| |
| : 5. *
| |
| * Calculate estimated time to reach 5.1 Operator calculate urrent l 200'F. temperature difference to be 90 F. !
| |
| 1 \
| |
| 5.2 Calculates time to reach 200 F to be 68 min.
| |
| Step was: Sat: Unsat
| |
| : 6. *
| |
| * Classify the event in accordance with 6.1 Operator determines that RCS EP G-1. thermocouple temperature will not exceed 200 F within one hour of RHR loss.
| |
| l 6.2 Determine loss of RHR <1 hr. but >l5 min.
| |
| 6.3 Classifies the event as an NUE #10 in accordance with EP G-1.
| |
| i Step was: Sat: Unsat Stop Time:
| |
| Total Time: (Enter total time on the cover page) i
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| ADMNRC-3 PAGE 4 OF 6 REV.O
| |
| | |
| JPM TITLE: JPM NUMBER: ADMNRC-3 EXAMINEE CUE SHEET j 1
| |
| 1 Initial Conditions: Unit 1 is in Mode 5,27 days into a refueling outage. Core reload has been completed, replacing 1/3 of the core with new fuel. Completed tensioning the RX vessel head while filling the RCS per OP A-2:III with level @ l14". A common mode failure on RHR pp breakers caused both RHR pps to trip 15 minutes ago. Pumps were shutdown 15 minutes ago. Core exit thermocouple readings are presently 110 F.
| |
| 1 1
| |
| Initiating Cue: The Shift Supervisor has directed you to classify the event and to report i the event classification to him. l Task Standard: Determine the proper event classification and report the classification to the Shift Supervisor.
| |
| l l
| |
| ADMNRC-3 PAGE 5 OF 6 REV.O
| |
| | |
| I i JPM TITLE: CLASSIFY A LOSS OF DECAY HEAT REMOVAL' JPM NUMBER: ADMNRC-3 ATTACHMENT 1, SIMULATOR SETUP O The simulator is not nceded for the performance of this JPM.
| |
| l t
| |
| i T
| |
| '1 ADMNRC-3 PAGE 6 OF 6 REV.0 l 1
| |
| l J
| |
| | |
| l PACIFIC GAS AND ELECTRIC COMPANY NIMBER OP AP' SD-0 DIABLO CANYON FOWER PUMr REVISION 7A PAGE 8 OF 38 TITLE: LOSS OF, ORINADEQUATE DECAY HEAT UNITS 1AND2 REMOVAL i
| |
| 4 APPENDIX B (Continued)
| |
| : 3. HEAT UP RATE PREDICTION MW X =
| |
| Degrees per Estimated Inventory Predicted Minute Decay Heat Factor Heat Up ,
| |
| Load Rate l 3.1 INVENTORY FACTOR - Degrees /MW Min ,
| |
| 107' O.52 108' O.45 Nozzle bams Installed.QB EQ Nozzle Dams Installed A@
| |
| SG Tubes Voided SG Tubes Not Voided 112' O.36 0.29 114' O.33 0.27 116' O.31 0.26 2 118' O.31 0.054 Upper Internals Removed (Use 2118' if Upper Imam =1= Installed) 120' O.06 130' O.03 138' O.02 <
| |
| : 4. ESTIMATED TIME TO REACH 200 DEGREES i
| |
| 200 Delta Tenp l Existing Temperature + Actual or =
| |
| Predicted Minutes to
| |
| _, Delta Temp Ileat Up Rate reach 200 {
| |
| l l
| |
| 01068807. DOE C2 8 F709.0927
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| |
| pS e n B E oh cd o iar in g e y nh r A
| |
| B V e igwt o E 4. T eM t lh Ri O ha es n
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| |
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| |
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| |
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| |
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| |
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| |
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| |
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| |
| | |
| PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP SD-0 DIABLO CANYON POWER PLANT REVISION 7A PAGE 7 OF 38 TITLE: LOSS OF, OR INADEQUATE DECAY HEAT UNITS 1AND2 REMOVAL APPENDIX B ESTIMATION OF DECAY HEAT AND HEATUP RATE
| |
| : 1. PREDICTED HEAT LOAD DECAY HEAT l 25 .
| |
| m I
| |
| 20 v
| |
| L g 15 I m \
| |
| ce <
| |
| p 10 \
| |
| 4 -
| |
| u U _____ _____ :sm.. 52"a==sg M 5 ----- ----- ------ - ----- ------ ---
| |
| A _____ _____ _____ _____ ____ m ..... ..m _____
| |
| 0 0 10 20 30 40 50 60 70 80 DAYS AFTER SHUTDOWN
| |
| : 2. REDUCTION FACTOR FOR REFUELED CORES MW X = MW Predicted Fraction of Previously Esthnated Heat Load Used assemblies Decay Heat Installed in Core
| |
| * Imd
| |
| * Use 1.0 if unknown 01068807. DOE '2 7 M09.0927
| |
| | |
| i CATEGORY: A.1 QUESTION # 1:
| |
| References Allowed: No During the implementation of an EOP the Shift Foreman informs die Shift Supervisor that he believes the procedure is inappropriate for the current plant conditions. The Shift Supervisor concurs with the Shift Foreman's assessment of the problem.
| |
| What two (2) checks should be performed per OPl.DCI1 " Conduct of Operations - Abnormal Conditions" by the Shift Foreman and Shift Supervisor prior to the Shift Supervisor authorizing the return to the diagnosis section of EOP E-0 for re-diagnosis instructions?
| |
| l 1
| |
| i l
| |
| QUESTION # 2: )
| |
| References Allowed: Yes Given the following:
| |
| Unit 1 is operating at 100% power.
| |
| . 0200 on January 2,1999 - CCP l-2 declared inoperable due motor problems.
| |
| . 0200 on January 5,1999 - CCP l-2 is still inoperable w/an estimated RTC of 12 hrs.
| |
| . 0200 on January 5,1999 - A Unit 1 Shutdown is initiated. i
| |
| * 0500 on January 5,1999 - CCP l-1 trips on OC due to pump bearing failure. }
| |
| e 0505 on January 5,1999 - PDP l-3 was placed in service and letdown restored. l 1
| |
| When is the unit REQUIRED to be in MODE 3 (date and time)? l l
| |
| STtIDENT COPY
| |
| | |
| CATEGORY: A.3 QUESTION # 1:
| |
| References Allowed: Yes Unit I has sustained a large break LOCA and fuel damage due to a 0,5g earthquake. The
| |
| - estimated exposure involved with obtaining an RCS sample is 4 REM.
| |
| Under these conditions, who can authorize the chemist, with a previous exposure history of 1.5 REM for the current year, to take the sample and receive the estimated exposure?
| |
| QUESTION # 2:
| |
| References Allowed: Yes You made two entries into the containment today and received the following exposure:
| |
| l'' entry: 124 mrem gamma; 0 mrem neutron 2"d entry: 42 mrem gamma; 44 mrem neutron The folbwing is your exposure history this yearjust prior to today's entries:
| |
| e 320 mrem DDE (Deep Dose Equivalent)
| |
| * 70 mrem CEDE (Committed Effective Dose Equivalent)
| |
| * 48 mrem SDE (Shal.'ow Dose Equivalent)
| |
| * 28 mrem CDE (Conanitted Dose Equivalent)
| |
| How much more radiation exposure can you receive without any signed extensions?
| |
| STUDENT COPY
| |
| | |
| c i
| |
| 8'
| |
| )
| |
| l CATEGORY: ' 'A.1 TOPIC: Procedure Usage KA: G2.1.20 (4.2)
| |
| Reference Allowed: NO l
| |
| 1 i
| |
| | |
| ==Reference:==
| |
| LPE-RULE, Obj.15, and OPl.DCl1, p.10.
| |
| i QUESTION # 1: )
| |
| During the implementation of an EOP the Shift Foreman informs the Shift Supervisor that he believes the procedure is inappropriate for the current plant conditions. The Shift Supervisor concurs with the Shift Foreman's assessment of the problem.
| |
| l 3
| |
| What two (2) checks should be performed per OPl.DCI1 " Conduct of Operations - Abnormal Conditions" by the Shift Foreman and Shift Supervisor prior to the Shift Supervisor authorizing .
| |
| the return to the diagnosis section of EOP E-0 for re-diagnosis mstructions? l l
| |
| l ANSWER:
| |
| : 1. The entry conditions for the procedure in use shall be reviewed. [0.5] l
| |
| : 2. Ifit is determined that the entry conditions are not met, the previous procedure in use shall be l reviewed to determine if a correct procedure transition was made. [0.5]
| |
| NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT b y f
| |
| /i n!hae g%s kuf;yl e= Pt) 44 r = y/
| |
| /r is e a e &=a a s a ussa s i
| |
| Admin-Package
| |
| | |
| PACIFIC GAS AND ELECTRIC COMPANY NUMBER OPI DC11 DIABLO CANYON POWER PLANT REVISION 12 PAGE 10 OF 15 TITLE: CONDUCT OF OPERATIONS-ABNORMAL PLANT CONDITIONS 5.5.9 If during the implementation of an EOP or AOP the Shift Foreman suspects that the procedure is inappropriate for the existing plant condition the following steps should be implemented:
| |
| : a. The Shift Supervisor'(or in the SS's absence, the STA) shall concur with the suspected problem.
| |
| : b. The entry conditions for the procedure in use shall be reviewed.
| |
| : c. If it is determined that the required entry conditions are not met, the previous procedure in use shall be reviewed to determine if a correct procedure transition was made.
| |
| : d. If entry into an EOP or Shutdown AOP has been made and, after performing Steps a, b, and c above, it still cannot be confirmed that the proper procedure is in use, the Shift Supervisor may authorize return to the diagnosis section of EOP E-0 or AP SD-0 for re-diagnosis instructions.
| |
| 5.5.10 Immediate operator actions for EOPs and AOPs should be performed in the following manner:
| |
| : a. The fact that entry into an EOP is necessary should be announced by '
| |
| the individual who observes the entry conditions.
| |
| : b. The Control Board Operators should immediately perform the required immediate actions. The results or conclusions from the immediate action verification should not be verbalized until requested by the pmcedure reader.
| |
| : c. If equipment is found in a condition other than expected, the condition should be corrected immediately, and the Shin Foreman shall be i notified of the problem as soon as possible. This applies to easily corrected conditions (e.g., Main Turbine not tripping on a Reactor Trip signal, components which have failed to respond to actuation signals) and is not intended to imply that all R.N.O. sections be memorized.
| |
| : d. When ready, the designated precedure reader should review the immediate actions with the crew in accordance with 5.5.3 of;his procedure.
| |
| 01165512.DOA IB 10 1001 1121
| |
| | |
| 1
| |
| , LESSON: EOP RULES OF USAGE - LESSON No.: LPE-RULE I General Rules, continued l
| |
| l Cases where The Shift Foreman can direct an Operator to perform an action prior to Operator can ,: aching that step in the procedure. According to OPl.DC11, if it is
| |
| ! deviate from oovious that the steps need to be performed, and:
| |
| l S'9"'A . The SFM consults with the SS or STA.
| |
| Obj ri . It is clear that the steps benefit event mitigation.
| |
| l . The SFM directs the steps to be performed.
| |
| ! . All steps performed are verified during the normal course of the procedure.
| |
| Justification for Tra.aitions between steps or procedures may occur when:
| |
| Transitions . directed by a procedural step (in either column).
| |
| Obj7 . a specified condition is met in a foldout page.
| |
| . a higher order procedure becomes necessary. l Eule ,
| |
| When a transition takes place within a procedure, all previous steps, notes
| |
| )
| |
| and cautions remain in effect.
| |
| If a transition is made to another procedure, notes and cautions of the previous procedure are no longer in effect, unless:
| |
| . it specifies it is to remain in effect, or
| |
| . a transition is subsequently made back to "the procedure and step in effect."
| |
| Rediagnosis - If, during the implementation of an AOP or EOP the Shift Foreman Objl5 suspects that the procedure is inappropriate for the existing plant conditions, the Shift Supervisor shall be informed and the following steps should be implemented.
| |
| . The Shift Supervisor must concur with the suspected problem. j
| |
| . The entry conditions for the procedure in use shall be reviewed.
| |
| . If it is determined that the required entry conditions are not met, the previous procedure in use shall be reviewed to determine if a correct procedure transition was made, and the proper procedure will then be implemented.
| |
| . In the case of the Optimal Recovery Guidelines
| |
| ^
| |
| Continued on nextpage l l
| |
| LPERULE. doc PAGE 11 oF 33 REV 7
| |
| | |
| I~
| |
| l CATEGORY: A.1
| |
| ! TOPIC: Procedure Usage l KA: G2.1.12 (4.0) ;
| |
| I Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| STO M8, Obj. 31 pg. 3-12, Technical Specifications 3.0.3,3.1.2.4 & 3.5.2 QUESTION # 2:
| |
| Given the following:
| |
| Unit 1 is operating at 100% power.
| |
| . 0200 on January 2,1999 - CCP 1-2 declared inoperable due motor problems.
| |
| . 0200 on January 5,1999 - CCP 1-2 is still inoperable w/an estimated RTS of 12 hrs.
| |
| . 0200 on January 5,1999 - A Unit 1 Shutdown is initiated. j e - 0500 on January 5,1999 - CCP l-1 trips on OC due to pump bearing failure.
| |
| . 0505 on January 5,1999 - PDP l-3 was placed in service and letdown restored. j l
| |
| When is the unit REQUIRED to be in MODE 3 (date and time)? j
| |
| . ANSWER:
| |
| MODE 3 is required by 0800 on January 5,1999
| |
| {
| |
| NOTE:. Although 2 CCPs inoperable places the unit in T.S. 3.0.3 and requires MODE 3 in 7 hours, T.S. 3.1.2.4 is more limiting since a required shutdown was already in progress with 3 )
| |
| hours remaining to MODE 3, therefore MODE 3 is required by 0800 on January 5,1999. I l
| |
| l 1
| |
| Candidate's Response: SAT UNSAT !
| |
| a n m vn 4 a ybaa a v m!! k yz=m = v l g3[ Qsjgg {V[m p#gDb .
| |
| j wM1 ; g ;; g h p[%g d
| |
| b 6 g l
| |
| Admin Package
| |
| | |
| Technical Specifications
| |
| ? l l Scenario Resolutions, continued l 1
| |
| i Scenario 3 Scenario Three resolution is listed below.
| |
| Obj 31
| |
| ! LCO 3.1.2.4 states that two charging pumps must be OPERABLE.
| |
| ACTION statement requires the operator to restore at least two charging pumps to OPERABLE status with 72 hours or have the plant in HOT STANDBY, borated to a shutdown margin greater than or equal to 1%
| |
| AK/K at 200*F within the next 6 hours, and to restore at least two charging pumps to OPERABLE status within the next 7 days or have the plant in cold shutdown within the next 30 hours.
| |
| l LCO 3.5.2 states that two ECCS subsystems must be OPERABLE. It also l states that each subsystem must have its own centrifugal charging pump. I l i ACTION statement requires the operator to restore the inoperable l subsystem to OPERABLE within 72 hours or have the plant in HOT l STANDBY within the next 6 hours and HOT SHUTDOWN within the following 6 hours.
| |
| Comment: This is an example in which there are more than one applicable LCO, yet neither refers to the other. .Ta 5tddition, this example shows that one LCO (3.1.2.4) may be more rest *ictive and specific than another (3.5.2). The operator must know the Technical Specifications well enough to find and consider all LCOs that are applicable to each individual situation.
| |
| Scenario 4 Scenario Four resolution is listed below.
| |
| Obj 31 LCO 3.5.1 states that each accumulator shall be OPERABLE with a boron concentration between 2200 ppm and 2500 ppm.
| |
| ACTION is not required because these concentrations are within the limits of the LCO.
| |
| l l
| |
| l M8. DOC 3 - 12 REV 6 l
| |
| | |
| SEE PSRC INTERPRETATION 90-09 92-01 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY
| |
| . LIMITING CONDITION FOR OPPtATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
| |
| 3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time-intervals. If the Limiting Condition for Operation is restored prior to Expiration of the specified time intervals, completion of the ACTION requirements is not required.
| |
| 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
| |
| : a. At least HOT STANDBY within the next 6 hours,
| |
| : b. At least HOT SHUTDOWN vithin the following 6 hours, and
| |
| : c. At least COLD SHUTDOWN within the subsequent 24 hours.
| |
| 'Where corrective measures are completed that permit operation under the ACTION
| |
| ' requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.
| |
| Exceptions of these requirements are stated in the individual specifications.
| |
| This specification is not applicable in MODE 5 or 6.
| |
| 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an I ;
| |
| unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements. Exceptions to these requirements are stated in the individual specifications.
| |
| 3.0.5 Limiting Conditions for Operation including the associated ACTION requirements shall apply to each unit individually un%ss otherwise indicated as follows:
| |
| l
| |
| : a. Whenever the Limiting Conditions for Operation refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously.' This will be indicated in the ACTION section;
| |
| : b. Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and
| |
| : c. Whenever certain portions of a specification contain operating parameter =, l Setpoints, etc., which are different for each unit, this will be identifiec ... l parentheses, footnotes or body of the requirement. ,
| |
| l 1
| |
| DIABLO. CANYON - UNITS') & 2 3/4 0-1 Amendment Nos. 55 and 54 J
| |
| | |
| SEE PSRC INTERPRETATION 94-07, 94-08, 96-01 1
| |
| REACTIVITY CONTROL SYSTEMS-CHARGING PUMPS - OPERATING r
| |
| LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, 3 and 4#.
| |
| ACTION:
| |
| With only one charging pum) OPERABLE, restore.at least two charging pumps to OPERABLE status within 72 iours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200'F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
| |
| SURVEILLANCE REOUIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE when tested pursuant to Specificatica 4.0.5. In addition, when the above required charging pumps include a centrifugal charging pump (s), verify)that, onrecircula differential pressure of greater than or equal to 2400 psid.
| |
| 4.1.2.4.2 All centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable
| |
| * at least once per 12 hours whenever the temperature of one or more of the Reactor Coolant System (RCS) cold-legs is less than or equal to 270'F by verifying that the motor breaker D.C.
| |
| control power is de-energized.
| |
| 1
| |
| #A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature'of one or more of the RCS cold legs is less than or equal to 270*F.
| |
| 4
| |
| *An inoperable pum) may be made OPERABLE for testing per Specification 4.0.5 provided the disciarge of the pump has been isolated from the Reactor Coolant System by an isolation valve with power removed from the valve operator, or by a sealed closed manual isolation valve.
| |
| DIABLO CANYON - UNITS 1 & 2 3/4 1-11 Unit 1 - Amendment No. 100 Unit 2 - Amer.dment No. 99
| |
| ,on,r.a. .. ,, van o so no.4i is inne
| |
| | |
| - Sie PSRC Interpretation 87-04, 90-U/, 9b-01 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - Tava GREATER THAN OR E00AL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two Emergency Core Coolirig System (ECCS) subsystems shall be OPERABLE i with each subsystem comprised of:
| |
| : a. One OPERABLE centrifugal charging pump,
| |
| 'b. One OPERABLE Safety Injection pump, I
| |
| : c. One OPERABLE Residual Heat Removal heat exchanger, ,
| |
| 1
| |
| : d. One OPERABLE Residual Heat Removal pump, and
| |
| : e. An~ OPERABLE flow path capable of taking suction from the Refueling j Water Storage Tank on a Safety Injection signal and manually transferring suction to the containment sump during the
| |
| ; recirculation phase' of operation.
| |
| l i APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTION:
| |
| : a. With one ECCS subsystem. inoperable, restore the inoperable j subsystem to OPERABLE status within 72 hours or be in at least DOT j STANDBY within the next 6 hours and in at least HOT SHUTDOWN l within the following 6 hours. !
| |
| l
| |
| : b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission onrsuant to Specification 6.9.2 within '
| |
| 90 days describing the circunstances of the actu tion and the ;
| |
| . total accumulated actuation cycles to date. The current value of the usage factor for each afrected safety injection nozzle shall ;
| |
| i be provided in this $pecial Report whenever its value exceeds l 0.70.
| |
| l
| |
| .DIABLO CANYON - UNITS 1 & 2 3/4 5-3 lL '32972702.4a TAB 12 3
| |
| | |
| 7--
| |
| i CATEGORY: A.3 TOPIC: Radiation Control (Limits and Control requirements)
| |
| K/A: G2.3.1 (3.0)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| RP9T," Legal Exposure Limits," Obj 3.2 pg.10 and EP RB-2," Emergency Exposure Guides," Rev. 4A, pg. 2,4 and Attachment 9.6 QUESTION # 1:
| |
| Unit I has sustained a large break LOCA and fuel damage due to a 0.5g earthquake. The estimated exposure involved with obtaining an RCS sample is 4 REM.
| |
| Under these conditions, who can authorize the chemist, with a previous exposure history of 1.5 REM for the current year, to take the sample and receive the estimated exposure?
| |
| ANSWER:
| |
| : 1. The Recovery .,.anager or Site Emergency Coordinator since the plant is in an en, q ,. .:cv I with conditions that warrant at least an Alert emergency declaration.
| |
| NOTE: Either answer will provide a satisfactory grade.
| |
| Candidate's Resnonse SAT UNSAT I
| |
| l s-wAesa6w* 6 n AAa
| |
| ! Admin Package
| |
| | |
| i .
| |
| 3.2 EMERGENCY EXPOSURE AUTHORIZATION The Recovery Manager (RM) or Site Emergency Coordinator (SEC) prior to turnover, has the unilateral authority and non-delegataole responsibility for authorizing an individual emergency worker to exceed normal 100FR20 exposure limits.
| |
| A couple of requirements have to be mednt before the guidelines can be put into effect, those are an Emergency classification of Alert or higher has been declared and an essential emergency action is required and cannot be performed without one or more workers potentially exceeding 10 CFE 20 annual exposure limits.
| |
| PERFORM EXERCISE 3.2 l
| |
| 1 I
| |
| i l
| |
| i I
| |
| l RP9T 10 REV1
| |
| | |
| l NUMBER EP RB-2 PACIFIC GAS AND ELECTRIC COMPANY REVISION 4A DIABLO CANYON POWER PLANT PAGE 2 OF 6 UNITS 1 AND 2 TITLE: EMERGENCY EXPOSURE GUIDES )
| |
| : 2. DISCUSSION 2.1 Authorization of emergency ezposure is an extraordinary measure, but justifiable under four sets of circumstances.
| |
| When the intended action requiring a potential overexposure to an emergency worker, is expected to result in;
| |
| * Saving or preserving the quality of a human life that would otherwise be lost
| |
| . Significant projected dose saving to others.
| |
| . Protection of valuable property.
| |
| . Sampling results required to redefine or adjust existing Protective Actions for the public or site personnel.
| |
| All of the above situations require that no reasonable method is immediate*g available (or readily apparent) to avoid exceeding the established annual limits and that every effort will be made to keep the emergency exposure ALARA.
| |
| 2.2 The emergency exposure guidelines implencai at DCPP are consistent with the Environmental Protection Agency (EPA) guicance for controlling doses to workers under emergency conditions. (Ref rence 10.3) 2.3 Authorizal dose limits for workers during emergencies are based on avoiding acute health effects and limiting the risk of delayed health effects.
| |
| 2.4 Since by their very nature, emergency exposures are not extensively planned, emergency worker exposures are gg controlled by Planned Special Exposures.
| |
| NOTE: Planned Special Exposures may be implemented during non-emergency situations including Recovery Operations. Refer to RPl.ID8, " Planned Special Exposures."
| |
| 2.5 Emergency exposures shall be deducted from the Planned Special Exposure Limit available to each worker.
| |
| 2.6 An emergency exposure should be authorized only once in an individual's lifetime and is in addition to any prior occupational exposure from normal or planned special exposures.
| |
| 2.7 The emergency exposure limits specified in this procedure are applicable to both in-plant team response activities and off-site field monitoring by PG&E ERO personnel.
| |
| 2.8 Emergency exposure above 25 rem TEDE shall require the volunta.7 consent of the authorized individual.
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| nnnwn, nnt w 2 i
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| L i i-NUMBER EP RB-2 PACIFIC GAS AND ELECTRIC COMPANY REVISION 4A l MN.dLO CANYON POWER PLANT PAGE 4OF6 UNITS 1 AND 2 TTILE: EMERGENCY EXPOSURE GUIDES l
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| : 4. RESPONSIBILITIES 4.1 The Recovery Manager (RM) or Site Emergency Coordinator (SEC) prior to turnover, has the unilateral autLrity and non-delegable responsibility for authorizing an individual emergency worker to exceed normal 10 CFR 20 exposure limits.
| |
| The RM/SEC is furthermore responsible for ensuring that the NRC is notified of any overexposure that may result.
| |
| 4.2 The TSC Radiological Advisor (RA) is respon**aole for evaluating the conditions requiring an emergency exposure authorizar'.on and advising the RM (or SEC) on its justification and when all prerequisite re;airemente have been met.
| |
| 4.3 The EOF Radiological Manager (ERM) is responsible for evaluating radiological conditions and exposures to off-sitt emergency response personnel and advising the RM when an emergency exposur.: authorization is justified.
| |
| 4.4 The Site Radiation Protectios Coordinator (SRPC) is responsible for identifying the necessity of obtaining :n r.nergency exposure authorization and in assisting with '
| |
| volunteer selecuor.as needed.
| |
| 4.5 The Emergency Maintenance Coordinator (EMC) is responsible for ensuring that the maximum protection and support is provided to those personnel dispatched from the OSC under the extraordinary conditions of emergency exposure.
| |
| 4.6 The emergency worker is responsible for knowing the potential health consequences of the emergency exposure and volunteering to accept the associated risks of exposures above 25 rem TEDE.
| |
| The emergency worker is responsible for mainaining his/her emergency exposure ALARA consistent with the successful completion of the emergency activity.
| |
| : 5. PREREOUISITES 5.1 Emergency classification of Alert or higher has been declared.
| |
| 5.2 An essential emergency action is required (refer to Attachment 9.6) and cannot be performed without one or more workers potentially exceeding 10 CFR 20 annual exposure limic t
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| | |
| l 10/07/93 page g cf 2 l DIABLO CANYON POWER PLANT EP RB-2 A'ITACHMENT 9.6 AND 1
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| TITLE: DCPP EMERGENCY EXPOSURE GUIDELINES The following table contams guidelines for use in authortzmg emergency exposures when lower doses are not practicable:
| |
| RADIOLOGICAL PROPERTY DOSE SAVING LIFESAVING ASSESSMENT SAVING TO TO .
| |
| SAMPLING POPULATION
| |
| * INDIVIDUAL
| |
| * Emergency Sampling Under Mitigating Damage Corrective Lifesaving Actions-> Emergency to Valuable Actions, Actions,1st Aid, Part of Body Conditions Property stop/ reduce a Search and rescue Irradiated release Whole Body 5 rem TEDE 10 rem TEDE 25 rem TEDE 25 rem TEDE Skin & any 50 rem SDE 100 rem SDE - 250 rem SDE 250 rem SDE Extremity Lens of the Eye 15 rem LDE 30 rem LDE 75 rem LDE 75 rem LDE Any Organ or 50 rem 100 rem 250 rem 250 rem Tissues (CDE+DDE) (CDE+DDE) (CDE+DDE) (CDE+DDE) ,
| |
| l NOTES: 1. Radiological Assessment Samoline, includes collection of atmospheric, liquid, and environmentai radiological activity samples as well as chemistry samples involving high activity or high radiation. Emergency exposure limits may be authorized for selected individuals, for emergency assessment functions, in addition to annual occupational dose to date.
| |
| : 2. Property Savine, for example, might be dispatching the Fire Brigade to extinguish a fire in a Very High Radiation Area to protect plant equipment though no immediate threat exists to compromising Plant Safety
| |
| : 3. Dose Savine to Pooulation, includes activities that justify a potential overexposure to a few workers in order to save even a small average dose in a large population. (May J also include Traffic Control for Evacuees or other Security Plan Functions.) j
| |
| : 4. Lifesavine to Individual, includes the activity of search and rescue in very high dose I rates or high airborne activity. l l
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| Extreme situations may occur in which a dose in excess of 25 rem TEDE would be l unavoidable for p.ithel Dose Saving to (Large) Population or Lifesaving to (An)
| |
| Individual. ,
| |
| An authorization of emergency exposure whh NO LIMITS may be made under those conditions, but only to volunteers who are ful.'7 aware of the risks involved, including the numerical levels of dose at which acute effects of radiation will be incurred and the numerical estimates of the risk of delayed effects.
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| 00030604. doe 3B 16 1
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| )
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| '69-10554 10/07/93 Page1ofI DIABLO CANYON POWER PLANT EP RB-2 AND TITLE: EMERGENCY EXPOSURE PERMIT Date: Time: _
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| Permit #:
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| Responder (s):
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| (Print)
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| RP Support: _
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| Description of Activity:
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| Special Hazards:
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| Special Instructions:
| |
| Anticipated TEDE Rate: (rem /hr) AUTHORIZED LIMIT: [] 5 rem TEDE (Check One) [] 10 rem TEDE Anticipated Stay Time: (hr) [] 25 rem TEDE
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| [] NO LIMIT Anticipated TEDE: (rem)
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| * Voluntary Consent: I hereby volunteer to perform the act.ivity described above and I acknowledge having received a radiological briefing. I am fully aware of the health risks associated with the anticipated i exposure. (Sign Below.) l l
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| i 1
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| Authorization of Site Emergency Coordinator or recovery Manager: Time:
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| NOTE: Forward completed form to the Director, SH&ES for record retention.
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| h 00030604. DOE 3B 18 I
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| l 1 ,
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| CATEGORY: A.3 TOPIC: Radiation Control (Limits and Control requirements)
| |
| K/A: G2.3.4 (3.1)
| |
| Reference Allowed: YES 1
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| | |
| ==Reference:==
| |
| GRPA400e, " Experienced Radworker," Obj.' 1, pg. 3 and RPl.lD6," Personnel Dose Limits & Monitoring Requirements," Att.10.1 QUESTION # 2:
| |
| You made two entries into the containment today and received the following exposure:
| |
| 1" entry: 124 mrem gamma; 0 mrem neutron 2"d entry: 42 mrem gamma; 44 mrem neutron The following is your exposure history this yearjust prior to today's entries:
| |
| e 320 mrem DDE (Deep Dose Equivalent)
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| * 70 mrem CEDE (Committed Effective Dose Equivalent)
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| * 48 mrem SDE (Shallow Dose Equivalent)
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| * 28 mrem CDE (Committed Dose Equivalent)
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| How much more radiation exposure can you receive without any signed extensions?
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| ANSWER: ;
| |
| 1400 mrem (1.4 rem) l NOTE: Current exposure is all DDE and totals 210 mrem. TEDE is CDE + DDE therefore the total exposure including today's 210 + 320 + 70 = 600 mrem. The maximum possible exposure l without a signed extension is the DCPP guideline of 2000 rnrem/ year. Therefore the remaining exposure allowed is 2000 - 600 = 1400 mrem.
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| Candidate's Respoggi g SA_T. == wgUgSATgm%. g y pgg y
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| & D '1 Admin Package
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| | |
| i ll 1.1 ACRONYMS USED FOR TRACKING EXPOSURE LIMITS The following terms and definitions are used to track personnel exposure to prevent violation of Legal Exposure Limits.
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| Table RP9-1 PERSONNEL EXPOSURE TERMS AND DEFINITIONS Acronym Term Definition CDE. Committed Dose Equivalent 'Ihe dose equivalent to an organ that will be received from an intake by an individual during the 50 year period following the intake, j
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| ~
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| CEDE Committed Effective Dose The product of the Committed Dose Equivalent Equivalent and a weighting factor to equate dose to the organ to the equivalent dose (risk) from uniform irradiation of the whole body.
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| DDE Deep Dose Equivalent That dose associated with exposure of the whole body (depth ofI cm).
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| LDE Lens Dose Equivalent The dose that applies to the external f
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| exposure of the lens of the eye (depth of 0.3 cm).
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| SDE Shallow Dose Equivalent The dose that applies to the external exposure of the skin or any extremity (depth of 0.007 cm).
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| TEDE Total Effective Dose Equivalent The sum of the DDE(for extemal ;
| |
| exposures) and the CEDE (for l intemal exposures). 1 TODE Total Organ Dose Equivalent The summed toal of the DDE (extemal whole body exposure) and the CDE (intemal dose to the organs or tissues from internal exposure from non-stochastic ALI),
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| when CDE is not zero.
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| DPW Declared Pregnant Woman A woman who has voluntarily '
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| informed DCPP,in writing, ofher '
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| pregnancy and the estimated dated '
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| ofconception.
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| RESTRICTED For DCPP the Restricted Area is AREA-' defined as the Protected area.
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| RPOT 4 REV1
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| | |
| LESSON NO: GRPA400-3GRPA400-3 LESSON: EXPOSURE LIh11TSEXPOSURE LIMITS OUTLINE KEYS / AIDS OBJECTIVE #C6 State the DCPP exposure limits.
| |
| INTRODUCTION The station exposure limits at DCPP are' different from federal limits in that station limits can be raised after review and approval by management.
| |
| NEED TO KNOW DCPP station exposure limits are 90% of the Show TP-7 federallimits.
| |
| Diablo Canyon Station limits:
| |
| TEDE administrative guideline 2.0 rem / year TEDE limil 4.5 rem / year
| |
| * Lens of the eye 13.5 rem / year
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| * Skia 45 rem / year
| |
| * Extremity limit 45 rem / year
| |
| * Organ limit 45 rem / year
| |
| * Declared pregnant worker ALARA#
| |
| * calendar year (Janitary 1" through December 31")
| |
| #As Low As Reasonably Achievable (Def'mitely less than the federal limit of 0.5 rem / term) J l
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| l 1
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| l t
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| Page 13 Rev 1010
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| I I
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| f l l
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| LESSON: EXPERIENCED RADWORKEREXPERIENCED RADWORKER LE COMPONENT DISPLAY KEYS / AIDS ]
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| l l DISPLAY OF THE LESSO_N i
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| l OBJECTIVE 1 (C6) 1 State the DCPP Exposure Limits.
| |
| NEED TO KNOW DCPP exposure limits are sat at 90% of the federal limit.
| |
| TEDE admin guideline 2.0 Rem / year TEDE (whole body) limit 4.5 Rem / year l Lens of the eye limit 13.5 Rem / year Skin limit 45 Rem / year Extremity limit 45 Rem / year l Organ limit 45 Rem / year Declared pregnant worker as low as reasonably achievable
| |
| * HEIZ
| |
| *The DCPP embryo / fetus protection program is contained in procedure RP1.ID10. Refer to the procedure or call 4325 for more information.
| |
| 1 Very few workers require an extension of exposure beyond the administrative guideline of 2.0 Rem. In most cases workers are rotated through high dose jobs to equalize exposure and keep all workers dose well below the limits. ;
| |
| 1 PRACTICE / FEEDBACK What is the station exposure limit for skin? (45 Rern/ year)
| |
| What are the station TEDE exposure limits? (2 Rem / year is the guideline and 4.5 Rem / year is the limit)
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| EXAMPLES / NON-EXAMPLES Some workers with special skills are extended above the administrative guideline of 2 Rem to perform jobs such as S/G nozzle dam installation.
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| Page 3 Rev M
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| r ,.
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| .r
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| /
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| /
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| l l
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| /
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| /
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| M J .
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| I f
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| I l
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| \
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| Appendix D Scenario Outline Fdrm ES-D-1 1
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| Facility: DCPP Units 1 & 2 Scenario No.: 1 Op-Test No.: Backup Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U )
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| SRO-U / SRO-1 l
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| Objectives: Evaluate the crew's ability to diagnose and respond to an Eagle 21 system failure .
| |
| Evaluate the crew's ability to diagnose and respond to a letdown pressure controller failure ]
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| l Evaluate the crew's ability to diagnose and respond to a S/G Level Channel failure Evaluate the crew's ability to diagnose and respond to a RCP # 1 seal failure Evaluate the crew in using EOPs during a RCP seal failure LOCA Evaluate the crew's ability to respond to a partial P-4 actuation l Evaluate the crew 's ability to diagnose and respond to a undercompensated IR channel Initial Conditions: 100% power, equilibrium xenon, Beginning of cycle (IC-1)
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| )
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| Tumover: 10 gpd leak on S/G 1-3, D/G 1-1 is 00S, CCW pump 1-3 is 00S MFW pp 1-2 Increased Vibrations - ramp unit to 50% for repair ASAP
| |
| , i Tiae Event Malf. Event Event I 1
| |
| min No. No. Type
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| * Description 0 1 n/a N, RO Ramp power to 50% for repair of MFW pp 12 l
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| 10 2 xmt I, BOP S/G LevelChannelLT-549 fails HIGH I mfw46 15 3 cnh cyc8 C, BOP Letdown pressure control valve PCV-135 fails CLOSED in auto, operatesin manual 23 4 ppl6a I, RO LCP Halt Protection Set 1, Rack 1 C, RO 33 5 rep 1b RCP 1-2 # 1 seal failure )
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| 36 ser1244 RCP 1-2 # 1 High Vibration cond on 6 rep 2b, M,ALL RCP 1-2 seal package failure.
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| trip rep 3b ppl5a C,SFM Rx Trip Breaker"RTA" fails to OPEN res3b Loop 2 RCS SB LOCA cond nis 3b I, RO IR NI 36is undercompensated evc2 C, RO Charging Pump 1-2 Overcurrent trip on Safety injection signal (N)ormal (R)eactivity (i)nstrument (C)omponent (M)ajor 0
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| | |
| SCENARIO 01 OVERVIEW .
| |
| The crew is directed to ramp to 50% power to remove MFP 1-2 from service (Event 1).
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| The RO will have to borate to ramp turbine and reactor power to 50%.
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| S/G Level transmitter LT-549 fails High (Event 2). The BOPCO and CO should identify I the failed channel and report to the SFM. The SFM should refer to annunciator
| |
| : response procedure PK09-15 for a Digital Feedwater Control System trouble and enter OP AP-5 for a Malfunction of Protection or Control channel. The SFM should inform the RO and BOPCO of the affects of the failure on AFW level controlif needed.
| |
| Letdown pressure controller PCV-135 fails closed while in the auto position (Event 3).
| |
| BOPCO should identify the failure of PCV-135 to control letdown pressure and report to j the SFM. SFM should refer to PK04-21 for letdown press / flow. SFM should direct that I the failed controller be placed in manual to regain letdown pressure control and ensure that the letdown relief valve re-seats or establishment of excess letdown.
| |
| Eagle 21 (LCP Halt) failure comes in (Event 4). The CO and BOPCO should identify the failure and report to the SFM. The SFM should refer to annunciator response procedure PK06-01 and enter OP AP-5. Crew should identify all instrumentation that is affected by the failure. LT-459 and PT-455 are failed as is and LT-459 and PT-455 should be deselected for control.
| |
| RCP 1-2 seal leakoff flow increases to 60 gpm (over 20 minutes) giving indications of a l RCP 1-2 # 1 seal problem (Event 5). SFM should refer to annunciator response procedure PK05-02. RCP parameters should be monitored and Operations Management should be informed. RCP 1-2 High Vibration (Event 5 continued). SFM should refer to annunciator response procedure PK05-05. SFM should direct a Reactor Trip, followed by the tripping of RCP 1-2 and the subsequent isolation of the RCP 1-2 # ,
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| 1 seal leakoff valve. I Following the transition to EOP E-0.1, RCP 1-2 seal package will fail, causing a 60gpm LOCA (Event 6), additionally Reactor Trip Breaker "A" fails to open. The LOCA results in decreasing pressurizer level accompanied by decreasing RCS pressure. The SFM should direct a manual Si prior to an automatic Si at 1850 psig and re-entry into EOP E-0 followed by a transition to EOP E-1. The SFM should direct the local action to open RTA. Charging Pump 1-2 will overcurrent trip on the SI signal and will not be available.
| |
| Approximately 10 to 15 minutes following the reactor trip, the intermediate range Ni should be low enough to energize the SR instruments except that N1-36 is undercompensated which results in only 1 of 2 intermediate Nis below P-6. The RO/ BOP should recognize the failure of SR instruments to energize and the SFM should direct the manual re-energizing of the SR detectors.
| |
| The scenario will be terminated following diagnosis of the RCS not intact and a transition out of EOP E-1 to EOP E-1.1.
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| i; \
| |
| l Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 1 Event No.: 1 Page _1_of_7 Event
| |
| | |
| == Description:==
| |
| Commence ramp to 50% power Time Position Applicant's Actions or Behavior BOP- Monitor Main Feedwater Pump 1-2 vibration readings RO Initiate Boration for ramp to 50% power
| |
| . Operate the makeup system mode select switch j
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| . Operate the boric acid integrator i
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| Setup DEHC
| |
| . Place MW feedbackin service
| |
| . Set MW reference ,
| |
| . Setload rate i
| |
| Commence ramp to 50%
| |
| Monitor Tavg and Tref and borate as necessary SFM Go to PK09 Review actions for increased MFW pp. Vibrations Review precautions and limitations of OP L-4 and conduct tumover briefing Direct RO to commence a ramp to 50% power at up to SMW/ min (Admin Umit)
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| Direct BOPCO to monitor MFW pp.1-2 bearing vibrations NUREG 1021 1 Interim Rev. 8, January 1997
| |
| | |
| r 1
| |
| l Appendix D Scenario Outline Form ES-D-2 l
| |
| Op-Test No.:_1 _ Scenario No.: 1 Event No.: 2 Page _2_ of _7_
| |
| Event Description : SG LevelChannel LT-549 FailHigh l
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| l Time Position ' Applicant's Actions or Behavior BOP Recognize and report LT 549 failure RO Monitor and report SG level trends and Digital Feedwater controls status I
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| 1 l
| |
| SFM Go to PK 09-15
| |
| . Dispatch Operator to Digital Feedwater Cabinet in Cable Spreading Room Go to OP AP-5
| |
| . Review attachment 4
| |
| . Determine that failure will affect SG level control on a Reactor Trip
| |
| . Initiate a bistable trip sheet (verbalize and initiate BS sheet actions)
| |
| Consult Technical Specification 3.3.1
| |
| . Trip inoperable channel within 6 hours Direct Control Room Asset Team to investigate failure of LT-549 i
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| l l
| |
| NUREG-1021 2- Interim Rev.8, January 1997
| |
| | |
| l l
| |
| i Appendix D Scenario Outline Form ES-D-2 i
| |
| Op-Test No.:_1 _ Scenario No.: 1 Event No.: 3 Page _3._ of _7_ j Event Description : Letdown Pressure control valve PCV-135 failure l Time Position Applicant's Actions or Behavior l
| |
| BOP Recognize and report Letdown flow decreasing to zero l
| |
| Recognize and report Letdown pressure normal and PCV-135 demand at zero %. [
| |
| Retums letdown flow to 75 gpm with PCV-135 in manual Recognize and report Letdown RV has lifted i I
| |
| Reduces Letdown pressure to reseat Letdown RV and retums letdown to normal in manual or isolates letdown and places excese letdown in service as directed.
| |
| f RO Monitors Letdown and Charging parameters Monitors PRT Parameters l Recognize and report PRT levelincreasing Placec turbine ramp on HOLD at DEHC panel l SFM Go to PK 04-21
| |
| . Determine that Letdown Pressure controller PCV-135 has failed in auto
| |
| . Direct BOPCO to controlletdown pressure using PCV-135 in manual
| |
| . Direct RO to monitor Letdown, Charging and PRT parameters
| |
| . Direct RO to place Turbine ramp on Hold at DEHC panel Go to OP B-1A:XII
| |
| * Direct BOPCO to reduce Letdown pressure to reseat Letdown RV or isolate letdown and establish excess letdovm Direct Control Room Asset team to investigate failure of PCV-135 NUREG-1021 3 Interim Rev. 8, January 1997
| |
| | |
| I Appendix D Scenario Outline Form ES-D-2 Op-Test No.:__1 Scenario No.: 1 Event No.: 4 Page _4_ of 7_
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| Event
| |
| | |
| == Description:==
| |
| Eagle 21 failure - LCP HALT on Protection Set 1, Rack 1 Time Position Applicant's Actions or Behavior BOP Recognize and report syinptoms of an LCP Halt
| |
| . Annunciators for channel set failure l
| |
| Take actions as directed by SCO I
| |
| RO Recognize and report symptoms of an LCP Halt on proter. tion set 1
| |
| . Annunet!ators for channel set failure
| |
| . Protection set 1 alarm on annunciator typewriter Check controlling systems controlling in AUTO
| |
| . LT-459 failed as is
| |
| . PT-455 failed as is
| |
| . Deselect 459 and 455 as controlling channels SFM Acknowledge reports from BOP /R0 (PK06-01 and 06-04)
| |
| Go to OP AP. 5 and direct operator recovery actions
| |
| . Direct RO to deselect 455 and 459 as pressurizer controlling channels Contact Nuclear Operator / Control Room Asset Team to investigate l Consult Technical Specifications 3.3.1 and 3.3.2
| |
| ! . 6 hour action to trip affected bistables for PZR press, PZR level and RCS flow NUREG-1021 4 Interim Rev. 8, January 1997
| |
| | |
| i -
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 1 Event No.: 5_ Page _5_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP 1-2 No.1 seat failure l Time Position Applicant's Actions or Behavior BOP Recognize and report high sealleakoff flow on RCP 1-2 I l
| |
| Y C hh$k;hY$Yh:db?hhSDbh!
| |
| OTQRCP$2&%&#fl4f8W AbhNbbhhh $%Wn ::h% l 1 Close No.1 sealleakoff valve after 3 to 5 minutes :
| |
| Monitor CCW to ensure adequate therma l barrier flow Complete immediate actions of EOP E-0
| |
| . Verify 4kV buses energized ;
| |
| I RO Acknowledge and report PK05-01 alarm input number 1393, RCP 1-1 Seal Leakoff Flow Hi Acknowledge and report PK05-05 RCP Vibration m -
| |
| '; % g. -d.
| |
| ,x W m, . , . ,.
| |
| lhitiate manuai_ reactor.thp. .. .yy ,, i b g' g;.
| |
| 5, ,
| |
| '...s .y n... 95 " 4 d.o V '7?Mr v; ,; W M
| |
| . x
| |
| !". eCribcalTaskbCM. . JW ?- ,' <
| |
| %se Complete immediate actions of EOP E-0
| |
| . Verify Rx tripped - reports RX trip Bkr. "A" closed
| |
| . Verify Turbine tripped
| |
| . Check Si actuation not required SFM Go to PK05-02, RCP 12 Go to PK05-05, RCP Vibration Diagnose No.1 sealfailure vn,. v, .n .n - c , . . . ,
| |
| ; Direct operators to perform a manualrea
| |
| , i: tor trip and the,n trip ~ RCR 1-2; V 4 yWm. . S e ~. wa,e ,
| |
| mw a n(([dJm;hj/DTM;j;,do
| |
| .7CriticalTask 4 :: wms < O':b, ' * " 'sCir + "7h " ' m Go to EOP E-0
| |
| . Directs Operator to locally open Rx Trip Bkr. "A" Go to EOP E-0.1 NUREG-1021 5 Interim Rev,8 January 1997
| |
| | |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 1 Event No.: 6__ Page _6_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP sealfailure (SBLOCA)
| |
| Time Position Applicant's Actions or Behavior BOP Recognize and report RCS Pressure Decrease Complete immediate actions of EOP E-0
| |
| . Veify 4kV buses energized Perfonn Appendix E; Secondary Auxiliary Status Control AFW flow to S/G 1-4 in manual due to level channel failure RO Recognize and report RCS Pressure Decrease l Perform Manual, Safety fnjec' hon, 4 1 ;[ g .
| |
| ;; 'Mjg@@ i
| |
| .~ Q3Qlg G .
| |
| K ai,; 3 Q9;' Qn Q+pQQ%f,s Complete immediate actions of E0P E-0
| |
| . Verify Rx tripped
| |
| . Verify Turbine tripped
| |
| . Check Si actuation required Recognize and report charging Pump 1-2 over current trip Recognize and report failure of both SR channels to energize
| |
| . Energize both SR channels (due to NI-368 failure)
| |
| " Critical Task NUREG-1021 6 Interim Rev.8, January 1997 u
| |
| | |
| n
| |
| ]
| |
| I s
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1 and 2_ Scenario No.: 1 Event No.: 6__ Page _7_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP sealfailure(Continued)
| |
| I l
| |
| I Time Pocition Applicant's Actions or Behavior SFM
| |
| % j$jlf(;hid' i Ndi$dslNNNiM$$@[s$$$N.$ $dsdMPSN%E$2 Re-enter EOP E-0 and direct immediate actions of E-0 .
| |
| * Direct STA to monitor CSFSTs Direct the following actions
| |
| . Direct energizing both SR detectors e Direct BOPCO to control AFW flow to SG 1-4 Transition to EOP E-1
| |
| . Review majoractions 1
| |
| . Assign Foldout Pageitems l
| |
| 4 v.
| |
| ]
| |
| Transition to EOP E-1.1
| |
| . Review major actions
| |
| . Assign Foldout Pageitems
| |
| . NUREG-1021- 7 Interim Rev. 8, January 1997
| |
| | |
| 1 i
| |
| dppendix D OperatorActions Fo'rm ES-D-1 l
| |
| Facility: DCPP Units 1 & 2 Scenario No.: 2 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U
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| _ SRO-U / SRO-U SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to a RCS makeup control failure Evaluate the crew's ability to diagnose and respond to a Radiation monitor failure Evaluate the crew's ability to diagnose and respond to a Thermal barrier leak Evaluate the crew's ability to diagnose and respond to a NI-42 Failure Evaluate the crew's ability to respond to a seismic event Evaluate the crew's ability to diagnose an ATWS and take timely actions l Evaluate the crew's ability to diagnose and respond to a stuck open PORV l Evaluate the crew in using the E0Ps during recovery from SB LOCA initial Conditions: 75% power, equilibrium xenon, end of cycle (IC-36)
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| Tumover: D/G 1-1 is 00S, for lube oil heater replacement Ramp to 100% power following repairs to #2 Htr Drip pump.
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| Time Event Malf. Event Event min No. No. Type
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| * Description 0 1 n/a N,SFM Ramp power to 100% following repairs to #2 Htr Drip pump R RO 5 2 viv cvc14 C, Ro FCV-111B VCT inlet from RCS makeup fails closed 12 3 xmtrms34 1, BOP Instrument failure of RM44 results in High rad alarm and CVI 20 4 nissb i,ALL Instrument failure of NIS-42 high 27 5 ccw3c C,ALL RCP 1-3 Thermal Barrier Leak (30 gpm) to CCW system 42 6 ppl5a M,ALL 0.35g Earthquake causes loss of startup power, Auto and manual ppl5b C, BOP Reactor trip blocked, PORV PCV-456 faits open with no power to block valve.
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| Post 7 res35 1, RO TE 413A Wide range Thot fails high Trip (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
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| SCENARIO 02 OVERVIEW The crew is directed to ramp to 100 % power (Event 1), #2 Htr. Drip pump repairs are complete with the pump warmed up and running. The RO will have to dilute to ramp turbine and reactor power to 100%.
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| During the ramp to 100% power, FCV-111B wi.Il fail to open in auto or manual (Event 2) causing RCS normal dilution flowpath to become inoperable. The SFM should refer to PK05-11 and enter AP-19 for malfunction of reacter control makeup system. SFM should direct local operators to investigate FCV-111B failure and direct the CO to use Alternate Dilution for RCS makeup.
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| Containment Exhaust radiation monitor RM-44A will fail high, over-ranged (Event 3),
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| causing an inadvertent containment vent isolation (CVI). SFM should refer to the annunciator response procedure PK011-21, High Radiation and BOPCO should be sent to control room radiation monitors to locate the source of the high alarm condition.
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| SFM should refer to annunciator response procedure PK01-17 and determine that RM-12, Containment Gaseous Radiation Monitor is out of service and direct the BOPCO to
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| . place the CFCU drain collection system in service.
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| Nuclear Instrument NI-42 will fail high (Event 4), CO should report the failure to the SFM. BOPCO should be sent to NI panels in the control room to investigate the failure.
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| SFM should enter OP AP-5 for the Ni failure and determine that NI-42 can be defeated.
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| SFM should direct the BOPCO to defeat NI-42 using OP AP-5 Attachment 4.1. BOPCO should defeat NI-42 at the NI Racks.
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| Reactor coolant pump 1-3 thermal barrier leakage (Event 5), ramps from 0 to 30 gpm causing increase in CCW head tank level, Radiation monitors RM-17A & B alarm and close of the CCW head tank vent RCV-16. CO and BOPCO should report to the SFM.
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| , The SFM should enter AP-1 for Excessive Reactor Coolant Leakage and directs CO and BOPCO in the determination of leak location. SFM enters AP-11 Malfunction of CCW System and directs control room and local actions to isolate thermal barrier for RCP-13.
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| Earthquake with a magnitude of 0.35g will occur (Event 6) causing the loss of startup power. An ATWS will exist due to the magnitude of the earthquake and the failure of the reactor to trip. SFM should direct the CO and BOPCO to perform the immediate actions for the ATWS and direct the opening of MG supply load center breakers to cause the reactor to trip. During the event PORV 456 will open and stay open, power to the block valve is not available, resulting in a SB LOCA. The steam space break results in increasing pressurizer level accompanied by decreasing RCS pressure. The SFM should direct a manual Si prior to an automatic Si at 1850 psig and re-entry into EOP E-0 followed by a transition to EOP E-1.
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| The scenario will be terminated following diagnosis of the RCS not intact, Appendix "C" required to be used due to failure of WR Thot TE-413A (Event 7) and a transition out of EOP E-1 to EOP E-1.2.
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| I
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| r l
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| Appendix D Scenario Outline Form ES-D-2 1
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| Op-Test No.:__1_ Scenario No :._2 Event No.: 1 Page _1_ of _7_
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| Event
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| == Description:==
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| Commence ramp to 100% power l Time Position Applicant's Actions or Behavior BOP Monitor plant parameters RO initiate dilution for ramp to 100% power Setup DEHC
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| . Place MW feedbackin service
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| * Setload reference e Setload rate Raise VPL(Valve Position Limit)
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| Commence ramp to 100%
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| SFM Review precautions and !!mitations of OP L-4 and conduct tumover briefing Direct RO to commence a ramp to 100% power at 3-5 MW/ min l
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| 1 l
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| . I 1
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| NUREG-1021 1 Interim Rev. 8. January 1997 L- !
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| 1 1
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| l l
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| Appendix D Scenario Outline Form ES-D-2 ,
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| l l
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| Op-Test No.:_1 and 2_ Scenario No.: 2 Event No.: _2 Page _2_ af _7_
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| l Event
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| == Description:==
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| VCT Makeup valve FCV-111B Failure l Time Position Applicant's Actions or Behavior BOP Recognize indications of CVCS makeup system failure and report status to SFM
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| . Verify correct Dilution lineup l
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| . Recognize and report failure of FCV -1118 to open when required Take actions as directed by SFM e Coordinate with RO to complete Dilution
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| . Attempts to manually open FCV-111B, reports failure of valve to open to SFM RO Recognize indications of CVCS makeup system failure and report status to SFM
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| = Acknowledge CVCS MAKEUP DEV!ATION alarm, PK05-11, and report to SFM
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| . Verify the makeup system switch alignment and integrator settings
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| . Uses attemate Dilution Path to dilute the RCS l
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| Recommend placing DEHC in hold if ramp was started SFM Acknowledge BOPC0/R0 on CVCS MAKEUP DEVIATION alarm and refer to PK05-11 Determine a malfunction of the reactor makeup control system
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| . Go to OP AP-19, Malfunction of Reactor Makeup Control System
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| . Direct BOPCO to Manually open FCV-111B Direct RO/ BOP to use Altemate Dilution per OP B-1 A:Vil l
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| Direct Control Room Asset team to investigate failure of FCV-111B NUREG-1021 ? Interim Rev. 8. January 1997
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| I ,
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| Appendix D Scenario Outline Form ES-D-2 l
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| Op-Test No.:__1 Scenario No.: 2 Event No.: 3 Page _3_ of _7_
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| Event
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| == Description:==
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| RM-44A fails high Time Position Applicant's Actions or Behavior l
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| BOP Recognize symptoms of a radiation monitor failure J
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| . High Rad & Rad Monitor failure alarms
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| . CVlinitiated
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| . RM44A overranged
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| . CVI Auto actions Place CFCU drain collection system in service RO Recognize symptoms of a radiation monitor failure
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| . High Rad & Rad Monitorfailure alarms
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| . CVl!aitiated Take actions as directed by SFM l
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| SFM Acknowledge reports from BOP /RO, refer to ARP PK02-06,11-21,11-22 & 01-17 Recognize that the CVI caused low flow to RM-11 & 12 Direct Control Room Asset team to investigate failure of RM-44A Consult Technical Specification 3.4.6.1.c j
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| . Direct CFCU drain collection system be placed in service
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| . 25 hour action to take grab samples of the containment atmosphere i
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| I NUREG-1021 3 Interim Rev. 8, January 1997
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| l I
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| Appendix D Scenario Outline Form ES-D-2 l
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| l Op-Test No.:_1_ Scenario No.: 2 Event No.: __4 Page _4_ of _7_
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| Event
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| == Description:==
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| NIS ChannelN42 Fails High Time Position Applicant's Actions or Behavior BOP Defeats NI 42 at NI panel per SFM direction per AP-5 Attachment 4.1 instructions
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| . Defeats Rod Stop l
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| . Defeats Power Mismatch
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| . Defeats QPTR e Defeats Comparator Reports Channel 42 defeated to SFM and R0 RO Recognizes and reports rod motion concurrent with NI 42 failure
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| . Places Rod Contblin Manuall
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| -"CriticalTask Recovers Tavg
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| . Using Rod Control- using 3 step pull and wait or
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| . Adjustment of Turbineload SFM Directs RO to place Rod Controlin Manual Enters OP AP-5
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| . Directs BOPCO to defeat NI-42 using attachment 4.1
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| . Directs RO to recoverTavg
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| . using 3 steps pull and waitor
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| . Adjustment of Turbineload Contact Control Room Asset team to troubleshoot and repair NI-42.
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| Consult Technical Specifications,
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| . 3.3.1 - 6 hours to place tripped condition
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| . 4.2.4.2 - QPTR - MIDS runs required within 12 hours NUREG-1021 4 Interim Rev. 8, January 1997
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| l i
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| i i
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| J Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1 and 2_ Scenario No.: 2 Event No.: __5 Page __5_ of _7_
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| Event
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| == Description:==
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| RCP 1-3 Thermal Barrier Leak Time Position Applicant's Actions or Behavior BOP Reco0nize and report increase in RCP 1-3 sealinjection flow Recognize and report increase in RM-17A & B indications !
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| I Recogni e and report ir uease in CCW head tank level l Recognize and report RCV-16 closed for CCW Head tank l h v. - >%
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| !Take .u, actisdisidiscled
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| . anu my p xb/SFMi@ . Mg: ~ j , ?*,.A, e,
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| 4 y'e 6 Isolate CCW 16 thermal barr;brs0using' FCV-750 - < f i -
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| wc . :. : .m m ,
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| s _ d,.
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| "CriticalTaskQ w [~f ' , . ' '.%R . '
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| .. a R0 Recognize and report increase in charging flow Determine leakage to CCW system of ~30gpm -
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| SFM May Go to PK05-03 for increasing RCP seal flow - determines no alarm for high flow Go to AP-1 for Excessive Reactor Coolant Leakage ,
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| I Direct RO to determineleak size Direct BOPCO to check possible leakage paths Direct BOPCO to Check CCW parameters e CCWHead tank Level e CCW Rad Monitors e RCV-16 _
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| Determine leak is in the CCW System and Go to AP-11, Section B Determine leak is a thermal barrier leak
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| - - , , , , n.n m,n.,,.- ,,wn- ~.w mun mwmx
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| .,-,n-mBOPCO W Dir6ct 16,Cidse CCW FCV-750 for'RCP Thermal bahidrsi ' ,P W'
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| .asac > , : m ir w n y w ~- c:n,: v w TCdticalTask}$+4WyseMy MMM ##08%!ikgp%Km a i c i Contact NSSS Asset team for repair of thermal barrier NUREG-1021 5 interim Rev. 8, aanuary 1997
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| r.
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| e l
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| i I
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| Appendix D Scenario Outline Form ES-D-2 l Op-Test No.:_1_ Scenario No.: 2 Event No.: 6&7 Page _6_ of _7_
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| l Event
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| == Description:==
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| ATWS on Earthquake with failed open PORV Time Position Applicant's Actions or Behavior BOP Performimmediate actions of EOP E-0 recognize failure of rods to insert and report status to SFM myammmmmmemmmrvemm~nrw-mwn:R mN %7 m y m u p p sv a m p t y g y w s p p;' /w# :: pten deenergisEbuse .
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| ELMMb$biSEl$di$$NS2$3!bddiNN25 NS I e1 $[n2 [i I
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| . re-energize buses 13D and 13E after rods are fully inserted Recognize and report PORV 456 open with low pressure
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| . Attempt to close PCV 456, reports valve failed as is and block valve without power l Perform Appendix E, Secondary and Auxiliaries status RO Performimmediate actions of E0P E-0
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| . recognize failure of rods to insert and report status to SFM e attempt manualreactortrip 55bMN
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| [w.mwng g$[h[i[I5t[a[$3g.48$fh y;,. N Y jp ,
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| ,g ,. ,
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| [$& b N N5 -
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| &..m.m,4m u .,.eucyy.glt_b 4 cwa m,.ww:aus )$,'I,I <al,*i,,u&
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| ,_( 1,'cl.,l 'll_f_,
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| _ lQv Ry
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| ~ "
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| s.
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| :[h?Q_]s)~~ wa% Q-
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| .um c - A u v:- n em ~ a e inform SFM when reactor trip breakers indicate open Recognize and report PORV 456 open with low pressure i
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| Recognize and report failure of WR Thot failing high TI-413A i
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| i i
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| l I
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| i l
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| I
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| ! NUREG-1021 6 Interim Rev. 8. January 1997
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| T l
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| l t
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| J Appendix D Scenario Outline Form ES-D-2 Op-Test No.: 1 Scenario No.: 2 Event No.: __6&7 Page _7_ of _7_
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| Event
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| == Description:==
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| ATWS on Earthquake with failed open PORV (Continued) l i
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| l Time Position Applicant's Actions or Behavior SFM Recognize failure of rods toinsert e v.mnm.,, ,c +
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| mmwnawnen-mmsxnmnwgpd,% ~v2a ~m
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| %w W
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| #gdirect RO to insert iods in manuall g!W:e r m
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| %wam yg g w .:.. g g e L a t, n.!@%
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| $e r' direct
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| . ;;- BOP,ydeenergize 480V buses 113D andg13Eid;?g 16 w1 m.**
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| _,Criti_ cal:Ta_sikid;m,.gw 9E.~uam y ygy
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| .&.4<
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| Lg.G,.w%_ww%,
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| 2 m.is
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| , g ; +; &an,;: #v ; W,
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| _M -- : % Aa _ - .-
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| e direct local opening of the reactor trip breakers e direct re-energizing buses 13D and 13E when rods are fully inserted Go to E-0 and direct immediate actions to be completed Direct RO/BOPRO to close PCV-456 and Block valve Direct BOP to perform Appendix E, Secondary and Auxiliaries status Transition to EOP E-1 based on Open PORV with no block valve e conduct tailboard prior to entering E-1
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| . Direct STA to monitor CSFSTs Determine RCS subcooling using Appendix "C" Direct transition to EOP 1.2 based on RCS pressure decreasing
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| . conduct tailboard prior to entering EOP E-1.2 i
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| )
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| l NUREG-1021 7 Interim Rev. 8, January 1997 l 1
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| ]
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| I Appendix D OperatorActions Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 3 Op-Test No.: 1A/1B !
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| I Examiners: Operators: SRO-U / SRO-U j SRO-U / SRO-U i
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| , SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to a 10% dump valve failure Evaluate the crew's ability to diagnose and respond to a SJAE Radiation monitor failure Evaluate the crew's ability to diagnose and respond to a Loss of DRPl indication Evaluate the crew's ability to diagnose and respond to a Loop 4 Delta T failure Evaluate the crew in using the AOP and E0P to respond to SG tube leak Evaluate the crew's ability to diagnose and respond to a SG tube rupture l Evaluate the crew's ability to respond to a seismic event Evaluate the crew 's ability to diagnose and respond to a failure of the MSIVs to close initialConditions: 2% power, enullibrium xenon, end of cycle (IC-40) i Tumover: Continue startup at step 6.5 of L-3 i
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| Time Event Malf. Event Event l min No. No. Type
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| * Description l l
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| 0 1 n/a N, BOP Transfer House loads to startup power j 5 2 xmt I,RO Loop 4 Delta T Fails high res142 10 3 ovr I, BOP SJAE RM-15R failure due to loss of power xdmiO33p 15 4 seli I,SFM 0.1g Earthquake causes loss of DRPI Indication and PY-15 {
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| dse rodi l dse eps29 20 5 mss 3 C, BOP PCV-20 fails full open, will close with backup air system M,ALL 25 6 res4b SG tube leak 0-200 gpm over 20 minutes l Cond 7 mss 7-10 C, BOP MSlVs failto close from control room ;
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| Post pn3 C,RO Pzr Spray Valve PCV-455A fails open, will close in manual Trip G!)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
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| 1 1
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| SCENARIO 03 OVERVIEW The crew is directed to continue the plant startup using OP L-3. First action for the crew will be for the BOPCO to transfer house loads to startup (Event 1) per step 6.5 of OP L-3. '
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| During the transfer to startup, Loop _4 DeltaT fails High (Event 2) causing alarms associated
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| <ith the failure. CO and BOPCO should report the failure to the SFM. The SFM should enter
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| ;P AP-5 for the failure and direct the CO to defeat the input to Delta T for the failed channel.
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| Steam Jet Air Ejector (SJAE) radiation monitor RM-15R will fail due to a loss of power (Event 3). CO and BOPCO should report to SFM the failure indications of the radiation monitor and the SFM should reference AR PK11-07 for the SJAE radiation monitor failure. The SFM should contact T_M to determine the cause of the alarm and an operator to determine the status of the RM-15R input to SPDS.
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| Earthquake with a magnitude of 0.01g will occur which will result in normal supply breaker for DRPI tripping open, and the loss of all DRPI indications, additionally PY-15 output bkr will fail open (Event 4). CO and BOPCO should report the loss of DRPI and the PY-15 to the SFM.
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| The SFM should refer to annunciator response procedure PK03-21 and determine the Tech Spec 3.0.3 applies and directs local operations to place DRPI on backup power supply. SFM i should mfer to OP AP-4 for loss of PY-15 and direct actions to restore PY-15. SFM may enter CP M-4 in response to the earthquake.
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| PCV-20,10% steam dump valve will fail open (Event 5) causing Tavg to decrease and the 40% dump valves to modulate closed. The SFM should direct the BOPCO to manually close PCV-20 at the controller, when it is determined that the controller is failed, the SFM should direct the BOPCO to use the backup air source to close the PCV. The SFM should direct the CO to control TAVG using the 40% dumps once PCV-20 is closed.
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| Steam Generator 1-2 will develop a tube leak (Event 6). CO and BOPCO should identify the indications of the tube leak and report them to the SFM. SFM should enter and take actions per AP-3 for a Steam Generator Tube Failure. CO should determine when leakage is greater -
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| than 50 gpm and report to the SFM. SFM should direct a Manual Safety injection and the performance of E-0 immediate actions. SFM may direct the BOPCO to perform an early isolation of SG 1-2 using E-3, Steam Generator Tube Rupture.
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| While performing E-3 the BOPCO should recognize the failure of the MSIVs to close (Event 7).
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| The failure should be reported to the SFM. The SFM should direct the actions of E-0 and transiun to E-3, When it is determined the MSIVs will not close remotely, local operators shoulo be dispatched to close the MSIVs locally. With isolation from intact SGs not possible the SFM should transition to ECA-3.1, SGTR with Loss of Coolant - Subcooled recovery desired. During performance of recovery actions PCV-455A fails open requiring the Reactor Operator to manually close the failed open valve.
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| Once the SFM tcansitions to ECA-3.1 the MSIVs will be closed and the scenario terminated when a cooldown is started.
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| c ;
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| l Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_. Scenario No.: 3 Event No.: 1 Page __1_ of _8_
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| l Event
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| == Description:==
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| Transfer House Loads to Startuo l
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| l Time Position Applicant's Actions or Behavior BOP Transfer Buses to Startup Power l
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| . 4 KV per OP-J6A:ll
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| . 12 KV per OP-J5:ll >
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| l i
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| R0 Monitorpl ant parameters j l
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| I SFM Prepare for Transfer of bases by reviewing OP L-3 with crew ,
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| l 1
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| Direct BOPCO to transfer 4 and 12 KV buses to startup NUREG 1021 1 Interim Rev. 8. January 1997
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| Appendix D Scenario Outline Form ES-D-2 l
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| l Op-Test No.: _1__ Scenario No.: 3 Event No.: 2 Page _2_ of _8_
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| Event
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| == Description:==
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| Loop 4 AT fails high Time Position Applicant's Actions or Behavior BOP Recognize and report loop 4 AT failure (TI-441 A) l R0 Take actions as directed by SFM
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| = Deselectloop 4 for ATinput SFM Go to PK04-01, RCL AT Deviation Go to OP AP-5 and direct operator recovery actions l
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| I Direct RO to defeat only AT forloop 4 i
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| Consult Technical Specification 3.3.1
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| . 6 hour action to trip bistables for OPAT and OTAT trips Direct Control Room Asset team to investigate failure Loop 4 AT t
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| NUREG-1021' 2 Interim Rev. 8. January 1997
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| 1 l
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| l l
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| Appendix D Scenario Outline Form ES-D-2 l Op-Test No.:_1_ Scenario No.: 3 Event No.: 3 Page _3_.of _8_
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| Event
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| == Description:==
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| SJAE Rad Monitor Failure RE-15R l
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| l Time Position Applicant's Actions or Behavior BOP Identify and report PK alarms and indications associated with the RE-15 R Failure
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| . RM-15R indications on RAD Monitor panel behind Vertical Boards
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| . Reports RM-15R with no power indications and RM-15 normal indicationi, 1
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| I RO Identify and report PK alarms and indications associated with .he RE-15 R Failure
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| . PK11-07, SJAE Rad monitor failure l l
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| l l
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| SFM Go to PK11-07, SJAE Rad monitor failure Direct operator actions
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| . BOPCO to SJAE monitor on Rad monitor panel behind Vertical Boards ,
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| . Outside operator to verify position of the SPDS selector switch in Cable Spreading Room n
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| . Consult ECGS - 39.2
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| . Restore the inoperable monitor to operable status within 30 days Direct Control Room Asset team to investigate failure of RM 15R NUREG-1021 3 Interim Rev. 8, January 1997
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| : m. ,
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| 1
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| . Appendix D _ Scenario Outline Form ES-D-2 )
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| ~
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| Op-Test No.:_1_ Scenario No.: -3 Event No.: 4 Page 4 of _8_
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| , Event
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| == Description:==
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| Small Earthouake with loss of DRPI and PY-15
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| ~
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| Time Position - Applicant's Actions or Behavior BOP Recognize and reports conditions for Loss of PY-15
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| . Numerous unrelated alarms
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| . Feedpump startup stations are dead
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| . MSR valve position Indications are dead Determines size of Earthquake and reports to SFM - 0.1 g R0 Recognizes and reports loss of Digital Rod Position Indication
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| . Loss of DRPI indication on Vertica! Board 2
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| . Confirms that no power change in progress
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| ]
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| SFM. Go to PK03-21 and directs actions for loss of DRPI
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| . Directs local operator to investigate loss of DRPI and place on Backup Per OPA-3:1 Consults Tech Specs
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| . Determines 3.0.3 applies until DRPI is restored Go to AP-4, Loss of Vita! or Non-vital Instrument AC, section B, and direct actions for loss of PY-15
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| . Direct local actions to restore PY-15 with use of backup power supply Direct BOPCO to determine size of earthquake at EFM instrument Direct Control Room Asset team to investigate failure of DRPI and PY-15 power supplies Go to CP-M-4 for Earthquake response l NUREG 1021 4 Interim Rev.8 January 1997 I
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| ~
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| j
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| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 3 Event No.: 5 Page _5_ of _8_
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| Event
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| == Description:==
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| Steam Generator 10% Steam Dumo failure - Open Time Position Applicant's Actions or Behavior i l
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| BOP Recognize and report SG 10% PCV 20 failed open l a
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| n1g - 1 QL 1
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| (^Tilda6tions-asdiredsd'bplS&lilli$:#O?
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| -79ucy i;mn,:w W.m ? A5?64 7 ' > IT::/-
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| zL,22:a )
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| M Atteinpt to blose PCV-20 manually at VB4 cbntrollef, r ;eport failure to close h W:yl
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| 'hhkttEihptl6'closePCNWithbb6kalNbhhlkshofthalve bosed , <
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| ' ^
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| @CriticaliaskM m e MM i :N% '
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| l RO Mondor Reactor Power and Report to SFM i
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| l SFM Directs actions for 10% dump valve failed open
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| . Direct RO to monitor Reactor Power mmo . . ; nm:.ww . . .m e. .. .n
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| : x. wcw g; ,, o
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| ,7:=f ne c.; Direct BOPCO to close PCV-20'with controller / determine c_ontrokr failed ?s -.
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| ;, ; ~ L i; W ^ , an.o . .r + ' mp -
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| th p
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| fj;gDirectBOl?COtoclosePCV-20with Back6fairsdpplysystemiggyi WCriticalIaskY c M d ' ' - - 2E N -
| |
| My^ ~ U!D <
| |
| Direct Control Room Asset team to investigate failure of PCV-20 controller NUREG-1021 5 InterimRev.8 January 1997 1.
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| | |
| L l
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| t l e l
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| I l j
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| l' l - Appendix D Scenario Outline Form ES-D-2 l
| |
| t i
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| 4 Op-Test No.:_1_ Scenario No.: 3 Event No.: ._6 Page _6_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| Steam Generator 1-2 tubeleak Time Position Applicant's Actions or Behavior BOP Acknowledge SJAE, Main Steamline high radiation (RM-72) and SIG Blowdown high radiation alarms and report to SFM
| |
| . Check radiation monitors j
| |
| . Check proper valve alignment for S/G blowdown isolation 4
| |
| Take actions as directed by SFM
| |
| = Start CCP pump and Shutdown PDP e Isolate letdown I
| |
| 1 R0 Recognize symptoms of a SGTR and report to SFM !
| |
| l
| |
| . Decreasing PZR level and increasing charging flow .
| |
| i Swap charging pumps from PDP to CCP l l
| |
| Calculate primary to secondary leak at greater than 50 gpm ,
| |
| Take actions as directed by the SFM w,my me , -mmmt .
| |
| M Perform a.ManualSafsty ,e ' w .mj ymry,y,,.m#,@m,-m W9' Q MMijEsN , " ~;b;lnjections?,-
| |
| *W' 0;' f N -
| |
| "Ie "
| |
| SFM Go to PK 11-06, SJAE High Rad Go to AP-3, Steam Generator Tube Leak
| |
| . Direct RO to determine leak rate
| |
| . Direct BOPCO toisolateletdown
| |
| . Direct swap of charging pumps
| |
| :Dir'ect
| |
| ..%.m ROlid Safety Irijic{and]dkio m -
| |
| E06E-0 (whenllesk is detdrminsd.16bE(5$.m)p 4m ..
| |
| ' h hii' W ei 7# %
| |
| WCriticalTssk5 , 4 au
| |
| ,- +
| |
| I l
| |
| l NUREG-1021 6 Interim Rev.8, January 1997
| |
| | |
| i 1
| |
| i i
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| 4 er Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 3 Event No.: 7 . Page _7_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| SG Tube Ruoturel MSIV failure to close 1
| |
| l Time Position Applicant's Actions or Behavior j BOP Complete E0P E-0immediate actions
| |
| . Verify allvital4kV buses energized Perform early isolation of ruptured S/G 1-2 per EOP E-3 e identifythe mptured S/G mn wmn .
| |
| m~.. ,
| |
| .J;isolaIw a
| |
| n,w.-.m,tured cfloWfromthe mp ,
| |
| SIG
| |
| . - (m-n,-mnclose FC,V-37/ECV-42)! .. .
| |
| i ?r -
| |
| , :i7.c,;: r,w;w^
| |
| . _ ma m"I,m L, hL m.e. . uu-~ % ' '{' j .)
| |
| ] 9
| |
| ,T Y\ A4,* t - '
| |
| ,; d ,' v
| |
| - 1
| |
| . recognize MSIV FCV-42 failure to close and report to SFM
| |
| . Attempt to Close all MSIVs, report failure of a!! MSIVs to clase Perform Appendix E, Secondary and Auxiliaries status RO Complete E0P E-0immediate actions
| |
| . Verify Rx trip
| |
| . Wify Turbine trip
| |
| . Check Si actuated l
| |
| Recognizi'EnirepodPZRISira yhlAPCf4554] ailed openUN;w~ 2 N SN -
| |
| gmpmonw
| |
| .gClose PZR Spray e ,no.
| |
| ww valve ;'msnuallfat.CC21 PCV455A y.
| |
| 4%n ..acw .
| |
| : a. * *;"n m#-5, >
| |
| g w> -% ,
| |
| i
| |
| ;**CptiC3lJggkj $NY/d'," , ?$ ;fM;? Q{ Y[' if ,
| |
| 3 RQ > ; [U$ @
| |
| i NUREG-1021 7 Interim Rev. 8 Jviuary 1997
| |
| | |
| m l
| |
| Appendix D Scenario Outline Form ES-D-2 l
| |
| 1 Op-Test No.:_1_ Scenario No.: 3 Event No.: 7 Page _8_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| SG Tube Ruoturel MSIV failure to close (Continued) -
| |
| l Time Position Appibaat's Actions or Behavior !
| |
| SFM Enter E-0 Direct RO/BOPCO to perform immediate actions ,
| |
| l l
| |
| , ~ .- . , 1
| |
| .~,.~-..a, ,. .. ,.
| |
| l Direct BOPCO to perioMig, !!'s'olatish. ofSGS2 .,- ,. - , ,i, s>
| |
| hD Nkh$l1hbjk,$bbbfh)erE-3 yy;s weu;gy , ..
| |
| steps ih
| |
| 'y ,'ysh; 3and4W b mh ; ~;.hj s
| |
| %w.m : ion ~to close all..MSIVs (AppendiiL) N-g Ldcalact " 1
| |
| .ODirect >w 1
| |
| ne a
| |
| . LYM .?l-') d rt ;$ . 2 Direct BOP to perform Appendix E, Secondary and Auxiliaries status
| |
| )
| |
| _n , nn . ~ w,,,, n. .. ., . .. v ,,
| |
| Direct RO to close failed open PZR Spray Vahie PCV 455A manually.
| |
| .? c . . , . . ~~ -
| |
| y-l
| |
| " CriticalTaskfM 'W7',i ' ,
| |
| . t. ,
| |
| & t,2 Transition to E0P E-3 based on high radiation (Steamline, SJAE, or Blowdown) e conduct tailboard prior to entering E-3
| |
| . may elect to conduct early isolation following immediate actions of EOP E-0
| |
| . Direct Monitoring of CSFSTs i Direct transition to E0P ECA-3.1 based on ALL MSIVs failed open (RNO for Step 3 of E0P E-3) e conduct tailboard prior to entering EOP ECA-3.1 e initiate RCS cooldown l
| |
| . NUREG-1021 8 Interim Rev. 8, January 1997
| |
| | |
| r Appendix D OperatorActions Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 4 , Op-Test No.: 1A/1B Exambers: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l i Objectives: Evaluate the crew's ability to diagnose and respond to a Condensate system disturbance Evaluate the crew's ability to diagnose and respond to a Pressurizer master controller failure Evaluate the crew's ability to diagnose and respond to a ASW pump 0. C. Trip Evaluate the crew in using the AOP and E0P to respond to full load rejection !
| |
| I Evaluate the crew's ability to diagnose and respond to a Loss of startup rawer Evaluate the crew's ability to diagnose and respond to a Loss of all AC event Evaluate the crew 's ability to diagnose and respond to a safety valve failure Initial Conditions: 100% power, equilibrium xenon, Beginning of cycle (IC-1)
| |
| Tumover: D/G 1-1 is 00S, for Lube oil heater replacement, CFCU 1-4 failed 10 minutes ago, TS have not been addressed Time Event Malf. Event Event min No. No. Type
| |
| * Description 0 1 none N, Co STP R-1 A Exercising Full Length Control Rods 12 2 cnd4 R, RO FCV-55 fails open l
| |
| 20 3 pzr4 t, RO Pressurizer Master pressure controller HC-455K fails High - manual j available !
| |
| 28 4 aswi C, BOP ASW pp 1-10.C. trip also trips 4Kv F Bus on 0.C., no auto start on l asw2 standby ASW pp1-2, can be started manually.
| |
| eps4c,1 j 37 5 syd1 M, ALL 100% full load rejection due to 500 Kv system disturbance l 38 6 syd1 M, ALL Loss ef startup power due to system disturbance I degib C, BOP Diesel generator 1-2 fails to start, can be started manually Post trip 7 cyc2 C,Ro CCP 1-2 fails to auto start on retum of Bus G Post trip mss 6d C,RO Main steam safety valve lead four fW 58 stuck open (N)ormal (R)eactivity (1)nstrument (C)omponent (M)ajor QQ /
| |
| | |
| p SCENARIO 04 OVERVIEW -
| |
| 4 Control operator performs of STP R-1 A (Event 1), Exercising full length control rods.
| |
| J Condensate system bypass valve FCV-55 will fail open (Event 2). The RO and BOPCO should recognize the symptoms and identify the failure and inform the SFM of l the transients affects on the plant. SFM should direct the BOPCO to attempt to close i FCV-55 from the control room. SFM should direct the CO to reduco turbine load to reduce reactor power and direct local isolation of FCV-55.
| |
| l Pressurizer master controller HC-455K fails high (Event 3), causing pressurizer spray I valves to open and pressurizer pressure to drop. CO should recognize the failure and i inform the SFM. The SFM should enter Annunciator Response procedure PK05-17 and j transition to OP AP-13 Malfunction of Reactor Pressure control system. Pressurizer I l spray valves should be manually closed prior to a low pressure Reactor Trip.'Pzr Master controller should be controlled in manual and spray valves returned to auto.
| |
| The running Aux. Salt Water pump 1-1 trips on Overcurrent (Event 4). The Overcuurent j trip also causes an Overcurrent Trip on 4Kv bus F causing the loss of the 4Kv bus. The i
| |
| , CO and BOPCO should recognize losses and report them to the SFM. The SFM should enter Annunciator Response procedure PK01-01 and OP AP-10 to restore the ASW system. The SFM should direct the CO/BOPCO to start the standby ASW pump.
| |
| The SFM should reference Annunciator Response procedure PK18-17 and PK18-22 l and direct the actions for the failure of the 4Kv bus F loss including the start of CCW j Pp.1-3 A 500 Kv system disturbance will cause a Full load rejection (Event 5). The CO and BOPCO should recognize loss and report to the SFM SFM should reference Abnormal procedure OP AP-2 and direct the actions of the CO/BOPCO per the procedure. Plant will not st:rvive the Load Rejection and Emergency procedure E-0 should be entered.
| |
| During performance of initial steps of E-0 Startup Power will be lost along with the a failure of DSL Gen 1-2 to stari(Event 6). SFM should recognize the entry conditions for ECA 0.0 and direct the CO/BOPCO in actions for Loss of All AC Power. SFM should direct the Manual starting of DSL Gen 1-2 and the restoration of AC power to Bus G. j SFM should direct the CO to start CCP1-2 which failed to auto start on bus restoration l SFM should determine that with one bus operable a transition to E-0 should be done and direct BOPCO to perform ECA 0.3 actions to restore additional buses.
| |
| During the initial Load rejection a Main Steam line safety valve will stick open (Event 7).
| |
| The BOPCO should determine that the 1-4 Steam generator is not intact and inform the SFM. The SFM should direct the RO to perform en early _ isolation of SG 1-4. SFM should determine the need to transition to EOP E-2 for the faulted SG. SFM should determine that a transition to EOP E-1 is required at the completion of EOP E-2.
| |
| , The scenario will be terminated following transition out of EOP E-1 to EOP E-1.1 Si
| |
| ! Termination.
| |
| | |
| r Appendix D Scenario Outline Form ES-D-2 Op-Test No.:._1_ Scenario No.: 4 Event No.: 1 Page _1._.of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| STP R-1 A Exercising Full Length Control Rods Time Position Applicant's Actions or Behavior BOP Monitors DRPlindications -
| |
| RO Performs STP R-1 A, exercising full length control rods
| |
| . Shutdown banks A, B, C, D.
| |
| . Control banks A, B, C, D Records data on Attachment 8.1 of STP R-1 A SFM- Review precautions and limitations of STP R-1 A and conduct tumover briefing Directs RO to commence exercising full iength control rods Provides direct oversight of control rod manipulations 1
| |
| I l
| |
| 1 NUREG-1021 1 Interim Rev. 8, January 1997
| |
| | |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:__1_ Scenario No.: 4 Event No.: 2 Page _2_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Condensate Bypass Valve FCV-55 fails open Time Position Applicant's Actions or Behavior BOP Report and acknowledge Feedwater Htrs Level Hi and MSR Shell Drips a: arms Recognize and report FCV-55 failed open Attempt to close FCV-55 as directed by SFM and report failure of valve to close RO Recognize and report increase in Reactor Power > 100%
| |
| Retum Rod Control system to Automatic Operate DEHC to reduce turbine load to reduce Reactor power to <100%
| |
| . Place MW feedbacksin service
| |
| . Enter ramp rate and target megawatts
| |
| . Press go when directed by SFM Monitor and report reactor power trend to SFM during ramp SFM Go to PK-10-16 Feedwater Heaters Level High
| |
| . Dispatch local operator to check feedwater heaters levels Direct RO to Place Rod Control to automatic Direct RO to reduce turbine load to maintain reactor power <100%
| |
| Direct actions to close FCV 55
| |
| . Direct 80PCO to close FCV-55 at VB-3
| |
| . Direct Local operators to investigate FCV-55 locally and manually isolate Contact Turbine Building Asset Team for investigation and repair NUREG-1021 2 Interim Rev. 8, January 1997
| |
| | |
| I l
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.: 1 Scenario No.: 4 Event No.: 3 Page _3_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Pressurizer Master Controller failure -HC-455K Time Position Applicant's Actions or Behavior BOP Performs checks on Pressurizer parameters as directed by SFM
| |
| . Monitor and report all pressurizer channels Channel check SAT
| |
| . Monitor and report PORVs and PZR safeties all closed RO P.wnize and report low Pressurizer Pressure alarms Recognize and report conditions of pressurizer controls
| |
| . PZR Htrs off
| |
| . PZR spray valves open
| |
| . Pn master controller HC-455K on CC2 above expected or at maximum demand
| |
| . Place HC-455K to manual and restores pressurizer pressure to NOP
| |
| ** CriticalTasb
| |
| ~ .
| |
| SFM Go to PK 05-17 Pressurizer Pressure Low Go to AP-13 Malfunction of Reactor Pressure Control System Direct BOP and RO to monitor Pressurizer control parameters Diagnose failure of HC-455k to control pressure in automatic
| |
| . Direct RO to Place'HC-455K to Manual and'rettirn Pressurizer pressure to NOP
| |
| ' ?CriticalTaskL / c 4
| |
| .l' Contact Control Room Asset Team for investigation and repair of HC-455K NUREG-1021 3 interim Rev. 8, January 1997
| |
| | |
| r 1 l
| |
| i l l l
| |
| Appendix D Scenario Outline Form ':S-D-2 Op-Test No.:_1_ Scenario No.: 4 Event No.: 4 Page _4_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| ASW PP1-10.C. trio and 4 Kv Bus F 0.C. trio Time Position Applicant's Actions or Behavior BOP. Recognize and report ASW 1-1 trip blue light on
| |
| . Inform SFM of pump trip
| |
| . Identify 4Kv Bus F Overcurrent trip j
| |
| . Identify need to place attemate equipment in service
| |
| [C6niinence plScin6?aitin'se i d;uipmsntiri s6riice~asdirbete' d by SFMF M ,wJ MMSSNhmpjp1-2fb
| |
| . , . . : .~ :- m' IhI ; ,
| |
| s
| |
| &k r
| |
| #: k'>' W y Vig;4 CCWgmpfi-3s l
| |
| ' l
| |
| " C'itica' l Taski . o.~
| |
| ~
| |
| r .x ''
| |
| Secure DieselGenerator 1-3 RO Recognize and report loss of 4Kv Bus F components
| |
| . DRPI
| |
| . CFCUs 1-1 & 1-2
| |
| . CCW pp 1-1
| |
| . ASW pp 1-1
| |
| . Battery Charger 1-1 SFM Acknowledge BOP / CO on loss of 4Kv bus F
| |
| . Reviews actions in Annunciator Response Procedure (PK18-16,17 and 22) for loss of 4kV bus F and loss of 480V bus F.
| |
| Direct starting / placing in service attemate equipment
| |
| . ASW pp 1-2 and CCW pp 1-3
| |
| . Place DRPI en back up
| |
| . Battery 1-1 to charger 1-21
| |
| . Securing of DieselGenerator 1-3 Contact Control Room / Electrical Asset Team for investigation and repair Bus F Refer to Technical Specifications 3.0.3
| |
| . Applies until DRPl is placed on Backup l
| |
| . Applies due to multiple Train A & B components 00S with Diesel Generator 00S NUREG-1021 4 Interim Rev. 8, January 1997
| |
| | |
| t -
| |
| 1 Appendix D Scenario Outline Form ES-D-2 ,
| |
| I 1
| |
| -- j Op-Test No.: _1_ Scenario No.:
| |
| 4 Event No.: 5 Page _5_. of _7_ l Event
| |
| | |
| == Description:==
| |
| 100% Fullload Rejection j l
| |
| 1 Time Position Applicant's Actions or Behavior i BOP Recognizes and reports all steam dump valves open Recognizes and reports ReactorTrip i
| |
| RO Recognizes and reports Main Generator Output bkrs open (Full Load rejection)
| |
| Monitors Plant Parameters l
| |
| * Rod Control- Rods stepping in in Automatic e Reactor Power decreasing
| |
| * Steam GeneratorWater Levels Recognizes and reports Reactor Trip SFM Go to AP-2 Full Load rejection Go to E-0 Reactor Trip NUREG-1021 5 Interim Rev. 8. January 1997
| |
| | |
| I 1
| |
| AppendiyD Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 4 Event No.: 6 Page _6_ of 7 Event
| |
| | |
| == Description:==
| |
| Loss of All AC Power and Bus G restoration Time Position Applicant's Acticns or Behavior BOP Perform immediate actions of E0P E-0 Identify loss of Startup power to all Buses Identify loss of all D/Gs resulting in a loss of power to all Buses Take action as directed by the SFM per EOP ECA-0.0 l
| |
| . Verify AFW flow 5
| |
| '.. . R$fdt5t5idown Relay and Sfait DS{ Gen 32
| |
| ** Criticaifaski Perform actions to crosstie 4 Kv Buses per ECA 0.3 RO Perform immediate actions of E0P E-0 l
| |
| . Verify Rx Trip
| |
| . Verify Turbine Trip
| |
| . 4kV buses energized Take action as directed by the SFM per EOP ECA-0.0
| |
| . Check RCS isisolated
| |
| . REstirt'CCP I1-2 after Bus fGirhtoration' :
| |
| ** CriticilTdsk SFM Directimmediate actions of EOP E-0 Acknowledge loss of all power and transitions to EOP ECA-0.0 Direct the RO I BOP as per the actions of EC/I U
| |
| . Verify Rx Trip
| |
| . Verify Turbine Trip
| |
| . Verify PORVs closed and Letdown isolated
| |
| . Verify AFW finw T Fiessi5h"uidbwri R$layfor'DS'L Gen 52 and Siart '
| |
| !*'* CriticilTaskI . .
| |
| ~
| |
| Direct B0PC0/R0 to v'erify A6tomatic loading of 4Kv bus G - 4-
| |
| .: ) birect RO to restart C'CP 1-25 ,
| |
| ** CriticalTask . ,
| |
| Transition to EOP E-0, Direct BOPCO to imp!ement ECA 0.3 NUREG-1021 6 Interim Rev. 8 Januay 1997
| |
| | |
| [~
| |
| l AppendhD Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 4 Event No.: 7__ Page _7_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Main Steam Line Safety Valve Failure Time Position Applicant's Actions or Behavior BOP Recognize and report main steam header break indications:
| |
| . Increasing steam flow on SG 1-4
| |
| . Decreasing steam pressure on SG 1-4 Perform EOP E-0 Immediate actions ;
| |
| . Verify power to AC emergency buses
| |
| . Check SI actuation RO Perform EOP E-0 Immediate actions
| |
| . Verify reactortrip
| |
| . Verify turbine trip
| |
| . Verify power to AC emergency buses
| |
| . Check SI actuation Recognize and report main steam header break indications:
| |
| . Increasing steam flow on SG 1-4 Isolate the' faulted SG as directed by the SFM (Close MSIV FCV44, AFW supply LCV113.109)
| |
| ** Critical Task-SFM Direct RO / BOP to check for steam leak indications
| |
| . Steam Generator pressures Enter EOP E-0 and direct operator actions Transitions to E0P E-2 and direct isolation of S/G '' 4 .
| |
| (may have bbenisolited harly) _
| |
| "CriticalTask , , a _ , _
| |
| l . conduct tailboard prior to entering E0P E-2 1
| |
| Direct transition to EOP E-1 based on Loss of Secondary Coolant
| |
| . conduct tailboard prior to entering EOP E-1 Direct transition to E0P E-1.1 based on ECCS flow reduction e conduct tailboard pour to entering EOP E-1.1 1
| |
| l l
| |
| I NUPEG-1021 7 Interim Rev. 8, January 1997 l
| |
| | |
| i i
| |
| l i
| |
| I l
| |
| i l
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| / 0 ,
| |
| -0 s ey /
| |
| l t
| |
| I l
| |
| t L_
| |
| | |
| y 1
| |
| REVISION 1 ES-301 Individual Walk-Through Test Outline Form ES-301-2 Facility: DCPP Units 1&2 Date of Examination: 01/25/99 E::am Level: __SRO(l) / SR0(U) Operating Test No.: 1 System /JPM Title / Type Codes
| |
| * Safety Planned Follow-up Questions:
| |
| Function K/A/G-Importance Description
| |
| : 1. RHR/ Transfer to CL recirculation '
| |
| IV a. 005000K4.11 (3.9) Time to reach CL recirculation (M, A, L, S) SRO(1)/ SRO(U) Shutdown b. 005000A2.02 (3.7) RHR solid plant resp to PCV failure
| |
| : 2. Cont Spray / Manually initiate Containment V a. G2.1.33(4.0) Determine system misalignment Spray conditions for entry into T.S. 303 i l
| |
| (D, L, S) SRO(l)/ SRO(U) Shutdown b. 026000K3.01 (4.1) Loss of CSP 1-1 effects on CCS
| |
| : 3. Instrument / R2B NI Calibration Vil a. 015000KS.04 (3.1) Determine effect on calorimetric due to venturierosion. I (D,A,S) SRO(U) ONLY 100% b. 015000K1.01 (4.2) Determine cause of Rx trip due to NImisoperation
| |
| : 4. ECCS / Respond to HI Accumulator level 11 a. 006000A2.03 (3.7) Determine accumulator inleakage (D, S) SRO(I)/ SRO(U) Shutdown b. 006000K3.01 (4.2) Required actions to cooldown/
| |
| depressurize if accumlators are unisolable from RCS
| |
| : 5. Elect. ICrosstie vital bus G and H VI a. 062000A2.01 (3.9) Mode 5 elect bus operation I
| |
| (D, S) SRO(U) ONLY Shutdown b. 062000A1.01(3.8) D/Gloadlimit
| |
| : 6. Rod Control / Verify misaligned rod is not I a. 001000K5.01 (3.7) Determine bank overlap stuck (D, S) SRO(U) ONLY 100 % b. 001000A2.15 (4.2) Determine QPTR
| |
| : 7. PZR / Respond to PZR LO pressure alarm lil a. 010000A1.09 (3.7) Determine tailpipe temperature (N, A, S) SRO(U) ONLY 100 % b. 010000K1.02 (4.1) Effect of Si on BU heater operation
| |
| : 8. AFW / Align attemate AFW from the FWST IV a. 061000K2.01 (3.3) Determine pwr supply to FCV-95 (D,R) SRO(U) ONLY Emergency b. 061000A1.04 (3.9) Determine time in Hot Standby
| |
| : 9. CCW/ Crosstie the CCW system
| |
| . Vill a. 008000A3.05 (3.1) Component isolation by phase B (N, R) SRO(1)/SRO(U) Emergency b. G2.1.52(3.8) CCWS LOCA response onloss of N,
| |
| : 10. EDG/Performlocalstartof a DieselGer.. VI a. 064000K1.04 (3.9) D/G operability (D, A, P) SRO(1)/ SRO(U) Abnormal b. G2.4.30(3.6) Missed surveillance reportability
| |
| * Type Codes: (D)irect from bank, (M)odified from bank, (N)ew, (A)ltemate path, (C)ontr01 room, (S)imulator, (L)ow-Power, (P)lant,(R)CA Y0 NUREG-1021 22 of 26 Interim Rev. 8, January 1997 a
| |
| | |
| C NUCLEA:< POWER GENERATION ,
| |
| DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: LJCNRC-1
| |
| | |
| ==Title:==
| |
| TRANSFER TO COLD LEG RECIRCULATION Examinee: _.
| |
| Evaluator: -
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes ,
| |
| Comments:
| |
| | |
| ==References:==
| |
| EOP E-1.3, Transfer to Cold Leg Recirculation, Rev.16 Alternate Path: Yes X No Time Critical: Yes X No Time Allotment: 20 minutes (must have recire flow to the RCS prior to RWST reaching 4%)
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| Critical Steps: 5,7-10,13-15,18-20,22-26 Job Designation: SRO K/A Number: 005000A4.01(3.4) I 1
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| l l
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| AUTHOR: JOHN BECERRA DA. .:.: 11/12/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
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| f l
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| JPM TITLE: TRANSFER To COLD LEG RECIRCULATION JPM NUMBER: LJCNf3C-1 INSTRUCTOR WORKSHEET l
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| Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken l by the examinee should be clearly demonstrated and verbalized to the l evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed.
| |
| Required Materials: None i
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| Initial Conditions: Unit I experienced a large break LOCA and RWST level decreased to less than 33%.
| |
| Initiating Cue: The Shift Foreman directs you to align RHR for cold leg recirculation, with the pump also aligned to the suction of the SI pumps and CCPs, per EOP E-1.3 Task Standard: RilR is aligned for cold leg recirculation and the RWST is isolated from the CCPs, SIPS and RHR pumps.
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| l I
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| l d
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| g.
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| A
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| , I LJCNRC-1 PAGE 2 OF 11 REV.0 l
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| r-1 JPM TrrLE: TRANSFER TO COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 INSTRUCTOR WORKSHEET Start Time:
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| Step Expected Operator Actions
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| : 1. Obtain the correct procedure. 1.1 Operator references EOP E-1.3.
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| Step was: Sat: Unsat )
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| l
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| : 2. Reset Safety Injection. 2.1 Operator reads CAUTION prior to Step 1.
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| 2.2 Verifies Safety Injection is reset by noting PK08-21 is OFF or ESF equipment monitor red light is OFF.
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| Step was: Sat: Unsat _
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| : 3. Reset both trains of Containment 3.1 Operator depresses Train A and B Isolation Phase A and B. Phase A reset push buttoas, verifies Phase A red lights OFF.
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| 3.2 Depresses Train A and B Phase B reset push buttons, verifies Phase B red lights OFF.
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| Step was: Sat: Unsat
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| : 4. Check ECCS status. 4.1 Operator verifies RHR Pps are stopped, VB1 green light ON.
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| Step was: Sat: Unsat I
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| 1
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJGNRC-1 PAGE 3 OF 11 REV 0
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| 1 JPM TITLE: TRANSFER To Coto LEG RECIRCULATION JPM NUMBER: LJCNBC-1 INSTRUCTOR WORKSHEET
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| : 5. ** Verify ASW and CCW are aligned 5.1 Operator checks both ASW pumps for two pumps through two heat running, VB1 red lights ON.
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| exchanger operation.
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| 5.2 Checks FCV-602 open, red lights ON.
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| 5.3 Positions FCV-603, to OPEN.
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| 5.4 Checks FCV-430 open, red light ON.
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| 5.5 Positions FCV-431, to OPEN.
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| 5.6 Checks reciprocating charging pump.
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| Stopped, VB2 green light ON.
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| Note: Only substeps 5.3 and 5.5 are critical.
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| Step was: Sat: Unsat
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| : 6. Dispatches operator to close breakers. 6.1 Operator dispatches opera' v to close breakers for:
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| 8976 (52-1H-20) 8980 (52-1F-31).
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| Cue: Operator dispatched.
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| I Step was: Sat: Unsat
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| : 7. *
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| * Close 8716A and B, RHR discharge 7.1 Operator selects CLOSE on control I
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| header crosstic valves. switches for 8716A and B.
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| 7.2 Verifies green lights ON and red lights OFF. (Not Critical)
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| Step was: Sat: Unsat
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Cr;tical Step.
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| LJCNRC-1 PAGE 4 0F 11 REV.0
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| F JPM TITLE: TRANSFER To COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 l lNSTRUCTOR WORKSHEET i
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| : 8. *
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| * Close 8700A and B, RHR pump 1 8.1 Operator positions control switches for and 2 normal suction valves. valves 8700A and 8700B to CLOSE.
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| 8.2 Verifies green lights ON and red
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| : lights OFF. (Not Critical)
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| Step was: Sat: Unsat
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| : 9. ** Open 8982B, RHR pump 2 suction 9.1 Operator cuts in series contactor for from containment recirculation sump. 8982B.
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| 9.2 Selects OPEN on control switch for 8982B.
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| 9.3 Verifies valve opens by red light ON and green light OFF. (Not Critical)
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| Step was: Sat: Unsat
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| : 10. ** Open FCV-364, RHR Heat 10.1 Operator positions FCV-364, to exchange 2 CCW outlet valve. OPEN.
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| 10.2 Verifies red light ON, green light OFF. (Not Critical) 10.3 Notes increase on CCW header B flow on FI-48. (Not Critical)
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| Step was: Sat: Unsat j
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| : 11. Checks Containment Recire Sump 11.1 Operator checks Containment Recire Level. Sump Level 2.5' on LI-940 & 941.
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| l
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| * l Step was: Sat: Unsat j 1
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| 1
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| * Denotes an entry required on the JPM cover sheet. I
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| *
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| * Denotes a Critical Step. ;
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| LJCNRC-1 PAGE 6 OF 11 REV.O
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| JPM TRLE: TRANSFER TO COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 INSTRUCTOR WORKSHEET
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| : 12. Start RHR pump 12. ._ .
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| ; 12.1 Operator selects START on RHR l pump 12 control switch.
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| 12.2 Verifies motor starts by amp reading and red start light ON.
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| 12.3 Determines RHR pump 1-2 tripped.
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| Red start light OFF and Blue light ON Step was: Sat: Unsat *
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| : 13. ** Close 8974A and B, SI pump 13.1 Operator cuts in series contactor for recirculation valves. 8974A and B.
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| 13.2 Selects CLOSE on control switches for 8974A and B.
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| 1 13..i Verifies valves closed by green lights ON and red lights OFF. (Not Critical)
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| Step was: Sat: Unsat
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| : 14. ** Close 8105 and 8106, centrifugal 14.1 Selects CLOSED on control switches charging pump recirculation valves. for 8105 and 8106.
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| 14.2 Verifies valves closed by green lights ON, and red lights OFF.
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| Step was: Sat: Unsat i
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-1 PAGE 6 OF 11 REV.0 l l
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| u
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| JPM TITLE: TRANSFER To COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 INSTRUCTOR WORKSHEET
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| : 15. *
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| * Open 8807A and B, RHR heat 15.1 Operator selects OPEN on control exchanger to SI pump 11. switches for 8807A and B.
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| l 15.2 Verifies valves open by red light ON and green light OFF. (Not Critical) 15.3 Notes increased charging pump flow and pressure. (Not Critical)
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| Step was: Sat: Unsat l
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| : 16. Verify CCPs running 16.1 Checks centrifugal charging pumps 11 and 12 are running, red lights ON.
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| Note: Procedure step 7.d. is NA. RIIR pp 1-2 is not running since it tripped i on OC.
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| Step was: Sat: Unsat
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| : 17. Verify Closed 8700A. 17.1 Operator locates and verifies 8700A closed.
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| Step was: Sat: Unsat
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-1 PAGE 7 OF 11 REV.0 u
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| j l
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| l JPM TrrLE: TRANSFER TO COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 INSTRUCTOR WORKSHEET 18.1 Operator cuts in series contactor for
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| : 18. ** Open 8982A, RHR Pp 1-1 suction 8982A.
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| from containment recirc sump.
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| 18.2 Selects OPEN on control switch for 8982A.
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| 18.3 Verifies valve open by red light ON and green light OFF. (Not Critical)
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| Step was: Sat: Unsat
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| : 19. ** Open FCV-365, CCW Outlet valve 19.1 Operator positions FCV-365, to from RHR Hx l-1. OPEN.
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| 19.2 Verifies red light ON, green light OFF. (Not Critical) i l
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| Step was: Sat: Unsat
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| * I l
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| : 20. ** Check RHR Pp 1-2 running. 20.1 Operator determines that the RNO column is appropriate since RHR Pp 1-2 tripped on OC.
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| Step wa.s: Sat: Unsat *
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| : 21. Check Containment Recire Sump 21.1 Operator checks Containment Recire Level. Sump Level greater than 92.5' on LI-940 and LI-941.
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| Step was: Sat: Unsat I
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| f
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| * Der,otes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-1 PAGS 8 CF 's 1 REV.0
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| r 1
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| , JPM TirLE: TRANSFER To COLD LEG RECIRCULATION JPM NUMBER: LJCNRC-1 l i INSTRUCTOR WORKSHEET
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| : 22. " Start RHR pump 1-1. 22.1 Operator selects START on RHR pump 1-1 control switch.
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| 22.2 Verifies motor starts by amp reading and red start light ON.
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| 22.3 Notes flow to the Reactor Vessel on FI-970A and FI-970B (Not Critical) 22.4 Notes no indication of cavitation on RHR pump 1-1.
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| Step was: Sat: Unsat
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| : 23. ** Open 8804A, RHR Hx l-1 to Chg 23.1 Operator selects OPEN on the control l Pps and SI Pps. switch for 8804A.
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| 23.2 Verifies valve opens by red light ON, green light OFF.
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| 23.3 Checks Chg Pp amps and Charging injection flow and S1 Pp flow.
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| 23.4 Throttle HCV-638 for maximum flow and RHR Pp 1-1 amps less than 57 amps.
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| Step was: Sat: Unsat
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| : 24. " Close 8805A and B, Charging Pp 24.1 Selects CLOSED on control switches suction form RWST. for 8805A and 8805B.
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| 24.2 Verifies valves closed by green lights ON and red lights OFF.
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| Step was: Sat: Unsat l
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| IJCNRC-1 PAGE 9 OF 11 REV.0
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| l l
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| JPM TITLE: TRANSFER To Coto LEG RECIRCULATION JPM NUMBER: LJCNF3C-1 !
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| INSTRUCTOR WORKSHEET l
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| : 25. ** Close 8976, SI Pp suction form 25.1 Selects CLOSED on control switch l' RWST. for 8976.
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| 25.2 Verifies valve closed by green light ON and red light OFF.
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| Step was: Sat: Unsat
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| : 26. ** Close 8980, RHR Pp suction form 26.1 Selects CLOSED on control switch RWST. for 8980.
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| 26.2 Verifies valve closed by green light ON and red light OFF.
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| Step was: Sat: Unsat Stop Time:
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| Total Time: (Enter total time on the cover page) i l
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| l l
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| l
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step. .
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| I LJCNRC-1 PAGE 10 OF 11 REV.O j
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| r --
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| JPM TrrLE: JPM NUMBER: LJCNBC-1 EXAMINEE CUE SHEET Initial Conditions: Unit 1 experienced a large break LOCA and RWST level decreased to less than 33%. ;
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| 1 1
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| Initiating Cue: The Shift Foreman directs you to align RHR for cold leg recirculation, .
| |
| with the pump also aligned to the suction of the SI pumps and CCPs, per j EOP E-1.3 Task Standard: RHR is aligned for cold leg recirculatian and the RWST is isolated from the CCPs, SIPS and RHR pumps.
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| I l
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| l l
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| l LJCNRC-1 PAGE 110F 9 REV, 0
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| JPM TITLE: TRANSFER TO COLD LEG RECIRCULATION JPM NUMBER: LJCN13C-1 ATTACHMENT 1 SIMULATOR SETUP O Type "init ljc027" on the expert screen command line. Click the BYPASS SW CliECK button on the expert screen to continue after control boards are aligned.
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| O Insert GCF pmp thr2 4,0,0,5,c,xyli237c,0. RIiR pump 1-2 will trip on overcurrent when the operator attempts to start it.
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| O This SNAP allows entry into EOP E-1.3 at Step 1. RWST level is 33% with both RIIR pumps OFF.
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| O Inform the examiner that the simulator setup is complete.
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| NOTE: This JPM is time critical. Going to RUN too soon may affect the examinee's ability to successfully complete this JPM.
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| O Go to RUN when the examinee references EOP E-1.3.
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| i l
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| )
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| I
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| , LJCNRC-1 PAGE 12 OF 9 REV.0
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| I
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| ; NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT i
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| i JOB PERFORMANCE MEASURE Number: LJCNRC-2
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| ==Title:==
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| INITIATE CONTAINMENT SPRAY MANUALLY Examince:
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| Evaluator: l Print Signature Date Results: Sat Unsat Total Time: minutes ;
| |
| Comments:
| |
| | |
| ==References:==
| |
| EOP FR-Z.1, Response to High Containment Pressure, Rev. 6 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 5 minutes Critical Steps: 3, 4 Job Designation: SRO 1 K/A Number: 026000A3.01(4.5)
| |
| AUTHOR: JOHN BECERRA DATE: 11/12/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BE NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| I 1
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| JPM TITLE: INITIATE CONTAINMENT SPRAY MANUALLY JPM NUMBER: L.'ONf3C-2 INSTRUCTOR WORKSHEET '
| |
| Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the I evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed.
| |
| Required Materials: None l
| |
| Initial Conditions: Unit 1 experienced a LOCA. EOP E-1 is in progress and Safety Injection is reset.
| |
| Initiating Cue: Containment pressure is 25 psig. The STA confirms a MAGENTA path on the Containment Critical Safety Function Status Tree. All higher priority cri'.ical safety functions have been addressed. The Shift Foreman directs you to manually initiate containment spray in response to the MAGENTA path on containment pressure.
| |
| TasP. Standard: Containment spray is initiated.
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| l l
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| LJCNRC-2 PAGE2OF7 REV.O
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| I l !
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| l JPM TrrLE: lNmATE CONTAINMENT SPRAY MANUALLY JPM NUMBER: LJCNSC-2 i INSTRUCTOR WORKSHEET l
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| ! I l
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| ( Start Time:
| |
| l
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| ; Step Expected Operator Actions
| |
| : 1. Obtain the correct procedure. 1.1 Operator references EOP FR Z.l. l i
| |
| Cue: Start with Step 3.
| |
| Step was: Sat:
| |
| * Unsat:
| |
| : 2. Check if containment spray is required. 2.1 Operator checks if EOP ECA-1.1 is the procedure in effect.
| |
| 1 2.2 Determines EOP E-1 is in progress !
| |
| (from initial conditions).
| |
| 2.3 Checks containment pressure greater i than 22 psig. l Step was: Sat: Unsat:
| |
| : 3. *
| |
| * Start the containment spray pumps. 3.1 Operator turns control switches to the START position for containment spray pumps 11 and 12.
| |
| 3.2 Verifies pumps running by amps returning to normal and red light ON.
| |
| Step was: Sat: Unsat: *
| |
| : 4. ** Align containment spra;< discharge 4.1 Operator determines that ECCS is by opening 9001 A and B. aligned for injection flow.
| |
| 4.2 Opens 9001 A and B.
| |
| Step was: Sat: Unsat:
| |
| 1
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
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| LJCNRC-2 PAGE 3 OF 7 REV.O
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| F I
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| i l
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| JPM TITLE: INITI'.TE CONTAINMENT SPRAY MANUALLY JPM NUMBER: LJCNRC-2 INSTRUCTOR WORKSHEET l 5. Verify spray add tank alignment for 5.1 Operator verifies open 8992,8994A, l l injection phase. and 8994B. I Step was: Sat: Unsat:
| |
| Note: Operator may look at flow on l PAM panel for verification of approximately 2600 gpm for each spray header.
| |
| Stop Time:
| |
| Total Time: (Enter total time on the cover page) l l
| |
| l l
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| 1 l
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| l
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| * Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
| |
| LJCNRC-2 PAGE 4 OF 7 REV.O
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| l I i 11 JPM Tm.E: JPM NUMBER: LJCNBC-2 EXAMINEE CUE SHEET Initial Conditions: Unit 1 experienced a LOCA. EOP E-1 is in progress and Safety Injection is reset.
| |
| Initiating Cue: Containment pressure is 25 psig. The STA confirms a MAGENTA path on the Containment Critical Safety Function Status Tree. All higher l priority critical safety functions have been addressed. The Shift Foreman l directs you to manually initiate contaimnent spray in response to the MAGENTA path on containment pressure.
| |
| Task Standard: Containment spray is initiated. ,
| |
| i I
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| l 1
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| LJCNRC-2 PAGE 5 OF 7 REV.O
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| JPM Tm.E: lNmATE CONTAINMENT SPRAY MANUALLY JPM NUMBER: LJCNf3C-2 ATTACHMENT 1. SIMULATOR SETUP O Initialize the simulator to IC-1 (100%, BOL). ;
| |
| O Enter drill file 6052 or manually insert the following:
| |
| Conunand Description ramp ccnmpmin=40,5,0,cjpplcipb Ensures enm pressure >22 psig forjpm j delm penm937 Removes monitor from last run l mony penm937 Monitors containment pressure mal ppl2a act,0,0,d,2 Inadvertent SI, Train A l mal ppl2b act,0,0,d,2 Inadvertent S1, Train B run 120 RUNS simulator for 2 minutes ovr xylil 18r act,1,0,60,c,fnispr.lt.10,15 Reset SI, Train A after 60 secs.
| |
| ovr xylil 19r act,1,0,60,c,fnispr.lt.10,l' Reset SI, train B after 60 secs.
| |
| mal rest act,3,1,61,c,fnispr.it.10,0 DBA LOCA ovr xreiO22h act,1,0,0,c,fnispr.lt.10,5 Reset MSR's i ovr xc3il36m act,1,0,0,c,fnispr.lt.10,5 Takes FWRVs to manual and close ovr xc3il 361 act,1,0,0,c,fnispr.lt.10,0 ovr xc3i137m act,1,0,0,c,fnispr.lt.10,5 ovr xc3il 371 act,1,0,0,c,fnispr.lt.10,0 ovr xc3il38m act,1,0,0,c,fnispr.lt.10,5 ovr xc3il381 act,1,0,0,c,fnispr.lt.10,0 ovr xc3il39m act,1,0,0,c,fnispr.lt.10,5 ovr xc3il391 act,1,0,0,c,fnispr.it.10,0 Svr xv3i2240 act,1,0,0,c,fnispr.it.10,0 Stops CND/BSTR PP l-2 ovr xv2i2600 act,1,0,0,c,fnispr.lt.10,0 Trips RCP 1-1 ovr xv2i2610 act,1,0,0,c,fnispr.lt.10,0 Trips RCP l-2 ovr xv2i2620 act,1,0,0,c,fnispr.lt.10,0 Trips RCP l-3 ovr xv2i2630 act,1,0,0,c,fnispr.lt.10,0 Trips RCP l-4 LJCNRC-2 PAGE 6 OF 7 REV.O
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| | |
| i JPM TITLE: INMATE CONTAINMENT SPRAY MANUALLY JPM NUMBER: LJCNf3C-2 ATTACHMENT 1, SIMUI.ATOR SETUP O Perform the following:
| |
| : 1. Verify MSR valves are closed.
| |
| : 2. Place FCV-53 and 54 in RECIRC.
| |
| : 3. Shut down condensate booster pump set 13, if running.
| |
| : 4. Select CSF-5 on SPDS.
| |
| O Inform the examiner that the simulator setup is complete.
| |
| O - Go to RUN when the examinee is given the cue sheet.
| |
| 1 I
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| \
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| I l
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| LJCNRC-2 PAGE 7 OF 7 REV.O
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| I l Nov 6 10:11 1998 ../istrfile/pge/drillsdir/drl_6052 Page 1 l
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| l
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| * JPM LJCNRC-2
| |
| * 11/06/98, dabl
| |
| * START CONTAINMENT SPRAY AFTER DBALOCA
| |
| * ensure enn press > 22 psig for jpm ramp cenmpmin=40,5,0, c,j pplcipb
| |
| * MONITOR CONTAINMENT PRESSURE delm pcnm937 mony penm937
| |
| * INADVERTANT SI, CLEAR AFTER 2 SEC.
| |
| mal pp12a act,0,0,d,2 mal pp12b act,0,0,d,2
| |
| * START SIMULATOR ON A TIMED RUN
| |
| . _._,.. gg.17 0* " " r ~ ~ ~' ~ ~ ~ ~
| |
| ~ 7'; 7"*""*'""'"'~ -
| |
| ~P e
| |
| * RESET SI 60 SECONDS AFTER INITIATED ovr XVII 118R act,1,0,60,c,fnispr.lt.10,15 fvb1130a ovr XVII 119R act,1,0,60,c,fnispr.lt.10,15 #vb1131a .
| |
| * INITIATE A DBA LOCA 61 SECONDS AFTER BELOW 10 PERCENT POWER mal resi act,3,1,61,c,fnispr.lt.10,0 .,., . .
| |
| 3 j
| |
| ~o .c ,
| |
| * RESET MSRS ovr XREIO22H act,1,0,0,c,fnispr.lt.10,5 #vb3164a
| |
| * TAKE FWRV'S.TO MANUAL AND CLOSE ov'r XC3I136M act,1,0,0,c,fnispr.lt.10,5 #cc3050c ovr XC3I136L act,1,0,0,c,fnispr.lt.10,0 #cc3050f I ovr XC3I137M act,1,0,0,c,fnispr.lt.10,5 #cc3051c '
| |
| ovr XC3I137L act,1,0,0,c,fnispr.lt.10,0 #cc3051f ovr XC3I138M act,1,0,0, c, fnispr. lf.10,5 #cc3052c ovr'XC3I138L act,1,0,0,c,fnispr.lt.10,0 #cc3052f ovr XC3I139M act,1,0,0,c,fnispr.lt.10,5 #cc3053c ovr'XC3I139L act,1,0,0,c,fnispr.lt.10,0 #cc3053f
| |
| * STOP CND/BSTR PP 1-2 ovr XV3I2240 act,1,0,0,c,fnispr.lt.10,0 fvb3062e
| |
| * TRIP RCPS ovr XV2I2600'sct,1,0,0,c,fnispr.lt.10,5 #vb2176e ovr XV2I2610_act,1,0,0,c,fnispr.lt.10,5 #vb2177e ovr XV2I2620 act,1,0,0,c,fnispr.lt.10,5 #vb2178e ovr XV2I2630 act,1,0,0,c,fnispr.lt.10,5 #vb2179e
| |
| . t ,,!, .lan t .I t , , ,m , ) , i i ., . 31a i t i .'] i l '% 1 a 6< A e - o
| |
| . I t. ,
| |
| i, - I 1. i, t ,
| |
| , t 1l i ,O /- . f r 't I- I f .}f ,O f. > ' ?
| |
| .*' f~ ,
| |
| .i . .4 l' . l t. , t) ,d ,f
| |
| ( r , , e i.
| |
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| $ ., . i t-(< 4 1 (8 /* I-
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| i NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT
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| ~ JOB PERFORMANCE MEASURE i
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| Number: LJCNRC-3 i
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| ==Title:==
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| MANUAL OPERATOR IIEAT BALANCE CALCULATION I i
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| Examinee:
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| i Evaluator:
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| I Print Signature Date Results: Sat Unsat Total Time: _ minutes Comments: Provided marked-up copy of STP R-2B2, copy of conversion tables, j ITS-90,91,92 Resistance Tables and OPHB2B Program printout (OPP l PC-4). l
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| )
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| ==References:==
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| STP R-2B2, Manual Operator Heat Balance, Rev. 8 Volume 9, Table T-IIC-27 (Vol. 9), Rev. 9 j Volume 9, Figure F-IIC-2, Rev. 25 l Alternate Path: Yes X No I Time Critical: Yes No X i Time Allotment: 25 minutes i
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| Critical Steps: 3,4 !
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| Job Designation: SRO j t
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| K/A Number: 015000A1.01(3.8 l
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| }.
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| i 4
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| I i
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| AUTHOR: JOHN BECERRA DATE: 11/12/98 l REVIEWED BY: NA DATE: NA !
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| JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
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| JPM TITLE: MANUAL OPERATOR HEAT BALANCE CALCULATION JPM NUMBER: LJCNRC-3 '
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| INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. Tlie examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin.
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| Required Materials: Calculator Initial Conditi:as: Unit 1 is at 100%. The PPC is not in service. STP R-2B2 data acquisition is complete.
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| Initiating Cue: The Shift Foreman has directed you to complete section 13," Data Reduction and Evaluations" of STP R-2B2. Procedure and data verification has been completed through Step 12.
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| Task Standard: Review STP R-2B2, starting at Step 13, and make gain adjustments, as necessary.
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| l l
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| l LJCNRC-3 PAGE 2 OF7 REv. O
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| JPM TITLE: MANUAL OPERATOR HEAT BALANCE CALCULATION JPM NUMBER: LJCNRC-3 INSTRUCTOR WORKSHEET l
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| l
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| ! Start Time:
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| i Step Expected Operator Actions l 1. Obtain the correct procedure. ' .1 Operator obtains STP R-2B2 from instructor.
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| Note: Operator may elect to review precautions, limitations and prerequisites prior to commencing with Step 13 of the procedure.
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| Cue: All required conditions are met.
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| Step was: Sat: Unsat: *
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| : 2. Performs data reduction and evaluation. 2.1 Operator inputs data into "OPHB2B" program.
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| une: Inform Operator that this will not actually be done and give him/her the printed output of OPHB2B. 4 I
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| 2.2 Attaches trend data, etc. I l
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| 1 2.3 Checks NIS % power within 4% of j Core % RTP. I 2.4 Checks NIS % power for each channel differs from Core % RTP and identifies N42 is > 2%.
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| Step was: Sat: Unsat:
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| l l
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-3 PAGE 3 OF 7 REV.0 1'
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| ! JPM TITLE: MANUAL OPERATOR HEAT BALANCE CALCULATION JPM NUMBER: LJCNR,C-3 lNSTRUCTOR WORKSHEET l
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| : 3. * *Detennines that NIS adjustment is 3.1 Operator notifies the SFM that N42 is required. > 2% above Core % RTP and that NIS adjustment is required.
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| Note: Only Step 3.1 is a critical element of Step 3.
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| 3.2 Contacts Reactor Engineering.
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| **********************************.s*
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| Cue: Rx. Engineering approves and recommends adjusting NI-42 ONLY at this time. j SFM gives permission to adjust loop 2 N42 only.
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| 3.3 Obtains Attachment 8.5 to make adjustment to N42 (loop 2).
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| Step was: Sat: Unsat:
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| : 4. Verify Control Rods in Manual. 4.1 Verify Rod Bank Selector switch is selected to MANUAL.
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| Step was: Sat: Unsat:
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| : 5. * *Makes gain adjustment to N42. 5.1 Operator reads and observes CAUTION.
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| 5.2 Records as found Pot setting for N42A.
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| 5.3 Records Indicated meter reading before adjustment. ,
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| 1 Note: Only steps 5.4 and 5.5 are required to satisfy the critical step 5.4 Unlocks gain pot for N42A.
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-3 PAGE 4 OF 7 REV.0 l
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| l
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| JPM TITLE: MANUAL OPERATOR HEAT BALANCE CALCULATION JPM NUMBER: LJCNRC-3 )
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| INSTRUCTOR WORKSHEET l 5.5 Adjusts meter reading 3.2% less than original value.
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| 5.6 Locks gain Pot. ,
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| l 5.7 Records indicated reading after adjustment.
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| 5.8 Records as left Pot setting.
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| Note: The following should be donc quickly as possible (per Att. 8.5) j 5.9 Records final meter readings.
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| 5.10 Records PPC reading.
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| Cue: PPC is out of sen' ice.
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| I 5.11 Reads Control console readings. l 5.12 Checks indicators agree within 1.8%.
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| 5.13 Checks allindicators agree within 2%.
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| . Note: If 5.11 and/or 5.12 above don't !
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| agree, Operator should indicate an AR is to be written.
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| Cue: Performance of a 2dheat balance is not required.
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| Step was: Sat: Unsat:
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| Stop Time: _
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| Total Time: (Enter total time on the cover page) ,
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| * Denotes an en'ry required on the JPM cover s
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| *
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| * Denotes a Critical Step. ~
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| LJCNRC-3 PAGE 5 OF 7 REV.O
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| JPM TITLE: JPM NUMBER: LJCNRC-3 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is at 100%. The PPC is not in service. STP R-2B2 data acquisition is complete.
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| Initiating Cue: The Shift Foreman has directed you to complete section 13 " Data Reduction and Evaluations" of STP R-2B2. Procedure and data verification has been completed through Step 12.
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| Task Standard: Complete STP R-2B2, starting at Step 13, and make gain adjustments, as necessary.
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| 1 I
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| 1 l
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| LJCNRC-3 PAGE 6 OF 7 REV.O
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| JPM TITLE: MANUAL OPERATOR HEAT BALANCE CALCULATION - JPM NUMBER: LJCNf3C-3 ATTACHMENT 1, SIMULATOR SETUP 0 Initialize the simulator to IC-01 (100%, BOL).
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| O Perform the following:
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| : 1. Go to RUN
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| : 2. Adjust the power range meters with the gain adjust to indicate the following:
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| N41 = 100.6 (6.04 turns)
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| N42 = 102.7 (7.15 turns)
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| N43 = 101.1 (6.62 turns)
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| N44 = 101.3 (6.33 tums)
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| O Freeze the simulator and inform the examiner that the simulator setup is complete.
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| O Go to RUN when the examinee is given the cue sheet.
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| l LJCNRC-3 PAGE 7 OF 7 REV.O
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| [ l l- :
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| <. *** ISSUED FOR USE BY: DATE: EXPIRES: ***
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| PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 NUCLEAR POWER GENERATION REVISION 9 DIABLO CANYON POWER PLANT PAGE 1 OF 11 SURVEILLANCE TEST PROCEDURE UNITS TITLE: MANUAL OPERATOR IIEAT BALANCE AND l
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| 06/12/98 I EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED
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| : 1. SCOPE l 1.1 Thir procedure gives direction for performing a manual secondary side heat balance. i The results of the heat balance (given in terms of reactor power level) are then used to I verify the calibration of the NIS power range channels.
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| 1.2 If the results of the heat balance indicate that a channel's calibration doer not meet the Acceptance Criteria, then a channel adjustment is required via Data Sheet 5. Occasions will occur where the Acceptance Criteria are met, but adjustment is desirable. In these situations, NIS adjustment may be made at the discretion of the Shift Foreman (SFM).
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| 1.3 With an evaluated amount of feedwater nozzle fouling, applying the Nozzle Fouling Factor (NFF) will decrease the calculated power to correct for nozzle fouling. To ensure that the calculated power is not overcorrected, when nozzle defouling has occurred or is expected to occur, the NFF will be verified acceptable or will be reset to 1.0.
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| : 2. RESPONSIBILITIES 2.1 SFM (Test Director), for obtaining data, data reduction, and reporting of results.
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| 2.2 Shift Technical Advisor (STA) for adjustment of nuclear instrumentation (if required).
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| 2.3 Power Production Engineer (PPE) (Reactor Engineering) for review of results and evaluation of NIS adjustment.
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| 2.4 Senior Power Production Engineer (Reactor Engineering) (Test Coordinator) for review of results involving NIS adjustment.
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| : 3. FREOUENCY 3.1 This procedure shall be performed daily when the reactor powei :evel is above 15 %
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| RATED THERMAL POWER and STP R-2B1 cannot be performed or as required by STP R-2Bl. Either this procedure or STP R-2B1 is required within 24 hours of exceeding 15 % RTP or prior to exceeding 30% RTP during Reactor Plant startup.
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| l 01051209. DOC 06 1
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| PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 2 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2 i
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| : 4. TECHNICAL SPECIFICATIONS 4.1 Technical Specification 4.3.1.1 requires that each reactor trip system instrumentation channel l- demonstrated OPERABLE per Table 4.3-1, Item 2.a. Successful performance of this procedure satisfies the channel surveillance requirement for the high trip setpoint of the Nuclear Power Range Channels. NIS channel adjustments are I required if NIS channels differ more than 2% from the heat balance calculated "% '
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| core power."
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| : 5. ACCEfrTANCE CRITERIA 5.1 If the power level indicated by any OPERABLE power range NIS channel differs from the calculated core power level (% RTP) by more than 2%, adjust the NIS channels to within i2% of the heat balance calculated "% core power."
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| 5.2 If the power level indicated by any OPERABLE power range NIS channel is more than 4% below the calculated core power level (% RTP), reduce power level by the amount the calculated core power level exceeds 100% and initiate an Action Request (AR).
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| Perform another STP R-2B2 and adjust the NIS channels to within i2% of the heat balance calculated "% core power."
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| : 6. REFERENCES 6.1 PG&E Drawing No. 663218-42, "W Nuclear and Temperature Instrumentation Calibration and Thermal Power Measurement."
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| 6.2 Technical Specifications, Table 4.3-1.
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| 6.3 AC R-2A, "High Accuracy Heat Balance."
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| 6.4 AC R-2B, " Operator Heat Balance."
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| 6.5 DPP PC-4, "OPHB2B Computer Program."
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| 6.6 AC DPP PC-4, "OPHB2B Computer Program."
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| 6.7 WCAP-11594, " Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Calculations for DCPP Units 1 and 2."
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| 6.8 AC R-14, " Periodic Adjustment of Feedwater Flow Nozzle Coefficients."
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| 6.9 PPC Manual,550-9600000-14.
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| 6.10 PEP 03-02, " Determination and Validation of the Nozzle Fouling Factor."
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| 6.11 OP O-15, " Control of Phn' Process Computer Addressable Points."
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| 6.12 NSSS Calculation N-212, " Low Power NIS Adjustment Uncertainty."
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| 01051209. DOC 06 2
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| . PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 3 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2 1
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| i l 7. APPENDICES 7.1 PME Instrumentation j
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| : 8. ATTACHMENTS l
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| 8.1 " Data Sheet i - NIS and Steam Pressure Data," 03/17/92 8.2 " Data Sheet 2 - RCS AT and Feedwater Temperature," 02/04/96 8.3 " Data Sheet 3 - Steam Generators 1 and 2 Feedflow D/P," 03/07/96 8.4 " Data Sheet 4 - Steam Generators 3 and 4 Feedflow D/P and S/G Blowdown Flow," l 03/07/96 l
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| 8.5 " Data Sheet 5 - NIS Power Range Channel Gain Adjustment," 04/04/97 )
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| 8.6 " Required NIS High Flux Trip Setpoint as a Function of Calorimetric Power,"
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| 05/22/96 8.7 "NFF Validity Check," 08/14/97 8.8 " Manual R-2B2 PPC Data Logging," 06/04/98
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| : 9. SPONSOR ,
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| Darell Farrer i
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| 010$1209. DOC 06 3
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| PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 4 OF 11 TITLE: MANIE OPERATOR IIEAT BALANCE UNITS 1AND2 START DATA SECTION UNIT l O/ERATING MODE I DATE/TIMETodai / 0000
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| : 10. PRECAUTIONS AND LIMITATIONS INITIALS 1
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| 10.1 The accuracy of the heat balance is adversely affected by plant transients. It is recognized that stable plant conditions may not exist for ideal performance of this test, so it is important to schedule and perform the test and interpret results with full consideration of the possible transient processes that may be present. If any of the prerequisites cannot be met, note reason in REMARKS. if.
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| 10.2 When this procedure is performed to adjust NIS powe. range channels, AT vos hould be within i0.5'F of Tarr. This will result in a more accurate NIS channel adjustment. Small changes in T4vo can appreciably affect the indicated NIS power. @
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| 10.3 Calculated power must not be overcorrected due to nozzle defouling. If the unit is curtailed and power was below 85 % RTP during the curtailment, the NFF must be verified upon return to 97% RTP to be acceptable or reset to 1.0.
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| 10.4 Once the PPC NFF value is reset to 1.0, Secondary System k
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| l Engineering approval is required to change the value from 1.0. N 10.5 Due to increasing feedwater flow uncertainty at low power, calorimetric power uncertainty increases as power decreases.
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| When NIS adjustments are performed based on low povier data, the NIS uncertainty is increased. Because the NIS adjustment is a gain adjustment, this uncertainty is amplified as power increases toward the trip setpoint. As a result, the trip setpoints must be controlled per Attachment 8.6 when NIS adjustments are performed at less than 45 % RTP. Nv PERF
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| : 11. PREREOUISITES 11.1 If temporary or alternate instrumentation is used, it must be in current calibration and meet or exceed the accuracy of the component it is replacing.
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| N/A [ ]
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| 11.2 Reactor thermal power is 215 % RTP. th
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| ()b 11.3 Generator and reactor power levels constant for one hour or longer. (Constant Power Level t1/2%.) (4 U,
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| 11.4 Water levels stable within operating bands for Steam Generators (S/G), VCT, pressurizer and condenser which imply STM/FW flow and makeup / letdown flow are balanced.
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| h 01051209. DOC 06 4
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| PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 5 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2 PERF 11.5 Axial Flux Difference within Target Band. MO 11.6 No short term reactivity changes '(such as control bank U movement or boration/ dilution) are occurring. (F7 e e ee e ee e e e eee e e e e e eeee eeeee ee eee e e e e e eee e ee e e e e ee e eeee ee se s e ee ee e e e e e e e e e eh e CAUTION: When this procedure is performed to adjust NIS power range channels, Tavashould be within i0.5*F of Tarp. <
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| eseeeeeeeeeeeeeeeeeeeeeeeeeeee. eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeese 11.7 Tavo within tl.5*F of Tasy. N/A [ ] d Tavo = 6'73.(o Tasr = OBA V NOTE: During plant temperature / power coastdown, Step 11.7 is N/A. Follow Operational Guidelines per OP L-4. '
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| NOTE: If the PPC is not available, mark Steps 11.8,11.9 and 11.10 N/A. t 11.8 Obtain the NFF value from the PPC.
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| N/A[
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| NFF, point KR2BNFF NOTE 1: The quality of point KR2BNFF must be " GOOD" or "S UB. "
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| NOTE 2: If the NFF value is 1.0, then mark Steps 11.9 and 11.10 N/A.
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| 11.9 Check the validity of the NFF: N/A [
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| 11.9.1 For a unit that is curtailed and power has been returned to 97% RTP, was power at or above 85% RTP throughout the curtailment?
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| YES [ ] NO [ ] N/A [ ]
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| 11.9.2 The quality of PPC point UMWRAT is " GOOD" and its value is less than or equal to 1.005.
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| s 1.005 YES [ ] NO [ ]
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| UMWRAT NOTE: If Step 11.9.2 is marked NO, then Step 11.9.3 may be skipped and marked N/A.
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| 11.9.3 If Step 11.9.1 is NO then check the validity of the NFF by performing Attachment 8.7.
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| NFF valid ? YES [ ] NO [ ] N/A [ ]
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| 01051209. DOC 06 5
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| J l
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| l PACIFIC GAS AND ELECTRIC COMPANY ' NUMBER STP R-2B2 ;
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| l DIABLO CANYON POWER PLANT REVISION 9 I- PAGE 6 OF 11 TITLE: MANUAL OPERATOR HEAT BALANCE UNITS 1AND2 l i
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| PERF VERIF 11.10 If either Step 11.9.1 or 11.9.3 is marked YES or N/A and Step 11.9.2 is marked YES, then. mark this step N/A.
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| l Otherwise, to 1.0. complete the following steps N/Ato
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| [ reset the NFF valu[e 11.10.1 Complete Attachment 9.2 of OP O-15, " Plant Process Computer Addressable Point Change Form." l CAUTION: Once the PPC NFF value is reset to 1.0, Secondary System Engineering approval is required to change the value from 1.0.
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| .........................................e.....e......................e................
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| 11.10.2 Reset the PPC NFF value, point KR2BNFF, to 1.0.
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| 11.10.3 Contact Secondary System Engineering and request them (PTES) to undate the PPC NFF value for this unit.
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| 11 Calculate the adjustment factor (AF) for input to the "OPHB2B" program:
| |
| 11.11.1 If the PPC is available, obtain point KR2BAF (quality must be " GOOD" or "SUB"):
| |
| AF =
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| KR2BAF N/A [
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| 11.11.2 If the PPC is not available, obtain the VAF from Volume 9 Table T-IIC-27 F-IIC-28 for Unit 2) and use this as the AF:
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| AF = 1.O I VAF N/A [ ] @t l
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| 01051209. DOC 06 6 j
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| PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 7 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2 PERF VERIF
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| : 12. PROCEDURE 12.1 Station data takers and synchronize acquisition of the data required on Attachments 8.1 through 8.4 at one minute intervals for a minimum of 5 minutes. (This includes placing all of the requisite PPC trend points on one minute digital trend. See Attachment 8.8 for guidance.) Larger intervals may be used for certain engineering tests. Q4 Start Time 0031 Oen. Output (MWe) Il31,7 b 12.2 Collect data sheets and PPC trend data, average the data, apply i correction factors where necessary, and record the final values. 7 12.3 Disconnect all temporary test equipment installed per this procedure or make a jumper log entry.
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| N/A[] U9 MO 12.4 Test performers and verifiers:
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| Name Sienature Date/ Time [nji n
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| RM M Tsk i olos @
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| h $rV14(L.E b % IOl[O h
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| /
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| /
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| 01051209. DOC 06 7
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| I I
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| l PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP T-2B2 l DIABLO CANYON POWER PLANT REVISION 9 l I
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| PAGE 8 Oc! 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1MD2 11R1
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| : 13. DATA REDUCTION AND EVALUATION l 13.1 Input the final data into the "OPHB2B" program (DPP PC-4) and f obtain a printout widi the heat balance results. ___ j 13.2 Attach all PPC trend data, completed data sheets, and l "OPHB2B" printout. . _ , _ ,
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| 13.3 NIS % power for each channel is not more than 4% below Core ;
| |
| % RTP. YES [ ] NO [ 7 13.3.1 If NIS % power is more than 4% below Core
| |
| % RTP, then: ;
| |
| N/A [ ] j
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| : a. Promptly notify t e SFM.
| |
| : b. Reduce power level below 100% calculated.
| |
| : c. Initiate an AR. I AR#
| |
| : d. Once below calculated 100%, run another STP R-2B2 to confirm results. _,
| |
| : e. Contact Reactor Engineering to assist ir die evaluation and obtain approval to adjur, NIS i channels. ,
| |
| : f. Adjust the NIS channels per Attachtrent 8.5. _
| |
| 13.4 The NIS % power for each channel does not differ ':icom Core %
| |
| RTP more than 2.0. 'iES [ ] NO [ ]
| |
| 13.5 If either Step 13.4 is NO or an NIS adjustment is Jesired, then complete the following:
| |
| N/A [ ]
| |
| 13.5.1 Notify the SFM.
| |
| 13.5.2 Contact Reactor Engineering to ass st in the evaluation and obtain approval to adjust NIS c.hannels.
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| 13.5.3 Adjust the NIS channels per Attaohment 8.5.
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| 01051209. DOC 06 8
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| i j .
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| PACIFIC GAS AND EUICTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 9 OF 11 TITLE: MANUAL OPERAT OR IIEAT BALANCE UNITS 1AND2 EEB.E 13.6 Ifim NIS adju.ttment w..is perfbnned AND calorimetric power is ten than 45% RTP, then are the NIS hi t;h flux trip serpoints set appropriately per Attachment 3.67 f
| |
| YES [ ] NO [ ] N/A [ ]
| |
| 13.6.1 if Step 13.6 is NO, then initiate an Info Only Tech
| |
| / Spec Tracking Sheet to ensnre that within 24 hours one of the following actions is corepleted:
| |
| : a. Reset the NIS high Flux trip setpoints as required by Attachment 8.6 bi. sed en the calorhnetric powerlevel from fae most recent STP R-2B1 or STP R-2B2.
| |
| : b. Increase reactor power rmd reperfonn STP R-2B1 or STP R-2B2 and adjust NIS if necessary. Verify that the NIS trip setpoints are acceptable per Attechmmt 8.6 based on the calorimetric power form the most recent STP R-2B1 or STP R-2'B2.
| |
| Tracking Sheet # ,_ N/A [ ]
| |
| 13.7 Subtract the Core % RTP frorn the s.verage PPC power (U1169) and edd this to 100% to cbtain the PPC MAX POWER value.
| |
| Record this on the CC2 famicoid and on the "Summs.ry of Daily Opcar.tions" Data Sheet N/A [ ]
| |
| ~
| |
| + 100 =
| |
| i Ull69 % RTP PPC MAX POWER REMARKS:
| |
| I Performed By: _
| |
| Date/ Time /
| |
| t 14 BEJYDJtATION j None 01051209. DOC - 06 9
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| '. PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9
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| ( PAGE 10 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2
| |
| [ PERF
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| : 15. REVIEW AND ROUTING 15.1 (STA) Verify M&TE usage recorded on the RT Work Order for this test.
| |
| N/A [ ]
| |
| 15.2 (SFM) Review completed procedure and based on test data, make determination of test acceptability. If malfunctions are noted, refer to the applicable LCO and initiate an AR.
| |
| AR#
| |
| REMARKS:
| |
| Signature: Date/ Time /
| |
| SFM 15.3 (PPE) Review procedure for completeness and acceptability.
| |
| NOTE: PPC data and OPHB2B output are Eqt required to be sent to RMS.
| |
| REMARKS:
| |
| Reviewed By: Date (PPE)
| |
| I I
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| f l
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| l 01031209, DOC 06 10
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| '. PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-2B2 DIABLO CANYON POWER PLANT REVISION 9 PAGE 11 OF 11 TITLE: MANUAL OPERATOR IIEAT BALANCE UNITS 1AND2 APPENDIX 7.1 PME INSTRUMENTATION This Appendix lists the PME Instrumentation whicliis or may be used to take Acceptance Criteria data in this STP.
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| TAG #
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| Loop 1 Loop 2 Loop 3 Loop 4 DESCRIPTION FI-510C FI-520C FI-530C FI-540C Feedwater D/P T0418A T0438A T0458A 'IT)478A PPC Feedwater Temperature P0400A PO420A P0440A P0460A PPC Steam Line Pressure,1 of 3 P0401A P0421A PO441A PO461A PPC Steam Line Pressure,2 of 3 PO402A PO422A P0442A P0462A PPC Steam Line Pressure,3 of 3 PI-514 PI-524 PI-534 PI-544 VB3 Steam Line Pressure FR-48 FR-49 FR-60 FR-51 Blowdown Chart Recorder Flow FI-153 FI-154 FI-155 FI-156 Blowdown Indicator Flow TAG # DESCRIPTION T2247A PPC Feedwater Heater l A Outlet Temperature T2249A PPC Feedwater Heater IB Outlet Temperature T2251A PPC Feedwater Heater IC Outlet Temperature N-41 Indicated Power, Power Range NIS Channel N41 N-42 Indicated Power, Power Range NIS Channel N42 N-43 Indicated Power, Power Range NIS Channel N43 N-44 Indicated Power, Power Range NIS Channel N44 i
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| l 1
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| l l
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| l l
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| 1 01051209. DOC O6 11
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| I (C*WUNCONTROLLED PROCEDURE YDO NOTUSEIO PERFORM WORK &ISSUEFOR;USE lL%l tv Axt.cuebarna::aa.i.u;ua.us -. asaam,ur xw caus: auaw.cxle (tc. a 03/17/92 Page1ofI DIABLO CANYON POWER PLANT STP R-2B2 ATTACHMENT 8.1 AND TITLE: DATA SHEET 1 - NIS AND STEAM PRESSURE DATA UNIT NO.
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| ! STATION: Control Room l NUCLEAR INSTRUMENTATION PPC AVE l
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| (NIS DRAWER Ma'ER) POWER l
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| N41A N42A N43A N44A U1169 TIME 1 I/Y), b 162 7 in1. C / 0l' 'I TIME 2 900. 5 102.8 101< 3 10 l 8 -
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| TIME 3 ( 00.$ /o2.% to/,4 IO) 3 TIME 4 90o,6 902 6 fol# /o f 9 TIME S too. 5 /oz. 0 Iol 3 101 d AVERAGE (00.b /02,7 IcI 4 lol N
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| ** METER CORR. FACTOR o ,[) d,6 -03 o,O N/A AVG + METER CORR. I00. (r /o2.7 I0le I / 0 /, 4 N/A "FROM VOLUME 9, FIGURE II-C-2 STEAM PRESSURE S/G 1 (P040_A)* S/G 2 (P042_A)* S/G 3 (P044_A)* S/G 4 (PG46_A)*
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| TIME 1 TIME 2 TIME 3 TIME 4 TIME 5 AVERAGE
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| _ = 0,1, or 2 i OR IF ' OF SERVICE: j i
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| Take one (1) reading at the beginning of the test at the time designated. j VB3 STEAM PRESStJRE (PSIG) l Loop 1PI514 Loop 2 P1524 Loop 3 PI 534 Loop 4 PI 544 YS * $ She 0] Tr 0 Q%$sb l
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| RECORDED BY: QO TIME /DATE: Uluf / C/02-C' / ,
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| 010512zz 06 12
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| 'i 2?
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| . :?f.mUNCONTROLLEDfROCEDUREdDO
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| .a .uatur.mzuw :xaums 02/04/ %
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| amavaw.a a.umn. cum.w.t NO]w:USE:TO PERFORM LuWORKorJS Page 1 of 2 DIABLO CANYON POWER PLANT STP R-2B2 ATTACHMENT 8.2 AND TITLE: DATA SHEET 2 - RCS _T AND FEEDWATER TEMPERATURE UNIT NO.
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| STATION: Control Room FEEDWATER TEMPERATURE (S/G INLET) ** (HTR #1 OUTLET)
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| T0418A T0438A T0458A T0478A T2247A T2249A T2251A TIME 1 TIME 2 ORIF TIME 3 ANY NOT TIME 4 AVAILABLE TIME S >
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| AVERAGE AVERAGE OF 3 =
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| ** Verify the #1 Feedwater Heater Bypass Valve is closed and not leaking.
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| OR IF THE PLANT COMPUTER IS OUT OF SERVICE:
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| If the Plant Computer is Out-of-Service, have Technical Maintenance hook up the Burns 1003 Series RTDs in the thermal wells of the Number 1 Heater Outlets (TX-249, TX-250 and TX-251). Fluke 8842A or 8810A are used to read resistance in ohms. Resistance to temperature conversion tables are provided j with each RTD; ensure that the serial numbers on the conversion tables match those of the RTDs. l Furthermore, these Burns 1003 Series RTD are four-wire RTD, ENSURE proper hookups have been done.
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| Complete page 2 of this data sheet. Attach the appropriate copies of the conversion tables to this procedure and perform the necessary interpolation. All temporary test equipment disconnected at the completion of the test, otherwise, a jumper-log entry is made. N/A [ ]
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| Verified by: Date:
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| RECORDED BY: TIME /DATE: /
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| 010512:z 06 13
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| yil[uta
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| * 1 UNCONTROLLEDuumu:n =%. PROCEDURE r (00 NOTVSE . uan TOPERFORM .=::aj....WORKQSSUEFOR r : +-m shacur..::~ .i STP R-2B2 (UNITS 1 AND 2)
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| ATTACHMENT 8.2 1 TITLE: DATA SHEET 2 - RCS _T AND FEEDWATER TEMPERATURE l N/A [ ]
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| FEEDWATER HEATER #1 OUTLET RTD RESISTANCES TX-249 TX-250 TX-251 Time 1 igg /4 /g r,24 /9 r., o g Time 2 pgG j[g ,qq, 7 g ; g g, j (,
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| Time 3 9gf,l9 y g7, zL g g5', Iy Time 4 jgp, / O l9 p, g6 pg.p,/o l
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| Time 5 /gr.26 igr, /3 l% T.12 .
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| 1 Average (OA) j3f f(, O f g.9 , 23 O j g f, Q R Data Taken by: Mkb Date: 7Je/o CONVERSION TABLE !
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| TX-249 TX-250 TX-251 (0) (F) (O) (F) (O) (*F) 04 T4 03 TA 0A T3 i Actual (gg,[g q p,q gg. 23 4361 lf Ll2 9%7 l l
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| RTD Serial #'s ll* TH I2- / (GCl40 / lisflilN Cony. Table #'s IT S'9 0 / 7T$~Cl) / ~ CTS ~92 j RTD Cal Due Dat:s Or / l.a h.e / 6b l' Fluke Serial #'s . (26 413 / i21,414 / st6 415-Fluke Cal Due Dates takr / O / L N /~
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| Performed by: N, Date: @el%
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| /
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| 010512zz 06 14
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| r lic sh:aw:?mu.w;Amayawksm:h.s
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| "* *l:lVNCONTROLLED PROCEDURERDO c' mac+ac.NOTUSE:TO ua:waum.:':. - ti un PERFORM WORKoIISSUEFO 03/07/96 Page1of1 DIABLO CANYON POWER PLANT STP R-2B2 ATTACHMENT 8.3 AND TITLE: DATA SHEET 3 - STEAM GENERATORS 1 AND 2 FEEDFLOW D/P UNIT NO.
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| STATION: 85 ft elevation Caged Pipe Rack Area (Mechanical Panels FI-510 and FI-520)
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| {
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| NOTE: Tap the Barton gauge face prior to reading. 1 l
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| l WARNING: If a Barton gauge needle is bouncing as high as 500 inches d/p, notify the SFM and SPPE (Reactor Engineering) prior to taking data. The gauge
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| ! may be unreliable.
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| l S/G 1 S/G 2 l
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| FI-510C FI-520C TIME 1 gq q 7 3g(,,o TIME 2 74 g- L{ p(,, g TIME 3 hq5 , 3 T/(,, $
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| TIME 4 q q y, 7 qq6, ']
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| TIME 5 jgy, g 3g(,, 9 ;
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| ! AVERAGE 395, g p(,, q l l Calibration Due Dates: FI-510C/NTobi, FI-520C/ db Nf RECORDED BY: YJ '
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| TIME /DATE: Tb
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| /
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| / 6/W ll l
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| .010512u 06 15
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| L PROCEDUREFDO v *. _ e m e e w.; ~ ~ NOTUSE - wTOPERFORM u m , . .uWORKofISSUEFOR a m u ..;s c u ra USEc.*7*
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| :a1 l 03/07/ % Page1of1 )
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| . DIABLO CANYON POWER PLANT STP R-2B2 ATTACHMENT 8.4 AND TITLE: DATA SHEET 4 - STEAM GENERATORS 3 AND 4 FEEDFLOW D/P AND S/G BLOWDOWN 1
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| (
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| UNIT NO. !
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| STATION: 100 ft. elevation Aux. Bldg. Penetration Area (Bartons are next to PM105) 115 ft. elevation (Blowdown indication inside PM205)
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| NOTE: Tap the Barton gauge face prior to reading.
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| WARNING: If a Barton gauge needle is bouncing as high as 500 inches d/p, notify the SFM and SPPE (Reactor Engineering) prior to taking data. The gauge may be unreliable.
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| CHART RECORDER PAPER,100% FULL SCALE = 200 GPM S/G 3 S/G 4 SGBD 1** SGBD 2" SGBD 3" SGBD 4" TIME F1-530C FI-540C 1 4<0. ') (4N. o 24.0 91. ? T4,2 R 3. 7 2 4W, I 4F.o 9t'. 4 del W 3. '7 rd a 3 4 40,4 %% o itt . t #1, i 23. x e3.4 ,
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| 4 4% tv. i RI lH,3 i n.4 5(3. S I 5 q v o0,'t '420 9 R1 S $1. ?.
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| t3,84 WIM l AVG M To. '2. '{25,6 !
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| GPM %9sc 8l,l 83.8 Tl,7 Calibration Due Dates: FI-530C/ IMe/- FI-540C/ kW If Blowdown Totalizers (FIl53 - FIl56) Used:
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| TIME (min) SGBD1 SGBD 2 SGBD 3 SGBD 4 START END DIFFERENCE FLOWRATE
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| * N/A Calibration Due Dates:
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| * Flowrate = Difference between Totalizer readings Difference in time (minutes)
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| ** Blowdown flow must be a' tained on scale on the flow recorder to be measured accurately.
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| RECORDED BY: , TIME /DATE: I*/4 / OI C lW l 010512zz 06 16
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| 4 GNUNCONTROL
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| >newww.:aa. LEO .z PROCEDUREEDO cac s.wantx=.:1;.n:.tn NOTUSEa.:. TO
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| .. PERFORM ~ . . . WORKorISSUE
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| =anas.= a FOR USE!:*&
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| 04/04/97 Page 1of1 DIABLO CANYON POWER PLANT.
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| STP R-2B2 ATTACHMENT 8.5 AND TITLE: DATA SHEET 5 - NIS POWER RANGE CHANNEL GAIN ADJUSTMENT l
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| UNIT NO. POWER LEVEL % RTP INITIAL / DATE / TIME 8.5.1 Heat Balance Performed: /
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| DATE TIME 8.5.2 Obtain Reactor Engineering approval to adjust NIS. / /
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| 8.5.3 Verify control rods in MANUAL. / /
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| 8.5.4 Complete the following Table for gne channel at a time.
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| CAUTION: The gain adjust pot must be turned slowly to prevent actuation of a rate trip bistaMe. If a rate trip bistable is actuated during this adjustment, reset by cycling the rate mode switch from NORMAL to RESET to NORMAL after completing the adjustment for that channel and before adjusting the next channel. DO NOT ADJUST NIS IF A CHANNEL IS OUT OF SERVICE. THIS MAY RESULT IN A REACTOR TRIP.
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| DATA N41A N42A N43A N44A
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| : a. Time l
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| : b. "As Found" Potentiometer Setting
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| : c. Indicated Meter Reading Before Adjustment, %
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| : d. Calculated Difference, % From OPHB2B Printout
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| : e. Unlock Gain Pot (Initials)
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| : f. Adjust Meter Reading [per PPC* or (c.- d.)}**
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| : g. Lock Gain Pot (Initials)
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| : h. Indicated Reading After Adjustment, %
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| : i. "As Left" Potentiometer Setting 8.5.5 As quickly as possible complete the following table.
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| Final Meter Reading (N41 A -+ N44A)
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| : a. Final PPC Reading (N0049A -+ N0052A)
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| Final Control Console Reading (N41B -+ N44B) b.Each NIS Channel's 3 Indicators Agree Within 1.8%t
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| : c. All Indicators in Step a Agree Within 2%t
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| : d. Reperform Heat Balance; STP R-2B1 or STP R-2B2 Results OK (Initials)
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| Adjust the pot to match the PPC NI and PPC calorimetric indications.
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| If the pot setting reaches 0.00 gr 10.00, have TM adjust the internal coarse gain control.
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| t If not acceptable, write an AR.
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| 8.5.6 Control rods as directed by SFM. MANUAL [] AUTO [ ] / /
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| REMARKS:
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| Adjustment By: DATE/ TIME /
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| Reviewed By: DATE/ TIME /
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| SFM Reviewed By: DATE SPPE (Reactor Engineering) 010512zz 06 17
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| tR c. UNCONTROLLED c u &i[w PROCEDUREXDONOTUSE x b u;L w.a A kLuaw TOPERFORM.WORKdrISSUEFOR s.<:2..mmt.,c:.t.wsum.c. USE11%1 i ,,:awa. : ; ..:n ..si,4.: s w..::.<bM 05/22/96 Page1of1
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| . DIABLO CANYON POWER PLANT STP R-2B2 ATTACHMENT 8.6 AND TITLE: REQUIRED NIS HIGH FLUX TRIP SETPOINT AS A FUNCTION OF l CALOR! METRIC POWER i
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| l Calorimetric Required NIS Power Trip Setpoint i
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| 2 45% RTP s 109% RTP l
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| 2 40% RTP s 107% RTP 2 35% RTP s 105% RTP a 30% RTP s 102% RTP 2 25% RTP s 97% RTP )
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| 2 20% RTP s 88% RTP 215% RTP s 72% RTP 010512zz 06 18
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| m u mia;.g;cin-.;.1s;4 UNCONTROLLED E.:>?fW:l PROCEDURE m ?DOm h uw< PERFORM NOTUSE!TO m gu:.e:a,e,;w mum.ER,JM WORKofISSUEFOR US c4.
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| Y1 08/14/97 Page 1 of 2
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| .- DIABLO CANYON POWER PLANT STP R-2B2 ;
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| , ATTACHMENT 8.7 AND TITLE: NFF VALIDITY CHECK 1 Determine the date and time of the most recent 100% RTP STP R-2B .
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| Last 100% STP R-2B performed: /
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| Date Time
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| : 2. Obtain the Secondary Calorimetric power and RCS loop Delta-T values at the time listed in paragraph 1. Use PPC archive function and MID/LONG term memory. {
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| U4300 T0403A T0423A T0443A T0463A
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| : 3. Log the PPC points listed below at 1 minute intervals for a total of 8-12 readings and average.
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| U4300 T0403A T0423A T0443A T0463A l AVERAG3 l
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| : 4. Using the average calorimetric power obtained in Step 3, determine the Enthalpy Change Weighting Factor (ECWF) using the values listed below. Interpolate as necessary.
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| % Power ECWF % Power ECWF % Power ECWF % Power ECWF 100 1.00000 97 1.00558 94 1.01137 91 1.01737 99 1.00183 96 1.00749 93 1.01334 98 1.00370 95 1.00941 92 1.01534 /
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| % POWER / ECWF
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| : 5. Calculate the power level adjustment coefficient (PLAC). ;
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| U4300 (from Sten 3)
| |
| * ECWF . ( )*( ). .
| |
| U4300 (from Step 2) ( ) PLAC
| |
| : 6. Complete the calculations in the following table to determine the adjusted loop Delta-T values.
| |
| values from step 2 times PLAC T0403A = * =
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| T0423A = * =
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| T0443A = * =
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| T0463A = * =
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| 010512n 06 19
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| Yy"guNCONTROLLED PROCEDURE s.a.x-m.u 3DO NOTUSE TOPERFORM WORKofJSSUEFOR U nx.ha.c.2rews.~ r2ersia.
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| 08/14/97 Page 2 of 2 STP R-2B2 (UNITS 1 AND 2)
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| ATTACHMENT 8.7 TITLE: NFF VALIDITY CHECK NOTE: Should any RCS loop Delta-T PPC point not be either " GOOD" or "DALM," then that loop is not to be included in the remaining calculations. Write N/A in each block affected and note in remarks.
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| : 7. Calculate the difference between the current and previous values.
| |
| values from step 6 values from step 3 T0403A = - =
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| T0423A = - =
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| T0443A = - =
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| T0463A = - =
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| : 8. Average the absolute values of the results of step 7.
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| Absolute value of results of step 7 T0403A T0423A T0443A i i
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| T0463A Average
| |
| : 9. Determine validity of the NFF. Is the average value in step 8 less than 0.4?
| |
| YES [ ] The NFF is valid NO [] The NFF is no longer valid and should be reset to 1.0 Performed by: / /
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| Signature Date Time Reviewed by: / /
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| Signature Date Time Remarks:
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| 010512zz 06 22
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| \
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| seesee. ..eeeeeeeeeeeeeeeeeeeeeeeeee seseeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeee**eeeee
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| ~
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| OPERATOR HEAT BALANCE FOR STP R-2B2, DPP PC-4, REV 3,3/2 .
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| seeeeeeeeeeeeeeeee.eee.ooeee.eeeeeeeeeeeeeeeeeeeeeese.eeeeeeee.eeeeeeeeeeeeeeeeeeeeeeee l "" INPUT DATA ""
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| l Date - Today DCPP Unit 1 1
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| Time- 0140 Verification OK l Parameter Loop 1 Loop 2 Loop 3 Loop 4 N41 N42 N43 N44 NIS power (%) 100.6 102.7 101.1 101.3 l Feedwater temp. (*F) 429.9 429.9 429.9 429.9 S/G pressure (psig) 772.8 772.1 775.0 783.3 Feedwater d/p (in. H2 O) 395.3 376.9 450.2 425.0 S/G blowdown (gpm) 84.0 81.1 83.8 83.7 Steam quality .99908 .99908 .99908 .99908 Generator output - 1128.7 megawatts Feedwater adjustment factor (AF) = 1
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| ***
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| * OUTPUT DATA ""
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| Parameter Loop 1 Loop 2 Loop 3 Loop 4 NIS power (%) 100.6 102.7 101.1 101.3 Cort E,TP(%) {
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| 99.5 99.5 99.5 99.5 i Difference (%) 1.1 3.2 1.6 1.8 Tctal NSSS thermal output = 333.0 Mwt or 99.49%
| |
| Total core thermal output = 3321.3 Mwt or 99.50%
| |
| Gross heat rate = 10020 BTU / kwhr
| |
| *For loop thermal outputs, print debugging values.
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| NIS channel 41 is within allowable tolerance.
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| NIS channel 42 exceeds the allowable limit of 2 %.
| |
| NIS adjustment required, notify Shift Foreman.
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| NIS channel 43 is within allowable tolerance.
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| NIS channel 44 is within allowable tolerance.
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| Remarks Performed by Date
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| ' Reviewed by Date
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| | |
| r NUCLEAR POWER GENERATION -
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| I* DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE i
| |
| i Number: LJCNRC-4
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| | |
| ==Title:==
| |
| RESPOND TO HIGH ACCUMULATOR LEVEL Examinee:
| |
| Evaluator:
| |
| Print Signature Date i l
| |
| Results: Sat Unsat Total Time: minutes Comments:
| |
| | |
| ==References:==
| |
| AR PK02-10, ACCUM LEVEL HI-LO, Rev.15 l AR PK02-05, ACCUM PRESSURE HI-LO, Rev.13 I l
| |
| Alternate Path: Yes No X Time Critical: Yes No X l
| |
| Time Allotment: 10 minutes Critical Steps: 4,5,10 Job Designation: SRO K/A Number: 006000A1.13(3.7)
| |
| -m .,
| |
| l AUTHOR: JOHN BECERRA DATE: 11/12/98 REVIEWED BY: NA DATE: NA l JPM COORDINATOR APPROVED BY: ,
| |
| NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| . JPM TIT 1.E: RESPOND TO HIGH ACCUMULATOR LEVEL JPM NUMBER: LJCNRC-4 lNSTRUCTOR WoRKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, md task standard. The examiner will then ask if any clarifications are needed.
| |
| Required Materials: None !
| |
| Initial Conditions: Unit 1 is at 100% power. Annunciator PK02-10 is in alarm and an investigation indicates a high level in accumulator 14. 4 Initiating Cue: You are directed to decrease accumulator 14 level. Normal RCS pressure and temperature exist.
| |
| Task Standard: Accumulator 14 level is restored and ALL accumulator alarms are clear.
| |
| l I
| |
| i l
| |
| ~
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| LJCNRC-4 PAGE 2 OF 7 REV.O
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| | |
| I JPM TIRE: RESPOND TO HiGH ACCUMULATOR LEVEL JPM NUMBER: LJCNRC-4 lNSTRUCTOR WORKSHEET I
| |
| l Start Time:
| |
| l Step Expected Operator Actions ;
| |
| l
| |
| : 1. Obtain the correct procedure. 1.1 Operator references AR PK02-10.
| |
| Step was: Sat: Unsat:
| |
| 1
| |
| : 2. Verify abnormal condition exists. 2.1 Operator checks LI-956 and LI-957 to (
| |
| verify the alarm is not due to an i i
| |
| instrument failure.
| |
| Cue: Technical Specificatior.
| |
| requirements do not have to be addre.ssed at this time.
| |
| 1 Step was: Sat: Unsat:
| |
| : 3. Check fill path isolated. 3.1 Operator locates the control s- J' Ses for:
| |
| e 8876D e 9357A e 9357B 3.2 Calls nuclear operator to clase or check closed 9352A, B, C, and D.
| |
| Cue: Nuclear operator reports 9352A through D CLOSED.
| |
| i 3.3 Operator locates the control switch for:
| |
| . 8878D 3.4 Verifies valves are closed.
| |
| Step was: Sat: Unsat: j i 1
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJCNRC4 PAGE 3 OF 7 REV.O
| |
| | |
| JPM TITLE: RESPOND TO HIGH ACCUMULATOR LEVEL JPM NUMBER: LJCNRC-4 lNSTRUCTOR WORKSHEET
| |
| : 4. ** Open accumulator drain to RCDT 4.1 Operator opens 8876D.
| |
| valve,8876D.
| |
| Step was: Sat: Unsat: *
| |
| : 5. *
| |
| * Monitor level decrease. 5.1 Operator monitors accumulator 14 level indication.
| |
| 5.2 Closes valve 8876D when the accumulator levelis approximately 67%.
| |
| Step was: Sat: Unsat:
| |
| : 6. Verify annunciator status. 6.1 Operator checks that annunciator windows PK02-05 and PK02-10 are not in alarm.
| |
| 6.2 Notes low pressure in accumulator 14.
| |
| Step was: Sat: Unsat:
| |
| : 7. Obtain the correct procedure. 7.1 Operator references AR PK02-05.
| |
| Step was: Sat: Unsat:
| |
| * l
| |
| : 8. Verify abnormal condition exists. 8.1 Operator checks PI-966 and 967 to verify the alarm is not due to an instrument failure.
| |
| Cue: Technical Specification action statements do not have to be addressed at this time.
| |
| Step was: Sat: Unsat: *
| |
| : 9. Check accumulator level. 9.1 Operator checks the level within Technical Specification limits.
| |
| 9.2 Determines the pressure decrease is due to the level decrease.
| |
| Step was: Sat: Unsat:
| |
| ' Denotes an entry required on the JPM cover sheet.
| |
| " Denotes a Critical Step.
| |
| LJCNRC-4 PAGE 4 OF 7 REV.O
| |
| | |
| I I
| |
| l i JPM TITLE: RESPOND TO HIGH ACCUMULATOR LEVEL JPM NUMBER: LJCNRC-4 l INSTRUCTOR WORKSHEET
| |
| : 10. ** Restore accumulator pressure. 10.1 Operator verifies nitrogen fill valve l
| |
| 8880 is open.
| |
| 10.2 Opens accumulator fill and vent isolation valve 8875D. )
| |
| 10.3 Closes 8875D after pressure returns to normal.
| |
| l i
| |
| 10.4 Verifies PK02-05 alarm is no longer i in alarm.
| |
| l 1 i Step was: Sat: Unsat: l l
| |
| 1
| |
| : 11. Monitor accumulator pressure. 11.1 Operator monitors accumulator 14 pressure indicators.
| |
| Cue: Pressure hat remained constant I for 20 minutes.
| |
| Step was: Sat: Unsat: I
| |
| : 12. Notify the Chemistry Deparunent. 12.1 Operator notifies the Chemistry ,
| |
| Department to sample accumulator l 14.
| |
| Note: Operator may perform this step after draining the accumulator (Step 5).
| |
| Step was: Sat: Unsat:
| |
| Stop Time:
| |
| Total Time: (Enter total time on the cover page) i
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJ3NRC-4 PAGE 5 OF 7 REV,0
| |
| | |
| JPM TITLE: JPM NUMBER: LJCNRC-4 EXAMINEE CUE SHCET Initial Conditions: Unit 1 is at 100% power. Anuunciator PK02-10 is in alarm and an investigation indicates a high level in accumulator 14.
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| Initiating Cue: You are directed to decrease accumulator 14 level. Normal RCS l pressure and temperature exist. I Task Standard: Accumulator 14 levelis restored and ALL accumulator alarms are clear.
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| i l
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| i l
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| LJCNRC-4 PAGE 6 OF 7 Rdv. O
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| JPM TITLE: RESPOND To HIGH ACCUMULATOR LEVEL JPM NUMBER: LJCNRC-4 ATTACHMENT 1, SIMULATOR SETUP j O Initialize the simulator to IC-1 (100%, BOL). !
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| 1 O Enter drill file 6054 or manually insert the following: l l
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| Command Description i
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| : 1. delm bsisace (4) Set up to monitor accumulator 14 level
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| : 2. mony bsisace (4) Increase accurr lator 14 level to 79%
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| : 3. ramp asisacw(4),55045,10,0 Raise accumulator 1-4 level
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| : 4. run 50 Freezes rfter 50 seconds O Inform the examiner that the simulator setup is complete.
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| O Go to RUN when the examinee is given the cue sheet.
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| )
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| e i
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| LJCNRC-4 PAGE 7 OF 7 REv. O
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| , NUCLEAR POWER GENERATION ,
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| DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: LJCNRC-5
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| | |
| ==Title:==
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| CROSSTIE OF VITAL BUS G TO H Examinee: l Evaluator:
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| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
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| l l
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| l
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| ==References:==
| |
| EOP ECA-0.3, Restore 4kV Buses, Appendix X, Rev. 8 l
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| Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 3,4,5,8,10,11 Job Designation: SRO K/A Number: 062000A4.01(3.1)
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| AUTHOR: JOHN BECERRA DATE: 11/12/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O c
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| | |
| JPM TITLE: CROSSTIE OF VITAL BUS G To H JPM NUMBER: LJCNRC-5 lNSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed.
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| Required Materials: None Initial Conditions: Reactor trip and safety injection occurred with loss of all off-site power.
| |
| Diesel generator 11 and diesel generator 13 failed due to a lube oil pressure problem. Diesel generator 12 is supplying 4kV bus G. CCW Pp 12 has failed resulting in a complete loss of CCW flow.
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| Initiating Cue: The Shift Foreman directs you to crosstie 4kV bus G to 4kV bus H per EOP ECA-0.3, Appendix X. Steps 1 and 2 have been completed. You are to begin at Step 3. The Site Emergency Coordinator has directed this implementation.
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| Task Standard: 4kV bus G on diesel generator 12 is crosstied to 4kV bus H.
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| LJCNRC-5 PAGE 2 OF 7 REV.0
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| , 1
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| =
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| JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: LJCNRC-5 lNSTRUCTOP WORKSHEET Start Time:
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| Step Expectet Operator Actions
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| : 1. Obtain the correct procedure. 1.1 Operator r :ferences ECA-0.3, Appendi 'C.
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| i Step was: Sa _ _ Unsat:
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| : 2. Cut in the DIR PWR. LOSS OF 2.1 Operator locates the toggle switch and j FIELD, & BKR OC PROT RLYS for cuts it in. I diesel generator 12.
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| Step was: Sat: Unsat:
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| : 3. ** Reset SI. 3.1 Operator checks PK08-21 " Safety injection Actuation" status.
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| 3.2 Operator locates SI reset pushbuttons.
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| 3.3 Manually depresses both pushbuttons.
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| I 3.4 Checks at least one of the following
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| . Monitor Light Box B " Safety Injection" red light OFF, OR
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| * PK08-21," Safety Injection Actuation" not ON.
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| Step was: Sat: Unsat:
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| : 4. *
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| * Cutout the auto transfer FCOs for 4.1 Operator locates auto transfer to 4kV and 12kV buses. startup FCOs for 4kV buses H, G, F, D and E, and 12kV buses D and E.
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| l 4.2 Places auto transfer FCOs mentioned above to CUTOUT position.
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| Note: Obsen es note.
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| Step was: Sat: Unsat:
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| * Denotes an entry required on the JPM cover sheet.
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| ** Denotes a Critical Step.
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| LJCNRC-5 PAGE 3 OF 7 REV.O
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| JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: LJCNRC-5 INSTRUCTOR WORKSHEET j
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| : 5. *
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| * Depress all auto transfer reset 5.1 Operator locates all auto transfer reset pushbuttons. pushbuttons near respective FCO cutout switches.
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| l 5.2 Depresses all auto transfer reset pushbuttons.
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| 5.3 Verifies blue lights OFF.
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| Step was: Sat: Unsat: l l
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| : 6. Verify OPEN all vital 4kV bus 6.1 Operator locates all vital 4kV bus I auxiliary feeder breakers. auxiliary feeder breaker indications. !
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| 6.2 Verifies all vital 4kV bus aux feeder breakers OPEN e 52-HH-13 OPEN e 52-HG-13 OPEN e 52-HF- 13 OPEN Step was: Sat: Unsat:
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| 1
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| : 7. Verify OPEN all vital 4kV bus startup 7.1 Operator locates all vital 4kV bus l feeder breakers. startup feeder breaker indications. i 7.2 Verifies all vital 4kV bus startup feeder breakers OPEN !
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| l e 52-HH-14 OPEN e 52-HG-14 OPEN e 52-HF-14 OPEN Step was: Sat: Unse.t: _
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| : 8. *
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| * Verify OPEN the 4kV startup feeder 8.1 Operator locates 4kV startup feeder breaker 52-HG-15. breaker control switch and indication for 52-HG-15.
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| 8.2 Verifies 4kV startup feeder breaker 52-HG-15 to 4kV Bus F, G and H OPEN.
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| Step was: Sat: Unsat:
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-5 PAGE40F7 REv.0
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| F
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| }
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| I
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| : JPM TITLE: CROSSTIE OF VITAL BUS G TO H JPM NUMBER: LJCNRC-5 INSTRUCTOR WORKSHEET
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| : 9. Verify OPEN the 4kV to 480 VAC bus 9.1 Operator locates the 4kV to 480 VAC feeder breaker for the deenergized bus bus feeder breaker fer the deenergized to be reenergized. bus to be reenergized.
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| 9.2 Verifies feeder breaker 52-HH-10 OPEN.
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| Unsat:
| |
| * Step was: Sat:
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| : 10. ** Close 4kV startup feeder breaker for 10.1 Operator locates the 4kV startup the deenergized bus being reenergized. feeder breaker for 4kV bus H, 52-HH-14.
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| 10.2 Inserts synchroscope key for 4kV bus H startup feeder breaker 52-HH-14.
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| 10.3 Turns synchroscope switch to ON.
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| 10.4 Closes 4kV ste; tup feeder breaker 52-HH-14 for 4kV bus H.
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| Step was: Sat: Unsat:
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| : 11. ** Close the 4kV startup feeder breaker 11.1 Operator locates the 4kV startup for the bus that will be supplying power feeder breaker for 4KV bus G, to the deenergized bus. 52-HG-14.
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| I 1.2 Inserts synchroscope key for 4kV bus G startup feeder breaker 52-HG-14.
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| I1.3 Turns synchroscope switch to ON.
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| L 11.4 CLOSES the 4kV startup feeder breaker 52-HG-14 for 4kV bus G.
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| I 1.5 Verifies running diesel generator l remains stable.
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| Step was: Sat: Unsat:
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| Stop Time:
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| l l
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| Tota! Time: (Enter total time on the cover page) j I
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-5 PAGE 5 OF 7 REV.O
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| \
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| JPM TITLE: JPM NUMBER: LJCNRC-5
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| , EXAMINEE CUE SHEET I
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| Initial Conditions: Reactvr trip and safety injection occurred with loss of all off-site power.
| |
| Diesel generator 11 and diesel generator 13 failed due to a lube oil pressure problem. Diesel generator 12 is supplying 4kV bus G. CCW Pp 12 has failed resulting in a complete loss of CCW flow.
| |
| I Initiating Cue: The Shift Foreman directs you to crosstie 4kV bus G to 4kV bus H per EOP ECA-0.3, Appendix X. Steps 1 and 2 have been completed. You l are to begin at Step 3. The Site Emergency Coordinator has directed this
| |
| ! implementation.
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| Task Standard: 4kV bus G on diesel generator 12 is crosstied to 4kV bus H.
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| I I
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| LJCNRC-5 PAGE 6 OF 7 REV.O
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| 'JPM TITLE: CROSSTIE OF VITAL BUS G To H JPM NUMBER: LJCNRC-5 ATTACHMENT 1, SIMULATOR SETUP O Initialize the simulator to the IC-1 (100%, BOL).
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| O Enter drill file 6055 or manually insert the following:
| |
| Command Description
| |
| : 1. mal degla act,2,0,0,d,0 Fails DG 11
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| : 2. mal degie act,2,0,0,d,0 Fails DG 13
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| : 3. mal sydl act,1,1,0,d,0 Loss of offset power
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| : 4. mal ppl2a act,0,0,0,d,2 Inadvertent SI, Train A
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| : 5. mal ppl2b act,0,0,0,d,2 Inadvertent SI, Train B
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| : 6. pmp ccw2 4,0,0,4,d,0 CCW pp 1-2 OC trip
| |
| : 7. loa afwl4 act,f,0,60,d,0 Opens knife switch for AFW pp 1-2
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| : 8. loa ess8 act,f,0,60,d,0 Opens knife switch for cont. spray pp 1-2
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| : 9. loa rhr10 act,f,0,60,d,0 Opens knife switch for RHR pp 1-2
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| : 10. loa ccw31 act,f,0,60,d,0 Opens knife switch for CCW pp 1-3
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| : 11. loa sis 3 act,f,0,60,d,0 Opens knife switch for SI pp 1-2
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| : 12. dse venl4 act,f,0,60,d,0 Opens breaker for CFCU 1-4
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| : 13. run 90 freezes simulator after 90 seconds O Inform the examiner that the simulator setup is complete.
| |
| O Go to RUN when the examinee is given the cue sheet.
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| LJCNRC-5 PAGE 7 OF 7 REV.O
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| NUCl EAR POWER GENERATION e DIAELO CANYON PO'NER PLANT 103 PERFORMANCE MEASURE Num'ocr: LJCNRC-6
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| ==Title:==
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| VERIFY A MISALIGNED ROD IS NOT STUCK Examinee:
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| Evaluator: l Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
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| 1
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| ==References:==
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| OP AP-12B, Control Rod Misalignment, Rev. 8A Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 10 minutes :
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| Critient Steps: 2,5,6 Job Designation: SRO K/A Number: 001000A2.03(4.2)
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| AUTHOR: JOHN BECERRA DATE: 11/12/98 RF. VIEWED BY: NA CaiC NA JPM COORDINATOR l APPROVED BY: NA DATE: NA TRAINING LEADER REv.0 1
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| l i
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| JPM Tm.E: VERIFY A MISALIGNED Roo is NOT STUCK JPM NUMBER: LJCNRC-6 lNSTRUCTOR WORKSHEET i
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| Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the
| |
| [ evaluator. The student will be given the initial conditions, initiating cue, l l and task standard. The examiner will then ask if any clarifications are l needed.
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| 1 Required Materials: None l
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| l 1 Initial Conditions: Unit 1 is at approximately 70% and was in the process of being returned to 100% following curtailment for condenser cleaning. The power i increase was stopped 5 minutes ago due to indications of a misaligned rod.
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| Initiating Cue: Rod K2, a control bank D, Group 2 rod, was observed to be inserted greater than 12 steps below the other rods in the group by DRPI indication. Procedure OP AP-12B, Control Rod Misalignment was implemented. You have been directed by the Shift Foreman to continue to address the control rod misalignment by verifying the rod is not stuck, per OP AP-12B, Step 13.
| |
| Task Standard: Verify that rod K2 is not stuck and report the status to the SFM.
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| l LJCNRC-6 PAGE 2 OF 6 REV.O
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| JPM TITLE: (ERIFY A MISALIGNED Roo IS NOT STUCK JPM NUMBER: LJCNRC-6 INSTRUCTOR WORKSHEET ,
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| Start Time:
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| Step Expected Operator Actions l
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| : 1. Obtain the co Tect procedure. 1.1 Operator references OP AP-12B.
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| Step was: Sat: Unsat:
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| * I I l 2. ** Place the bank selector to the 2.1 Operator reads the CAUTION before l affected bank. Step 13.a.
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| 2.2 Turns the ROD BANK / MODE SELECT switch to CB D position.
| |
| )
| |
| Step was: Sat: Unsat: *
| |
| : 3. Record the bank step counter position. 3.1 Operator observes the indicated step counter positions of control bank D groups 1 and 2 and records the values j observed. I Step was: Sat: Unsat: *
| |
| : 4. Record the plant process computer 4.1 Operator displays the control bank D position for the affected bank, position from the plant computer and j records the value.
| |
| Note: Operator may call up the control bank D address as listed in Appendix B of OP AP-12B (U0052), or record the value of U0052 from an existing group display.
| |
| Step was: Sat: Unsat:
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| : 5. *
| |
| * Drive the affected bank IN several 5.1 Operator inserts rods for control bank steps. D for several steps until a DRPI change is observed.
| |
| Step was: Sat: Unsat:
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| !
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJCNRC-6 FAGE 3 OF 6 REV.O
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| I l
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| 1 l JPM TITLE: VERIFY A MISAUGNED Roo is NOT stuck JPM NUMBER: LJCNRC-6 lNSTRUCTOR WORKSHEET
| |
| : 6. ** Determine rod status. 6.1 Operator informs SFM that ROD is not stuck.
| |
| I Step was: Sat: Unsat:
| |
| j Stop Time:
| |
| Total Time: (Enter total time on the cover page) l l
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| l l
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| * Denotes an entry required on the JPM cover sheet. 1
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJCNRCf3 PAGE 4 OF 6 REV.0 1
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| l JPM TITLE: VERIFY A MISALIGNED Roo is Nor STUCK JPM NUMBER: LJCNRC-6 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is at approximately 70% and was in the process of being returned to 100% following curtailment for condenser cleaning. The power increase was stopped 5 minutes ago due to indications of a misaligned rod.
| |
| Initiating Cue: Rod K2, a control bank D, Group 2 rod, was observed to be inserted 1 greater than 12 steps below the other rods in the group by DRPI J indication. Procedure OP AP-12B, Control Rod Misalignment was implemented. You have been directed by the Shift Foreman to continue to address the control rod misalignment by verifying the rod is not stuck, per OP AP-12B, Step 13.
| |
| Task Standard: Verify that rod K2 is not stuck and repon the status to the SFM.
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| l l
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| i LJCNRCs6 PAGE 6 0F 6 REV.0
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| JPM TITLE: VERIFY A MISAUGNED Roo is NOT STUCK JPM NUMBER: LJCNRC-6 J.mCHMENT 1 SIMULATOR SETUP O Initialize the simulator to IC-2 (75%, BOL).
| |
| O Verify step counters and RBU are updated to bank D at 190 steps.
| |
| O Enter drill file 6056, or manually insert the foliowing:
| |
| Command Description l 1. mal rod 4 act,1,k2,0,d,merfpa(14).gt.207 Sticks Rod K-2. Clears when bank D l group 2 is greater than 207 steps
| |
| : 2. mal res5 act,1254,120,0,d, Borates to get rods above K2 l
| |
| mcrfpa(14).gt.207
| |
| : 3. frz when mcrfpa(14).gt.207 Freezes sim when bank D is greater than j 207 steps
| |
| : 4. run Freezes sim when bank D is greater than 207 steps O Place Rod Control Switch to Manual after simulator goes to freeze.
| |
| O Inform the examiner that the simulator setup is complete.
| |
| O Go to RUN when the examinee is given the cue sheet.
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| l t
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| LJCNRC-6 PAGE 6 OF 6 REV.O
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| i l .
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| Nuctuaa POWER GENERATION ,
| |
| l D%BLO CANYON POWER PLANT k Joe PERFORMANCE MEASURE l
| |
| Number: LJCNRC-7 Title- IESPOND TG A PRESSURIZER LOW PRESSURE ALARM '
| |
| l l Examince:
| |
| Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes i Comments:
| |
| | |
| ==References:==
| |
| AR PK05-17, PZR PRESSURE LOW, Rev. 9 OP AP-13, Malfunction of Reactor Pressure Control System, Rev. 4A i Alternate Path: Yes X No Time Critical: Yes No X Time Allotment: 10 minutes Critical Steps: 9,10 l
| |
| Job Designation: SRO K/A Number: 010000A4.01(3.5), A4.02(3.4)
| |
| I AUTHOR: JOHN BECERRA DATE: 11/12/98 l
| |
| l REVIEWED BY: NA DATE: NA f JPM COORDINATOR l
| |
| APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
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| | |
| r JPM TITLE: RESPOND TO A PRESSURIZER LOW PRESSURE ALARM JPM NUMBER: LJCNRC-7 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the l performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed.
| |
| Required Materials: None Initial Conditions: Unit 1 is operating at 50% power and ramping at 3 MW/ min to full power following repairs made to MFP l-2. All systems are aligned for normal full power operation.
| |
| Initiating Cue: PK05-17, PZR PRESSURE LOW, alarm is received. Respond to the alarm.
| |
| . Task Standard: All actions required to address the low pressure condition are complete.
| |
| LJCNRC-7 PAGE2OF7 REV.O
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| JPM TITLE: RESPOND TO A PRESSURIZER LOW PRESSURE ALARM JPM NUMBER: LJCNRC-7 lNSTRUCTOR WORKSHEET Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. Obtain the correct procedure Note: Operator may take actions necessary to place the plant in a stable condition without procedural guidance.
| |
| Appropriate procedure (s) should then be referenced to verify actions. i 1.1 Operator references AR PK05-17, PZR PRESSURE LOW.
| |
| Step was: Sat: Unsat *
| |
| : 2. Perform channel check. 2.1 Operator observes pressudzer pressure channels (PI-455,456,457, 474) indication to determine if a channelis affected Step was: Sat: Unsat i
| |
| : 3. Check pressurizer control system linear 3.1 Operator observes pressunzer ;
| |
| potentiometer. pressure control system 10 turn potentiometer, HC-455K, at 7.88 turns (S.P. = 2235 psig). i Step was: Sat: Unsat
| |
| : 4. Obtain the correct procedure. 4.1 Operator references OP AP-13, Malfunction of Reactor Pressure Control System.
| |
| Step was: Sat: Unsat
| |
| : 5. Stop any load changes. 5.1 Operator verifies no load changes are being performed.
| |
| 5.2 Operator depresses the HOLD'outton on DEHC Step was: Sat: Unsat
| |
| ' Denotes an entry required on the JPM cover sheet.
| |
| " Denotes a C 'ical Step.
| |
| ~
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| LJCNRC-7 PAGE 3 OF 7 REV O
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| I l
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| JPM Tm.E: RESPOND TO A PRESSURLIER LOW PRESSURE ALARM JPM NUMBER: LJCNRC-7 lNSTRUCTOR WORKSHEET
| |
| : 6. Check pressurizer PORVs. 6.1 Operator verifies pressure less than 2335 psig 6.2 Verifies pressurizer PORVs CLOSED, green lights ON.
| |
| Step was: Sat: Unsat
| |
| : 7. Check pressurizer safeties. 7.1 Operator verifies sonic flow indicators approximately 0 7.2 Verifies tail pipe temperature is less than or equal to 185 .
| |
| Step was: Sat: Unsat
| |
| : 8. Check pressurizer spray valves 8.1 Operator observes spray valve PCV-CLOSED. 455A CLOSED, green light ON and PCV-455B OPEN, green light OFF.
| |
| 8.2 Observes pressurizer pressure less than 2260 psig.
| |
| 8.3 Takes manual control of PCV-455B and attempts to close the valve.
| |
| 8.4 Observes spray valve PCV-455B still OPEN, green light OFF.
| |
| Step was: Sat: Unsat I
| |
| i
| |
| ! l l
| |
| l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJCNRC-7 PAGE 4 OF 7 REV.O
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| | |
| c JPM TITLE: RESPOND TO A PRESSUREER LOW PRESSURE ALARM JPM NUMBER: LJCNRC-7 lNSTRUCTOR WORKSHEET I
| |
| : 9. " TRIP the Reactor. 9.1 Operator locates the Reactor Trip -
| |
| switch on either CC-1 or VB-2 9.2 Turns switch to the TRIP position.
| |
| 9.3 Verifies all control rods inserted, DRPI rod bottom lights, red light ON I
| |
| I 9.4 Verifies Reactor Trip breakers OPEN, '
| |
| RTB A&B green lights ON, red light OFF.
| |
| 9.5 Verifies power decreasing, NI-41 through NI-44, less than 5% and decreasing.
| |
| 9.6 Verifies Turbine trip, all governor and ,
| |
| stop valve indicate CLOSED, green I light ON 9.7 Verifies 4kV vital buses energized Step was: Sat: Unsat
| |
| : 10. " Stop RCP(s) supplying the failed 10.1 Operatorlocates the STOP/ START spray valve. switch for RCP l-2.
| |
| 10.2 Turns switch to Stop position and verifies RCP tripped, red light OFF, green light ON, amp meter ZERO.
| |
| 1 Note: Inform Examinee that the JPM is l complete once the immediate actions of EP E-0 are completed.
| |
| Step was: Sat: Unsat Stop Time:
| |
| Total Time: (Enter total time on the cover page)
| |
| ' Denotes an entry required on the JPM cover sheet.
| |
| " Denotes a Critical Step.
| |
| LJCNRC-7 PAGE 5 OF 7 REV.O
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| r-JPM TITLE JPM NUMBER: LJCNRC-7 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is operating at 50% power and ramping at 3 MW/ min to full power following repairs made to MFP l-2. All systems are aligned for normal full power operation.
| |
| Initiating Cue: PK05-17, PZR PRESSURE LOW, alarm is received. Respond to the alarm.
| |
| Task Standard: All actions required to address the low pressure condition are complete.
| |
| I l
| |
| 1 l
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| LJCNRC-7 PAGE 6 OF 7 REV.O
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| | |
| F JPM TITLE: RESPOND TO A PRESSURIZER LOW PRESSURE ALARM JPM NUMBER: LJCNRC-7 ATTACHMENT 1, SIMULATOR SETUP O Initialize the simulator to IC-3 (50%, BOL).
| |
| O Place DER MW feedback IN O Set reference load to 1000 MW D Establish a 3 MW/ min ramp O Enter drill file 6057 or manually insert the following:
| |
| Command Description
| |
| : 1. cnv res2 2,0.3,0,0,d,0 Fails PCW455B partially open (valve will not close in AUTO or MANUAL)
| |
| : 2. run 10 Runs simulator for 10 seconds O Inform the examiner that the simulator setup is complete.
| |
| O Go to RUN when the examinee is given the cue sheet.
| |
| l LJCNRC-7 PAGE 7 OF 7 REV.0 l
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| p NUCLEAR POWER GENERATION -
| |
| DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: LJPNRC-1
| |
| | |
| ==Title:==
| |
| ALIGN ALTERNATE AFW FROM THE FIRE WATER STORAGE TANK Examinee:
| |
| Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments: Requires PG&E "909" Key l
| |
| l
| |
| | |
| ==References:==
| |
| OP D-1:V, Auxiliary Feedwater System - Altemate Auxiliary Feedwater Supplies, Rev.12 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 20 minutes Critical Steps: 2,3,6 Job Designation: SRO K/A Number: 061000K4.01(4.2)
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| AUTHOR: JOHN BECERRA DATE: 11/12/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
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| l
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| , JPM T'rLE: AUGN ALTERNATE AFW FROM THE FIRE WATER STORAGE TANK JPM NUMBER: LJPNRC-1 lNSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the l evaluator. The student will be given the initial conditions, initiating cue, -
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| and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step l with which to begm.
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| 1 Required Materials: PG&E "909" key. The operator must have a "909" key in possession or be able to obtain one.
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| Initial Conditions: The Unit 1 AFW pumps are supplied from the raw water reservoir due to the CST being less than 10%. l l
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| Initiating Cue: A seismic event occurs with a resultant loss of AFW. The Shift Foreman directs you to line up the next preferred water source to the AFW system and vent the pumps per OP D-1:V. ,
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| 1 l
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| Task Standard: AFW pump suction is supplied from the preferred water source and the pumps are vented.
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| l l
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| LJPNRC-1 PAGE 2 OF 5 REV.0
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| , JPM TITLE: AllGN ALTERNATE AFW FROM THE FIRE WATER STORAGE TANK JPM NUMBER: LJPNRC-1 lNSTRUCTOR WORKSHEET Start Time:
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| Step Expected Operator' Actions
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| : 1. Check open FP-0-4. 1.1 Operator locates FP-0-4 in the FWST l vault. l l
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| 1.2 Checks valve open. I Step was: Sat: Unsat
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| : 2. *
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| * Open FP-0-306 and FP-0-307. 2.1 Operator locates FP-0-306 and 307, located above the MU water pumps.
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| Note: These valves are normally chained and locked closed.
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| 2.2 Unlocks and opens the valves.
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| Step was: Sat: Unsat l
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| l
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| : 3. ** Close MU-0-1557. 3.1 Operator locates MU-0-1557 in the ,
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| overhead by the RWST vault,100' I elevation.
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| 3.2 Closes MU-0-1557.
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| Step was: Sat: Unsat
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| : 4. Check open FCV-436 and 437. 4.1 Operator locates FCV-436 and 437 in AFW pump room 11,100' elevation.
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| 4.2 Open FCV-436 and 437.
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| Note: FCV-436 and 437 would be open already based on the initial conditions given in the JPM.
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| Cue: Valves are open.
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| Step was: Sat: Unsat
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| * Denotes an entry required on the JPM cover sheet.
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| ** Denotes a Critical Step.
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| LJPNRC-1 PAGE 3 OF S REv. O
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| 1 l
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| l JPM TITLE: ALIGN ALTERNATE AFW FROM THE FIRE WATER STORAGE TANK JPM NUMBER: LJPNRC-1 lNSTRUCTOR WORKSHEET
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| : 5. Notify control room of alignment. 5.1 Operator calls control room and informs them of alignment.
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| Cue: You are to continue with step 6.2.6 I
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| Note: FW-1-121,159,180, AFW pump j suction valves from the CST, l would already be closed based on initial conditions given in the JPM.
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| Step was: Sat: Unsat j l
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| : 6. ** Vent the AFW pumps to remove 6.1 Operator locates the AFW pump vents:
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| trapped air. AFW ppl1 - FW-1-127
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| - FW-1-128 AFW pp12 - FW-1-165 AFW ppl3 - FW-1-186 by their associated pumps.
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| 6.2 Removes vent caps and opens vent valves to vent pumps until'. rapped air is removed. l Cue: The pumps are vented.
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| 6.3 Shuts vent valves and repleces vent caps.
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| Step was: Sat: Unsat Stop Time:
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| Total Time: (Enter total time on the cover page)
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJPNRC-1 PAGE 4 0F 5 REv.0
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| JPM TITLE: AUGN ALTERNATE AFW FROM THE FIRE WATER STORAGE TANK JPM NUMBER: LJPNRC-1 lNSTRUCTOR WORKSHEET l
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| {
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| I Initial Conditions: The Unit 1 AFW pumps are supplied from the raw water reservoir due to the CST being less than 10%.
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| i Initiating Cue: A seismic event occurs with a resultant loss of AFW. The Shift 1 Foreman directs you to line up the next preferred water source to the I AFW system and vent the pumps per OP D-1:V.
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| l Task Standard. AFW pump suction is supplied from the preferred water source and the i pumps are vented.
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| l l
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| l l
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| * Denotes an entry required on the JPM cover sheet.
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| *
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| * Denotes a Critical Step.
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| LJPNRC-1 PAGE 5 OF 5 REV.O
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| / *** ISSUED FOR USE BY: DATE: EXPBES:
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| NUMBER OP D4:V -
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| PACIFIC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION REVISION 12 PAGE 1 OF 6 DIABLO CANYON POWER PLANT
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| . OPERATING PROCEDURE UNIT TITLE: AUYtrJARY FEEDWATER SYSTEM - ALTERNATE AUXILIARY FEEDWATER SUPPLIES l Tb5 91 l l
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| EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED
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| : 1. SCOPE l 1.1 This procedure is referred to whenever :he Condensate Storage Tank (CST) is less than 10% and alternate Auxiliary Feedwater (AFW) to the steam generators is required.
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| : 2. DISCUSSION 2.1 This procedure provides the manwy instructions for supplying alternate soure.s of feedwater to the AFW System. 'Ibese sources of APW are listed in preferred order of use.
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| : 3. RESPONSIBILITIES 3.1 Shift Foreman is responsible for system ahgnment and operation of equipment required by this procedure.
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| : 4. PREREOUISITES 4.1 The following systems or components should be in servme as applicaS' to the specifi, water supply being used:
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| 4.1.1 Main Steam 4.1.2 Plant Electrical 4.1.3 Fire Water Storage Tank 4.1.4 Condensate Storage Tank 4.1.5 Condenser Hotwell 4.1.6 Makeup Water System 4.1.7 Portable Diesel Powered Fire Pumps
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| : 5. PRECAUTIONS AND LIMITATIONS 5.1 This procedure shall not be used except in the event the condensate storage tank is not available as the normal source of AFW System supply.
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| 5.2 Vo!ume 9 Section IF contains data on the CST volume required considering time maintaining hot standby and plant coo'.down rate.
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| 5.3 AFW Pp recirc returns to the CST at 50 gpm per pump and must be accounted for when makmg; " time remaining" calculations for AFW supplies other than the CST. ,
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| 00873712.DoA 02 I I 0923.0847
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| i PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP D-1:V -
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| DIABLO CANYON POWER PLANT REVISION 12 __..
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| PAGE 2OF6 TITLE: AUXILIARY FEEDWATER SYSTEM - ALTERNATE UNIT I AUXILIARY FEEDWATER SUPPLIES
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| : 6. INSTRUCTIONS 6.1 TRANSFER OF AUXILIARY FEEDWATER PUMP SUCTION SUPPLY TO THE ,
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| RAW WATER RESERVOIR 6.1.1 Locally open FCV-436 for AFW Pp No. I fully (100' Elev., AFW Pp 1-1 room).
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| I 6.1.2 locally open FCV-437 for AN Pp Nos. 2 and No. 3 fully (100' Elev.,
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| AFW Pp 1-1 room).
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| 6.1.3 Close AFW Pp suctions from CST: ,
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| 4 AFW Po 4 1 FW-1-121 2 FW-1-159 3 FW-1-180 NOTE: Continue to monito Raw Water Supply'and CST level since AFW Pp recire returns to the CST.
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| 6.1.4 If suction to the AFW Pps was lost, vent the AFW Pp castings as necessary to remove trapped air: l AFW Po 1 FW-1-127 FW-1-128 2 FW-1-165 3 FW-1-186 6.2 TRANSFER OF AUXILIARY FEED PUMP SUCTION SUPPLY FROM THE CST TO THE FWST 6.2.1 OPEN or Check OPEN FP-0-4 (115' Elev., in FWST vault).
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| 6.2.2 IF AFW Pump supply is being swapped from the CST to the FWST, THEN do steps a and b in sequence, OTHERWISE do step b before step a.
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| : a. CLOSE MU-0-1557 (100' Elev., chain operated, in hallway near RWST vault).
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| : b. OPEN FP-0-306 and FP-0-307 (chain operated, nonnally locked closed 100' Elev., above MU Water pps).
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| 6.2.3 Locally open FCV-436 and 437 fully (100' Elev., AFW Pp 1-1 room).
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| 6.2.4 Notify the Control Room that the suction for the AFW Pps is now available from the FWST.
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| 00873712.DoA 02 2 I o923.0847
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| NUMBER OP D-ItV -
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| PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT REVISION 12 ' _.
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| PAGE 3OF6 TITLE: AUXILIARY FEEDWATER SYSTEM - ALTERNATE UNIT 1 AUXILIARY FEEDWATER SUPPLIES 6.2.5 Close AFW Pp suctions from CST:
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| AFW Po 1 FW-1-121 2 FW-1-159 3 FW-1-180 6.2.6 If suction to the AFW Pps was lost, vent the AFW Pp casings as necessary to remove trapped air:
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| AFW Po 1 FW-1-127 FW-1-128 2 FW-1-165 3 FW-1-186 6.3 MAKEUP TO THE CONDENSATE STORAGE TANK FROM THE CONDENSER HOTWELL If the condensate pumps are available, refill the CST from the condenser hotwell via the condensate reject path.
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| 6.3.1 Place to " CONT ONLY," LCV-12 Selector Switch.
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| 6.3.2 Take manual control of LCV-12 Cond Hotwell Rejection Cont at VB3 and fully open LCV-12.
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| 6.3.3 Open or verify open FCV-230.
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| 6.3.4 Start a condensate and booster pump set.
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| 6.3.5 Monitor increasing CST level (VB3) and decreasing hotwel' level (VB3).
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| I 6.3.6 Monitor condensate and booster pump amps.
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| 6.3.7 Use normal AFW Pp suction supply.
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| 00873712.DOA 02 3 1 0923.o847
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| NUMBER OP D-1:V -
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| . PACIF1C GAS AND ELECTRIC COMPANY '
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| DIABLO CANYON POWER PLANT REVISION 12 -
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| PAGE 4 OF 6 TITLE: AUXILIARY FEEDWATER SYSTEM - ALTERNATE UNIT I AUXILIARY FEEDWATER SUPPLIES Am ge. '
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| 6.4 SUPPLY AUXILIARY FEEDWATER PUMP SUCTION FROM THE MAKEUP WATER TRANSFER TANK VIA THE MAKEUP WATER TRANSFER PUMPS Refer to Attachment 9.2 of this procedure.
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| NOTE: AFW Pp flow will be limited to Makeup Water Transfer Pump capacity.
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| 6.4.1 Stop the Makeup Water Transfer Pumps.
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| 6.4.2 Open or check open MU-0-696, Makeup Demin Outlet Iso! Valve to Transfer Tank 0-1.
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| 6.4.3 Open or check open MU-0-788, Makeup Demin Outlet to Transfer Tank 01 Throttle Valve.
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| 6.4.4 Open or check open MU-0-698, Makeup Demin Outlet Isol Valve to Transfer Pp Suction.
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| 6.4.5 Open or check open MU-0-715 and 0-710, Makeup Water Transfer Pp Suction Isol Valve from Transfer Tank 0-1.
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| 6.4.6 Close or check closed MU-0-712 and MU-0-713, Makeup Water Transfer Pp 0-1 Disch Isol Valves.
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| 6.4.7 Close or check closed MU-0-709 and MU-0-708, Makeup Water Transfer Pp 0-2 Disch Isol Valves.
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| 6.4.8 Close or Check closed MU-0-714 and MU-0-711, Makeup Water Transfer Pump Suetions from U-1 and U-2 CST's.
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| 6.4.9 Open or check open MU4-727, Makeup Water Transf:r Pp 0-1 Disch Isol Valve to RAW WTR RESERVOIR, CCW Surge Tanks or Makeup Demineralizers and MU-0-728, Makeup Water Transfer Pp 0-2 Disch Isol Valve to RAW WTR RESERVGIR, CCW Surge Tanks or Makeup Demin.
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| 6.4.10 Open or check open MU-0-743, Makeup Water Transfer Pps Disch Hdr Isol Valve to RAW WTR RESERVOIR.
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| 6.4.11 Open or check open MU-0-278, Crosstie valve octween 6" and 8" headers.
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| 6.4.12 Close or check closed MU-277.
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| 6.4.13 Close or check closed MU-0-280, 8" header Isol Valve downstream of domestic water treatment system tap.
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| 6.4.14 Open or check open MU-0-283, Strainer 97 Inlet Isol Valve.
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| 6.4.15 Open or check open MU-0-284, Strainer 07 Outlet Isol Valve.
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| 6.4.16 Open or check open MU-0-1557, Isol Valve in the 8" header.
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| 00873712.DOA 02 4 1 0923.0847
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| I NUMBER OP D 1:V -
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| - PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT REVISION 12 _.
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| PAGE 5 OF 6 TITLE: AUXILIARY FEEDWATER SYSTEM - ALTERNATE UNIT 1 AUXILIARY FEEDWATER SUPPLIES 6.4.17 Open or check open MU-1-297, Isol Valve to AFW Pps suction from 8" header.
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| 6.4.18 Notify the Control Room that the suction for the AFW Pps is now available from the Makeup WaterTransfer Tank.
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| 6.4.19 locally open or verify fully, open FCV-436, AFW p No.1 RAW WTR SUPPLY and FCV-437, AFW Pp No. 2 and No. 3 RAW WTR SUPPLY (100' Elev, AFW Pp 1-1 room).
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| 6.4.20 .ctart the Makeup Water Transfer pumps to supply AFW Pp suction.
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| 6.4.21 Close AFW Pp suctions from CST: l AFW Po 1 FW-1-121 2 FW-1-159 3 FW-1-180 i NOTE: Continue to monitor CST 1evel since AFW Pp recirc returns to the CST.
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| 6.4.22 If suction to the AFW Pps was lost, vent the AFW Pp casings as necessary to remove trapped air:
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| AFW Po 1 FW-1-127 I I
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| FW-1-128 j 2 FW-1-165 3 FW-1-186 l
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| l l
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| 00873712.DOA 02 5 I o923.0847
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| | |
| NUMBER OP D-ItV -
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| PACIFIC GAS AND ELECTRIC COMPANY '
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| DIABLO CANYON POWER PLANT REVISION 12 _.
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| PAGE 6OF6 !
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| TITLE: AUXILIARY FEEDWATER SYSTEM - ALTERNATE UNIT 1 AUXILIARY FEEDWATER SUPPLIES 6.5 AUXILIARY FEED PUMP SUCTION SUPPLY FROM THE FIRE WATER STORAGE TANK USING THE PORTABLE DIESEL POWERED FIRE WATER i 1
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| PUMP 6.5.1 Lineup the fire water tank to the AFW Pp suction as per Section 6.2 of this procedure.
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| 6.5.2 Refer to OP K-2A:III, " Alternate Methods of Pressurizing and Filling the Fire Water System" to fill the FWST using the fire water yard loop from the i Main Condenser Hotwell.
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| 6.5.3 If the Condenser Hotwell is not available, refer to OP K-2A:Ill to establish a j i
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| siphon from the Raw Water Reservoir to the makeup water header to the AFW Pp suction.
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| NOTE: Upon depletion of the Raw Water Reservoir, refill the Raw Water Reservoir from Diablo Creek using the portable diesel driven fire water pump (refer to Drawing 107031, Sheet 26).
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| 6.5.4 Upon exhaustion of all other water sources, refer to OP K-2A:III to establish makeup to the fire water tank from the Auxiliary Sahwater System.
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| : 7. REFERENCES 7.1 Makeup Water System Description, Chapter F-3a.
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| 7.2 OP K-2A:III, " Alternate Methods of Pressurizing and Filling the Fire Water System."
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| 7.3 OVIDs- l 7.3.1 106703, Shdet 3. l 7.3.2 106716, Sheets 11 and 16.
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| : 8. RECORDS None
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| : 9. ATTACHMENTS 9.1 "AFW Supply From FWST," 06/05/%
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| 9.2 "AFW Supply From Makeup Transfer Tank," 06/30/95
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| : 10. SPONSOR C. P. Dunlap 00873712.DOA 02 6 1 0923.0847
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| . 06/05/96 Page1of1 -
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| DIABLO CANYON POWER PLANT -
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| ( OP D-1:V .
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| ,, ATTACHMENT 9.1 TITLE: AFW SUPPLY FROM FWST l
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| FIRE WATER TANK l
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| l PVST VAULT FP44 I sts- S.o.
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| To Fire Pump Suchon MU w r' I
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| 100' WATER PW r)p's
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| '' OE XFER Pp4 N M_U.
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| MIN.
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| N 325 427 W1-3981W Package W 7 '
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| h MO41567 W RAhVWATER L 2 HEADER i
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| "P -/-x FCV436 W12p?
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| ><A FCV436 "P l
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| -X h2-2ps 1W BEHIND AFWPp 1W BEHND AFW Pp TWUNNE TUR8ME UNIT 1 2
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| No. 3 2 "Xpcy437 Fbe Wubw Tank To O ''
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| 'J
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| -\ E3 2
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| FCV h MU41667IS AN8"VALVELOCATED AT *112' LEVEL DETWEEN THE MU WATER XFER PUMPS AND THE PW PUMPS. CHAM OPERATED FROM THE tw LEVEL 00873701 i
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| 00873712.DOA 02 7 I 0923.0847
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| i NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: LJPNRC-2
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| ==Title:==
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| CROSS TIE CCW SYSTEM BETWEEN UNITS Examinee:
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| Evaluator:
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| Print Signature Date Results: Sat Unsat . Total Time: minutes Comments:
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| ==References:==
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| OP AP SD-4, Loss of Component Cooling Water, Appendix D, Rev.I1 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 1, 3 Job Designation: SRO K/A Number: 008000A202 (3.5)
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| . . %OR:
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| JOHN BECERRA DATE: 11/12/98 REVIEWED BY NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
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| : JPM TITLE: CROSS TIE CCW SYSTEM BETWEEN UNITS JPM NUMBER: LJPNRC-2 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin.
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| Required Materials: 909 Key Initial Conditions: PK01-07, CCW SYS SURGE TK LVL/MK-UP and PK01-14, CCW SUE 7E TANK PRESSURE are in alarm due to a leak in the CCW surt .aak. The CCW system has been walked down and determined to be intact with the exception of the leak which is limited to the surge tank. The Unit 2 CCW system is available for cross-tie.
| |
| Initiating Cue: You are directed by the Shift Foreman to Cross-Tie CCW between .
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| l Units, per OP AP SD-4, Appendix D, beginning at step 2.
| |
| Task Standard: CCW system is cross-tied.
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| LJPNRC-2. DOC PAGE 2 OF 5 REV.O
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| JPM TITLE: CROSS TIE CCW SYSTEM BETWEEN UNITS JPM NUMBER: LJPNRC-2 i INSTRUCTOR WORKSHEET j 1
| |
| 1 Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. *
| |
| * Isolate Surge Tank on Unit 1 1.1 Operator locates CCW-51 (100' Pen Area GW), unlocks and closes.
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| Cue: CCW-51 is closed.
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| ***********************************A 1.2 Operator locates CCW-52 (700' Pen Area GW), unlocks and closes. l Cue: CCW-52 is closed.
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| Step was: Sat: Unsat J
| |
| : 2. Isolate Makeup to Surge Tank on 2.1 Operator locates CCW-61 and closes Unit 1. (100' Pen Area GW).
| |
| Cue: CCW-61 is closed.
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| 2.2 Operator locates CCW-64 and closes (100' Pen Area GW).
| |
| Cue: CCW-64 is closed.
| |
| Note: CCW-61 & 64 are scaled valves.
| |
| The operator should discuss completing a scaled valve change form after completing the task.
| |
| Step was: Sat: Unsat
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
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| LJPNRC-2. DOC PAGE 3 OF 5 REV.O
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| | |
| JPM TITLE: CROSS TIE CCW SYSTEM BETWEEN UNITS JPM NUMBER: LJPNRC-2 lNSTRUCTOR WORKSHEET
| |
| : 3. *
| |
| * Cross-Tie CCW between Units. 3.1 Operator locates CCW-1-356 and checks open (78' ASDR Mezzanine).
| |
| Cue: CCW-1-356 is open.
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| 3.2 Operator locates CCW-2-356 and opens (78' ASDR Mezzanine).
| |
| Cue: CCW-2-356 is open.
| |
| 3.3 Operator checks makeup capability adequate by checking Unit 2 CCW Surge Tank is maintaining level.
| |
| Cue: Unit 2 Control Room reports Unit 2 CCW Surge Tank levelis 54% and constant.
| |
| Step was: Sat: Unsat Stop Time:
| |
| Total Time: (Enter total time on the cover page)
| |
| I l
| |
| l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJPNRC-2. DOC PAGE4OF5 REV.O
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| | |
| JPM TrrLE: JPM NUMBER: LJPNRC-2 EXAMINEE CUE SHEET l
| |
| Initial Conditions: PK01-07, CCW SYS SURGE TK LVUMK-UP and PK01-14, CCW l SURGE TANK PRESSURE are in alarm due to a leak in the CCW surge tank. The CCW system has been walked down and determined to be intact with the exception of the leak which is limited to the surge tank. The Unit 2 CCW system is available for cross-tie.
| |
| Initiating Cue: You are directed by the Shift Forernan to Cross-Tie CCW between ,
| |
| Units, per OP AP SD-4, Appenuix D, beginning at st 'p 2.
| |
| I Task Standard: CCW system is cross-tied. l i
| |
| l l
| |
| i i
| |
| LJPNRC-2. DOC PAGE 5 OF 5 REv. O
| |
| | |
| PACIFIC GAS AND ELEC. JC COMPANY NUMBER OP AP SD4 DIABLO CANYON POWER PLANT REVISION 11 - ~
| |
| PAGE 22 OF 23 TITLE: LOSS OF COMPONENT COOLING WATER UNITS 1AND2 APPENDIX D LOSS OF SURGE TANK /CCW SYSTEM CROSS TIE BETWEEN UNITS l
| |
| ACTION / EXPECTED RESPONSE RESPONSE NQT OBTAINED
| |
| : 1. CHECK CCW System Status: GO To Step 13, page 8.
| |
| : a. Check damage limirad to surge tank e Surge tankisolable from l
| |
| CCW tvstem
| |
| * CCW System otherwise intact
| |
| : b. Check opposite 'Init CCW System is intact and available for crossaie
| |
| : c. Check Header C will remmin in service on bodt Units
| |
| : 2. ISOLATE Surne Tank and Makeup on Affected Unit:
| |
| : a. UNLOCK and CLOSE CCW-51, Surge Tank to Header B isolation (100' Pen Are GW)
| |
| : b. UNLOCK and CLOSE CCW-52, Surge Tank to Header A isolation (100* Pen Area GW)
| |
| : c. CLOSE CCW 41, LCV-69 downstream isolation (100' Pen Area GW)
| |
| : d. . CLOSE CCW-64, LCV-70 downstream isolation (100' Pen Area GW)
| |
| Ot070411.DOA 02 22 1005.0304
| |
| | |
| 1 NUMBER OF AP SD-4 PACB1C GAS AND ELEL stIC COMPANY DIABLO CANYON POWER PLANT REVISION 11 _.
| |
| PAGE 23 OF 23 UNITS 1AND2 i TITLE: LOSS OF COMPONENT COOLING WATER APPENDIX D (Commued) 1
| |
| : 3. CROSS-TIE CCW Between Units:
| |
| : a. OPEN CCW-1-356, CCW Remrn a. Open CCW-1-204, WGC 0-1 Seal Cooler l
| |
| from Aux Stm Drain Receiver Remrn. (55' Pipe Tunnel West End) 98' ASDR Mensmne)
| |
| : b. OPEN CCW-2-356, CCW Remrn b. Open CCW-2-204;WGC 0-1 Seal Cooler from Aux Sem Drain Receiver Remrn. (55' Pipe Tunnel West End)
| |
| G8' ASDR Mensmne)
| |
| : c. Check unaffected Unit's Surge Tank c. Isolate Cross-tie path until leakage is M=intaining Level and Makeup isolated.
| |
| Capability ada=='a
| |
| : 4. GO TO STEP 13. name 8 i
| |
| 1 5 ,
| |
| l 2005.0304 OltnO411.DOA 02 23
| |
| | |
| s NUCLEAR POWER GENERATION ,
| |
| D:ABLO CANYON POWER PLANT JOB PERFORMANCE MEASURE Number: LJPNRC-3
| |
| | |
| ==Title:==
| |
| PERFORM A LOCAL START OF A DIESEL GENERATOR Examinee:
| |
| Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
| |
| l l
| |
| l l
| |
| | |
| ==References:==
| |
| OP AP-8A, Control Room Inaccessibility, Establishing Hot Standby, ;
| |
| Rev.9B Alternate Path: Yes X No l Time Critical: Yes No X Time Allotment: 10 minutes Critical Steps: 1,2,3,4,7,8 Job Designation: SRO K/A Number: 064000A4.01(4.3)
| |
| AUTHOR: JOHN BECERRA DATE: 11/12/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| . JPM TITLE: PERFORM A LOCAL START OF A DIESEL GENERATOR JPM NUMBER: LJPNRC-3 lNSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the l performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner wi:1 then ask if any clarifications are needed. The examinee may be given the procedure and told the step l with which to begin. I Required Materials: None Initial Conditions: A fire in the vertical boards required an evacuation of the control room.
| |
| Plant control was established from the Hot Shutdown Panel.
| |
| Initiating Cue: The Shift Foreman directs you to establish local control and start diesel generator 12 in accordance with OP AP-8A, Appendix B, Step 6. l l
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| Task Standard: Diesel generator 12 is running.
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| LJPNRC-3 PAGE 2 OF 6 REV.O
| |
| :. JPM TrrLE: PERFORM A LOCAL START OF A DIESEL GENERATOR JPM NUMBER: LJPNRC-3 lNSTRUCTOR WORKSHEET Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. *
| |
| * Place the droop switch on the Note: Operator should ensure excitation cubicle to the ISOC mode. procedure NOTE is complied with. ]
| |
| 1 1.1 Operator locates the diesel generator 12 droop switch on the excitation i cubicle. f
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| )
| |
| 1.2 Places the droop switch to the ISOC 1 position. l Step was: Sat: _ Unsat: l 1
| |
| : 2. *
| |
| * Verify AUTO / TEST selector switch 2.1 Operejor locates the AUTO / TEST on the local control panelis to the selector switch on the diesel generator I TEST position. 12 local control panel.
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| 1 I
| |
| 2.2 Places the AUTOffEST selector switch to the TEST position.
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| Step was: Sat: Unsat:
| |
| : 3. *
| |
| * Place the diesel generator control 3.1 Operator locates the diesel generator selection switch on the excitation 12 control selection switch on the cubicle to the LOCAL position. excitation cubicle.
| |
| 3.2 Places the diesel generator 12 control selector switch to the LOCAL position.
| |
| Step was: Sat: Unsat:
| |
| : 4. *
| |
| * Place the Appendix R fuse selector 4.1 Operator locates the 43DC-12 switch switch 43DC-12 to the BACKUP on the diesel generator 12 local
| |
| ' position. control panel.
| |
| 4.2 Places the Appendix R fuse selector switch to the BACKUP position.
| |
| Step was: Sat: Unsat:
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| " Denotes a Critical Step.
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| LJPNRC-3 PAGE 3 OF 6 REv. O
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| | |
| JPM TITLE: PERFORM A LOCAL START OF A DIESEL GENERATOR JPM NUMBER: LJPNRC-3 INSTRUCTOR WORKSHEET I
| |
| : 5. Depress the alarm relay reset push 5.1 Operator locates alarrn relay reset button. push button on diesel generator 12 local control panel.
| |
| 5.2 Depresses alarm relay rest pushbutton.
| |
| l Step was: Sat: Unsat: *
| |
| ]
| |
| : 6. Place the START /STOP switch on the 6.1 Operator locates the START /STOP local control panel to tne START switch at the diesel generator 12 local position. control panel.
| |
| 6.2 Places the START /STOP switch to the START position Cue: The diesel generator did not start.
| |
| 1 Step was: Sat: Unsat:
| |
| : 7. *
| |
| * Place the 125VDC control power 7.1 Operator locates control power transfer switch, EQD-12, to the transfer switch EQD-12 opposite the BACKUP position. diesel generator 12 local control panel.
| |
| 7.2 Moves the transfer switch down to the BACKUP PWR SUP position.
| |
| Cue: The normal power supply light is OFF and the backup power supply light is ON.
| |
| Step was: Sat: Unsat:
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
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| LJPNRC-3 PAGE 4 OF 6 REv. O
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| | |
| 1 JPM TITLE: PFRFORM A LOCAL START OF A DIESEL GENERATOR JPM NUMBER: LJPNRC-3 INSTRUCTOR WORKbHEET l
| |
| , i l !
| |
| ! 8. ** Attempt to restart the diesel 8.1 Operator locates the START /STOP l l generator. switch at the diesel generator 12 local control panel.
| |
| l 8.2 Places the START /STOP switch to the START position.
| |
| Cue: The diesel generator is running )
| |
| with normal operating parameters.
| |
| Step was: Sat: Unsat:
| |
| Stop Time:
| |
| Total Time: (Enter total time on the cover page) l l
| |
| 1 l
| |
| i I
| |
| (
| |
| l l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| LJPNRC-3 PAGE 5 OF 6 REV.O j l
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| | |
| )
| |
| JPM TITLE: JPM NUMBER LJPNRC-3 '
| |
| EXAMINEE CUE SHEET '
| |
| Initial Conditions: A fire in the vertical boards required an evacuation of the control room.
| |
| Plant control was established from the Hot Shutdown Panel.
| |
| Initiating Cue: The Shift Foreman directs you to establish local control and start diesel generator 12 in accordance with OP AP-8A, Appendix B, Step 6.
| |
| Task Standard: Diesel generator 12 is running.
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| l l
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| l LJPNRC-3 PAGE 6 OF 6 REV.O
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| | |
| PACIFIC GAS AND EL TRIC COMPANY NUMBER OP AP-8A DIABLO CANYON POWER PLANT REVISION 9B_..
| |
| PAGE 33 OF 43 TITLE: CONTROL ROOM INACCESSIBILITY.
| |
| UNIT ' 1 ESTABLISHING HOT STANDBY APPENDIX B (Continued) l ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: The Control Selection switch at the D/G excitation panel gg.1 be switched to the " LOCAL" position prior to operation of the Appendix R Fuse Selector switt.h (located on the D/G Control Panel) to the "Back-up" position. Failure to do so could result in a loss of all local D/G control.
| |
| l
| |
| : 6. ESTABTEH Local Control and Start Aonlicable Diesel Generator (s):
| |
| : a. Place Droop Switch on excitation cubicle to ISOC mode
| |
| : b. Verify AITTO/ TEST selector switch on local panelin TEST position
| |
| : c. Place D/G Control Selection switch on excitation cubicle to LOCAL position
| |
| : d. Place the Appendix R Fuse Selector switch (43DC-11,43DC-12, 43DC-13) to the BACKUP position
| |
| : e. Depress the Alarm Relay reset push button
| |
| : f. Place engine START /STOP switch on l the local panel to START position
| |
| : g. Verify D/G started g. Place the 125V DC Control Power !
| |
| Transfer Switch (EQDll,12,13) to its BACKUP position so that the backup l power supply light is lit. Attempt to restart the D/G.
| |
| 01052409. doe 02 33 0629.1041
| |
| | |
| E-s l JOB PERFORMANCE MEASURE TEST ITEM QUESTION # 1a:
| |
| Reference Allowed: YES Given the following:
| |
| . _ A design basis Large Break LOCA occurred on Unit 2.
| |
| l . Initial RWST level was 98% )
| |
| l
| |
| . Assume all equipment operates as designed.
| |
| Determine the approximate time for RWST level to decrease to the cold leg i recirculation transfer criteria level.
| |
| 1 1
| |
| QUESTION # 1b:
| |
| Reference Allowed: YES Current plant conditions:
| |
| . Mode 5 I e Solid Plant Conditions e RCS temperature 168 'F e RCS pressure 200 psig I
| |
| * 1-1 RHR system in operation
| |
| . 1-2 RHR system in standby lined up to the RCS The air supply line to CVCS-1-PCV-135, Letdown Press Control Valve, is broken off by a maintenance worker.
| |
| : 1. Predict the RCS pressure response.
| |
| -2. Give your reasons for your response.
| |
| STUDENT COPY
| |
| | |
| 1 1
| |
| JOB PERFORMANCE MEASURE TEST ITEM i
| |
| )
| |
| QUESTION # 2a:
| |
| Reference Allowed: YES Unit 1 is currently in Mode 4 with preparations in progress to enter Mode 3.
| |
| While performing STP l-1D, MODES 1,2, and 3 Monthly Checkt:st, valve CS-1-8992, Spray Additive Tank to Eductor, is found closed.
| |
| Attempts to open the valve have failed. l What, if any Tech. Spec. actions apply?
| |
| l QUESTION # 2b:
| |
| 1 Reference Allowed: NO
| |
| ' Unit 2 experienced a DBA LOCA and Containment Spray has actuated.
| |
| Containment Spray Pump 2-1 tripped on overcurrent.
| |
| : 1. What is the minimum assumed equipment available to protect the Containment function under these circumstances?
| |
| : 2. What is the Basis for this?
| |
| STUDENT COPY
| |
| | |
| I JOB PERFORMANCE MEASURE TEST ITEM QUESTION # 3a:
| |
| Reference Allowed: YES Explain the effect on reactor power calorimetric as the throat of the main feedwater flow venturies erode.
| |
| l QUESTION # 3b:
| |
| Reference Allowed: YES Unit 1 is operating at 50% power due to a curtailment to clean Circulating Water Tunnels. A Control Room / Electrical technician is performing STP l-2C1, " Removal of Power Range Channel from Service" for NI-41. During performance of the STP the technician mistakenly pulls the instrument fuses fo; NI-42, realizing his mistake he reinserts the fuses for NI-42 and immediately pulls the instrument fuses for NI-41. A Reactor Trip occurs immediately after the fuses for NI-41 were removed.
| |
| Explain what caused the Reactor Trip?
| |
| l i
| |
| STUDENT COPY
| |
| | |
| j JOB PERFORMANCE MEASURE TEST ITEM QUESTION # 4a:'
| |
| Reference Allowed: YES The following conditions exist:
| |
| . ECCS Accumulator 1-1 level 66% at 1730
| |
| . ECCS Accumulator 1-1 level 73% at 1830 -
| |
| Determine the leak rate from the RCS into the 1-1 Accumulator.
| |
| l QUESTION # 4b:
| |
| Reference Allowed: No An earthquake caused a small break LOCA to occur on Unit 1 along with the loss of 4kV bus G. EOP E-0 and E-1 have been completed. A plant cooldown and depressurization is in progress IAW EOP E-1.2, " Post LOCA Cooldown and Depressurization."
| |
| Describe what action (s) will need to be taken, regarding the ECCS Accumulators, in order to continue with the RCS cooldown and depressurization if 4kV bus G cannot be restored.
| |
| STUDENT COPY :
| |
| | |
| JOB PERFORMANCE MEASURE TEST ITEM QUESTION # Sa:
| |
| Reference Allowed: YES Unit 2 is in Mode'5, fuel in the core, with RCS loops filled, following a refueling outage.
| |
| Given the following equipment status:
| |
| . ASW Pump 2-1 Operable, not running
| |
| . ASW Pump 2-2 Running
| |
| . RHR Punip 2-1 Operable, not running
| |
| . RHR Pump 2-2 Running
| |
| . CCW Pump 2-1_ Running
| |
| . CCW Pump 2-2 Running
| |
| . CCW Pump 2-3 Operable, not running
| |
| . CCP 2-1 DC Open
| |
| . CCP 2-2 Operable, not running l
| |
| . PDP 2-3 Cleared The Control Room / Electrical Asset Team desires to run STP M-13H on Bus H, causing the diesel to start and its 4KV bus to load onto the diesel automatically.
| |
| What if any contingency actions should the Shift Foreman direct prior to the start of the test based on current plant conditions and equipment !!neup?
| |
| QUESTION # 5b:
| |
| Reference Allowed: YES Following a reactor trip from 100% power, off-site power has been lost and only D/G 2-3 is available to supply the Unit 2 4KV busses. In accordance with ECA-0.3 (Restore 4KV Busses) Appendix X (Crosstie of Vital Bus), D/G 2-3 is now supplying Vital Bus F and Vital Bus G.
| |
| Assuming zero overload hours for this year, what would be the maximum amount of time that D/G 2-3 could be operated at 2.96 MWs?
| |
| STUDENT COPY
| |
| | |
| i JOB PERFORMANCE MEASURE TEST ITEM !
| |
| 1 l
| |
| QUESTION # 6a:
| |
| Reference Allowed: YES Unit 2 Control Rods are being inserted in the MANUAL mode during a reactor shutdown.
| |
| Assuming Control Bank D rods are at 110 steps, how many steps must rods be inserted
| |
| - before Control Bank B rods begin to insert? -
| |
| QUESTION # 6b:
| |
| Reference Allowed:' YES i
| |
| Unit 1 is at 100% power with the following power range excore detector normalized '
| |
| values:
| |
| N41 N42 N43 N44 UPPER 0.96 0.95 0.96 0.99 LOWER 1.01 1.02 1.01 1.03
| |
| : 1. Calculate the OPTR.
| |
| : 2. Does the calculated QPTR exceed the Tech Epecs?
| |
| STUDENT COPY
| |
| | |
| p I
| |
| JOB PERFORMANCE MEASURE TEST ITEM l
| |
| l QUESTION # 7a: )
| |
| i ' Reference Allowed: YES 1
| |
| i i The plant is operating at 100% power when a pressurizer safety valve inadvertently lifts and sticks open.
| |
| With PRT pressure at 100 psig just before the rupture disk blows, what would be the approximate tail pipe temperature of the safety valve?
| |
| QUESTION # 7b:
| |
| Reference Allowed: YES Unit 1 is at 100% power with pressurizer heater group 1-2 being powered from breaker 52-1G-72.
| |
| What effect would a Safety injection signal have on pressurizer heater group 1-27 l
| |
| l l
| |
| STUDENT COPY
| |
| | |
| ' JOB PERFORMANCE MEASURE TEST ITEM QUESTION # 8a: - ,
| |
| Reference Allowed: YES The green light on FCV-95 for AFW Pp 1-1 is not lit and the operator suspects the l
| |
| power supply breaker may be tripped. j l
| |
| . What breaker feeds power to the green light?
| |
| I 1
| |
| I QUESTION # 8b: ;
| |
| Reference Allowed: YES Unit 2 was at 100% power when a reactor trip and loss of offsite power occurred.
| |
| The Condensate Storage Tank is at 53% and it will take 6 hours to cool down to 350 F. !
| |
| How long can the plant stay at Hot Standby before cooling Down?
| |
| l 1
| |
| l STUDENT COPY
| |
| | |
| l l
| |
| l JOB PERFORMANCE MEASURE TEST ITEM
| |
| . QUESTION # 9a:
| |
| Reference Allowed: YES While the Control Room Asset Team was performing testing in Train B of SSPS, an inadvertent Phase B occurred on Train B.
| |
| What effect did this have on the CCW system?
| |
| QUESTION # 9b:
| |
| Reference Allowod: No i Unit 1 Component Cooling Water (CCW) head tank pressure has dropped to less than i i
| |
| 19 psig due to a problem with the cover gas system.
| |
| What sequence of events could lead to flashing and water hammer in the CCW system if CCW head tank pressure is not restored?
| |
| : STUDENT COPY
| |
| | |
| l JOB PERFORMANCE MEASURE TEST ITEM QUESTION # 10a: )
| |
| Reference Allowed: YES Unit 1 was in Mode 5, with ALL D/Gs operable.
| |
| The SFM directed a Nuclear Operator to place D/G 1-1 on its backup DC source due to loss of the normal source.
| |
| What is the operability status of all the Unit 1 Diesel Generators?
| |
| I QUESTION # 10b:
| |
| Reference Allowed: YES Unit 1 is at 100% power, an operator was directed to start DG 1-1 using the undervoltage starting method, per STP M-9A DG 1-3 was started instead due to the ;
| |
| operator going to the wrong 4 kV bus to perform the required start.
| |
| What are the reportability requirements , if any for this Diesel Generator etart?
| |
| STUDENT COPY
| |
| | |
| I
| |
| \
| |
| s
| |
| ! JOB PERFORMANCE MEASURE TEST ITEM l
| |
| Task
| |
| | |
| ==Title:==
| |
| Transfer to CL Recirculation K/A
| |
| | |
| ==Reference:==
| |
| 005000K4.11 (3.9) 1 Reference Allowed: YES l
| |
| | |
| ==Reference:==
| |
| Vol. 9 p. IE-9.2b; STG B3, Obj.12, pp. 2-7,2-14,2-35; and j STG 12, Obj. 7, p. 3-9 !
| |
| QUESTION # 1a:
| |
| . A design basis Large Break LOCA occurred on Unit 2.
| |
| . Initial RWST level was 98%.
| |
| . Assume all equipment operates as designed.
| |
| Determine the approximate time for RWST level to decrease to the cold leg recirculation transfer criteria level.
| |
| ANSWER:
| |
| RWST @98% = 475538 gal.
| |
| RWST @33% = 177665 gal.
| |
| . Level difference = 297873 gal.
| |
| . CCP 550 gpm ea. (2) = 1100 gpm
| |
| . SIP 650 gpna ea. (2) = 1300 gpm l
| |
| . RHR 4500 gpm ea. (2) = 9000 gpm
| |
| . CSP 2466 gpm ea. (2) = 4932 gpm
| |
| . Total flow = 16332 gpm ,
| |
| Level difference divided by Total flow: 297873 /16332 = 18.2 minutes i 2 minutes.
| |
| Candidate's Response: SAT UNSAT ANSWER KJEY
| |
| | |
| ~~ ' ' "
| |
| 'DI1850 ANYON POWER PLANT bPERA'TIO S DATA STORAGE TANK VOLUME DATA ,
| |
| i REFUELING WATER STORAGE TANK ,
| |
| PERCENTINDICATED VS. VOLUME 1 i Gallom i Gallons i Gallons i Gallons 0 26438 25 141004 50 255571 75 370137 1 31021 26 145587 51 260153 76 374719 2 35603 27 150170 52 264736 77 379302 I 3 40186 28 154752 53 269318 78 383885 l 4 44769 29 159335 54 273901 79 388467 5 49351 30- 163918 55 278484 80 393050 6 53934 31 168500 56 283066 81 397633 7 58517 32 173083 57 287649 82 402215 8 63099 33 177665. 58 292232 83 406798 9 67682 34 182248 59 2 % 814 84 411381 10 72265 35 186831 60 301397 85 415 % 3 11 76847 36 191413 61 305980 86 420546 12 81430 37 195996 62 310562 87 425129 13 86012 38 200579 63 315145 88 429711 14 90595 39 205161 64 319728 89 434294 15 95178 40 209744 65 324310 90 438877 16 99760 41 214327 66 32889,3 91 443459 17 104343 42 218909 67 333476 92 448042 18 108926 43 223492 68 338058 93 452624 19 113508 44 228075 69 342641 94 457207 20 118091 45 232657 70 347224 95- 461790 21 122674 46 237240 71 351806 96 466372 22 127256 47 241823 72 356389 97 470955 23 131839 48 246405 73 360971 98 475538 24 136422 49 250988 74 365554 99 480120 100 484703 Total Contained Volume: 484703 gallons Unmeasured Volume: 26438 gallons Unusable Volume: 19780 gallons Page IE-9.2b Rev. 4 Date 04/25/95 I
| |
| 01268604.DOA 93 I I
| |
| | |
| , Major Components High Head injection System, continued 4
| |
| CCPs The diagram below illustrates the flow characteristics of the CCPs.
| |
| Ob)12
| |
| -m emo-
| |
| _ % com 2105)
| |
| - 70 mm-I ((1732) 3000 -
| |
| - 63 s
| |
| f (13o0) som-
| |
| - 50 #
| |
| , E
| |
| ,, case) -e g 1 iom- r g (ca) -so [
| |
| - 2o h ''*
| |
| Im am am a sm Fkm in GaNons Per Minute Ref. e5321o 145 ECC-05 The table below lists the design tiow characteristics of the CCPs.
| |
| Characteristic Flow rate Pressure Shutoff head 0 gpm 2592 psig Design flow rate 150 gpm 2550 psid Runout flow 550 gpm 650 psid Indications 'the table below lists the indications associated with the high head portion of the ECCS.
| |
| Ob) 9 Indication Can be read at ... Normal reading FI-917 Charging injection header VB2 0*
| |
| flow PI-947 Charging injection header pressure
| |
| * This number represents the expected reading when the unit is at 100 % power; dunng SI conditions the reading will vary in accordance with the RCS pressure.
| |
| l B3. DOC 2-7 REV 6
| |
| | |
| E Emergency Cire Cooling sysum Safety injection Pumps, continued i
| |
| Physical Each unit has two 100% capacity pumps with the following characteristics.
| |
| description l
| |
| Ob]12 Characteristic Details Type 10 stage, horizontal centrifugal pump.
| |
| Pump drive 4 kV 400 hp induction motor Design Pressure 1700 psig Design Temperature 300 F Shutoff head 1559 psig Design flow / head 425 gpm O 2500 ft/1150 psid Runout flow / head - 650 gpm @ 1500 ft/675 psid SIP operating The diagram below illustrates the flow characteristics of the SIPS.
| |
| characteristics Obj 12
| |
| )
| |
| I (1 - -",
| |
| !<Q-2- -"
| |
| ~*
| |
| j I N- s j -
| |
| -m [
| |
| _a
| |
| _ a _ ,o i
| |
| . . . . , i ico ao no ao soo no >
| |
| Pass en GeNons Per Wowta h
| |
| ECC-08 "
| |
| Continued on ncn page i
| |
| i B3. DOC 2 - 14 REV 6
| |
| | |
| Major Ccmponrnts Low Head injection System, continued I
| |
| \
| |
| l RHR pump The operating characteristics for the RHR pumps are shown below.
| |
| cperating characteristics Ob]12 e% -a p ,- N -=
| |
| 1
| |
| -ng
| |
| . m- -.]
| |
| l
| |
| [nm -a, !
| |
| I R-
| |
| ]
| |
| -= [
| |
| E~
| |
| }"
| |
| a a, . a~~~ a &
| |
| ECC-14 The table below lists *he design flow characteristics of the RHR pumps.
| |
| )
| |
| i Characteristic Flow rate Pressure Shutoff head 0 gpm 173 psid Design flow rate 3000 gpm 160 psid Runout flow 4500 gpm 130 psid B3. DOC 2 - 35 REV 6
| |
| | |
| 5 system Operations Emergency Operations Alignments - . A Spray actuation signal of either a HI HI containment pressure signal OR Spray a MANUAL spray actuation will cause:
| |
| Actuation + CS-8994A and B, spray add tank to eductors to open.
| |
| Obf4,18 + CS-8992, Spray add tank outlet (normally open) to receive an open signal.
| |
| . A Spray actuation signal WITH a Safety Injection Actuation will cause:
| |
| + Spray pumps to start and to pump flow through the eductors,
| |
| + The eductors to draw fluid from the spmy add tank and mix it in the Pumps.
| |
| . Approximately 15 gpm will flow through each spray nozzle in containment (a:2466 gpm per train).
| |
| . This diagram shows the spray system in operation.
| |
| o.: us ac: =c E A n!: Ja omna :
| |
| must j sa==5 3una3 -
| |
| g;;;;;;7,%,
| |
| - (a m. .J l
| |
| ~4 e e a
| |
| ::(Q
| |
| : , 5-
| |
| $ N 3 g mas 2
| |
| ; O CSS-IS Equipment that change status on Spray Action Actuation..*
| |
| CS-8994A and B. Spray add tank to eductors Open NOTE:
| |
| * CS-8992. Spray add tank outlet also receives an open signal. however it is normally open.
| |
| _ Equipment that change status on Spray Actuatien Action WITH a Safety Injection Actuation signal Containment Spray pumps Start CS-9001 A and B. Spray add pumps to Containment Open Continued on nen page l
| |
| ) m n^o 39 REV 6 1
| |
| | |
| i i
| |
| 1 JOB PERFORMANCE MEASURE TEST ITEM !
| |
| l Task
| |
| | |
| ==Title:==
| |
| Transfer to CL Recirculation !
| |
| K/A
| |
| | |
| ==Reference:==
| |
| 005000A2.02 (3.7)
| |
| Reference Allowed: YES Refeience: STG B1 A, Obj.16, pp. 2.1-25,2.1-26, & 2.1-27 .
| |
| l QUESTION # 1b:
| |
| Current plant conditions: ,
| |
| . Mode 5
| |
| . Solid Plant Conditions
| |
| . RCS temperature 168 F I
| |
| . RCS pressure 200 psig
| |
| . 1-1 RHR system in operation j
| |
| . 1-2 RHR system in standby lined up to the RCS j The air supply line to CVCS-1-PCV-135, Letdown Press Control Valve, is broken off by a maintenance worker.
| |
| : 1. Predict the RCS pressure response.
| |
| : 2. Give your reasons for your response.
| |
| ANSWER:
| |
| : 1. RCS pressure will decrease [0.5 pts]
| |
| : 2. PCV-136 controls RCS pressure in a solid plant lineup and fails open on loss of air.
| |
| As PCV-135 opens, letdown flow from the RHR system will increase, causing RCS inventory and pressure to decrease. [0.5 pts)
| |
| NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT f -
| |
| I 19'h R % Q . -
| |
| i4 f7f;N F3M TE%h M ya P
| |
| %,e if EFESin V pg
| |
| #ey h $f l Th - @M hEF !69%\ $ " f'T If hi k kTN ti 4k [#bba $ Nb N $\bms b
| |
| | |
| Major Componems . Letdown system l
| |
| l
| |
| ; Letdown Pressure Control Valve, PCV-135 1
| |
| l i
| |
| l Purpose The purpose of PC%135 is to:
| |
| Obj7 . maintain backpressure downstream of the letdown orifices to prevent flashing and cavitation from occurring in the orifices or letdown hea.
| |
| exchanger,and
| |
| . control RCS pressure when the plant is in a solid water condition and RHR is supplying letdown. .
| |
| i I
| |
| Location PCV-135 is located in the Letdown heat exchanger room, 85' elevation of Obj8 the Auxiliary building.
| |
| Manual bypass PCV-135 has a manual bypas.; valve, CVCS-8409, used to control letdown Obj7 pressure when PCV-135 is out of service.
| |
| Diagram The diagram below illustrates the PCV-135 in the letdown system.
| |
| PT.135 l g.................,
| |
| PCV.135 Letdown ; :>< >c :
| |
| Letdown Heat Exchanger x 8409 CVC-16 Physical The characteristics of PCV-135 are:
| |
| description . air-to-close e fail open
| |
| . globe valve Continued on next page l
| |
| B1A. DOC 2.1 - 25 REV 6
| |
| | |
| E 2 j I
| |
| I Chemicci and Volume Control System Letdown Pressure Control Valve, PCV-135, continued Controls Hand Controller HC-135A is located on VB-2.
| |
| Ob]17 Control Operation I2tdown Pressure Controller HC-135A Auto / Manual station l
| |
| Letdown pressure is controlled by throttling PCV-135. It is controlled with {
| |
| its manual / automatic hand station (HC-135A). The table below lists the functions of the station.
| |
| Part Function Output meter Indicates the magnitude of the control signal that is l being sent to PCV-135. It is a demand signal.
| |
| (0% = closed,50% = half open.100% = full open)
| |
| Potentiometer Provided to adjust the Reference Ietdown pressure when the controller is in automatic. One full turn equals 60 psi on a 0-600 psig ten-turn scale.
| |
| AUTO Used to select the automatic output of the controller as j Pushbutton the control signal for PCV-135. PT-135 is used as the I setpoint signal. 4 MANUAL Used to disconnect the automatic output of the Pushbutton controller from PCV-135, and substitute a manual control signal.
| |
| INCREASE / Used to vary the manual control signal during manual DECREASE operation.
| |
| Pushbuttons Continued on nextpage 1
| |
| B1A. DOC 2.1 - 26 REV 6 l I
| |
| I
| |
| | |
| l l
| |
| Major Componems . Leidown System Letdown Pressure ControlValve, PCV-135, Continued Logic The logic associated with HC-135A Auto / Manual operation is described in Obf10 the following table.
| |
| If the Auto / Manual Then...
| |
| controller ...
| |
| Potentiometer is turned clockwise The Reference I.etdown pressure setpoint is increased.
| |
| Potentiometer is turned counter- The Reference Letdown pressure 1
| |
| ' clockmre setpoint is decreased. I AUTO ytshbutton is pushed The automatic output is selected as the control signal to PCV-135. j MAN pushbutton The automatic output of the i l
| |
| controller is disconnected, and a mann=1 control signal coming directly from the hand controller is substituted.
| |
| If the MAN pushbutton is pushed Then...
| |
| . and...
| |
| INCREASE pushbutton is pushed The valve is opened.
| |
| (Pressure 4 on PI-135)
| |
| DECREASE pushbutton is pushed The valve is closed.
| |
| (Pressure i on PI-135)
| |
| Indications The following indications are available for PCV-135.
| |
| Ob]11 Auto / Manual switch Indication Meaning Normal Status Potentiometer A 10-turn dial (60 psi per turn). A 5.83 turns setting of 0 turns corresponds to a (350 psig) reference pressure of 0 psig; a setting of 10 turns corresponds to a reference pressure of 600 psig.
| |
| Output meter Indicates the magnitude of the control 50-60 %
| |
| signal that is actually being applied to the pressure control components.
| |
| BIA. DOC 2.1 - 27 REV 6
| |
| | |
| r: a l
| |
| JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Manually initiate Containment Spray KIA
| |
| | |
| ==Reference:==
| |
| 026000G2.1.33 (4.0)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| STP l-1D; T/S 3.6.2.2; and PSRC Interpretation 87-04; STG 12, Obj. 22, Pg. 4-1 QUESTION # 2a:
| |
| Unit 1 is currently in Mode 4 with preparations in progress to enter Mode 3.
| |
| While performing STP l-1D, MODES 1,2, and 3 Monthly Checklist, valve CS-1-8992, Spray Additive Tank to Eductor, is found closed.
| |
| Attempts to open the valve have failed.
| |
| What, if any Tech. Spec. actions apply?
| |
| ANSWER:
| |
| Enter T/S 3.0.3.
| |
| Candidate's Resnon ef; SAT UNSAT l
| |
| l t
| |
| t
| |
| ? 5 ' !aw A% k: -" 1 k 's N "
| |
| Lh
| |
| * wo uj v S h5 ab %
| |
| as s s ums a l l
| |
| | |
| 06/07/98 Page 3 of 25 STP 1-1D (UNIT 1)
| |
| ^ {
| |
| \
| |
| ATTACHMENT 11.1 TITLE: MODES 1,2. AND 3 MONTHLY CHECKLIST APPL TECH SPEC PERP MODE REFERENCE ' A. CHECK / VERIFICATION FROM CONTROL ROOMI ELECTRICAL ROOMS 1,2,3 4.7.3.1.a ACChriABLE (Cont'd) yJS F_Q
| |
| : e. Valve FCV-364 CCW from residual HX-2 OPERABLE or open. [] []
| |
| : f. Valve FCV-365 CCW from residual HX-1 OPERABLE or open. [] []
| |
| : g. Verify valve CCW-1-LCV-69, surge tank makeup water, OPERABLE. [] []
| |
| : h. Verify valve CC :-t-LCV-70, surge tank makeup water, OPERABLE. [] []
| |
| NOTE: Verify CCW-1-LCV-69 and CCW-1-LCV-70 OPERABILITY by verifying 1 STP V-3H11 has been performed in the l last three months. j 1.2.3 4.6.2.1.a 4. Verify the status of the following contamment spray system valves.
| |
| ACCEPTABLE
| |
| : a. Valve 9001A Cont. Spray Pump 1 discharge ;
| |
| OPERABLE and is closed. [] [] l
| |
| : b. Valve 9001B Cont. Spray Pump 2 discharge OPERABLE and is closed. [] []
| |
| : c. Valve 9003A RHR to Cont. Spray Train 1 OPERABLE and is closed. [] []
| |
| : d. Valve 9003B RHR to Cont. Spray Train 2 OPERABLE and is closed. [] []
| |
| 1,2,3 4.6.2.2.a 5. Verify the status of the following spray 1 additive system valves. I ACCEPTABLE l Y.Es EQ
| |
| : a. Valve 8994A spray additive OPERABLE and is closed. [] []
| |
| : b. Valve 8994B spray additive OPERABLE and is closed. [] []
| |
| : c. Valve 8992 spray additive tank outlet is open. [] []
| |
| l 00319754. doz 06 5 1 0611.0635
| |
| | |
| o SEE PSRC INTERPRETATION 87-04 CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION l
| |
| 3.6.2.2 .The Spray Additive System shall be OPERABLE with:
| |
| : a. A spray additive tank with a contained volume of between 2025 and 4000 gallons of between 30 and 32% by weight NaOH solution, and l b. Two spray additive eductors each capable of adding Na0H solution from the chemical additive tank to a containment Spray System pump flow.
| |
| 6PPLICABILITY: MODES 1, 2, 3, and 4.
| |
| ACTION:
| |
| With the S ray Additive System inoperable, restore the system to OPERABLE status within 72 ours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.
| |
| SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE: l l a.
| |
| At least once peror31automatic power-operated, days by)inverifying the flowthat patheach valve that is (manual, not locked, sealed, or otherwise secured in position, is in its correct position;
| |
| : b. At least once per 6 months by:
| |
| l 1) Verifying the contained solution volume in the tank, and
| |
| : 2) Verifying the concentration of the NaOH solution by chemical analysis.
| |
| : c. At least once each REFUELING INTERVAL by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray actuation test signal; and
| |
| ! d. At least once per 5 years by verifying both spray additive and RWST full flow from the test valve 8993 in'the Spray Additive System.
| |
| DIABLO CANYON - UNITS 1 0 2 3/4 6-12 Unit 1 - Amendment No. 118 Unit 2 - Amendment No. 116 32972804.4A III TAB 13 11 L
| |
| | |
| 1 10/6/94 Paga 1 of y' 3 PACIFIC GAS AND ELECTRIC COMPANY NUC12AR POWER GENERATION IDAP II3.ID3 ATTACHMENT 6.1 TITLE: REQUEST FOR TECHNICAL SPECIFICATION INTERPRETATION I REQUEST i
| |
| PSRC INTERPRETATION N NasR: 87-04 REVISION: 1 TECHNICAL SPECIFICATION: 3.5.2. 3.G.2.1. 3.G.2.2. 3.7.1.2 QUESTION / CONCERN: To anmure the comrability of safetv-related systems. ;
| |
| valves in the flow nath =a-t be either correctiv nositioned. or canable of actuatina to the correct nosition when reauired. A - 11 n**=har of these valves. if incorrectiv nositianad. can ==ke both trains in a safetv-related system inanarable. Guidance was reauested to identify all indiviM=1 valves ehnt. if mi-namitinn=d. would r-- .:'=r a safetv-related system 4nanarable and violate a Technfem1 sneelfication Y.4=4 tina condition for coeration.
| |
| SUBMITTED B .
| |
| DATE: % h REVIEWED BY SE I DI e _ ,
| |
| DATE:
| |
| REVIEWED BY RE TORY (# 4 ULNCE: - DATE: 7, Is 7
| |
| II PSRC INTERPRETATION Placina any one of the followina valves in the incorrect nosition. durine an ooeratine mode in which the associated system is reouired to be ooerable. would result in an aman =ntion in the accident an=1vais not beina met and cause both tra4n= of the safetv-related system to be 4 . -=rable .
| |
| This action is a violation of the Technical Snecification f.4=4 tina {
| |
| Condition for coeration and reouires entry into Technical Snecification l 3.0.3.* Entry into Teehniem1 Snecification 3.0. 3 is rer.,eri.able i= '=r 10CTR50.73 f a) (2) (i) (B)
| |
| *Except for T.S. 3.7.1.2 PSRC APPROVED YES: NO: DATE: b APPROVED: _ % DATE:.2 / [
| |
| PLAlff MANAGER '
| |
| EFFECTIVE DATE: h DISTRIBITTION REQUIRED YES: NO:
| |
| ('
| |
| | |
| l t
| |
| PhM L& I%
| |
| l W/.
| |
| Valve Number Y.glve Function Reauired Valve Modes Technical Position Specification 8105 CCP 1 and 2 Recirc Open 1,2,3 3.5.2 Line Isolation 8106 CCP 1 and 2 Recire Open 1,2,3 3.5.2 .c Line Isolation
| |
| * 8703* RHR to RCS Hot Closed 1,2,3 3.5.2 ..
| |
| Legs 8802A* SI to RCS Hot Legs Closed 1,2,3 3.5.2 88028* SI to RCS Hot Lega Closed 1,2,3 3.5.2 3
| |
| l 8809A* RER to RCS Cold Open 1,2,3 3.5.2 Legs )
| |
| 8809B* RHR to RCS Cold j Open 1,2,3 3.5.2 '
| |
| Legs 8835* SI to RCS Cold .Open 1,2,3 3.5.2 Legs 8974A* SI Pump Recire to Open 1,2,3 3.5.2 .
| |
| RWST (
| |
| 8974B* SI Pump Recirc to Open 1,2,3 3.5.2 l RWST 8976* RWST to SI Pumps Open 1,2,3 3.5.2
| |
| {
| |
| 8980* RWST to RHR Pumps Open 1,2,3 3.5.2 l
| |
| 8982A* Containment Sump Closed 1,2,3 3.5.2 i to RHR .
| |
| 8982B* Containment Sump Closed 1,2,3 3.5.2 to RHR )
| |
| l 8992* Spray Additive Open 1,2,3,4 3.5.2, j Tank to Eductor 3.6.2.2 1 8716A RHR Crossover Line Open 1,2,3 3.5.2 '
| |
| 8716B RHR Crossover Line Open 1,2,3 3.5.2 9003A RHR to Containment Closed 1,2,3 3.5.2 Spray 9003B RHR to Containment Closed 1,2,3 3.5.2 Spray 8804A RHR to Charging Closed 1,2,3 3.5.2 Pump 8804B RHR to SI Pump Closed 1,2,3 3.5.2 8741 RHR to RWST Manual Closed 1,2,3 3.5.2 Valve SI 1 RWST to ECCS Open 1,2,3,4 3.5.2, Manual valve 3.6.2.1 MU-671 CST to AFW Pumps Open 1,2,3 3.7.1.2 Manual Valve (3.0.3 is not applicable)
| |
| *These ECCS ve.lves are identified in Surveillance Requirement 4.5.2. I
| |
| | |
| Section 4.0 Integrated Operations Limiting Specifications Tech Specs The following is a listing of Tech Specs identified for the Containment i Obj 2, 21, 22 Spray system. Please refer to a copy of DCPP Technical Specifications for I the details of the LCO, applicability, action (s), surveillance (s) and basis for each.
| |
| i T/S Item 3/4.3.2 ESF Actuation System Instrumentation 3/4.6.2.1. Containment Spray system 3/4.6.2.2 Spray Additive system 3/4.6.3 Containment Isolation Valves PSRC What valves in the Containment Spray system, if not correctly positioned Interpretation would render the Containment Spray system inoperable?
| |
| Valve No. Required position CS-8992, Spray add open tank to eductor CS-9003A, RHR to closed Containment Sp ay CS-9003B, RHR to closed Containment Spray Conunued on next page
| |
| : 12. DOC 4-1 REV 7
| |
| | |
| p l
| |
| JOB PERFORMANCE MEASURE TEST ITEM l
| |
| Task
| |
| | |
| ==Title:==
| |
| Manually initiate Containment Spray K/A
| |
| | |
| ==Reference:==
| |
| 026000K3.01 (4.1)
| |
| Reference Allowed: NO
| |
| | |
| ==Reference:==
| |
| STG 12, Obj. 6, p.1-6 j QUESTION tt 2b:
| |
| i I
| |
| Unit 2 experienced a DBA LOCA and Containment Spray has actuated.
| |
| Containment Spray Pump 2-1 tripped on overcurrent. i
| |
| : 1. What is the minimum assumed equipment available to protect the Containment function under these circumstances?
| |
| : 2. What is the Basis for this? ,
| |
| i ANSWER:
| |
| : 1. Per accident analysis, one train of Containment Spray and two.of five Containment Fan Cooler Units are required. [0.5)
| |
| : 2. To maintain the Containment pressure below its design value of 47 psig. [0.5]
| |
| NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT
| |
| % I ,4 F%g $ 7 I[IiCU EQ (f jif (N Th
| |
| )k a , Ej jV h s ]! bN 44f,8@rmNIY[
| |
| 7 M j $r d"$1i! i% Y id' ba 3 h df h ba d !
| |
| l
| |
| | |
| CS System l
| |
| .c Design Information Design basis The design bases are tabulated below.
| |
| Obj 5,6 The Contairanent Spray system is designed Reason to...
| |
| Operate in conjunction with the containment The spray is used to help condense the steam fan cooler units (CFCUs). resulting from a LOCA or MSLB.
| |
| 1 One train of Containment Spray and two of. The CFCUs further reduce the temperature of five CFCUs provide sufficient heat removal to the air and water vapor mixture.
| |
| maintain containment pressure below its design value of 47 psig following a design basis LOCA or MSLB.
| |
| Provide post-accident containment heat This design accommodates all postulated removal, and is therefore designated Design events.
| |
| Class I.
| |
| Provide a time delay on starting CSS pumps: . The spray piping needs to remain dry until
| |
| . A minimum time delay of 23 seconds after the maximum seismic loading has j subsided after an earthquake,
| |
| . A maximum delay of 41 seconds . Cooling is needed in the containment atmosphere in a LOCA.
| |
| A Containment Spray piping fill time of 39 to 44 seconds is required for flow to reach the The piping in containment is normally dry.
| |
| spray nozzles.
| |
| Continued on nen page
| |
| : 12. DOC 1-6 REV 6
| |
| | |
| r j
| |
| 1 JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Manual heat balance calculation K/A
| |
| | |
| ==Reference:==
| |
| 015000KS.04 (3.1)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| LSTP-R2, Pg.18, STP R2B QUESTION # 3a:
| |
| Explain the effect on reactor power calorimetric as the throat of the main feedwater flow venturies erode.
| |
| l ANSWER:
| |
| The calorimetric calculation would indicate a lower than actual power level.
| |
| NOTE: (Since power is directly proportional to flow and measured flow is directly proportional to delta-P, venturi throat errosion will cause the delta-P to decrease therefore measured flow decreases resulting in a lower calculated power.)
| |
| Candidate's Response: SAT UNSAT 1
| |
| i .!) h~ ;h h l
| |
| * N. i i% 2 %m %/
| |
| i, kegh{ # t 4d> g&amd; $n %*Dg anni n w- ~ :
| |
| d, $]1 i i. s3 EEm2 ad
| |
| | |
| i I
| |
| LESSON: HEAT BALANCE LESSON Ne - }STP-R2 Inaccuracies and Their Effects l S/G blowdown Failure to subtract the S/G blowdown flow will cause:
| |
| CALCUu nD to be higher than D ACTUAL
| |
| + The equation term s assumes that steam flow equals feed flow, so s would appear to be higher than it actually is.
| |
| . The NIS would then appear to be indicating low.
| |
| . Adjusting the NI to agree with the D CALCUunD would cause:
| |
| + Core PowerACmAt to be less than Core PowertNDICAnD Venturi Feed flow venturi flow blockage due to some corrosion product deposition I corrosion will cause: i blockage
| |
| , g CALCUunD to be higher than ACTUA1.
| |
| + The venturi AP would be higher indicating a higher s .
| |
| . The NIS would then appear to be indicating low.
| |
| . Adjusting the NI to agree with the D CALCUunD would cause:
| |
| + Core PowerACTUAL to be less than Core PowerisoiCAno Venturi erosion Feed flow venturi throat opening being badly worn (eroded) will cause:
| |
| DCALCUuTED to be lower thanACTUA1.
| |
| + The venturi AP would be lower indicating a lower s .
| |
| . The NIS would then appear to be indicating Ingh.
| |
| . Adjusting the NI to agree with the CALCUunD would cause:
| |
| + Core PowerACTUALt o be greater than Core PowerisoiCAnD l l
| |
| Letdown flow Letdown flow is diverted. l
| |
| . This condition only affects the primary system heat balance performed m !
| |
| STP R-2C.
| |
| i t
| |
| LSTPR2. Doc PAGE 18 OF 22 REv3
| |
| | |
| i i
| |
| JOB PERFORMANCE MEASURE TEST ITEM l Task
| |
| | |
| ==Title:==
| |
| Manual heat balance calculation !
| |
| l l- K/A
| |
| | |
| ==Reference:==
| |
| 015000K1.01 (4.2)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| STG B4, Obj. 50, Pg. 3-20, OlM B-4-2 QUESTION # 3b:
| |
| Unit 1 is operating at 50% power due to a curtailment to clean Circulating Water Tunnels. A Control Room / Electrical technician is performing STP l-2C1, " Removal of Power Range Channel from Service" for NI-41. During performance of the STP the j technician mistakenly pulls the instrument fuses for N1-42, realizing his mistake he reinserts the fuses for NI-42 and immediately pulls the instrument fuses for NI-41. A Reactor Trip occurs immediately after the fuses for NI-41 were removed.
| |
| Explain what caused the Reactor Trip?
| |
| ANSWER:
| |
| Pulling the instrument fuses for NI-42 generated a power range negative rate trip which is not automatically reset when power is restored. This action makes the power range negative rate trip a 1/3 coincidence. Pulling the NI 41 instrument fuses satisfied the coincidence for the power range negative rate trip.
| |
| Candidate's Response: SAT UNSAT L ANSWER KEY:
| |
| | |
| f'L i
| |
| Excore Nuclear Instrumentation System ,
| |
| Abnormal Operations, continued PR channel The following table describes some possible failures of PR equipment and failures the effects on plant / system operation.
| |
| Obj50- ,
| |
| If Then PR Chanocl no trip under any alignment, however, if this occurs fails low during reactor operation, Tech Spec actions will apply, a Channel Deviation alarm and a Negative Rate trip will be generated from the failed channel, however, no reactor trip will occur due to the 2 of 4 logic required for the trip.
| |
| . Operators should reset the Rate trip at the NI drawer to return the logic to a 2 of 3 coincidence, otherwise it {
| |
| will be a less desirable 1 of 3 logic. I the permissives P-8, P-9 and P-10 will still function, but ,
| |
| will work on a 2 of 3 remaining channels instead of the
| |
| ]
| |
| original 2 of 4 design logic. )
| |
| PR Channel a High Power and Positive Rate (depending on rate of fails high failure) trips un be generated from the failed channel, however, no reactor trip will occur due to the 2 of 4 logic required for the trips.
| |
| . Operators should reset the Rate trip at the NI drawer to return the logic to a 2 of 3 coincidence, otherwise it will be a less desirable 1 of 3 logic.
| |
| Other effects are:
| |
| . Channel Deviation alarm.
| |
| . C-2, PR rod stop, only 1 of 4 channels required.
| |
| . Input to Rod Control System will generate an automatic rod insertion signal.
| |
| the permissives P-8, P-9 and P-10 will still function, but will work on a 1 of 3 remaining channels instead of the original 2 of 4 design logic.
| |
| Continued on nextpage B4. DOC 3 - 20 REV 7
| |
| | |
| lIliil
| |
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| |
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| |
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| |
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| |
| ip w r dy r e ip a eB mp w t r
| |
| e UI ri e r o p
| |
| r e o g tT n l
| |
| n inl t w n n
| |
| c I it t .
| |
| dv e e o
| |
| p e u ak GC ne aL" m l o
| |
| h t
| |
| h u _
| |
| r d N ef c
| |
| r o t
| |
| r s
| |
| n t
| |
| n o
| |
| c k
| |
| e c
| |
| e r V un oo si t i
| |
| f o
| |
| f o
| |
| lo b -
| |
| s s e o
| |
| i ns s s r oo o o a sp c ce L L s x
| |
| t l t
| |
| n eh a f t n e
| |
| E S mr ue f
| |
| en r p o is g
| |
| / r e u ip -
| |
| B t w wt _
| |
| so o t r _
| |
| InP pan
| |
| > e t
| |
| fd h _
| |
| n on t
| |
| l e se n lat mr ue s sp oe h e
| |
| nu e igp t
| |
| r ws s or i Ld W SI n i nPF mA0 2% G
| |
| | |
| i L JOB PERFORMANCE MEASURE TEST ITEM l
| |
| Task
| |
| | |
| ==Title:==
| |
| Respond to high Accumulator level KIA
| |
| | |
| ==Reference:==
| |
| 006000A2.03 (3.7)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| DCPP Volume 9B, Pg.1.E. 9.1.b; LPA-1, Obj. 2, Pg. 8 QUESTION # 4a:
| |
| The following conditions exist:
| |
| . ECCS Accumulator 1-1 level 66% at 1730
| |
| . ECCS Accumulator 1-1 level 73% at 1830 Determine the leak rate from the RCS into the 1-1 Accumulator.
| |
| ANSWER:
| |
| * ACUMM 1-1 @73% = 6527 gal. j e ACUMM 1-1 @66% = 6397 gal. !
| |
| * Level difference = 130 gal.
| |
| 1 130 gal. / 60 min. = 2.2 gal / min (i 0.2 gal / min)
| |
| Candidate's Response: SAT UNSAT l ANSWER KEY'
| |
| | |
| F V
| |
| Diablo Canyon Power Plant Unit No(s) 1 And 2 ,I t ,, v. .
| |
| | |
| ==Title:==
| |
| Scaling Calculation for Accumulator Volume e .. -
| |
| Acoumulators Volume vs.' Percent inaicated
| |
| % Gallons % Gallons % Gallons . % Gallons 0 5177 25 5639 50 6101 75 6564 1 5195 26 5657 51 6120 ' 76 6582 2 5213 27 5676 52 6138 ! 77 6601 3 5232 28 5694' 53 6157 ! 78 6619 4 5250 29 5713 54 6175l 79 6638 5 5269 30 5731- 55 6194 l 80 6656-6 5287 31 5750 56 81 6675 6212l 6693 7 5306 32 '5768 57 6231 1 82 8 5324 33 5787 58 6249 83 6712 9 5343 34 5805 59 6268 84 6730 10 5361 35 5824 60 6286 85 '6749 11 5380 36 5842 51 6305 , 86 6767 12 5398 37 5861 62 6323 87 6786 13 5417 38 5879 63 6342 ! 88 6804 14 5435 39 5898~ 64 6360 ! 89 6823 15 5454 40 5916 65 6379 ' '
| |
| 90 6841 16 5472 41 5935 66 6397 '' 91 6860 17 5491 42 5953 67 6416 92 6878
| |
| : 18. 5509 43 5972 68 6434 93 6897 19 5528 44 6990 69 6453 94 6915 20 5546 45 6009, 70 6471 95 6933
| |
| . 21 5565 46 .6027 71 6490 96. 6952 22 5583 47 .6046 72 .fB508 ; 97. 6970 23 5602 48 $6064, 73 6527 j 98 6989 24 5620 49 6083 74 6545 i 99 7007
| |
| * ,. .100 7026 q
| |
| i Total internal Volume 10099 gallons
| |
| , Unmeasured Volum6 l5177 gallons Undrainable Volume (Top of Pump Suction)c N/A
| |
| .;L Maximum Contained Volume (To Bottom of Overflow) N/A 8
| |
| 4 5 $
| |
| ns< stwte
| |
| ~
| |
| a o !
| |
| .p t 9.t.b- Rey # :
| |
| D & shs/* l
| |
| < v. . T . . e.
| |
| l
| |
| | |
| LESSON: EXCESSIVE REACTOR COOLANT SYSTEM LEAKAGE LESSON No.: LPA-1 RCS Leakage Paths, continued ECCS . Symptoms of RCS boundary valve leakage into the ECCS are shown Obj 2, 3 below.
| |
| . Leakage into the ECCS via check valve back leakage will most likely be so minor that OP AP-1 will not be entered.
| |
| + Normally discovered while performing STP V-SA1 & 2, ECCS Check Valve Leakage.
| |
| When the RCS leaks into and the following alarms / indications the ECCS, the following may exist ... I symptoms may exist ... l Increasing pipe pres.sure . PI-919 & 923, SI Sys Pump Press
| |
| . Piping reliefs to PRT show leakage.
| |
| Increasing SI accumulator . PI-960 to 967, Accumulator Press I pressure and level . LI-950 to 957, Accumulator Level
| |
| . PK02 05, ACCUM PRESSURE HI-LO
| |
| . PK02-10. ACCUM LEVEL HI-LO RCPs . Seal packages have designed leakage rates that are addressed in STG A-6.
| |
| Obj 2, 3 . With excessive leakage, the RCP seal packages have the following leak symptoms.
| |
| + # 1 seal return goes to the charging pump suction.
| |
| + # 2 and # 3 seal leakoffs go to the RCDT and the Containment sump.
| |
| When the RCPs leak, the~ and the following alarms / indications following symptoms may may exist ... ;
| |
| exist ...
| |
| Irregular RCP seal ieakoff . FR-156 to 159, RCP Seal # 1 RTN Flew return flow . PK05-01 to 04, RCP NO. I to 4 Abnormal seal injection FI-144,143,115,116; RCP Seal # 1 inj flow flow Coatinued on next page
| |
| ~
| |
| LPA1. doc PAGE 8 oF 23 REV6
| |
| | |
| f 1
| |
| l j JOB PERFORMANCE MEASURE TEST ITEM l
| |
| Task
| |
| | |
| ==Title:==
| |
| Respond to high Accumulator level KIA
| |
| | |
| ==Reference:==
| |
| 006000K3.01 (4.2)
| |
| Reference Allowed: No
| |
| | |
| ==Reference:==
| |
| EOP E-1.2, " Post LOCA Cooldown and Depressurization"; LPE-1B, Obj.
| |
| 13,pg.
| |
| QUESTION #4b:
| |
| An earthquake caused a small break LOCA to occur on Unit 1 along with the loss of 4kV bus G. EOP E-0 and E-1 have been completed. A plant cooldown and depressurization is in progress IAW EOP E-1.2, " Post LOCA Cooldown and Depressurization."
| |
| Describe what action (s) will need to be taken, regarding the ECCS Accumulators, in order to continue with the RCS cooldown and depressurization if 4kV bus G cannot be restored.
| |
| I ANSWER: -
| |
| Associated Accumulators will have to be vented to containment since they cannot be isolated from the RCS as the cooldown and derassurization continue.
| |
| Candidate's Response: SAT UNSAT 1
| |
| l l
| |
| J'%111%d. v f
| |
| e 6 il u a: %A B!
| |
| | |
| -u PACIFIC-GAS.Ne St.EGinn 6carans 1
| |
| 1 6 ;REVIS!0NJ10 DIABLO' CANYON P0hidt PLANT 24 0F 38 SE E e s . . .
| |
| , PAGE s.'! / ku "i .
| |
| TITLE: POST LOCA C00%IDEPRES$1RIZATIII RESPONSE NOT 08TAINED
| |
| ' ACTION / EXPECTED RESPONSE- 1
| |
| . 25. CHECK If Accumulators should j_
| |
| y .
| |
| Be Isolated: ,
| |
| * a. RCS Subcooling based on a. E , RCS WR Pressure core exit T/Cs - GREATER .'" GREATER'THAN l 175:PSIG, THAN 20*F (5ubcooled Margin Monitor YI-31 or THEN- GO T0; Step 26 (Next '
| |
| Appendix,C) Paae).
| |
| E RCS WR, Pressure LESS j THAN 175 PSIG THEN GO TO Step 25b.
| |
| i
| |
| .............................. )
| |
| 4 .. m PZR Level-GREATER- - - - b. RETURN TO Step'13 -
| |
| - b. ~Verify THAN 12%-[36%] ~ ~ ~ ~ ~ ~
| |
| (Page 10)'..: , 3. - !
| |
| ,. ~ o I
| |
| : c. Check Power to Accumulator c.Restoregowerto A B
| |
| ~
| |
| b i 8808A 52-1F-46 AND 52.lF-46R 8808B 52-lG-07e E 52-lG-07R i
| |
| 8808C 52-lH-14 AND 52-lH-14R
| |
| ,, 8808D 52-10-05 E 52-lG-05R
| |
| : d. Close 8808 A, B, C and D ~ d. E Any Accum Outlet Isol Viv,'does NOT close, THEN Perform the following:
| |
| o, i
| |
| : 1) Close 8880,1v .Accum N2 l Isol Fill (
| |
| V
| |
| '~ ^* 2) Open 8875 A, C or D <
| |
| f' Accum Vent an ,N2 , Fill, I to affected Accum.
| |
| t L. - . . . . - 3) Open HCV-943,-- Accum-N2 '-
| |
| Vent Control Valve, WHEN iAccum is idepressurized, THEN jCloseHCV-943.
| |
| : 4) Close 8875 A, B, C or D.
| |
| o
| |
| | |
| f' LESSON: ECCS Flow REDUCTioM LESSON No.: LPE-1 B ,
| |
| E-1.2, Post LOCA Cooldown and Depressurization, continued E-1.1 Actions Review EOP E-1.1 with the students and emphasize the following items.
| |
| j & Step Basis . Action / Expected Response and Response Not Obtained basis Obj l3,40 . Procedure transitions l
| |
| . Other discussion items in the list below l
| |
| STEPS 25 M!cCELLANEOUS STEPS Item Basis and Discussicu
| |
| . SI accumulators are isolated or vented to prevent discharge into the RCS when RCS subcooling criteria is satisfied.
| |
| Isolate . Either the contents are no longer required because of l Accumulators operator control actions, or the contents are l
| |
| ; considered to be discharged, and isolation prevents nitrogen injection into the RCS. l
| |
| + Nitrogen could collect in high places and produce either a "hard" PZR bubble or cause gas binding in l the S/G tubes.
| |
| l . Manufacturers typically recommend that diesels not Stop D/Gs be run extensively unless carrying load.
| |
| . D/Gs should auto-start on SI signal, but will not load if off-site power is available.
| |
| . CCW is used for RCP motor oil cooling as well as thermal barrier cooling.
| |
| Cool RCPs + Seal injection is only for seal cooling.
| |
| + These normal cooling alignments should be established if they are not present.
| |
| Continued on next page I
| |
| l l
| |
| LPE18. Doc PAGE 37 oF 53 REV5
| |
| (
| |
| | |
| JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Crosstie vital bus G and H K/A
| |
| | |
| ==Reference:==
| |
| 062000A2.01 (3.9)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| OlM p. J-1-1; STG J6A, Obj 30, pg. 2-19 and 2-34; STP M-13H, pg. 9,20 QUESTION # Sa:
| |
| Unit 2 is in Mode 5, fuel in the core, with RCS loops filled, following a refueling outage.
| |
| Given the following equipment status:
| |
| . ASW Pump 2-1 Operable, not running
| |
| . ASW Pump 2-2 Running
| |
| . RHR Pump 2-1 Operable, not running
| |
| . RHR Pump 2-2 Running
| |
| . CCW Pump 2-1 Running
| |
| . CCW Pump 2-2 Running
| |
| . CCW Pump 2-3 Operable, not running
| |
| . CCP 2-1 DC Open
| |
| . CCP 2-2 Operable, not running
| |
| . PDP 2-3 Cleared l 1
| |
| The Control Roora/ Electrical Asset Team desires to run STP M-13H on Bus H, causing the diesel to start and its 4KV bus to load onto the diesel automatically.
| |
| What if any contingency actions should the Shift Foreman direct prior to the start of the test based on current plant conditions and equipment lineup?
| |
| ANSWER:
| |
| Shift Foreman should direct the manual restarting of RHR Pp. 2-2 immediately after the bus transfer or select RHR Pp. 2-1 for decay heat removal.
| |
| NOTE: Either answer is acceptable to receive a SATISFACTORY grade.
| |
| j F3 #% "" M !NZ! EPM &
| |
| F , Can%
| |
| [!! didate's RAsconse:
| |
| ~*
| |
| [ SAT bl .., UN.SATN N #[ NE
| |
| [j"yf S V'Q(f J;) & g \ W~' );"$
| |
| !bea I??% h"m~ W e %f t y! % h) Ev Fll & u iu % thna a
| |
| | |
| ({;(
| |
| p23
| |
| - - V: Bx xtnk eBr e a
| |
| s mror y r P 22 mpp 0TunnuAAAACaTh 85' 0730 Tt '
| |
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| |
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| |
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| |
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| |
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| |
| sPP P t BB 2 55 A2 N 22
| |
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| |
| I- J CC dr 2- um 1 2
| |
| x x hay H1y 7g2q dm u x xut B 24 1
| |
| (
| |
| DPp V08T0' BA Au A' ' uA nCkT e st r 0*30 0Tta9 c l D22 2 us l450 0
| |
| ( QBB i DDJMDJD0, 1
| |
| V 1 n
| |
| * sPP P a 8,1,1.7,1,1, t i 1 iw _- _" V2 k
| |
| uCCW d t
| |
| - " Vk k 1 i
| |
| BRRC n V1 2223345 S ,, 5 5 C N22222222 2 2Vk V
| |
| k I 1 '
| |
| /
| |
| V 4 ) )
| |
| 0 3
| |
| k 0
| |
| 0 5
| |
| ' (
| |
| I
| |
| )B)A )
| |
| B(A(B
| |
| &2(2 - -
| |
| 3- H c
| |
| )
| |
| - A222 (2 2 p)0 V 2 s s r G2 - B ( -
| |
| f X
| |
| m
| |
| ' D2 A@-
| |
| (
| |
| (
| |
| I HPp2 pP p 2pP B 48 w i s l k
| |
| n rr I s
| |
| uWP R PW&
| |
| BFSH 1 VC(t Aa l e u i a ( ACRSC B B m-f 2 i n X r v
| |
| r fTM ia 2
| |
| U A
| |
| x u
| |
| e (
| |
| I )2 -
| |
| B212 1
| |
| 2G 2V e
| |
| 1 G
| |
| - - (
| |
| p2 p2 p0 2)
| |
| G P pP 2 -
| |
| P8 v " h G(
| |
| D B
| |
| @ (
| |
| I s
| |
| uWP RPPW!a BSSHCDCi 4
| |
| t I
| |
| ACRCPCV O I[
| |
| )
| |
| (
| |
| n I
| |
| Y 1 I
| |
| ) 1 - F o ( A22 2 V I
| |
| N (G g(,1 3) - ( pp1 p10 D2( A@
| |
| I - I F PP2 P28 -
| |
| i ( s p4 t 2 2 Nc) E uWWPWPla BSFCCt i u
| |
| t s ( I AACCSV s u s X (
| |
| t B r (
| |
| mV m b
| |
| i f k x03 u2 U
| |
| S V U f
| |
| Xf y
| |
| 4 )
| |
| )
| |
| r S k 2
| |
| S m(X ) I
| |
| ) 1 1
| |
| - F t )
| |
| 1
| |
| ) e 2 )
| |
| A1
| |
| (
| |
| 1 pp1 p12 1
| |
| V i
| |
| s m G3)-
| |
| D1 A@
| |
| (
| |
| - )
| |
| I FPP1 P16 sWWPWPa u
| |
| o4 l BSFCCt i D
| |
| l t eY )
| |
| (
| |
| I AACCSv l
| |
| I[
| |
| a )
| |
| I
| |
| )2 - 1
| |
| - 2G -
| |
| c h 2-
| |
| )
| |
| B1
| |
| (
| |
| p1 p2 1- 1 1V p0 i
| |
| t r g.n w ik r
| |
| r r
| |
| a 1 )
| |
| G(
| |
| D B@ - )
| |
| I G P pP 1 sWPRPPWl u
| |
| BSSHCDCi P8 4
| |
| t a
| |
| u,n m r I ACRCPCV c g f
| |
| X x
| |
| e e
| |
| e 5
| |
| s u2 - G ) )
| |
| A1 B)A s )
| |
| l . u ' I
| |
| )
| |
| (A()3 B2(2B - H Ed r a
| |
| 2 2 '_ B "X )
| |
| G1-
| |
| ) . )
| |
| &1 21- -(2 A p1 p- p)0 1
| |
| V h
| |
| c y
| |
| (
| |
| D1 A B@.
| |
| (
| |
| - )
| |
| I (P p P 1p P B H
| |
| sWP uFSHtRC(it PW& Aa l 8
| |
| 4 it "Y )
| |
| (
| |
| I BACRSC V w V S V k
| |
| 4 1
| |
| 1 x x x p r V
| |
| k k
| |
| 5 2Vk 24 1
| |
| - PpPr V08T0*
| |
| BA Au A' u xu uA n t
| |
| Ck7h e 8e s
| |
| 0 2 l
| |
| D1 1 1 at
| |
| * 0,3 0Ta9i 4
| |
| 2 sPP P sVs l450 81 ,1.7,0 0 1 tn 1 lo t 1 0
| |
| 5 7
| |
| 1 l
| |
| uCCW BRRC )WDc
| |
| ) uh B Bsd an a ,
| |
| it DDJyDJD0J, V1 22p33455 1
| |
| ,1. ,
| |
| ,P 2
| |
| " 23 NC N1 1 1 S 1 1 1 1 1 d"
| |
| 1 1 s x x u p pp V 8
| |
| B 2y y3s 1 13 2- PPP 0 BAunxutuAneCk7 l
| |
| E1 1 1 r r 8T0AA *
| |
| * 7 a9 a a e r
| |
| m1- sPP P sDss t t l450 730 071 1 i1 Ttn1 d
| |
| w i id G w t a
| |
| f 1 X
| |
| x u
| |
| I uCCW BRRC )MEm
| |
| ) u BB B2 dd tr nn HCC N 1 }i a 8,1,1 iEE1,NE1,E F.,
| |
| t V1 2223345 1 1 1 1 1
| |
| ,1 ,
| |
| MM A Zk %0 1
| |
| b%4
| |
| | |
| [l Major Components i
| |
| Vital 4 kV Buses F, G, and H, continued l Bus F, G, and The electrical loads on the Vital 4 kV Buses F, G, and H are shown in the H loads following table.
| |
| Load No. Bus F No. Bus G No. Bus H Auxiliary 11 52-HF-8 12 52-HG-6 1 Saltwater pump (21) (22) l (ASWP)
| |
| Aux Feedwater 13 52-HF-9 ,
| |
| 12 52-HH-8 pump (AFWP) (23) (22) l Centrifugal 11 52-HF-11 12 52-HG-9 i Charging pump (21) (22)
| |
| (CCP)
| |
| Positive 13 52-HG-11 Displacement (23)
| |
| (reciprocating) pump (PDP) or CCP)
| |
| Component 11 52-HF-12 12 52-HG-12 13 52-HH-12 ,
| |
| Cooling Water (21) (22) (23) !
| |
| pump (CCWP)
| |
| Residual Heat 11 52-HG-8 12 52-HH-11 Removal pump -
| |
| (21) (22)
| |
| (RHRP)
| |
| Safety Injection 11 52-HF-15 12 52-HH-15 pump (SIP) (21) (22) 480 V MCC IF 52-HF-10 1G 52-HG-10 1H 52-HH-10 Transformer (2F) (2G) (2H)
| |
| Containment Spray 11 52-HG-7 12 52-HH-9 Pump (CSP) (21) (22)
| |
| Continued on nextpage J6A. DOC 2-19 REV 9
| |
| | |
| '4kV System '
| |
| l l Vital 4 kV Buses F, G, and H, continued L
| |
| l l
| |
| Automatic A automatic transfer to the diesel generators will occur if either of the ;
| |
| i transfer to following conditions occur: )
| |
| Diesel , automatic transfer from auxiliary to startup is required and startup power i Generator log,c i s not available, or
| |
| . the vital 4 kV bus is on startup power and startup power becomes unavailable.
| |
| If one of the above conditions occurs, the sequence for automatic transfer to the diesel generator is as follows:
| |
| . all motor supply breakers on the Vital 4 kV bus are tripped, !
| |
| . the diesel generator starts and comes up to rated speed and voltage, l
| |
| . the auxiliary and startup feeder breakers are tripped and locked out, e after the vital 4 kV bus voltage has been at zero for greater than 2.0 l seconds, the diesel generator feeder breaker closes, e the vital bus loads are sequenced on the bus by the sequencer timers.
| |
| Details of the automatic transfer are described in J-15 Trips The vital 4 kV bus breakers have the following trips:
| |
| . overcurrent !
| |
| . differential overcurrent 86G1 lockout relay (Aux Feeder only) l 86G11 lockout relay (Aux Feeder only)
| |
| {
| |
| The diesel generator feeder breakers have the following trips:
| |
| . diesel generator differential 4 kV bus differential
| |
| . overcurrent*
| |
| . reverse power *
| |
| . engine shutdown
| |
| . diesel generator loss-of-field *
| |
| ' Note: These trips are normally cut-out (non-seismic relays).
| |
| Continued on nextpage l
| |
| i J6A. DOC 2-34 REV 9
| |
| | |
| I-t
| |
| $f m } eqf4Y!. g l PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP M-13H ,
| |
| DIABLO CANYON POWER PLANT REVISION 16 l PAGE 9 OF 69 TITLE: 4KV BUS H NON-SI AUTO-TRANSFER TEST UNITS 1AND2 START DATA SECTION l UNIT OPERATING MODE DATE/ TIME /
| |
| : 10. PRECAUTIGilS AND LIMITATIONS INITIALS 10.1 Various 480V loads connected to Vital Bus H will lose power for a short period of time. Certain pump or compressor motors will )
| |
| not restart on a loss of power. A survey of the ventilation systems, auxiliary control board and radiation sampling monitor should be made after the transfer in order to restart any equipment which is needed for plant operations. STP M-9M contains a convenient check list of vital 480V loads which may i be used if desired.
| |
| 10.2 If RHR Pump 2 is in service to provide decay heat removal, it I will be tripped during the transfer to Diesel as part of the load shed scheme but will not auto-start as it is not part of the auto-transfer scheme. Consideration should be made to either
| |
| . manually restart RHR Pump 2 after the transfer or select RHR Pump 1 for decay heat removal.
| |
| 10.3 Spent Fuel Pit Pump 2 is powered from 480V Bus H. It has a local pushbutton control with seal-in contact. The contactor will drop out on momentary loss of bus voltage which will happen during the performance of this test. Put Spent Fuel Pit Pump 1 in service and secure Spent Fuel Pit Pump 2 or take operator action to restart the pump after each auto-transfer.
| |
| 10.4 When timing the CFCUs, the fan coolers must be running in HIGH speed and aligned to start in HIGH speed to obtain the proper auto-transfer timer measurement in Part 3 or Part 4 of this test, while also ensuring the control circuit starts the CFCU in LOW speed regardless of switch position.
| |
| 10.5 Safety Injection Pump 2 is tested in this procedure for its load shed feature. The pump will not auto-start on auto-transfer to Diesel as it is not part of the auto-transfer scheme.
| |
| 10.6 Containment Spray Pump 2 is tested in this procedure for its load shed feature. The pump will not auto-start on auto-transfer to Diesel as it is not part of the auto-transfer scheme.
| |
| 10.7 This test shall be cord # red a " Partial Test" if any auto-transfer to Diesel load breaker (AFW Pump 2, CCW Pump 3) is placed in TEST for this test. This is because the intent of Tech Spec Diesel Surveillance 4.8.1.1.2.b.5) is not completely satisfied.
| |
| I 00911516. doc 06 9
| |
| | |
| ?Y$$.?'$$$h?hh$l?Yh PACIFIC GAS AND ELECTRIC COMPANY
| |
| ? YY$f f $b NUMBER STP M-13H DIABLO CANYON POWER PLANT REVISION 16 PAGE 20 OF 69 TITLE: 4KV BUS H NON-SI AUTO-TRANSFER TEST UNITS 1 AND 2 PERF 11.16 Residual Heat Removal Pump 1 is available for decay heat removal.
| |
| NOTE: RHR Pump 1 is not associated with Bus H but is required for this test to ensure that in case the load shed relay cannot be reset during the test and the trip sigual to RHR Pump 2 1 f
| |
| is maintained, RHR Pump I can be put in service for decay heat removal. This prerequisite can be waived per SFM's direction ;
| |
| or if the test is performed during FJIR window.
| |
| I1.17 Residual Heat Removal Pump 215. available for operations or the breaker is "available" to be placed in TEST and CLOSED.
| |
| (If the breaker is available in TEST, do not place it in TEST until l; Part 3 of the test.) l
| |
| [ ] RHR Pump 2 is available far operations. !
| |
| I
| |
| [ ] RHR Pump 2 motor breaker is available to be placed in j TEST. I 1
| |
| NO TE 1: This is to test its Ic>ad shed feature. !
| |
| NOTE 2: If RHR Pump 2 is cleared and if a ground buggy is installed in the RHR Pump switchgear for personnel safety ;
| |
| reason, it is recommended t' tat the pump breaker be installed !
| |
| outside the RHR Pump switchgear with the " umbilical cord" j feeding the control signal from the switchgear to the pump i breaker. This is considered an acceptable way to place the j breaker in the TEST posit'.on for the purpose of this test and )
| |
| maintain personnel safety for possible maintenance work at the j pump motor.
| |
| 00911516. DOC 06 20
| |
| | |
| p i
| |
| l JOB PERFORMANCE MEASURE TEST ITEM 1 l
| |
| i Task
| |
| | |
| ==Title:==
| |
| Crosstie vital bus G and H l K/A
| |
| | |
| ==Reference:==
| |
| 062000A1.01 (3.8) {
| |
| l Reference Allowed: YES l
| |
| | |
| ==Reference:==
| |
| EOP ECA-0.3, Appendix Q: Lesson Plan LPECAO, p. 40, Obj.17 j QUESTION # 5b:
| |
| Following a reactor trip from 100% power, off-site power has been lost and only D/G 2-3 is available to supply the Unit 2 4KV busses. In accordance with ECA-0.3 (Restore 4KV Busses) Appendix X (Crosstie of Vital Bus), D/G 2-3 is now supplying Vital Bus F and Vital Bus G.
| |
| Assuming zero overload hours for this year, what would be the maximum amount of time that D/G 2-3 could be operated at 2.96 MWs?
| |
| ANSWER:
| |
| 6 hours i1 hour ;
| |
| 1 Candidate's Response: SAT UNSAT 4
| |
| d are ,:
| |
| a y
| |
| t ma ay?y a u f.
| |
| == v
| |
| ^kk kh,b hem N kla I h bbem b
| |
| | |
| PACIFIC GAS AND ELECTRIC COMPANY NUMBER E0P ECA-0.3 DIABLO CANYON POWER PLANT REVISION 5 ,
| |
| PAGE 26 0F 32 TITLE: RESTORE 4KV BUSES UNIT 2 APPENGIX 0 DIESEL GENERATOR LOAD LIMITS GENERATOR OPERATING HOURS / YEAR 10000 , , ,i l ; N'm !i, 6
| |
| 3 i ,
| |
| ,g
| |
| \ g g
| |
| \ \ l I hg ll h< ,
| |
| 1 1000,
| |
| )\g
| |
| \, 1 a
| |
| (
| |
| g i
| |
| \
| |
| 3 100,
| |
| \g ABNORMALOPERATION[
| |
| 1 OPERATING ENVELOPE \
| |
| \ J l \
| |
| . \,
| |
| 10 , g ;
| |
| i \
| |
| t
| |
| \
| |
| \
| |
| \
| |
| ~
| |
| 1 2500 2600 2700 2800 2900 3000 3100 GENERATOR (KW)
| |
| | |
| r-LESSON: LOSS OF ALL VITAL AC POWER RESPONSE LESSON No.: LPECA-0 ECA-0.3 Appendices j Appendix A . Appendix A provides guidance for loading emergency buses during blackout Objl7 conditions. If power is lo;t to a vital bus and then restored prior to resetting the SI signal the equipment will automatically load on the bus and !
| |
| the equipment should be verified operating. If the SI signal has been reset, the D/G may require manual. loading.
| |
| Appendix II Appendix H provides guidance for using a diesel from one unit, via cross-tie, to energize a bus on the other unit. In the event that the only source of power is a D/G from the non-casualty unit it will be necessary to cross"e units.
| |
| Appendix N Appendix N provides guidance for making plant auxiliaries available by energizing non-vital buses with a diesel. Although not required, they may aid in plant recovery.
| |
| Appendix Q Appendix Q provides diesel generator load limits including a graph displaying diesel generator load limits (Generator Operating Hours / Year vs.
| |
| Generator KW), and a table displaying equipment load ratings.
| |
| 1 Appendix X Two vital buses must be energized in order to have one ECCS train l complete. Appendix X provides guidance for using one vital bus to l energize a second vital bus.
| |
| Site Emergency Coordinator (or designee) approval is required prior to implementation of this app :ndix. j LPECAO. Doc PAGE 40 oF 78 REV. 6
| |
| | |
| f I
| |
| JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Verify misaligned rod is not stuck l K/A
| |
| | |
| ==Reference:==
| |
| 001000K5.01 (3.7)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| OlM p. A-3-4; STG A3A, Obj.11, Pgs. 2-4G and 2-41 QUESTION # 6a:
| |
| Unit 2 Control Rods are being inserted in the MANUAL mode during a reactor shutdown.
| |
| Assuming Control Bank D rods are at 110 steps, how many steps must rods be inserted before Control Bank B rods begin to insert?
| |
| ANSWER:
| |
| 138 steps i 1 step Candidate's Response: SAT UNSAT l
| |
| f-A;lh , $g, . , _ tr , _ f.f
| |
| | |
| - 2 _
| |
| 1 6 _
| |
| - - - ]
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| s
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| - - - 4)6 g 8 S p n eJ
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| \ 'N p 4( it a
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| t
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| - Da e l
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| S l - &er s r Ng rl .
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| - CvO ee _
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| e 4 )S t e -
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| - nh v 8 3(S u w
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| - - - o b O - - -
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| 6 Cmu 5 )4 t
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| p 3( S inh UT k -
| |
| Calr of pi n
| |
| t
| |
| - an
| |
| - &e l rU a
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| BvO N%x> e .
| |
| - vp .
| |
| 6)3 _
| |
| B -
| |
| 5 2(S Ola ke nv r
| |
| l BaO o - - - _
| |
| : 8) 2
| |
| - 2 S k r S 2( n t
| |
| n Ba
| |
| &e lr p
| |
| ./
| |
| . B
| |
| )
| |
| a o -
| |
| AvO x' 8 )1
| |
| (
| |
| [
| |
| C -
| |
| ~
| |
| 2 1( S
| |
| ,/
| |
| 8 8 0 2 2 0 2 1 1 s
| |
| p t
| |
| e S
| |
| 3k $1m
| |
| | |
| 7 Rod Controi System l Bank Overlap Unit l
| |
| l Purpose The purpose of the Bank Overlap Unit is to ensure that the control bank
| |
| ' Obj 6,11 rods move with proper overlap and in proper sequence. Bank overlap is only used when the Rod Bank Selector switch is selected to the AUTO and MAN position because the other positions are for individual bank control.
| |
| In the discussions of the slave and master cyclers, the Bank Overlap Unit l was responsible for directing the master cycler to send "Go" pulses to the proper slave cyclers which supply current orders to the proper power cabinets. The Bank Overlap Unit also supplies the group select multiplexing signals to ensure that the proper groups of rods move.
| |
| Counter . The Bank Overlap Unit has a counter which counts up (rods out) or down (rods in) and starts from 0.
| |
| . The value represents the total Control Bank steps withdrawn.
| |
| . The counter value is displayed in the Logic Cabinet.
| |
| DCPP Bank The objective of sequencing is to move the rod banks in a pre-selected Overlap order. When the control rods are on the bottom, they move out (beginning with bank A, and followed by B, C, and D) in order. They also must stop traveling when they reach their top limit. I The overlap will vary to accomniodate changes in the rod full-out position. l For example, when the rod full-out position is 228 steps, the overlap !
| |
| switches will be set to give 100 steps overlap.
| |
| Bank Starts when.. Stops at... j A X 228 Steps ;
| |
| (228 total steps)
| |
| B- Bank A at 128 steps 228 Steps (128 total steps) (356 total steps)
| |
| C Bank B at 128 steps 228 Steps (256 total steps) (484 total steps)
| |
| D Bank C at 128 steps 228*
| |
| : u. _ (384 total steps) (612 total steps) _
| |
| * If moved in MANUAL, in AUTO, centrol interlock C-1I stops Bank D at 220 steps.
| |
| Continued on nextpge A3A. DOC 2 -40 REV 6 t
| |
| | |
| I Major Components L
| |
| l Bank Overlap Unit, continued l
| |
| DCPP Bank Overlap (continued) l The figure below illustrates the bank overlap.
| |
| 1
| |
| - - - - - - ~ ~ ~ ~ ~ -
| |
| 228 - - - - - - - - ~ ~ ~ ~ ~ ~ - - - ------- ---
| |
| I d 4 N N !
| |
| --~~~-- -- --- -- ~~
| |
| 128 - -
| |
| Steps .A&B- : B&C C&D eda @ eda Neqa
| |
| \
| |
| v % O . o NJ
| |
| / / s
| |
| /
| |
| ~
| |
| / Y 128 228 256 356 384 484 612 (S1) (S2) (S3) (S4) (SS) (SS)
| |
| Bank Overlap Unit Counters
| |
| [( ) Bank Overlap Unit Thumbwheel Settings)
| |
| RCL-6(A-3-4)
| |
| Components The table below describes the components of the Bank Overlap Unit.
| |
| Component Stage Description
| |
| . "Go" pulses from the master cycler j Inputs . "In/Out" information from the supervisory i I
| |
| circuit
| |
| . The counter counts the number of "Go" pulses Counter Process to determine the total number of rod steps.
| |
| + Pulses are added for "Out" steps and subtracted for "In" steps Output The total rod step count is compared to the thumbwheel switches.
| |
| Continued on next page A3A. DOC 2 - 41 REV 6
| |
| | |
| JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Verify misaligned rod is not stuck KIA
| |
| | |
| ==Reference:==
| |
| 001000A2.15 (4.2) l Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Tech Spec 1.27 and 3.2.4; and, STP R-25, Calculation of Quadrant Power Tilt Ratio; LPWRDST, Obj. 5, Pgs.10 & 11 QUESTION # 6b:
| |
| Unit 1 is at 100% power with the following power range excore detector normalized !
| |
| values:
| |
| N41 N42 N43 N44 UPPER 0.96 0.95 0.96 0.99 LOWER 1.01 1.02 1.01 1.03
| |
| : 1. Calculate the OPTR.
| |
| : 2. Does the calculated OPTR exceed the Tech Specs?
| |
| ANSWER:
| |
| : 1. QPTR = UPPER N-44 at 1.0259 [0.75)
| |
| : 2. YES (0.25]
| |
| NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT ANSWER KEY
| |
| | |
| m
| |
| )
| |
| SEE PSRC INTERPRETATION 94-11 DEFINITIONS OPERATIONAL MODE - MODE 1.22 An OPERATIONAL MODE (i.e.. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1 1.2.
| |
| PHYSICS TESTS 1.23 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Comission.
| |
| PRESSURE BOUNDARY LEAKAGE 1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage, except steam generator tube leakage, through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.
| |
| PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that precessing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other requirements ,
| |
| governing the disposal of solid radioactive waste.
| |
| PURGE - PURGING 1.26 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating I condition, in such a manner that replacement air or gas is required to purify the confinement. )
| |
| i' OUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the '
| |
| maximum lower excore detector calibrated output to the average of the lower excore detector I calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average. ;
| |
| RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3338 MWt for Unit 1 and 3411 MWt for Unit 2.
| |
| DIABLO CANYON - UNITS 1 & 2 1-5 AMENDMENT NOS. 67 AND 66 32975802,4a II TAB 4 6 January 22. 1992
| |
| | |
| e l POWER DISTRIBUTION LIMITS 3/4.2.4 0UADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION
| |
| '3.2.4-.THE QUADRANT POWER TILT RATIOa sh'll not exceed 1.02.
| |
| ( APPLICABILITY: H00E 1 AB0VE 50% OF RATED THERMAL POWER *.
| |
| ACTION:
| |
| : a. With the 00ADRANT POWER TILT RATIO determined to exceed 1.02 but less than or
| |
| : j. equal to 1.09:
| |
| : 1. Calculate the OVADRANT POWER TILT RATIO at least once per hour until either:
| |
| a) The OVADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
| |
| : 2. Within 2 hours either:
| |
| a) Reduce the OVADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints l within the next 4 hours.
| |
| : 3. Verify that.the OVADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to'less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the rext 4 hours; and 4, Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50%
| |
| of RATED THERMAL POWER may proceed provided that the OVADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
| |
| : b. With the OVADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
| |
| : 1. Calculate the 0UADRANT POWER TILT RATIO at least once per hour until either.
| |
| *See Special Test Exceptions Specification 3.10.2 DIABLO CANYON - UNITS 1.& 2 3/4 2-18 Amendment Nos. 37 and 36 32975501.4A- .I TAB 9 -18 ~ Effective at.end of Unit l' Cycle 3
| |
| | |
| [Q' "L!UNCONTROLLEOPROCEDUREFDO ww mawwzwuxw.wamsaya;p.uu.mma:wau, N07!USETO PEREORM sun, WORKdQSSUEFOR 'USE;??9e PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 NUCLEAR POWER GENERATION REVISION 16 DIABLO CANYON POWER PLANT PAGE 1 OF 8 SURVEILLANCE TEST PROCEDURE UNITS TITLE: CALCULATION OF QUADRANT POWER TILT RATIO AND 06/03/98 l l EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED
| |
| : 1. SCOPE 1
| |
| 1.1 This procedure provides a method for determining the QPTR directly from the plant I 1
| |
| computer or calculating the QPTR using the upper and lower Power Range detectors.
| |
| 1.2 The Technical Specifications definition of QPTR is the ratio of the maximum upper (or lower) excore detector calibrated output to the average of the corresponding upper (or lower) excore detector calibrated outputs, whichever is greater.
| |
| : 2. RESPONSIBILITIES 2.1 Senior Power Production Engineer (Reactor Engineering) (Test Coordinator), for providing technical guidance, review, reporting of results, and updating the Master Surveillance Test Schedule.
| |
| 2.2 Shift Foreman, for obtaining test data, data reduction and for determination of Quadrant Power Tilt Ratio.
| |
| 2.3 Technical Maintenance (TM) Foreman, for providing support for utilization of the Power Range NIS Channel individual detector chamber current meters if necessary.
| |
| : 3. FREOUENCY 3.1 When the plant is in MODE I above 50% RTP, this procedure shall be used to i l
| |
| calculate the Quadrant Power Tilt Ratio (QPTR) as follows:
| |
| 3.1.1 At a nominal 7 day frequency if the power range QPTR alarm (Annunciator PK03-10 upper / lower QPTR) is OPERABLE.
| |
| 3.1.2 At a nominal 12 hour frequency if the alarm is INOPERABLE. This includes the time during NIS calibrations (STP I-2D).
| |
| NOTE: Above 75% RTP, this procedure shall not be used to calculate QPTR with an INOPERABLE NIS channel unless the individual detector chamber currents for the INOPERABLE NIS channel can be obtained and the affected detector has been declared OPERABLE per Step 12.2. If an NIS channel is to be taken out of service (e.g., during STP I-2D), then this procedure should be performed just prior to removing the channel from service in order to avoid having to calculate QPTR with an INOPERABLE NIS channel if at all possible.
| |
| 00191516. DOC 06 1
| |
| | |
| @w!?"UNCONTROLLEDPROCEDURE5DONOTUSE70iPERFORM 6uce mec- muumga.s'-m m e m m;4a m WORKb7I.wm m mswSSUEFOR;USE@
| |
| wam PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R 25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 2OF8 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2
| |
| : 4. TECHNICAL SPECIFICATIONS 4.1 Technical Specification 4.2.4.1 requires that with power above 50% RTP, QPTR be calculated at least once per 7 days.when the QPTR alarm is OPERABLE and at least once per 12 hours when the alarm is INOPERABLE.
| |
| 4.2 Technical Specification 4.2.4.2 requires that with one NIS channel INOPERABLE and power above 75% RTP, QPTR be calculated using the Movable Incore Detector System (MIDS) at least once per 12 hours. Since only the detector currents (read locally on the NIS drawer meters) are required for QPTR determination, this Tech Spec has been interpreted to require use of the MIDS (performed using STP R-11) only when one or both of an NIS channel's detector currents are unavailable (Reference 6.2). If the detector currents are available for all channels (detectors are considered OPERABLE per Step 12.2), then these may be used to calculate QPTR and satisfy Tech Spec 4.2.4.2 in lieu of the MIDS.
| |
| 4.3 Technical Specification 3.3.1 (Table 3.3-1, Action 2.c) requires that with one NIS channel INOPERABLE, EITHER THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85 % of RATED THERMAL POWER within 4 hours; or the QUADRANT POWER TIL T RATIO is monitored per specification 4.2.4.2 when TIIERMAL POWER is greater than or equal to 50% of RATED THERMAL POWER.
| |
| : 5. ACCEPTANCE CRITERI A 5.1 In MODE I above 50% RTP, the value of the calculated QUADRANT POWER TILT RATIO shall not exceed 1.02. If the QPTR exceeds 1.02 and power is above 50% RTP the Technical Specifications ACTION Requirements of Spec. 3.2.4 must be invoked.
| |
| : 6. REFERENCES 6.1 STP R-ll, " Quadrant Power Tilt Ratio Detennination By Movable Incore Detectors." i 6.2 Memo from T. L. Grebel to C. R. Groff, " Determination of Excore Detector Operability for QPTR Surveillance Requirement," March 30,1993.
| |
| : 7. APPENDICES.
| |
| None
| |
| : 8. ATTACHMENTS 8.1 " Calculation of Quadrant Power Tilt Ratio - Manual Method," 05/30/95
| |
| : 9. SPONSOR D. Farrer l 00191516. DOC 06 2
| |
| | |
| gQM Q @QQgQQhfQQQSEgEgRyg [gg g g Q PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 3OF8 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 i START DATA SECTION ;
| |
| UNIT OPERATING MODE DATE/ TIME /
| |
| : 10. PRECAUTIONS AND LIMITATIONS INITIALS 10.1 Once Power Range NIS Calibrations have commenced (STP I-2D), MANUAL calculations will be required until ALL FOUR Power Range NIS Channel calibrations have been COMPLETED. Until all Power Range NIS Channels are !
| |
| calibrated to the same revision of Table IIC-4, the PPC channels for the detectors which have not been calibrated will no longer be in current calibration. As such, the QPTR calculated by the PPC will be erroneous and must not be used.
| |
| 10.2 If a Power Range NIS Channel is out of service for any reason, the OPERABILITY of the Power Range NIS Channel must be evaluated and determined sufficiently in advance to allow for the j warmup of the MIDS. Six hours of notification to the Reactor Engineers is required to ensure that sufficient time is available to perform and analyze the data for a flux map.
| |
| 10.3 Should it be necessary to restore a Power Range NIS Channel to OPERABLE for the express purpose of performing this STP, sufficient time must be allocated to ensure that the TM Foreman can evaluate the status of the channel and have the Power Range NIS Channel individual chamber current meters restored to service. This time must also include sufficient time for any required warm up periods.
| |
| PERF
| |
| : 11. PREREOUISITES 11.1 The plant is in MODE 1.
| |
| l l
| |
| 00191sl6. DOC 06 3
| |
| | |
| 1 I
| |
| i m wmagz mm waammuungsmunema mmeAma.md
| |
| . '%aUN.CONTROLLEOPROCEDURE:?00NOTUSE!TOPERFORMWORK6flSSUEFOR~USE7 i PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 4OF8 i TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 l
| |
| l PERF j I
| |
| : 12. PROCEDURE l
| |
| 12.1 Record approximate power level.
| |
| % RTP l 12.2 If an NIS Power Range Channel is INOPERABLE, can the channel be restored o a point where individual detector chamber l currents may be ret ' The TM Foreman should be consulted in 1
| |
| making this determi .iori. Check N/A if all four channels are !
| |
| fully OPERABLE or .er is less than or equal to 75% RTP. !
| |
| YES [ ] NO [ ] N/A [ ]
| |
| 12.2.1 If Step 12.2 is NO, contact Reactor Engineering to i determine QPTR using MIDS (STP R-11), unless Y STP R-11 is already in progress. )
| |
| N/A [ ]
| |
| 12.2.2 If Step 12.2 is YES, then have the affected detector j current indications declared OPERABLE by the Shift l
| |
| Foreman before continuing. j N/A [ ] l NOTE: This procedure may be performed by either of two methods: l plant computer (Step 12.3) and manual (Step 12.4). The plant computer method is preferred. If the plant computer is not available, STP I-2D is in progress or an NIS channel is INOPERABLE, the manual method MUST l be used. !
| |
| l 12.3 QPTR by Plant Process Computer (PPC) method, i N/A [ ]
| |
| NOTE: This method uses a dedicated group display of points to help determine QP'TR.
| |
| 12.3.1 At a PPC SDS terminal call up addressable points U1261-U1268 in order (such as, by using GRPDIS and displaying group OP R-25 on a 60 second update frequency).
| |
| 12.3.2 Verify that all data point qualities are acceptable (for this test, acceptable point qualities are GOOD, liALM, LALM,11WRN, LWRN and INiiB). If not, mark this step and Steps 12.3.3 and 12.3.4 N/A (explain in REMARKS) and skip to Step 12.4.
| |
| N/A [ ]
| |
| 00191516. Doc 06 4
| |
| | |
| V% '?RNCONTROLLEUPROCEDURE awwmeewmanuane?~DO,M0 FUSE TOPERFORM sa.waway a WORKbnISSU,me n.: rauamm, EFOR wUSE57G
| |
| .m u PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R 25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 5 OF 8 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 l
| |
| ! PERF 12.3.3 Record values from the SDS screen for the following points:
| |
| N/A[]
| |
| CURRENT POINT ID DESCRIPTION VALUE U1261 RADIAL FLUX TILT UP QUAD 1/ N-43A U1262 RADIAL FLUX TILT UP QUAD 2 / N-42A U1263 RADIAL FLUX TILT UP QUAD 3 / N-44A l U1264 RADIAL FLUX TILT UP QUAD 4 / N-41 A U1265 RADIAL FLUX TILT LO QUAD 1/ N-43B U1266 RADIAL FLUX TILT LO QUAD 2 / N-42B U1267 RADIAL FLUX TILT LO QUAD 3 / N-44B U1268 RADIAL FLUX TILT LO QUAD 4 / N-41B 12.3.4 Record the POINT ID and the CURRENT VALUE of the highest reading from Step 12.3.3. This is the Quadrant Power Tilt Ratio (QPTR).
| |
| N/A [ ]
| |
| POINT ID: CURRENT VALUE:
| |
| 4 J
| |
| 00191516. DOC 06 5
| |
| | |
| ga%wmaa.maammw" G WNCONTROLLEDPROCEDURE5D0NOFUSE70 x e unamuw;aw.:WORKFrTSSUEFOR PERFORM waxapuuw e mawn.w USEM,.Q.J PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 6 OF 8 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 PERF NOTE: Attachment 8.1 may be filled out by hand, or the approved Microsoft Excel spreadsheet (located on Control Room computers) may be used. If using the spreadsheet, affix the computer generated copy of Attachment 8.1 to the back of this procedure and sign.
| |
| 12.4 QPTR by the Manual Method: Determine QPTR by recording the currents and performing the calculations for Attachment 8.1 (5 significant figures are sufficient).
| |
| N/A [ ]
| |
| 12.4.1 Record QPTR from Attachment 8.1, Step 13.
| |
| QPTR 12.5 REMARKS: !
| |
| l
| |
| )
| |
| 12.6 Test performers and verifiers:
| |
| Name Sicnature Date/ Time Lnjt !
| |
| /
| |
| /
| |
| /
| |
| /
| |
| 1 00191516. DOC 06 6
| |
| | |
| I
| |
| , h m-mmm I?iquMCONTROLLED,PROCEDURERDOMOTUSET0PERFORMMORKbrISSUEFOR m.m m m m m m . m e m m u .m aUSEy%
| |
| l PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 l DIABLO CANYON POWER PLANT REVISION 16 l PAGE 7OF8 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 PERF
| |
| : 13. REVIEW AND ROUTING 13.1 (SFM/STA) Evaluate results:
| |
| 13.1.1 Verify QPTR (from Step 12.3.4 or 12.4.1) less than or equal to 1.02. YES [ ] NO [ ] N/A [ ] j If NO, and power level is above 50% RTP, comply l
| |
| with the Action Statement in Tech Spec 3.2.4.
| |
| 13.1.2 All 4 NIS power range channels and QPTR alarm are OPERABLE. This procedure satisfies Tech Spec 4.2.4.1.a. YES [ ] NO [ ] N/A []
| |
| CAUTION: If NIS power range channel is inoperable for greater than 12 hours, comply with Action 2.c of Tech Spec 3.3.1.
| |
| 13.1.3 QPTR Alarm INOPERABLE and either all four power range NIS channels OPERABLE or reactor power limited to less than or equal to 75% RTP. This procedure satisfies Tech Spec 4.2.4.1.b.
| |
| YES [ ] NO [ ] N/A [ ]
| |
| 13.1.4 One NIS power range channel INOPERABLE, but detector and ammeter OPERABLE. This procedure satisfies Tech Spec 4.2.4.2. YES [ ] NO [ ] N/A []
| |
| 13.1.5 If neither Step 13.1.2,13.1.3 or 13.1.4 are satisfied, l this procedure does not satisfy a Tech Spec surveillance requirement but may be used to provide ,
| |
| data for STP R-il. YES [ ] NO [ ] N, A [ ]
| |
| 13.2 (SFM) Review completed procedure. If Step 13.1.1 is not l satisfied, refer to the applicable LCO and initiate or reference an Action Request.
| |
| A/R #
| |
| REMARKS: ,
| |
| Signature: Date/ Time /
| |
| Shift Foreman 00191516.DrC Of 7
| |
| * O' SKhane qAgauumumaaha3mawressuumawc MMCONTRDLLEDPROCEDURE=DGNOTUSET0 -
| |
| PERFORM WORKuYISSUEEDR -4 USE*22]
| |
| __ PACIFIC GAS AND ELECTRIC COMPANY NUMBER STP R-25 DIABLO CANYON POWER PLANT REVISION 16 PAGE 8 OF S TITLE: CALCULATION OF QUADRANT POWER TILT RATIO UNITS 1AND2 PERF 13.3 PPE (Reactor Engineering) verify the Master Surveillance Test Schedule has been updated.
| |
| 13.4 PPE (Reactor Engineering) review test for completeness and acceptability.
| |
| REMARKS:
| |
| Reviewed By: Date/ Time /
| |
| PPE (Reactor Engineering) 1 1
| |
| 1 00191516. DOC 06 8
| |
| | |
| ywygescournoitroesociount=ooworuseroeraronswoaxugssueronusm wy nsa m,am. a,um g na aa ega w w m a ucra;saa m m s w w g 05/30/P5 Page 1 of-2 DIABLO CANYON POWER PLANT STP R-25 ATTACilMEh t' 8.1 AND TITLE: CALCULATION OF QUADRANT POWER TILT RATIO - MANUAL METHOD N41 N42 N43 N44
| |
| : 1. RECORD channel AFDs. If any channel AFD is not -- I available, mark N/A for that channel. I
| |
| : 2. CALCULATE average AFD of values in Step 1.
| |
| If a channel AFD is N/A, then calculate average AFD for the other 3 channels.
| |
| : 3. RECORD reactor power (%RTP)
| |
| Recommend U4300 or U4300A05 or
| |
| ((U1169 + PPC Max)
| |
| * 100) 4, CALCULATE Core Average Axial Offset: (IAO)
| |
| (Step 2
| |
| * 100 + Step 3)
| |
| : 5. RECORD Indicated Currents (MICROAMPS)
| |
| (See Notes: I and 2)
| |
| : a. Upper Detector
| |
| : b. lower Detector
| |
| : 6. RECORD Core Average Axial Offset (IAO) from Step 4 in table below.
| |
| : 7. RECORD Vol. 9 (pg. II-20,20a) Table IIC-4 Detector Current Data (Slope and Intercept) in table below (see Note 3).
| |
| : 8. CALCULATE and RECORD in table below:
| |
| Normalized detector current = IAO
| |
| * m + b (From Step 4) (Slope) (Intercept) Normalized detector j NI IAO
| |
| * m + b = current i 1
| |
| 41 upper (8a) 1 41 lower (8b) 42 upper (8a) 42 lower (8b) 43 upper (8a) 43 lower (8b) 44 upper (8a)
| |
| , 44 lower (8b) i 00191516.IX)C 06 9 1
| |
| | |
| { ~.:
| |
| VQQmtraum m wnema n gauum e wsw e ggamsm2w.
| |
| Es INCON7ROLLEDPROCEDURE5DONOTUSE?TDPERFDRMWORKElISSUEFOR;USEG,f 05/30/95 Page 2 of 2 l STP R-25 (UNITS 1 AND 2)
| |
| ATTACIIMENT 8.1 TITLE: CALCULATION OF QUADRANT POWER TILT RATIO - MANUAL METiiOD 1
| |
| : 9. CALCULATE Normalized values:
| |
| : a. Upper Detector Normalized Value N41 N42 N43 N44 l (Step Sa. divided by Step 8a.)
| |
| Lower Detector Normalized Value b.
| |
| (Step Sb. divided by Step 8b.)
| |
| : 10. CALCULATE average of normalized values: 1
| |
| : a. Upper Detector Average Normalized Value l (Average of 4 values in Step 9a.) i
| |
| : b. Lower Detector Average Normalized Value !
| |
| (Average of 4 values in Step 9b.) .
| |
| 1
| |
| : 11. CALCULATE Quadrant Tilts:
| |
| : a. Upper Detector Tilts (Step 9a. divided by Step 10a.) ,
| |
| I
| |
| : b. Lower Detector Tilts (Step 9b. divided by Step 10b.)
| |
| l l
| |
| : 12. RECORD Upper and Lower Maximum Tilt:
| |
| : a. Upper Detector - Max value of Step 11a.
| |
| : b. Lower Detec:or - Max value of Step lib.
| |
| l
| |
| : 13. RECORD QPTR:
| |
| Largest value in Step 12.
| |
| Performed by: Date/ Time /
| |
| Reviewed by: Date NOTES:
| |
| : 1. OPERABLE detectors only. If Voltmeters installed per STP R-13/R-13B are used for determining individual detector currents, then record M&TE Calibration Data in REMARKS.
| |
| l 2. Use caution when .ecording values: Use 500 MICROAMP scale.
| |
| : 3. During recalibration (STP I-2D), use the new 100% full power detector currents (from the latest l
| |
| i revision of Table IIC-4) for all NIS channels, including those not yet recalibrated.
| |
| 00191516. DOC 06 10
| |
| | |
| LESSON: POWER DISTRIBUTION CONCERNS LESSOM No.: LPWRDST .
| |
| Quadrant Power Tilt Ratio (QPTR) l l
| |
| 1 l
| |
| Definition .The Quadrant Power Tilt Ratio is defined as the greater of the ratio of the )
| |
| Obf 1 maximum upper detector output to the average upper detector output and )
| |
| the maximum lower detector output to the average lower detector output. )
| |
| l I
| |
| QITR The upper and lower QPTR are calculated by the following formulas:
| |
| calculation Upper Detector nuGuEsT Upper QPTR =
| |
| Upper Detector AWRAGE l
| |
| l Lower QPTR =
| |
| Lower Detector AWRAGE The higher of these two ratios is the QPTR.
| |
| 1 QPTR Given the following data from the NIS. '
| |
| calculation example N41 N42 N43 N44 Upper detectors 99 % 98 % 99 % 96 %
| |
| Lower detectors 101 % 100 % 102 % 99 %
| |
| 99 % 99 %
| |
| Upper QPTR = f = = 1.010 99% + 98% + 99% + 96%' 98 %
| |
| s 4 s 102 % t02%
| |
| Lower QPTR = 1.015
| |
| '101% + 100% + 102% + 99%' 100.05 %
| |
| < 4 >
| |
| QPTR = 1015 -
| |
| Method of Unlike AFD, the operator has little control over QPTR. Maintaining the control control rods in proper alignment is about the only control the operator has.
| |
| t i Continued on next page l
| |
| LPWRDST. Doc PAGE 10 OF 18 REV1
| |
| | |
| t l
| |
| LESSON: POWER DISVRIBUTION CONCERNS LESSON No.: LPWRDST ,
| |
| Quadrant Power Tilt Ratio (QPTR), continued l QPTR cffects Large tilts are typically caused by major problems, such as:
| |
| Obj l . improper fuel loading.
| |
| . undetected stuck or misaligned rod.
| |
| . gross deviations from the design enrichment
| |
| . catastrophic flow or temperature variations.
| |
| Of the effects given above, the only one the operator may be able to correct is the stuck or misaligned rod.
| |
| Misaligned rod A stuck or misaligned rod in one quadrant will cause the power production in all quadrants to be affected (all upper and lower tilts can be affected by a single rod that is significantly misaligned). l QPTR limits QPTR is limited to a value of 1.02 by the Technical Specifications.
| |
| QPTR QPTR is verified to within its limits by the performance of STP R-25, !
| |
| verification " Calculation of QPTR ". This STP provides two methods of performing the calculations for QPTR:
| |
| . Calculation by the PPC (preferred method)
| |
| . Manual calculation LPWRDST. Doc PAGE 11 oF 18 REV 1 l
| |
| | |
| l
| |
| (. !
| |
| JOB PERFORMANCE MEASURE TEST ITEM l
| |
| ; Task
| |
| | |
| ==Title:==
| |
| Respond to PZR LOW pressure alarm i
| |
| l KIA
| |
| | |
| ==Reference:==
| |
| 010000A1.09 (3.7)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Steam Tables; Mollier Diagram; and, STG A4A p. 3-4, Obj. 25 QUESTION # 7a:
| |
| The plant is operating at 100% power when a pressurizer safety valve inadvertently lifts and sticks open.
| |
| With PRT pressure at 100 psig just before the rupture disk blows, what would be the approximate tail pipe temperature of the safety valve?
| |
| i ANSWER:
| |
| 340'F i 10 F Candidate's Response: SAT _ UNSAT b.
| |
| | |
| F i
| |
| l,.
| |
| Table 3. Superhealed Steam-Conunued !
| |
| p -
| |
| fbal (Sat. Tempi Sat Sat ' Temperature -Degree 5 f ahrenhes!
| |
| Water Steam 354 400 458 $40 $50 600 700 800 000 1000 1100 1200 1380 1400 "
| |
| Sh j
| |
| at 3796 8796 137 96 18796 237 96 287 96 347 % 487 % $87 96 68796 787.96 88796 987.96 1087 96 b 1312 041 e 0 0l?57 5 471 5 801 6 218 6 622 7 018 7 408 7 794 8 560 9 319 10 075 10 829 11.581 12 331 13 081 13 829 h
| |
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| |
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| |
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| |
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| |
| I.
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| |
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| |
| a . . t 129 Sh 8 73 58 73 108 73 158 73 208 73 258 73 358 73 458 73 558 73 658 13 758 73 858 73 958 73 1058 13 ,t v 0 01789 3 7275 3 1815 4 0786 4 3610 4 6341 4 9009 51637 5 6813 6 1928 6 7006 7 2060 7 7096 8.2119 8 7130 9 2134 13412h h 312 58 1190 4 !!95 6 1224 1 1251 2 1277 4 1302 9 1328 2 13784 1428 8 1479 8 1531 4 1583 9 16371 1691 3 1746 2 s-t 0 4919 1.5879 1 5943 1 6286 16592 1 6872 11132 11376 1 7829 1 8246 I 8635 09001 1.9349 1 9680 1 9996 2.0300 ,i l
| |
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| |
| 047.331 318 95 1891 7 !!93 4 1222.5 1249 9 1276 4 1302 1 1327 5 1377 9 14284 1479 4 1531.1 158J 6 1636 9 1691.1 17461 !
| |
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| |
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| |
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| |
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| |
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| |
| 26 92 76 92 126 92 176 92 226 92 326 92 426 92 52692 626 92 726 92 826 92 926 92 1026 92 '1 188 v 0 01827 2 5312 2 6474 2 8508 3 0433 3.2286 3 4093 3 7621 4 1984 4 4508 4 7907 5 1789 5 4657 5 8014 6 1363 i 0 73.081 h 34619 11 % 9 1243 8 1243 4 1271 2 1297 9 1324 0 1375 3 1426 3 14 77.7 1529 7 1542 4 1635 9 1690.2 17453
| |
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| |
| Sh e 0 01833 ?4030 2247 7241 12241 17247 22247 32247 42247 $2247 622 47 72247 82247 92247 l02241 $,
| |
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| | |
| p system Opdration 4
| |
| Normal Operation, continued Effects of Equipment operating effects en system indications are summarized below.
| |
| operation Although numerous possibilities exist for the initial conditions of operating Obj 4, 5, 25, a componen:, the information below assumes operation from a normal 50 configuration.
| |
| Effect of operating ... Consequence l The operator should expect to see ... ;
| |
| PCV-455A/B proportional heaters increasing their output, I backup heaters turning on, slow decrease in pressure depending upon how for the spray valve is open.
| |
| PCV-455C/456/474 rapid pressure reduction, reactor trip, and safety injection if valve has failed open. PRT pressure rises, PORV tailpipe temperature increases to saturation temperature for PRT pressure. After valve is closed, the temperature will gradually return to ambient.
| |
| RCS-8000A/B/C no consequence to plant operation. Alarm alerts operator to situation.
| |
| Pressurizer heaters no immediate change in plant operation. Heaters while they are will burn out and lose effectiveness in I uncovered controlling pressure. i DC fuses for . amber light deenergizes ]
| |
| PCV-455C/456/474 . control switch is inoperable !
| |
| removed Cycling PCV-456 in PRT level increase mode 5 3 RCPs (on spray) The valve in the loop without an RCP will lose its effectiveness. The valve should be closed in manual to prevent the valve opening on high pressure and "short circuiting" spray flow away from the Pressurizer.
| |
| Proportional heaters Pressure may increase to the point where sprays begin to modulate open Backup heaters Spray valves modulate open to control pressure PORV low setpoint no consequence protection c/o switch Continued on next page A4A. DOC 3-3 REV 8
| |
| | |
| f Pressurizer, Pressurizer and level Control system '
| |
| l Normal. Operation, Continued l
| |
| l Effects of operation (continued) i Effect of operating ... Consequence The operator should expect to see ...
| |
| Recorder selector no consequence 1 switch
| |
| - Control channel no consequence selector switch Master pressure . heaters turning off, sprays opening,455C controller opening, pressure lowering
| |
| . heaters turning on, pressure rising Control power switch . no consequence for vital bus breaker to heaters . switch allows breaker operation Control power switch . no consequence for innvital bus breaker to heaters . switch allows breaker operation RCS-8000A/B/C . valve becomes inoperable l breaker RCS-RV-8010A/B/C rapid pressure reduction, reactor trip, and safety l injection if valve has failed open. PRT pressure i rises, PORV tailpipe temperature increases to i saturation temperature for PRT pressure. After valve is closed, the temperature will gradually return to ambient.
| |
| HC-459D PDP speed rises, charging flow rises, Pressurizer level rises.
| |
| A4A. DOC 3-4 REV 8
| |
| | |
| i p
| |
| I JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Respond to PZR LOW pressure alarm K/A
| |
| | |
| ==Reference:==
| |
| 010000K1.02 (4.1)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Dwg 458858, Schematic Diagram Pressurizer Heaters; and '
| |
| STG A4A, p.2.1-24, Obj. 22 i
| |
| QUESTION # 7b:
| |
| Unit 1 is at 1009'o power with pressurizer heater group 1-2 being powered from breaker 52-1G-72. j What effect would a Safety injection signal have on pressurizer heater group 1-2? ,
| |
| l ANSWER:
| |
| Safety injection signal trips the pressurizer heater vital breaker to prevent overloading the vital bus.
| |
| Candidate'ti Response: SAT UNSAT
| |
| .M fl dT{h' jft f 1 7 [@M Q &
| |
| I hj@fk @Sp22 s33 kqS* FTM4 4-bglfjd{g%]
| |
| l
| |
| ! SEP
| |
| %29 ik Y[/y bd% dp[%
| |
| B
| |
| $f#m b dY
| |
| | |
| i I
| |
| ' Pressurizer, Pressurizer Pressure and Level Control System -
| |
| Backup Heater HSDP Controls, Continued Backup heater automatic actions (continued)
| |
| )
| |
| The table below describes the conditions under which the Backup Heaters are deenergized.
| |
| . Switch Positions Condition I Transfer switch in CONT RM . Backup Heater Control switch on CC-1 Norrnal power aligned, is placed in OFF.
| |
| . MPC output 215.6% when the Backup Heater Control switch is in AUTO-after-OFF. I
| |
| . Level deviation-high reduced to less than
| |
| +5% of span (LC-459E) when the Backup Heater Control switch is in AUTO-after-OFF.
| |
| . Low Pressurizer level,17% of span I (LC-459C or 460C), when the Backup Heater Control switch is in AUTO-after-OFF or AUTO after-ON.
| |
| . Overcurrent.
| |
| Transfer switch in CONT RM . Backup Heater Control switch on CC-1 Vital power aligned to OFF. {
| |
| . Safety Injection signal trips the breaker to prevent overloading of the vital bus; the breaker cannot be reclosed until the SI signal is reset.
| |
| l Group 2 . Backup Heater Local Control switch on Transfer switch in LOCAL HSDP to OFF (HSDP)
| |
| Normal power aligned . Low Pressurizer level,17% of span (LC-459C or 460C)
| |
| Continued on nextpage A4A. DOC 2.1 - 24 REV 8
| |
| | |
| P f JOB PERFORMANCE MEASURE TEST ITEM l
| |
| Task
| |
| | |
| ==Title:==
| |
| Align alternate AFW from the FWST K/A
| |
| | |
| ==Reference:==
| |
| 061000K2.01 (3.3)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Dwg 455060, Schematic Diagram Auxiliary Feedwater Pump Control; STG D1, Pg. 2-24, Obj. 24 QUESTION # 8a:
| |
| The green light on FCV-95 for AFW Pp 1-1 is not lit and the operator suspects the power supply breaker may be tripped.
| |
| What breaker feeds power to the green light?
| |
| ANSWER:
| |
| Breaker 72-1230 Candidate's Response: SAT UNSAT b $1 iY e,h (el B 9% iy hl d"n!! 1%dl Nify hajd 3=t' y=
| |
| : h. W((% tha s
| |
| (W=m V)j
| |
| | |
| Auxiliary Feedwater System l
| |
| l TDAFW Pump Steam Supply Valve FCV-95 l
| |
| Purpose The purpose of FCV-95 is to provide isolation of the steam supply to the Obj l3 TDAFW pump and maintain it in a standby condition.
| |
| Location FCV-95 is located in the 115' penetration area.
| |
| Obj 8 See figure on page 2-22 Power supplies FCV-95 is powered from DC Pnl 12.
| |
| Obj 24 Description FCV-95 is a DC powered MOV. j
| |
| . It is designed to operate during a loss of all AC. !
| |
| I l
| |
| Controls . FCV-95 has control capability from the
| |
| + Control Room on VB-3 Obj 25 + HSD Pnl I
| |
| + local handwheel 1
| |
| . FCV-95 transfer relay has control capability from 480V Bus G.
| |
| In the Control Room: '
| |
| Control for valve: Operation CLOSE/OPEN 3 position, spring return to neutral.
| |
| At the HSD Pnl:
| |
| Control for valve: Operation l CLOSE/OPEN 3 position, spring return to neutral.
| |
| l CONT RM/ LOCAL 2 position, maintained. l i
| |
| At 480 V Bus G: ;
| |
| Control for transfer relay: Operation l CUTIN / CUTOUT 2 position, maintained.
| |
| TRIP / RESET 2 position, maintained.
| |
| Continued on nextpage l
| |
| DI. DOC 2 - 26 REV 8
| |
| | |
| i s
| |
| JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Align alternate AFW from the FWST K/A
| |
| | |
| ==Reference:==
| |
| 061000A1.04 (3.9)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Vol. 9 page 1F-1, Required Condensate Storage Capacity as a Function of Hot Standby Tirne and Cooldown Time QUESTION # 8b:
| |
| i Unit 2 was at 100% power when a reactor trip and loss of offsite power occurred. !
| |
| The Condensate Storage Tank is at 53% and it will take 6 hours to cool down to 350 'F.
| |
| ' How long can the plant stay at Hot Standby before cooling Down?
| |
| ANSWER:
| |
| 3.1 hours i0.2 hours Candidate's Response: SAT UNSAT
| |
| * ! i'
| |
| [h1 P
| |
| Phib4d, O # Y C 7a If \ . )47)
| |
| | |
| E .
| |
| i
| |
| - - I l-OIABLO CANYON POWER PLANT OPERATION DATA REQUIRED CONDENSATE STORAGE CAPACITY AS A FUNCTION OF l HOT STANDBY TIME AND C00LOOWN i l l TIME (RATED MWt=3423) B
| |
| , ' l
| |
| {
| |
| } {
| |
| . i 4
| |
| . t i , , i ,
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| . i s 30 u 2 3 4 5 6 0 1 Page IF-1 TIME AT HOT STAND 8Y (HOURS) FIGURE ID-1 i
| |
| l REVISION 1 DATE 4/6/81
| |
| | |
| q l JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Crosstie the CCW system K/A
| |
| | |
| ==Reference:==
| |
| 008000A3.05 (3.1)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| Dwg. 445651, Separation and Color Code Instrument and Control '
| |
| Engineered Safety Features; and, dwg. 437608, Schematic Diagram Component Cooling System Motor Operated Valves; STG F2, Pg. 2-42, Obj.10 QUESTION # 9a:
| |
| While the Control Room Asset Team was performing testing in Train B of SSPS, an inadvertent Phase B occurred on Train B.
| |
| What effect did this have on the CCW system?
| |
| ANSWER:
| |
| : 1. FCV-363, RCP Oil Coolers CCW Return, closed [0.5]
| |
| : 2. FCV-357, RCP Thermal Barrier CCW Return, closed [0.5]
| |
| l NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT i
| |
| ANSWER KEY
| |
| | |
| [h
| |
| ' Component Cooling Water System I
| |
| Non-vital Header C, continued 1
| |
| i Automatic - The following table describes the automatic actions of Header C valves:
| |
| Actions - ;
| |
| Obj 10, il l Signal The following valve (s) close:
| |
| l Phase A FCV-361, l Excess Letdown HX CCW Flow Control Valve FCV-355, CCW Hdr C Supply valve FCV-356, CCW Supply to RCP and RV Support coolers )
| |
| FCV-363, Phase B RCP Lube Oil cooler CCW Flow Control valve FCV-749, RCP Lube Oil cooler CCW Return valve (
| |
| FCV-750, RCP Thermal Barrier CCW Return valve FCV-357, RCP Thermal Barrier CCW Return valve High-High flow in the FCV-357, j line RCP Thermal Barrier CCW Return valve ;
| |
| F2. DOC 2 - 42 REV 7
| |
| | |
| I J l
| |
| l JOB PERFORMANCE MEASURE TEST ITEM Task
| |
| | |
| ==Title:==
| |
| Crosstie the CCW system K/A
| |
| | |
| ==Reference:==
| |
| 008000G2.1.32 (3.8)
| |
| Reference Allowed: No
| |
| | |
| ==Reference:==
| |
| STG F2, pp. 2-16,2-17, and 2-18, Obj. 32; ECG 14.1 1
| |
| QUESTION # 9b: j Unit 1 Component Cooling Water (CCW) head tank pressure has dropped to less than {
| |
| 19 psig due to a problem with the cover gas system.
| |
| What sequence of events could lead to flashing and water hammer in the CCW system if CCW head tank pressure is not restored?
| |
| ANSWER:
| |
| A Large Break LOCA coincident with a Loss of Offsite Power that results in double l sequencing of 4Kv buses. I NOTE: Both parts must be answered correctly to receive a SATISFACTORY grade.
| |
| Candidate's Response: SAT UNSAT
| |
| %4 4 g4 3 B b w % 29 b. 4,6 rg'=
| |
| d r 5 arca a va n 6' b n aum a r= a()
| |
| | |
| f L
| |
| ' Component Cooling Water System Surge Tank j i
| |
| l l
| |
| 1 Purpose The purpose of the surge tank is to:
| |
| Objl9,32 . . accommodate surges from thermal expansion and contraction
| |
| . provide makeup for system leakage
| |
| . accommodate leakage into the CCW system
| |
| . maintain net positive suction head for the CCW pumps
| |
| . casure water to at least one vital cooling loop in the event of a rupture on )
| |
| one of the headers. l
| |
| . ensure sufficient static head to prevent CCW from flashing in the CFCUs during LOCA coincident with loss of offsite power (N2 Pressurization 4 system).
| |
| Loution The location of the surge tank and its related valves is described in the l Obj 6 following table. l Equipment Location Surge tank The roof of the auxiliary building at 1 the 163'- 4" elevation !
| |
| LCV-69, Train A Surge Tank )
| |
| Makeup valve !
| |
| CCW-62, LCV-69 Bypass valve Aux Building 100' penetration area LCV-70, Train B Surge Tank Makeup valve CCW-65, LCV-70 Bypass valve Continued on nextpage I
| |
| l l
| |
| I l
| |
| F2. DOC 2 - 16 REV 7
| |
| | |
| r Major Components Surge Tank, continued Surge Tank A diagram of the surge tank is shown below,
| |
| ^*"'***
| |
| diagram
| |
| 'C PCV 2M2
| |
| ,@MMD RCV.1e a I Swy l O [to 750 gen O
| |
| i i es 70 ua.up : : u a.up e2 65 susoi
| |
| < : B Header
| |
| );s2cso)
| |
| : A H o= CCW-06 Physical The surge tank is equipped with a vertical divider plate that establishes two description compartments. The divider plate ensures that water is available to at least one of the vital headers in case of a rupture.
| |
| One compartment is connected to loop A return piping; the other is I
| |
| connected to loop B return piping.
| |
| The divider does not reach to the top of tank. Therefore, when the level of j the tank is high, water from one side of the tank can spill over to the other side.
| |
| Characteristic Details Type Horizontal cylindrical tank l Materials Carbon steel Capacity 10,750 gallons Divider Plate 37.5% oflevel Continued on nextpage
| |
| . F2. DOC 2 - 17 IGV 7
| |
| | |
| Component Cooling Water System .
| |
| Surge Tank, continued 1
| |
| Purpose of During analysis of CCW system heat loading, it was discovered that a I Surge Tank postulated scenario could result in flashing in the CCW piping. l Pressurization . In the event of a Large Break LOCA coincident with a loss of off-site I power, the CFCUs would coast down. At some point during the coast I down the CFCUs would sequence on the diesel generators before the CCW pumps would start.
| |
| + The resulting heat up of the CCW fluid, due to the high temperature inside containment, would cause flashing of the fluid and subsequent l water hammer when the CCW pumps sequenced on.
| |
| In this scenario the integrity of the CCW system could be challenged. To '
| |
| preclude flashing and subsequent water hammer, the CCW surge tank will be pressurized to increase the static head on the system.
| |
| Surge Tank The Surge Tank Pressurization system is show below.
| |
| Pressurization
| |
| " ' " -~
| |
| System g _,_
| |
| yy y C "" p.
| |
| l .n _L_ .. l.n _ _ 1 7. E """' "*
| |
| - - . . -'6' -- , _
| |
| ]
| |
| =
| |
| l 1 i i !
| |
| -e -- CCW-19 Stage Description 85# N2 Normal source of N2 to pressurize the Surge Tank N2 Bottles l'' backup for source of N2 Inst. Air Last backup to pressurize the Surge Tank if N2 is not available PCV-2021A/B Pressure control and backup to reduce N2 bottle pressure to 75 psig PCV-2020A Controls Surge Tank N2 or air pressure at 20 psig for outsurges from the tank or for normal makeup PCV-2020B Backup for PCV2020A PCV-2002 Controls Surge Tank pressure at 25 psig for insurges to the tank due to heatup or inleakage Continued on nextpage F2. DOC 2 - 18 REV 7
| |
| | |
| c l
| |
| i
| |
| . unh 'S . -.u . M teT W , m3 CCW Surge Tank Pressurization System
| |
| 'p!
| |
| k l 14 1 l .
| |
| 14.0 i
| |
| _ _. COMPONENT
| |
| ..-n.. .,.COOL.ING
| |
| . . . . . -WATE~RiCCW)- - . - .
| |
| l
| |
| ~'
| |
| 14.1 ~ . s. CCW Suree Tank Pressndmilon System. !
| |
| t .".
| |
| ECG 14.1 The'CCW surge tank pressurization' system shall be.OPERABLEld '';
| |
| . .' ;1: m ,y APPLICABILITY: MODES 1,2,3, AND'4 '
| |
| l w .a c l & b a g m '. m saya g>hs m '""' s '
| |
| ACTIONS ' to v.
| |
| CONDITION REQUIRED ACTION COMPLETION --
| |
| . TIME -
| |
| A. ,,CCW surge, tank _
| |
| A.lg. Restore CCW pressurization. 12 hours "
| |
| i Pressurization sys, tem , ,; , a 4,systerp to OPERABLE status ww .s - ,a w . qM.I.'gm A a e: .m:4 m e ;:J' ' b**' ' '' 2
| |
| ~} '" ''
| |
| ''*V i fHj,Q ,:.,; t -
| |
| : u. r: . .e . pw X D tlk lin'u .. ~
| |
| g,,.. ~,a.. . , . . . . .; J '
| |
| b ''' '
| |
| 230kV offsite power '.g: tu ,g ;iqn no '.
| |
| source OPERABLE or at c zero voltage a 4: 4"'+
| |
| B. CCW surge tank M:..,'.Upon completion of Action
| |
| "' pressurization system B.3 or B.4 below, TS 3.0.3 may be inoperable exited
| |
| .. , o j ;,un,rt a- 'i -
| |
| ', a ..'n.a ca ya WO3 - 0- i id'd ' 'd A N' l,
| |
| $n!.m source 230kV offsite power , o degraded (as - co B.1D .tDeclare both trains of the
| |
| : b."t dCCW system ineper.ble '
| |
| 'idui A Eidtely #
| |
| "~
| |
| indicated by Power
| |
| " "' T 7' " # ' .5"'
| |
| Control)
| |
| L
| |
| *, . f t' B.2 Comply with TS 3.0.3 Immediately m; '-
| |
| 6 9au. o -
| |
| U
| |
| ~ ,
| |
| ,". ' , a.: , .
| |
| , .a O .5ds ri 'n - > I ' u "' 6 ' '"
| |
| "' 7.'
| |
| . o c ., . l3 B.3irt) Take'a~ctions''tiirestore either ;
| |
| lP"r5'oI,to'.e.xit1'88 '.3 "c"
| |
| ''4, .m'.
| |
| mr. .
| |
| 2 "rS 3:0,.3 i. ''
| |
| ..p.g ,
| |
| [.,. .i. .. uo o *11 the 1 CCW pressurization -
| |
| system or the 230kV system u,; . , !
| |
| to OPERABLE status
| |
| .. ; :.o - .
| |
| r-g %;4-
| |
| -*. ea rei -
| |
| ,.1 .
| |
| . -, a k ,e ,
| |
| , B.4 xNOpen the vital 4kV bus trip Prior'to exiting ''#"~' T. ,[
| |
| t 1
| |
| cut-out switches for TS'313" "
| |
| l i .
| |
| . a wt fer to,startup n,Mmo :, .imi MUEM S
| |
| ["'.
| |
| ,gg .. .. . . . .f4, 4 .HG-14, gius a c:a. e!E.4:3 j . y- " " ' - '
| |
| 'HH-14)
| |
| , o us .. wX; to it.1 .
| |
| , . ,' y' ~ . ~ 3v .g.:G: ,
| |
| n N. i',, i ks -
| |
| , v.
| |
| Diablo Canyon Units 1 &2 ,
| |
| ~
| |
| W ev.22 R '#
| |
| 01337502.DOA 31
| |
| .' N9-@
| |
| | |
| m
| |
| ,, . CCW Surge Tank Pressurization System 14.1 SURVEILLANCE REQUIREMENTS SURV W ANCE FREQUENCY 14.1.1 Verify the CCW surge tank is pressurized at greater than or : 12 hours l
| |
| j equal to 17 psig. (STP I-1A) , ,
| |
| , 14.1.2 Verify the CCW surge tank pressurizatiop system leakage does 24 months not exceed 2.0 scfm. (STP M-130)
| |
| BASES' The addition of the CCW surge tank pressurization system 1,rovides assurmce that boiling in the containment fan cooler units (CFCU) will not occur as a result of a design basis accident with a concurrent loss of offsite power (LOOP). Sufficient pressure is ==3nraiaM in the surge tank to provide sufficient static head in the CFCUs to prevent boiling until the CCW pumps are re-energized using the emergency diesel generators (EDG). After the CCW pumps restart, the dynamic head developed by the flow through the system is sufficient to maintain the fluid in the CFCUs in a s'ub-cooled condition.
| |
| If the pressurization system is not available, it has been shown that with a limited displacement loss of coolant accident (LOCA) break size, determined using the methodology of NUREG-0800 (ref. 3) concurrent with a LOOP, boiling in the CFCUs will not occur prior to the restart of the CCW pumps on the EDGs.
| |
| The assumption of a design basis accident with a non-concurrent LOvP is beyond the licensing basis of i
| |
| DCPP. If, however, such a LOOP were to occur and the CCW surge tank remained pressurized, boiling wou?d 20t occur. If the CCW surge tank were not pressurized, then it is possible that boiling and void formation could occur for both limited displacement pipe breaks and double guillotine LOCAs and main j steam line breaks (MSLB) with a non-concurrent LOOP.
| |
| A probabilist ci risk assessment (PRA) analysis was used to determine a risk acceptable allowed outage 1 time (AOT) for the CCW surge tank pressurization system based on the licensing and non-licensing (non-concurrent LOOP) basis scenarios diene =M above.
| |
| . n. .
| |
| 1 The AOT was determined by equating the added risk of having the CCW surge tank pressurization system out-of-service (OOS) to a value of " acceptable" increase in risk. The level of " acceptable" increase in risk used was set at 1.0E-07 as suggested in EPRI TR-1053% (ref. 6) when considering added core damage probability which leads to an increase in the large, early release (outside containment) frequency.
| |
| The PRA analysis conservatively assumed that:
| |
| o Large (> 6 inches), medium (betwcen 2 and 6 inches), and small (less than 2 inches) LOCA initiating events result in coprainment conditions for a period of 24 hours which, in conjunction with a subsequent failure or degradation of the 230 kV power supply to the vnal busses, fail the CCWS leading to core damage. ,
| |
| o MSLBs inside coprainment result in containment conditions for a period of one hour which, in I conjunction with a suh=nt failure or degradation of the 230 kV power supply to the vital busses, fail the CCWS leading to core damage. i o Failure of the CCWS, in addition to leading to core damage leads to a Jr.rge early release outside containment.
| |
| l Diablo Canyon Units 1 & 2 Rev.2 l 01337502.DOA 2
| |
| | |
| d-CCW Surge Tank Pressurization System 14.1 The PRA analysis assumed as best estimates that:
| |
| * The probability of 230 kV becoming degraded over a 24 hour period is based on data at DCPP from August 1,1995 to the present, that indicates 230 kV is degraded approximately 1.2 times per year.
| |
| After completion ofinstallation of new 230 kV transformers with automa*.ic load tap changers in 1998, the occurrences of degraded 230 kV will drop to zero.
| |
| * The probability.of a loss of 230 kV over a 24 hour period is based on a combination of DCPP and indatry experience, and is estimated to occur at a frequency of 0.15 events per year.
| |
| Based on the above, the PRA determined that a risk acceptable AOT for the CCW surge tank pressurization systept being OOS is 14 hours. After installation of the new 230 kV transformers 1998, the calculated risk acceptable AOT will increase to 93 hours. The PRA calculated AOT included the risk associated with the " double sequencing" condition discussed above. It is, therefore, inclusive of the licensing and non-licensing basis scenarios as well as degraded offsite power supply conditions.
| |
| Although the PRA analysis indicated an AOT of 14 hours (93 hours after 1998) was " risk acceptable", a conservative AOT of 12 hours has been established for the CCW surge tank pressurization system. The 12 hour AOT should be sufficient to restore the CCW surge tank pressurization system to operable status and it minimizes the period of operation without the system.
| |
| If the CCW surge tank pressunzation system is inoperable with the 230kV system degraded, flashing could oc::ur under accident conditions which could render both trains of the CCW system inoperable unless double sequencing of the CCW pumps is prevented. In that case, within one hour, either CCW surge tank pressure or the 230kV system will be restored or actions will be taken to prevent double sequencing. To prevent double sequencing, the vital bus transfer to startup cutout switches will be opened to force the vital busses to transfer directly to the EDGs, effectively causing the vital busses to respond as if a LOOP occurred. Once the cutout switches are opened, or either the 230kV system or the CCW surge tank pressurization system are returned to service, the one hour action statement may be .
| |
| exited. If the above actions are not completed within one hour, the plant will be shut down.
| |
| REFERENCES
| |
| : 1. Operability Evaluation 96-03
| |
| : 2. Design Change Packages M-049284 and M-050284
| |
| : 3. NRC Srmndard Review Plan (NUREG-0800), Section 3.6.2, Branch Technical Position MEB 3-1, Paragraph B.3.a.(3)
| |
| : 4. PG&E System Operations Instruction O-23
| |
| : 5. Operating Order OP O-27, " Coordination of 500/230kV System Activities"
| |
| : 6. Electric Power Research Institute Probabilistic Safety Assessment Applications Guide (EPRI TR-1053%), August 1995 Approved om / 6/2 /p y ' VP&PM, DCPP Effective Date -
| |
| Diablo Canyon Units 1 & 2 Rev.2 01337502.DOA 3
| |
| | |
| r l
| |
| l l
| |
| JOB PERFORMANCE MEASURE TEST ITEM l
| |
| 1 1
| |
| Task
| |
| | |
| ==Title:==
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| Perform local start of a Diesel Generator i K/A
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| | |
| ==Reference:==
| |
| 064000K1.04 (3.9) l Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| AR PK16-04, Diesel 11 Control UV; STG J6B, Obj. 58, Pg. 4-1 QUESTION # 104:
| |
| Unit 1 was in Mode 5, with ALL D/Gs operable.
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| l The SFM directed a Nuclear Operator to place D/G 1-1 on its backup DC source due to loss of the normal source.
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| What is the operability status of all the Unit 1 Diesel Generators?
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| ANSWER:
| |
| D/G 1-1 is inoperable. D/Gs 1-2 and 1-3 are still operable.
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| Candidate's Response: SAT UNSAT r
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| 'i !!
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| * r
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| : =h yg e
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| [
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| | |
| ~ . - . _ . - _ m, _ _ -. _ , m._ ._ ._ m L\*CLUNCON[ROLLEDEROCEQURFiDONO1,USEi!OEEREORMWORKb,d,SSUEEORD_SE28 PACIFIC GAS AND ELECTRIC COMPANY NUMBER AR PK16-04 NUCLEAR POWER GENERATION REVISION 7 DIABLO CANYON POWER PLANT PAGE 1 OF 4 ANNUNCIATOR RESPONSE UNIT TITLE: DIESEL 11 CONTROL UV l
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| APPROVED: AUTO INC 05/14/96 07/16/96 DATE EFFECTIVE DATE l
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| : 1. LOGIC DIAGRAM 27GHDC 27H23 m 3
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| 40L ,m O -
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| 3OL m
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| - 27Gs2 O .
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| sat m O m 60L '
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| O ALARM ,
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| DCF11 DCC1A11 ;
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| DCR1-11 y
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| * DCR2 -
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| l 43DC-11/SS 4300118/AS 3 0
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| : 2. ALARM INPUT DESCRIIFTION DEVICE ALARM ANNUNCIATOR TYPEWRITER NUMBER INPUT PRil4 TOUT SETPOINT DCF11 253 Dsl 1-1 DC Control UV Aux relay DCCI A11 253 Dsl 1-1 DC Control UV Aux relay DCRI-il 253 Dsl 1-1 DC Control UV Aux relay DCR2-11 253 Dsl 1-1 DC Control UV Aux relay o49519ZZ. DOC 16 1
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| | |
| hbh$ .$h h h$h$5hbhI5 5hh$@[$h h $$
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| NUMBER hh b AR PK16-04 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT REVISION 7 PAGE 2OF4 TITLE: DIESEL 11 CONTROL UV UNIT 1 l
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| DEVICE ALARM ANNUNCIATOR TYPEWRITER NUMBER INPUT PRINTOUT SETPOINT 43DC-11/SS 236 Dsl 1-1 on Backup DC Pwr Supply and/or Aux Relay Backup Fuse l l
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| 43DCllB/AS 236 Dsl 1-1 on Backup DC Pwr Supply and/or Aux Relay '
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| Backup Fuse 27 GHDC 268 Dsl 1-1 Ckt Prot Relay UV Aux relay l 27H23 855 Dsl 1-1 Auxiliaries UV or OC Aux relay 40L 855 Dsl 1-1 Auxiliaries UV or OC 4.28 amps l
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| 3OL 855 Dsl 1-1 Auxiliaries UV or OC 21.2 amps i 27G62 855 Dsl 1-1 Auxiliaries UV or OC Aux relay 80L 855 Dsl 1-1 Auxiliaries UV or OC 21.2 amps 60L 855 Dsl 1-1 Auxiliar;cs UV or OC 21.2 amps
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| : 3. PROBABLE CAUSE 3.1 Loss of normal DC control voltage.
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| 3.2 Backup DC control power and/or backup Appendix R fuses selected. l 3.3 Protection relay UV. )
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| 3.4 Diesel Auxiliary equipment UV or OC.
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| : 4. AUTOMATIC ACTIONS None
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| : 5. OPERATOR ACTIONS 5.1 Diesel DC Control UV NOTE: Diesel 11 is started from SSPS Train A and B.
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| 5.1.1 Check the normal DC power supply, bkr 72-1313.
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| : a. If the breaker is closed, consider selecting backup DC control power.
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| Go to Step 5.1.1.d.
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| : b. If the breaker is open, reclose the breaker. If normal power is not regained go to Step 5.1.1.d.
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| : c. If the breaker is in the trip free position:
| |
| : 1. Attempt to find out why the breaker tripped.
| |
| : 2. If the cause cannot be found, attempt to reset the breaker once.
| |
| : 3. If the breaker trips open again, follow Step 5.1.1.d.
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| 049519ZZ. Doc ' 16 2
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| | |
| D$$$$$$$15fbb6]&RR[0}$$0hkobN#G0555 NUMBER PACIFIC GAS AND ELECTRIC COMPANY AR PK1644 DIABLO CANYON POWER PLANT REVISION 7 l
| |
| PAGE 3OF4 TITLE: DIESEL 11 CONTROL UV UNIT 1
| |
| : d. If backup DC control power is needed:
| |
| NOTE: In Modes 1-4 the operability of the backup DC source (Battery 12) may be affected if DG l-1 DC control power is selected to backup. In non-emergency situations consult plant management prior to selecting backup DC control power. In Modes 5 and 6 the battery can sustain the simultaneous starting currents of 2 diesel generators, l therefore backup DC control power may be selected without affecting the operability of Battery 12. i 1
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| : 1. If the diesel generator is needed immediately, place DG l-1 in '
| |
| manual (Modes 1-4 only) and select backup DC control power.
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| DG l-1 may be manually started in accordance with OP J-6B:lV after DG 1-2 has been started.
| |
| : 2. When backup DC control power is selected check the "Dsl 1-1 DC control UV" alarm input clears. If the "Dsl 1-1 DC Control UV" alann input does not clear it is likely that a control power fuse is blown. Place the local / remote switch to LOCAL, select TEST on the 43-11 Mode Control switch (Modes 1-4 only), and place the Appendix R fuse selector switch 43DC-11/SS in BACKUP. Check that the "Dsl DC Control UV" alarm input clears.
| |
| 5.1.2 If the normal power supply is restored verify that the "Dsl 1-1 DC Control UV" alarm input clears. If the "Dsl 1-1 DC Control UV" alarm input does not clear it is likely that a control power fuse is blown. Place the local / remote switch to LOCAL and place the Appendix R fuse selector switch 43DC-II/SS in BACKUP. Check that the "Dsl DC Control UV" alarm input clears.
| |
| 5.2 Diesel 1-1 on Backup DC Power Supply and/or Backup Fuse 5.2.1 The Appendix R fuse selector switch,43DC-II/SS, should be in the normal position when remote control of the diesel generator is selected. Backup on 43DC-II/SS should only be selected after the diesel generator has been aligned for local control (local / remote selector switch to LOCAL).
| |
| NOTE: In Modes 1-4 the operability of the backup DC source (Battery 12) may be affected if DG l-1 control power is selected to backup. In non-emergency situations consult plant inanagement prior to selecting backup DC control power. In Modes 5 and 6 the battery can sustain the simultaneous staning currents of 2 diesel generators, therefore, backup DC control power may be selected without affecting the operability of Battery 12.
| |
| l If DG l-1 has backup DC control power selected all starting and control 5.2.2 circuits are fully functional. However, DG l-1 is inoperable in this configuration because DG 1-1 and DG l-2 control power supplies are not independent.
| |
| o19519ZZ. DOC 16 3
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| | |
| &llkk5hbkhhh8hh!$$j1$)EN0h.[5$Ej0]$b$0]$8b8&b$Nl5.50$$55 PACIFIC GAS AND ELECTRIC COMPANY NUMBER AR PK16-04 DIABLO CANYON POWER PLANT REVISION 7 PAGE 4OF4 '
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| TITLE: DIESEL 11 CONTROL UV UNIT I 5.3 Diesel protective relay UV:
| |
| 5.3.1 Check the protective relay power supply, bkr 72-1316.
| |
| 5.3.2 Deterraine the cause for protective relay loss of power.
| |
| l 5.3.3 If power cannot be restored, the three normally cut-out relays will not be '
| |
| available and the generator differential relay will not provide protection.
| |
| 5.4 Diesel auxiliaries UV or OC:
| |
| 5.4.1 Check contactor panel A power supply 52-lH-23.
| |
| : a. If contactor panel A is lost, the diesel is inoperable due to loss of both crankcase exhausters.
| |
| 5.4.2 Check contactor panel B power supply 52-lG-62. I
| |
| : a. If contactor panel B is lost, monitor start and turbo air pressures. l 5.4.3 Determine which component brought in the alarm off of the local alarm panel.
| |
| 5.4.4 Check turbo air and starting air pressures (refer to STP M-9A to determine adequate pressure for diesel operability).
| |
| 5.5 Refer to Tech Spec 3. 8.1.1, 3.8.1.2.
| |
| t Gl9519ZZ. DOC 16 4
| |
| | |
| ~ __
| |
| Section 4.0 Integrated Operations Limiting Specifications I
| |
| Tech Specs The following is a listing of Tech Specs identified for the D/G system.
| |
| Obj 6, 53, 54, Please refer to a copy of DCPP Technical Specifications for the details of 58 the LCO, applicability, action (s), surveillance (s) and basis for each.
| |
| T/S Item I 3/4.3.3.2 Engineered Safety Features Actuation System i Instrumentation 3/4.7.8.1.1 A.C. Sources (Operating) 3/4.3.8.1.2 A.C. Sources (Shutdown)
| |
| Safe shutdown The D/G system provides safe shutdown functions by providing emergency I power to equipment directly used in core cooling, (i.e. RHR pumps, CCW pumps, ASW pumps). The D/G equipment related to safc shutdown include:
| |
| . Two of the three D/Gs and associated equipmat
| |
| . One of the two FO transfer pumps 0:se of the two day tank LCVs per day tank
| |
| . One of the two FO storage tanks
| |
| .: .unsequences of losing the D/G function during a safe shutdown include the inability to provide long term core cooling.
| |
| Continued on nextpage I
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| I l
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| J6B. DOC 4-1 REV 9
| |
| : c. j I~
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| JOB PERFORMANCE MEASURE TEST ITEM :
| |
| l Task
| |
| | |
| ==Title:==
| |
| Perform local start of a Diesel Generator i l
| |
| K/A
| |
| | |
| ==Reference:==
| |
| 064000G2.4.30 (3.6) l Reference Allowed: 1 YES I
| |
| L
| |
| | |
| ==Reference:==
| |
| Administrative procedure X11.lD2, Regulatory Reporting Requirements and Repo'iing Process; LADM-3, Pg.11, Obj. 4 l
| |
| QUESTION 10b:
| |
| Unit 1 is at 100% power, an operator was directed to start DG 1-1 using the undervoltage starting method, per STP M-9A. DG 1-3 was started instead due to the l operator going to the wrong 4 kV bus to perform the required start.
| |
| What are the reportability requirements , if any for this Diesel Generator start?
| |
| ANS'NER:
| |
| Non-emergency 4 hour report i
| |
| Candidate's Response: SAT UNSAT ,
| |
| ~
| |
| [ g T I
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| | |
| r.
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| r i . 06/16/98 h hbhhkhhh $5kMlbtldibkE 5 5555 BE55 Page 6 of 10 l
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| XII.ID2 f ATTACllMENT 8.6 TITLE: 10 CFR 50.72 EVENT REPORT GUIDELINES l
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| 10 CFR 50.72 CRITERIA SPECIFIC EVENT GUIDELlNE C. Non-Emereenev Events: Four Hour Reports
| |
| : 1. Any event or condition that results in 1. See Attachment 8.7 manual or automatic actuation of any Engineered Safety Feature (ESF),
| |
| including the Reactor Protection System (RPS), except when:
| |
| l
| |
| : a. The actuation results from and is Any time that ESF " Systems" are actuated part of a preplanned sequence when the actuation is specifically identified as i during testing or reactor definitely expected to occur in a work order, i l
| |
| operation. procedure, or control room log, there are no I reporting requirements.
| |
| : b. The actuation is invalid and:
| |
| : 1) Occurs while the syso:m is Events involving the initiation of invalid ESF properly removed from actuation signals should be reported only when service; the associated ESF equipment is actuated or changes state. If the equipment actuated by
| |
| : 2) Occurs after the safety the invalid ESF signal is already in the function has been already condition required by the ESF funct,on, i the completed; or event should not be reported as an ESF
| |
| : 3) Involves only the following actuation.
| |
| specific ESFs: i For example, an invalid reactor trip actuation l Control Room Emergency after the control rods have been fully inserted ventilation system; into the core would not be reportable.
| |
| Containment ventilation 110 wever, a reactor trip actuation from a valid system; signal is reportable even if the control rods are fully inserted into the core.
| |
| Fuel Handling Building ventilation system; or Valid ESF actuations are those actuations that result from " valid signals" or from intentional Auxiliary Building manual initiation, unless part of a preplanned ventilation system. test.
| |
| 50.72(b)(2)(ii) i i
| |
| l l
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| 01293909. DOC 1A 54 071631956
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| LESSON: REPORTING REQUIREMENTS LESSON No.: LADM-3 10 CFR 50.72 Notification Requirements, continued l Four-hour The following seven (7) non-emergency events are reportable to the NRC, Notifications - via the' ENS, as soon as practical and in all cases, withinfour hours.
| |
| Non-emergency . Use examples of each non-emergency event from Attachment 8.6.
| |
| Events Obj 3, 4, 7, 8 Four-hour Notifications (Non-emergency Events)
| |
| Any event or condition that results in manual or automatic actuation of l E3F, including RPS (see Attachment 8.7), except when: j
| |
| . Actuation is part of a pre-planned sequence of events and:
| |
| + The actuation must be listed as " definitely expected" to occur in a work order, procedure or control room log.
| |
| + The acication is an intended result of the procedure.
| |
| . Actuation is invalid and:
| |
| + Occurs while system is properly removed from service,
| |
| + Occurs after the safety function has been completed, or l + Involves L following ventilation systems: Control Room, Containment, FHB, or Auxiliary Building. l Any event or condition that alone could have prevented the fulfillment of l the safety function of structures or systems that are needed to:
| |
| . Establish and maintain safe shutdown,
| |
| . Remove decay heat,
| |
| . Control the release of radioactive materials, or
| |
| . Mitigate the consequences of an accident.
| |
| Any event or conditionfound while shutdown that, had it beenfound while operating, would have resulted in the plant or safety barriers being:
| |
| . Seriously degraded.
| |
| . In an unanalyzed condition that significantly compromises safety.
| |
| Airborne radioactivity released to an unrestricted area exceeds 20 times the limits of 10 CFR 20, if averaged over 1 hour (also classifiable as a NUE).
| |
| Liquid effluent released to an unrestricted area exceeds 20 times the limits of 10 CFR 20 for all radionuclides except tritium and dissolved noble gasses, if averaged over 1 hour (also classifiable as a NUE).
| |
| Any event requiring the transport of a radiologically contaminated person to an offsite medical facility fc: treatment.
| |
| Continued on nextpage LADM3. Doc . PAGE 11 oF 24 REV 6
| |
| | |
| l l
| |
| m ''
| |
| s JAf]4Y U &
| |
| 1
| |
| | |
| U. S. Nuclear Regulatory Commission Site-Specific Written Examination Applicant Information 1 Name: Region: IV i Date: Facility / Unit: Di5blo Canyon 1 & 2 License Level: SRO ReactorType: Westinghouse Start Time: Finish Time: l I
| |
| Instructions i I
| |
| i Use the answer sheets provided to document your answers. Staple i this cover sheet on top of the answer sheets. The passing grade I requires a final grade of 80.00 percent. Examination papers will be collected four hours after the examination starts.
| |
| Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
| |
| Applicants Signature )
| |
| Results Examination Value 100 Points Applicant's Score Points A'pplicant's Grade Percent
| |
| | |
| I Appendix E NRC Policies and Guidelines for Taking NRC Examinations PART A - GENERAL GUIDELINES
| |
| : 1. Cheating on any part the examination will result in a denial of your ,
| |
| application and/or action against your license. !
| |
| : 2. If you have any questions conceming the administration of any part of the
| |
| - examination, do not hesitate asking them before starting that part of the exam.
| |
| : 3. SRO applicants will be tested at the level of responsibility of the senior licensed shift position (i.e., shift supervisor, senior shift supervisor, or whatever the title of the position may be).
| |
| : 4. You must pass every part of the examination to receive a license or to continue performing license duties. Applicants for an SDO-upgrade license may require remedial training in order to continue their RO duties if the examination reveals deficiencies in the required knowledge and abilities.
| |
| : 5. The NRC examiner is not allowed to reveal the results of any part of the examination until they have been reviewed and approved by NRC management. Grades provided by the facility licensee are preliminary until approved by the NRC. You will be informed of the official examination results about 30 days after all the examinations are complete.
| |
| PART B - WRITTEN EXAMINATION GUIDELINES
| |
| : 1. After You complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received er given assistance in completing this examination.
| |
| : 2. To pass the examination, you must achieve a grade of 80.00 percent or greater. Every question is worth one point.
| |
| : 3. For an initial examination the time limit for completion of this examination is four hours.
| |
| : 4. You may bring pens and calculators into the examination room. Use only dark ink to ensure legible copies.
| |
| : 5. Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identification.
| |
| : 6. Mark your answers on the answer sheet provided and do not leave any i question blank. Use only the paper provided and do not write on the back sides of pages. If you decide to change your original answer. draw a single ,
| |
| line through the error, enter the desired answer, and initial the change.
| |
| | |
| p Appendix E PART B -WRITTEN EXAMINATION GUIDELINES
| |
| : 7. If the intent of a question is unclear, ask questions of the examiner or facility the designated facility instructor only.
| |
| : 8. Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or p'ossibility of cheating.
| |
| : 9. When you complete the examination, assemble a package including the examination questions., examination aids, answer sheets, and scrap paper and give it the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.
| |
| : 10. After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner, if you are found in this area while the examination is still in progress, your license may be denied or revoked.
| |
| l
| |
| : 11. Do you have any Questions?
| |
| l 1
| |
| I i
| |
| | |
| ANSWER SHEET FOR SENIOR REACTOR OPERATOR WRITTEN EXAMINATION Name:
| |
| : 1. @@@@ 18. @@@@ 35. @@@@ l l
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| c ,
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| ANSWER SHEET FOR SENIOR REACTOR OPERATOR l WRITTEN EXAMINATION Name:
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| 2
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| 1 EXAMINATION Senior Reactor Operator Written Exam Question: 001 As reactor thermal power is increased, the rod insertion limits (RILs) are required to be progressively higher. From WHICH ONE (1) of the following parameters is the control bank low alarm determined?
| |
| : a. auctioneered high Tavg
| |
| : b. auctioneered high Tref
| |
| : c. auctioneered high Ni power level
| |
| : d. auctioneered high delta T i
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| Page: 1 I
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| EXAMINAT!ON Senior Reactor Operator Written Exam i
| |
| I Question: 002 )
| |
| The following initial conditions exist
| |
| )
| |
| Ta vg -Tref. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 0 F PT 5 0 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 %
| |
| Non-Linear Gain Unit output (T/M test) .. 2 F Rod Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . M a n u al Control Room / Electrical Team is testing the auto rod control system, simulating a signal that reactor power has just decreased while turbine power power remained constant.
| |
| Us.ng the drawing provided, state the speed rods would move out if AUTO is inadvertently selected?
| |
| : a. 8 steps / min
| |
| : b. 40 steps / min
| |
| : c. 48 steps / min
| |
| : d. 72 steps / min f
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| t Page: 2
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| % ~
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| 1 EXAMINATION Senior Reactor Operator Written Exam Question: 003 1
| |
| Given the following:
| |
| . The unit is at 30% power. ,
| |
| . RCP 1-2 trips. I
| |
| . NO operator action is taken.
| |
| WHICH ONE (1) of the following describes the INITIAL unit response to the RCP trip? i
| |
| : a. A reactor trip will NOT occur and S/G 1-2 water level will INCREASE.
| |
| : b. A reactor trip will NOT occur and S/G 1-2 water level will DECREASE.
| |
| I
| |
| : c. A reactor trip WILL occur and S/G 1-2 water level will 1.NCREASE.
| |
| : d. A reactor trip WILL cccur and S/G 1-2 water level will DECREASE.
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| Page: 3
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| l t
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| l EXAMINATION !
| |
| Senior Reactor Operator Written Exam Question: 004 Given the following:
| |
| . RCS Pressure - 335 psig
| |
| . RCS Ter.iperature - 340 F
| |
| . Steam Generator Pressures -150 psig
| |
| . A bubble exist in the Pressurizer !
| |
| . No RCPs are running WHICH ONE (1) of the following correctly predicts the response of RCS temperature and pressure if a Reactor Coolant Pump is started?
| |
| RCS Temperature RCS Pressure
| |
| : a. increases increases
| |
| : b. Increases Decreases
| |
| : c. Decreases increases
| |
| : d. Decreases Decrear ;
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| F Page: 4
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| | |
| r EXAMINATION Senior Reactor Operator Written Exam
| |
| ]
| |
| Question: 005 Given the following Unit 1 plant conditions;
| |
| * Emergency boration is initiated to restore adequate SDM.
| |
| . The preferred Emergency Boration flowpath alignment results in 0 gpm flow due to a clogged Boric acid tilter.
| |
| To successfully initiate emergency boration WHICH ONE (1) of the following methods should be used?
| |
| : a. Swap charging pump suction to RWST.
| |
| : b. Open CVCS-8104, Emergency Borate Valve.
| |
| : c. Locally Open CVCS-8471, Manual Emergency Borate Valve.
| |
| : d. Locally open CVCS-8476, Boric Acid Transfer pump crosstie.
| |
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| l Page: 5
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| f EXAMINATION Senior Reactor Operator Written Exam Question: 006 Given the following:
| |
| . Unit 2 is at 100% power
| |
| . STP M-9A is being conducted on D/G 2-1
| |
| . D/G 2-1 is paralleled to 4Kv Bus G and loaded to 2560 Kw e A spurious Si actuation occurs With no operator actions what would be the status of the 4Kv buses?
| |
| : a. Buses F, G & H energized by their respective D/Gs.
| |
| : b. Buses F & H energized from startup power and Bus G is denergized.
| |
| : c. Buses F, G & H energized from startup power.
| |
| : d. Buses F & H energized from startup power and Bus G is energized by D/G 2-1.
| |
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| EXAMINATION Senior Reactor Operator Written Exam Question: 007 Unit 1 is in Mode 6, with refueling in progress. The Control Room Asset Team has been instructed to reduce the Hi Flux Trip Setpoints for all Power Ranges (PR) to 25%
| |
| for Physics Testing. Permission is requested to work two (2) power range Nis simultaneously to save time.
| |
| WHICH ONE (1) of the following describes what effect if any, this could have on the current status of the plant?
| |
| : a. No effect, PR channels are not needed in this mode.
| |
| : b. One source range channel will be de-energized.
| |
| : c. Both source range channels will be de-energized.
| |
| : d. Subcooled Margin Monitor LoLo alarm wiil be disabled.
| |
| I
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| \
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| l Page: 7
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| | |
| r EXAMINATION Senior Reactor Operator Written Exam Question: 008 Unit 1 is at 90% power when a DRPI General Warning occurs on rod H-8. The rod bottom light for rod H-8 is NOT on.
| |
| WHICH ONE (1) of the following describes the operability of DRPI and the accuracy of determining the position of rod H-87
| |
| : a. inoperable; no measurable accuracy
| |
| : b. inoperable; accuracy is .!-4, +/-10
| |
| : c. operable; accuracy is +/-4, +/-10
| |
| : d. operable; accuracy is +/-12 1
| |
| 1 i
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| Page: 8
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 009 A reactor startup is in progress with power at 5.0E-11 amps. The Source Range High Flux Trip has not been blocked.
| |
| WHICH ONE (1) of the following describes Reactor Protection System response if an INSTRUMENT POWER fuse fails open on N-31 with the Source Range Level Bypass Switch in the positions indicated?
| |
| Switch in NORMAL Switch in BYPASS
| |
| : a. No Trip No Trip
| |
| : b. Reactor Trip No Trip
| |
| : c. No Trip Reactor Trip
| |
| : d. Reactor Trip Reactor Trip Page: 9
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 010 Unit 2 is operating at 100% power with the incore Temperature Monitoring display for the hottest in-core thermocouple reading 622 F. Temperature in the area of the Reference Junction boxes for the thermocouples rises 15 F over the duration of the shift. Reactor power level remains constant over the duration of the shift. I I
| |
| WHICH ONE of the following correctly describes how core exit thermocouple (incore l Temperature Monitor) readings are affected by the temperature change in the area of j the of the reference junction boxes? l
| |
| : a. Will read higher due to higher voltage differential between the metals at the cold junction.
| |
| : b. Will read lower due to lowered voltage differential between the metals at the hot junction,
| |
| : c. Will remain the same because the temperature error is compensated for by the incore Temperature Monitor Processor system.
| |
| : d. Will remain the same since temperature change does NOT affect the signal from the reference junction.
| |
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| Page: 10 c.
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 011 Given the following:
| |
| . The plant tripped from 100% power due to a Main Generator / Grid Fault; offsite power is lost.
| |
| . ALL EDGs started normally.
| |
| . EOP E-0.2, " Natural Circulation Cooldown", is being implemented.
| |
| . During the step to verify CRDM fans are all running, the BOP reports that the PPC indication shows that the CRDM fans are NOT running.
| |
| WHICH ONE (1) of the following is a reason for the lack of CRDM fan indications?
| |
| : a. The Redundant (containment penetration protection) supply breaker needs to be reclosed following the momentary loss of power.
| |
| : b. The momentary loss of power to 4kV buses F, G, and H has deenergized the MCCs supplying the CRDM fans.
| |
| : c. The CRDM fan MCCs are deenergized until non-vital 4kV buses D and E are re-energized to supply their respective MCCs.
| |
| : d. The CRDM fans controllers require the fans be restarted manually following the transfer of power to the EDGs.
| |
| 4 Page: 11 i i'
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 012 '
| |
| Given the following:
| |
| . Unit 1 is operating at 75% power.
| |
| . D/G 1-3 cleared for maintenance, when a loss of offsite power and safety injection occurs.
| |
| . The CFCU Speed Selector switch positions are as follows:
| |
| * CFCU 1-1,1-3 & 1-4 running in HIGH speed
| |
| * CFCU 1-2 & 1-5 are OFF
| |
| * All CFCU control switches are selected to LOW WHICH ONE (1) of the following is the status of the CFCUs based on the given conditions?
| |
| : a. CFCUs 1-3,1-4 and 1-5 running in slow and started by the ESF Load Sequencing timer, CFCUs 1-1 and 1-2 are not running.
| |
| : b. CFCUs 1-1 thru 1-5 are all running in slow started by the Auto Transfer Timer.
| |
| l
| |
| : c. CFCUs 1-1 and 1-2 are running in slow and started by the ESF Load l Sequencing Timer, CFCU1-4 is running in high and started by the Auto ,
| |
| Transfer Timer, and CFCUs 1-3 and 1-5 are not running.
| |
| : d. CFCUs 1-1,1-2 and 1-4 are running in slow and started by the Auto Transfer Timer, and CFCUs 1-3 and 1-5 are not running.
| |
| Page: 12
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam 4
| |
| Question: 013 l
| |
| Given the following: !
| |
| . The plant is operating at 100% power.
| |
| . Containment Spray pump 1-1 is isolated from the spray header and is running in recirculation back to the RWST for surveillance testing.
| |
| . A Main Steam Line break inside containment occurs resulting in a containment pressure of 23 psig. -
| |
| WHICH ONE (1) of the following describes the expected status of the containment spray system?
| |
| : a. CSP 1-1 continues to run in recirc; CSP 1-2 STARTS; ONLY the spray additive tank outlet valves OPEN.
| |
| : b. CSP 1-1 is tripped and then sequences back ON along with CSP 1-2; both CSP discharge valves and spray additive tank outlet valves OPEN.
| |
| : c. CSP 1-1 is tripped; CSP 1-2 STARTS; CS additive tank outlet valves and CSP 1-2 pump discharge valves are OPENED.
| |
| : d. CSP 1-1 continues to run in recirc; CSP 1-2 STARTS; both CSP discharge valves and spray additive tank outlet valves OPEN.
| |
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| Page: 13
| |
| | |
| EXAMINATION -
| |
| )
| |
| Senior Reactor Operator Written Exam Question: 014 Unit 1 is operating with the following conditions:
| |
| . Reactor power 100%
| |
| . 2 Condensate pumps running
| |
| . 2 Circulating Water pumps are running in this situation, WHICH ONE (1) of the following conditions would result in a trip of a Main Feedwater Pump, assuming no operator actions are taken?
| |
| : a. A spurious Reactor Trip.
| |
| : b. A Condensate pump trips on overcurrent, STBY pump starts.
| |
| : c. A Steam Generator Level channel fails low.
| |
| : d. A heater #2 drain tank pamp; trips on overcurrent.
| |
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| Page: 14
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 015 WHICH ONE (1) of the following correctly describes the basis for the Auxiliary Feedwater Technical Specification?
| |
| : a. Maintain the capability to cooldown the RCS to RHR initiation conditions following a complete loss of off-site power.
| |
| : b. Ensure the capability to cooldown and maintain the RCS at <500 F for 8 hours in the event of a complete loss of off-site power, assuming failed fuel,
| |
| : c. Remove decay heat and maintain the RCS at HSB conditions for 24 hours following a complete loss of off-site power.
| |
| : d. Provide the RCS heat removal capability necessary to prevent a challenge to the pressurizer safety valves during a full power ATWS, followed by a complete loss of off-site power.
| |
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| Page: 15
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 016 Diesel Generator 2-3 is shutdown and in automatic when PK18-04 " Diesel 23 Control Under Voltage" alarms. Investigation of the alarm reveals that the normal DC power supply broaker 72-2116 is tripped and will.not reclose.
| |
| What is the affect of this failure on D/G 237
| |
| : a. Will start and load on to the 4Kv bus if the D/G is started in manual from the control room,
| |
| : b. Will not start in auto or manually from the control room, but can be started locally if the Appendix R fuses are selected to backup position.
| |
| : c. Will not start in auto or manually and load onto the 4Kv bus until the DC control power supply is swapped to the backup DC source.
| |
| : d. Will start in auto or manually, but no field flash will be available, D/G will not load onto 4Kv Bus.
| |
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| Page: 16 L
| |
| | |
| -c - - _
| |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 017 A Safety injection was caused by SSPS testing. Plant recovery actions are being taken and the SI signal has been reset.
| |
| WHICH ONE (1) of the following actions will restore the Reactor Coolant Drain Tank l (RCDT) Pump discharge flowpath (FCV-253 and 254) to the Liquid Radwaste (LRW) l system.
| |
| : a. Reset Phase A and reset LRW isolation,
| |
| : b. Reset Phase A, Reset LRW lsolation, Open FCV-253 and 254 from the local control panel.
| |
| Reset Phase A, Open FCV-253 and 254 from the local panel.
| |
| c.
| |
| : d. Reset LRW isolation, open FCV-253 and 254 from the local panel.
| |
| }
| |
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| Page: 17
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 018 The following Radiation Monitor indications exist:
| |
| . RM-14/87 - NR/RM i 4 status light OFF and ER/RM87 status light ON
| |
| . RM-29: Warn and High Alarm lights are ON
| |
| . All other radiation monitors indicate normal reading Based on the above indications, what accident has occurred?
| |
| : a. Gas Decay Tark Rupture j
| |
| : b. S/G Tube Leak
| |
| : c. Main Steam Line Break, downstream of the MSIVs
| |
| : d. Incore Seal Table Leak I
| |
| i 1
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| i Page: 18
| |
| | |
| l EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 019 WHICH ONE (1) of the following will cause the Fuel Handling Building Ventilation System to automatically swap to the lodine Removal Mode?
| |
| : a. RM 58 ALERT alarm actuates with RM 58 Bypass switch in NORMAL.
| |
| : b. RM 59 HIGH alarm actuates, with RM 59 Bypass switch in BYPASS.
| |
| : c. RM 58 HIGH ala m actuates, with RM 58 Bypass switch in NORMAL.
| |
| l
| |
| : d. RM 59 ALERT alarm actuates with RM 59 Bypass switch in BYPASS.
| |
| )
| |
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| i Page: 19
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 020 WHICH ONE (1) of the following components is credited in the FSAR as being designed to mitigate the pressure rise in the RCS following a turbine trip from 100%
| |
| power without a reactor trip?
| |
| : a. PZR Steam Space Volume
| |
| : b. PZR PORVs
| |
| : c. PZR Spray Valves
| |
| : d. PZR Code Safety Valv. is.
| |
| l
| |
| ~'
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| Page: 20
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator W (tten Exam Question: 021 Unit 1 has been operating at 100% power for an extended period of time . Twelve hours ago, residual heat removal (RHR) heat exchanger 1-1 was declared inoperable to perform maintenance on HCV-638. The. Unit 1 cultrol operator has just reported ,
| |
| that Centrifugal Charging pump 1-1 has tripped on overcurrent.
| |
| WHICH ONE (1) of the following describes the allowances and/or limitations im?csed by the Technical Specifications for placing Unit 1 in Hot Standby?
| |
| : a. Be in Hot Standby in 6 hours
| |
| : b. Be !n Het Standby in 7 hours
| |
| : c. Be in Hot Standby in 60 hours
| |
| : d. Be in Hot Standby in 66 hours l
| |
| \
| |
| Page: 21 I i
| |
| t.
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 022 Given the following:
| |
| . Unit 1 is in MODE 4.
| |
| . The Low Setpoint Protection Switches for Power Operated Relief Valves (PORVs) PCV-455C and PCV-456 are in the CUTIN position.
| |
| WHICH ONE (1) of the following plant conditions will result in the LTOP arming and actuation of the primary PORVs?
| |
| Wide Range Wide Range RCS Pressure Cold Lea RCS Tem _p
| |
| : a. 428 psig 250 F
| |
| : b. 456 psig 265 F
| |
| : c. 439 psig 283 F
| |
| : d. 462 psig 315 F l
| |
| 1 Page: 22
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam i
| |
| Question: 023' The following plant conditions exist on Unit 2:
| |
| * Reactor Power 50% and stable
| |
| . Rod Controlin AUTO
| |
| . All PZR controls in AUTO e PZR Master Pressure Controller (PC-455K) output at 25%
| |
| A loss of power to the PZR proportional heaters causes PZR pressure to decrease at a rate of 1 psig/ min.
| |
| How much time will elapse before the PZR Backup heaters turn on?
| |
| : a. 10 min.
| |
| : b. 15 min.
| |
| : c. 20 min.
| |
| : d. 25 min.
| |
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| (
| |
| l Page: 23
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 024 Given the following: )
| |
| . A large break LOCA occurred 10 minutes ago.
| |
| . Subcooling margin is MINUS 20 F e All ECCS equipment is operating properly
| |
| . Pressurizer level has just increased rapidly from off-scale low to 50%
| |
| WHICH ONE (1) of the following is the cause of this rapid increase in pressurizer level?
| |
| : a. A pressurizer vapor space leak has developed,
| |
| : b. Voiding is occurring in the reactor vessel head.
| |
| : c. Si Flow is refilling the pressurizer.
| |
| : d. Temperature variations on the reference leg of the pressurizer level instrument.
| |
| 1 l
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| Page: 24 j
| |
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| 3 l
| |
| EXAMINATION '
| |
| i Senior Reactor Operator Written Exam Question: 025 The following plant conditions exist:
| |
| . Unit 1 is at 100% power
| |
| . Pressurizer level control U460A is selected for L1459/LT 461
| |
| . Pressurizer level channel LT 459 has just failed low WHICH ONE (1) of the following describes the plant response to the pressurizer level.
| |
| channel failure? ASSUME no operator action is taken.
| |
| : a. Letdown isolation occurs, VCT level increases, Charging flow decreases, and actual PZR level decreases.
| |
| : b. Charging flow remains the same, Letdown isolation does not occur, backup heaters turn on.
| |
| : c. Letdown isolation occurs, Charging flow increaces, actual PZR level increases and Reactor Trips on High PZR level.
| |
| : d. Charging flow decreases, Letdown isolation occurs, heaters de-energize and Reactor Trips on Low Pressurizer pressure.
| |
| j Page: 25
| |
| | |
| I I
| |
| EX. AMINATION Senior Reactor Operator Written Exam Question: 026 !
| |
| Given the following conditions on Unit 1: j
| |
| . The plant is at 100% power
| |
| . Solid-state protection system (SSPS) testing is in progress
| |
| . Reactor trip breaker "B" (RTd) is OPEN.
| |
| . Bypass breaker "G"(BYB) is CLOSED.
| |
| . Reactor trip breaker "A" (RTA) is CLOSED.
| |
| . Bypass breaker"A"(BYA) is OPEN.
| |
| WHICH ONE (1) of the following is the response of Breaker RTA immediately after Breaker BYA is manually taken to CLOSE?
| |
| : a. RTA shunt trip coil and undervoltage trip coil will be energized.
| |
| : b. RTA shunt trip coil and undervoltage trip coil will be deenergized.
| |
| : c. RTA shunt trip coil wiH be deenergized and RTA undervoltage trip coil will be energized.
| |
| : d. RTA shunt trip coil will be er'ergized and RTA undervoltage trip coil will be deenergized.
| |
| l Page: 26
| |
| | |
| I EXAMINATION Senior Reactor Operator Written Exam l
| |
| Question: 027 Given the following:
| |
| . Unit 1 is at 9% power diaring a plant startup.
| |
| * Control systems are aligned normally. '
| |
| - WHICH ONE (1) of the following provides the feed forward signal to the steam generator level control system?
| |
| : a. Wide range steam generator level
| |
| : b. Auctioneered high power range nuclear power
| |
| : c. Steam flow /Fced flow mismatch ,
| |
| : d. Derivative of the arbitrated narrow range steam generator levels I
| |
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| Page: 27
| |
| | |
| ~
| |
| f EXAMINATION l Senior Reactor Operator Written Exam l
| |
| Question: 028 WHICH ONE (1) of the following reasons correctly identifies why the Containment Vacuum /F'ressure Relief Isolation valves are limited to 50 degrees open during Containment venting operations?
| |
| : a. Ensure that Containment peak pressure does not exceed the design pressure during a LOCA 1
| |
| : b. The valves will close on a Containment Vent Isolation signal, but do not I automatically close on a Containment Phase "A" Isolation signal. l l
| |
| : c. The valves have not been demonstrated to be capable of closing dunng a steam break inside containment.
| |
| : d. The valves have not been demonstrated to be capable of closing during a LOCA inside containment.
| |
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| Page: 28
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| EXAMINATION Senior Reactor Operator Written Exam t
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| Question: 029 The following conditions exist:
| |
| . A LOCA has occurred e Safety inje:: tion has actuated e Phase B isolation has actuated
| |
| . 4 Ku buses are being powered from their respective diesels
| |
| . All equipment has had time to sequence on.
| |
| WHICH ONE (1) of the following is the expected response of the Spent Fuel Pool SFP cooling system without any operator actions?
| |
| : a. SFP pump 1-1 restarts, SFP temperature increases due to CCW flow isolation to the SFP heat exchanger.
| |
| : b. SFP pump 1-2 restarts, SFP temperature decreases due to increased CCW flow.
| |
| : c. Selected SFP pump restarts, SFP temperature decreases due to increased CCW flow.
| |
| : d. Neither SFP pump rest. arts, SFP temperature increases due to CCW flow isolation to the SFP heat exchanger.
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| i EXAMINATION -
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| Senior Reactor Operator Written Exam Question: 030 l
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| You are the SRO in Containment for Fuel Handling Operations with the following conditions:
| |
| . MODE 6 '
| |
| . Fuel handling in progress
| |
| . A fuel assembly is dropped back into the core during removal from the core.
| |
| . Bubbling is observed from the dropped assembly ,
| |
| After suspending Core alterations and ensuring the fuel assembly is in a safe position, your NEXT step / action per AP-21 " Irradiated Fuel Damage" should be to direct:
| |
| I
| |
| : a. the control room to establish Containment Closure.
| |
| : b. refueling crew members to get off the manipuletor crane.
| |
| : c. refueling crew members to determine extent of fuel damage. !
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| : d. refueling crew members to close SFS-50, Fuel Transfer tube gate valve.
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| p EXAMINATION Senior Reactor Operator Written Exam Question: 031 Given the following l
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| 2 The plant has been at 100% power for the past 200 days.
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| 1
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| * A Steam Line Break has just occurred.
| |
| WHICH ONE (1) of the following describes the effect on SHUTDOWN MARGIN (SDM)?
| |
| : a. SDM decreases due to the negative Isothermal Temperature Coef.'icient.
| |
| : b. SDM decreases due to the positive Moderator Pressure Coefficient.
| |
| : c. SDM increases due to the negative Moderator Temperature Coefficient.
| |
| : d. SDM increases due to the negative differential boron worth. j l
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| ' 1 EXAMINATION l Senior Reactor Operator Written Exam j l
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| Question: 032 Unit 1 is operating at 100% power with the following conditions:
| |
| * Normal electric distribution lineup l
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| e 4 Kv breaker inspection reveals that startup feeder breaker for Bus G (52-HG-14) is I inoperable.
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| What action (s) if any are required by Technical Specifications?
| |
| : a. Demonstrate the cperability of the remaining A. C. source to 4 KV bus G and the operability of Diesel Generator 1-2. ;
| |
| l
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| : b. Demonstrate the operability of all the remaining A. C. sources and the operability of all Diesel Generators.
| |
| : c. Demonstrate the operability of the remaining A. C. cources to 4 Kv Buses H &
| |
| F and the operability of Diesel Generators 1-1 and 1-3.
| |
| : d. Since startup power was not lost to the 230 Kv switchyard, no technical specifications apply to this situation.
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| EXAMINATION Senior Reactor Operator Written Exam Question: 033 WHICH ONE (1) of the following initially supplies the generator field current during the startup of Diesel Generator 1-17
| |
| : a. Diesel DC control power, supplied by 125 VDC bus 1-2.
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| : b. Diesel DC control power, supplied by 125 VDC bus 1-3.
| |
| : c. A permanent magnet generator (PMG).
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| l
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| : d. Diesel DC control power, supplied by 125 VDC bus 1-1.
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| EXAMINATION Senior Reactor Operator Written Exam )
| |
| l l
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| Question: 034 WHICH ONE (1) of the following describes what is provided by the " live zero" function of a check source? 4 It provides:
| |
| : a. a constant source fnr calibrating the discriminator circuit.
| |
| : b. a minimum indication below which a circuit failure is indicated.
| |
| : c. a minimum indication below which meter calibration is required. I
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| : d. a minimum indication when the " check source" function is used.
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| Page: 34
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| EXAMINATION Senior Reactor Operator Written Exam Question: 035 Unit 1 plant heatup is in progress with RCS temperature at 420 F.
| |
| Fire system engineer reports to the Shift Foreman that the fire door between the centrifugal charging pumps and the positive displacement charging pump is noi,-
| |
| functional. Additiona"y the engineer reports that no other fire system impairments ;
| |
| exist.
| |
| Which ONE (1) of the following actions is required?
| |
| I
| |
| : a. Enter Tech Spec 3.0.3 due to no operable charging pumps.
| |
| : b. Establish a continuous fire watch on at least one side of the fire door within 15 minutes,
| |
| : c. Establish an hourly fire watch patrol with in 1 hour.
| |
| : d. Establish continuous fire watches on both sides of the door within 1 hour.
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| Page: 35
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| I EXAMINATION Senior Reactor Operator W" <1 Exam Ouestion: .036 l l
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| Given the following conditions on Unit 1:
| |
| . Unit le in Mode 1
| |
| . Containment temperatures indications are: l (1 TE-85 = 118'F u TE-87 = 119'F f
| |
| a TE-89 = 121*F u TE-91 = 12'3*F
| |
| . ' PK01-16 Containment Environment is in alarm Actions to reduce containment internal temperature below the Tech Spec limit are required due to:
| |
| : a. 2 out of 4 TE's being greater than 120 F.
| |
| : b. Auctioneered Hi temperature greater than 120 F.
| |
| : c. Average temperature greater than 120 F.
| |
| : d. selected TE at Panel 199 greater than 120 F. ,
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| EXAMINATION Senior Reactor Operator Written Exam l
| |
| Question: 037 -
| |
| Unit 1 is operating at 100%. A seismic event occurs causing the following:
| |
| * Reactor Trip
| |
| . Safety injection l
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| . Large Break LOCA
| |
| . 4kv Bus F differentiallockont l WHICH ONE (1) of the following actions must be performed in EOP E-0, " Reactor Trip '
| |
| or Safety injection", in regards to the CCW/ASW system?
| |
| : a. Place ctandby Component Cooling Water heat exchanger inservice, j
| |
| : b. Start the standby Component Cooiing Water pump.
| |
| : c. Cross-tie Auxiliary Saltwater with Unit 2.
| |
| : d. Isolate the non-vital Component Cooling Water header.
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| EXAMINATION Senior Reactor Operator Written Exam Question: 038 Unit 1 is in Mode 3 , Hot Shutdown with the following conditions:
| |
| . Cooldown to Mode 4 was stopped until after shift turnover is complete.
| |
| . Steam Dump Selector Switch is in Steam Pressure Mode Bypass interlock selected on both selector switches
| |
| )
| |
| . HC-507 is in automatic with pot set to control RCS temperature at 500 F I
| |
| . Circ Water Pump 1-1 inservice
| |
| . Condenser back pressure is 1"Hg. ABS What would be the steam dump response to Main Steam Header pressure transmitter PT-507 failing high? (Assume No Operator Action) i
| |
| : a. Groups 1 & 2 would fully open.
| |
| 1
| |
| : b. Steam dumps would not respond due to P-12 interlock. >
| |
| : c. Group 1 would modulate open to maintain RCS temperature at 500 F.
| |
| : d. Group 1 would fully open.
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| Page: 38
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 039 Given the following:
| |
| . Stator Cooling Water Conductivity is above its normal range at 4 S/cm; the HIGH CONDUCTIVITY alarm is illuminated. l
| |
| . A problem with a Stator Cooling Water flow switch (FS1) has caused it to indicate a low flow condition.
| |
| . The unit initiates a runback to 15% of rated amps.
| |
| What is the MAXIMUM time the unit will operate with NO further operator action?
| |
| (Time frame is from time of stator cooling water flow switch trip until unit trip).
| |
| : a. 5 seconds
| |
| : b. 45 seconds
| |
| : c. 3 minutes
| |
| : d. 60 minutes l
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| Page.39
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| EXAMINATION '
| |
| ' Senior Reactor Operator Written Exam
| |
| =_
| |
| Question: 040 Given the following:
| |
| e The Unit is in Mode 5 for a refueling outage.
| |
| . RHR is in service with HCV-637 (RHR Hx No.2 Outlet Flow Control) & 638 (RHR Hx No.1 Outlet Flow Control) at 10% OPEN and HCV-670 (RHR Hx Bypass) at 75% OPEN.
| |
| i e Malfunctions in the Control Air systems have caused the RHR valves to go to their FAILED position.
| |
| WHICH ONE (1) of the following describes how the RHR system and RCS will respond?
| |
| : a. HCV-637, HCV-638, & HCV-670 will all fail OPEN and RCS temperature will DECREASE.
| |
| : b. HCV-637, HCV-638 will fail OPEN; HCV-670 will fail CLOSED, RCS temperature will DECREASE.
| |
| : c. HCV-637, HCV-638 will fail CLOSED; HCV-670 will fail OPEN, RCS g temperature will INCREASE.
| |
| : d. HCV-637, HCV-638, & HCV-670 will all fail CLOSED and RCS temperature will INCREASE. ;
| |
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| Page: 40
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| - mm mm memummmi mu mi in n iii .
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| mi isimummimumm ssi mm aus
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| EXAMINATION Senior Reactor Operator Written Exam Question: 041 Given the following;
| |
| . Unit 1 is at 50% power with Rod Control in Automatic
| |
| . Failure in the Rod Control System caused control rods to step out 10 steps.
| |
| * Rod control was placed in manual e RCS beration was used to return RCS temperature to program WHICH ONE (1) of the following conditions h'ss resulted from the rod withdrawl event?
| |
| : a. The Rod Insertion Limit (RIL) has increased.
| |
| : b. DNBR has increased.
| |
| 1
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| : c. Available shutdown margin has increased.
| |
| : d. Quadrant power tilt ratio (QPTR) has increased.
| |
| 1 Page: 41
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 042 j
| |
| (' Given the following on Unit 1:
| |
| . Unit is ramping to 100%
| |
| . Reactor power is at 90% with Control rods in automatic
| |
| . A Control Bank D group 1 rod drops into the core without causing a reactor trip; no trip is required.
| |
| . Operators have implemented OP AP-120, " Dropped Control Rod".
| |
| WHICH ONE (1) of the following describes the required actions to establish initial recovery conditions 20 minutes following the dropped rod even+7
| |
| : a. No action is required.
| |
| : b. Reduce turbine load to reduce Reactor power to ~85%.
| |
| : c. Reduce turbine load to reduce Reactor power to less than 50%.
| |
| : d. Set Tavg 1.5 F above Tref by withdrawing control bank rods as necessary, i
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| I Page: 42
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 043 Unit 1 is operating at 100% power when Main Turbine trip occurs. While performing i the actions of E-0.1, " Reactor Trip Response," the control operator notices the following control rod status;
| |
| * M8, no rod bottom light, no rod position indication light
| |
| . H4, rod position indication light indicating 8 steps
| |
| . H12, rod position indication light indicating 6 steps What action is required?
| |
| : a. No actions required
| |
| : b. Emergency Borate 900 gallens l I
| |
| : c. Emergency Borate 1800 gallons
| |
| : d. Emergency Borate 2700 gallons !
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 044 EOP E-1, " Loss of Reactor or Secondary Coolant", Step 13, " Check if Transfer to Cold Leg Recirculation is Required," does not allow continuation of the procedure until Refuel Water Stage Tank (RWST) level is.less than 33%.
| |
| WHICH ONE (1) of the following describes the basis for this delay?
| |
| : a. To ensure most of the boric acid available in the RWST has been flushed through the core.
| |
| : b. To ensure there is sufficient water in the containment recirculation sump to provide adequate suction head for the ECCS pumps.
| |
| : c. To ensure most of the water available in the RWST has been used for core cooling.
| |
| l
| |
| : d. To ensure sufficient boric acid and sodium hydroxide mixing to rnintain the proper pH of water in the containment recirculation sump.
| |
| Page: 44
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam 1
| |
| l Question: 045 A Loss Off Coolant Accident (LOCA) outside containment has resulted in RCS Subcooling dropping to 0 F.
| |
| 1 When attempting to determine if a LOCA outside containment has been isolated, l WHICH ONE (1) of the following is the PRIMARY indication that the completed actions ;
| |
| have been successful?
| |
| : a. ECCS flow decreasing
| |
| : b. Containment Recirculation Sump Levelincreasing
| |
| : c. Pressurizer levelincreasing
| |
| : d. RCS pressure increasing l
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| Page: 45
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| | |
| EXAMINATION
| |
| - 1 Senior Reactor Operator Written Exam Question: 046 I Given the following conditions:
| |
| 1
| |
| . Unit 2 has experienced a Safety injection
| |
| . EOP E-0, " Reactor Trip or Safety injection", is in progress I
| |
| . RM-10, Aux Bldg Control Board Radiation Monitor is in alarm
| |
| . WR RCS Pressure is slowly decreasing l l
| |
| WHICH ONE (1) of the following is the correct procedure transition for the above l conditions:
| |
| : a. Continue in E-0, " Reactor Trip or Safety injection and REFER to OP AP-17,
| |
| " Charging Line Break."
| |
| : b. GO TO E-1 " Loss of Reactor or Secondary Coolant."
| |
| : c. GO TO ECA-1.2, "LOCA Outside Containment"
| |
| : d. GO TO EOP E-1.2, " Post LOCA Cooldown and Depressurization. )
| |
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| Page: 46
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| EXAMINATION Senior Reactor Operator Written Exam Question: 047 Assume that RCS Pressure is stable at 1700 psig following a reactor trip with a Safety injection.
| |
| WHICH ONE (1) of the following describes the effect on subcooling when the first safety injection pump is secured?
| |
| : a. Subcooling will remain constant because pressure is above the shutoff hend ,
| |
| for the S1 Pump. !
| |
| : b. Subcooling will decrease because RCS pressure will drop below the shutoff head for the S1 Pump.
| |
| : c. Subcooling will decrease because Si flow will be reduced to the output of the I remaining S1 Pump.
| |
| : d. Subcooling will remain constant because charging pump flow will increase to maintain RCS pressure.
| |
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| Page: 47
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 048 With Unit 1 in Mode 1 at 100% power:
| |
| WHICH ONE (1) of the following conditions would require the tripping of all Reactor Coolant Pumps (RCPs)
| |
| : a. Inadvertent phase B train A actuation,
| |
| : b. Degraded CCW flow causes # 1 seal leakoff temperatures to indicate 225*F on all RCPs
| |
| : c. RCS pressure decreases to 1275 psig during cooldown steps in EOP E-3,
| |
| " Steam Generator Tube Rupture"
| |
| : d. Loss of all ch,4ging pumps.
| |
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| Page: 48
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 049 Crew is performing E-0.2, " Natural Circulation Cooldown", Step 15c which requires the crew to verify all CRDM fans are running. 1 What is the basis for this step? ,
| |
| : a. To prevent overheating and damage to the DRPI coils and CRDM coil windings.
| |
| : b. To allow cooldown of RCS to be done at an increased rate of 50 F per hour. l
| |
| : c. To allow a lower minimum subcooling requirement during the RCS cooldown and depressurization
| |
| : d. To eliminate PTS concerns in the reactor vessel head during the RCS cooldown and depressurization 1
| |
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| Page: 49
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 050 Unit 1 has the following conditions:
| |
| . Emergency Boration is required due to stuck rods
| |
| . Emergency Boration via normal makeup, results in 25 gpm flow with the Boric Acid pp in high speed.
| |
| . Emergency Boration is accomplished by using CVCS-8104 WHICH ONE (1) of the following describes the method us=3d to determine the total number of gallons of boric acid added?
| |
| : a. F1-113A (Emergency Boration Flow meter on Vertical Board 2) and the duration the valve is open.
| |
| : b. FR-110 (Boric Acid and Primary water flow recorder on Control Console 2).
| |
| : c. YlC-110 (Boric Acid Integrator on Control Console 2).
| |
| : d. XFIT-113 (Emergency Boration Flow Transmitter in Cable Spreading Room) anc the duration the valve is open.
| |
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| )
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 051 Given the following conditions on Unit 1:
| |
| . EOP ECA-0.0, " Loss of All Vital AC Power is in progress"
| |
| . Local operations are in progress to isolate the RCP seals What is the PRIMARY reason for isolating the RCP Thermal Barrier CCW Return isolation valve (FCV-357) ?
| |
| : a. Prevents subsequent steam binding in the CCW system and prepares the plant for recovery
| |
| : b. Prevents damage to the RCP shaft when CCW flow is re-established.
| |
| : c. Prevents damage to the RCP seal package when CCW flow is re-established
| |
| : d. Isolates a potential RCS leakage path in cace of a thermal barrier rupture.
| |
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| | |
| EXAMINATION Senior Reactor Operator Written Exam l Question: 052 Given the following:
| |
| . An ATWS has occurred on Unit 1 and EOP FR-S.1," Response to Nuclear Power Generation /ATWS", is in progress.
| |
| . While implementing step # 7, " Check if the Following Trips Have Occurred", a Safety injection occurs which results in all rods inserting.
| |
| WHICH ONE (1) of the following actions should the SFM perform?
| |
| : a. Immediately exit FR-S.1, and perform EP E-0, " Reactor Trip or Sl", since the reactor has tripped.
| |
| : b. Remain in FR-S.1, while concorrently verifying the immediate actions of EP E-0, " Reactor Trip or SI". I
| |
| {
| |
| : c. Transition to EP E-0, " Reactor Trip or SI", complete the immediate actions, and then return to FR-S.1. ;
| |
| : d. Return to Step 1 of FR-S.1, " Verify Reactor Tripped" and transition to EP E-0
| |
| " Reactor Trip or SI".
| |
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| EXAMINATION !
| |
| Senior Reactor Operator Written Exam Question: 053 Given the following:
| |
| . Operators are pedorming EOP ECA-2.1," Uncontrolled Depressurization of All Steam Generators."
| |
| l
| |
| . Cooldown rate is 200 F per hour. !
| |
| 1
| |
| . Steam Generator levels range from 1% to 3% NR level. I WHICH ONE (1) of the following is the reason for maintaining a MINIMUM of 25 gpm AFW flow to each steam generator in this condition? 1 i
| |
| : a. To provide AFW pump minimum recirc flow requirements.
| |
| l
| |
| : b. To prevent S/G tube dryout while ensuring the minimum detectable feed flow is maintained. 3 I
| |
| : c. To minimize the potential water hammer by maintaining a minimum flow l through the feed ring J tubes.
| |
| : d. To conserve CST inventory until the end of the blowdown phase.
| |
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| Page: 53
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| | |
| r, EXAMINATION Senior Reactor Operator Written Exam Question: 054 WHICH ONE (1) of the following ore ; tor actions will result in Pressurized Thermal Shock (PTS) conditions following a ciaam Generator Tube Rupture (STGR) concurrent with a loss of offsite power (LOOP)?
| |
| : a. Overcooling the RCS by 50*F when using the 10% steam dumps to cool down !
| |
| the RCS to establish subcooling.
| |
| : b. Allowing the Safety injection Accummutators to inject while cooling down and depressurizing l
| |
| : c. Delaying the termination of Safety injection after the termination criteria have been met.
| |
| : d. Allowing the ruptured Steam Generator pressure to INCREASE while cooling down the RCS to establish subcooling.
| |
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| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 055 Unit 2 is shutdown in Mode 3 in preparation for a refueling outage. The following ;
| |
| conditions existed just prior to a complete loss of condenser vacuum:
| |
| . RCS pressure is 2235 psig l
| |
| . All four RCP's running 1
| |
| . Tavg being maintained by the steam dumps at 547'F WHICH ONE (1) of the following temperatures should Tavg stabilize with NO operator action?
| |
| : a. 543 F l l
| |
| : b. 547 F '
| |
| : c. 551 F
| |
| : d. 554*F I
| |
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| Page: 55
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 056 WHICH ONE (1) of the following is the reason for the order that the valve positions are checked in step 3, " CHECK RCS is Isolated", of ECA-0.0, " Loss of All AC Power"?
| |
| : a. They are listed according to control board indication.
| |
| : b. Those most likely to have an RNO corrective action outside the control room are last.
| |
| : c. Those most likely to fail in a loss of AC are listed first.
| |
| : d. They consider capacity of outflow lines and potential for inventory loss.
| |
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| Page: 56 l
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| I
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam i
| |
| Question: 057 Given the following initial conditions:
| |
| . Unit 1 is at 100% power and ramping to 50% power for condenser tube cleaning.
| |
| How will the loss of the MANUAL power supply to HC-459D, " Pressurizer Master Level Controller' affect pressurizer level control during the ramp?
| |
| 4
| |
| : a. No actions should be necessary due to level decrease in PZR due to change in Tavg.
| |
| I
| |
| : b. Throttle open HCV-142, "RCP Seal Flow control valve" to provide 'PZR level decrease.
| |
| l c. Throttle charging pump discharge valve FCV-128 " CCP flow control valve" to support a PZR level decrease. I
| |
| : d. Adjust HC-459D controller output in anticipation of the PZR level decrease. l I
| |
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| Page: 57 L
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| f: ,
| |
| EXAMINATION Senior Reactor Operator Written Exam ,
| |
| 1 Question: 058 I How would the liquid radwaste system respond if RE-18, " Liquid Radwaste Rad l Monitor", were to come into alarm during a discharge of a Floor Drain Receiver? !
| |
| : a. RCV-18 closes and FCV-477 opens. Flow is directed to the Floor Drain Receiver that is selected for fill.
| |
| : b. RCV-18 closes and FCV-477 opens. Flow is directed to the Equipment Drain l Receiver selected for fill. I
| |
| : c. RCV-18 opens and FCV-477 closes. The tank that is on discharge will swap to recirculation.
| |
| : d. RCV-18 closes. The running Floor Drain Receiver pump will receive a trip signal.
| |
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| l-t Page: 58
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| | |
| EXAMINAT!ON Senior Reactor Operator Written Exam j Question: 059 Given the following:
| |
| i Unit 1 is at 100% power and Unit 2 is 100% power.
| |
| Unit 1 SCO is completing the Shift Watch List Minimum Shift Crew requirements have been met and a Licensed Responder is needed for the plant Fire Brigade.
| |
| WHICH ONE (1) of the following individuals can be the Licensed Responder for the plant Fire Brigade?
| |
| : a. Balance of Plant Control Operator l
| |
| : b. Work Control Shift Foreman
| |
| : c. Shift Supervisor
| |
| : d. Shift Technical Advisor I
| |
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| i Page: 59 I
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| | |
| p EXAMINATION Senior Reactor Operator Written Exam Question: 060 WHICH ONE(1) of the following sets of indications are available on the Hot Shutdown Panel?
| |
| : a. Auxiliary Feedwater Flow, Containment Pressure, Charging Header Flow
| |
| : b. Emergency Boration Flow, Raw Water Reservoir Level, Pressurizer Pressure I
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| : c. Pressurizer Level, Refueling Water Storage Tank Level, Volume Control Tank Level
| |
| : d. Letdown Flow, Charging Pump Discharge Pressure, Wide Range RCS Pressure i
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| 1 EXAMINATION '
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| 1 Senior Reactor Operator Written Exam I I
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| Question: 061 Operators at the Hot Shutdown Panel (HSDP) have been given the direction to maintain a constant actual Steam Generator (SG) level of 85% wide range.
| |
| Using OP AP -8B " Cmtrol Room inaccessibility - Hot to Cold Shutdown, Figure 2 ,
| |
| determine WHICH ONE (1) of the following indicates how indicated SG level at the HSDP must change as SG pressure decreases from 1000 psig to 200 psig.
| |
| : a. SG indicated level at the HSDP decreases to 63%
| |
| l
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| : b. SG indicated level at the HSDP remains constant at 63% j 1
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| l c. SG indicated level at the HSDP remains constant at 74% I i'
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| : d. SG indicated level at the HSDP increases to 74%
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| =
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| l EXAMINATION '
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| Senior Reactor Operator Written Exam "
| |
| Question: 062 Given the following conditions on Unit 1:
| |
| . . Surveillance test on containment pressure transmitter PT-934 is in progress using STP l-12-P934.
| |
| e PC934A " Containment Hi Pressure" bistable is in the Tripped position e PC934B " Containment Hi Hi Pressure" bistable is in the Bypassed position WHICH ONE (1) of the following reflects the new coincidence, based on the above channel status, required to cause a Containment HI pressure Safety injection and Containment Hi HI pressure Phase B logic,
| |
| : a. Containment Hi Pressure 2/3 and Containment Hi Hi pressure is 3/4
| |
| : b. Containment Hi Pressure 1/2 and Containment Hi HI pressure is 1/3
| |
| : c. Containment Hi Pressure 1/2 and )
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| Containment HI HI pressure is 2/3
| |
| : d. Containment Hi Pressure 1/2 and '
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| Containment Hi HI pressure is 2/4 l
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| 4 Page: 62
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 063 Given the following conditions on Unit 2:
| |
| * Reactor Trip and Safety injection Has occurred
| |
| . Degraded Core Cooling conditions exist due to several malfunctions in the ECCS system.
| |
| . Crew is responding to the event using FR-C.2, " Response to Degraded Core !
| |
| Cooling" .
| |
| . During Step 12 to depressurize all intact Steam Generators to 140 psig, a valid ,
| |
| RWST low level alarm occurs.
| |
| WHICH ONE (1) of the following is the proper procedure transition for this event?
| |
| : a. GO TO t .A-1.1 " Loss of Emergency Coolant Recirculation"."
| |
| 'b. GO TO EOP E-1.3, " Transfer to Cold Leg Recirculation", complete the steps in E-1.3 and return to FR-C.2
| |
| : c. GO TO EOP E-1, " Loss of Reactor or Secondary Coolant'
| |
| : d. Complete FR-C.2 and then GO TO EOP E-1.3, " Transfer to Cold Leg Recirculation" I l
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| l Page: 63
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 064 Unit 1 is operating at 100% power with steady state conditions. At 0800 hours on January 5*, Chemistry reports the following RCS DOSE EQUIVALENT l-131 sample results for the past 4 hours:
| |
| . 0400 0.75 microcuries/ gram
| |
| . 0500 2.15 microcuries/ gram
| |
| . 0600 45.6 microcuries/ gram
| |
| . 0700 80.0 microcuries/ gram Using Reactor Coolant System Technical Specifications 3.4.8, WHICH ONE (1) of the following is the action required based on the chemistry reports?
| |
| : a. Be in at least HOT STANDBY with Tavg less then 500'F by 1100.
| |
| : b. Restore the Dose Equivlant 1-131 within the limits by 0500 January 7*, or be in HOT STANDBY by 1100 en January 7*.
| |
| : c. Be in at least HOT STANDBY with Tavg less then 500 F by 1300.
| |
| : d. Restore the Dose Equivlant 1-131 within the limits by 0700 January 7*, or be in I
| |
| HOT STANDBY by 1300 on January 7*.
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| 1 Page: 64
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| t EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 065 A reactor trip has occurred from 100% power. During performance of EP E-0, " Reactor Trip or Sl" immediate actions, it is determined that Governor valve 3 and it's associated stop valve failed to close.
| |
| What is the concern if this malfunction is uncorrected?
| |
| : a. Over-heating of the last stage of blades on the low pressure turbines
| |
| : b. Over-speeding of the main turbine
| |
| : c. Rupture of LP turbine rupture disks
| |
| : d. An Uncontrolled cooldown of the RCS l
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| l Page: 65
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| E ,
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| l EXAMINATION Senior Reactor Operator Written Exam l l
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| Question: 066 WHICH ONE (1) of the following indicates that the Pressurizer Relief Tank (PRT) l rupture disk has ruptured following a pressurizer PORV failing OPEN?
| |
| : a. PRT temperature increasing
| |
| : b. Reliefline temperature decreasing
| |
| : c. PRT low level
| |
| : d. Pressurizer level decreasing l
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| Page: 66
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| EXAMINATION Senior Reactor Operator Written Exam Question: 067 The RCS is in MODE 6 when a large leak occurs.
| |
| The running RHR Pump is lost due to cavitation.
| |
| Which procedure will provide direction for the restoration of RHR7
| |
| : a. OP AP SD-2, Loss of RCS Inventory.
| |
| : b. OP AP AP-24, Shutdown LOCA.
| |
| : c. OP AP SD-5, Loss of Residual Heat Removal
| |
| : d. OP AP SD-0, Loss of, or inadequate Decay Heat Removal l
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| Page: 67 ,
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| | |
| I EXAMINATION Senior Reactor Operator Written Exam Question: 068 1 During performance of EOP E-1.2, " Post LOCA CcMdown and Depressurization,"it is I desirable to have only one Reactor Coolant Pump (RCP) running.
| |
| WHICH ONE (1) of the following describes the reasons that only one RCP is !
| |
| desirable?
| |
| : a. One RCP is desired for letdown flow and RCS heat transport to the Steam Gen 3rators. Additional RCPs would add unnecessary heat load.
| |
| : b. One RCP is desired for spray flow and RCS heat transport to the Steam Generators. Additional RCPs would add unnecessary heat load.
| |
| : c. One RCP is desired for spray flow and RCS mixing of boric acid. Additional RCPs would add unnecessary heat load.
| |
| : d. One RCP is desired for spray flow and RCS heat transport to the Steam Generato% Aiditional RCPs would cause Pressurizer level to decrease.
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| Page: 68 i ___ _ _ _ _ _ _ _
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| 1 i
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam a
| |
| Question: 069 While performing the actions for a Loss of Emergency Coolant Recirculation, a RED path condition is identified for the Containment Status Tree.
| |
| WHICH ONE (1) of the following reasons describes why the Containment Spray Pumps are operated within the guidelines of ECA-1.1, " Loss of Emergency Coolant Recirculation"instead of using the guidelines contained in FR-Z.1," Response to High Containment Pressure."
| |
| : a. Ensures that the maximum heat removal system capacity that is available is used to reduce the containment pressure.
| |
| : b. ECA-1.1 pump operating criteria is more restrictive, ensuring continuous containment spray system operation to reduce containment pressure.
| |
| : c. ECA-1.1 pump operating criteria is less restrictive, permitting reduced containment spray operation to conserve RWST inventory.
| |
| : d. Provide a more rapid means of verifying automatic actuation of the containment spray system.
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| I Page: 69
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| EXAMINATIO: '
| |
| Senior Reactor Operator Written Exam Question: 070 Unit 2 is cooling down for an outage. Wide Range RCS Temperature is 325 F. The Balance of Plant operator reports the following indications.
| |
| . Abnormal fluctuations in RHR flow in both trains.
| |
| . Both RHR pumps have 5-10 amp swings
| |
| . PRT levelis increasing
| |
| * Pressurizer Levelis decreasing What is the most likely source of the problem.
| |
| : a. RV-8707, RHR Pump Suction Line Relief Valve from RCS Loop 4 Hot Leg is stuck open
| |
| : b. RV-8856A, RHR HX 2-1 Discharge Relief Valve is stuck open
| |
| : c. RV-8708, RHR Pumps Discharge to RCS Loops 1 & 2 Relief Valve is stuck open ;
| |
| : d. FCV-641 A, RHR Pump 2-1 Recirculation Valve is stuck open.
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| I Page: 70
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam i
| |
| Question: 071 Unit 1 is operating at 100% power all controls in automatic. Pressurizer Pressure control is selected to PT-457 for control and PT-456 as the backup channel.
| |
| The following sequence of alarms and actions occur:
| |
| . PK05 PZR Pressure Low e Pressurizer Backup Heaters on
| |
| .- PKD4 Protect Channel Activated - PZR Press Lo 1/4
| |
| . PK05 PZR Hi Pressure WHICH ONE (1) of the following describes the affect of these alarms and indications on the plant if no operator actions are taken?
| |
| : a. PORVs will cycle to maintain pressure s 2335 psig.
| |
| : b. PORVs will open at 2335 psig and close at 2185 psig to maintain pressure,
| |
| : c. Pressurizer sprays will modulate to maintain pressure s 2310 psig.
| |
| : d. Hi Pressure Reactor trip will occur.
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| I Page: 71
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| )
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| 1 EXAMINATION '
| |
| Senior Reactor Operator Written Exam l
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| l l
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| Question: 072 l Given the following:
| |
| * Unit 2 is in Mode 2.
| |
| * A reactor startup is in progress.
| |
| . Source range channel N-31 has just failed low. ;
| |
| 1 WHICH ONE (1) of the following conditions require the suspension of positive reactivity j per Technical Specifications? j f
| |
| : a. Any time reactor power is less than 10" amps.
| |
| : b. Any time reactor power is less than the source range high flux trip setpoint. l
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| : c. Any time reactor power is less than the point of adding heat.
| |
| : d. Any time the reactor power is less than 104 amps.
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| I' age: 72
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| (
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 073 Unit one was operating at 100% power when a Steam Generator tube leak occurred.
| |
| A manual Safety injection was initiated due to a calculated leak rate of 200 gpm.
| |
| The conditions just prior to the Safety injection were:
| |
| . RCS pressure 2200 psig and decreasing.
| |
| . S/G pressures at 800 psig.
| |
| The conditions following the Safety Injection were:
| |
| . RCS pressure 1700 psig and decreasing.
| |
| . S/G pressures at 1000 psig.
| |
| Primary to secondary leakage following the trip is approximately:
| |
| : a. 67 gpm
| |
| : b. 100 gpm
| |
| : c. 141 gpm
| |
| : d. 200 gpm Page: 73 e
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 074 WHICH ONE (1) of the following indications would confirm that a Steam Generator Tube Rupture was occurring?
| |
| : a. Pressurizer level DECREASE witti affected S/G steam flow DECREASING as feed Sow DECREASES. j l
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| : b. Pressurizer pressure DECREASE with affected S/G steam flow GREATER '
| |
| than feed flow.
| |
| : c. Pressurizer level DECREASE with affected S/G steam flow EQUAL to feed flow. i 1
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| : d. Pressurizer pressure DECREASE with affected S/G steam flow LESS than l feed flow.
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| i Page: 74
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 075 Unit 1 has experienced a feedwater line break inside containment on Steam Generator 1-1 and a totalloss of feedwater. FR-H.1 has been entered and feed and bleed of the RCS has been initiated. Shortly after opening the PORVs, the Turbine Driven Auxiliary Feedwater pump is returned to service and a source of feedwater is available. The operators are directed to restore a steam generator for a heat sink per FR-H.1 with the following plant conditions:
| |
| Indication Loop 1 Looo2 Looo 3 Looo4 S/G WR level (%) 0 12 7 3 S/G pressure (psig) 0 650 675 645 RCS hot leg temp 545 551 552 548
| |
| . Containment pressure = 3.5 psig
| |
| . Core exit T/Cs are stable at an average value = 560 F Which steam generator should be fed first?
| |
| : a. S/G 1-1 I
| |
| : b. S/G 1-2
| |
| : c. S/G 1-3
| |
| : d. S/G 1-4 Page: 75
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 76 '
| |
| During a loss of heat sink from 100% power, the Steam Generatcr wide range level indicators decrease below 23% on all Steam Generators with RCS pressure starting to increase.
| |
| The operators initiate bleed and feed by initiating Safety injection and opening the PORVs, but only one PORV opens. All attempts to open the other two PORVs fail.
| |
| If the plant remains in this condition without any additional actions taken, the most likely effect on having only one PORV open is:
| |
| : a. No effect occurs; bleed and feed remains effective.
| |
| : b. Reactor vessel water level remains higher with the reduction in bleed.
| |
| : c. Core uncovery occurs as a higher RCS repressurization decreases ECCS flow.
| |
| : d. RCS depressurizes to a lower pressure with less of an inventory loss.
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| i Page: 76 1
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 077 LHUT 1-1 ruptures while the Auxiliary Building Ventilation system is selected to the
| |
| " Buildings Only" mode.
| |
| To ensure that the Aux Bldg ventilation flow is adequately filtered WHICH ONE (1) of the following actions should be done?
| |
| : a. Place BOTH units in Safeguards Only with an "S" signal.
| |
| : b. Place Unit 1 in Safeguards Only and Unit 2 in Buildings and Safeguards.
| |
| I
| |
| : c. Place Unit 1 in Building and safeguards with no changes in the Unit 2 ventilation alignment. j
| |
| : d. Place Both units in Buildings and Safeguards.
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| Page: 77
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| EXAMINATION '
| |
| Senior Reactor Operator Written Exam I Question: 078 Unit 1 Reactor is at 100% power. The Control oporator reports that PK 02-11. "CVI Mode 6 Cutin"is in alarm. Upon futher investigation the Balance of Plant Operator discovers that the Containment Vent isolation (CVI) selector switch is in the MODE 6 position.
| |
| WHICH ONE (1) of the following describes the operability of CVl?
| |
| : a. CVI is INOPERABLE, CVI actuation can only be caused by the radiation monitors
| |
| : b. CVI is OPERABLE, the slave relays have a different power supply,
| |
| : c. CVI is OPERABLE, the master relay have a different power supply,
| |
| : d. CVI is INOPERABLE, No CVI actuation can occur.
| |
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| Page: 78
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 079 One minute after a coincident reactor overpower accident /large break LOCA occurred, containment pressure was noted to be steady at 12 psig and containment radiation was noted to be steady a 2E+5 rad /hr. Th.e use of adverse containment parameters was directed by the Shift Foreman.
| |
| WHICH ONE (1) of the following sets of current conditions would allow discontinuing the use of Adverse Containment Conditions?
| |
| Containment Radiation Integrated Radiation Pressure Level Level (PSIG) (Rad /HR) (RAD)
| |
| : a. 2.0 1.5E+5 3.5E+5
| |
| : b. 2.5 1.5E+4 1.5E+5
| |
| : c. 1.5 5E+4 1.5E+6
| |
| : d. 3.5 SE+3 6.5E+4 Page: 79
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam l
| |
| I Question: 080 Unit 2 experienced a loss of 'nstrurnent Air. The Shift Foreman has directed the response to the loss by using OP AP-9, " Loss of Instrument Air". During the performance of the procedure Main Feedwater Regulation valves failed closed causing a Reactor Trip.
| |
| WHICH ONE (1) of the following describes the proper procedure usage of E-0
| |
| " Reactor Trip or SI" and/or OP AP-9, for this event?
| |
| : a. Enter E-0 and refer to AP-9
| |
| : b. Stay in AP-9 and refer to E-0
| |
| : c. Return to AP-9 after completing immediate actions of E-0
| |
| : d. Enter E-0 and implement AP-9 in parallel l
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| l Page: 80
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| | |
| l EXAMINATION -
| |
| l l
| |
| Senior Reactor Operator Written Exam Question: 081 Unit 2 is operating with the following plant conditions: ;
| |
| . 75% power
| |
| . All control systems in automatic, except Rod Control in manual
| |
| . PDP Charging Pump 2-3 in service
| |
| . PZR level Hi/Lo control alarm (PK05-22) is in alarm. (Input 543 - PZR Lvl Dev High from REF backup Htrs on)
| |
| Based only on the above indications WHICH ONE (1) of the following could be the cause for the deviation alarm?
| |
| : a. Condensate valve FCV-55 failing open
| |
| : b. Train A HI Letdown heat exchanger room temperature failing hi on 2/3 sensors.
| |
| : c. 10% steam dump valve PCV-21 failing open
| |
| : d. First Stage impulse Pressure transmitter PT-505 fails high.
| |
| Page: 81
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| | |
| EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 082 Using Refueling Operations Technical Specifications, determine WHICH ONE (1) of the following combinations of initial enrichment (%) and assembly burnup (MWD /MTU) shown below, would be allowed to be stored only in region 1 of the spent fuel pool?
| |
| (Fuel pellet diameter is 0.3225 inches) 10i tial Enrichment MWD /MTU
| |
| : a. 3.5 25,000
| |
| : b. 4.0 30,000
| |
| : c. 4.5 35,000
| |
| : d. 4.65 1,000 Page: 82 I
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| | |
| EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 083 Given the following:
| |
| . A reactor trip occurred coincident with a loss of offsite power.
| |
| . EOP E-0.4, " Natural Circulation Cooldown with Steam Void in Vessel (without RVLIS)", is in progress.
| |
| . RCS pressure is 1600 psig.
| |
| e RCS temperature is 450 F.
| |
| WHICH ONE (1) of the following is the reason for equalizing charging and letdown flows during the subsequent depressurization?
| |
| : a. Allows Pressurizer level to be used for monitoring void growth,
| |
| : b. Assures RCS total mass does not drop below minimum conditions assumed in FSAR analysis.
| |
| : c. Pressurizer level is not accurate during these conditions and flow matching assures the pressurizer will not go solid.
| |
| : d. Assures the regenerative heat exchanger is not overstressed due to large fluctuations in charging and letdown.
| |
| I Page: 83
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| | |
| c;-
| |
| EXAMINATION Senior Reactor Operator Written Exam Question: 084 WHICH ONE (1) of the following tasks can the Work Control Shift Foreman authorize
| |
| - with verbal concurrence of a Unit Shift Foreman?
| |
| : a. Performance of STP R-1 A, " Exercising Full Length Control Rods"
| |
| : b. Performance of STP M-21C, " Main Turbine Valve testing" i
| |
| : c. P.emove DG 1-2 from service for Governor replacement
| |
| : d. Placement of CVCS Deborating Demin 1-2 in service Page: 84
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 085 Unit 1 experienced an instrument failure which caused the control rods to insert 20 steps and Tavg to decrease 3*F below Tref. Control Rods were placed in manual.
| |
| WHICH ONE (1) of the following methods should be used by the Control Operator to recover Tavg once the instrument failure is fixed ?
| |
| : a. Withdraw controls rods as necessary, stopping at least every 3 steps to veriff indications of temperature and power.
| |
| : b. Withdraw rods as necessary while continuously monitoring indications of temperature and power.
| |
| : c. Withdraw rods as riecessary, stopping every full step to monitor indications of temperature and power.
| |
| : d. Place Rod controlin Auto and monitorindications of temperature and power.
| |
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| | |
| EXAMINATION Senior Reactor Operator Written Exam J
| |
| Question: 086 WHICH ONE (1) of the following would NOTrequire the prompt notification of the Shift Chemistry Technician by the Shift Foreman?
| |
| : a. Increasing SI Accumulator level by 4%
| |
| : b. Initiating Steam Generator Blowdown overboard
| |
| : c. Starting the Nash Vacuum pump
| |
| : d. Cutting in additional Steam Jet Air Ejectors l
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| Page: 86 l l
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| | |
| - q EXAMINATION Senior Reactor Operator Written Exam Question: 087 Given the following conditions on Unit 1:
| |
| . A refueling outage has just been completed
| |
| . Preparations are being made to heat up the RCS and conduct a Reactor startup.
| |
| . RCS temperature is 180 F
| |
| . RCS pressure is 450 psig
| |
| . RCS Boron Concerstration is 2050 ppm
| |
| . RHR Cooling has just been secured WHICH ONE (1) of the following describes a condition that requires containment integrity be established?
| |
| : a. Increasing RCS pressure to 550 psig.
| |
| : b. Increasing RCS temperature to 205*F
| |
| : c. Reducing boron concentration to 1980 ppm
| |
| : d. Shifting from RHR cooling to RCP operation for SG cooling l
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| l i
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| Page: 87
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| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam l Question: 088 While performing EOP E-0.4 " Natural Circulation Cooldown with steam Void in Vessel (Without RVLIS)" Step 1 "Try to Start an RCP," Pressurizer level is increased to j greater than 57% on Unit 1 and greater than 75% on Unit 2. '
| |
| WHICH ONE (1) of the following describes the basis for the difference in pressurizer level?
| |
| : a. Larger upper head volume on Unit 1
| |
| : b. Larger upper head volume on Unit 2
| |
| : c. Higher Natural Circulation Flowrate on Unit 2 l
| |
| : d. Higher Natural Circulation Flowrate on Unit 1 1 I
| |
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| Page: 88 l
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| | |
| n-EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 089 Unit 1 DFWCS backup power supply needs replacement. 1 A formal communication has been written to allow the MFW Regulation Valves and MFW pump controllers to be placed in manual for the duration of the replacement.
| |
| WHICH ONE(1)'of the following should be prepared and attached to the formal communication before work is allowed to begin.
| |
| : a. Licensing Basis impact Evaluation (LBIE)
| |
| : b. Operability Evaluation (OE)
| |
| : c. Licensing Basis impact Evaluation (LBIE) Screen
| |
| : d. Prompt Operability Assessment (POA) i i
| |
| l Page: 89
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| | |
| l EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 090 Unit 1 is in Mode 5 with the following electrical equipment status:
| |
| . Start-up Power is cleared for Transformer work
| |
| . Aux. Power and Crosstie capability is operable
| |
| . Diesel Generator 1-1 is cleared
| |
| . Diesel Generator 1-2 & 1-3 are operable Using the attached reference (s) determine WHICH ONE (1) of the following situations is required to allow Maintenance to place Diesel Generator 1-3 on its backup DC power source 1
| |
| : a. After lY 13 is made available
| |
| : b. After Diesel Generator 1-1 is made available
| |
| : c. After Mode 6 transition
| |
| : d. After water level in reactor cavity greater than 23' l
| |
| l 1
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| Page: 90 m
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 091 What is the basis for requiring two (2) Reactor Coolant Pumps (RCPs) to be started prior to RCS temperature exceeding 350 F7 (Assume reactor trip breakers are CLOSED and rod drive MG sets are Energized)
| |
| : a. Ensures conditions assumed in the accident analysis are satisfied for a bank withdrawl from suberitical accident.
| |
| : b. Ensures RCP seal package will NOT experience thermal shock from pump starts at higher RCS temperatures,
| |
| : c. Ensures boron concentration in the pressurizer and the RCS will be equaiized I with pressurizer spray flow. I i
| |
| : d. Ensures thermal shock to the pressurizer is minimized by providing adequate l spray line flow l l
| |
| Page: 91
| |
| | |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam j l
| |
| l
| |
| . Quertic. : 092 l i
| |
| l WHICH ONE (1) of the following is the responsibility and/or duty of ONLY the Refueling SRO?
| |
| : a. Verification of all refueling prerequisites,
| |
| : b. Delegating supervisory responsibility for the supervision of core alterations to a designated operations representative.
| |
| : c. Observation and direct supervision of all fuel handling in the FHB.
| |
| : d. Determining the cause of high radiation or criticality alarms during refueling operations.
| |
| )
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| Page: 92 l l
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| l
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| | |
| F 1 EXAMINATION ,
| |
| Senior Reactor Operator Written Exam j Question: 093 Given the following:
| |
| . An operator is performing a whole body frisk using a portable frisker, RM-14/HP-210.
| |
| . Background radiation count rate is at the MAXIMUM allowed for using the frisker.
| |
| WHICH ONE (1) of the following is the count rate (background + actual) at which the operator is considered to be contaminated?
| |
| : a. 100 counts per minute
| |
| : b. 200 counts per minute
| |
| : c. 300 counts per minute
| |
| : d. 400 counts per minute i
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| I l
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| Page: 93
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| | |
| t EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 094 WHICH ONE (1) of the following conditions would require the termination of a containment vent or purge?
| |
| : a. Failure of either RM44A or B while in Mode 6 with movement of irradiated fuel in containment.
| |
| : b. Failure of either RM44A or B while in Mode 4.
| |
| : c. Failure of either RM44A or B while in Mode 6 during core alterations.
| |
| : d. RM-44A out of service and RM-44B failure in Mode 5.
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| i i
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| )
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| Page: 94
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| | |
| p,r -
| |
| EXAMINATION '
| |
| Senior Reactor Operator Written Exam Question: 095 An area in the Auxiliary Building has the following conditions; Dose Rate 10 mrem /hr Airborne 1-131 1.5 DAC Surface contamination 800 dpm/100 cm2 gamma.
| |
| WHICH ONE (1) of the following is the correct posting for this area?
| |
| : a. Radiation Area only.
| |
| : b. Surface Contamination Area and Airborne Radioactivity Area.
| |
| : c. Airborne Radioactivity Area only,
| |
| : d. Radiation Area and Airborne Radioactivity Area.
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| l l
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| l Page: 95 j
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| | |
| EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 096 While operating at 100% power, a Main Steamline break on S/G 1-4 occurred. Si actuated, and E-0 " Reactor Trip or Safety Injection," and E-2, '' Faulted Steam Generator Isolation," were completed without incident. A transition to E-1, " Loss of Reactor or Secondary Coolant," has been' performed. You are now in E-1, Step 2,
| |
| " Check if Steam Generators are intact."
| |
| The following plant parameters exist:
| |
| . Secondary radiation monitors show NO alarms, NO upward trends, and NO upward spikes
| |
| . Containment pressure is 23 PSIG
| |
| . All RCS Cold Leg Temps are 250*F and DECREASING
| |
| . All RCS Hot Leg Temps are 253*F and DECREASING e RCS Pressure is 1000 PSIG and DECREASING
| |
| . S/G 1-1,1-2,1-3 pressures are 700 PSIG and STABLE
| |
| . S/G 1-4 pressure is 300 PSIG and DECREASING Based on the above parameters the operators should:
| |
| : a. Go to FR-Z.1, " Response to High CNMT Pressure."
| |
| : b. Return to E-2, " Faulted Steam Generator Isolation."
| |
| : c. Complete E-1, then go to FR-P.2, " Response to Anticipated Pressurized l Thermal Shock Conditions." I
| |
| : d. Go to FR-P.1, " Response to imminent Pressurized Thermal Shock l Conditions."
| |
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| Page: 96 {
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| l
| |
| | |
| e EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 097 During the response to a loss of all A/C power, a maximum rate S/G depressurization is performed to 240 PSIG.
| |
| How does this secondary depressurizatiori minimize the RCS inventory loss?
| |
| : a. It will cause the RCP No.1 seal to go from a film riding to face rubbing seal thus reducing RCS leakage.
| |
| : b. It ensures that NO RCP seal failures will occur.
| |
| : c. It reduces RCS pressure and seal AP and thus RCS leakrate.
| |
| : d. It will cause the No.1 seal to "uncock" and reseat properly. l l
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| l 1
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| Page: 97
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| | |
| E l EXAMINATION -
| |
| Senior Reactor Operator Written Exam Question: 098 Unit 1 is in a refueling outage with the following conditions:
| |
| . RVRLIS is at 107' 6" following nozzle dam installation
| |
| . Reactor Vessel Head is detensioned
| |
| . All equipment required by the outage safety plan is available
| |
| . A loss of ALL AC power occurred and recovery actions are in progress
| |
| . RCS Temperature is increasing AP SD-0 "Lc as of, or inadequate Decay Heat Removal" directs the operators to control RCS temperature increase by WHICH ONE (1) of the following methods?
| |
| : a. Decay heat removal using Natural Circulation
| |
| : b. Decay heat removal using feed and bleed
| |
| : c. Decay heat removal using fill and spill
| |
| : d. Decay heat removal using Reflux cooling i
| |
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| |
| Page: 98 L
| |
| | |
| EXAMINATION -
| |
| Senior Reactor Operator Written Exam i
| |
| Question: 099 Given the following conditions on Unit 1:
| |
| . Reactor Trip due to a Seismic induced LOCA
| |
| . All RCPs have been tripped
| |
| . Containment Spray has been actuated
| |
| . SFM has transitioned to EOP-E2 " Faulted SG lsolation" l The STA notes the following:
| |
| . Intermediate Range Nis - 10E-03 Amps, -0.3 dpm
| |
| . CETs - 580 F
| |
| . RVLIS Full range - 63%
| |
| . All SG NR Levels - Off scale low
| |
| . Aux Feedwater flow - 865 gpm total
| |
| . Containment Pressure - 16 psig
| |
| . Containment WR Sump Level- 95 ft
| |
| . Pressurizer level- Off scale low WHICH ONE (1) of the following identifies the monitoring frequency required for the Critical Safety Function Status Trees?
| |
| : a. Continuous
| |
| : b. Every 5 minutes
| |
| : c. Every 15 minutes
| |
| : d. Every 30 minutes
| |
| )
| |
| 1 Page: 99 l
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| ( )
| |
| | |
| EXAMINATION Senior Reactor Operator Written Exam Question: 100 WHICH ONE (1) c the following responsibilities may be delegated by the Interim Site Emerg. .,y Coordinator (ISEC)? -
| |
| : a. Approval of emergency dose limits necessary to save a life.
| |
| : b. Classification of an emergency event.
| |
| : c. Notification of the Nuclear Regulatory Commission.
| |
| : d. Ass'gnment of plant personnel to positions in the Site Emergency Organization.
| |
| i l
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| Page: 100 L_
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| | |
| 2
| |
| , SCENARIO 01 OVERVIEW' The crew is directed to ramp to 50% power to remove MFP 1-2 from service (Event 1).
| |
| The RO will have to borate to ramp turbine ar.d reactor power to 50%.
| |
| S/G Level transmitter LT-549 fails High (Event 2). The BOPCO and CO should identify the failed channel and report to the SFM. The SFM should refer to annunciator respense procedure PK09-15 for a Digital Feedwater Control System trouble and enter
| |
| - OP AP-5 for a Malfunction of Protection or Control channel. The SFM should inform the
| |
| ~
| |
| RO and BOPCO of the affects of the failure on AFW level control if needed.
| |
| Letdown pressure controller POV-135 fails closed while in the auto position (Event 3).
| |
| BOPCO should identify the failure of PCV-135 to control letdown pressure and report to the SFM. SFM should refer to PK04-21 for letdown press / flow. SFM should direct that the failed controller be placed in manual to regain letdown pressure control and ensure that the letdown relief valve re-seats or establishment of excess letdown.
| |
| Eagle 21 (LOP Halt) failure comes in (Event 4). The CO and BOPCO should identify the failure and report to the SFM. The SFM should refer to annunciator response procedure !
| |
| PK06-01 and enter OP AP-5. Crew should identify all instrumentation that is affected by the failure. LT-459 and PT-455 are failed as is and LT-459 and PT-455 should be deselected for control.
| |
| RCP 1-2 seal leakoff flow increases to 60 gpm (over 20 minutes) giving indications of a j RCP 1-2 # 1 seal problem (Event 5). SFM should refer to annunciator response ;
| |
| procedure PK05-02. RCP parameters shoulu be monitored and Operations Management should be informed. RCP 1-2 High Vibration (Event 5 continued). SFM should refer to annunciator response procedure PK05-05. SFM should direct a Reactor Trip, followed by the tripping of RCP 1-2 and the subsequent isolation of the RCP 1-2 # ;
| |
| 1 sealleakoff valve.
| |
| Following the transition to EOP E-0.1, RCP 1-2 seal package will fail, causing a 60gpm LOCA (Event 6), additionally Reactor Trip Breaker "A" fails to open. The LOCA results in decreasing pressurizer level accompanied by decreasing RCS pressure. The SFM should direct a manual SI prior to an automatic Si at 1850 psig and re-entry into EOP E-0 followed by a transition to EOP E-1. The SFM should direct the local action to open RTA. Charging Pump 1-2 will overcurrent trip on the SI signal and will not be available.
| |
| Approximately 10 to 15 minutes following the reactor trip, the intermediate range N1 should be low enough to energize the SR instruments except that NI-36 is undercompensated which results in onle i of 2 intermediate Nis below P-6. The RO/ BOP should recognize the failure of SR instruments to energize and the SFM should direct the manual re-energizing of the SR detectors.
| |
| The scenario will be terminated following diagnosis of the RCS not intact and a transition out of EOP E-1 to EOP E-1.1.
| |
| | |
| s l Appendix D Scenario Outline Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 1 Op-Test No.: Backup Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U I SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to an Eagle 21 system failure Evaluate the crew's ability to diagnose and respond to a letdown pressure controller failure Evaluate he crew's ability to diagnose and respond to a S/G Level Channel failure Evaluate the crew's ability to diagnose and respond to a RCP # 1 seal failurc
| |
| ! Evaluate the crew in using E0Ps during a RCP seal failure LOCA Evaluate the crew's ability to respond to a partial P-4 actuation Evaluate the crew 's ability to diagnose and respond to a undercompensated IR channel InitialConditions: 100% power, equilibrium xenon, Begir,ning of cycle (IC-1) l Tumover: 10 rpd leak on S/G 1-3, D/G 1-1 is 00S, CCW pump 1-3 is 00S MFW pp 1-2 increased Vibrations - ramp unit to 50% for repair ASAP Time Event Malf. Event Event l min No. No. Type
| |
| * Description j 0 1 n/aa N, RO Ramp power to 50% for repair of MFW pp 1-2 10 2 xmt I, BOP S/G Level Channel LT-549 fails HIGH mfw46 15 3 cnh cyc8 C, BOP Letdown pressure conuol valve PCV-135 fails CLOSED in auto, operates in manual 23 4 ppl6a I, RO LCP Halt Protection Set 1, Rack 1 33 5 rep 1b C, RO RCP 1-2 # 1 sealfallure 36 ser 1244 RCP 1-2 # 1 High Vibration I
| |
| cond on 6 rep 2b, M,ALL RCP 1-2 seal package failure.
| |
| l trip rep 3b pplSa C,SFM Rx Trip Breaker'RTA" fails to OPEN res3b Loop 2 RCS SB LOCA cond nis 3b 1, RO IR NI 36is undercompensated cyc2 C, RO Charging Pump 1-2 Overcurrent trip on Safety injection signal (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor i
| |
| | |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1 _ Scenario rJo.: 1 Event No.: 1 Page ._1_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Commence ramp to 50% power i
| |
| Time Position Applicant's Actions or Behavior BOP Monitor Main Feedwater Pump 1-2 vibration readings l l
| |
| RO initiate Boration for ramp to 50% power
| |
| . Operate the makeup system mode select switch l
| |
| . Operate the boric acid integrator l Sttup DEHC l
| |
| . Place MW feedbackir service
| |
| . Set MW reference !
| |
| . Setload rate Commence ramp to 50%
| |
| i Monitor Tavg and Tref and borate as necessary l SFM Go to PK09 Review actions for increased MFW pp. Vibrations Review precautions and limitations of OP L-4 and conduct tumover briefing Direct RO to commence a ramp to 5ti.;, power at up to SMW/ min (Admin Limit)
| |
| Direct BOPCO to monitor MFW pp.1-2 bearing vibrations NUREG-1021 1 InterimRev.8 January 1997
| |
| | |
| I, l a
| |
| Appendix D Scenario Outline Form ES-D-2 )
| |
| Op-Test No.:_1 _ Scenario No.: 1 Event No.: 2 Page _2_ of _7._
| |
| Event Description : SG LevelChannel LT-549 Fails High Time Position Applicant's Actions or Behavior BOP Recognize and report LT 549 failure R0 Monitor and report SG level trends and Digital Feedwater controls status i
| |
| SFM Go to PK 09-15
| |
| . Dispatch Operator to Digital Feedwater Cabinet in Cable Spreading Room Go to OP AP 3
| |
| . Reviewattachment4
| |
| . Determine that failure will affect SG level control on a Reactor Trip
| |
| . Initiate a bistable trip sheet (verbalize and initiate BS sheet actions) l Consult Technical Specification 3.3.1
| |
| . Trip inoperable channel within 6 hours i
| |
| Direct Control Room Asset Team to investigate failure of LT-549 l
| |
| NUREG-1021 2 Interim Rev. 8, January 1997
| |
| | |
| ]
| |
| {
| |
| Appendix D Scensrio Outline Form ES-D-2 Op-Test No.:_1 _ Scenario No.: 1 Event No.: 3 Page _3_ of _7._
| |
| Event Description : Letdown Pressure control valve PCV-135 failure l
| |
| Time Position Applicant's Actions or Behavior i
| |
| BOP Recognite and report Letdown flow decreasing to zero Recognize and report Letdown pressure norma! and PCV-135 demand at zero %.
| |
| Retums letdown flow to 75 gpm with PCV 135 in manual Recognize and report Letdown RV has lifted Reduces Letdown pressure to reseat Letdown RV and retums letdown to normal in manual or isolates letdown and places excess letdown in service as directed.
| |
| RO Monitors Letdown and Charging parameters Monitors PRT Parameters Recognize and report PRT level increasing Places turbine ramp on HOLD at DEHC panel SFM Go to PK 04-21
| |
| . Determine that Letdown Pressuru controller PCV-135 has failed in auto
| |
| . Direct BOPCO to control letdown pressure us5g PCV-135 in manual
| |
| . Direct RO to monitor Letdown, Charging and PRT parameters
| |
| . Direct RO to place Turbine ramp on Hold at DEHC panel Go to OP B-1 A:XII .
| |
| . Direct BOPCO to reduce Letdown pressure to reseat Letdown RV or isolate letdown and )
| |
| establish excessletdown j Direct Control Room Asset team to investigate failure of PCV-135 ,
| |
| l l
| |
| I i
| |
| NUREG-1021 - 3 Interim Rev. 8, January 1997 :
| |
| | |
| or 9
| |
| Appendix D- Scenario Outline Form ES-D-2 i
| |
| Op-Test No.:__1 Scenario No.: 1 Event No.: 4 Page _4_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Eagle 21 failure - LCP HALT on Protection Set 1, Rack 1 l
| |
| Time Position Applicant's Actions or Behavior BOP Recognize and report symptoms of an LCP Halt
| |
| . Annunciators for channel set failure i
| |
| Take actions as directed oy SCO RO Recognize and report symptoms of an LCP Halt on protection set 1 j e Annunciators for channel set failure {
| |
| . Protection set 1 alarm on annunciator typewriter l
| |
| Check e ' '. rolling systems controlling in AUTO l i
| |
| i
| |
| . LT-459 failed as is l
| |
| . PT-455 failed as is
| |
| ]
| |
| . Deselect 459 and 455 as controlling channels )
| |
| l l
| |
| SFM Acknowledge reports from BOP /RO (PK06-01 and 06-04)
| |
| Go to OP AP-5 and direct operator recovery actions
| |
| . Direct RO to deselect 455 and 459 as pressurizer controlling channels l I
| |
| l Contact Nuclear Operator / Control Room Asset Team to investigate Consult Technical Specifications 3.3.1 and 3.3.2
| |
| ' . 6 hour action to trip affected bistabbs for PZR press, PZR level and RCS flow NUREG-1021 4 Interitr Rev. 8, January 1997
| |
| | |
| ' Appendix D Scenario Outline Form ES-D-2 Op Test No.:__1_ Scenario No.: 1 Event No.: 5__ - Page _5_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP 1-2 No.1 sealfailure Time Position Applicant's Actions or Behavior )
| |
| i
| |
| ' BOP Recognize and report high seal leakoff flow on RCP 1 2 l sJK hyb i n A LM yij Q # $,<$ $ M j g g M Q Q e.vd%&g
| |
| -f nu %
| |
| l jd$g%gy
| |
| ^ ]e$5 stw hi$$,tj%g%gyit&h$
| |
| y
| |
| % d Jh&m.:yyN@y% N Close No.1 seal leakoff valve after 3 to 5 minutes Monitor CCW to ensure adequate thermal barrier flow Compleic iminediate actions of EOP E-0 I
| |
| . Verify 4kV buses energized l
| |
| RO Acknowledge and report PK05-01 alarm input number 1393, RCP 1-1 Seal Leakoff Flow Hi Acknowledge and report PK05-05 RCP Vibration yurg:w a
| |
| @Al"llde'5i$iualW;wngw$!" EliiXMN.x .idThe$$a2 me n 87a: f?!@
| |
| S@is c @J1; y 9 ry[s@ M
| |
| *g%;:#u
| |
| ) "CriticalTaskiMMM MY ? fa <
| |
| 1 Complete immediate actions of EOP E 1
| |
| )-
| |
| . Verify Rx tripped - reports RX trip Bkr. "A" closed
| |
| . . VerifyTurbinetripped e Ct rk Si actuation not required SFM Go to PK05-02, RCP 1-2 Go to PK05-05, RCP Vibration i Diagnose No.1 sealfailure SmanualN we
| |
| ,0isciBishiiilhdti%
| |
| ?NEloillTask#dg%Aggp w
| |
| gam 9gSpgbM.gzhggp&pr g , ads $w$$dd$nWRCP3 ggmeMaft q pnWgs w@h M =
| |
| Go to EOP E-0
| |
| . - Directs Operator to locally open Rx Trip Bkr. "A".
| |
| Go to EOP E-0.1 NUREG-1021 1 5 Interim Rev. '. Jar = > 1997
| |
| | |
| . 1 i
| |
| 1 9
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 1 Event No.: 6__ Page _6_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP sealfailure (SBLOCA)
| |
| Time Position Applicant's Actions or Behavior BOP Recognize and report RCS Pressure Decrease Complete immediate actions of E0P E-0
| |
| . Verify 4kV buses energized Perform Appendix E; Secondary Auxiliary Status Control AFW flow to S/G 1-4 in manuai due to level channel failure j
| |
| RO Recognize and report RCS Pressure Decrease Perform ManualSafetyinjedtiod; ''f >
| |
| .7
| |
| ** CriticalTasle- .3 Complete immediate actions of E0P E-0
| |
| . Verify Rx tripped
| |
| . Verify Turbine tripped e Check Si actuation required Recognize and report charging Pump 1-2 over current trip Recognize and report failure of both SR channels to energize
| |
| . Energize both SR channels (due to NI-36B failure)
| |
| "CriticalTask NUREG 1021 6 Interim Rev. 8, January 1997
| |
| | |
| ~
| |
| l'^ :
| |
| Ayandix D Scenario Outline Form ES-D-2 I
| |
| Op-Test No.:_1 and 2_ Scenario No.: 1 Event No.: 6__ Page _7_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP sealfailure(Continued)
| |
| Time Position Applicant's Actions or Behavior SFM ;Diiect 'a mahual'SI). ' Aw @ ' . . hs [> N :: 4- y ,
| |
| "Criticahhabk}i [b[] kM h ,< s, [
| |
| Re-enter E0P E-0 and direct immediate actions of E-0
| |
| . Direct STA to monitor CSFSTs Direct the following actions
| |
| . Direct energizing both SR dete', tors
| |
| * Direct BOPCO to control ArW flow to SG 1-4 Transition to E0P 61
| |
| . Review major actions
| |
| . Assign Foldout Page items j Transition to EOP E-1.1
| |
| . Review major actions
| |
| . Assign Foldout Pageitems
| |
| ]
| |
| i l
| |
| NUREG-1021 7 Interim Rev. 8, January 1997
| |
| | |
| .. i I SCENARIO 02 OVERVIEW '
| |
| The crew is directed to ramp to 100 % power (Event 1), #2 Htr. Drip pump repairs are complete with the pump warmed up and running. The RO will have to dilute to ramp turbine and reactor power to 100%.
| |
| During the ramp to 100% power, FCV-111B will fail to open in auto or manual (Event 2) causing RCS normal dilution ficwpath to become inoperable. The SFM should refer to PK05-11 and enter AP-19 for malfunction of reactor control makeup system. SFM should direct local operators to investigate FCV-111B failure and direct the CO to use Alternate Dilution for RCS makeup.
| |
| Containment Exhaust radiation monitor RM-44A will fail high, over-ranged (Event 3),
| |
| causing an inadvertent containment vent isolation (CVI). SFM should refer to the annunciator response procedure PK011-21, High Radiation and BOPCO should be sent to control room radiation monitors to locate the source of the high alarm condition.
| |
| SFM should refer to annunciator response procedure PK01-17 and determine that RM-12, Containment Gaseous Radiation Monitor is out of service and direct the BOPCO to f place the CFCU drain collection system in service. )
| |
| Nuclear Instrument NI-42 will fail high (Event 4), CO should report the failure to the SFM. BOPCO should be sent to NI panels in the control room to investigate the failure.
| |
| SFM should enter OP AP-5 for the Nl failure and determine that N142 can be defeated.
| |
| SFM should direct the BOPCO to defeat NI-42 using OP AP-5 Attachment 4.1. BOPCO should defeat Nl-42 at the NI Racks.
| |
| Reactor coolant pump 1-3 thermal barrier leakage (Event 5), ramps from 0 to 30 gpm l causing increase in CCW head tank level, Radiation monitors RM-17A & B alarm and
| |
| ! close of the CCW head tank vent RCV-16. CO and BOPCO should report to the SFM.
| |
| l The SFM should enter AP-1 for Excessive Reactor Coolant Leakage and directs CO l l
| |
| and BOPCO in the determination of leak location. SFM should enter AP-11 Malfunction of CCW System ana directs control room and local actions to isolate thermal barrier for RCP-13.
| |
| Earthquake with a magnitude of 0.35g will occur (Event 6) causing the loss of startup power. An ATWS will exist due to the magnitude of the earthquake and the failure of the l reactor to trip. SFM should direct the CO and BOPCO to perform the immediate actions j: for the ATWS and direct the opening of MG supply load center breakers to cause the 4 reactor to trip. During the event PORV 456 will open and stay open, power to the block i valve is not available, resulting in a SB LOCA. The steam space break results in increasing pressurizer level accompanied by decreasing RCS pressure. The SFM should direct a manual SI prior to an automatic Si at 1850 psig and re-entry into EOP E-0 followed by a transition to EOP E-1.
| |
| The scenario will be terminated following diagnosis cf the RCS not intact, Appendix "C" required to be used due to failure of WR Thot TE-413A (Event 7) and a transition out of EOP E-1 to EOP E-1.2.
| |
| | |
| l
| |
| :o l Appendix D OperatorActions Form ES-D-1 l l . j Facility: DCPP Units 1 & 2 Scenario No.: 2 Op-Test No.: 1A/1B j Examiners: Operators: SRO-U / SRO-U ,
| |
| SRO-U / SRO-U )
| |
| SRO-U / SRO-l Objectives: Evaluate the crew's ability to diagnose and respond to a RCS makeup control failure Evaluate the crew's ability to diagnose and respond to a Radi@n monitor failure Evaluate the crew's ability to diagnose and respond to a The. _ 2rrierleak Evaluate the crew's ability to diagnose and respond to a ? .: Fe e Evaluate the crew's ability to respond to a seismic event Evaluate the crew's ability to diagnose an ATWS and take tirnely a 7ns l Evaluate the crew's ability to diagnose and respond to a stuck open PORV l Evaluate the crew in using the EOPs during recovery from SB LOCA i Initial Conditions: 75% power, equilibrium xenon, end of cycle (IC-36)
| |
| Tumover: D/G 1-1 is OOS, for tube oil heater replacement l Ramp to 100% power following repairs to #2 Htr Drip pump.
| |
| Time Event Malf. Event Event min No. No. Type
| |
| * Description 0 1 n/a N,SFM Ramp power to 100% following repairs to #2 Htr Drip pump R, RO variable 2 vtv evc14 c, RO FCV-111B VCT inlet from RCS makeup fails closed 12 3 xmt rms34 1, BOP instrument failure of RM-44 results in High rad alarm and CVI i 20 4 nis6b 1,ALL Instrument failure of NIS-42 high 27 5 ccw3c C,ALL RCP 1-3 Thermal Barrier Leak (30 gpm) to CCW system 42 6 pp15a M,ALL 0.35g Earthquake causes loss of startup power and 500 Kv power, ppl5b C, BOP Auto and manual Reactor trip blocked, PORV PCV-456 fails open with no power to block valve.
| |
| Post 7 res35 I, RO TE 413A Wide range Thot fails high Trip ;
| |
| 1 1
| |
| (N)ormai (R)eactivity (1)nstrument (C)cmponent (M)ajor
| |
| | |
| E .
| |
| l.
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| 1 s i
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| t .
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| Appendix D S narioOutline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 2 Event No.: 1 Page _1_ of _7_
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| Event
| |
| | |
| == Description:==
| |
| Commence ramp to 100% power Time Position Applicant's Actions or Behavior BOP Monitor plant parameters RO Initiate dilution for ramp to 100% power
| |
| . Set up makeup control sys+em for dilution in batch mode (100-200 gals.)
| |
| Setup DEHC
| |
| . Place MW feedbackin service
| |
| . Setload reference
| |
| . Setload rate Raise VPL(Valve Position Limit)
| |
| Commence ramp to 100%
| |
| SFM Review precautions and limitations of OP L-4 and conduct tumover briefing Direct RO to commence a ramp to 100% power at 3-5 MW/ min i
| |
| l l
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| I j
| |
| I l
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| l NUREG-1021 1 Interim Rev. 8, January 1997 l l
| |
| t- J
| |
| | |
| e ._
| |
| .Y Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1 and 2_ Scenario No.: 2 Event No.: __2 Page _2_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| VCT Makeup valve FCV-111B Failure _
| |
| l l
| |
| Time Position Applicant's Actions or Behavior i BOP Recognize indications of CVCS makeup system failure and report status to SFM l
| |
| . Verify correct Dilution lineup l
| |
| . Recognize and report failure of FCV-111B to open when required i
| |
| l l
| |
| Take actions as directed by SFM
| |
| . Coordinate with RO to complete Dilution j
| |
| . Attempts to manually open FCV-111B, reports failure of valve to open to SFM RO Recognize indications of CVCS makeup system failure and report status to SFM e Acknowledge CVCS MAKEUP DEVIATION alarm, PK05-11, and report to SFM
| |
| . Verify the makeup system switch alignment and integrator settings
| |
| . Uses attemate Dilution Path to dilute the RCS l
| |
| Recommend placing DEHC in hold if ramp was started 1
| |
| SFM Acknowledge BOPCOIRO on CV0: : 4AKEUP DEVIATION alarm and refer to PK05-11 Determine a malfunction of the reactor makeup control system
| |
| . Go to OP AP-19, Malfunction of Reactor Makeup Control System
| |
| . Direct BOPCO to Manually open FCV-111B Direct Control Room Asset team to investigate failure of FCV-111B (Cue: SS direct SFM to Altemate dilute and continue with the ramp)
| |
| Direct RO/ BOP to use Altemate Dilution per OP B-1 A:Vil & continue ramp MOREG-1021 2 Interim Rev. 8, January 1997
| |
| | |
| i I
| |
| i Appendix D Scenario Outline' Form ES-D-2 l l
| |
| Op-Test No.:__.1 Scenario No.: 2 Event No.: .3 Page _3_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RM-44A fails high I Time Position ~ Applicant's Actions or Behavior
| |
| ,m._
| |
| BOP Recognize symptoms of a radiation mon l tor failure
| |
| . High Rad & Rad Monitorfailure alarms I e CVlinitiated j
| |
| . RM-44A overranged
| |
| . CVI Auto actions Place CFCU drain collection system in service R0 Recognize symptoms of a radiation monitor failure
| |
| . High Rad & Rad Monitorfailure alarms
| |
| . CVlinitiated i
| |
| )
| |
| Take anions as directed by SFM SFM Acknowledge reports from BOP /RO, refer to ARP PK02-06,11-21,11-22 & 01-17 i l
| |
| Recognize that the CVI caused low flow to RM-11 & 12 Direct Control Room Asset team to investigate failure of RM-44A Consult Technical Specification 3.4.6.1.c
| |
| . Direct CFCU drain collection system be placed in service
| |
| . 24 hour action to take grab samples of the containment atmosphere NUREG-1021 3 Interim Rev. 8. January 1997
| |
| | |
| Appendix D Scenario Outline Form ES-D-2 Op Test No.:_1_ Scenario No.: 2 Event No.: ._4 Page _4_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| NIS Channel N42 Fails High Time Position Applicant's Actions or Behavior BOP Defeats NI 42 at NI rsnel per SFM direction per AP-5 Attachment 4.1 instructions e Defeats Rod Stop e Defeats Power Mismatch e Oefeats QPTR e Defeats Comparator Reports Channel 42 defeated to SFM and RO RO Recognizes and reports rod motion concurrent with NI 42 failure na m y .n mbi, n .:PlacesRodConti mi iniManualgs*gm ob . .m.7, , R ; M ,SidMc . ,g 'm:.,.w,gw&
| |
| ," Ar$?f 8 Lu : W- 1: "~t N n t ;.
| |
| idCriticalTask!Q
| |
| * e , _ :nw .g #@[9fW $} j& 3 & .-h@@1W S,a- >WhL gM M$%M l Recovers Tavg e Using Rod Control- using 3 step pull and wait or e Adjustment of Turbine load SFM Directs RO to place Rod Control in Manual Enters OP AP-5 e Directs BOPCO to defeat NI-42 using attachment 4.1
| |
| . Directs RO to recoverTavg e using 3 steps pull and wait or e Adjustment of Turbineload Contact Control Room Asset team to troubleshoot and repair NI-42.
| |
| ConsultTechnical Specifications, e 3.3.1 - 6 hours to place tripped condition
| |
| . 4.2.4.2 - QPTR - MIDS runs required within 12 hours NUREG-1021 4 Interim Rev. 8. January 1997
| |
| | |
| E. E- )
| |
| >w Appendix D Scenario Outline . Form ES-D-2 '
| |
| Op-Test No.:_1 and 2_ . Scenario No.: '2 Event No.: __5 Page _5_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| RCP 1-3 Thermal Bamer Leak '
| |
| Time Position Applicant's Actions or Behavior BOP Recognize and report increase in RCP 1-3 seal injection flow Recognize and report increase in RM-17A & B indications i
| |
| Recognize and report increase in CCW head tank level l Recognim and report RCV 16 closed for CCW Head tank - )
| |
| j r
| |
| l
| |
| )
| |
| s
| |
| , TMaictionissidisikidUjhTliik$$w$M&[x p,,, yp3 3INiW WMsWN$5d.- e - "g a N$$MEM g
| |
| w%. , egbiltsnisiswsddPOV-760nf N disolateCCWliithodi la6 ~ W,e w gy I l
| |
| [ h kAsk h h h d k k N b b b h us NNNbh %
| |
| l RO Recognize and report increase in charging flow l
| |
| Determine leakage to CCW system of ~30gpm
| |
| , SFM May Go to PK05-03 for increasing RCP seal flow - determines no alarm for high flow l
| |
| l Go to AP-1 for Excessive Reactor Coolant Leakage
| |
| ! Direct RO to determineleak size Direct BOPCO to check possible leakage paths l
| |
| Direct BOPCO to Check CCW parameters L
| |
| ! *' CCW Head tank Level l
| |
| e CCW Rad Monitors
| |
| . . RCV-16 Determine leak is in the CCW System and Go to AP-11, Section B Determine leak is a thermal barrier leak EEEkN awng N NdiNIIidNN@k;gpdb
| |
| + A. n ww$[ eIbis! awppym Wq&W s M M S w1bGliodifik99ig :
| |
| Contact NSSS Asset team for repair of thermal barrier
| |
| ' NUREG-1021 - 5' interim Rev. 8, January 1997
| |
| . J
| |
| | |
| E .
| |
| i..r s
| |
| l
| |
| : l. .
| |
| Appendix D Scenario Outline Form ES-D-2 l
| |
| ! Op-Test No.:_1_ Scenario No.: 2 Event No.: 6&7 Page _6_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| ATWS on Earthquake with failed open PORV Time Position Applicant's Actions or Behavior BOP Performimmediate actions of E0P E-0 recognize failure of rods to insert and report status to SFM nymmvmm yynmrm mmemr
| |
| . mewnsm,13Dand13Earimmmamw m
| |
| ps ar?tze buses;ynen dre %wayp% port 3Mp% i toM"A status SF '~
| |
| %% m,lp,l;TaskMi%u*L % JW MM wpyy
| |
| (?s;w' %
| |
| * JnorM ink % nM* k .: w 6 Ri anasM:&
| |
| MCri 13 w ~tca M , 6,, 2 o wn cxwhkswaa:k.z s~ u.: ~a ~
| |
| e re-energize buses 13D and 13E after rods are fully inserted Recognize and report PORV 456 open with low pressure l
| |
| . Attempt to close PCV 456, reports valve failed as is and block valve without power Perform Appendix E, Secondary and Auxiliaries status !
| |
| RO Performimmediate actions of EOP E-0 e recognize failure of rods to insert and report status to SFM e attempt manualreactor trip ymp ~nme w n w n ,r w w w ~gm v m-n rn w wcw m m ~'n iezinsertcontrolrodsinmanuali C - ; RITWWi -
| |
| L ', % 4a ln-_ ;-}, 4,wl4)gf' f; gg;ar,y ,m *.aam ' ' fi
| |
| ,Q Q a ge< , z %%g .m * [;a m_nnam:ad e inform SFM when reactor trip breakers indicate open Recognize and report PORV 456 open with low pressure Recognize and report failure of WR Thot failing high TI-413A l
| |
| ! . NUREG-1021 6 Interim Rev. 8. January 1997, L
| |
| | |
| O Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 2 Event No.:__6 & 7 Page _7_ of _7._
| |
| Event
| |
| | |
| == Description:==
| |
| ATWS on Earthquake with failed open PORV (Continued)
| |
| Time Position Applicant's Actions or Behavior SFM Recognize failure of rods toinsert vmevmmmw gm pg y gyraernw5?insertrods>in
| |
| ;en d manualy;sm?p
| |
| : Wa vm w4
| |
| - w. - gg?1m w a w', a.
| |
| p # rect,RO1mw w p m wr,ae1m SNAP; s c%.blm4
| |
| :M diiect BOB lo de energize 480V biisss 13D'andp sb %q$&J % ..- ? 3Elt:
| |
| g; 55$;pa ygggggggg ggggggggsg
| |
| ,Jashiyj;;gygggggggy;qgg s y gpg
| |
| ,,;g a yn u e direct local opening of the reactor trip breakers e direct re-energizing buses 13D and 13E when rods are fully, inserted Go to E-0 and direct immediate actions to be completed Direct R0lBOPRO to close PCV-456 and Block valve Direct BOP to perform Appendix E, Secondary and Auxiliaries status Transition to EOP E-1 based on Open PORV with no block valve e conduct tailboard prior to entering E-1
| |
| . Direct STA to monitor CSFSTs l
| |
| Determine RCS subcooling using Appendix 'C' l I
| |
| Direct transition to EOP 1.2 based on RCS pressure decreasing !
| |
| e conduct tailboard prior to entering E0P E-1.2 l
| |
| l l
| |
| NUREG-1021 7 Interim Rev. 8, January 1997
| |
| | |
| 1 a
| |
| SCENARIO 03 OVERVIEW a
| |
| The crew is directed to continue the plant startup using OP L-3. First action for the crew will be for the BOPCO to parallel the Turbine Generator to the Grid (Event 1) per step 6.33 of OP L-3.
| |
| Loop 4 DeltaT fails High (Event 2) causing alarms associated with the failure. CO and BOPCO should report the failure to the SFM. The SFM should enter OP AP-5 for the failure and direct the CO to defeat the input to Delta T for the failed channel.
| |
| Steam Jet Air Ejector (SJAE) radiation monitor RM-15R will fail due to a loss of power I (Event 3). CO and BOPCO should report to SFM the failure indications of the radiation monitor and the SFM should reference AR PK11-07 for the SJAE radiation monitor failure. The SFM should contact TM to determine the cause of the alarm and an operator to determine the status of the RM-15R input to SPDS.
| |
| Earthquake with a magnitude of 0.01g will occur which will result in normal supply breaker for DRPI tripping open, and the loss of all DRPI indications, additionally PY-15 output bkr will fail open (Event 4). CO and BOPCO should report the loss of DRPI and the PY-15 to the SFM.
| |
| The SFM should refer to annunciator response procedure PK03-21 and determine the Tech Spec 3.0.3 applies and directs local operations to place DRPI on backup power supply. SFM should refer to OP AP-4 for loss of PY-15 and direct actions to restore PY-15. SFM may enter !
| |
| CP M-4 in response to the earthquake.
| |
| I PCV-20,10% steam dump valve will fail open (Event 5) causing Tavg to decrease and the 40% dump valves to modulate closed. The SFM should direct the BOPCO to manually close PCV-20 at the controller, when it is determined that the controller is failed, the SFM should direct the BOPCO to use the backup air source to close the PCV. The SFM should direct the CO to control TAVG using the 40% dumps once PCV-20 is closed.
| |
| Steam Generator 1-2 will develop a tube leak (Event 6). CO and BOPCO should identify the indications of the tube leak and report them to the SFM. SFM should enter and take actions per AP-3 for a Steam Generator Tube Failure. CO should determine when leakage is greater than 50 gpm and report to the SFM. SFM should direct a Manual Safety injection and the i performance of E-0 immediate actions. SFM may direct the BOPCO to perform an early isolation of SG 1-2 using E-3, Steam Generator Tube Rupture.
| |
| While performing E-3 the BOPCO should recognize the failure of the MSIVs to close (Event 7).
| |
| The failure should be reported to the SFM. The SFM should direct the actions of E-0 and transition to E-3. When it is determined the MSIVs will not close remotely, local operators l
| |
| should be dispatched to close the MSIVs locally. With isolation from intact SGs not possible the SFM should transition to ECA-3.1, SGTR with Loss of Coolant - Subcooled recovery desired. During performance of recovery actions PCV-455A fails open requiring the Reactor I Operator to manually close the failed open valve.
| |
| )
| |
| Once the SFM transitions to ECA-3.1 the MSIVs wili be closed and the scenario terminated when a cooldown is started.
| |
| l 1
| |
| | |
| Appendix D OperatorActions Form ES-D-1 Facility: DCPP Units 1 & 2 Scenario No.: 3 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l
| |
| . Objectives: Evaluate the crew's ability to diagnose and respond to a 10% dump valve failure Eva?uate the crew's ability to diagnose and respond to a SJAE Radiation monitor failure Evaluate the crew's ability to diagnose and respond to a Loss of DRPI indication Evaluate the crew's ability to diagnose and respond to a Loop 4 Delta T failure Evaluate the crew in using the AOP and EOP to respond to SG tube leak Evaluate the crew's ability to diagnose and respond to a SG tube rupture Evaluate the crew's ability to respond to a seismic event Evaluate the crew 's ability to diagnose and respond to a failure of the MSIVs to close Initial Conoitions: 12% power, equilibrium xenon, end of cycle (1C-22)
| |
| Tumover: Continue startup at step 6.33 of L-3 Time Event Malf. Event Event min No. No. Type
| |
| * Description 0 1 n/a N, BOP Parallel Turbine Generator to the Grid 15 2 xmt I,RO Loop 4 Delta T Fails high res142 20 3 ovr I, BOP SJAE RM-15R failure due to loss of power xdmiO33p 25 4 sei1 I,SFM 0.1g Earthquake causes loss of DRPI Indication and PY-15 dse rodi dsceps29 ,
| |
| 33 5 mss 3 C, bop PCV 20 fails full open, will close with backup air system l 38 6 res4b M,ALL SG tubeleak 0-200 gpm over 20 minutes l Cond 7 mss 710 C, BOP MSIVs failto close from control room !
| |
| Post pzr3 C,Ro Pzr Spray Valve PCV-455A fails open, will close in manual Tnp
| |
| ]
| |
| l
| |
| * (N)ormal (R)eactivity (1)nstrument (C)omponent (M)ajor l 1
| |
| I i
| |
| | |
| ro .,
| |
| i l ,
| |
| i Appendix D Scenario Outline Form ES-D-2 Op-Test No.: _1_ Scenario No.:
| |
| _ 3 Event No.: 1 Page _.1_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| Parallelina the Main Generator to the Grid '
| |
| Time Position Applicant's Actions or Behavior l
| |
| BOP Parallelthe Generator to the Grid
| |
| . Using OP C-3:ll step 6.3 l
| |
| l R0 Monitor primary plant parameters i i
| |
| l l <
| |
| l l
| |
| SFM Prepare for paralleling the Main Generator to the Grid by reviewing OP L-3 with crew Direct BOPCO to parallel the Main Generator to the Grid per OP C-3:ll step 6.3 NUREG-1021 ~1 Interim Rev. 8, January 1997 -
| |
| | |
| Appendix D Scenaric Outline Form ES-D-2 Op-Test No.:_1,_ Scenario No.: 3 Event No.: 2 Page _2_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| Loop 4 AT fails high Time Position Applicant's Actions or Behavior BOP Recognize and report loop 4 AT failure (TI-441 A)
| |
| RO Take actions as directed by SFM e Deselectloop 4 for ATinput SFM Go to PK04-01, RCL AT Deviation Go to OP AP-5 and direct operator recovery actions Direct RO to defeat only AT forloop 4 Consult Technical Specification 3.3.1
| |
| . 6 hour action to trip bistables for OPAT and OTAT trips Direct Control Room Asset team to investigate failure Loop 4 AT NUREG 1021 2 Interim Rev. 8, January 1997
| |
| | |
| i i
| |
| 1 Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 3 Event No.: 3 Page _3_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| SJAE Rad Monitor Failure RE 15R !
| |
| Time Position Applicant's Actions or Behavior l
| |
| BOP Identify and report PK alarms and indications associated with the RE-15 R Failure !
| |
| . RM-15R indications on RAD Monitor panel behind Vertical Boards
| |
| . Reports RM-15R with no power indications and RM-15 normal indications l
| |
| l i
| |
| RO identify and report PK alarms and indications associated with the RE-15 R Failu e
| |
| . PK11-07, SJAE Rad monitor failure )
| |
| SFM Go to PK11-07, SJAE Rad monitor failure Direcioperator actions
| |
| , . BOPCO to SJAE monitor on Rad monitor panel behind Vertical Boards
| |
| . Outside operator to verify position of the SPDS selector switch in Cable Spreading Room l
| |
| Consult ECGS -39.2
| |
| ! . Restore the inoperable raonitor to operable status within 30 days Direct Control Room Asset team to investigate failure of RM 15R NUREG-1021 3 Interim Fw. 8, Januar; 5997
| |
| | |
| m eY Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_.1_ Scenario No.: 3 Event No.: 4 Page 4 of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| Small Earthauake with loss of DRPI and PY-15 Time Position Applicant's Actions or Behavior 4
| |
| BOP Recognize and reports conditions fc- Loss of PY-15
| |
| . Numerous unrelated alarms
| |
| . Feedpump startup stations are dead
| |
| . MSR valve position Indications are dead Determines size of Earthquake and reports to SFM - 0.1 g RO Recognizes and reports loss of Digital Rod Position Indication
| |
| . Loss of DRPI indication on Vertical Board 2
| |
| . Confirms that no power change in progress SFM Go to PK03-25 and directs actions for loss of DRPl
| |
| . Directs local operator to investigate loss of DRPI and place on Backup Per OPA-3:1 Consults Tech Specs
| |
| . Determines 3.0.3 applies until DRPI is restored Go to AP-4, Loss of Vital or Non-vital Instrument AC, section B, and direct actions for loss of PY-15
| |
| . Direct local actions to restore PY-15 with use of backup power supply Direct 80PCO to determine size of carthquake at EFM instrument Direct Control Room Asset team to investigate failure of DRPI and PY-15 power supplies Go to CP-M-4 for Earthquake response I
| |
| d-NUREG-1021 4 Interim Rev. 8 January 1997
| |
| | |
| g 'I Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 3 Event No.: 5 Page _5_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| Steam Generator 10% Steam Dumo failure - Open Time Position Applicant's Actions or Behavior BOP Recognize and report SG 10% PCV 20 failed open
| |
| !TaksMil66sddifict dbySFMF: INR NL74 "dFinksMi@?9s
| |
| : 9. . myx;g_wcnn4 pW n,n;Attemptto n closeanuaByPCV-20 m_5eK ~j%Q b u.a - m.y c a - , -
| |
| 7-a . .. ;%M :;.:m e
| |
| , f.3 Attempt 16 ws .close PCV-20 with Back'u; pair s@pplyfreport valve ~# is closed '
| |
| - -. _m .. . .- ..
| |
| **CNticaltahk9?h/ 4 MMcM!' t.,8 8 a; , Ad6 RO Monitor Reactor Power and Report to SFM SFM Directs actions for 10% dump valve failed open a
| |
| e Direct RO to monitorReactor Power n,gm ,n wn :n. -,- ,
| |
| w . mcm,,7,p VDirectBOPCOtoclosePCV-2,0withcontroHerlm x ~ u- - determine ,
| |
| . :, ,c contronerfailed?/
| |
| 'A, n.
| |
| @;39[DifictBOPCOtocids,eiCf20 CriticalTaikW" ' 4: 7 "BM:kup; air'supplyiystek ~;'g' J^]..,yg 1 ViWM Direct Control Room Asset team to investigate failure of PCV-20 controller NUREG-1021 5 Interim Rev.8 January 1997
| |
| | |
| [ .
| |
| ]
| |
| i Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.: 3 Event No.: .__6 Page _6_. of _8__ j Event
| |
| | |
| == Description:==
| |
| Steam Generator 1-2 teheleak Time Position Applicant's Actions or Behavior i l
| |
| BOP Acknowledge SJAE, Main Steamline high radiation (RM-72) and S/G Blowdown high radiation I alarms and report to SFM
| |
| . Check radiation monitu
| |
| . Check proper valve alignment for S/G blowdown isolation ;
| |
| l l
| |
| Take actions as directed by SFM l
| |
| . Start CCP punip and Shutdown PDP j e isolate letdown RO Recognize symptoms of a SGTR and report to SFM
| |
| . Decreasing PZR level and increasing charging flow Swap charging pumps from PDP to CCP Calculate primary to secondary leak at greater than 50 gpm Take actions as directed by the SFM w wm gem m n7,, wn, ne en - m.m ,7,,n
| |
| * l , ' ':pi sh Perform a Manual Safe. .ty _ Injection : " - ~ o:
| |
| , 7 M, , ,^ .
| |
| "Criticalfask_k $; -
| |
| SFM Go to PK 11-06, SJAE High Rad Go to AP-3, Steam Generator Tube Leak
| |
| . Direct RO to determineleak rate l . Direct 80PC0 toisolateletdown
| |
| . Direct swap of charging pumps Direct RO io' Safety inject and go'to EOP E 0 (when leak"is determined to be N0 gpm)b.
| |
| s '. . . u. . ;
| |
| '"CriticalTask? ,c. < o-i NUREG-1021 6 Interim Rev. 8, January 1997 L-
| |
| | |
| J Appendix D Scenario Outline Form FS-D-2
| |
| . I Op Test No.:_1_ Scenario No.: .3 Event No.: 7 Page _7_ of _8_ '
| |
| Event
| |
| | |
| == Description:==
| |
| SG Tube Ruoturel MSIV failure to close -
| |
| {
| |
| l i
| |
| Time Position Applicant's Actions or Behavior BOP Complete E0P E-0 immediate actons '
| |
| . Verify allvital4kV buses energized Perform early isolation of ruptured S/G 1-2 per EOP E-3
| |
| . identify the ruptured S/G mam -m - , a mmngenwr-se isolatellowfromfl ne$wwwmmmie ruptu SGEm s er~CY37lFCV32)+n
| |
| ** ;9 W' ,;
| |
| - mygg(close F 33 a ,
| |
| g o a mn,gTask g PCriM.a.n.wa+,w nwn- g*wnha ld40 m.s:akW:gm:0
| |
| :$wamxsaa V. Eku mMmms c p?in L. ~ .ML..anauhat
| |
| -u.
| |
| . recognize MSIV FCV-42 failure to close and report to SFM e Attempt to Close all MSIVs, report failure of all MSIVs to close i
| |
| i Perform Appendix E, Secondary and Auxiliaries status i
| |
| 1 RO Complete EOP E-0immediate actions :
| |
| . Verify Rx trip i
| |
| . . VerifyTurbinetrip
| |
| . Check Si actuated i
| |
| Recogninsidi **dtfdj#TP %
| |
| EciodM$@M@VdiiMN5541 ICritickT5skh$w$$Uji6EM$$Idii/SNi2@$P
| |
| % o. . , N9m ~%ggggMy 0$PMOldgyh>g+4:eq@dffjghw;g@,
| |
| Ah -C -
| |
| failed %q
| |
| $h
| |
| @il/E
| |
| %7 4 5 ; Jg;y$ fp-NUREG-1021. 7 Interim Rev. 8.Jaiuary 1997
| |
| | |
| ?g 1
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:_1_ Scenario No.; 3 Event No.: 7 Page _8_ of _8_
| |
| Event
| |
| | |
| == Description:==
| |
| SG Tube Ruoturel MSIV failure to close (Continued)
| |
| Time Position Applicant's Actions or Behavior SFM Enter E-0 Direct RO/BOPCO to perform immediate actions lDidiiit 50PdOltd}ieifoiin Esrly isolati3Edf SG 112WE-3 steps 3 and 44 3%.gumempm j ; .
| |
| -- E , 'eV s
| |
| [.e@
| |
| n/.n. Diriedt .
| |
| .y BOPCO aw.sy n, to ,Wnw Q#;;ggb close w call MSIVs}a ,
| |
| (74
| |
| +
| |
| a mmms# - x~; 'y-f.E Direct Locsl action to'close all MSIVs (Appendix L)k > < ;T
| |
| %x.m .ye4
| |
| ..o %. yn, , e - - m ,
| |
| w". Critical.Ta.s.kE n W-ra a 3"a.auL=n.;
| |
| . . au.
| |
| a.;.-
| |
| .w , n ., w -
| |
| Direct BOP to perform Appendix E Secondary and Auxiliades status e ~ ,m .. ,. .. m y m .-m. - . . _.
| |
| Direct ROl,to close' failed open PZR SpfaWalve.PCV 455A manually ;
| |
| MddticaNTahk/ -W'' . ,
| |
| Transition to EOP E-3 based on high radiation (Steamline, SJAE, or Blowdown) e conduct tailboard prior to entering E-3
| |
| . may elect to conduct eady isolation following immediate actions of E0P E-0
| |
| . Direct Monitoring of CSFSTs Direct transition to E0P ECA-3.1 based on ALL MSIVs failed open (RNO for Step 3 of E0P E-3) e conduct tailboard prior to entering EOP ECA-3.1 e initiate RCS cooldown l
| |
| NUREG-1021 8 Interim Rev. 8, January 1997 L~
| |
| | |
| b 4 i
| |
| SCENARIO 04 OVERVIEW 4
| |
| Control operator performs of STP R-1 A (Event 1), Exercising full length control rods.
| |
| Condensate system bypass valve FCV-55 will fail open (Event 2). The RO and BOPCO should recognize the symptoms and identify the failure and inform the t'FM of the transients affects on the plant. SFM should direct the BOPCO to attempt to c:aae FCV-55 from the control room. SFM should direct the CO to reduce turbine load to reduce reactor power and direct local isolation of FCV-55.
| |
| Pressurizer master controller HC-455K fails high (Event 3), causing pressurizer spray valves to open and pressurizer pressure to drop. CO should recognize the failure and inform the SFM. The SFM should enter Annunciator Response procedure PK05-17 and transition to OP AP-13 Malfunction of Reactor Pressure cord.ml syste-.. Pressurizer spray valves sh ould be manually closed prior to a low pressure Reactor Tip. Pzr Master controller should be controlled in manual and spray valves retumeo 'o auto.
| |
| The running Aux. Salt Water pump 1-1 trips on Overcurrent (Event 4). The Overcuurent trip also causes an Overcurrent Trip on 4Kt bus F causing the loss of the 4Kv bus. The CO and BOPCO should recognize losses and report them to the SFM. The SFM should enter Annunciator Response procedure PK01-01 and OP AP-10 to restore the ASW system. The SFM should direct the CO/BOPCO to start the standby ASW pump.
| |
| The SFM should reference Annunciator R 3sponse procedure PK18-17 and PK18-22 and direct the actions for the failure of the 4Kv bus F loss including the start of CCW Pp.1-3 A 500 Kv system disturbance will cause a Full load rejection (Event 5). The CO and ;
| |
| BOPCO should recognize loss and report to the SFM. SFM should reference Abnormal procedure OP AP-2 and direct the actions of the CO/BOPCO per the procedure. Plant !
| |
| - will not survive the Load Rejection and Emergency procedure E-0 should be entered.
| |
| During performance of initial steps of E-0 Startup Power will be lost along with the a failure of DSL Gen 1-2 output Bkr to close (Event 6). SFM should recognize the entry conditions for ECA 0.0 and direct the CO/BOPCO in actions for Loss of All AC Power.
| |
| SFM should direct the Manual closing of DSL Gen 1-2 output Bkr and the restoration of AC power to Bus G. SFM should direct the CO to start CCP1-2 which failed to auto start on bus restoration. SFM should determine that with one bus operable a transition to E-0 should be done and direct BOPCO to perform ECA 0.3 actions to restore additional buses.
| |
| During the initial Load rejection a Main Steam line safety valve will stick open (Event 7).
| |
| The BOPCO should determine that the 1-4 Steam generator is not intact and inform the SFM. The SFM should direct the RO to perform an early isolation of SG 1-4. SFM should determine the need to transition to EOP E-2 for the faulted SG. SFM should determine that a transition to EOP E-1 is required at the completion of EOP E-2.
| |
| The scenario will be terminated following transition out of EOP E-1 to EOP E-1.1 Si Termination.
| |
| | |
| t i Appendix D OperatorActions Form ES-D-1 4
| |
| Facility: DCPP Units 1 & 2 Scenario No.: 4 Op-Test No.: 1A/1B Examiners: Operators: SRO-U / SRO-U SRO-U / SRO-U SRO-U / SRO-l .
| |
| Objectives: Evaluate the crew's ability to diagnose and respond to a Condensate system disturbance l Evaluate the crew's ability to diagnose and respond to a Pressurizer master controller failure
| |
| ]
| |
| Evaluate the crew's ability to diagnose and respond to a ASW pump 0. C. Trip j Evaluate the crew in using the AOP and EOP to respond to full load rejection i Evaluate the crew's ability to diagnose and respond to a Loss of startup power l Evaluate the crew's ability to diagnose and respond to a loss of all AC event Evaluate the crew 's ability to diagnose and respond to a safety vaive failure j i
| |
| InitialConditions: 100% power, equilibrium xenon, Beginning of cycle (IC-1) !
| |
| Tumover: D/G 1-1 is 00S, for Lube oil heater replacement, CFCU ' 4 failed 10 minutes ago, TS j have not been addressed Time Event Malf. Event Event min No. No. Type
| |
| * Description 0 1 none N. Co STP R-1 A Exercising Full Length Control Rods 12 2 cnd4 R Ro FCV-55 fails open ,
| |
| 20 3 pzr4 1, Ro Pressurizer Master pressure controller HC-455K fails High - manual I available 28 4 aswi C, bop ASW pp 1-10.C. trip also trips 4Ky F Bus on 0.C., no auto start on j asw2 standby ASW pp1-2, can be started manually. !
| |
| eps4c,1 37 5 syd1 M,ALL 100% full load rejection due to 500 Kv system disturbance 38 6 syd1 M ALL Loss of startup power due to system disturbance ,
| |
| eps16 C, bop Diesel generator 1-2 output Bkr fails to close, can be closed manually I Post trip 7 cyc2 C,Ro CCP 12 fails to auto start on retum of Bus G ,
| |
| l Post trip : mss 6d C,Ro Main steam safety valve lead four RV-58 stuck open
| |
| * (N)ormal (R)eactivity (I)nstrument (C)omponent (M)ajor
| |
| | |
| D-i Appendix D Scenario Outline Form ES-D-2 1 l
| |
| Op-Test No.:__i_ Scenario No.: 4 Event No.: 1 Page _1_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| STP R-1 A Exercising Full Length Control Rods Time Position Applicant's Actions or Behavior BOP Monitors DRPIindications RO Performs STP R-1 A, exercising full length control rods
| |
| . Shutdown banks A, B, C, D j l
| |
| . Control banks A, B, C, D 1
| |
| Records data on Attachment 8.1 of STP R-1 A SFM Review precautions and limitations of S TP R-1 A and conduct tumover briefing Directs RO to commence exercising full length control rods Provides direct oversight of control rod manipulations NUREG-1021 1 Interim Rev. 8, January 1997
| |
| | |
| X e
| |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:__1_ Scenario No.: 4 Event No.: 2 Page _2_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Condensate Bypass Valve FCV-55 fails open Time Position Applicant's Actions or Behavior BOP Report and acknowledge Feedwater Htrs Level Hi and MSR Shell Drips alarms Recognize and report FCV-55 failed open I Attempt to close FCV-55 as directed by SFM and report failure of valve to close I 1
| |
| l RO Recognize and report increase in Reactor Power > 100%
| |
| Retum Rod Control system to Automatic l Operate DEHC to reduce turbine load to reduce Reactor power to <100% l
| |
| . Place MW feedbacksin service 3
| |
| . Enter ramp rate and target megawatts
| |
| . Press go when directed by SFM I
| |
| Monitor and report reactor power trend to SFM during ramp I
| |
| \
| |
| SFM Go to PK-10-16 Feedwater Heaters Level High l
| |
| . Dispatch local operator to check feedwater heaters levels Direct RO to Place Rod Control to automatic Direct RO to reduce turbine load to maintain reactor power <100%
| |
| Direct actions to close FCV-55
| |
| . Direct BOPCO to close FCV-55 at VB-3
| |
| . Direct Local operators to investigate FCV-55 locally and manually isolate Contact Turbine Building Asset Team for investigation and repair NUREG-1021 2 Interim Rev. 8. January 1997
| |
| | |
| Appendix D Scenario Outline Form ES-D-2 Op-Test No.:._._1_ Scenario No.: 4 Event No.: 3 Page _3_ of _7_
| |
| l Event
| |
| | |
| == Description:==
| |
| Pressurizer Master Controller failure -HC-455K Time Position Applicant's Actions u Behavior BOP Performs checks on Pressurizer parameters as directed by SFM e Monitor and report all pressurizer channels Channel check SAT
| |
| . Monitor and report PORVs and PZR safeties all closed RO Recognize and report Low Pressurizer Pressure alarms Recognize and report conditions of pressurizer controls l 4
| |
| . PZR Htrs off 4 1
| |
| . PZR spray valves open !
| |
| . Pzr master controller HC-455K on CC2 above expected or at maximum demand j t
| |
| : Place HC-455K'iman0al'is rehtorss pressurizer pressure tb NOR Ese MQS;Y ,
| |
| ..m .~,m- . w. ~ . , , -
| |
| a a "CriticalTaskGTC, w -a, mm 2-7B M. . :U * -
| |
| NNsb sc:1 ~ * ,
| |
| . Y ;ll6l$?%%%l$l,p ;;c cO?% J & &: a k&f s?nwn?mW , '5l?
| |
| SFM Go to PK 05-17 Pressurizer Pressure Low Go to AP-13 Malfunction of Reactor Pressure Control System Direct BOP and 10 to monitor Pressurizer control parameters Diagnose failure of HC-455k to control pressure in automatic Direct R+0ilo
| |
| .e PldNC455K a O, IMindal as , , ss , ,,
| |
| . - h$g.k #y MMk$v si Mhhhh!
| |
| "OnticalTEsihNNN~h hh [$5$M$NretUrn
| |
| ;hhkhpkh .' Pres Contact Control Room Asset Team for investigation and repair of HC-455K I
| |
| NUREG-1021 3 Interim Rev. 8..hiuary 1997
| |
| | |
| n >
| |
| l p u a .. a ..~, l 1
| |
| l4 i
| |
| 'I' Appendix D Scenario Outline Form ES-D-2 1
| |
| Op-Test No.:_1_ Scenario No.: 4 Event No.: 4 Page _4_ of _7_ l l
| |
| Event
| |
| | |
| == Description:==
| |
| ASW PP1-10.C. trio and 4 Kv Bus F 0.C. trio 1 Time Position Applicant's Actions or Behavior-BOP Recognize and report ASW 1-1' trip blue light on e' Inform SFM of pump trip
| |
| . Identify 4Kv Bus F Overcunenttrip
| |
| . Identify need to place altcmate equipment in service
| |
| 'dommimoe'plai:inhTallenilsiglii
| |
| %gwp fASdR ag m . i$$w$$gu;@;w$pineshi$seir E Y '
| |
| % en0CWpumph%
| |
| W d I f $ $p $ $ %g @w ggggh.;nw,p.p::M, 4
| |
| w n_, ' &g::,NW%
| |
| n? pt w y"+,
| |
| %. 4''
| |
| aem by w. ~m . . , ~gxx ICdlica!Tsk$$m$l$$$dN!@I$w$55GQN,%a TP i 'Oh D Secure DieselGenerator1-3 RO Recognize and report loss of 4Kv Bus F components
| |
| . DRPI e CFCUs 1-1 & 1-2
| |
| . CCW pp 1-1
| |
| . ASW pp 1-1
| |
| . Battery Charger 1-1
| |
| ^ :
| |
| SFM Acknowledge BOP / CO on loss of 4Kv bus F
| |
| . Reviews actions in Annunciator Response Procedure (PK18-16,17 and 22) for loss of 4kV bus F and loss of 480V bus F.
| |
| t Direct starting / placing in service altemate equipment
| |
| ! . ASW pp 1-2 and CCW pp 1-3
| |
| . Place DRPIon back up
| |
| . Battery 1-1 to charger 1 . Securing of DieselGenerator 1-3 Contact Control Room /Electncal Asset Team for investigation and repair Bus F Refer to Technical Specifcations 3.0.3 I . Applies until DRPI is placed on Backup
| |
| .. Applies due to multiple Train A & B components OOS with Diesel Generator 00S NUREG-1021 4- Interim Rev.8, January 1997
| |
| | |
| r 1
| |
| :o ,
| |
| Appendix D Scenario Outline Form ES-D-2 l
| |
| Op-Test No.:_.1_ Scenario No.: 4 Event No.: 5 Page _5_ of _7_ l Event
| |
| | |
| == Description:==
| |
| 100% Fullload Rejection Time Position App 0cantb Actions or Behavior BOP Recognizes and reports all steam dump valves open Recognizes and reports ReactorTrip RO Recognizes and reports Main Generator Output bkrs open (Full Load rejection)
| |
| Monitors Plant Parameters
| |
| . Rod Control- Rods stepping in in Automatic
| |
| . Reactor Power decreasing
| |
| . Steam GeneratorWater Levels Recognizes and reports Reactor Trip -
| |
| ! SFM Go to AP-2 Full Load rejection l
| |
| Go to E-0 Reactor Trip NUREG-1021 5 Interim Rev. 8, January 1997 l
| |
| | |
| A?
| |
| A Appendix D Scenario Outline Form ES-D-2 op-Test No.:_1_ Scenario No.: 4 Event No.: 6 Page _6_ of _7_
| |
| Event
| |
| | |
| == Description:==
| |
| Loss of All AC Power and Bus G restoration Time Position Applicant's Actions or Behavior DOP Performimmediate actions of EOP E-0 identify loss of Startup power to all Buses identify loss of all DIGS resulting in a loss of power to all Buses Take action as directed by the SFM per E0P ECA-0.0
| |
| . Verify AFW flow n -- m-nn. ~.ww -,n , n , v.wan 9 e p,,?;Clo,s,,,52-HG u . emm w u p g m .. . w 05 Diesel Genera,torOutput Bkr/ manually; e ' g L Og, . v- -,
| |
| a ~
| |
| iCriticalTssif5iWldWM 3Mdg. r ,. hN
| |
| % W6 -
| |
| Perform actions to crosstie 4 Ky Buses per ECA 0.3 RO Performimmediate actions of E0P E-0 l . Verify Rx Tnp
| |
| . Verify Turbine Trip ;
| |
| . 4kV buses energized Take action as directed by the SFM per E0P ECA-0.0
| |
| * Check RCSisisolated
| |
| -_ ,- -- n:.,w,nym,mm, n,w 3ok n,,Res, n,
| |
| tart CCP 1-2 after, Busy restoration W wwd aang m. u u ::3 < v : w. 710-e um.,,KRO >-
| |
| W mn n +_ v.
| |
| '"CribcalTaskhdgh ~ WP i : ' 1A a ;*M 1" ' ,
| |
| w W+ d M iMM %
| |
| l SFM Directimmediate actions of E0P E-0 l Acknowledge loss of all power and transitions to EOP ECA-0.0 i
| |
| Direct the RO I BOP as per the actions of ECA-0.0 l
| |
| . Verify Rx Trip
| |
| . Verify Turbino Trip
| |
| . Verify PORVs closed and Letdown Isolated
| |
| . Verify AFW flow
| |
| -- ., .nnn,, --n . ,w. ,m Close52
| |
| &,m..,[,n.-, M_NT 7,W M ~ ,s f urm 4HG:05, Diesel Generator Ou,tput Bkrman
| |
| &_ Critical;TasNgpgggfpegg ;gggggwgp a gie u M w w cenguiu msmgah ggeg,s
| |
| 'yggggwux Direct BOPCOlRO c kiverifyAutdisstic v.waw w. 16edi6sof 4KinM GMMSBMMsi@@@
| |
| f e w 0 sect R O . a~CCR;1-2 pWiE4W MAaiyisi@ JM@@@$@s4pv Ng?W $d9, p!!pggumsgg$
| |
| WR Transition to EOP E-0, Direct BOPCO to implement ECA 0.3 NUREG 1021 6. Interim Rev. 8. January 1997
| |
| | |
| r ,
| |
| 4 l
| |
| i-i l .
| |
| l l Appendix D Scenario Outline Form ES-D-2 j l
| |
| Op Test No.:_1_ Scenario No.: 4 Event No.: 7__ Page _7_ of _7_
| |
| ]
| |
| Event
| |
| | |
| == Description:==
| |
| Main Steam Line Safety Valve Failure Time Position Applicant's Actions or Behavior l l
| |
| BOP Recognize and report main steam header break indications:
| |
| l
| |
| . . Increasing steam flow on SG 1-4
| |
| . Decreasing steam pressure on SG 1-4 Perform E0P E-0immediate actions i
| |
| e Verify power to AC emergency buses i e Check SI actuation RO Perform E0P E-0 immediate actions e Verify reactor trip e - Verify turbine trip e Verify powerto AC emergency buses e Check Si actuation e
| |
| Recognize and report main steam header break indications:
| |
| e increasing steam flowon SG 1-4
| |
| ~
| |
| ilsolate m- thle ,- fiuitsd' SGiaidiiecteci
| |
| .a ;,-,by the SFM (Close IWSIV FCV44/AFW xcm.m.supplylCViT3,10 w
| |
| s 7, '
| |
| , . , f - .
| |
| gl . .l ~ * , l '
| |
| SFM Direct R0 / BOP to check for steam leak indications
| |
| . Steam Generator pressures Enter EOP E-0 and direct operator actions Tran m sitionsto
| |
| .mem:EOP E-2andduectisolation,of
| |
| : w. ,yv S/G 1-4 ,+m $= ,T. : 'i % a
| |
| 'w
| |
| '(may)aviiibeen1solated y)pp op A; - nM@if,-@g % f~ Ms W
| |
| &gpg&%sEMW.aahuakaanons&wamauxWb.sa'
| |
| : O. &Haasukwua-e conduct tailboard prior to entering EOP E-2 j- Direct transition to EOP E-1 based on Loss of Secondary Coolant l . conduct tailboard prior to entering EOP E-1 i
| |
| Direct transition to E0P E-1,1 based on ECCS flow reduction e conduct tailboard prior to entering EOP E-1.1.
| |
| NUREG-1'0 21 7 interim Rev. 8. January 1997
| |
| | |
| 1 , ,
| |
| REVISION 1 ES-301 Administrative Topics Outline Form ES-3011 Facility: DCPP Units 1&2 Date of Examination: _ _01/25/99 Examination Level: __SRO Operating Test Number i Administrative Describe method of eva!9ation:
| |
| Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions A.1 Conduct of G2.1.23 (4.0) Abiftty to perform specific system integrated plant procedures Operations during allmodes of plantoperation.
| |
| (JPM ADMNRC-1) Review an (faulted) ECP G2.1.20 (4.2) Abilityto execute procedure steps ,
| |
| (Ques) Required steps when SFM suspects current procedure is inappropriate for plant conditions.
| |
| G2.1.12 (4.0) Ability to apply technical specifications for a system (Ques) Application of mode transition times and allowed out of service time (A0T)
| |
| A.2 Equipment Control G2.2.13 (3.8) Knowledge of tagging and clearance procedures (JPM ADMNRC-2) Review completed (faulted) clearance for FCV-95 A.3 Radiation Control G2.3.1 (3.0) Knowledge of 10CFR20 and related facility radiation control f requirements.
| |
| (Ques) Authorizationforreceiving emergencyexposurelimits G2.3.4 (3.1) Knowledge of radiation exposure limits and contamination control including permissible levels in excess of those authorized. j (Ques) Interpret DCPP administrative radiation control requirements A.4 Emergency Plan G2.4.41 (4.1) Knowledge of the emergency a:V . level thresholds and classifications j I
| |
| (JPM ADMNRC-3) Classify a loss of decay heat removal NUREG-1021 21 OF 26 Interim Rev. 8, January 1997
| |
| | |
| NUCLEAR POVER GENERATION DIABLO CANYON PO9fER PLANT JOB PERFORMANCE MEASURE Number: ADMNRC-1 :
| |
| l
| |
| | |
| ==Title:==
| |
| REVIEW AN ESTIMATED CRITICAL POSITION l Examinee:
| |
| Evaluator:
| |
| Print Signature Date i
| |
| Results: Sat Unsat Total Time: minutes Comments: Provide the " Student Copy" of the completed STP R-17 Attachment 8.2 for review.
| |
| | |
| ==References:==
| |
| STP R-17, Estimated Critical Position, Res.12A Vol. 9B: Fig. R17-1F-3, Rev. I1; Fig. R17-IT-2, Rev.13; Fig. R17-lT-3, Rev. 8; Fig. R4-1 F-4, Rev. b; R17-1 F-2, Rev. 8; R17-lT-4, Rev. 9 l
| |
| Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 25 minutes Critical Steps: 2 - 9, and 11 - 18 Job Designation: SRO
| |
| /
| |
| K/A Number G2.1.23 (4.0)
| |
| AUTHOR: JOHN BECERRA DATE: 11/17/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REv.O
| |
| | |
| f 1 i
| |
| l JPM TITLE: REVIEW AN ESTIMATED CRmCAL PoSmON JPM NUMBER: ADMNRC-1 INSTRUCTOR WORKSHEET Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are '
| |
| needed. The examinee may be given the procedure and told the step with which to begin. ;
| |
| l Required Materials: Calculator and a complete STP R-17, Attachment 8.2 (ECP).
| |
| ]
| |
| l Initial Conditior.s: Unit 1 tripped from 100% power 28 hours ago and is in IIOT )
| |
| STANDBY with shutdown banks withdrawn. Core burnup is 6000 l MWD /MTU !
| |
| Initiating Cue: Reactor Engineering delivered an ECP to the control room. A reactor startup is scheduled for two hours from now. You are to review and approve the ECP.
| |
| l Task Standard: The ECP is reviewed and approved after finding and correcting the three (3) technical errors.
| |
| 1 NOTE: Recalculations due to errors are not considered " technical" errors.
| |
| l ADMNRC-1 PAGE 2 OF 8 REv. O
| |
| | |
| i JPM TITLE: REVIEW AN ESTIMATED CRITICAL POSITION JPM NUMBER:- ADMNRC-1 lNSTRUCTOR WORKSHEET Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. Obtain the correct procedure. 1.1 Operator references STP R-17 and Volume 9.
| |
| Step was: Sat: Unsat *
| |
| : 2. *
| |
| * Check reference power defect. 2.1 Operator verifies the reference power defect to be +1165 pcm from Tabic R17-l F-3.
| |
| Step was: Sat: Unsat *
| |
| : 3. *
| |
| * Check reference xenon worth. 3.1 Operator verifies the reference xenon worth to be -2901 pcm from Table R17-lT-2.
| |
| Step was: Sat: Unsat
| |
| * I
| |
| : 4. *
| |
| * Check estimated xenon worth at 4.1 Operator verifies the estimated xenon !
| |
| criticality, worth to be -1885 pcm from Table R17-lT-2. l Step was: Sat: __ __ Unsat
| |
| * l
| |
| : 5. *
| |
| * Calculate net change in xenon worth. 5.1 Operator calculates the net change in xenon worth to be +1016 pcm.
| |
| 5.2 Determines Step C.4 in error due to incorrect sign.
| |
| Step was: Sat: Unsat *
| |
| : 6. ** Check reference samarium worth. 6.1 Operator determines the reference samarium worth to be -943 from Table R17-lT-3 by interpolation.
| |
| Step was: Sat: Unsat *
| |
| ' Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
| |
| ADMNRC-1 PAGE 3 OF 8 REV.O
| |
| | |
| JPM TITLE: REVIEW AN ESTIMATED CRITICAL POSITION JPM NUMBER:- ADMNRC-1 INSTRUCTOR WORKSHEET
| |
| : 7. *
| |
| * Check estimated samarium worth at 7.1 Operator determines the estimated criticality. samarium worth to be -1069 pcm from Table R17-1T-3 by interpolation.
| |
| 7.2 Determines Step C.6 in error due to transposition error (1096 vice 1069).
| |
| Step was: Sat: Unsat
| |
| * I
| |
| : 8. ** Calculate net change in samarium 8.1 Operator recalculates the net change worth. in samarium worth to be -126 pcm.
| |
| Step was: Sat: Unsat *
| |
| : 9. ** Check reference integral rod worth. 9.1 Operator verifies the reference integral rod worth to be 0 to -5 pcm from Figure R4-1F-4. l I
| |
| Note: Operator must use Figure R4-1F-4 for hot full power reference conditions vs. Table R17-1T-4, which is for hot zero i power.
| |
| l l Step was: Sat: Unsat *
| |
| : 10. Check critical rod position. 10.1 Operator agrees with a critical rod l position of control bank D at 150 steps.
| |
| Step was: Sat: Unsat *
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
| |
| ADMNRC-1 PAGE 4 OF 8 REV.O
| |
| | |
| l
| |
| * JPM TITLE: REVIEW AN ESTIMATED CRITICAL POSITION JPM NUMBER:- ADMNRC-1 ;
| |
| INSTRUCTOR WORKSHEET
| |
| {
| |
| i
| |
| : 11. ** Check integral rod worth. I1.1 Operator verifies the integral rod ;
| |
| worth for control bank D at 150 steps i to be -202 pcm from Table R17-lT-4.
| |
| Note: Operator must use Table R17-1T-4 for this step and not Figure R4-IF-4 which was used in Step 9.
| |
| I 1.2 Determines Step C.10 in error due to misread table. I Step was: Sat: Unsat
| |
| * I
| |
| : 12. *
| |
| * Calculate net change in integral rod 12.1 Operator recalculates the net change I worth. in ir.tegral rod worth to be -197 pcm.
| |
| 1 Step was: Sat: Unsat
| |
| * l
| |
| : 13. ** Calculate the net reactivity changes. 13.1 Operator recalculates the sum of the net reactivity changes to be +1858.
| |
| I Step was: Sat: Unsat *
| |
| : 14. ** Calculate preliminary change in 14.1 Operator recalculates preliminary boron concentration. change in boron concentration to be
| |
| +206 ppm.
| |
| Step was: Sat: Unsat *
| |
| : 15. ** Calculate preliminary boron 15.1 Operator recalculates the preliminary concentration. boron concentration to be 1506 ppm.
| |
| Step was: Sat: Unsat
| |
| * l l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
| |
| ADMNRC-1 PAGE 5 or 8 REV.O
| |
| | |
| JPM TITLE: REVIEW AN ESTIMATED CRmCAL POSITION JPM NUMBER:- ADMNRC-1 INSTRUCTOR WORKSHEET
| |
| : 16. ** Check actual ditTerential boron 16.1 Operator determines the actual l worth. differential boron worth to be -7.7 ;
| |
| pcm/ ppm from Figure R17-lF-2.
| |
| Note: -7.7 pem/ ppm is based on 6,000 MWD /MTU. Thus the value is l taken 1/4 of the way between the ;
| |
| BOL and EOL cun es.
| |
| Step was: Sat: Unsat *
| |
| : 17. ** Calculate the boron concentration 17.1 Operator recalculates the estimated change to offset the net reactivity change in boron concentration to be change. +241 ppm.
| |
| Step was: Sat: Unsat *
| |
| : 18. " Calculate the estimated critical boron 18.1 Operator recalculates the estimated !
| |
| concentration. critical boron concentration to be 1541 ppm.
| |
| Step was: Sat: Unsat *
| |
| : 19. Analyze acceptance criteria. 19.1 Operator checks the YES box for rod ;
| |
| insertion limit criteria.
| |
| 19.2 Checks the YES box for withdrawal limits.
| |
| Step was: Sat: Unsat
| |
| : 20. Complete Shift Foreman review. 20.1 Operator completes Item A.6.
| |
| I 20.2 Signs name and enters the date and time.
| |
| 1 Step was: Sat: Unsat Stop Time:
| |
| Total Time: (Enter total time on the cover page)
| |
| * Denotes an entrv required on the JPM cover sheet.
| |
| " Denotes a Critical Step.
| |
| ADMNRC-1 PAGE 6 OF 8 REv. O
| |
| | |
| JPM TITLE: JPM NUMBER: ADMNRC-1 EXAMINEE CUE SHEET Initial Conditions: Unit I tripped from 100% power 28 hours ago and is in HOT STANDBY with shutdown banks withdrawn. Core burnup is 6000 MWD /MTU Initiating Cue: Reactor Engineering delivered an ECP to the control room. A reactor startup is scheduled for two hours from now. You are to review and approve the ECP.
| |
| Task Standard: The ECP is reviewed and approved after finding and correcting the three 1 (3) technical errors. l l
| |
| NOTE: Recalculations due to errors are not considered " technical" errors.
| |
| l l
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| l l
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| ADMNRC-1 PAGE 7 OF 8 REV.O L
| |
| | |
| O NUCLEAR POWER GENERATION j DIABLO CANYON POWER PLANT
| |
| ' l JOB PERFORMANCE MEASURE Number: ADMNRC-2 !
| |
| I
| |
| | |
| ==Title:==
| |
| REVIEW A CLEARANCE FOR TECHNICAL ACCURACY I l
| |
| 1 Examinee.
| |
| i Evaluator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
| |
| | |
| ==References:==
| |
| Operator Valve Identification Diagram,106704, Sheet 4 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 15 minutes Critical Steps: 2 Job Designation: SRO K/A Number G2.2.13 (3.8)
| |
| AUTHOR: JOHN BECERRA DATE: 11/11/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| r- !
| |
| JPM TITLE: REVIEW A CLEARANCE FOR TECHNICAL ACCURACV JPM NUMBER: ADMNRC-2 INSVRUCTOR WORKSHEET
| |
| )
| |
| I Directions: No plant controls or equipment are to be operated during the performance of this Job Performance Measure. All actions taken by the examinee should be clearly demonstrated and verbalized to the evaluator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin. '
| |
| Required Materials: Clearance NBR: 00034803, " Install and Remove Test Equipment on FCV-95 Initial Conditions: Unit 1 is 100% power, steady state.
| |
| Initiating Cue: Perform a review of all clearance points on Clearance NBR: 00034803, Install and Remove Test Equipment on FCV-95, for technical errors.
| |
| 1 Task Standard: Find and correct three (3) technical clearance errors within the clearance points.
| |
| 1 I
| |
| i l
| |
| l l
| |
| ADMNRC-2 PAGE 2 OF 5 REV.0 1
| |
| | |
| r JPM TITLE: REVIEW A CLEARANCE FOR TECHNICAL ACCURACY JPM NUMBER: ADMNRC-2 INSTRUCTOR WORK. SHEET Start Time:
| |
| Step Expected Operator Actions
| |
| : 1. Obrain the correct reference material.
| |
| Cue: Provide operator with clearance sheet.
| |
| 1.1 Operator selects Operator Valve Identification Diagram (OVID).
| |
| 1.2 Selects Section 106704, Sheet 4.
| |
| Note: Operator may obtain valve and breaker number using optional j reference material.
| |
| Note: Clearance Legend TAGS:
| |
| MAN-ON-LINE (M)
| |
| CAUTION (C)
| |
| CONTROL BOARD (CBC)
| |
| POSITIONS:
| |
| RACKED OUT (RO)
| |
| RACKED IN (RI)
| |
| CLOSED (CL)
| |
| OPEN (OP)
| |
| OPEN & UNCAPPED (OPU)
| |
| Step was: Sat: Unsat
| |
| * i l !
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| ADMNRC-2 PAGE 3 OF 5 REV.O
| |
| \
| |
| 4 t
| |
| | |
| JPM TITLE: REVIEW A CLEARANCE FOR TECHNICAL ACCURACY JPM NUMBER: ADMNRC-2 INSTRUCTOR WORKSHEET
| |
| : 2. " Identify MS-1-FCV-95 clearance 2.1 Determines point #4 tag should be a erTors. MAN-ON-LINE (M) tag.
| |
| 2.2 Determines point #5 breaker should be 52-1F-30.
| |
| 2.3 Determines point #7 valve (MS FCV-37) position should be closed.
| |
| Step was: Sat: Unsat
| |
| * l Stop Time: l Total Time: (Enter total time on the cover page) l i
| |
| 1 l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| l ADMNRC-2 PAGE 4 OF 5 REV.O
| |
| | |
| I l
| |
| $ . JPM TITLE: JPM NUMBER: ADMNRC-2 EXAMINEE CUE SHEET Initial Conditions: Unit 1 is 100% power, steady state.
| |
| Initiating Cue: Perform a review of all clearance points on Clearance NBR: 00034803, Install and Remove Test Equipment on FCV-95, for technical er ars.
| |
| Task Standard: Find and correct three (3) technical clearance errors within the clearance i points.
| |
| l l
| |
| l ADMNRC-2~ PAGE 5 OF 5 REV.O
| |
| | |
| i NUCLEAR POWER GENERATION DIABLO CANYON POWER PLANT Joe PERFORMANCE MEASURE i
| |
| Number: ADMNRC-3
| |
| | |
| ==Title:==
| |
| CLASSIFY A LOSS OF DECAY HEAT REMOVAL Examince:
| |
| Evalaator:
| |
| Print Signature Date Results: Sat Unsat Total Time: minutes Comments:
| |
| | |
| ==References:==
| |
| OP AP SD-0, Rev. 7A EP G-1, Accident Classification and Emergency Plan Activation, Rev.28 Alternate Path: Yes No X Time Critical: Yes No X Time Allotment: 5 minutes Critical Steps: 2,3,4,5,6 Job Designation: SRO IUA Number: G2.4.41 (4.1)
| |
| AUTHOR: JOHN BECERRA DATE: 11/11/98 REVIEWED BY: NA DATE: NA JPM COORDINATOR APPROVED BY: NA DATE: NA TRAINING LEADER REV.O
| |
| | |
| b 'O JPM TITLE: CLASSIFY A LOSS OF DECAi HEAT REMOVAL JPM NUMBER: ADMNRC-3 lNSTPUCTOR WORKSHEET Directions: No ph.:n controls or equipment art to be operated during the performance of this Job Performance Measure. All actions taken by tb . .uninee should be clearly demonstrated and verbalized to the
| |
| : c. .sator. The student will be given the initial conditions, initiating cue, and task standard. The examiner will then ask if any clarifications are needed. The examinee may be given the procedure and told the step with which to begin.
| |
| Required Materials: Calculator Initial Conditions: Unit 1 is in Mode 5,27 days into a refueling outage. Core reload has been completed, replacing 1/3 of the core with new fuel. Completed tensioning the RX vessel head while filling the RCS per OP A-2:llI with level @ l14". A common mode failure on RHR pp breakers caused both RHR pps to trip 15 minutes ago. Pumps were shutdown 15 minutes ago. Core exit thermocouple readings are presently 110 F.
| |
| Initiating Cue: The Shift Supervisor has directed you to classify the event and to report the event classification to him.
| |
| , Task Standard: Determine the proper event classification and report the classification to the Shift Supervisor.
| |
| ADMNRC-3 PAGE 2 OF 5 REV.0
| |
| | |
| JPM TITLE: CLASStFY A LOSS OF DECAY HEAT REMOVAL JPM NUMBER: ADMNRC-3
| |
| , INSTRUCTOR WORKSHEET Start Time:
| |
| 3,,p Expected Operator Actions
| |
| : 1. Obtain correct procedure. Note: Operator may reference EP G-1 prior to performing Appendix B.
| |
| 1.1 Operator references OP AP SD series procedures, Appendix B.
| |
| Note: All six of the OP AP SD series procedures contain Appendix B.
| |
| Step was: Sat: Unsat *
| |
| : 2. *
| |
| * Determine the predicted heat load. 2.1 Operator estimates heat load from Appendix B curve for 27 days after shutdown to be 6 MW (i0.5 MW).
| |
| Step was: Sat: Unsat *
| |
| : 3. ** Determine the reduction factor for 3.1 Operator uses predicted heat load refueled cores. from previous step.
| |
| 3.2 Uses 0.67 as the fraction of previously used assemblies installed in the core (0.66 to 0.67L 3.3 Calculates estimated decay heat load to be 4MW (3.6 - 4.4).
| |
| Step was: Sat: Unsat *
| |
| : 4. ** Predict the heat up rate. 4.1 Operator uses estimated decay heat load from previous step.
| |
| 4.2 Determines the inventory factor to be 0.33.
| |
| 4.3 Calculates predicted heat up rate to be 1.32 F/ min (1.22 to 1.42 F/ minute)
| |
| Step was: Sat: Unsat *
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| *
| |
| * Denotes a Critical Step.
| |
| ADMNRC-3 PAGE 3 OF 5 REV.0
| |
| | |
| r JPM TITLE: CLASSIFY A LOSS OF DECAY HEAT REMOVAL JPM NUMBER: ADMNRC-3 l INSTRUCTOR WORKSHEET
| |
| : 5. *
| |
| * Calculate estimated time to reach Operator calculates current
| |
| ~
| |
| 5.1 200 F. temperature difference to be 90 F.
| |
| 5.2 Calculates time to reach 200 F to be 68 min.
| |
| Step was: Sat: Unsat *
| |
| : 6. *
| |
| * Classify the event in accordance with 6.1 Operator determines that RCS EP G-1. thermocouple temperature will not exceed 200 F within one hour of RHR loss.
| |
| 6.2 Determine loss of RHR <1 hr. but >l5 min.
| |
| 6.3 Classifies the event as an NUE #10 in accordance with EP G-1.
| |
| Step was: Sat: Unsat
| |
| * Stop Time:
| |
| Total Time: __ (Enter total time on the cover page) l
| |
| * Denotes an entry required on the JPM cover sheet.
| |
| ** Denotes a Critical Step.
| |
| ADMNRC-3 PAGE 4 OF 5 REv.0 l
| |
| 1
| |
| | |
| . JPM TITLE: JPM NUMBER: ADMNRC-3
| |
| . EXAMINEE CUE SHEET Initial Conditions: Unit 1 is in Mode 5,27 days into a refueling outage. Core reload has been completed, replacing 1/3 of the core with new fuel. Completed tensioning the RX vessel head while filling the RCS per OP A-2:III with level @ l14". A common mode failure on RIIR pp breakers caused both RHR pps to trip 15 minutes ago. Pumps were shutdown 15 minutes ago. Core exit thermocouple readings are presently 110 F.
| |
| Initiating Cue: The Shift Supervisor has directed you to classify the event and to report the event classification to him.
| |
| Task Standard: Determine the proper event classification and report the classification to the Shift Supervisor.
| |
| ADMNRC-3 PAGE 5 OF 5 REv. O
| |
| | |
| t CATEGORY: A.1
| |
| .. QUESTION # 1:
| |
| References Allowed: No During the implementation of an EOP the Shift Foreman informs the Shift Supervisor that he
| |
| ' believes the procedure is inappropriate for the current plant conditions. The Shift Supervisor concurs with the Shift Foreman's assessment of the problem.
| |
| What two (2) checks should be perfonned per OPl.DCl 1 " Conduct of Operations - Abnormal Conditions" by the Shift Foreman and Shift Supervisor prior to the Shift Supervisor authorizing the return to the diagnosis section of EOP E-0 for re-diagnosis instructions?
| |
| ' Admin Package
| |
| | |
| )
| |
| CATEGORY: A.1 QUESTION # 2:
| |
| References Allowed: Yes Given the following:
| |
| Unit 1 is operating at 100% power.
| |
| e 0200 on January 2,1999 - CCP l-2 declared inoperable due motor problems.
| |
| . 0200 on January 5,1999 - CCP 1-2 is still inoperable w/an estimated RTS of 12 hrs.
| |
| e 0200 on January 5,1999 - A Unit 1 Shutdown is initiated.
| |
| e 0500 on January 5,1999 - CCP l-1 trips on OC due to pump bearing failure.
| |
| -e 0505 on January 5,1999 - PDP l-3 was placed in service and letdown restored.
| |
| When is the unit REQUIRED to be in MODE 3 (date and time)?
| |
| i i
| |
| i Admin Package
| |
| | |
| I
| |
| ~ CATEGORY: A.3 QUESTION # 1:-
| |
| i References Allowed: Yes Unit I has sustained a large break LOCA and fuel damage due to a 0.5g earthquake. The estimated exposure involved with obtaining an RCS sample is 4 REM.
| |
| Under these conditions, who can authorize the chemist, with a previous exposure history of 1.5 REM for the current year, to take the sample and receive the estimated exposure?
| |
| i Admin Package
| |
| | |
| CATEGORY: A.3 QUESTION # 2:
| |
| References Allowed: Yes You made two entries into the containment today and received the following exposure: .
| |
| {
| |
| l'' entry: 124 mrem gamma; O mrem neutron 2"d entry: 42 mrem gamma; 44 mrem neutron j The following is your exposure history this yearjust prior to today's entries: !
| |
| = 320 mrem DDE (Deep Dose Equivalent) l
| |
| * 70 mrem CEDE (Committed Effective Dose Equivalent) e- <18 mrern SDE ' (Shallow Dose Equivalent)
| |
| )
| |
| e 28 mrem CDE (Committed Dose Equivalent) l How much more radiation exposure can you receive without any signed extensions?
| |
| , l l
| |
| i l
| |
| l l
| |
| Admin Package
| |
| | |
| 1 CATEGORY: A.1 TOPIC: Procedure Usage j KA: G2.1.20 (4.2)
| |
| Reference Allowed: NO
| |
| | |
| ==Reference:==
| |
| LPE-R ULE, Obj.15, and OPl .DC l 1, p.10. '
| |
| QUESTION # 1:
| |
| During the implementation of an EOP the Shift Foreman informs the Shift Supervisor that he believes the procedure is inappropriate for the current plant conditions. The Shift Supervisor concurs with the Shift Foreman's assessment of the problem.
| |
| 1 What two (2) checks should be performed per OPl.DCI1 " Conduct of Operations - Abnormal !
| |
| Conditions" by the Shin Foreman and Shin Supervisor prior to the Shift Supervisor authorizing the return to the diagnosis section of EOP E-0 for re-diagnosis instructions?
| |
| ANSWER:
| |
| I
| |
| : 1. The entry conditions for the procedure in use shall be reviewed. [0.5]
| |
| : 2. Ifit is determined that the entry conditions are not met, the previous procedure in use shall be reviewed to determine if a correct procedure transition was made. [0.5]
| |
| l
| |
| /
| |
| NOTE: Subparts 1 and 2 must be answered correctly to receive a SATISFACTORY grade.
| |
| 1 Candidate's Response: SAT UNSAT (
| |
| l l
| |
| i Admin Package
| |
| | |
| O CATEGORY: A.1 TOPIC: Procedure Usage KA: G2.1.12 (4.0)
| |
| Reference Allowed: YES
| |
| | |
| ==Reference:==
| |
| STG M8, Obj. 31 pg. 3-12, Technical Specifications 3.0.3,3.1.2.4 & 3.5.2 I
| |
| QUESTION # 2:
| |
| Given the following:
| |
| i Unit 1 is operating at 100% power.
| |
| . 0200 on January 2,1999 - CCP l-2 declared inoperable due motor problems. !
| |
| e 0200 on January 5,1999 - CCP l-2 is still inoperable w/an estimated RTS of 12 hrs.
| |
| e 0200 on January 5,1999 - A Unit 1 Shutdown is initiated.
| |
| e 0500 on January 5,1999 - CCP l-1 trips on OC due to pump bearing failure. ,
| |
| e 0505 on January 5,1999 - PDP l-3 was placed in service and letdown restored.
| |
| When is the unit REQUIRED to b>.: in MODE 3 (date and time)?
| |
| ANSWER: )
| |
| l MODE 3 is required by 0800 on January 5,1999 NOTE:. Although 2 CCP.c inoperable places the unit in T.S. 3.0.3 and requires MODE 3 in 7 hours, T.S. 3.1.2.4 is more limiting since a required shutdown was already in progress with 3 !
| |
| hours remaining to MODE 3, therefore MODE 3 is required by 0800 on January 5,1999. I Candidate's Response: SAT UNSAT l
| |
| l 1
| |
| t-Admin Package l
| |
| | |
| 1 p.
| |
| - CATEGORY: A.3 TOPIC: Radiation Control (Limits and Control requirements)
| |
| K/A: G2.3.4 (3.1) {
| |
| I Reference Allowed: YES )
| |
| | |
| ==Reference:==
| |
| GRPA400e," Experienced Radworker," Obj.1, pg. 3 and RPl.ID6," Personnel Dose Limits & Monitoring Requirements," Att.10.1 QUESTION # 2:
| |
| 1 You made two entries into the containment today and received the following exposure: i l'' entry: 124 mrem gamma; 0 mrem neutron ,
| |
| 2"d entry: 42 mrem gamma; 44 mrem neutron l The following is your exposure history this yearjust prior to today's entries:
| |
| . 320 mrem DDE (Deep Dose Equivalent)
| |
| . 70 mrem CEDE (Committed Effective Dose Equivalent) e 48 mrem SDE (Shallow Dose Equivalent) e 28 mrem CDE (Committed Dose Equivalent)
| |
| IIow much more radiation exposure can you receive without any signed extensions?
| |
| ANSWER:
| |
| 1400 mrem (1.4 rem)
| |
| NOTE: Cunent exposure is all DDE and totals 210 mrem. TEDE is CDE + DDE therefore the total exposure including today's 210 + 320 + 70 = 600 mrem. The maximum possible exposure without a signed extension is the DCPP guideline of 2000 mrem / year. Therefore the remaining exposure allowed is 2000 - 600 = 1400 mrem.
| |
| Candidate's Response: SAT UNSAT l
| |
| Admin Package
| |
| | |
| -r - i
| |
| . i CATEGORY: A.3 TOPIC: Radiation Control (Limits and Control requirements)
| |
| K/A:}}
| |
