IR 05000275/1999301
| ML20196K531 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/25/1999 |
| From: | Hurley L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9904010020 | |
| Download: ML20196K531 (120) | |
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{{#Wiki_filter:r j# '% UNITED STATES f a
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9 NUCLEAR REGULATORY COMMISSION
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REGION IV
, 811 RYAN PLAZA DRIVE, SulTE 400 %, ARLINGTON, TExA5 760118064 March 25,1999 NOTE TO: NRC Document Control Desk Mail Stop O-5-D-24 FROM: Laura Hurley, Licensing Assistant Operations Branch, Region IV SUBJECT:
OPERATOR LICENSING EXAMINATIONS ADMINISTERED ON JANUARY 25-28,1999, 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: ltem #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 Code A070.
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.
,n 9904010020 990325 PDR ADOCK 05000275 V PDR 4
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l, ATTACHMENT 3 FINAL WRITTEN EXAMINATIONS AND ANSWER KEYS I
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U. S. Nuclear Regulatory Commission ) Site-Specific I Written Examination Applicant Information Name: Region: IV Date: Facility / Unit: Diablo Canyon 1 & 2 , License Level: SRO Reactor Type: Westinghouse Start Time: Finish Time: Instructions { Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of 80.00 percent. Examination papers will be
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collected four hours after the examination starts.
i Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicants Signature i Results Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent u_ J
, Appandix E , NRC Policies and Guidelines for Taking NRC Examinations ;
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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 queMions concerning 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 blRC 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 or 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 ine 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 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.
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Appendix E ., PART B -WRITTEN EXAMINATION GUIDELINES .7. If the ir- .t 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 possibility 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 I on the examination cover sheet indicating that the work is your own and that you have neither g'vsn nor received assistance in completing the l 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.
11. Do you have any Questions? l
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ANSWER SHEET FOR SENIOR REACTOR OPERATOR WRITTEN EXAMINATION l Name: 1. @@@@ 18. @@@@ as. @@@@ 2. @@@@ 19. O@@@ 3e. @@@@ l l 3. @@@@ 20. @@@@ 37. @@@@ 4. @@@@ 21. @@@@ 38. @@@@ s. @@@@ 22. @@@@ 39. bb@@ e. @@@@ 23. @@@@ . 40. @@@@ 7. @@@@ 24. @@@@ 41. @@@@ 8. @@@@ 25. @@@@ 42. @@@@ 9. b@@@ 26. bbbb 43. b@@b 10. @@@@ 27. @@@@ 44. @@@@ 11. @@@@ 28. @@@@ 45. @@@@ 12. @@@@ 29. bbbb 46. bbbb 13. @@@@ 30. @@@@ 47. @@@@ 14. 8@@@ 31. @@@@ 48. @@@@ 1s. 8@@@ 32. @@@@ 49. b@bb 18. @@@@ 33. O@@@ so. @@@@ 17. bb@b 34. bb@b GRADED BY DATE
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WRITTEN EXAMINATION Name: l 51. b@@@ 68. @@@@ 85. b@@@ I 52. b@b@ 69. bbbb 86. bbbb 53. b@@@ 70. bbbb 87. b b b b 54. b@@@ 71. bbbb 88. 'b b b b ! 55. bbbb 72. bbbb 89. bbb@ 56. b@@@ 73. b@@@ 90. bbbb 57. b@@@ 74. b b b b 91. bbbb ! 58. b@@b 75. b@b@ 92. bbbb 59. b b'@ b 76. bbbb 93. bbbb 60. b@b@ 77. b@@@ 94. bbbb 61. bbbb 78. bbbb 95. bbbb 62. b@@@ 79. bbbb 96. bbbb es. b@b@ 80. b@b@ 97. b b b b 64. bbbb 81. bbbb 98. bbbb 65. bbbb 82. bbbb 99. b b b b 66. bbbb 83. bbbb 100. b b b b 67. bbbb 84. bbbb ! ,
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EXAMINATION ' Senior Reactor Operator Written Exam Question: 001 As reactor thermal power is increased, the rod insertion limits (RILs) are required to be j progressively higher. From WHICH ONE (1) of the following parameters is the control j bank low alarm determined? a. auctioneered high Tavg b. auctioneered high Tref c. auctioneered high Ni power level d. auctioneered high delta T ) l l Page: 1
EXAMINATION ' Senior Reactor Operator Written Exam Question: 002 The following initial conditions exist:
. Tavg -Tref. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . OF . PT505.................................. 30 % . Non-Linear Gain Unit output (T/M test) 2F e Rod Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Manual Control Room / Electrical Team is testing the auto rod control system, simulating a signal that reactor power has just decreased while turbine power remained constant.
WHICH ONE (1) of the following is the speed rods would move out if AUTO is inadvertently selected? OlM drawing is provided, a. 8 steps / min b. 40 steps / min c. 48 steps / min ' d. 72 steps / min I
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EXAMINATION ' Senior Reactor Operator Written Exam Question: 003 Given the following:
. The unit is at 30% power.
. RCP 1-2 trips.
. NO operator action is taken.
WHICH ONE (1) of the following describes the INITIAL unit response to the RCP trip? 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. j c. A reactor trip WILL occur and S/G 1-2 water level will INCREASE. 4 d. A reactor trip WILL occur and S/G 1-2 water level will DECREASE.
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EXAMINATION ' Senior Reactor Operator Written Exam Question: 004 Given the following:
. RCS Pressure - 335 psig . RCS Temperature - 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?
j RCS Temperature RCS Pressure a. Increases increases i b. Increases Decreases c. Decreases increases d. Decreases Decreases . Page: 4
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EXAMINATION ' i Senior Reactor Operator Written Exam Question: 005 Given the following Unit 1 plant conditions; e Emergency boration is initiated to restore adequate SDM.
. The preferred Emergency Boration flo~ .e .ch alignment results in 0 gpm flow due to a clogged Boric acid filter.
To successfully initiate emergency boration WHICH ONE (1) of the following methods should be used? j i 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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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 . 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.
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b. Buses F & H energized from startup power & Bus G is de-energized.
c. Buses F, G & H energized from startup power.
d. Buses F & H energized from startup power & Bus G is energized by D/G 2-1. I I l i i l i Page: 6 j
i EXAMINATION ' Senior Reactor Operator Written Exam Question: 007 Unit 1 is in Mode 6, with refueling in progress. The Control Roorn 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 Nls 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 will be disabled.
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EXAMINATION ' Senior Reactor Operator Written Exam Question: 008 Unit 1 is at 90% power when a DRPl 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 steps, +/-10 steps c. operable; accuracy is +/-4 steps, +/-10 steps i d. operable; accuracy is +/-12 steps I i l t
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EXAMINATION Senior Reactor Operator Written Exam l 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 I 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 i Page: 9 .
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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 boxe's for the thermocouples rises 15 F over the duration of the shift. Reactor power level remains constant over the duration of the shift.
WHICH ONE of the following correctly describes how core exit thermocouple (Incore Temperature Monitor) readings are affected by the temperature change in the area of the of the reference junction boxes? 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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EXAMINATION Senior Reactor Operator Written Exam l l 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 the reason for the CRDM fans NOT running? 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 de-energized the MCCs supplying the CRDM fans.
c. The CRDM fan MCCs are de-energized 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.
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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 through 1-5 are a!! running in slow started by the Auto Transfer Timer.
c. CFCUs 1-1 and 1-2 are running in slow and started by the ESF Load 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.
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' ' EXAMINATION Senior Reactor Operator Written Exam Question: 013 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 STA'.TS; 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 valvas OPEN.
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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 pump; trips on overcurrent.
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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 tower.
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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EXAMINATION I Senior Reactor Operator Written Exam l Question: 016 Diesel Generator 2-3 is shutdown and in automatic when PK18-04 "Desel 23 Control Under Voltage" alarms. Investigation of the alarm reveals tisat the normal DC power supply breaker 72-2116 is tripped and will not reclose.
What is the affect of this failure on D/G 23? 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.
l 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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EXAMINATION Senior Reactor Operator Written Exam Question: 017 A Safety injection wsc 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 (RCDT) Pump discharge flowpath (FCV-253 and 254) to the Liquid Radwaste (LRW) system.
a. Reset Phase A and reset LRW lsolation.
b. Reset Phase A, Reset LRW lsolation, Open FCV-253 a7d 254 from the local control panel. , I c. Reset Phase A, Open FCV-253 and 254 from the local panel.
d. Reset LRW lsolation, open FCV-253 and 254 from the local panel.
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' ' EXAMINATION Senior Reactor Operator Written Exam Question: 018 The following Radiation Monitor indications exist:
. RM-14/87 - NR/RM-14 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? l l
a. Gas Decay Tank Rupture ! I b. S/G Tube Leak l c. Main Steam Line Break, downstream of the MSIVs d. Incore Seal Table Leak i l
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l EXAMINATION Senior Reactor Operator Written Exam Question: 019 l WHICH ONE (1) of the following will cause the Fuel Handling Building Ventilation ! System to automatically swap to the lodine Removal Mode? 4 a. RM 58 ALERT alarm actuates with RM 58 Bypass switch in NORMAL. i b. RM 59 HIGH alarm actuates, with RM 59 Bypass switch in BYPASS.
c. RM 58 HIGH alarm actuates, with RM 58 Bypass switch in NORMAL.
d. RM 59 ALERT alarm actuates with RM 59 Bypass switch in BYPASS. ]
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EXAMINATION Senior Reactor Operator Written Exarn 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 i b. PZR PORVs c. PZR Spray Valves d. PZR Code Safety Valves
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i 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 control operator has just reported j that Centrifugal Charging pump 1-1 has tripped on overcurrent. j WHICH ONE (1) of the following describes the allowances and/or limitations imposed by the Technical Specifications for placing Unit 1 in Hot Standby? ECCS T.S. 3.5.2 is provided.
a. Be in Hot Standby in 6 hours b. Be in Hot Standby in 7 hours c. Be in Hot Standby in 60 hours d. Be in Hot Standby in 66 hours i i
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EXAM! NATION 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 Leo RCS Temo a. 428 psig 250 F b. 456 psig 265 F c. 439 psig 283 F d. 462 psig 315 F l
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EXAMINATION Senior Peactor Operator Written Exam Question: 023
- The following plant conditions exist on Unit 2: . Reactor Power 50% and stable . Rod Controlin AUTO e All PZR controls in AUTO i . 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 i rate of 1 psig/ min.
How much time will elapse before the PZR Backup heaters turn on? j a. 10 min.
b. 15 min.
' c. 20 min.
d. 25 min.
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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 . All ECCS equipment is operating properly . 73ressuri7.er 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.
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EXAMINATION Senior Reactor Operator Written Exam Question: 025 The following plant conditions exist:
. Unit 1 is at 100% power- . Pressurizer level control IJ460A is selected for LT 459/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 increases, 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. !
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Senior Reactor Operator Written Exam Question: 026 - Given the following conditions on Unit 1:
. The plant is at 100% power . Solid-state protection system (SSPS) testing is in progress . Reactor trip breaker "B" (RTB) is OPEN.
. Bypass breaker "B" (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 CLOSE7 I a. RTA shunt trip coil and undervoltage trip coil will be energized.
b. RTA shunt trip coil and undervoltage trip coil will be de-energized.
c. RTA shunt trip coil will be de-energized and RTA undervoltage trip coil will be energized.
d. RTA shunt trip coil will be energized and RTA undervoltage trip coil will be de-energized.
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Senior Reactor Operator Written Exam Question: 027 Given the following:
. Unit 1 is at 9% power du.ing 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 / Feed flow mismatch d. Derivative of the aibitrated narrow range steam generator levels Page: 27
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Senior Reactor Operator Written Exam Question: 028 WHICH ONE (1) of the following reasons correctly identifies why the Containment Vacuum / Pressure 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 b. The valves will close on a Containment Vent isolation signal, but do not ; automatically close on a containment Phase "A" Isolation signal. l I c. The valves have not been demonstrated to isa capable of closing during a
. steam break inside containment. 1 d. The valves have not been demonstrated to be capable of closing during a LOCA inside containment.
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EXAMINATION - Ser.ior Reactor Operator Written Exam Question: 029 The following conditions exist:
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. A LOCA has occurred l I
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. Safety injection has actuated
. Phase B isolation has actuated
. 4 Kv 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 l l flow.
c. Selected SFP pump restarts, SFP temperature decreases due to increased CCW flow.
d. Neither SFP pump restarts, SFP temperature increases due to CCW flow ;
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Question: 030 ) l Given the following: I
. Unit 1 is in MODE 6 l '
. Fuel handling in progress j . A fuel assembly is dropped back into the core during removal from the core.
. Bubbling is observed from the dropped assembly l J WHICH ONE (1) of the following is NOT a required action per OP AP-21," Irradiated Fuel Damage," under these circumstances? l l ' a. Verify proper ventilation alignments.
b. Close the Fuel Transfer tube gate valve.
c. Determine the extent of fuel damage.
l d. Evacuate personnel from the manipulator crane.
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EXAMINATION - Senior Reactor Operator Written Exam Question: 031 Given the following:
* The plant has been at 100% power for the past 200 days.
- A Steam Line Break has just occurred.
WHICH ONE (1) of the following describes the effect on SHUTDOWN MARGIN (SDM)? a. SRA cacreases due to the negative isothermal Temperature Coefficient.
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.
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- EXAMINATION - Senior Reactor Operator Written Exam Question: 032 Unit 1 is operating at 100% power with the following conditions:
. Normal electric distribution lineup . 4 Kv breaker inspection reveals that startup feeder breaker for Bus G (52-HG-14) is inoperable.
What action (s) if any is(are) required by Technical Specifications? Electrical Power Systems T. S. 3.8.1.1 is provided a. Demonstrate the operability of the remaining A. C. source to 4 Kv bus G and the operability of Diesel Generator 1-2.
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. sources 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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I 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. A permanent magnet generator (PMG).
b. Diesel DC control power, supplied by 125 VDC bus 1-1. I
c. Diesel DC control power, supplied by 125 VDC bus 1-2.
l d. Diesel DC control power, supplied by 125 VDC bus 1-3.
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.. EXAMINATION - Senior Reactor Operator Written Exam Question: 034 WHICH ONE (1) of the following describes what is provided by the " live zero" function of a check source? It provides: a. a constant source for calibrating the discriminator circuit.
b. a minimum indication below which a circuit failure is indicated.
c. a minimum indication below which the monitor switches to the pulse mode.
d. a minimum indication when the " check source" function is used.
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l - EXAMINATION - Senior Reactor Operator Written Exam i 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 non-functional. Additionally the engineer reports that no other fire system impairments exist.
'Which ONE (1) of the following is the minimum required action? 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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EXAMINATION - Senior Reactor Operator Written Exam
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Question: 036 Given the following conditions on Unit 1:
. Unit is in Mode 1 . Containment temperatures indications are:
a TE-85 = 118 F a TE-87 = 119 F a TE-89 = 121*F a TE-91 = 123 F e 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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i l EXAMINATION - Senior Reactor Operator Written Exam
Question: 037 ) Unit 1 is operating at 100% A seismic event occurs causing the following:
. Reactor Tnp
. Safety injection
. Large Break LOCA
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. 4kv Bus F differentiallockout WHICH ONE (1) of the following actions must be performed in EOF E-0, " Reactor Trip or Safety injection", in regards to the CCW/ASW system?
a. Place standby Component Cooling Water heat exchanger in service. 1 b. Start the standby Component Cooling Water pump.
c. Cross-tie Auxiliary Saltwater with Unit 2.
d. Isolate the non-vital Component Cooling Water header.
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i 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 . Circ Water Pump 1-1 in service . 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)
a. Groups 1 & 2 would fully open.
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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EXAMINATION - Senior Reactor Operator Written Exam Question: 039 Given the following:
. Stator Cooling Water Conductivity is above its normal range at 4 pS/cm; the HIGH CONDUCTIVITY alarm is illuminated.
. A problem with a Stator Cooling Water flow switch (FS1) has caused it to indicate a low flow condition.
. The unit initiates a sunback 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 s
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EXAMINATION - Senior Reactor Operator Written Exam Question: 040 Given the following:
. 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.
. 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 temperature will INCREASE.
d. HCV-637, HCV-638, & HCV-670 will all fail CLOSED and RCS temperature will INCREASE.
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l l EXAMINATION - Senior Reactor Operator Written Exam Question: 041 Given the following;
. Unit 1 is at 50% pcwer 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 boration was used to return RCS temperature to program WHICH ONE (1) of the following conditions has resulted from the rod withdrawal event? a. The Rod Insertion Limit (RIL) has increased.
b. DNBR has increased.
c. Available shutdown margin has increased.
, d. Quadrant power tilt ratio (QPTR) has increased.
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EXAMINATION - Senior Reactor Operator Written Exam Question: 042 Given the following on Unit 1:
. Unit is ramping to 100% . Reactor power is at 909o 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-12C, " Dropped Control Rod".
WHICH ONE (1) of the following describes th9 required actions to establish initial i recovery conditions 20 minutes following the dropped rod event? I a. No action is required.
b. Reduce turbine load to reduce Reactor power to ~85%. c. Reduce turbine load to reduce Reactor pow n to less than 50%. . I d. Set Tavg 1.5*F above Tref by withdrawing control bank rods as necessary.
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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 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 6 steps . H12, rod position indication light indicating 6 steps What action is required?
a. No actions required b. Emergency Borate 900 gallons c. Emergency Borate 1800 gallons d. Emergency Borate 2700 ga!!ons l Page: 43 l
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EXAMINATION - Senior Reactor Operator Written Exam
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QuesGox 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 boiic acid avai!able 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.
d. To ensure sufficient boric acid and sodiurn hydroxide mixing to maintain the proper pH of water in the containment recirculation sump.
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EXAMINATION - Senior Reactor Operator Written Exam Question: 045 A Loss Off Coolant Accident (LOCA) outside containment has resulted in RCS Subcooling dropping to 0 F.
When attempting to determine if a LOCA outside containment has been isolated, in accordance with ECA-1.2 "LOCA Outside Containment," 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 Level increasing c. Pressurizer levelincreasing d. RCS pressure increasing l I l
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EXAMINATION , Senior Reactor Operator Written Exam Question: 046 Give o the following conditions:
. Unit 2 has experienced a Safety injection
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. EOP E-0, " Reactor Trip or Safety injection", is in progress
. RM-10, Aux Bldg Control Board Radiation Monitor is in alarm
. WR RCS Pressure is slowly decreasing WHICH ONE (1) of the following is the correct procedure transition for the above conditions: l
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a. Continue in E-0, " Reactor Trip or Safety injection and REFER to OP AP-17, j
" Charging Line Break."
b. GO TO E-1 " Loss of Reactor or Secondary Coolant." j 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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. 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 remein constant because pressure is above the shutoff head ' for the Si Pump.
b. Subcooling will decrease because RCS pressure will drop below the shutoff head for the St Pump.
c. Subcooling will decrease.because SI flow will be reduced to the output of the remaining St Pump.
d. Subcooling will remain constant because charging pump flow will increase to maintain RCS pressure.
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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 sealleakoff 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 charging pumps.
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EXAMINATION , Senior Reactor Operator Written Exam
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Question: 049 Crew is performing E-0.2, " Natural Circulation Cooldown", Step 15c which requires the crew to verify all CRDM fans are running.
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.
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
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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 makedp, results in 25 gpm flow with the Boric Acid pump in high speed.
. Emergency Boration is accomplished by using CVCS-8104 WHICH ONE (1) of the following describes the method used 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.
l 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) and the duration the valve is open.
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I 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 closing FCV-357, RCP Thermal Barrier CCW Return isolation valve? a. Prevents subsequent steam binding in the CCW system and prepares the plant for recovery.
b. Prevents inadvertent operation of the valve when power is restored to the bus.
c. Prevents damage to the RCP seal package when CCW flow is re-established.
d. Isolates a potential RCS leakage path in case of a thermal barrier rupture.
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Senior Reactor Operator Written Exam ' i Question: 052 l
Given the followi'g: _ . 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?
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a. immediately exit FR-S.1, and perform EF E-0, " Reactor Trip or SI", since the j reactor has tripped.
b. Remain in FR-S.1, while concurrently verifying the immediate actions of EP E-0, " Reactor Trip or SI".
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
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EXAMINATION
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Senior Reactor Operator Written Exam Question: 053 Given the following:
. Operators are performing EOP ECA-2.1, " Uncontrolled Depressurization of All Steam Generators."
. Cooldown rate is 200'F per hour.
. Steam Generator !avels range from 1% to 3% NR level.
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? a. To provide AFW pump minimum recire flow requirements.
b. To prevent S/G tube dryout while ensuring the minimum detectable feed flow is maintained.
c. To minimize the potential water hammer by maintair:ing a minimum flow through the feed ring J tubes.
d. To conserve CST inventory until the end of the blowdown phase.
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EXAMINATION - t Senior Reactor Operator Written Exam Question: 054 WHICH ONE (1) of the following operator actions will result in Pressurized Thermal Shock (PTS) conditions following a Steam 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 Accumulators to inject while cooling down and depressurizing c. Delaying the termination of Safety injection after the termination criteria have been met, f 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 . All four RCP's running . 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 b. 547 F i c. 551 F d. 554 F i I l
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EXAMINATION - Senior Reactor Operator Written Exam Question: 056 WHICH ONE (1) of the following is the reason for the order in which the valve positions are listed in step 3," CHECK RCS is isolated," of ECA-0.0," Loss of All AC Power?" a. Those most likely to fail in a loss of AC are listed first.
b. Those most likely to have an RNO corrective action outside the control room are listed last.
c. They are listed according to the way they are found on the control board, from left to right.
d. They are listed according to the capacity of outflow and potential for inventory loss.
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EXAMINATION - l Senior Reactor Operator Written Exam 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 MA'NUAL power supply to HC-459D, " Pressurizer Master Level , Controller," affect pressurizer le' control during the ramp? I a. No operator actions should be necessary.
b. Throttle open on HCV-142, "RCP Seal Flow control valve."
c. Throttle closed on FCV-128, "CCP flow control valve." l
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d. Adjust HC-459D, " Pressurizer Master Level Controller."
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EXAMINATION - Senior Reactor Operator Written Exam Question: 058
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i How would the liquid radwaste system respond if RE-18, " Liquid Radwaste Rad j Monitor", were to come into alarm during a discharge of a Floor Drain Receiver? j 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 Receiver selected for fill. ;
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c. RCV-18 opens and FCV-477 closes. The tank that is on discharge will swap to I recirculation.
d. RCV-18 closes. The running Floor Drain Receiver pump will receive a trip I signal.
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EXAMINATION - Senior Reactor Operator Written Exam Question: 059 Given the following: 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 b. Work Control Shift Foreman c. Shift Supervisor d. Shift Technical Advisor i Page: E9
EXAMINATION - Senior Reactor Operator Written Exam Question: 060 WHICH ONE(1) of the following sets ofindications are available on the Hot Shutdown Panel? a. Auxilian/ Feedwater Flow, Containment Pressure, Charging Header Flow l b. Emergency Boration Flow, Raw Water Reservoir Level, Pressurizer Pressure c. Pressurizer Level, Refueling Water Storage Tank Level, Volume Control Tank Level d. Letdown Flow, Charging Pump Discharge Pressure, Wide Range RCS ! Pressure ; i i i
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EXAMINATION - Senior Reactor Operator Written Exam i
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Question: 061 Following a Reactor trip and Control Room evacuation, operators at the Hot Shutdown Panel (HSDP) have been given the direction to maintain Steam Generator (S/G) levels at 62% on LI-501 through LI-504 while cooling down the RCS from 547 F to 350 F.
WHICH ONE (1) of the following describes ACTUAL S/G level response during the cooldown? OP AP-8B, " Control Room inaccessibility - Hot to Cold Shutdown," Figure 2 is provided.
I a. levelincreases from 62% to 75%. b. level remains constant at 62%. c. level remains constant at 84%. d. level decreases from 84% to 68%. l I
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Senior Reactor Operator Written Exam i Question: 062 Given the following conditions on Unit 1: i
. Sunteillance test on containment pressure transmitter PT-934 is in progress using STP l-12-P934.
. PC934A " Containment Hi Pressure" bistable is in the Tripped position
. 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 H1 pressure is 3/4 b. Containment Hi Pressure 1/2 and l Containment HI HI pressure is 1/3 c. Containment Hi Pressure 1/2 and i Containment Hi HI pressure is 2/3 d. Containment Hi Pressure 1/2 and Containment Hi HI pressure is 2/4 l l l
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i EXAMINATION - Senior Reactor Operator Written Exam i l Question: 063
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)1 . Reactor Trip and Safety injection Has occurred . Degraded Core Cooling conditions exist due to several malfunctions in the ECCS 1 system. j e 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 ECA-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' I d. Complete FR-C.2 and then GO TO EOP E-1.3, " Transfer to Cold Leg Recirculation"
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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 l WHICH ONE (1) of the following is the action required based on the chemistry reports? i Reactor Coolant System T.S. 3.4.8 is provided. !
a. Be in at least HOT STANDBY with Tavg less then 500 F by 1100.
b. Restore the Dose Equivalent 1-131 within the limits by 0500 January 7*, or be in HOT STANDBY by 1100 on January 7*. c. Be in at least HOT STANDBY with Tavg less then 500 F by 1300.
d. Restore the Dose Equivalent I-131 within the limits by 0700 January 7*, or be in HOT STANDBY by 1300 on January 7*. l
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. EXAMINATION - Senior Reactor Operator Written Exam Question: 065 A reactor trip has occurred from 100% power. Governor valve 3 and it's associated stop valve are stuck open. During performance of EP E-0," Reactor Trip or SI" immediate actions, a MANUAL turbine trip failed to close the governor and stop valves.
With no further operator action (s), WHICH ONE (1) of the following plant responses I will mitigate this accident? l a. Safety injection due to RCS low pressure.
{ b. Turbine Trip due to overspeed. i c. Unit Trip due to anti-motoring.
d. Safety injection due to S/G low pressure. i
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EXAMINATION - Senior Reactor Operator Written Exam
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Question: 066 I l Given the following:
) . Unit 1 was at 100% power.
. A malfunction caused pressurizer safety valve 801'C to open and stick open.
. A Reactor Trip and Safety injection occurred.
. The RCS rapidly depressurized to saturation conditions.
. Pressurizer level initially dropped and then began to increase rapidly. '
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WHICH ONE (1) of the following describes the relationship between pressurizer level , and RCS inventory under these conditions? l a. Level is NOT an accurate indication of inventory, because the pressurizer level channels are calibrated for normal operating temperature and pressure.
b. Level is NOT an accurate indication ofinventory, because RCS voiding may result in a rapidly increasing pressurizer level.
c. Level is an accurate indication of inventory, because hydraulic pressure would force any voids into the pressurizer steam space and out the safety valve.
d. Level is an accurate indication of inventory, because voiding would occur in the pressurizer first due to reaching saturation conditions before the RCS.
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Question: 067 The RCS is in MODE 6 when a largo leak occurs.
The ru eng RHR Pump is los! due to cavitation.
Which procedure will provide direction for toe 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 Page: 67 _ _ _ _ _ _ _ _ - - _ _ _ - _
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EXAMINATION - Senior Reactor Operator Written Exam Question: 068 During performance of EOP E-1.2, " Post LOCA Cooldown and Depressurization," all RCP's have been stopped.
A PORV is being used to depressurize the RCS to restore PZR level.
WHICH ONE (1) of the following describes the reason for restoring PZR level to greater than 17% [50%)? a. To ensura that a reduction in subcooling does not occur when an RCP is started.
b. To have sufficient inventory such that PZR level does not drop off scale when an RCP is started, c. To prevent letdown isolation when an RCP is started.
d. To ensure adequate PZR steam space to absorb pressure fluctuations when an RCP is started.
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l EXAMINATION Senior Reactor Operstor Written Exam 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 vf 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 EXAMINATION , Senior Reactor Operator Written Exam Question: 070 Given the following:
. RHR in service at reduced inventory conditions.
. OP AP SD-2," Loss of RCS Inventory," has been implemented due to pump cavitation on the running pump.
WHICH ONE (1) of the following could cause an increase in RCS level during performance of this procedure? a. Opening valve 8980, RHR suction from the RWST b. Venting the RHR system c. Closing valve 8702, RHR suction from the RCS d. Cycling a pressurizer PORV l i Page: 70
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EXAMINATION Senior Reactor Operator Written Exam 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:
. PK05-17 - PZR Pressure Low . Pressurizer Backup Heaters on . PK04-06 - Protect Channel Activated - PZR Press Lo 1/4 . PK05-16 - 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 maintan 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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EXAMINATION Senior Reactor Operator Written Exam Question: 072 Given the following: . Unit 2 is in Mode 2.'
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. ' A reactor startup is in progress and control banks are being withdrawn.
. Technical Specifications would require the suspension of rod withdrawalif Source range channel N-31 failed low and ' a. reactor power is less than 10 ' amps.
b.' reactor power is less than the source range high flux trip setpoint.
c. reactor power is less than the point of adding heat.
d. reactor power is less than 10-8 amps.
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, EXAMINATION Senior Reactor Operator Written Exam Question: 073 l Unit one was operating at 100% power when a Steam Generator tube leak occurred.
l 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 i d. 200 gpm l Page.73
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 with affected S/G steam flow DECREASING as feed flow DECREASES.
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.
d. Pressurizer pressure DECREASE with affected S/G steam flow LESS than feed flow.
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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 I operators are directed to restore a steam generator for a heat sink per FR-H.1 with the following plant conditions.
Indication Lo_op_1_ Loop 2 Looo3 Loop 4 l t S/G WR level (%) 0 12 7 3 S/G pressure (psig) 0 650 675 645 RCS hot leg temp ( F) 545 553 554 Sc. 3
. Containment pressure = 3.5 psig . Core exit T/Cs are stable at an average value = 560 F Wnich steam generator should be fed first?
a. S/G 1-1 b. S/G 1-2
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EXAMINATION Senior Reactor Operator Written Exam Question: 76 During a loss of heat sink from 100% power, the Steam Generator 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 of 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. j l c. Core uncovery occurs as a higher RCS repressurization decreases ECCS l flow.
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;. RCS depressurizes to a lower pressure with less of an inventory loss.
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r -. EXAMINATION
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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 s iould 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, c. Place Unit 1 in Building and safeguards with no changes in the Unit 2 ventilation alignment.
d. Place Both units in Buildings and Safeguards.
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EXAMINATION Senior Reactor Operator Written Exam Question: 078 Unit 1 Reactor is at 100% power. The Control operator reports that PK 02-11, "CVI Mode 6 Cutin"is in alarm. Upon further 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. , t
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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 at 2E+5 rad /hr. The 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 SE+4 1.5E+6 d. 3.5 SE+3 6.5E+4 I
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EXAMINATION Senior Reactor Operator Written Exam Question: 080 Unit 2 experienced a loss of Instrument A . 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 I
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EXAMINATION Senior Reactor Operator Written Exam l Question: 081 Unit 2 is operating with the following plant conditions: . 75% power j . 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 i 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 I i d. First Stage Impulse Pressure transmitter PT-505 fails high.
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EXAMINATION Senior Reactor Operator Written Exam Question: 082 WHICH ONE (1) of the following combinations of initial enrichment (%) and assembly burnup (MWD /MTU) shown below, would be allowed to be stored in Region 1 or Region 2 of the spent fuel pool with NO restrictions? Refueling Operations T.S.
3.9.14.1 and 3.9.14.3 provided.
(Fuel pellet diameter is 0.3225 inches) Initial Enrichment MWD /MTU a. 3.5 22,000 b. 4.0 24,000 c. 4.5 36,000 d. 4.8 37,000 Page: 82
l 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.
. 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 over stressed due to large fluctuations in charging and letdown.
Page: 83
_ 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" c. Remove DG 1-2 from service for Governor replacement
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d. Placement of CVCS Deborating Dernin 1-2 in service
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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 instr'Jment failure is fixed ? a. Withdraw controls rods as necessary, stopping at least every 3 steps to verify indications of temperature and power.
i b. Withdraw rods as necessary while continuously monitoring indications of I temperature and power. I c. Withdraw rods as necessary, stopping every full step to monitor indications of temperature and power.
d. Place Rod control in Auto and monitor indications of temperature and power.
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EXAMINATION Senior Reactor Operator Written Exam Question: 086 WHICH ONE (1) of the following would NOT require 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 Page: 86
m 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 e RCS pressure is 450 psig
. RCS Boron Concentration is 2050 ppm . RHR Cooling has just been secured i WHICH ONE (1) of the following describes a condition that requires containment integrity be established?
l a. Increasing RCS pressure to 550 psig.
a 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
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I EXAMINATION Senior Reactor Operator Written Exam I 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 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 d. Higher Natural Circulation Flowrate on Unit 1
Page: 88
EXAMINATION Senior Reactor Operator Written Exam Question: 089 Unit 1 DFWCS backup power supply needs replacement.
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
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,.- . 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 WHICH ONE (1) of the following situations is required to allow Maintenance to place Diesel Generator 1-3 on its backup DC power source? Outage Safety Scheduling, AD8.DC55, Attachment 7.2 is provided.
l a. After lY 13 is made available l b. After Diesel Generator 1-1 is made available c. After Mode 6 transition I d. After water level in reactor cavity greater than 23' l l l l l Page: 90 t-
EXAMINATION Senior Reactor Operator Written Exam ! Question: 091 What is the basis for requiring two (2) Reactor Coolant Pumps (RCPs) to be started i ' 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 withdrawal from subcritical accident.
b. Ensures RCP seal package will NOT experience thermal shock from pump starts at higher RCS temperatures.
! l c. Ensures boron concentration in the pressurizer and the RCS will be equalized ! with pressurizer spray flow.
, d. Ensures thermal shock to the pressurizer is minirnized by providing adequate j spray line flow l l r l Page: 91
EXAMINATION Senior Reactor Operator Written Exam Question: 092 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. Ensuring the evacuation of the refueling crew in the event of a valid high radiation alarm.
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EXAMINATION Senior Reactor Operator Written Exam 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 l l Page: 93
EXAMINATION Senior Reactor Operator Written Exam Question: 094 WHICH ONE (1) of the fc!!owing 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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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 Radioa.:tivity Area.
c. Airborne Radioactivity Area only.
d. Radiation Area and Airborne Radioactivity Area.
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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 1 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 . 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 Thermal Shock Conditions."
d. Go to FR-P.1, " Response to imminent Pressurized Thermal Shock Conditions."
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EXAMINATION l Senior Reactor Operator Written Exam Question: 097 1 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 depressurization minimize the RCS inventory loss? a. It will cause the KCP 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.
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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_" Loss 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 .
.b
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EXAMINATION Senior Reactor Operator Written Exam Question: 099 Given the following conditions on Unit 1: . Reactor Trip due to a Seismic induced LOCA q . All RCPs have been tripped ! . Containment Spray has been actuated . SFM has transitioned to EOP E-1 " Loss of Reactor or Secondary Coolant" The STA notes the following: . Intermediate Range Nls - 10E-03 Amps, -0.3 dpm . CETs - 580 F e 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- 96 ft . Pressurizer level- Off scale low WHICH ONE (1) of tha 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 Page: 99 s
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EXAMINATION Senior Reactor Operator Written Exam Question: 100 WHICH ONE (1) of the following responsibilities may be delegated by the Interim Site , Emergency 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. Assignment of plant personnel to positions in the Site Emergency Organization. I i l l l
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REFERENCE MATERIAL
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' NRC LICENSE WRITTEN EXAMINATION JANUARY 1999 ,
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l Auto Rod Control Block Diagram l Power Range NI's EEO4 z z z z 1P 1P 1P 1P Power Range l High Auctioneer Rate m Non-Linear u Variable ;
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Turbine y Comparator Gain Unit 4 Gain Unit Impulse Pressure (PT-505)
---* TREF - > Total 2 '
Auctioneered ' Error High T-Avg 1P Speed & Direction Programmer
---------______________________________________________________.
Non-Linear Gain Unit Variable Gain Unit Speed and Direction Programmer Gain 72
Speed Out
'5 8 r Error e i :
1 J 1! 3 5 _ , 1.5 50 % 100 % Turbine Power F Error l in i A-3-2 Rev 15 a
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SGe PSRC Interpretation 87-04, 90-07, 96-01 EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T g GREATER THAN OR E0 VAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two Emergency core Cooling System (ECCS) subsystems shall be OPERABLE with each subsyston comprised of: a. One OPERABLE centrifugal charging pump, b. One OPERABLE Safety Injection pump, c. One OPERABLE Residual Heat Removal heat exchanger,
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d. One OPERABLE Residual Heat Removal pump, and e. An OPERABLE flow path capable of taking suction from the Refueling Water Storage Tank on a Safety Injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILJJY: MODES 1, 2, and 3.
ACTION: a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be-in at least HOT.
STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
b. In the event the ECCS is actuated and injects water into the l Reactor Coolant System, a Special Report shall be prepared and
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submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of ' the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
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DIABLO CANYON - UNITS 1 & 2 3/4 5-3 32972702.4a TAB 12 3 uo_
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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: a. At least once each 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed: Valve Number Valve Function Valve Position 8703 RHR to RCS Hot Legs Closed 8802A Safety Injection Closed to.RCS Hot Legs 8802B Safety Injection Closed to RCS Hot Legs 8809A RHR to RCS Cold Legs Open 8809B RHR to RCS Cold Legs Open 8835 Safety Injection Open to RCS Cold Legs 8974A Safety Injection Open Pump Recir. to RWST 8974B Safety Injection Open Pump Recir. to RWST 8976 RWST to Safety Open Injection Pumps , 8980 RWST to RHR Pumps Open 8982A Containment Sump to Closed RHR 8982B Containment Sump to Closed RHR 8992 Spray Additive Tank Open to Eductor
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8701 RHR Suction Closed 8702 RHR Suction Closad l b. .At least once per 31 days by: 1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
DIABLO CANYON - UNITS 1 & 2 3/4 5-4 32972702.4a TAB 12 4 C_
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS (Continuedl i l c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed: 1) For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2) At least once daily of the areas affected within containment by containment entry and during the final entry when CONTAINMENT INTEGRITY is established.
d. At least once each REFUELING INTERVAL by a visual inspection of the containment :: amp and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion; 4 e. At least once each REFUELING INTERVAL by: 1) Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection actuation test signa!. 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal: a) Centrifugal charging pump, b) Safety Injection pump, and c) Residual Heat Removal pump, f. By verifying that each of the following pumps develops the
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indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5: 1) Centrifugal charging pump 2 2400 psid,
2) Safety Injection pump 2 1455 psid, and l 3) Residual Heat Removal pump 2 165 psid.
DIABLO CANYON - UNITS 1 & 2 3/4 5-5 Unit 1 - Amendment No. lQR1 118 32972702.4a Unit 2 - Amendment No. 1R\\ 116'
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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) g. By verifying the correct position of each electrical and/or mechanical position'stop for the following ECCS throttle valves: 1). Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2) At 1. east once each REFUELING INTERVAL.
Charging Injection Safety Injection Throttle Valves Throttle Valves _ 8810A 8822A 88108 88228 8810C 8822C 8810D 88220 l h. By perfonning a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the i subsystem flow characteristics and verifying that: 1) For centrifugal charging pumps, with a single pump running: j I a) The sum of injection line flow rates, excluding the i highest flow rate, is greater than or equal to 299 j gpm, and j
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DIABLO CANYON - UNITS 1 & 2 3/4 5-6 Unit 1-AmendmentNo.551\XRi\lQll118 32972702.4a TAB 12 6
. ' EMERGENCY CORE COOLING SYSTEMS Reissued September 19, 1991 SURVEILLANCE RE0VIREMENTS (Continued)
b) The total flow rate through all four injection lines I is less than or equal to 461 gpm, and c) The difference between the maximum and minimum
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injection line flow rates is less than or equal to 15.5 gpm, and d) The total pump flow rate is less thar, or equal to 560
.9Pm.
2) For safety injection pumps, with a single pump running: a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 427 gpm, and I b) The total flow through all four injection lines is less than or equal to 650 gpm, and c) The difference between the maximum and minimum ! injection line flow rates is less than or equal to ! 20.0 gpm, and ! The total pump flow rate is less than or equal to 675
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d) gpm.
i. By performing a flow test, during shutdown, following compl'etion ; ci modifications to the RHR system that alter the system flow
characteristics, and verifying that with a single pump running, L and delivering to all four cold legs, a total flow rate greater than or equal to 3976 gpm.
l l DIABLO CANYON - UNITS 1 & 2 3/4 5-6a Amendment Nos. 65 and 64 32972702.4a TAB 12 6 September 5, 1991
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10/6/94 ft PACIFIC GAS AND ELECTRIC COMPANY Pcga 1 of f pf NUCLEAR POWER GENERATION IDAP II3.ID3 ATTACHMENT 6.1 TITLE- REQUEST FOR TECHNICAL SPECIFICATION INTERPRETATION I REQUEST PSRC INTERPRETATION NUMBER: 8'7-04 REVISION: 1 TECHNICAL SPECIFICATION: 3.5.2. 3.6.2.1. 3.6.2.2. 3.7.1.2 QUESTION / CONCERN: To ensure the coarability of safetv-related systems.
valves in the flow each must be either correctiv nositioned. or canable of actuatina to the correct oosition when reauired. A small nu=h=r of these valves. if incorrectiv nositioned. can make both trains in a safety-related system inocerable. Guidance was reouested to identify all individual valves that. if mismositioned. would rendar a cafetv-related system inocerable and violate a Terhnient Soecification t4=4 tina Condition for Ooeration.
SUBMITTED B # -
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DATE: %/ 7 CIJ REVIEWED BY SE I DI e ._ . DATE: REVIEWED SY RE ATORY COMPLIANCE: [- f 4 DATE: 7,h , de II PSRC IfCERPRETATION Placino any one of the followina valves in the incorrect oosition. durina an couratino mode in which the associated system is reauired to be coerable. would result in an assumotion in the accident an=1vais not beina met and esuse both trains of the safetv-related system to be inonerable.
'"h i s action is a violation of the Technical Soecification L4m4eina Condition for Ooeration and reouires entry into Terhnical Snecification 3.0.3.* Entry into Technical Soecification 3.0.3 is rer,crtable nneler J 0CFR50. 73 (a) (2) (i) (B)
*Except for T.S. 3.7.1.2 PSRC APPROVED YES: NO: DATE: b APPROVED- _
h PLANT MANAGER DATE: 2!/2fL d EFFECTIVE DATE: b h DISTRIBUTION REQUIRED YES: NO: i- , s. !
e - g, Valve Number Valve Function Recuired Walve Modes Technical Position SoecificaricD 8105 CCP 1 and 2 Rocire Open 1,2,3 3.5.2 Line Isolation-8106 CCP 1 and 2 Recire Open 1,2,3 3.5.2 Line Isolation i 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 8802B' SI to RCS Hot Legs Closed' 1,2,3 3.5.2 8809A* RRR to RCS Cold Open 1,2,3 3.5.2 Legs 8809B* RHR to RCS Cold Open 1,2,3 J.5.2 Legs 8835* SI to RCS Cold Open 1,2,3 3.5.2 l Legs 8974A* SI Pump Recire to Open 1,2,3 3.5.2 RWST 8974B* SI Pump Recire to Open 1,2,3 3.5.2 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 i 8982A* Centainment Sump Closed 1,2,3 3.5.2 ! to RHR 8982B* Containment Sump Closed 1,2,3 3.5.2 to RHR j 8992* Spray Additive Open 1,2,3,4 3.5.2, ! Tank to Eductor 3.6.2.2 8716A RHR CroFsover Line Open 1,2,3 3.5.2 8716B RHR Crossover Line Open 1,2,3 3.5.2 9003A RER to Containment Closed 1,2,3 3.5.2 Spray 9003B RER 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 RNR 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 EC2S valves are identified in Surveillance Requirement 4.5.2.
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WJ;X@Mhr~E/88 (25) PACIFfC GAS AND ELECTRIC COMPANY NUCLEAR POWER GENERATION l DIABLO CANYON POWER PLANT UNITS 1 AND 2 ' TITLE: REQUEST FOR PSRC LICENSING 00CUMENT INTERPRETATION I REQUEST PSRC INTERPRETATION NUMBER: 90-07 REVISION: O LICENSE DOCUMENT: Technical Soecification 3.5.2e and 3.5.3d QUESTION / CONCERN: When the ECCS containment recirculation sumo uocer
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cratine assembiv access hatch is ooen durino Modes 1 throuch 4. what ; i actinn thould be taken? j
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i . W T.9C. b 4
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SUBMITTED BY: m DATE: S/9/96 REVIEWED BY DEPARTMENT NEAD: DATE: 7 /) REVIEWEDBYREGULATORYCOMPLINCE:/7 d fids DATE: 6/9/ O d
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II PSRC INTERPRETATION
, Technical Soecification 3.0.3 is to be entered whenever the ECCS containment j
i recirculation sumo access hatch is opened durino Modes 1 throu9h 4. This is l ha t a ri nn the NRC resconse (Chron 1485981 to PG&E's Reply to Notice of Viniatinn NRC enforcement Action 89-241 (PG1E Letter No. OCL-90-070).
f%a a ttsr her4\ l i i PSRC APPROVED YES:. / NO: DATE: .#/d3/rv i
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APPROVED: DATE: PLANT MANAGER
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1 10/6/94 Paga 1 of 1 PACIFIC GAS AND ELECTRIC COMPANY ] J NUCLEAR POWER GENERATION l IDAP XI3.ID3 i ATTACHMENT 6.1 i l TITLE: REQUEST FOR TECHNICAL SPECIFICATION INTERPRETATION I REQUEST PSRC INTERPRETATION NUMBER: 96-01 REVISION: 0 TECHNICAL SPECIFICATION: 4.1.2.3.1. 4.1.2.4.1. 4.5.2f.1) J QUESTION / CONCERN: The above Technical Soecifications (TS) recuire j verification that the CCP's develoo creater than or ecual to 2400 esid on recirculation flow when tested oursuant to TS 4.0.5. Does this testina j have to be done usino the CCP minimum recirculation flowcath, or may j testine be done usino an alternate flowcath if the recuired differential cressure is met at a flow rate ecual to or creater than the recirculation flow rate? SUBMITTED BY: M o. DATE: 222!9d REVIEWED OY SECTION DIREC'"OR: 5 MNNshnec DATE: Z- 0
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REVIEWED BY REGULATORY COMPLI : - u
% 3TE : - ~2 %
II PSRC INTERPRETATION The intent.of this TS is to' ensure that the CCPs are tested in accordance with TS 4.0.5 and to ensure continued conformance to the assumotions of the accident analyses. Accordinalv. the above TS are met by verifyino the differential cressure is creater than er eg al to 2400 osid by testino at a flow rate creater than or ecual to the flow rate develooed when coeratino a CCP solelv en its minimum flow recirculation flowcath.
P5RC APPROVE" NO: * ] ' q S.
ES: t DATE: APPROVED: ' .W DATE: e N3 M PLANT 1M"AGER EFFEC"'IVE DATE : ' / 7 5 f d'
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DISTRIBUTION REQUIRED YES: L NO: 01275401.01 1
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See PSRC Interpretation 85-03, 86-02 89-07, 95-03 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING , LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE: a.- between the offsite I
~Two independent transmission circuits network and the(one with Onsite delayed Class access)ibution 1E Distr System, and b.- Three separate and independent diesel generators, each with:
1. A se arate engine-mounted fuel tank containing a minimum volume of 250 allons of fuel, and 2. Two supply trains of the Diesel Fuel Oil Storage and Transfer System containing a minimum combined storage of 33,000 gallons of fuel for one unit o operation *geration* and 65,000 gallons of fuel for two unit APPLICABILITY: MODES 1.-2, 3, and 4.
ACTION: i a. With one offsite circuit of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. ! sources by performing Specification 4.8.1.1.la. within I hour and at least ' once per B hours thereafter. If each of the diesel generators have not been successfully tested within the past 24 hours demonstrate its OPERABILITY by perfonning Specification separately forcircuit each to such diesel generator within 24 hours. Restore4.8.1.1.2a.2)ffsite the o
OPERABLE status within 72*** hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
The performance of Technical Specification Surveillance Requirement 4.8.1.1.3.e requires one fuel oil storage tank to be removed from service to be drained and cleaned. During this surveillance the diesel generator fuel oil storage requirement for one unit operation,in Modes 1 through 4 and one unit operation in Mode 6 with at least 23 feet of water above the reactor vessel flange or 000 gallons. The tank being cleaned may with the reactor be inoperable for upvessel to 10 days. defueled For t is 35,he duration of tank cleaning,' te onsite fuel oil storage of 24 000 gallons will be maintained. Prior to removal of a tank from service, the offsite circuits required by Technical
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Specification 3.8.1.1.a will be verified to be OPER4BLE.
The. performance of modifications to the diesel fuel ?il storage and transfer system requires one fuel oil storage tank at a time to be drained and replaced with a new storage tank. During this period, the diesel generator fuel oil storage requirement for two unit operation in Modes 1 - 4, or for one unit operation in Modes 1 - 4 and one unit in Mode 5 or 6 is 35,000 gallons. A total of up to 120 days may be required to complete the replacement of both tanks. For the duration of the tank replacement, temporary onsite storage of 30,000 gallons will be maintained. Prior to removal of a tank from service, the offsite circuits required by Technical Specification 3.8.1.la. will be verified to be OPERABLE.
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For Unit 1 Cycle 8, the allowed outage time may be extended to 120 hours on a one-time basis for installation of auxiliary transfonner 1-1.
DIABLO CANYON -UNITS 1 & 2- 3/4 8-1 Unit 1 - Amendment No. 111 Unit 2 - Amendment No. 108 32973007.4a 1 TAB 15 1- March 8, 1996
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See PSRC Interpretations 95-01, 97-08
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) ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) b. With a diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the A.C. offsite sources by , performing Specification 4.8.1.1.la within I hour and at least once per 8 ! hours thereafter; and if the diesel generator became inoperable due to any I cause other than preventive maintenance or testing, demonstrate the OPERABILITY of the remainin Specification 4.8.1.1.2a.2)gwithin OPERABLE 24 hoursdiesel generators
*; restore by performing the diesel generator to OPERABLE status within 7 days or be in at least HOT STANDBY within' the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
c. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specification 4.8.1.1.la. within 1 hour and at least once per 8 hours thereafter; and if the diesel generator became inoperable due to any -cause other than preventive maintenance 'or testing, demonstrate the OPERABILITY of the remainin generators by performing Specification 4.8.1.1.2a.2)g OPERABLE within 8 hours; diesel restore at least one of the inoperable sources to OPERABLE status within 12 l hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore the other A.C. power source (offsite circuit or diesel generator) to OPERABLE status in
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i accordance with ACTION a. or b., as appropriate with the time requirement ; of that ACTION' statement based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel OPERABILITY per Specification 4.8.1.1.2a.2) perfomed under this ACTION statement for OPERABLE diesels or a restored to OPERABLE diesel satisfies the diesel ! generator test requirement of ACTION a. or b. ' d. With one diesel generator inoperable in addition to ACTION b. or c. above verify that: 1. All required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generators as a source of emergency power are also OPERABLE, and 2. When in MODE 1, 2, or 3 that at least two ~ auxiliary feedwater pumps are OPERABLE.
If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- This test is required to be completed regardless of when the inoperable diesel generator is restored to operability. l DIABLO CANYON - UNITS 1 & 2- 3/4 8-2 Unit 1 - Amendment No. 109 Unit 2 - Amendment No. 108 32973007.4a I . TAB 15 2 January 3, 1996
ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) e. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of three diesel generators by performing the requirements of Specification 4.8.1.1.2a.2) within 8 hours, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. Following restoration of one offsite source, follow ACTION a. with the time requirement of that ACTION statement based on the time of initial.. loss of the remaining inoperable offsite A.C. circuit. A successful test (s) of diesel generator OPERABILITY per Specification 4.8.1.1.2a.2) performed under this ACTION statement for the OPERABLE diesel generators satisfies the diesel generator test requirement of ACTION a.
f. With two or more of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing the requirements of Specification 4.8.1.1.la. within 1 hour and at least once per 8 hours thereafter; ensure at least two of the diesel generators are OPERABLE within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. With two diesel { generators OPERABLE follow ACTION b. with the time requirement of that j ACTION statement based on the time of initial loss of the remaining inoperable diesel generator. A successful test of diesel generator OPERABILITY per Specification 4.8.1.1.2a.2) performed under this ACTION statement for a restored to OPERABLE diesel generator satisfies the diesel generator test requirement of ACTION b.
g. With one supply train of the Diesel Fuel Oil Storage and Transfer System inoperable, restore the inoperable system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and be in HOT SHUTDOWN within the following 6 hours.
h. With both supply trains of the Diesel Fuel Oil Storage and Transfer System inoperable, restore at least one supply train, including the common storage system, to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and be in COLD SHUTDOWN within the following 30 hours. l l l l I
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DIABLO CANYON - UNITS 1 & 2 3/4 8-2a Amendment Nos. 15 and 14 l 32973007.4a I TAB 15 3 July 24, 1987 ; I
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ELECTRICAL POWER SYSTEMS' SURVEILLANCE REQUIREMENTS _
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4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class IE Distribution System shall be: a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and b. Demonstrated OPERABLE at least once per REFUELING INTERVAL during shutdown by: 1) Transferring 4 kV vital bus power supply from the nonnal circuit to the alternate circuit (manually and automatically) and to the delayed access circuit (manually), and 2) Verifying that on a Safety Injection test signal, without loss of offsite power, the preferred, imediate access offsite power source energizes the emergency busses with pennanently connected loads and energizes the auto-connected emergency (accident) loads through sequencing timers.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:* 1) Verifying the fuel level in the engine-mounted fuel tank, 2) Verifying the diesel starts from standby condition and accelerates to at least 900 rpm in less than or equal to 10 seconds. The generator voltage and frequency shall be 4160 + 240/-375 volts and 60 1.2 Hz within 13 seconds after the start signal. The diesel generator shall be started for this test by using one of the following signals: a) Manual, or b) Simulated loss of offsite power by itself (Startup bus undervoltage), or c) A Safety Injection actuation test signal by itself.
- All diesel generator starts for the purpose of this surveillance test may be preceded by an engine prelube period. Further, all surveillance tests, with the exception of once per 184 days, may also be preceded b gradual acceleration and/or gradual loading > 150 sec)y aswarmup procedures recommended by the (e.g.,
manufacturer so that the mechanical stress and wear on the diesel engine is minimized.
DIABLO CANYON'- UNITS 1 & 2 3/4 8-3 Unit 1 - Amendment 44\S(\S$11X51127 32973007.4a 1 TAB 15 4 Unit 2 - Amendment 45\$515R11R41125
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3) Verifying the generator is synchronized, loaded to greater than or equal to 2484 kW in less than or equal to 60 seconds, and operates for greater than or equal to 60 minutes, 4) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses, and 5) Verifying the diesel engine protective relay trip cutout switch is returned to the cutout position following each diesel generator test.
b. At least once per REFUELING INTERVAL during shutdown, by: 1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for
~ this class of standby service; 2) Verifying that the load sequence timers are OPERABLE with each load sequence timer within the limits specified in Table 4.8-2; 3) Verifying the generator capability to reject a load of greater than or equal to 508 kW while maintaining voltage at 4160 + 240/-375 volts and frequency at 60 ! 3 Hz; 4) Verifying the generator capability to reject a load of greater than or equal to 2484 kW without tripping. The generator voltage shall not exceed 4580 volts during and following the load rejection; 5) Simulating a loss of offsite power by itself, and:
a) Verifying de-energization of the emergency busses and load shedding from the emergency busses, and i b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds, energizes the required auto-connected loads through sequencing timers and operates for greater than or equal to 5 minutes while its generator is loaded with the pennanent and auto-connected loads. After energization of these loads, the steady state voltage ~and frequency of the emergency busses shall be maintained at 4160 + 240/-375 volts and 60 1.2 Hz during this test.
" DIABLO CANYON - UNITS 1 & 2 3/4 8-4 Unit 1 - Amendment Mi\S(1\Sai 126 32973007.4a I TAB 15 5 Unit 2 - Amendment 441\(51\ sri 124
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continusd) 6) Verifying that on a Safety Injection test signal with'out loss of offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 + 240/-375 volts and 60 1 1.2 Hz within 13 seconds after the auto-start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test; 7) Simulating a loss of offsite power in conjunction with a Safety Injection test signal, and: a) Verifying de-energization of the emergency busses and load shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with pemanently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through sequencing timers and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization of these loads, the steady state voltage and frequency of the emergency busses shall be maintained at 4160 + 240/-375 volts and 60 1 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engi" overspeed, low lube oil pressure and generator differential, ar bypassed when the diesel engine trip cutout switch is in the cutout position and the diesel is aligned for automatic operation.
8) Verifying the diesel generator operates for at least 24 hours. During the first 2 hours of this test, the diesel generator shall be loaded to greater than or equal to 2750 kW and during the remaining 22 hours of this test, the diesel generator shall be loaded to greater than or equal to 2484 kW. The generator voltage and frequency shall be 4160 + 240/-375 volts and 60 1.2 Hz within 13 seconds after the start signal. For Units 1 and 2 Cycle 7: Within 5 minutes after completing this 24 hour test, perfom Specification 4.8.1.1.2b.5)b);* 9) Verifying that the auto-connected loads to each diesel generator do not exceed the maximum rating of 2750 kW; 10) Verifying the diesel generator's capability to:
For Units 1 and 2 Cycle 7: If Specification 4.8.1.1.2b.5)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead the diesel generator may be operated at 2484 kW for 1 hour or until operating temperature has stabilized.
k l l DIABLO CANYON - UNITS 1 & 2 3/4 8-5 Unit 1 - Amendment No. 105 a Unit 2 - Amendment No. 104 I 32973007.4a I TAB 15 6 ae 26, 1995
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ELECTRICAL POWER SYSTEMS
I SURVEILLANCE REQUIREMENTS (Continued) a) Synchronize its isolated bus with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.
11). Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injection signal opens the auxiliary transformer breaker and automatically . sequences the emergency loads onto the diesel generator; and 12) Verifying that the shutdown relay lockout feature prevents diesel generator starting only when required: a) Generator differential current-high, or b) Engine lube oil pressure-low, or c) Emergency stop button actuated, or d) Overspeed trip actuated.
c. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all diesel cenerators simultaneously, during shutdown, and verifying that all liesel generators accelerate to at least 900 rpm in less than or equal to 10 seconds.
d. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day tank e. For Units 1 and 2, Cycle 8 and after: At least once per REFUELING INTERVAL by verifying the diesel generator starts and accelerates to at least 900 rpm in less than or equal to 10 seconds. The generator voltage and frequency shall be 4160 +240/-375 volts and 60 t 1.2 Hz within 13 seconds after the start signal. This test shall be perfonned within 5 minutes of shutting down the diesel generator after
- the diesel generator has operated for at least 2 hours at a load of greater than or equal to 2484 kW.
- 4.8.1.1.3 The Diesel Fuel Oil Storage and Transfer System shall be demonstrated i OPERABLE: ' a. At least once per 31 days by: 1 Verifying the fuel level in the fuel storage tank, and 2 Verifying that each fuel transfer pump starts and transfers fuel from 4 the storage system to each engine-mounted tank via installed lines.
b. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks; c. By sampling new fuel oil in accordance with ASTM-D4057 prior to addition to the storage tanks and: DIABLO CANYON - UNITS 1 & 2- 3/4 8-6 Unit 1 - Amendment No. 151\lR51 126 Unit 2 - Amen w nt No. 191\lq(1 124 32973007.4a I TAB 15 7
ELECTRICAL POWER SYSTEMS I SURVEILLANCE REQUIREMENTS (Continued) . 1) By verifying in accordance with the tests specified in ASTM-D975-81 l prior to addition to the storage tanks that the sample has:
.a) An API Gravity of within 0.3 degrees at 60*F, or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not detemined by comparison with the supplier's !
certification; I c) A flash point equal to or greater than 125'F; and d) A clear and bright appeLrance with proper color when tested in accordance with ASTM-D4176-82 or a water and sediment content of less than or equal to 0.05 volume percent when tested in accordance with ASTM D1796-83.
2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-D975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82.
d. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method l A; , e. At least once per 10 years by: 1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and 2) Performing a visual examination of accessible piping during an operating pressure leak test.
4.8.1.1.4 Reoorts - All diesel generator failures, valid or non-valid, shall be reported as a Special Report within 30 days to the Commission pursuant to Specification 6.9.2. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures (on a per diesel generator basis) in the last 100 valid tests is greater than or equal to 7, the report shall be l supplemented to include the additional information recommended in Regulatory Position l C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
! ! DIABLO CANYON - UNITS 1 & 2 3/4 8-7 Amendment Nos. 95 and 94 32973007.4a I TAB 15 8 September 23, 1994 _ _ _ _ _ _
. TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE Number of Failures in Number of failures in Last 20 Valid Tests * Last 100 Valid Tests * Test Freauency 51 55 At least once per 31 days 2 2** 26 At least once per 7 days i i
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* Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the last 20 and 100 tests and failures are determined on a per diesel generator basis. For the purpose of this schedule, only valid tests conducted after the completion of the preoperational test requirements of Regulatory Guide 1.108, Revision 1 August 1977, shall be included in the computation of the "Last 20 Valid Tests" and the "Last 100 Valid Tests." For the purpose of determining , .the required test frequency, the previous test failure count may be reduced to zero l if the specific cause for the diesel unreliability has been identified and resolved; appropriate post maintenance operation and testing have been completed; and if i acceptable reliability has been demonstrated. The reliability criterion shall be the 3 successful completion of 14 consecutive tests in a single series. These tests shall be in accordance with Specification 4.8.1.1.2a.2).
- The associated test frequency shall be maintained until seven consecutive failure '
free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.
l DIABLO CANYON - UNITS 1 & 2 3/4 8-8 Amendment Nos. 15 and 14 ; 32973007.4a I TAB 15 9 July 24, 1987 j
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69 86s9C, 12/66 (89) Page 2 c? i PACIFjC GAS AND ELECTR8C COMPANY l DEPARTMENT OF NUCLEAR PLANT OPERAT80NS O!ASLO CANYON POWEQ PLANT UN8T NOS. 1 AND 2 T' LE: RE0' JEST FOR ASRC LICENSING DOCUMENT INTERPRETATION PSRC INTERPRETATION tlUMBER 85-03 REVISION O LICENSE DOCUMENT: Technical Specification 3.8.1.1 and 3.8.1.2 (t,g:s g g ; QUESTION / CONCERN: When an IST or Auto Transf er Timer listed in Table 4.3.' l 1s found to be out of tolerance, whatsif any Technical Soecificatie . A m: statement (s) should be followed? )
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buBf'ITTED SY: f.. T. Womack DATE: 3/22/85 ATTACHl;EliTS (Y / h PSRC INTERPRETATION k'ith an IST Timer found to b'e outside the range of acceptable settints specified in Table 4.8-2. the corresponding diesel senerator shall be declared inoperable for MODES 1-4 and the correspondinz ACTIO! stateme-t(s) f ellowed . With an AUTO Transfer Timer found to be outside the tante of acceptable settings specified in Table 4.8-2, the correspondine diesel generator shall be declared inoperable for all MODES.
. PSRC APPROVED: YES X No DATE: 5-22-85 APPROVED: fh DATE: /i / f
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u,,MAusu / / '.
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x wafs-me urr-s guy _ e -o PACIFZC GAS AND ELdCTR8C COMPANY DEPARTMENT OF NUCLEAR PLANT OPERATZONS DIABLO CANYON POWER PLANT UNIT NOS. 1 AND 2-TITLE: RE00EST FOR PSRC LICENSING DOCUMENT INTERPRETATION PSRC INTERPRETATION NUMBER: 86-02 -REVISION: 0 LICENSE DOCUMENT: Technical Specification 3.8.1.1 QUESTION / CONCERN: Technical Specification 3.8.1.1 requires that on the loss of the offsite circuit (s) and/or diesel generator (s), the remaining diesel _ generator (s) must be verified operable per Specification 4.8.1.1.2a.2). This specification requires the diesel generator to be started and timed to its rated speed, voltage and. frequency. If the remaining diesel generator (s) have started and loaded.on a bus, must the diesel generator (s) be shutdown and restarted to verify operable per Specification 4.8.1.1.2a.2)?
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SUBMITTED BY:- M 2220^- ~ DATE: REVIEWED BY REGULATORY COMPLIANCE: %, <[ DATE: M17/N
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REQUEST NUMBER:
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PSRC INTERPRETATION If a diesel generator starts and loads on a bus, it is not necessary to shut-down the diesel generator and perfonn Specification 4.8.1.1.2a.2). The diesel generator is verified operable since it is performing its intended function.
However, if the bus is restored with a degradation of offsite sources, and the diesel is shutdown, it will be necessary to perfonn the accelerated periodic surveillance required by the Action Statement. ,
PSRC APPROVED YES: M NO: DATE: 8/22 /8 6
.. APPROVED: gud DATE: Z7[% J
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7 ' PLANT MANAGER
'DC0183 4VI'
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OfABLO CANYON POSER PLANT UMT N05.1 AND 2
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TfTLE: REOUEST FOR PSRC LICENSfNG DOCUMENT INTERPRETATf0N
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PSRC INTERPRETATION NUMBER: 89-07 REVISION: n LICENSE DOCUMENT: Technical Soecification 3.8.1.1 QUESTION / CONCERN: What constitutes comoliance with TS 3.8.1.1. Action a.
if an ina.n.na.nt ci rmi t en nm the nfreit, trancmiccinn netwnek it innnarable with resoect to en1v one akv vital bus (e.a. the bus G startuo feeder breaker is racked out)?
, -
SUBMITTED BY: S. R. Fridlev - DATE: T 4 /rv REVIEWED BY REGULATORY COMPLIANCE: O DATE: / REQUEST NUMBER: PSRC INTERPRETATION I! one t- .uit from the offsite transmission network to only one of the vital 4kv busses is inoperable, it is necessary to perform TS 3.8.1.1 Action a, only for the affected 4kv bus and affected diesel generator. It is not necessary to verify power sucolies or start the diesel generators on the
'
_ unaffected 4ky busses.
PSRC APPROVED YES: ! NO: DATE: O lbf89
^ \
APPROVED: W DATE: 8 b c .c - -
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NUCLEAR POWER GENERATION - XI3.ID3 ATTACHMENT 6.1-TITLE: REQUEST - TECHNICAL SPECIFICATION INTERPRETATION
--
.
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I REQUEST PSRC INTERPRETATION NUMBER: 95-03 REVISION: 1 TECHNICAL SPECIFICATION: TS 3.3.1.1, "AC SOURCES - OPERATING" QUESTION / CONCERN: When should the 230 kV system be considered incoerable?
.
SUBMITTED BY: vh-,
. DATE: Y N REVIEWEDBYSEbiMOIRECTOR, \. b < .
-
DATE: _'I ik #//h REVIEWED BY REGULATORY SERVICES: "I ile DATE: y II FSRC INTERPRETATION The'230 kV system shculd be considered inocerable when the OCPP Shift Suoervisor has been notified in accordance with Transmission Oceratina 3r0cedure 0-23. "Oceratinc ' Instructions For Deliable Transmission rvice For Se 3f abic Canyon F.P.". cf system incoerability by the Diablo Canyon Switchina Canter. Grid Oceratiens Schedulinc. or the Grid Shift Supervisor.
SSRC APPROVED YES: Y NO: DATE: Oli/96 n , APPROVED: N -h DATE: li 4j Plani MANAGER EFFECTIVE DATE: 4h CISTRIEUTION REydIRED YE3: NO: , . .. .. . _ _ - -- - ---- -
p. ] NUCLEAR POWER GENERATION XI3.ID3 ASTACHMENT 6.1 TITLE: RE00EST FOR TECHNICAL SPECl?ICATION INTERPRETATION _ _ _ _ _ . _ I REQUEST PS?.C INTERPRETATION NUMBER: 95-03 REVISION: 1 TECHNICAL SPECIFICATION: TS 3.3.1.1, "AC SOURCES - OPERATING" QUESTION / CONCERN: When should the 230 kV system be considered inccerable? __ SUBMITTED BY: .k-, DATE: U 'N REVIEWED BY SECT M OIRECTOR, b /e . _ DATE: '\- i h # [.1 // KEVIEWED BY REGULATORY SERVICES: CC OATI: Y!' 2-
!/ .
II PSRC INTERPRETATION The 230 kV' system shculd be considered inocerable when the OCPP Shift Suoervisor has been notified in accordance with Transmission Oceratina Pr0cedure 0-E3, "Oceratine Instructions For Reliable Transmission Service For
:iablo Canyon P.P.".
Of-system incoerability by the Diablo Canyon Switchinc ' Canter. Grid Oceraticns Schedulinc. or the Grid Shif t Suoervisor. i
?SRC APPROVED YES:_'/ NO: DATE: Ol9/96 APPROVED: N .h PLANT .vANAGER DATE: I49/
EFFECTIVE DATE: 44l e GISTRIEUTION REydIRED YES: NO:
. . . . . . . . . . . . . . . . . . . . .
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*:* UNCONTROLLEDPROCEDUREiDONOTUSETOPERFORMWORKn. .. _._,..,___._ l ISSUEFOR'UXi**
PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-8B DIABLO CANYON POWER PLANT REVISION 8B PAGE 19 OF 35 TITLE: CONTROL ROOM INACCESSIBILITY - IIOT UNIT 1 STANDBY TO COLD SIIUTDOWN FIGURE 3 STEAM GENERATOR LEVEL (WID5 RANGE) CORRECTION CURVES HOT SHUTDOWN PANEL LI-501,502,503,504 (COLD CALIBRATED) 100 m n I H 90 S/G i K I I PRESSURES ll /j 80 1200 PSIG x I [[ 1000 PSIG -% % [[ / 600 PSIG x \ /[ [2 200 PSIG x x N N !b M J 60 50 PSIG N x \ \M/ kb
$ ' NN x MCK#.l y* N ?[// in fjW nn 5 40 s MI !
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# / in IH ' '" # IN 0 ! Ih 0 10 20 30 40 50 60 70 80y\ \1( 90100 83 85 INORMAL)
OPERATING ACTUAL LEVEL BAND 01058408. DOC 02 19
- - _ _ _ _ - _ _ _ - _ _ - _ -
REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to: a. Less than or equal to 1 microcurie / gram DOSE EQUIVALENT I-131, and b. Less than or equal to 100/5 microcuries/ gram of gross radioactivity.
{ j APPLICABILITY: MODES 1, 2, 3, 4, and 5. l l ACTION: MODES 1, 2, and 3*: a. With the specific activity of the reactor coolant greater than 1 { microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less l than 500*F within 6 hours; and b. With_the specific activity of the reactor coolant greater than i 100/E microcuries/ gram of gross radioactivity, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours.
MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 micro-curie / gram DOSE EQUIVALENT I-131 er greater than 100/E microcuries/ gram of gross radioactivity, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.
SURVEILLANCE RE0VIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be wit,'.in the limits by performance of the sampling and analysis program of Table 4.4-4.
- With Tavg greater than or equal to 500*F.
DIABLO CANYON - UNITS 1 & 2- 3/4 4-25 Amendment Nos. 18 and 17 (next page is 3/4 4-27) August 31, 1987 32972605.4A I TAB 11 28
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20 30 40 50 80 70 80 90 100 l l PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 CI/ GRAM DOSE EQUIVALENT I-131 J DIABLO CANYON - UNITS 1 & 2 3/4 4-27 32972605.4A I TAC 11 29
TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM MODES IN WHICH TYPE OF MEASUREMENT SAMPLE AND ANALYSIS SAMPLE AND AND ANALYSIS FRE00ENCY ANALYSIS REOUIRED 1. Gross Radioactivity At least once per 1,2,3,4 Determination ** 72 hours 2. Isotopic Analysis for 1 per 14 days 1 DOSE EQUIVALENT I-131 Concentration 3. Radiochemical for E 1 per 6 months * 1 Determination *** 4. Isotopic Analysis for a) Once per 4 hours, If, 2#, 3#, 4#, 5# Iodine Including I-131, whenever the I-133, and I-135 specific activity exceeds 1 Ci/ gram DOSE EQUIVALENT I-131 or 100/E Ci/ gram of gross radioactivity, and b) One sample 1,2,3 between 2 and 6 hours following a THERMAL POWER change exceeding 15% of the RATED THERMAL PCWER within a '-hour period.
! DIABLO CANYON - UNITS 1 & 2 3/4 4-28 32972605.4A I TAB 11 30
_ TABLE 4.4-4 (Continued) TABLE NOTATIONS
# Until the specific activity of the Reactor Coolant System is restored within its limits.
i Sample to be takei, after a mi.nimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
> ** A gross radioactivity analysis shall consist of the quantitative I measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines. The total specific activity shall be the ei of the degassed beta gama activity and the tttal of all identified gaseous activities in the sample within 2 hours after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides.
A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radiciodines, which is identified in the reactor coolant. The specific activities .for these individual radionuclides shall be used in the determination of E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.
t DIABLO CANYON - UNITS 1 & 2 3/4 4-29 1 32972605.4A I TAB 11 31
REFUELING OPERAT10NS
.
3/4 9.14 SPENT FUEL ASSEMBLY STORAGE SPENT FUEL POOL REGION 2
. LIMITING CONDITION FOR OPERAil0N l 3.9.14.1 The following conditions shall be met forstorage of fuel assemblies in Region 2 of the spent fuel pool:
a. The combination of initial enrichment, fuel pellet diameter, and cumulative bumup of the assemblies is within the acceptable area of Figure 3.9-2: or - b. The assemblies are put into a checkerboard pattem with water cells or non-fissile material.
APPLICABILITY: .Whenever fuel assemblies are in the spent fuel pool.
ACTION: a. With the requirements cf the above specification not satis 5ed, suspend all movement of fuel assemblies and crane cperations (with loads in the fuel storage area) except to perform the following: move the non-complying fuel assemblies into compliance with the'above specification or Specification
'3.9.14.3. Until the requirements of the above specification and Specification 3.9.14.3 are satisfied,-
bcron concentration cf the spent fuel pool shall be verified to be greater than er ecual to 2000 ppm at least cnce per 8 hours.
b. The provisions of Spec!5 cations 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 2.9.14.1 The cumulative bumup of each spent fuel assembly stcred in Region 2 shall be determ ned by analysis of its bumup histcry, prier to storage in Regica 2. A complete reccrd of initial enrichment, fuel pe!!et diameter, and the cumulative bumup analysis shall be maintained for the time pericd that each fuel assembly remains in Region 2 cf the spent fuel peci l' t OlABLO CANYON UNITS 1 & 2 3/4917' Unit 1 - Amendment No.116 32973102.4a TA3 16 13 Unit 2 - Amendment No.114
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:lABLO CANYON . UNITS 1 & 2 3/4 9-18 Unit 1. Amendment No.104 l 2 m 3.03.a na u - i Unit 2. Amendment No.103 l
July 7,1995 i j
.'REFUEUNG OPERATlONS - - SPENT FUEL ASSEMBLY STORAGE SPENT FUEL POOL REGION i - LIMITING CONDITION FOR OPERATION 3.9.14.3 The following conditions shall be met for storage of fuel assemblies in Region 1 of the spent fuel pool a. The initial enrichment is 4.5 weight percent U-235 orless; or b .
The initial enrichment is from 4.5 up to a maximum of 5.0 weight percent U-235, and any of the following conditions are met: 1) The combination ofinitial enrichment and cumulative bumup of the assemblies is within the acceptable area of Figure 3.9-3; or 2) The assemblies initially contained a minimum of a nominal 36 mg/in. per assembly of the isotope B 10 integrated in the fuel rods; cr 3) The assemblies are put in a checkerboard pattem with any of the following: a) waterce!!s,or i b) assemblies that initially contained a minimum of a nominal 72 mg/in. per assembly of the isotope 8-10 integrated in the fuel rods, or { i
c) partially irradiated fuel of at least 8000 MWD /MTU cumulative bumup; cr 4) The assemblies are put into a pattem with altemate rows of fuel assemblies and water cells.
A::UCASIUTY: Whenever fuel Essemblies are in Region 1 of the scent fuel pool.
ACSCN: a. With the requirements of the above speci5 cation not satisfied, suspend all movement of fuel assemblies and crane ccerations (with loads in the fuel storage area) except to perform the felicwing: move the non-ccmplying fuel assemblies into a pattem that complies with requirements of the above spec!5 cation or Speci5 cation 3.9.14.1. Until the requirements of the above specification and Specification 3.9.14.1 are satisfied, boron concentration of the spent fuel pool shall be verified to be greater than er equal to 2000 ppm at least once per 8 hours.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
CIABLO CANYON -UNITS 1 & 2 3/4 9-20 Unit 1 - Amendment No.116 32 n2103.4a ns is 21 Unit 2 - Amendment No.114
-.
-
REFUELING OPERATIONS SPENT FUEL ASSEMBLY STORAGE SPENT FUEL POOL REGION 1 SURVEILLANCE REQUIREMENTS.
. . . . 4.9.14.3 :The cumulative bumup of each fuel assembly stored in Region I shall be determined by analysis of its i bumup history, prior to' storage in Region 1. A complete record of initial enrichment, initial integral baron content, and the cumulative bumup analysis shall be maintained for the time period that the fuel assembly remains in Region i 1 of the spent fuelpool.
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i DIABLO CANYON POWER PLANT JANUARY 1999 EXAMINTION I SENIOR REACTOR OPERATOR WRITTEN EXAM ANSWER KEY 1. d 26.d 51.a 76.c 2. b 27.a 52.b 77.a 3. b 28.d 53.b 78.a 4. a -20.d # 54. c 79.b
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UNITED STATES
/ t NUCLEAR REGULATORY COMMISSION i ~
f REGloN IV
\7 611 RYF 8LAZA DRIVE, sulTE 400 ll oN, texas 76011-8064 y,g ARL'
l February 16, 1999
Gregory M. Rueger, Senior Vice President I and General Manager l Nuclear Power Generation Bus. Unit ) Pacific Gas and Electric Company l Nuclear Power Generation, B32 77 Beale Street,32nd Floor P.O. Box 770000 ; San Francisco, California 94177 i SUBJECT: NRC INSPECTION REPORT NO. 50-275/99-301; 50-323/99-301
Dear Mr. Rueger:
From January 25-28,1999, an NRC inspection was conducted at your Diablo Canyon Nuclear Power Plant, Units 1 and 2, reactor facilities. The enclosed report presents the scope and l results of that inspection. l The inspection included an evaluation of six applicants for senior operator licenses. We determined that all applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
During the inspection, your staff submitted greater than 5 percent of the written examination questions for consideration of answer key changes post administration of the examination. In accordance with NUREG -1021, ES 501, Section C.2.c, the NRC requests that you explain why i so many post-examination changes were necessary and what actions will be taken to preclude i similar recurrence on future examinations. l ( In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room (PDR).
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, John L. Pellet, Chief / Operations Branch j[ Division of Reactor Safety ) Docket Nos.: 50-275;50-323 License Nos.: DPR-80; DPR-82 03 k OfYO ^ N f/
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Pacific Gas and E!sctric Company -2-
Enclosure:
NRC Inspection Report No.
50-275/99-301;50-323/99-301
REGION IV== Docket Nos.: 50-275;50-323 License Nos.: DPR-80; DPR-82 Report No.: 50-275/99-301;50-323/99-301 Licensee: Pacific Gas and Electric Company Facility: Diablo Canyon Nuclear Power Plant, Units 1 and 2 Location: 71/2 miles NW of Avila Beach Avila Beach, California Dates: January 25 to 28,1999 Inspectors: T. O. McKernon, Chief Examiner, Operations Branch R. E. Lantz, Examiner, Operations Branch Accompanying Desirce Smith, Examiner in Training, Region lil Personnel: Approved By: John L. Pellet, Chief, Operations Branch Division of Reactor Safety ATTACHMENTS: Attachment 1: SupplementalInformation j Attachment 2: Post Written Examination Comments Attachment 3: Written Examination and Answer Key - l $4/)72/ln/G ~ $(-f
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2-EXECUTIVE SUMMARY
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Diablo Canyon Nuclear Power Plant, Units 1 and 2
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NRC Inspection Report No. 50-275/99-301; 50-323/99-301
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NRC examiners evaluated the competency of six senior operator applicants for issuance of , operating licenses at the Diablo Canyon Power Plant, Units 1 and 2. The licensee developed the initial license examinations using NUREG-1021, " Operator Licensing Examination - Standards for Power Reactors," Interim Revision 8. NRC examiners reviewed, approved, and ' administered the examinations. The initial written examinations were administered on , January 25,1999. The NRC examiners administered the operating tests on January 26-28, l- 1999.
, Operations l '
* . All six applicants demonstrated the requisite knowledge and skills to satisfy the j requirements of 10 CFR Part 55 and were issued senior operator licenses 1 (Section O4.1).
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*- Overall,' licensed operator applicants performed well during the examination. Operators demonstrated good 3-way communications practices, peer checking, and crew briefs.
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l No generic performance weaknesses were identified (Section O4.2).
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* The licensee's initial examination submittal was con;idered acceptable for administration requiring only minor enhancement suggestions. However, subsequent post-written examinations resulted in, the licensee commenting on six written examination questions, which required justification by the licensee and an explanation of how future post- - ,
examination comments will be minimized (Section 05.1). l l l Plant Support j
* Housekeeping and material condition of the plant observed coincident to plant walkthroughs was good (Section F8.1).
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Summary of Plant Status Both units operated at essentially 100. percent power for the duration of this inspection.
. 1. Operations' s 04 Operator Knowledge and Performance 04.1 ~ Initial Written Examination a a. Inspection Scope The licensee developed the written examination with dedicated training instructors on the _ security agreement and used facility training and operations staff ori security agreement to validate the examination. The licensee proctored the administration of the written examination to the license applicants on January 25,1999. The licensee staff; proposed grading for the written examinations, analyzed the proposed results, and _ { presented their evaluation and draft resultant comments for examination revision to the j chief examiner on January 28,1999. The licensee formally transmitted the examination comments to the NRC on February 1,1999.
b. Observations and Findinas
. The minimum passing score was 80 percent. All applicants (six senior operators)
passed with scores ranging from 80.8 to 88.8 percent, with an average score of I 85.6 percent. The NRC specifically notified the licensee leaming services ! representative of one individual, who passed with a score of 80.8 percent, for consideration of additional enhancement or remedial training. The grades reflected the results after incorporation of the accepted examination changes recommended by the licensee.as a result of post-examination question analysis were incorporated. The NRC also revised one additional question based on post-examination analysis.
The licensee provided comments and the appropriate references for six questions as described in Attachment 2. Three questions were recommended for deletion: SRO Question 29 because of depth of knowledge; SRO Question 75 because three of the l choices were correct answers, and Question 94 because of depth of knowledge. l Questions 42,50, and 57 were, revised to accept two correct answers. The chief examiner reviewed and accepted some of the recommendations based on the technical merits of each recommendation and the material references provided by the licensee. I However, the NRC did not accept the recommendation to delete Questions 29 and 75. I
- Question 29.was not deleted because the correct choice did not require detailed knowledge of electrical schematics, but rather whether or not the spent fuel pump automatically started following an accident. Choice d was the only correct answer.
. Question 75 was not deleted but accepted with two possible correct choices, b or d.
- Since the auxiliary feedwater pump draws its steam supply from Steam Generators 1-2
' and 1-3, the operator's decision to choose one of the two steam generators is valid.
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- , -4- . However, Steam Generator 1-2 (choice b) should be selected rather than Steam Generator 1-3 (choice c) because pressure is lower and level is higher.
Choice d, Steam Generator 1-4, which is not hot and dry, is also a correct answer in accordance with Procedure FR-H.1. The licensee's submitted examination comments are included as Attachment 2 to this inspection report.
~ The NRC also reviewed other questions missed by a majority of examinees and
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determined Question 60 to have two possible correct answers (c.hoices b or d).
Choice d was considered a correct choice because the wording of the distractor was not specific enough to discount it as a correct answer.
The licensee's post-examination test analysis indicated that more than half of the applicants missed the same ten questions. Six of the questions were submitted for comment. The chief examiner and the licensee determined that there were no significant inter-relationships to indicate generic weaknesses in knowledge or ability.
The licensee stated that all missed questions would be reviewed with the individuals as part of the training department's remediation prior to assuming shift watch, c. Conclusions All six applicants demonstrated the requisite knowledge and skills to satisfy the requirements of 10 CFR Part 55 and were issued senior operator licenses.
04.2 Initial Operatina Test a. 'Insoection Scooe The examination team administered the various portions of the operating test to the six applicants between January 26-28,1999. Each applicant participated in three dynamic simulator scenarios and received a walkthrough test, which consisted of five system job performance tasks (except for the one senior reactor operator-instant applicant, who performed ten tasks), together with two followup questions for each system. Additionally, each applicant was tested on five subjects in four administrative areas by answering two questions or performing one task for each subject.
i I b. Observations and Findinas The examiners observed effective communications and good peer checks of control
' board activities during the dynamic simulator scenarios. Good status updates and crew briefs were practiced. Good plant and component awareness was observed during the walkthrough portion of the operating tests. The crews utilized effective three-way
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Applicants displayed good knowledge of the location and operation of local plant l components. The applicants responded accurately to the walkthrough followup 4 questions, which indicated a depth of associated system knowledge.
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5-c. Conclusions All applicants passed all sections of the operating test. Operators demonstrated good 3-way communications practices and good peer checks during the dynamic simulator scenarios. Overall, operators performed well during the examination.
05 Operator Training and Qualification 05.1 Initial Licensina Examination Devetooment The licensee developed the initial licensing examination in accordance with NUREG-1021. ,
05.1.1 - Examination Outline a. Insoection Scope The licensee submitted the initial examination outline on September 25,1998. The examiners reviewed the submittal against the requirements of NUREG-1021.
b. Observations and Findinas The chief examiner provided only minor enhancement suggestions related to a balanced mix of malfunctions and power maneuvers in the dynamic sirnulator scenarios. Some other minor enhancements were suggested to the scenarios to ensure that senior operator applicants were evaluated in exercising the facility's technical specifications.
c. Conclusions The licensee's examination outline was acceptable. Minor enhancements suggested by the chief examiner were incorporated.
I 05.1.2 Examination Packaae a. Inspection Scope The licensee submitted the initial examination package on November 20,1998. The chief examiner reviewed the submittal against the requirements of NUREG-1021.
b. Observations and Findinos The licensee submitted 100 draft written examination questions. The chief examiner provided comments or questions on 13 questions. In resolving these comments, the .
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licensee revised or replaced 10 questions. The remaining questions were found to be j satisfactory. The majority of the chief examiner's comments were enhancements and ! not considered substantive. The examinations were acceptable for administration as submitted. Additional review of the examination against the audit examination resulted , in other changes to the job performance measures and some of the scenarios to avoid l I
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any duplication. As discussed in Section 04.1, following post-examination review, one question was deleted and credit for multiple answers on five questions was allowed.
- The failure to make these changes would not have invalidated the examinations or, degraded their discriminatory value. The examinations were considered acceptable for administration as submitted. However, because the licensee submitted greater than 5 percent of the questions for comment the licensee was requested to respond with information related to changes in their examination development process, which will improve future examinations and preclude similar recurrences.
The licensee submitted one set of operating tests, which included a total of ten job performance measures, one administrative tests, three scenarios, and one backup scenario. The submitted scenarios were considered acceptable for administration. H l However, some enhancement suggestions were incorporated during NRC validation to add better balance to the scenarios. The submitted facility walkthrough subsection of the examination discriminated at the required level. Some enhancement suggestions were incorporated to better facilitate the test administration and eliminate any duplication with the audit examination. While some enhancements and revisions to the operating tests were made, the number of revisions was few and the changes did not impact administering the examination.
Final revisions to the examination were completed prior to the examination. The licensee's training department and operations department provided excellent support . during the development and administration of the examination. )
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c. Conclusions The licensee's initial examination submittal was considered acceptable for administration requiring only minor enhancement suggestions. However, the licensee commented on six written examination questions following, which required justification by the licensee and an explanation of how future post-examination comments will be minimized 05.2 Simulation Facility Performance The examiners observed simulator performance with regard to fidelity during the examination validation and administration. The simulation facility supported the examination administration well. No problems were observed.
IV. Plant Support F8 Miscellaneous Fire Protection issucs F8.1 General Comments The examiners observed good plant housekeeping and condition of external panel and equipment coincident with the inplant walkthrough portion of the examination. The facility was reasonably clean, well lighted, and the floors were clear and free of debris.
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- -7-V. Manaaement Meetinas X1 Exit Meeting Summary The exarniners presented the inspection results to members of the licensee management at the conclusion of the inspection on January 28,1999. The licensee acknowledged the findings presented.
The licensee did not identify any information or materials examined as proprietary during the inspection.
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ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee S. Kettlesen, Supervisor, Licensing G. Goelzer, Acting Operations Director T. Blake, Learning Services Director D. Bums, Training instructor R. Jett, Training Leader J. Haynes, Training Leader J. Molden, Operations Manager B. Garrett,' Operations Director-J. Becerra, Instructor . l
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ATTACHMENT 2 Facility initial License Written Examination Comments l l l !
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ATTACHMENT 2., Pacific Gas and
,,& : llecqicfonspany David H. Oatley Diablo Canyon Power Plant Vce President-Diablo Canyon PO Bon 56 Operations and Plant Manager Avda Beach. CA 93424 sos 54ssooo January 29,1999 PG&E Letter DCL-99-012 Thomas O. McKernon, Chief Examiner U.S. Nuclear Regulatory Commission, Region IV 611 Ryan Plaza Dr., Suite 400 Arlington, TX 76011-8064 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 NRC License Written Examination Formal Comments dear Mr. McKernon:
In accordance with NUREG 1021, Interim Revision 8, PG&E is providing the enclosed formal comments on the written examination administered to Diablo Canyon Power Plant license candidates on January 25,1999.
PG&E appreciates the NRC staff efforts during the entire examination and review cycle.
If you have any questions, please contact Roger Jett, Operations Training
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Supervisor, at (805) 545-3439.
Sincerely, f_. M O David H. Oatley Enclosures cc: Timothy M. Blake David L. Burns Roger L. Jett David L. Proulx TLH/1753
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DATE: EXPIRES: ** j
***; ISSUED FOR USE BY:
PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-12C NUCLEAR POWER GENERATION REVISION 8 DIABLO CANYON POWER PLANT PAGE 1 OF 13 ABNORMAL OPERATING PROCEDURE UNITS l
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TITLE: DROF!;ED CONTROL ROD g n/s/91 EFFECTIVE DATE
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PROCEDURE CLASSIFICATION: QUALITY RELATED 1. SCOPE 1.1 This procedure provides instructions for plant operation when a control rod becomes disengaged from its drive mechanimm and drops into the core.
2. SYMNOMS 2.1 Control rods stepping out (if select switch is in auto) 2.2 Rapid drop in TAVG indication 2.3 Rapid drop in reactor power 2.4 Rod bonom light (DRPI panel)_ 2.5 Possible power range high flux rate status light and P-250 printout 2.6 Possible Main Annunciator Alarms-2.6.1 PWR RNGE DEV/QPTR (PK03-10) a. Pwr Rnge lower Quadrant Power Tilt b. NIS Pwr Rnge Channel Flux Deviation c. Pwr Rnge Upper Quadrant Power Tilt 2.6.2 DRPI Failure / Rod Bottom (PK03-21) a. Rod Position Indicator Rod Bottom 2.6.3 TAVG DEVIATION FROM REF (PK04-03) a. TAVG Deviation TAVG-TREF Lo 2.6.4 Rod Cont Urgent Failure (PK03-17) a. Rod Cont Sys Urgent Failure 2.6.5 P-250 Rx Alm Axial Flux / Rod Pos (PK03-25)
. a. P-250 Computer Axial Offset Alarm b. P-250 Computer Rod Position Dev or Rod Bank Sequence 00095808.DoA 02_ 1 1030.01 %
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-12C
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DIABLO CANYON POWER PLANT REVISION 8 PAGE 2 OF 13 TITLE: DROPPED CONTROL ROD UNITS 1AND2 ,,..
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-. ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 1. ONLY One Control Rod Dropped Trip the Reactor and GO TO EOP E-0, REACTOR TRIP OR SAFETY INJECTION 2. PLACE Rod Controlin MANUAE, 3. STOP Any Load Channe In Progress AND Allow Conditions To Stabilize: a. Reactor is critical. a. Place the reactor in Mode 3 by fully inserting the control rods. GO TO OP L-5, PLANT COOLDOWN FROM MINIMUM LOAD TO COLD SHUTDOWN 4. ADJ1!SITurbine Load To Match TAVG AND TREF 5. CHECK Axial Flux Difference Within Refer to Tech Spec 3.2.1.
Tech Spec Lhnits l
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6. Calculate OPTR ner STP R-25: a. Verify LESS than 1.09. a. Refer to Tech Spec 3.2.4 action b. i b. Verify LESS than 1.02. b. Refer to Tech Spec 3.2.4 action a.
7. VERIFY Rod Control System Had No Urnent Failggg: a. Verify Rod Cont Urgent Failure 1. Do not attempt to move rods or reset the (PK03-17) - OFF Urgent Failure.
2. Contact MS for trouble shooting.
3. Refer to AR PK03-17, ROD CONT URGENT FAILURE.
00095808.DoA 02 2 1030.0106
, PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-12C
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DIABLO CANYON POWER PLANT REVISION 8 - PAGE 3 OF 13 TITLE: DROPPED CONTROL ROD UNITS 1AND2 i
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ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED 8. FIND And Correct Cause Of The Dropped Rod: NOTE: See Appendix A for typical powe.- abinet fuse arrangement. Refer to STP R-1B, Attachment 7.4 (7.5 on Unit 2) for specific tuse locations.
a. Dispatch an operator to the affected rod control cabinet to check indicator fuses for the lift, moveable, and stationary grippers b. Contact MS to initiate troubleshooting and repairs c. Rehr to Tech Specs 3.1.1.1 and 3.1.3.1 NOTE: The lift, stationary, or moveable gripper coils can have a blown fuse and not have an urgent failure alarm because the regulation failure cards look at auctioneered high current from all four coils.
9. CONTACT Reactor Enmineerina Renarding the Dropped Rod to Obtain: a. Guidance on rate of control rod movement during recovery b. Power level at which recovery should be performed 10. Record Adeaunte Data to Track: a. How long rod has been dropped.
b. Movement of other control rods during the subsequent recovery
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00095808.DOA 02 3 1030.0106
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-12C
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REVISION 8 ' DIABLO CANYON POWER PLANT PAGE 4 OF 13 - TITLE: DROPPED CONTROL ROD UNITS 1AND2 !
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ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED ; 11. ESTABLISH Initial Recoverv ) I Conditions: a. uce reagtor power to WS WAN a. Check the time since the rod 50%. Contmue with S*.ep 11.c when l dropped - LESS THAN ONE hour power is LESS THAN 50%. b. Reduce, power as necessary such that the steady state power level attained after the rod is recovered is l LESS THAN 90% Reactor Power I c. Set TAVG 1.5'F below TREF by inserting control bank rods as necessary 12. PREPARE For Rod Withdrawal: a. Select the affected rod bank on the Bank Selector Switch b. Record the step counter position on the affected group Bank G mup Step c. Reset the step cower to zero v.. the affected group only d. Locally open lift coil disconnect switches on all rods in the affected bank exceot the droceed rod. (Lift Coil Disconnect Cabinet 115' Elev I Aux Bldg) e. Check dropped rod is in a control e. GO TO Step 12.g.
bank
-THIS STEP CONTINUED ON NEXT PAGE- '
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-6 NUCLEAR POWER GENERATION REVISION 11 DIABLO CANYON POWER PLANT PAGE 1 OF 6 ABNORMAL OPERATING PROCEDURE UNITS i l TITLE: EMERGENCY BORATION AND APPROVED: 06/04/97 06/OS/97 DATE EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED I 1. SCOPE ! 1.1 This procedure covers situations which require emergency boration and the . Sods for accomplishing this operation. Various options of emergency boration are discussed in this procedure.
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1.2 The preferred option is using the VCT Makeup System. The next option is borating I through the emergency boration valve (CVCS-8104). The next alternate option is the I use of the RWST. The use of manual emergency borate valve CVCS-8471 is too )' involved and takes so much time that it is ONLY USED as the LAST option.
NOTE: Emergency boration is defined as a flow GREATER THAN 30 GPM of 7,000 to 7,700 PPM boron or equivalent. Channel inaccuracies have been included where appropriate yielding flow values of greater than 30 GPM.
2. SYMPTOMS Any one of the following conditions requires emerge:.O imation of the specified amount: 2.1 Control rods inserted below the low-low insertion limit when critical.
ROD BANK LO LO INSERTION LIMIT (PK03-14) j 2.2 Failure of any 2 control rods to fully insert following a reactor trip as indicated by rod position indication and rod bottom lights.
2.3 Uncontrolled Reactor Coolant System cooldown following a reactor trip with no ESF action.
2.4 Uncontrolled er unexplained reactivity increase as indicated by: 2.4.1 Unexplained control rod insertion.
2.4.2 Increasing TAVG or nuclear power with no increased load demand.
2.4.3 Unexpected increasing count rate when shutdown.
2.5 When boration is required and normal boration thrcugh the VCT makeup system is not possible.
2.6 Shutdown margin less than acceptable minimum limits pa.t Tech Spec 3.1.1.1, 3.1.1.2 and 3.9.1.
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00307111. DOC 02 1
' ** UNCONTROLLEDPROCEDURE-DONOTUSE TOPERFORM WORKNISSUE$0$US$5,$
' PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-6 DIABLO CANYON POWER PLANT REVISION 11 PAGE 2OF6 TITLE: EMERGENCY BORATION UNITS 1AND2 ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED ) NOTE 1: 900 gallons of 4% boric acid provides 100 ppm INCREASE TO THE RCS, BOL. Calculated l values may be used in place of this thumbrule.
NOTE 2: If Letdown is NOT in service, then it will be necessary to cool down 50 F per hour while I injecting Boric Acid at 30 gpm in order to maintain a constant pressurizer level.
1. INITIATE Emereency Boration: i a. Verify GREATER THAN 55 gpm a. GO TO OP AP-17, LOSS OF CHARGING. l charging flow to the RCS j b. Place VCT make up control in BORATE position c. Set boron Pow controller HC-110 pot c. Increase demand manually to 100% on setting to 9.5 turns HC-110.
d. Set integrator for desired gallons of boric acid. Refer to Appendix A for boration requirements c. Place M/U controller 1/MU in e. Perform the following: START position - Adjust HC-110 pot Verify BA Transfer Pp - HIGH 1) setting to obtain GREATER THAN SPEED 32 GPM of boric acid flow 2) E VCT pressure GREATER l THAN 30 PSIG, THEN Vent the VCT by opening l l CVCS-8101 until LESS THAN 30 PSIG 3) E Boric acid flow remains ; LESS THAN 32 GPM, ' l THEN GO TO Step 2.
f. GO TO Step 3 ooymit. ooc 02 2
*** UNCONTROLLEDPROCEDURE-DONOTUSETOPERFORMWORKoGSS0$$5NNE$d .
PACIFIC GAS AND ELECTRIC COMPANY NUMBER OP AP-6 DIABLO CANYON POWER PLANT REVISION 11 PAGE 3OF6 TITLE: EMERGENCY BORATION UNITS 1AND2 l ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: Emergency Boration Flowmeter FI-113 may peg high at 50 GPM. XFIT-113 in the Cable Spreading room may be used for higher flowrates or to determine total gallons of boric acid added via the Emergency Boration flowpath.
2. INITIATE Alternate Boration Method a. OPEN CVCS-8104 and verify Perform one of the following in order of greater than 33 GPM Emergency preference: Boration Flow 1) Swap Charging Pp suction to the RWST.
a. OPEN 8805A AFLD 8805B.
b. CLOSE LCV-112B Ah'il LCV-112C. 1 I c. VERIFY GREATER THAN 1 105 GPM charging flow.
l na 2) Locally OPEN CVCS-8471 (100' Blender Room).
3. CIIECK Sufficient Boric Acid Available: In Service Boric Acid Tank level a. Stop the Boric Acid Transfer Pp not GREATER THAN required gallons of Boric aligned to the blender.
Acid per Appendix A b. Locally OPEN CVCS-8476, Boric Acid Transfer Pp crosstie. (100' Behind Suction i to BA Transfer Pp 1-1/2-2).
WHEN Sufficient BA inventory restored, THEN Realign the system per OP B-lC:ll,4% BORIC ACID SYSTEM - PLACE IN SERVICE.
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- UNCONTROLLEOPROCEOURE-00NOTUSETOPERFORMWORKorISSUEFORUSEdsj 1/98 FOLDOUT PAGE FOR EOP FR-H.1 Page1of1 1.0 SECONDARY INTEGRITY CRITERI A l E Any S/G Pressure is decreasing in an Uncontrolled manner or has completely depressurized, AND has NOI been isolated, unless it is needed for RCS cooldown, IEEN GO TO EOP E-2, FAULTED STEAM GENERATOR ISOLATION, Step 1.
2.0 BLEED AND FEED CRITERIA l E a. WR S/G Level in any 3 S/Os LESS THAN 23% [34%), AND ALL NR S/G Levels are LESS THAN 6% [16%), OR b. PZR Pressure is GREATER THAN 2335 PSIG due to a loss of secondary heat sink,
.THEN STOP ALL RCPs l
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AND Initiate Bleed and Feed, Steps 12 through 18.
3.0 RESTART SAFEGUARDS EQUIPMENT AFTER LOSS OF OFFSITE POWER l E Offsite Power is lost AFTER Si RESET, THEN o Restart Safeguards equipment as necessary oE in recirculation mode, TIIEN CCPs should be held in STOP/ RESET until RHR is in service.
o REFER TO APPENDlX A for guidance.
4.0 COLD LEG RECIRCULATION SWITCHOVER CRITERION l E RWST Level decreases to LESS THAN 33%, IHEN GO TO EOP E-1.3, TRANSFER TO COLD LEG RECIRCULATION.
5.0 CONTMT SPRAY INITIATION CRITERIA l E Contmt Pressure is GREATER THAN 22 PSIG, IEEN Initiate Contmt Spray.
6.0 ESTABLISillNG FEED TO A HOT DRY STEAM GENERATOR CRITERIA l A " Hot Dry" S/G is a S/G with Hot Leg temperature GREATER THAN 550 F AND WR level LESS TilAN 7% [17%). Feeding a Hot Dry S/G should be performed only when another intact S/G is NOIavailable for cooldown. (When depressuriz-ing a S/G to inject a low pressure water source per Step 18, all S/Gs do not have to be Hot and Dry prior to feeding.)
oE Hot Leg temperatures are increasing, TilEN Feed ONE Hot Dry S/G at MAXIMUM rate.
. WEEN Hot Leg temperature is LESS THAN 550 F THEN Check for SGTR. Use another S/G if a SGTR exists.
oE Hot Leg temperatures are stable or decreasing, IHEN IMPLEMENT EOP FR-H.S.
7.0 AFW SUPPLY SWITCHOVER CRITERION I IE CST level decreases to WSS THAN 10%, IEEN IMPLEMENT OP D-1:V, ALTERNATE AFW SUPPLIES.
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Radioictiva Gaseous Efflu nt Monitoring Instrumentition 39.4
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39.0 INSTRUMENTATION 39.4 Radioactive Gaseous Effluent Monitorina Instrumentation ECG 39.4 The Radioactive Gaseous Effluent monitoring instrumentation channels shown.in Table 39.4-1 shall be OPERABLE * with their alam/ trip setpoints set to ensure the limits of the Radiological :
-Monitoring and Controls Program (AP A-81) are not exceeded. The !
Alarm / Trip Setpoints of these channels shall be detemined and adjusted in accoroa.1ce with the methodology and parameters in the offsite dose calculation procedure (CAP A-8).
APPLICABILITY: In accordance with Table 39.4-1.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME One or more required Perform Required Actions as As specified in radiation monitors specified in Conditions applicable ACTION channels listed in referenced
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in Table 39.4-1. conditions.
Table 39.4-1 inoperable.
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* As described in the Diablo Canyon Power Plant Technical Specifications.
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Diablo. Canyon Units 1 & 2 Rev. 7 39-4R7.48 39-15
Radiccctiva Gaseous Efflu:nt Monitoring Instrumentction
, 39.4 Table 39.4-1 Radiological Gaseous Effluent Monitoring Instrumentation REQUIRED ECG 39.4 NiseER OF . REQUIRED ACTION FUNCTION CHANNELS MODE CONDITION 1. GASEDUS RADWASTE SYSTEM Noble Gas Activity Monitor -
Providing Alars and Automatic , 1 At all A times Termination of Release (RM-22) 2. Plant Vent System a. Noble Gas Activity Monitor - 1 per Unit At all Providing Alars B times R'4-14 or 14R b. Iodine Sampler (the cartridge and filter only, 1 At all C times associatedwith):
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RF-24 or RF-24R c. Particulate Sampler 1 At all (the cartridge and filter times C only,associatedwith): RF-28 or RF-28R d. Plant Vent Flow Rate Monitor 1 At all D times FR-12 (Fed frca FT-12 or FT-12R) e. { Iodine Sampler Flow Monitor: 1 At all D times FT-813 or FT-814 ! 3. Containment Purge System TS 3.3.2 TS 3.3.2 TS 3.3.2 (In Accordancs With Tech Spec) TS 3.3.3.1 TS 3.3.3.1 TS 3.3.3.1 Noble Gas Activity Monitor.- 2(1) per 1,2,3,4,6 Providing Alars and Automatic Unit i (2) ' Termination of Release Unit 1: RM-44A and 448 Unit 2: RM-44A and 448 (1) Only 1 channel required in Mode 6.
-(2) During CORE ALTERATIONS * or movement of irradiated fuel within containment.
- As described in the Diablo Canyon Power Plant Technical Specifications.
Diablo Canyon Units 1 & 2 39-4R7.4B Rev. 7 39-16
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INSTRUMENTATION I 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LINITING CONDITION FOR OPERATION I 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip i Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
' APPLICABILITY: As shown in Table 3.3-3. , l ACTION: a. With an ESFAS Instrumentation Channel or Interlock Trip Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Values column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.
b. With an ESFAS Instrumentation CLannel or Interlock Trip Setpoint less ; conservative than the value shown in the Allowable Values column of Table 1
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3.3-4, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the Engineered Safew Feature Actuation System Instrumentation Surveillance Requirements specified ir. Table 4.3-2.
A.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 24 months. Each test shall l include at least one train such that both trains are tested at least once per 48 months and one channel per function such that all channels are tested at least cnce per N times 24 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of ~ Channels" column of Table 3.3-3. . l l l l
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DIABLO CANYON - UNITS 1 & 2 3/4 3-14 Unit 1 - Amendment No. M \ M 119
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32972307.4a 1 TAB 10 15
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TABLE 3.3-3 (Continued) TABLE NOTATIONS
# Trip function may be blocked in this MODE below the P11 (Pressurizer Pressure f I
Interlock) Setpoint.
ff Trip function automatically blocked above P-11 (Pressurizer Pressure Interlock) l Setpoint and is automatically blocked below P-11 when Safety Injection on Steam Line Pressure-Low is not blocked.
dif For Mode 3, the Trip Time Delay associated with the Steam Generator Water Level-Low-Low channel must be less than or equal to 464.1 seconds.
ACTION STATEMENTS ACTION 14 - With the number of OPERABLE' channels one less than the Minimum Channels < OPERABLE requirement, restore the inoperable channel to OPERABLE status I within 6 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.
ACTION 15 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, declare the affected Emergency Diesel Generator (s) inoperable and comply with the ACTION statements of Specification 3.8.1.1; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1.
ACTION 16 - With the number of OPERABLE Channels one less than the Total Number of Channels, declare the affected Emergency Diesel Generator (s) inoperable I and comply with the ACTION statements of Specification 3.8.1.1; however, one channel may be bypassed for up to 2 hours for surveillance testing ,. per Specification 4.3.2.1. ( ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1.
ACTION 18 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves (RCV-11,12,FCV660,661,662,663,664) are maintained closed.
ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
ACTION 20 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied: a. The inoperable channel is placed in the tripped condition within 6 hours, and b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel or one additional channel may be bypassed for up to 4 hours for surveillance testing per Specification 4.3.2.1.
DIABLO CANYON - UNITS 1 & 2 3/4 3-21 Unit 1 - Amendment No. R\S %1QR127 Unit 2 - Amendment No. (%S h lqh 125 32972507.4a I TAB 10 22
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. BSTRUMENTATION
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3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1- The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits.
APPLICABILITY: As shown in Table 3.3-6.
ACTION: a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours or declare the channel inoperable. - b. With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.
c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS _ 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3.
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i l l DIABLO CANYON - UNITS 1 & 2 3/4 3-26 Amendment Nos. 55 and 54 i 32972507.4a I TAB 10 37 June 11, 1990 l l i
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SEE PSRC INTERPRETATION 88-02
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TABLE 3.3-6 (Continued) ACTION STATEMENTS ACTION 30 - With less than the Minimum Channels OPERABLE requirement, operation may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alam Setpoint or an individual qualified in radiation protection procedures with a radiation dose rate monitoring device is provided in the fuel storage pool area. Restore the incperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool areas.
ACTION 31 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1. The provisions of Specification 3.0.4 are not applicable.
ACTION 32 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.12.
ACTION 33 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.9.
ACTION 34 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the Control Room Ventilation System in a recirculation mode with the HEPA filter and charcoal adsorber bank in operation.
DIABLO CANYON - UNITS 1-& 2 3/4 3-38 Amendment Nos. 55 and 54 32972507.4a I TAB 10 39 June 11, 1990 }}