IR 05000361/1998301
| ML20236Q569 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20236Q556 | List: |
| References | |
| 50-361-98-301, 50-362-98-301, NUDOCS 9807200345 | |
| Download: ML20236Q569 (19) | |
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l ENCLOSURE
! U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos.: 50-361; 50-362 License Nos.: NPF-10; NPF-15 Report No.: 50-361/98-301;50-362/98-301 Licensee: Southern California Edison C Facility: San Onofre Nuclear Generating Station, Units 2 and 3 Location: 5000 S. Pacific Coast Hw San Clemente California Dates: June 14-19,1998 Inspector (s): S. L. McCrory, Reactor Engineer, Examiner / inspector, Chief Examiner T. O. McKernon, Reactor Engineer, Examiner / Inspector M. E. Murphy, Reactor Engineer, Examiner / Inspector R. E. Lantz, Reactor Engineer, Examiner / inspector Approved By: J. L. Pellet, Chief, Operations Branch ATTACHMENTS:
Attachment 1: Supplemental information Attachment 2: Facility Licensee Post-Examination Comments Attachment 3: Final Written Examination and Answer Key l
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i 9807200345 980715 hDR ADOCK 05000361 m [y
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-2-EXECUTIVE SUMMARY San Onofre Nuclear Generating Station, Units 2 and 3 NRC Inspection Report 50-361/98-301; 50-362/98-301 NRC examiners evaluated the competency of eight senior operator applicants for issuance of operating licenses at the San Onofre Nuclear Generating Station facility. The NRC developed the initial license examinations using NUREG-1021, " Operator Licensing Examination Standards for Power Reactors," Interim Revision 8. The chief examiner administered the initial written examinations to all applicants on June 14,1998. The NRC examiners administered the operating tests on June 15-18,199 Ooerations
- All applicants passed the written examination (Section 04.1).
. . Seven of eight applicants passed all sections of the operating test. One applicant failed
~ the dynamic simulator portion as a result of poor supervisory performance during response to a small-break-loss-of-coolant accident scenario (Section 04.2).
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Two of the four crews were significantly challenged by the small-break-loss-of-coolant accident scenario. Also, applicant response to a diesel generator vital electrical bus failure to energize was inconsistent and did not meet facility licensee expectations. However, based on review of overall applicant examination performance, the chief l examiner determined that the performance weaknesses displayed were the result of I individual knowledge and ability weaknesses rather than generic or broad training weaknesses (Section 04.2).
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-3-Reoort Details Summary of Plant Status The units operated at essentially 100 percent power for the duration of this inspectio . Operations 04 Operator Knowledge and Performance l 0 Initial Wntten Examination /
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' Insoection Scoce On June 14,1998, the chief examiner proctored the administration of the written examination to eight senior operator license applicants. The facility licensee provided post-examination comments (Attachment 2) at the end of the examination week. The chief examiner reviewed the comments for technical adequacy. The chief examiner concurred on all recommendations, and revised the written examination answer ke The chief examiner graded the written examinations on June 26,1998.
i l ' Observations and Findinas
-The minimum score required to pass the examination was 80 percent. All applicants passed with scores ranging from 82 to 95 percent, with an average score of 88.2 Questions 27,37,38,39,44,58, and 72 were missed by at least half of the applicant The chief examiner reviewed these and related written items. This review determined l that there were no significant relationships among the missed questions to indicate generic weaknesses in knowledge or abilit Conclusions All applicants passed the written examination. Commonly missed questions were due to isolated knowledge deficiencies rather than broad or systemic weaknesse O4.2 Initial Ooeratina Test Insoection Scone
- The examination team administered the various portions of the operating test to the eight applicants on June 15-18,1998. Each applicant participated in two dynamic simulator scenarios and received a walkthrough test, which consisted of ten system tasks together with two followup questions for each system. Additionally, each applicant j was tested on five subjects in four administrative areas with a combination of '
administrative tasks and questions.
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-4-l l Observations and Findinas The examiners observed consistently good three-way communications and supervision of control panel activities during the dynamic simulator and dynamic walkthrough portions of the operating test. Crew briefings were informative and organized and not I excessive in duration.
l l Two of the four crews were significantly challenged by a dynamic scenario that included l a small-break-loss-of-coolant accident with only one charging pump available for high pressure injection. One of the crews was slow to initiate plant cooldown and momentarily lost subcooling, which, while adequate, impeded their ability to depressurize l below the discharge pressure of the low pressure injection systems. The other crew responded to the accident for over 40 minutes without initiating a controlled cooldown and depressurization. The crew lost subcooling and repeatedly expressed confusion while attempting to identify a success path in the functional recovery portion of the emergency operating procedures. The examiners determined that the performance of the applicant in the control room supervisor position for this scenario was unsatisfactory, which resulted in a failure of this portion of the operating examinatio Applicant response to a task to energize a vital electrical bus from its associated diesel generator was inconsistent. One of the job performance measures cued the applicants that a reactor trip had occurred and that Vital Bus 2A04 was de-energized with Diesel Generator 2G002 running with its output breaker open. The applicant was then cued to promptly energize Vital Bus 2A04 from Diesel Generator 2G002. The condition simulated a diesel generator start following a safety injection actuation signal, but with low running voltage and frequency. The appropriate action was to increase voltage and frequency to allow the diesel generator output breaker to shut automaticall The condition produced the annunciator alarm "2A04/3A04 TIE FDR 2A0417 OC/ MANUAL" because the tie breaker shifted to manual control as a result of the safety injection signal. Two applicants observed the actuated annunciator window and immediately concluded that an electrical fault existed on the bus. These applicants placed the diesel generator in maintenance-lockout to shut it down since its cooling water <
pump was not running because the pump was connected to the de-energized bus. At ,
least two applicants observed that the simulator did not indicate a safety injection actuation (which was the result of a cuing error in the simulator setup and initiating cue) '
and, therefore, that the tie breaker should not be in manual. These applicants concluded that the conflicting indications meant that an electrical fault was likely and, therefore, '
l secured the diesel generator by placing it in maintenance lockout. The remainder of the applicants adjusted voltage and frequency to cause the diesel generator's output breaker I to shu The chief examiner discussed this performance with the facility licensee staff. The facinty licensee staff stated that operators were expected to adjust the voltage and frequency to attempt to energize the bus in an emergency situation, with the indications provided, and rely on the fault isolation circuitry to protect the diesel generator if an
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l j~ electrical fault existed. The chief examiner determined that applicants who secured the
!- diese! generator solely due to the annunciator actuation demonstrated unsatisfactory performance and that applicants who adjusted voltage and frequency to cause the output breaker to shut, or who diagnosed the inconsistencies between the simulator indications and electrical distribution condition, and then secured the diesel generator, demonstrated acceptable performanc Conclusions Seven of eight applicants passed all sections of the operating 'est. One applicant failed l l
the dynamic simulator portion as a result of poor supervisory performance during response to a small-break-loss-of-coolant accident scenari Two of the four crews were significantly challenged by the small-break-loss-of-coolant accident scenario. Also, applicant response to a diesel generator vital electrical bus failure to energize was inconsistent and did not meet facility licensee expectation However, based on review of overall applicant examination performance, the chief t examiner determined that the performance weaknesses displayed were the result of individual knowledge and ability weaknesses rather than generic or broad training l weaknesses.
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05 Operator Training and Qualification 05.1 initial Licensino Examination Devetooment The NRC developed the initial licensing examination in accordance with guidance l
provided in NUREG-1021. The facility provided a majority of the reference materials in l electronic media, which was often difficult to use due to poor cross-references or indices l to correlate the file name to its content. Additionally, a large amount of material related ( to Unit 1, which has been in decommissioning for several years, was included but not l readily identified as related to Unit 1. Otherwise, the material adequately supported j development of all sections of the examination.
i l The chief examiner and another member of the examination team validated the operating
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test onsite during the week of April 27,1998. The chief examiner provided the written examination to the facility licensee for technical review on June 1,1998. The facility licensee provided several minor editorial comments to improve question wording or to correct terminology. The facility licensee identified only a few technical errors, most of which were due to design or procedure changes not reflected in the reference material used by the examination develope During the examination week, one of the job performance measures had to be replaced i due to a major procedure revision that had occurred between the operating test validation and the week of administration. The chief examiner approved the use of l another task, of comparable complexity and that evaluated the same safety functio l l
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I from the facility licensee's test item bank. Since the replacement task related to the same systems as the original task, there was no need to revise the prescripted followup question O5.2 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. The examiners observed no problem V. Management Meetings i
X1 Exit Meeting Summary The examiners presented the inspection results to members of the licensee management at the conclusion of the inspection on June 18,1998. The licensee acknowledged the findings presente The licensee did not identify as proprietary any information or materials examined during the inspection.
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ATTACHMENT 1 SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee R. Krieger, Vice President Nuclear Generation M. Jones, Manager, Operations G. Gibson, Manager, Compliance R. Sandstrom, Manager, Training G. Cook, Supervisor, Compliance l K. Rauch, Supervisor, Operations Training l B. Boos, Nuclear Oversight Department l
T. Prey, Compliance ,
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ATTACHMENT 2 FACILITY LICENS;' POST-EXAMINATION COMMENTS
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$OL.lHIRN CalllORNIA R. W. Krieger
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EDISON .
Nudear Generation
An (D150'~ a %TTR\ 4ilo\ Al Comr.my June 25,1998 Mr. E. W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Subject: Docket Nos. 50-361 and 50-362 Comments on Written Operator License Examination San Onofre Nuclear Generating Station, Units 2 and 3 In accordance with NUREG-1021, Chapter ES402 (Section E), enclosed are Southern California Edison's (SCE's) comments on the written portion of the NRC Operator License examinations administered at SONGS Units 2 and 3 during the week of June 15,199 If you have any questions regarding this matter, please contact me or Mr. R. L. Sandstrom, Manager of Training, at (949) 368-8387, S'ncerely,
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(>.l w Enclosures cc: NRC Document Control Desk J. A. Sloan (NRC Senior Resident inspector, Units 2 and 3)
S. L. McCrory (NRC Lead Examiner, Region IV)
i l' O tkn 128 Nn Clemente, C \ 42e74-0128 714 3e80233 F.n714Te8el8)
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ENCLOSURE COMMENTS ON THE OPERATOR LICENSE EXAMINATION l l
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Question The unit was operating at 100% when condenser vacuum was lost due to a ruptured LP turbine diaphragm. The reactor was manually tripped and the SPTAs were initiate Plant conditions are as follows:
- All CEAs are fully inserted
- Pzrlevelis 20% '
- Pzr press is 1875 psia
- S/G water levels are 30% and MFW RTO has actuated
- Containment temperature is 110*F Containment pressure is 1.3 psig Based on the above plant conditions, what operator action is required to be taken in the SPTAs? Override AFW system to feed both S/Gs at greater than 200 gpm Operate SBCS to maintain Tavg between 545-555'F Verify normal containment cooling is functioning properly Operate ADVs to maintain Tavg between 545-555'F
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Time to complete: 4 l l Topic: (N5644) RO/SRO, Lesson Plan 2EO712 Objective Cross Reference: OBJECTIVE Comment: SO23-12-1 Author / Reviewer: Miller, D.Date taught: 12/05/92 KA000051 AK3012.8/ Accept Both A and D as correct answer With a loss of condenser vacuum, Main Feedwater can not supply feedwater to the Steam Generators. Main Feedwater pumps are steam driven and discharge to the Main Condenser. To meet RCS Heat Removal Safety Function, we need greater than 40% level or 200 GPM Auxiliary Feedwater flow. The Auxiliary Feedwater valves will have to be overridden and throttled to 200 GPM flow to each Steam Generator. D is still correct since it is required by the SPTAs to maintain Heat Removal when the Steam Bypass Control System (SBCS) is not available which, also due to the loss of vacuum, is the case in this scenari Reference: SO23-12-1
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NUCLEAR ORGANIZATION EMERGENCY OPERATING INSTRUCTION S023-12-1 UNITS 2 AND 3 REVISION 15 PAGE 7 0F 25
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STANDARD POST TRIP ACTIONS OPERATOR ACTIONS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINEO 8 VERIFY RCS Heat Removal criteria satisfied: I VERIFY feedwater adequate: ) IF any S/G level less than 21%,
1) Level in both S/Gs THEN ENSURE EFAS - actuate greater than 40% NR 2) ENSURE MFW pump Miniflow Valves AND in MODULAT J Feedwater - available to 3) OPERATE feedwater systems to I both S/G raise level in at least one S/G to 40% NR at greater than or f OR equal to 200 GP ) Auxiliary Feedwater flow to I each S/G - greater than or equal to 200 GP OR 3) Main Feedwater flow to each j S/G:
a) At least one MFW Pump
- operating, b) RTO - actuated to both S/Gs.
l c) MFW Pump discharge to l S/ Differential l
t pressure - greater than 25 PSI PREVENTS /Ghighlevel:
1) CLOSE MFW Block Valves:
HV-4047 HV-405 ) VERIFY both S/G levels 2) ENSURE Main Feedwater RTO
- less than 80% N actuate ,
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.' NUCLEAR ORGANIZATION EMERGENCY OPERATING INSTRUCTION 5023-12-1 UNITS 2 AND 3 REVISION 15 PAGE 8 0F 25'
STANDARD POST TRIP ACTIONS OPERATOR ACTIONS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINEO 8 VERIFY RCS Heat' Removal criteria satisfied: (Continued) VERIFY heat removal adequate: c. 1) OPERATE SBCS to maintain Tc
- between 545'F and 555' ) Tc - less than 555* ) IF SBCS NOT maintaining T ) S/Gpressures
- approximately 1000 PSI THEN OPERATE ADVs to maintain Tc
- between 545'F and 555' VERIFY Tc ) ENSURE SBCS valves - close greater than 545'F 2) ENSURE ADVs - close OR 3) CHECK S/G Safety Valves - close controlle ) ENSURE S/G Blowdown Valves
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E-088 HV-4054 E-089 HV-4053 5) a) ENSURE Moisture Separator Reheater isolation valves
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HV-2703 HV-2721 HV-270 b) IF MSR isolation valves CANNOT be verified closed AND Tc uncontrolled, THEN CLOSE MSIV >
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. NUCLEAR ORGANIZATION EMERGEmCY OPERATING INSTRUCTION S023-12-1 UNITS 2 AND 3 REVISION 15 PAGE 9 0F 25
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STANDARD POST TRIP ACTI)NS OPERATOR ACTIONS ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINEO 8 VERIFY RCS Heat Removal criteria satisfied: (Continued) d. (Continued)
6) IF S/G levels (WR) rising, THEN a) VERIFY RTO response 4
- satisfactory AND EVALUATE feeding via 3 Condensate Pumps and 1 MFW Pum b) VERIFY AFW Valves - close c) IF Steam-Driven AFW Pump l
operating,
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l THEN OVERRIDE and STOP
! Electric AFW Pump d) ENSURE step Ba criteria for adequate feedwater 1
- satisfie ) IF Tc less than 500*F, THEN ENSURE at least one RCP
- stoppe VERIFYS/Gpressures-greater ) ENSURE MSIS - actuate than 740 PSI ) RECORD MSIS on Attachment 1 WORKSHEE )
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. Ousstion 5 A Shutdown From Outside the Control Room is being performed per SO23-13-2. Plant conditions are as follows:
- RCS pressure is 2050 psia and s'owly lowerin T- hot is 575* F and slowly lowerin T-cold is 555' F and stabl Pressurizer level (actual)is 21% and slowly lowerin J
- Charging is not establishe l
- S/G E089 level (actual)is 58% narrow rang Which one of the following statements is conect concerning Natural Circulation for the current plant conditions per procedure? Natural Circulation core cooling is established and may be verifie ] Natural Circulation is established but cannot be verified until Pressurizer Level is greater than 30%. Natural Circulation cannot be verified until manual control of the ADV's is established and T-cold is lowerin Natural Circulation is not established because the plant parameters are outside the guidelines.
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B Time to complete: 4 Topic: (N0476) RO/SRO, Lesson Plan 2AO702, Objective 1.3.2
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Cross Reference: Objective 1.3.2 l Comment: SO23-13-2 Author / Reviewer. Jones. Exam Review GroupDate taught: 10/15/92 KA 000A13AK2.2 3.4/3.6
! Answer A is correct instead of Answer B
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i Justification: The reference below states the condition > necessary to verify Natural Circulation. Step 1.3.5 states that pressurizer level must be greater than 40% with Charging available QB greater than 10% until charging is available. Answer A meets the latter of the two conditions, so Natural Circulation can be verified. The two substeps of ;
1.3.5 discuss what pressurizer level should be doing when charging and do not apply to !
when charging is not availabl Reference: SO2313-2, " Shutdown Outside the Control Room," Attachment 15, " Natural Circulation Guidelines"
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NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTION S023-13-2 UNITS 2 AND 3 REVISION 5 PAGE 103 0F 218
, ATTACHMENT 15 NATURAL CIRCULATI0h' VERIFICATION GUIDELINES (Coatinued) Natural Circulation and Adequate Core Cooling exist if the following conditions are present:
l NOTE: 1. Adequate Core Cooling and Natural Circulation can also exist outside of the limits stated belo Inadequate Core Cooling should NOT be based exclusively on one parameter.
! That should lower for the first 2 minutes following RCP trip. For the next 5 to 7 minutes Thot should rise and then begin to stabilize / lower as controlled by Steam Generator
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. steaming rate.
CAUTION Maintain Saturation Margin greater than 20*F Thor to prevent void formation in the Reactor Vessel Hea CAUTION Naintain Saturation Margin less than 200*F Tcold to minimize potential for Pressurized Thermal Shoc .
I 1. That - Not Risin . Tcold - Not Risin . Operating loop Delta T - Less than 58'F [2].
1. RCS Saturation Margin - Greater than 20*F Tho (from Figure 1) - Less than 200*F Teoid.
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[2] Due to instrument loop uncertainties, e telta T between 25'F and 91*F may not be a true indication that natver.: circulation exists. (Ref. 4.4.2)
ATTACHMENT 15 PAGE 2 0F 8
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NUCLEAR ORGANIZATION ABNORMAL OPERATING INSTRUCTION S023-13-2 UNITS 2 AND 3 REVISION 5 PAGE 104 0F 218
, ATTACMMENT 15 l
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NATURAL CIRCULATION VERIFICATION GUIDELINES (Continued)
NOTE: An unexplained rapid rise in PZR level while lowering RCS pressure, or a lowering of PZR
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level while charging, may indicate void formation in the Reactor Vessel Hea . PZR level [1] - Greater than 40% with Charging available, DE Greater than 10% until Charging is available, &HQ:
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.1 Not rising at a rate greater than can be explained by charging rat .2 Not lowering while chargin . At least one S/G level [1] - Greater than 55% N .4 1E Natural Circulation and Adequate Core Cooling is NOT confirmed by Section 1.3, IHEN perform the following:
1. Raise the Steam Generator steaming rat . Raise Steam Generator actual (corrected) level to 80% N . Ensure charging rate is maintaining PZR actual (corrected)
level 240%.
1. Stop Auxiliary Spray flo i i
NOTE: To prevent vuid f ormation in the Upper Reactor !
Vessel Head during Natural Circulation cooldown, an RCS Saturation Margin in the center of Figure 1, is optimu CAUTION Do Not exceed 2275 psia PZR pressure while attempting to establish optimum Saturation Margi . Energize all available PZR Heaters and move RCS Saturation Margin towards the Optimum curve of Figure [1] Actual (corrected) level. 11 a seismic event has occurred, then value must be obtained from the EVSD.
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ATTACHMENT 15 PAGE 3 0F 8
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Qu2stion 7 Unit 2 is operating at 100% power when alarm 50A31 " PRESSURIZER SAFETY VALVE OUTLET TEMP Hi" comes i What is the MINIMUM RCS pressure at which operation can continue with this condition and i
the bases for that pressure limit?
1 psia - A reactor trip would result in RCS pressure going to SIAS setpoin psia - DNBR limits could be exceeded during normal power maneuver psia - A reactor trip would result in RCS pressure going to SIAS setpoin psia - DNBR limits could be exceeded during normal power maneuver Answer C
Reference: SO23-15-50.A. lesson plan 2N1703 KA 000008G2.1.10 2.7/ Answer A is correct instead of Answer C l
Justification: The Annunciator Response Procedure referred to below states "A Reactor j Trip at pressures < 2200 psia may initiate SIAS, therefore DO NOT reduce RCS pressure to less than 2200 psia" (under these conditions).
Reference: SO23-15-50.A2, window 50A31 J
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. NUCLEAR ORGANIZATION ALARM RESPONSE INSTRUCTION S023-15-50.A2 UNITS 2 AND 3 REVISION 5 PAGE 9 0F 84
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ATTACHMENT 3 50A31 PZR RELIEF VALVE OUTLET TEMP HI APPLICABILITY PRIORITY REFLASH ASSOCIATED WINDOWS Modes 1-4 WHITE YES NONE
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INITIATING NOUN NAME SETPOINT VALIDATION PMS ID LINK #
DEVICE INSTRUMENT U2/U3 l
2(3)TSH 0107 Pressurizer Relief Line 150af 2(3)TI-0107 T107 644/666 2(3)TSH-0108 Temperature Switch High 2(3)TI-0108 T108 645/667 REQUIRED ACTIONS: l 1.1 if Pressurizer Safety Valves have operated, then refer to the actions of 5023-15-50.A for window 50A14, CORRECTIVE ACTIONS:
SPECIFIC CAUSES SPECIFIC CORRECTIVE ACTIONS 2.1 Leaking Pressurizer 2.1 If Pressurizer Safety Valves are leaking, Safety Valves then perform the following:
CAUTION A Reactor Trip at pressures
< 2200 psia may initiate SIAS, therefore do not reduce RCS pressure below 2200 psi . Obtain the CRS's permission to lower j RCS pressure to 2200 psi !
2. Slowly reduce RCS pressure to 2200 psia using 2(3)PIC-0100, Pressurizer Pressure Controlle . GO TO S023-13-14, Reactor Coolant Lea .2 High Ambient Containment 2.2 If high Containment A'mbient Temperature, Temperature then request for the I&C department to reset the alarm setpoint to ambient temperature
+25'F.
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NUCLEAR ORGANIZATION ALARM RESPONSE INSTRUCTION S023-15-50.A2 l UNITS 2 AND 3 REVISION 5 PAGE 10 0F 84
, ATTACHMENT 3 50A31 PZR RELIEF VALVE OUTLET TEMP HI (Continued) ASSOCIATED RESPONSES:
3.1 Notify the CRS/SS and the STA to review Tech. Spec. LCO 3.4.1, LC0 3.4.10 and LCO 3.4.13, ano initiate an EDMR/LC0AR, as require .0 COMPENSATORY ACTIONS:
DEVICE NUMBER SPECIFIC COMPENSATORY ACTIONS 4.1 2(3)TI-0107 4.1 Monitor Pressurizer Relief Line temperature 2(3)TI-0108 at least once every 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> .0 REFERENCES:
5.1 NRC Commitments:
5. Technical Specifications f 5.2 Operating Instructions:
5. S023-13-14, Reactor Coolant Leak 5.3 Drawings: 1 5. P& ids:
.1 40111B, Reactor Cooling System l 5. Elementaries: l j
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.1 34556, Cabling Block Diagram Spec 200 Pwr Sply & Nest A NSSS L 140 I
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