ML021790027
| ML021790027 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/21/2002 |
| From: | Hackenberg J - No Known Affiliation |
| To: | Bissett P NRC Region 1 |
| References | |
| Download: ML021790027 (5) | |
Text
May 21, 2002 Mr. Paul Bissett Senior Operations Engineer Division of Reactor Safety 475 Allendale Road King of Prussia, PA 19406
Subject:
Submittal of Facility Comments concerning SRO NRC Examination May 2002 Oyster Creek Nuclear Generating Station NRC Docket Number 50-219 In accordance with NUREG 1021, Revision 8, "Operating Licensing Examination Standards for Power Reactors", Oyster Creek Nuclear Generating Station administered the NRC SRO initial written examination on May 20, 2002. Post exam analysis resulted in identifying four exam questions with two correct answers. The exam question numbers are 33, 64, 70 and 91. Facility comments and justification for the proposed changes were submitted to the Lead Examiner by separate attachment.
In accordance with NUREG 1021, Revision 8, Section ES-402, this letter is submitted in support of the proposed facility comments.
Please contact Greg Young at (609) 971-4196 with any questions concerning this letter or the submitted examination changes.
Respectfully, Jesse Hackenberg Operations Training Manager Oyster Creek Nuclear Generating Station (Delivered only to Paul Bissett, Chief Examiner, NRC Region 1)
Enclosure:
OytrCek20 R
IiilLcnigEa
- 33. The plant was in the process of reducing power to conduct MSIV surveillance testing. The CRO reports the following:
RPV level has increased to 165 inches TAF
- Feed water flow is greater than steam flow Alarm J-5-d, MFRV LOCK UP (A) has annunciated The auxiliary plant operator reported that an airline in the Feed Pump area has broken and is leaking air.
Based on the above conditions, 'which one of the following is the correct initial response?
- a. Implement ABN-3200.35, Instrument Air System Failure Abnormal Operations and repair the broken instrument air line.
- c. Implement OPS-3024.14, Feedwater System - Diagnostic and Restoration Actions, and place MFRV "A" in local manual operation.
- d. Initiate a manual scram lAW ABN-3200.01, Reactor Scram, before a turbine trip occurs at 175 inches TAF.
Facility Comment:
As submitted, answer "C" was indicated to be the correct answer based on the guidance contained in RAP J-5-D, "MFRV LOCK UP". RAP Manual Corrective Action #1 indicates that an Equipment Operator be directed to manually locally control the MFRV. Answer "C" is correct but does require some time to establish.
Answer "D" is to direct a manual scram before the automatic scram setpoint is exceeded.
Station Procedure 106, Conduct of Operations, specifies in step 5.1.1 that the SRO has the duty and authority to shut down the reactor under any of a series of conditions including, "when operating parameters are trending such that an automatic scram is imminent or inevitable."
Since no information is given in the question regarding the rate of increase other than indications that level is rising and will continue to rise, answer "D" is also correct.
Recommended Resolution:
Accept both answers "C" and "D" as correct.
33 NRC Initial Licensing Exam Oyster Creek 2002
Oyster Creek 2002 NRC Initial Licensing Exam
- 64. Given the following:
"* The plant is operating at 50% power
"* The turbine inlet pressure is 995 psig
"* The reactor pressure is 1010 psig Reactor power is being raised to 100 percent with control rods and recirculation flow.
Which one of the following is correct if NO operator action is taken concerning the turbine controls?
- a. The Turbine will trip on "25% Load Trip Not Reset" before reaching 100% power.
- b. The Turbine Bypass Valves will open before reaching 100% power.
- c.
An APRM upscale rod block will occur before reaching 100% power.
- d.
A high pressure reactor scram will occur before reaching 100% power.
Facility Comment:
As submitted, answer "D" is correct because if no adjustments are made to the pressure setpoint, during power ascension, pressure in the reactor will increase and result in a reactor scram.
At Oyster Creek, it is also a common occurrence to receive APRM upscale rod blocks during routine power ascensions.
"* The stem of the question does not provide any information related to the rod pattern and the sequence of control rod and recirc system adjustments.
"* As submitted, the justification for answer "C" indicated that APRM rod blocks are expected.
"* At Oyster Creek, the APRMs have "PUSH-RESET" alarms that must be routinely reset during the power ascension to prevent receiving rod blocks.
"o This action is directed in Step 6.55 of Station Procedure 201, Plant Startup.
"o Station Procedure 403.2, Operation of the LPRM-APRM System During Startup and Power Operation, precaution 4.2 states, "Failure to acknowledge the APRM ALARM LEVEL - PUSH-RESET indicator each time it illuminates will result in a rod block if power is allowed to increase."
"* Procedural guidance and operational experience have demonstrated that answer "C" is also a correct answer.
Recommended Resolution:
Accept both answers "C" and "D" as correct.
64 NRC Initial Licensing Exam Oyster Creek 2002
NRC Initial Licensing Exam
- 70. Given the following:
The first plant startup of the year is in progress Reactor power is 7%
Rod withdrawals are in progress to increase power All systems are aligned and operating normally for this power level When the reactor operator selects the next control rod within the group, the Rod Worth Minimizer (RWM) failed.
Based on the above conditions, which one of the following is correct?
- a. The RWM may NOT be bypassed and the plant startup may NOT continue until the RWM is returned to service lAW Procedure 409, Operation of the RWM.
- b.
The RWM may be bypassed and the startup may continue lAW Procedure 201, Plant Startup. A second licensed operator and additional Reactor Engineering assistance are NOT required..
- c.
The RWM may be bypassed and the startup may continue lAW Procedure 201, Plant Startup after a second licensed operator is stationed. No additional Reactor Engineering assistance is required.
- d.
The RWM may be bypassed and the plant startup may continue lAW Procedure 201, Plant Startup after a second licensed operator and a Reactor Engineer are stationed.
Facility Comment:
As submitted, answer "C" was indicated to be the correct answer based on the guidance contained in step 5.3 of Station Procedure 201, Plant Startup. This answer is correct because the procedure does not require specific Reactor Engineering assistance under these conditions.
Answer "D" is similar to answer "C", except that it does not address Reactor Engineering assistance, but rather whether a Reactor Engineer is required to be stationed. In the same procedure, 201, in the Precautions and Limitations section, step 4.12 requires a Reactor Engineer to be present in the control room during all core reactivity changes. Answer "V" is also correct.
Recommended Resolution:
Accept both answers "C" and "D" as correct.
70 Oyster Creek 2002
Oyster Creek 2002 NRC Initial Licensing Exam
- 91. The maintenance department has prepared a Temporary Modification (TM)
Package and submitted it to Engineering. The Temporary Modification requires installation of blank flanges to isolate a CRD pump for 4 days for maintenance.
As the Shift Manager conducting a review of the TM, in addition to the Maintenance Risk Assessment and a 50.59 Screening Review, what is the highest additional review required by Station Procedure 108.8, Temporary Modification Control.
- a.
Second independent 50.59 screening review.
- b.
Nuclear Safety Review Board (NSRB) review.
- c.
Plant Operations Review Committee (PORC) review
- d.
Nuclear Regulatory Commission (NRC) review Facility Comment:
A stem focus issue exists in that the stem does not indicate the results of the Screening Review that was conducted. Based on Procedure 108.8, Temporary Modification Control:
", If the candidate believes that the Screening Review indicated that a full 50.59 evaluation was not required, then Answer "A" is correct since the only additional review required would be the second review when the component is a "Critical Component or System" in accordance with step 8.2.5. Critical Components or systems are listed in Procedure 108.9, Equipment Alignment and Verification, and include the Control Rod Drive System.
"* If the candidate believes that the Screening Reviews were completed in the stem and that an additional review is required, this indicates that the reviews must have required a full 50.59 evaluation. Under these conditions, answer "C" is correct because a PORC review is required in accordance with step 8.2.6 of the same procedure when a full 50.59 evaluation is required.
Recommended Resolution:
Accept both answers "A" and "C" as correct.
91 NRC Initial Licensing Exam Oyster Creek 2002