IR 05000280/2009301
| ML13220B227 | |
| Person / Time | |
|---|---|
| Site: | Surry, 05523694  |
| Issue date: | 08/14/2009 |
| From: | Widmann M Division of Reactor Safety II |
| To: | Heacock D Virginia Electric & Power Co (VEPCO) |
| SECY RAS | |
| References | |
| 55-23694-SP, ASLBP 13-925-01-SP-BD01, RAS 24803 | |
| Download: ML13220B227 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION CCS-120
REGION II
SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931 August 14, 2009 United States Nuclear Regulatory Commission Official Hearing Exhibit Mr. David Charlissa C. Smith In the Matter of:
President and Chief Nuclear Officer (Denial of Senior Reactor Operator License)
Virginia Electric and Power Company ASLBP #: 13-925-01-SP-BD01 Docket #: 05523694 Innsbrook Technical Center Exhibit #: CCS-120-00-BD01 Identified: 7/17/2013 5000 Dominion Boulevard Admitted: 7/17/2013 Withdrawn:
SUBJECT:
SURRY POWER STATION - NRC OPERATOR LICENSE EXAMINATION REPORT 05000280/2009301 AND 05000281/2009301
Dear Mr. Heacock:
During the period July 20 - 23, 2009, the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Surry Power Station. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests with those members of your staff identified in the enclosed report.
The written examination was administered by your staff on July 29, 2009.
All applicants passed both the operating test and written examination. There were two post-examination comments concerning the written examination. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm.adams.html (the Public Electronic Reading Room).
If you have any questions concerning this letter, please contact me at (404) 562-4550.
Sincerely,
/FJE RA for/
Malcolm T. Widmann, Chief Operations Branch Division of Reactor Safety Docket Nos.: 50-280 and 50-281 License Nos.: DPR-32 and DPR-37 cc w/Encl.: (See page 2)
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolutions 3. Simulator Fidelity Report
_________________________ X SUNSI REVIEW COMPLETE FJE OFFICE RII:DRS RII:DRS RII:DRS SIGNATURE MAB /RA/ FJE /RA for/
NAME MBates MWidmann DATE 08/14/2009 08/14/2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
VEPCO 2
REGION II==
Docket No.: 50-280, 50-281 License No.: DPR-32, DPR-37 Report No.: 05000280/2009301, 05000281/2009301 Licensee: Virginia Electric and Power Company Facility: Surry Power Station Location: 5850 Hog Island Rd.
Surry, VA 23883 Dates: Operating Test - July 20 - July 23, 2009 Written Examination - July 29, 2009 Examiners: M. Bates, Chief Examiner, Senior Operations Engineer F. Ehrhardt, Senior Operations Engineer Approved by: Malcolm T. Widmann, Chief Operations Branch Division of Reactor Safety Enclosure 1
SUMMARY OF FINDINGS ER 05000280/2009301, 05000281/2009301, 07/20-23/2009 & 07/29/2009; Surry Power Station; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 9, Supplement 1, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
The operating tests were developed by Surry Power Station staff and the written examination was developed by members of the NRC.
The NRC administered the operating tests during the period of July 20 - 23, 2009. Members of the Surry Power Station training staff administered the written examination on July 29, 2009. All Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. Five applicants were issued licenses commensurate with the level of examination administered.
There were two post-examination comments.
No findings of significance were identified.
Enclosure 1
REPORT DETAILS 4. OTHER ACTIVITIES 4OA5 Operator Licensing Initial Examinations a. Inspection Scope The operating tests were developed by members of the Surry Power Station staff and the written examination was developed by members of the NRC. All examination material was developed in accordance with the guidelines contained in Revision 9, Supplement 1, of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." The NRC examination team reviewed the proposed examination.
Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021 and incorporated into the final version of the examination materials.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.
The NRC examiners evaluated five Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. The examiners administered the operating tests during the period July 20 - 23, 2009. Members of the Surry Power Station training staff administered the written examination on July 29, 2009. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Surry Power Station, met the requirements specified in 10 CFR Part 55, Operators Licenses.
b. Findings No findings of significance were identified. The NRC determined, using NUREG-1021, that the licensees examination submittal was within the range of acceptability expected for a proposed examination.
Five applicants passed both the operating test and written examination and were issued licenses.
Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted two post-examination comments concerning the written examination. A copy of the final written examination and answer key, with all changes incorporated, and the licensees post-examination comments may be accessed in the ADAMS system (ADAMS Accession Number ML092260186, ML092260242 and ML092260259).
Enclosure 1
4OA6 Meetings Exit Meeting Summary On July 23, 2009 the NRC examination team discussed generic issues associated with the operating test with Mr. B. L. Stanley, Director Nuclear Safety & Licensing, and members of the Surry Power Station staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
KEY POINTS OF CONTACT Licensee personnel L. Baker, Superintendent, Shift Operations A. Barbee, Manager, Training J. Dillich, Assistant Plant Manager J. Ford, Nuclear Training K. Grover, Manager, Operations S. Irwin, Nuclear Training P. Kershner, Station Licensing W. Marshall, Nuclear Training B. Stanley, Director Nuclear Safety & Licensing D. Wilson, Supervisor, Nuclear Training NRC personnel J. Nadel, Resident Inspector Enclosure 1
FACILITY COMMENTS AND NRC RESOLUTIONS A complete text of the licensees comments can be found in ADAMS under accession number ML092260259.
SRO QUESTION 86 Licensee Comment:
The licensee contends that actions of Technical Specification (TS) 3.1.A.4 are not required to be taken and actions of TS 3.1.A.5 are required to be taken, thereby making B the only correct answer.
The licensee contends that TS 3.1.A.4 is intended to apply only to reactor coolant loops, and that reactor coolant pump status does not impact TS 3.1.A.5. The licensee contends that the position of reactor coolant loop stop valves primarily determines whether or not the loop is in service and that the presence (or absence) of forced coolant flow has no impact on determining whether or not a reactor coolant loop is in service.
The licensee supports their conclusion by stating that a separate TS (TS 3.17) exists to address the impact of reactor coolant loop stop valves. TS 3.17 lists several items which are required to be met in order to return a loop to service, none of which require a reactor coolant pump to be operating.
The licensee also supports their position with Section 4.2 of the Updated Final Safety Analysis Report which lists the following criteria which are required to return a reactor coolant loop to service:
2. Prevent opening of a cold-leg stop valve unless:
a. The hot-leg stop valve has been opened a specified time.
b. The loop bypass valve has been opened a specified time.
c. Flow has existed through the relief line for a specified time.
d. The cold-leg temperature is within 20oF of the highest cold-leg temperature in other loops and the hot-leg temperature is within 20oF of the highest hot-leg temperature in the other loops.
The licensee states that these conditions required for returning a loop to service do not include any items that suggest that forced flow through the loop is required to consider the loop to be in service.
Lastly, the licensee contends that the question statement is worded such that it asks for whether or not action statements of the LCO are required to be performed, and TS 3.1.A.4 does not contain any actions to be performed.
NRC Discussion:
The NRC agrees with the licensees assessment of the question. The NRC agrees that actions of Technical Specification (TS) 3.1.A.4 are not required to be taken and actions of TS 3.1.A.5 are required to be taken.
Enclosure 2
Based on supporting information provided by the licensee, the determination of a reactor coolant loop being in service is not dependent on forced flow existing within that loop. If the loop stop valves are fully open, the loop would be considered in service and the conditions of LCO 3.1.A.4 would be met, thus not requiring any action statements to be performed.
Lastly, the NRC also agrees that TS 3.1.A.4 does not contain any action statements to be performed, regardless of whether the reactor coolant loop is considered to be in service.
NRC Resolution:
B is the only correct answer.
SRO QUESTION 93 Licensee Comment:
The licensee contends that alarm 1D-C6, PZR PWR RELIEF VV LO AIR PRESS, can annunciate on either low backup air bottle pressure (1000 psig) or low backup air system pressure downstream of the pressure regulator (80 psig). Due to backup air bottle pressure being greater than 1000 psig, the cause of the alarm would have to be low backup air pressure downstream of the regulator, which would render the PORV inoperable and the conditions of TS 3.1.A.6 would not be met. The TS actions of restoring the PORV backup air supply within 14 days OR be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> would be required. Therefore, C is the only correct answer.
The licensee contends that air bottle pressure was provided as 1050 psig in the stem of the question, which is above the 1000 psig setpoint to cause 1D-C6 to annunciate. The licensee further states that the alarm would have to be caused by a low backup air system pressure downstream of the regulator since the potential cause of low air bottle pressure can be ruled out due to the given air bottle pressure of 1050 psig. As specified in 1D-C6 and in TS 3.1.A.6 Bases, the PORV is required to be declared inoperable if backup air pressure downstream of the regulator is less than 80 psig.
NRC Discussion:
The NRC agrees with the licensees contention that the conditions of TS 3.1.A.6 were not met due to low backup air pressure downstream of the regulator and that the required actions of restoring the PORV backup air supply within 14 days OR be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> would be required.
The NRC agrees that the alarm can annunciate due to low backup air bottle pressure less than or equal to 1000 psig, low backup air system pressure of less than or equal to 80 psig downstream of the regulator, or an instrument failure. The low backup air bottle pressure can be ruled out as the cause of the alarm because air bottle pressure of 1050 psig was provided in the question stem. No conditions in the stem were provided that would indicate that an instrument failure existed for the pressure switch downstream of the regulator, which could cause the 1D-C6 alarm to be lit. Therefore, the cause of 1D-C6 being lit would be a low backup air pressure downstream of the regulator.
Enclosure 2
1D-C6 leads the operator to Step 4 in the presence of a low backup air system pressure downstream of the regulator. Step 4 directs the operator to declare pressurizer PORVs inoperable and start the 14 day clock in accordance with TS 3.1.A.6.f.
NRC Resolution:
C is the only correct answer.
Enclosure 2
SIMULATION FACILITY REPORT Facility Licensee: Surry Power Station Facility Docket Nos.: 05000280/05000281 Operating Tests Administered on: July 20 - July 23, 2009 This form is to be used only to report observations. These observations do not constitute audit or inspection findings and, without further verification and review in accordance with Inspection Procedure 71111.11 are not indicative of noncompliance with 10 CFR 55.46. No licensee action is required in response to these observations.
No simulator fidelity or configuration items were identified.
Enclosure 3