ML072910020
| ML072910020 | |
| Person / Time | |
|---|---|
| Site: | Millstone  (DPR-065) |
| Issue date: | 10/04/2007 |
| From: | Gerald Bichof Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 06-0841B | |
| Download: ML072910020 (8) | |
Text
Dominion Nudear Connecticut, Inc.
IF Dm i nion 5000 Dominion Boulevard, Glen Allen, Virginia 23060 Dominiolif Web Address: www.dom.com October 4, 2007 U.S. Nuclear Regulatory Commission Serial No.
06-0841B Attention: Document Control Desk MPS ILic/WDB R1 One White Flint North Docket No.
50-336 11555 Rockville Pike License No.
DPR-65 Rockville, Maryland 20852-2738 DOMINION NUCLEAR CONNECTICUT. INC.
MILLSTONE POWER STATION UNIT 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING INSTRUMENTATION TECHNICAL SPECIFICATION CHANGES Dominion Nuclear Connecticut, Inc. (DNC) submitted a proposed license amendment to modify the Technical Specification Action and Surveillance Requirements for instrumentation identified in Millstone Power Station Unit 2 Technical Specifications 3.3.1 and 3.3.2 on November 8, 2006 (Serial No. 06-0841).
DNC responded to a request for additional information (RAI) regarding this proposed change on May 4, 2007 (Serial No. 06-0841A).
During a conference call on August 21, 2007, the NRC requested additional information. The response to this RAI is provided in Attachment 1 to this letter. to the letter contains a calculation requested in the conference call.
The signatures of the individuals who prepared, reviewed, and approved the calculation have been removed for this docketed submittal. Attachment 3 provides revised marked up Technical Specification pages as discussed in the conference call.
The additional information provided in this letter does not affect the conclusions of the significant hazards consideration discussion in DNC's original submittal dated November 8, 2006.
In accordance with 10 CFR 50.91(b), a copy of this response is being provided to the State of Connecticut.
Serial No. 06-0841 B Docket No. 50-336 Response to Request for Additional Information Page 2 of 3 Should you have any questions about the information provided or require additional information, please contact Ms. Margaret A. Earle at (804) 273-2768.
Sincerely, Gerald T. Bischof Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO
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The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering, of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
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4 Serial No. 06-0841B Docket No. 50-336 Response to Request for Additional Information Page 3 of 3 Attachments:
- 1. Response to Request for Additional Information
- 2. Calculation ZPMDrift-0426012, Rev 0, dated 9/14/07, "Zero Power Mode Drift Analysis in Support of LBDCR 06-MP2-036"
- 3. Marked up Technical Specification Pages
Enclosure:
CD Containing MP2 Instrumentation Schematic Diagrams Document Components:
001 Schematic Diagram 25203-39047-Sh 1.pdf; 1,869,277 bytes; publicly available 002 Schematic Diagram 25203-39047-Sh 2.pdf; 2,036,912 bytes; publicly available 003 Schematic Diagram 25203-39047-Sh 10.pdf; 1,989,574 bytes; publicly available 004 Schematic Diagram 25203-39069-Sh 18.pdf; 1,175,779 bytes; publicly available 005 Schematic Diagram 25203-39069-Sh 19.pdf; 1,322,495 bytes; publicly available 006 Schematic Diagram 25203-39069-Sh 23C.pdf; 250,303 bytes; publicly available 007 Schematic Diagram 25203-39256-Sh 32.pdf; 201,156 bytes; publicly available 008 Schematic Diagram 25203-39256-Sh 58.pdf; 725,435 bytes; publicly available Commitments made in this letter: None.
cc:
U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. D. Hughey NRC Project Manager Millstone Units 2 and 3 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop O-8B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring & Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127
Serial No. 06-0841B Docket No. 50-336 ATTACHMENT I INSTRUMENTATION TECHNICAL SPECIFICATION CHANGES RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2
t Serial No. 06-0841B Docket No. 50-336 Response to Request for Additional Information Page 1 of 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION During a conference call on August 21, 2007, the NRC requested additional information from Dominion Nuclear Connecticut, Inc. (DNC) regarding a proposed change to the Millstone Power Station Unit 2 (MPS2) instrumentation technical specifications.
The requested information is necessary in order for the NRC staff to complete its review.
The information requested is provided below.
NRC Question No. 1.
The License Amendment Request (LAR), dated November 8, 2006, proposes to change the first sentence in Millstone 2 Technical Specification (TS), Surveillance Requirements (SR) 4.3.1.1.2 for Reactor Protective (RPS) Instrumentation and SR 4.3.2.1.2 for Engineered Safety Feature Actuation System (ESFAS) Instrumentation from "The logic for the bypass shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation" to "The bypass function and automatic bypass removal function shall be demonstrated OPERABLE during a CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup."
The Millstone 2, TS Table 4.3-1, Reactor Protective Instrumentation Surveillance Requirements, and TS Table 4.3-2 Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements, specify monthly CHANNEL FUNCTIONAL TEST for each Functional Unit. Thus, the current wordings of SR 4.3.1.1.2 and SR 4.3.2.1.2 require monthly demonstration of the operability of the bypass channel logic.
The proposed TS change will effectively replace the bypass channel monthly CHANNEL FUNCTIONAL TEST by CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup.
It is stated in the LAR that the bases for the similar requirements contained in NUREG 1432 were reviewed and determined to be applicable to MPS2.
NUREG 1432, SR 3.3.1.7, specifies, "Once within 92 days prior to each reactor startup" for automatic bypass removal functions but not for bypass channels. Provide justifications for extending SR for bypass channels from monthly to once within 92 days prior to each reactor startup.
It is also stated in the LAR, "The allowance to conduct this test within 92 days of startup is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation," which is referenced in NUREG 1432 and is applicable to MPS2." Provide justifications why this portion of NUREG CEN-327 is applicable to MPS2 and if necessary provide the schematic diagrams of the bypass and automatic bypass removal circuits.
Serial No. 06-0841B Docket No. 50-336 Response to Request for Additional Information Page 2 of 4 DNC Response As discussed in the conference call on August 21, 2007, the Reactor Protective System (RPS) Instrumentation and Engineered Safety Feature Actuation System (ESFAS)
Instrumentation contain both trip channel bypasses as well as operating bypasses. The trip channel bypasses will continue to be demonstrated operable through performance of the channel monthly CHANNEL FUNCTIONAL TEST required by SR 4.3.1.1.1 and SR 4.3.2.1.1. The operating bypass function and automatic bypass removal function of the operating bypasses will be demonstrated operable during a
CHANNEL FUNCTIONAL TEST once within 92 days prior to each reactor startup.
Also as discussed in the conference call, the justification for extending the surveillance test interval for the operating bypasses from monthly to once within 92 days prior to each reactor startup is provided in the response to Question 2.
Copies of the applicable schematic diagrams of the operating bypasses and the automatic bypass removal circuits are provided on the enclosed compact disc. These schematics show that the logic of the operating bypasses and the automatic bypass removal circuits are directly related to the RPS and ESFAS channels evaluated under CEN-327.
NRC Question No. 2. to the LAR letter dated May 4, 2007, refers to the Calculation PA79-219-00767GE, Rev 01.
This calculation does not provide the necessary information to conclude that the proposed changes to SR 4.3.1.1.2 and SR 4.3.2.1.2 will not have any adverse effect on plant safety, specifically the drift evaluation related to increasing the operability check for bypass channels from monthly to 92 days prior to each reactor startup. The NRC letter dated November 6, 1989, "NRC Evaluation of CEOG Topical Report CEN-327, "RPS/ESFS Extended Test Interval Evaluation" states, "The licensees must confirm that they have reviewed instrument drift information for each instrument channel involved and have determined that drift occurring in that channel over the period of extended STI will not cause the setpoint value to exceed the allowable value as calculated for that channel by their setpoint methodology.
Each licensee should have onsite records of the as-found and as-left values showing actual calculations and supporting data for planned future staff audits. The records should consist of monthly data over a period of 2 to 3 years with the current plant-specific setpoint methodology used to derive the safety margins." Provide the drift evaluation to justify changing the CHANNEL FUNCTIONAL-TEST frequency for the bypass channels from monthly to 92 days prior to each reactor startup.
Serial No. 06-0841 B Docket No. 50-336 Response to Request for Additional Information Page 3 of 4 DNC Response DNC Calculation ZPMDrift-0426012, Rev 0, dated 9/14/07, "Zero Power Mode Drift Analysis in Support of LBDCR 06-MP2-036," is provided as Attachment 2.
This calculation confirms that the wide range neutron flux instrument rack drift occurring over the period of extended STI will not cause the Zero Power Mode (ZPM) operating bypass channel setpoint value to exceed the allowable value. Accordingly, the wide range neutron flux ZPM bypass bistable trip upper limit will be set at less than or equal to 7.413E-05%.
This calculation concludes that the change of the CHANNEL FUNCTIONAL TEST frequency for the ZPM operating bypass channels from monthly to 92 days prior to each reactor startup is acceptable. Onsite records of the as-found and as-left values showing actual calculations and supporting data are available for staff audits. The records consist of monthly data over the period from January 2005 through July 2007.
The operating bypasses and their setpoints are not modeled in the plant safety analysis and are not considered analytical limits. The instrument uncertainty is used as a basis for establishing the setpoint value such that the setpoint value will remain below the allowable value over the period of the surveillance interval.
NRC Question No. 3.
Does the Millstone Unit 2 control room currently have logarithmic displays or are they being added as a modification associated with this TS submittal?
DNC Response The Millstone Unit 2 control room currently has logarithmic displays.
NRC Question No. 4.
Proposed ACTION 8.a appears to be internally inconsistent. The word "channel" should be "channels" as this ACTION applies to the condition of two inoperable automatic bypass removal channels.
DNC Response DNC concurs that proposed ACTION 8.a should read as follows:
Serial No. 06-0841B Docket No. 50-336 Response to Request for Additional Information Page 4 of 4
- a. disable the bypass channels within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or A revised insert reflecting this change is included in Attachment 3.
Question No. 5.
-Table 3.3-1, Table Notation (f) should be revised similar to the proposed change to Table Notation (a) to accurately reflect the parameter being measured by referring to logarithmic power instead of 'THERMAL POWER' and replacing '%RTP' with '%' where appropriate.
DNC Response DNC proposes changing Table 3.3-1, Table Notation (f) as follows:
Current (f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is >
5% of RATED THERMAL POWER.
Proposed (f) AT Power input to trip may be bypassed when logarithmic power is < 1E-04%
and the bypass shall be capable of automatic removal whenever logarithmic power is < 1 E-04%. Bypass shall be removed prior to raising logarithmic power to a value _> 1E-04%.
The parameter used for the input to the AT trip input to the operating bypass for the affected functional unit (i.e., Power Level -
High) is the associated wide range logarithmic nuclear instrumentation channel.
Accordingly, the discussion related to proposed changes 4 and 5, in Section 4.0, 'Technical Analysis,' of Attachment 1 to the November 8, 2006 submittal is also applicable to the above proposed change.
A revised marked up Technical Specification page and associated insert is included in.