RS-05-008, Additional Information Supporting Main Steam Line Flow - High Instrumentation Amendment Request
| ML050390330 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/21/2005 |
| From: | Simpson P Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-05-008 | |
| Download: ML050390330 (12) | |
Text
Exelkn.
Exelon Generation 4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com Nuclear RS-05-008 January 21, 2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265
Subject:
Additional Information Supporting Main Steam Line Flow - High Instrumentation Amendment Request
References:
(1)
Letter from P. R. Simpson (Exelon Generation Company) to U. S. NRC,
'Technical Specifications Changes Related to Main Steam Line Flow-High Isolation Instrumentation," dated June 10, 2004 (2)
Letter from U. S. NRC to C. M. Crane (Exelon Generation Company),
"Quad Cities Nuclear Power Station, Units I and 2 - Request for Additional Information Regarding Main Steam Line High Flow Instrumentation Amendment Request," dated November 29, 2004 In Reference 1, Exelon Generation Company, LLC (EGC) requested a change to the Technical Specifications (TS) of Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed TS change revises the Main Steam Line (MSL) Flow-High surveillance requirements (SRs) and allowable value (AV) specified in TS Table 3.3.6.1-1, "Primary Containment Isolation Instrumentation," and Table 3.3.7.1-1, "Control Room Emergency Ventilation (CREV) System Isolation Instrumentation," to reflect planned design changes in instrumentation.
In Reference 2, the NRC requested additional information to complete the review of this license amendment. This request for additional information (RAI) pertained to EGC's use of Instrumentation, Systems and Automation Society (ISA) 67.04 Part 2 Method 3 to establish the setpoint AVs. Attachment 1 of this letter provides EGC's response to the RAI.
As agreed upon during a teleconference between the NRC (Larry Rossbach, Evangelos Marinos, et al.) and EGC (Patrick Simpson, David Gullott, et al.) on January 10, 2005, the response to Question A involves adding a Note to TS Tables 3.3.6.1-1 and 3.3.7.1-1. The details of this note are discussed in Attachment 1. Attachment 2 of this letter provides a revised mark up of Technical Specification pages 3.3.6.1-5 and 3.3.7.1-4 submitted in Reference 1 with this noted added. Attachment 3 of this letter provides the typed version of these same two TS pages.
January 21, 2005 U. S. Nuclear Regulatory Commission Page 2 EGC has reviewed the information supporting a finding of no significant hazards consideration that was previously provided to the NRC in Attachment 1 of Reference 1. The addition of the note to TS Tables 3.3.6.1-1 and 3.3.7.1-1 does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.
Should you have any questions concerning this letter, please contact Mr. David Gullott at (630) 657-2819.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21st day of January 2005.
Respectfully, Patrick R. Simpson Manager - Licensing : Response to Request for Additional Information : Marked Up Technical Specifications Pages : Retyped Technical Specifications Pages
ATTACHMENT I Response to Request for Additional Information NRC QUESTION A The staff has determined that setpoint Allowable Values (AV) established by means of Instrumentation, Systems and Automation Society (ISA) 67.04 Part 2 Method 3 do not provide adequate assurance that a plant will operate in accordance with the assumptions upon which the plant safety analyses have been based. These concerns have been described in various public meetings. The presentation used in public meetings in June and July, 2004, to describe the staff concerns is available on the public website under ADAMS accession number ML041810346'.
The staff is currently formulating generic action on this subject. It is presently clear, however, that the staff will not be able to accept any requested Technical Specification (TS) changes that are based upon the use of Method 3, unless the method is modified to alleviate the staff concerns. In particular, each setpoint limit in the TSs must ensure at least 95 percent probability with at least 95 percent confidence that the associated action will be initiated with the process variable no less conservative than the initiation value assumed in the plant safety analyses. In addition, the operability of each instrument channel addressed in the setpoint-related TSs must be ensured by the TSs. That is, conformance to the TS must provide adequate assurance that the plant will operate in accordance with the safety analyses. Reliance on settings or practices outside the TS and not mandated by them is not adequate.
The staff has determined that AV computed in accordance with ISA Method 1 or 2 do provide adequate assurance that the safety analysis limits will not be exceeded. The staff has also determined that an entirely different approach, based upon the performance of an instrument channel rather than directly upon the measured trip setting, can also provide the required assurance. This alternative approach, designated Performance-Based TSs, sets limits on acceptable nominal setpoints and upon the observed deviation in the measured setpoint from the end of one test to the beginning of the next. This approach has been accepted for use at R. E. Ginna Nuclear Power Plant (GNPP), and is discussed in a safety evaluation (SE) available via ADAMS as accession number ML041180293. The referenced SE is specific to GNPP, and is cited here only as a general example for other plants. It is up to the licensee to modify the approach as necessary to meet the indicated objectives for the particular plant(s) in question. In addition, licensees are welcome to propose alternative approaches that provide the indicated confidence, but such alternative approaches must be presented in detail and must be shown explicitly to provide adequate assurance that the safety analysis assumptions will not be violated.
The Nuclear Energy Institute (NEI) has indicated an intent to submit a white paper concerning this matter for U. S. Nuclear Regulatory Commission consideration. Receipt of
'Go to www.nrc.gov, click on 'Electronic Reading Room,' then Documents in ADAMS," then 'Web-Based Access.' then 'Advanced Search," and enter the Accession number into the Accession Number box near the top of the page. Click on the Search" button near the bottom of the page. Click on the icon under "Image File' on the search results page. NOTE: You will need Adobe Acrobat Reader to open this file. NOTE: Public access to ADAMS has been temporarily suspended so that security review of publicly available documents may be performed and potentially sensitive information removed. Please check the NRC Web site for updates on the resumption of ADAMS access
ATTACHMENT 1 Response to Request for Additional Information that white paper is anticipated in late November 2004. Licensees may choose to endorse whatever approach and justification is described in that white paper, or to act independently of the NEI. If the NEI approach is found to be acceptable to the staff, it will be necessary for each licensee who chooses to use it to affirm that the salient conditions, practices, etc. described in it are applicable to their individual situations.
Please indicate how you wish to proceed in regard to the Setpoint-Related TS changes addressed in your request. Following are some examples of acceptable actions:
- 1. Demonstrate that the approach that you have used to develop the proposed limits provides adequate assurance that the plant will operate in accordance with the safety analyses. Show that operability is ensured in the TSs.
- 2. Suspend consideration of setpoint-related aspects of your request pending generic resolution of the staff concern.
- 3. Revise your request to incorporate Method 1, Method 2, or Performance-Based TSs.
- 4. Revise your request to incorporate some other approach that you demonstrate to provide adequate confidence that the plant will operate in accordance with the safety analyses and show that operability is ensured in the TSs.
QUESTION A RESPONSE Exelon Generation Company, LLC (EGC) plans to upgrade the Main Steam Line (MSL) Flow-High primary containment isolation system (PCIS) instrumentation at Quad Cities Nuclear Power Station (QCNPS) from pressure switches to pressure transmitter/trip units. The current instrumentation used for the MSL Flow - High Functions in Technical Specification (TS)
Sections 3.3.6.1 and 3.3.7.1 utilizes pressure switches, which are extremely sensitive to vibration, difficult to calibrate, and tend to have more drift than other types of currently available instrumentation. Since the pressure switches provide the logic actuation contacts for PCIS and Control Room Emergency Ventilation (CREV) isolation actuations, a false indication may initiate a spurious half Group 1 and CREV isolation signal. Recently, the current instrumentation has resulted in two Licensee Event Reports (LERs) based on as-found setpoints exceeding the TS allowable value (AV) (LER 2-03-005, 1-04-001).
The existing Barton pressure switches will be replaced with Rosemount Model 1153 Series B Alphaline pressure transmitters and Rosemount Model 710DU trip units. The Rosemount units have a higher reliability and, thus, will serve to mitigate spurious isolations and produce better overall performance relative to the existing instrumentation. This replacement is intended to upgrade the material condition of QCNPS.
Installation of this equipment upgrade requires the revision of the AV for the MSL Flow-High function in TS Sections 3.3.6.1 and 3.3.7.1. Reference 1 describes the EGC setpoint methodology used for determining the revised AV. The setpoint methodology utilizes Topical Reports and/or calculation methodologies that have been previously reviewed by the NRC (Reference 2). The NRC approved the use of this setpoint methodology in a Safety Evaluation (Reference 3).
ATTACHMENT I Response to Request for Additional Information The setpoint methodology determines the uncertainties associated with instruments, instrument loops, and instrument setpoints and applies these uncertainties to the determination of instrument loop accuracies, allowable values,'and calculated setpoints. This methodology is consistent with ISA-RP67.04.02-2000 (Reference 4).
This methodology involves establishing a setting tolerance for all TS setpoints. The setting tolerance is the uncertainty associated with the calibration procedure allowances used by technicians in the calibration process. The setting tolerance is a tighter band around the trip setpoint than the AV range (i.e., more conservative). Programs exist at QCNPS to ensure that instrument channels and calibrated setpoints will not be left outside of a specific setting tolerance. This practice resets the as-left trip setpoint within the calculated setting tolerance and near the trip setpoint value.
EGC encourages the NRC to work closely with the Nuclear Energy Institute while formulating any generic recommendations on this subject. Until a final resolution is reached, EGC will modify TS Tables 3.3.6.1-1 and 3.3.7.1-1 with a Note regarding as-left trip setpoints for the specified instruments. This Note states:
"...[Instrument Function] is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint."
The addition of this note to the TS is based on the teleconference between the NRC (Larry Rossbach, Evangelos Marinos, et al.) and EGC (Patrick Simpson, David Gullott, et al.) on January 10, 2005.
NRC QUESTION B The amendment request proposes to add surveillance requirements 3.3.6.1.3 and 3.3.7.1.3 for the trip units with a surveillance interval of 92 days. Explain the basis for the proposed trip unit calibrations and their frequency.
QUESTION B RESPONSE The additional surveillance test requirements are consistent with NRC-approved testing recommendations for TS actuation instrumentation. In the early 1980s, General Electric (GE) and the Boiling Water Reactor Owners' Group (BWROG) embarked on a program to evaluate testing requirements. This effort resulted in a series of GE Topical Reports (i.e.,
improvements in instrumentation allowed outage time/surveillance test intervals - AOT/STI) that provided the bases for a quarterlyltest interval. QCNPS applied for AOT/STI in Reference 5, which was subsequently approved by the NRC in a Safety Evaluation (SE) (Reference 6).
The proposed change includes an additional 92-day surveillance for the MSL Flow-High trip units. This is consistent with GE Topical Report NEDC-31677P-A, 'Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," dated July 1990 (Reference 7). Reference 7 was approved by the NRC in a SE dated June 18, 1990 (Reference 8). In addition, Section 5.7.3 of Reference 9 specifically discusses the calibration intervals for trip units and concludes a "3 month calibration interval for analog trip units can be
ATTACHMENT 1 Response to Request for Additional Information supported by current information." The NRC approved Reference 9 on July 15, 1987 (Reference 10). The proposed test interval is also consistent with other QCNPS TS trip functions that employ trip unit devices. For these reasons, the proposed surveillance requirements 3.3.6.1.3 and 3.3.7.1.3 for the MSL Flow-High trip units are acceptable.
References
- 1. NES-EIC-20.04, "Analysis of Instrument Channel Setpoint Error and Instrument Loop Accuracy"
- 2. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Request for Technical Specifications Changes for Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2, to Implement Improved Standard Technical Specifications,' dated March 3, 2000
- 3. Letter from U. S. NRC to 0. D. Kinsley (Exelon Generation Company), Issuance of Amendments," dated March 30, 2001
- 4. ISA-RP67.04.02-2000, -Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation," approved January 1, 2000
- 5. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, 'Proposed Technical Specifications Change Surveillance Test Intervals and Allowed Outage Time for Protective Instrumentation," dated December 22, 1999
- 6. Letter from U. S. NRC to 0. D. Kinsley (Exelon Generation Company), Issuance of Amendments," dated March 28, 2001
- 7. General Electric Topical Report NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation," dated July 1990
- 8. Letter from C. Rossi (U. S. NRC) to S. Ployd (BWROG), "GE Topical Report NEDC-31677P," dated June 18, 1990
- 9. General Electric Topical Report NEDC-30851 P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," dated March 1988 1 0. Letter from A. Thadani (U. S. NRC) to T. Pickens (BWROG), "GE Topical Report NEDC-30844," dated July 15, 1987
ATTACHMENT 2 Marked Up Technical Specifications Pages
Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page I of 3)
Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REOUIREMENTS VALUE
- 1. Main Steam Line Isolation
- a. Reactor Vessel Water Level-Low Low 1.2.3 2
D SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 2 -55.2 inches
- b. Main Steam Line Pressure-Low 1
2
- c. Main Steam Line Pressure-Timer 2
2 per MSL E
SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 E
SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 D
SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 2 831 psig S 0.331 seconds24.
S 2i44 -si
- d. Main Steam Line Flow-High 1,2,3
- e. Main Steam Line Tunnel Temperature-High 1,2.3 2 per trip string D
SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7 S 1981F
- 2. Primary Containment Isolation
- a. Reactor Vessel Water Level-Low 1,2,3 2
SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 2 3.8 inches
- b. Drywell Pressure-High
- c. Drywell Radiation-High 1.2.3 1.2,3 2
1 G
SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 F
SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 s 2.43 psig S 70 R/hr (continued)
(c) Function I.d is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.
Quad Cities 1 and 2
- 3. 3.6. 1-5 Amendment No. 202/-198 l
CREV System Isolation Instrumentation 3.3.7.1 Table 3.3.7.1-1 (page 1 of 1)
Emergency Ventilation (CREV) System Isolation Instrumentation Control Room APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE
- 1. Reactor Vessel Water Level-Low 1,2,3, (a) 2 C
SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6 2 3.8 inches
- 2. Drywell Pressure-High
- 3. Main Steam Line Flow-High 1.2,3 2
C B
SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6 SR 3.3.7.1.1 SR 3.3;.7.1.2.._
SR 3.3.7.1.5+
SR 3.3.7.1.6 s 2.43 psig I1 13.3. 7. 1.
1.2,3 2 per MSL
- 4. Refueling Floor Radiation-High
- 5. Reactor Building Ventilation Exhaust Radiation-High 1.2.3.
(a)
,(b) 1,2.3.
(a),(b) 2 2
B B
SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6 SR 3.3.7.1.1 SR 3.3.7.1.2 SR 3.3.7.1.4 SR 3.3.7.1.6 s 100 mR/hr s 9 mR/hr (a)
(b)
During operations with a potential for draining the reactor vessel.
During CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment.
(c) Function 3 is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.
Quad Cities 1 and 2 3.3.7. 1-4 Amendment No. 202J198 1
ATTACHMENT 3 Retyped Technical Specifications Pages
Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 3)
Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REOUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REOUIREMENTS VALUE
- 1.
Main Steam Line Isolation
- a.
Reactor Vessel Water Level-Low Low 1,2.3 2
- b.
Main Steam Line Pressure-Low
- c.
Main Steam Line Pressure-Timer I
2 2
D SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 E
SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 E
SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 2 831 psig S 0.331 seconds 2 -55.2 inches
- d.
Main Steam Line Flow-High 1.2.3 2 per MSL D
SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 D
SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7 s 248.1 psidec)
- e.
Main Steam Line Tunnel Temperature-High 1,2.3 2 per trip string s 198°F
- 2.
Primary Containment Isolation
- a.
Reactor Vessel Water Level-Low 1.2,3 2
G SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.6 SR 3.3.6.1.7 2 3.8 inches
- b.
Drywell Pressure-High
- c.
Drywell Radiation-High 1,2,3 1,2,3 2
1 G
SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.7 F
SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.6 SR 3.3.6.1.7 s 2.43 psig s 70 R/hr (continued)
(c)
Function 1.d is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.
Quad Cities 1 and 2 3.3.6. 1-5 Amendment No.
CREV System Isolation Instrumentation 3.3.7.1 Control Room Table 3.3.7.1-1 (page 1 of 1)
Emergency Ventilation (CREV) System Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED.
REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE
- 1. Reactor Vessel-Water 1.2.3.
2 C
SR 3.3.7.1.1 2 3.8 inches Level-Low (a)
SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6
- 2. Drywell Pressure-High 1,2,3 2
C SR 3.3.7.1.2 s 2.43 psig SR 3.3.7.1.4 SR 3.3.7.1.6
- 3. Main Steam Line 1,2,3 2 per MSL B
SR 3.3.7.1.1 s 248.1 psidec)
Flow-High SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.5 SR 3.3.7.1.6
- 4. Refueling Floor 1.2.3.
2 B
SR 3.3.7.1.1 s 100 mR/hr Radiation-High SR 3.3.7.1.2 (a),(b)
SR 3.3.7.1.4 SR 3.3.7.1.6
- 5.
Reactor Building 1,2.3, 2
B SR 3.3.7.1.1 s 9 mR/hr Ventilation Exhaust SR 3.3.7.1.2 Radiation-High (a),(b)
SR 3.3.7.1.4 SR 3.3.7.1.6 (a)
(b)
(c)
During operations with a potential for draining the reactor vessel.
During CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment.
Function 3 is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established setting tolerance band of the Nominal Trip Setpoint.
Quad Cities 1 and 2 3.3.7. 1-4 Amendment No.