ML110480489
| ML110480489 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/11/2011 |
| From: | Gillespie T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML110480489 (20) | |
Text
T. PRESTON GILLESPIE, Jr.
Duke Vice President RPEnergy Oconee Nuclear Station Duke Energy ONO VP / 7800 Rochester Hwy.
Seneca, SC 29672 864-873-4478 864-873-4208 fax T.Gillespie@duke-energy. com February 11,2011 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001
Subject:
Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2 and 3 Renewed Facility Operating Licenses Numbers DPR-38, -47, -55; Docket Number 50-269, 50-270 and 50-287; Response to Request for Additional Information Regarding License Amendment Request to Change Technical Specification Surveillance Requirement Frequencies to Support 24-Month Fuel Cycles License Amendment Request No. 2010-001, Supplement 1 On May 6, 2010, Duke Energy Carolinas, LLC (Duke Energy) submitted a LAR requesting Nuclear Regulatory Commission (NRC) approval for certain Oconee Nuclear Station (ONS) TS Surveillance Requirement frequencies that are specified as "18 months" by revising them to "24 months" in accordance with the guidance of Generic Letter 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle." The NRC electronically transmitted a Request for Additional Information (RAI) to Duke Energy on November 26, 2010. The Enclosure provides Duke Energy's response to the RAI and revises information submitted in the initial LAR as indicated in the enclosure.
If there are any questions regarding this submittal, please contact Boyd Shingleton of the Oconee Regulatory Compliance Group at (864) 873-4716.
I declare under penalty of perjury that the foregoing is true and correct. Executed on February 11,2011.
Sincerely, T. Preston Gillespie, Jr., Vice President Oconee Nuclear Station
Enclosure:
Duke Energy Response to NRC Request for Additional Information 4
www. duke-energy, corn
U. S. Nuclear Regulatory Commission February 11,2011 Page 2 cc w/
Enclosure:
Mr. Victor McCree, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. John Stang, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, D. C. 20555 Mr. Andy Sabisch Senior Resident Inspector Oconee Nuclear Site Ms. Susan E. Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.
Columbia, SC 29201
ENCLOSURE Duke Energy Response to NRC Request for Additional Information
February 11,2011 Page 1 Enclosure Duke Energy Response to NRC Request for Additional Information (RAI)
RAI I The licensee stated "Any necessary revisions to setpoint calculations and calibration procedures to incorporate results of the statistical analysis of the historical As Found/ As Left (AFAL) data will be completed prior to implementation" for the following SR in Attachment 6 and the first Commitment in Attachment 5 of the LAR:
SR 3.3.1.7 page 21 SR 3.3.5.4 page 25 SR 3.3.8.3 page 27 SR 3.3.10.2 page 32 SR 3.3.11.3 page 34 SR 3.3.28.2 page 38 To ensure compliance with TSTF-493, Revision 4, or RIS 2006-17, provide the necessary calculations, affected calibration settings and calibration procedures for all instrumentation where projected drift values are outside the existing design allowances.
Duke Energy Response to RAI I An instrument drift analysis was performed for the applicable functions of the six noted surveillances per the methodology included in Attachment 7 of the LAR. The results of these analyses determined drift for a bounding interval of 30 months which is represented as the extended cycle analyzed drift (ADE). The analyzed drift values were then compared to the 30-month acceptable limit as determined from the setpoint uncertainty calculation for each function.
For all occurrences where ADE exceeded the 30-month acceptable limit (i.e., projected drift exceeded design allowances), revisions to the applicable setpoint uncertainty calculation were performed to determine impacts to the total loop uncertainty and any related setpoints or decision points. In addition, a review of existing as-found calibration tolerances, channel functional tests (if applicable) and channel check limits (if applicable) were performed to identify any changes needed as a result of the increased drift. The attached evaluation flow chart (Attachment 1) illustrates this process.
The attached table (Attachment 2) provides a comparison of ADE versus the 30-month acceptable limit for all the surveillance functions in question. For those functions where ADE >
30-month acceptable limit (AL30); the referenced drift analysis, setpoint uncertainty calculation and calibration procedure are provided as requested. For the Reactor Protective System (RPS) and Engineered Safeguards Protective System (ESPS) functions, the calibration procedure for one channel is provided and is typical for all channels.
The following input is applicable to the table information and compliance with TSTF-493, Revision 4 or Regulatory Information Summary (RIS) 2006-17:
There are some functions for which a drift analysis was not applicable. Justifications for these functions are provided in the referenced section of calculation OSC-9852 as noted in the AFAL Analysis column.
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 2 The scope of TSTF-493, Revision 4 or RIS 2006-17 is limited to the RPS and ESPS functions included in SR 3.3.1.7 and SR 3.3.5.4. For these functions, the current Allowable Values (AV) specified in Technical Specifications (TSs) are based on the channel functional tests performed on a 45 day staggered basis for RPS and every 92 days for ESPS. The channel functional tests only include the cabinet hardware.
As such, instrument drift associated with these tests are unaffected by transition to 24 month cycles. The ADE specified for the RPS and ESPS functions is applicable to the process sensors calibrated on a 18 month frequency and does impact the total loop uncertainty for each function. For the RPS and ESPS functions where ADE exceeded the AL30, revision of the associated setpoint uncertainty calculation has been completed and no instances were found requiring changes to the current nominal plant setpoint or associated AV. Refer to the response to RAI 2 for additional information.
The notice of availability for TSTF-493, Revision 4 occurred after the Oconee Nuclear Station (ONS) submittal of the LAR for 24 month cycles. Given this and the fact that no AVs specified in the TSs require change as a result of transitioning to 24 month cycles, ONS has not adopted the TSTF.
The uncertainty calculation revisions for PAM SR 3.3.8.3 functions 4, 12 and 14 have not been approved as of this date but will be completed prior to implementation as stated in the LAR.
RAI 2
If the projected drift values are within the existing design allowances in question 1, provide a comparison table and/or a sample diagram to illustrate the setpoint calculation values of current and 24-month fuel cycles. The information to be provided should include but not be limited to the instrument function name, TS allowable value (AV), Limiting Trip Setpoint (LTSP), Nominal trip setpoint (NSTP), total loop uncertainties, margin, As-Found tolerance (AFT), and As-Left tolerance (ALT).
Duke Energy Response to RAI 2 The attached table (Attachment 3) includes the requested information for those functions in the six noted surveillances where the analyzed drift for 30 months was bounded by existing design allowances or ADE < AL30 as specified in Attachment 2. All the values were obtained from the noted references or as specified in the table notes.
The following input is applicable to the table information:
Only SR 3.3.1.7 (RPS) and SR 3.3.5.4 (ESPS) have AVs specified in the TSs and these AVs are based on the channel functional tests as noted in the response to RAI 1.
The ONS setpoint methodology does not require the calculation of a LTSP, therefore only the NTSP employed in the plant equipment, the allowable value as specified in TSs and the calibration as-found setting tolerance are provided.
For the RPS and ESPS functions, the margin identified is that between the AV and the as-found calibration setting tolerance. For increasing setpoints, margin between the
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 3 nominal trip setpoint (NTSP) and the analytical limit (AL) can be determined by adding the total loop uncertainty to the NTSP (or subtracting for decreasing setpoints) and comparing that value to the AL.
For the non RPS and ESPS surveillances, no AVs are specified in the TSs and no ALs are applicable. Therefore, applicable process limits and setpoints (decision points) are provided. Note that some functions have multiple decision points only one of which is provided for clarity.
At ONS, as-found calibration tolerances are conservatively set equal to as-left calibration tolerances. Therefore the setting tolerances specified are for both as-found and as-left conditions.
By procedure, any instrumentation found out of the specified setting tolerance (OOT) is re-calibrated within the required tolerance prior to return to service.
Engineering evaluations are required for all OOT conditions exceeding notification limits.
Notification limits are specified in each calibration procedure. The default notification limit is two times the specified procedure setting tolerance or as specified in the calibration procedure for other reasons (TSs, setpoint uncertainty calculation, site directives).
RAI 3
SR 3.3.1.7 - The licensee stated "the instrumentation had six failures of the TS functions that would have been detected solely by the periodic performance of the SR" in the fourth paragraph of page,20 to Attachment 6 of the LAR. The licensee also indicated those failures were repetitive, but stated, at the end of page 22, "Considering the total number of Rosemount transmitters in the various systems in all three units, the total number of failures identified is small." What's the total number of Rosemount 1154GP transmitters in all three Units and justify why extension of surveillance frequency to 24-months is valid.
Duke Energy Response to RAI 3 The total number of Rosemount 1154GP transmitters installed in all three units is 64. Of these 64 applications, 52 support an 18 month TS surveillance included in the LAR for extension to 24 months and for which a surveillance history review was performed to identify failures. The six failures identified for SR 3.3.1.7 on pages 21 and 22 of Attachment 6 of the LAR were the only failures associated with the 52 applications. As noted on page 2 of Attachment 6 of the LAR, the surveillance history review included a minimum of five 18 month cycles or approximately 65,000 transmitter operating hours per application. Therefore, the number of failures identified is considered small.
Also, there are 95 additional Rosemount 1154 series transmitters installed in all three units (Models 1154DP, 1154HP, 1154HH and 1154SH) and 60 additional Rosemount 1153 series transmitters installed in all three units. The Rosemount model 1153 and 1154 transmitters are all Class 1E nuclear service qualified devices with the same operating principle and similar construction. ONS has not identified any failure trends of these devices due to time-based
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 4 degradation that would impact extension of the surveillance interval to 24 months.
RAI 4
SR 3.3.8.3 - The licensee indicated in pages 29 and 30 of Attachment 6 of the LAR that only one good drift data was obtained for the RVLIS Head Level instrument loop.
Provide a justification regarding how the drift for this instrument loop is established with high degree of confidence.
Duke Energy Response to RAI 4 The statement in question from Page 30 of Attachment 6 of the LAR is "The data shows continued improvement with each procedural change and the final calibration, which is the only one with both procedural changes, shows the RVLIS Head Level instrument loops performing well within expected calibration limits". As the two procedural changes noted had a significant effect on calibration results, performance of an AFAL drift analysis with like data (i.e., calibrations performed with both procedure changes implemented) was not possible due to an inadequate sample size.
Based on the above, the drift of the RVLIS Head Level instrumentation for a worst case calibration frequency of 30 months was established based on the AFAL drift analysis results obtained for the RVLIS Hot Leg Level instrumentation. The RVLIS Hot Leg Indication extended cycle Analyzed Drift is a reasonable estimate of the RVLIS Vessel Head Level Indication extended cycle Analyzed Drift because both loops share the same equipment, location and function. The only difference between the loops is instrument span. The RVLIS Hot Leg Level AFAL drift error determined for the maximum 30-month calibration frequency is +/- 3.92% of span. This value has been incorporated in the RVLIS instrument uncertainty calculation for both the Hot Leg and Head level instrumentation. Previously, the RVLIS instrument uncertainty calculation included an equivalent drift allowance of +/- 2.46% of span for a nominal 18 month calibration frequency. Therefore the drift allowance included in the uncertainty analysis increased by 1.61 times the 18 month drift allowance. This increase exceeds the typical Square Root Sum of the Squares (SRSS) extrapolation of drift of 1.29 times or (30/18)1/2 for moderate time dependency and provides a high degree of confidence that the drift established in the uncertainty analysis is both reasonable and bounding.
RAI 5
SR 3.3.16 - The licensee stated, in the last paragraph of page 35 to Attachment 6 of the LAR, "This instrument loop is calibrated every 12 months online," and "There are no uncertainty calculations for the Reactor Building Purge - High Radiation instrument loops." Since this online calibration is a TS surveillance requirement, describe this online calibration and the setpoint calculation in more detail.
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 5 Duke Energy Response to RAI 5 Background from TS 3.3.16 Bases:
The RB Purge Isolation-High Radiation Function closes the RB purge valves. This action isolates the RB atmosphere from the environment to minimize releases of radioactivity in the event an accident occurs.
The radiation monitoring system measures the activity in a representative sample of air drawn in succession through a particulate sampler, an iodine sampler, and a gas sampler. The TS Limiting Condition for Operation (LCO) addresses only the gas sampler portion of this system (RIA-45).
The trip setpoint is chosen sufficiently below hazardous radiation levels to ensure that the consequences of an accident will be acceptable, provided the unit is operated within the LCO at the onset of an accident or transient and the equipment functions as designed. The mode of applicability for TS 3.3.16 is during movement of recently irradiated fuel assemblies within containment.
Calibration of RIA-45:
Two different calibration procedures are used to meet SR 3.3.16.3. One procedure is used to calibrate the instrumentation necessary to extract the sample from the process duct, route the sample through the detectors, monitor and control the sample flow rate, and return the sample to the process stream. The components consist of pumps, valves, gauges, flow meters, RM-80 processing unit including internal power supplies and tubing. The calibration tolerances employed are based on vendor specifications. The calibration of this instrumentation includes the following activities:
RM-23L Display Test Monitor In-Service Test Subassembly Test and Power Fail Power Fail Trip Setpoint Test and Adjustment Purge Test Mass Flow Meter Calibration
" Pressure Gauge Check Vacuum Gauge and Sample Vacuum Gauge Checks
" Analog Output Test/Adjustment Cleaning Sight Glass Rotometers The other procedure calibrates the radiation monitor detector. The calibration tolerances employed are based on vendor specifications coupled with operating experience. The calibration of the detector includes the following activities:
A loss of count test is performed.
A detector RD-59 Temperature Control Test is performed.
A Calibration Source (CI-36) is obtained. The pump is turned off. Inlet and outlet valves are closed. The source is installed on a Lucite source positioner and placed in a sample chamber facing the detector face.
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 6 After 30 minutes, the most current 10 minute average for AS FOUND Measured Gross counts per minute is recorded.
Error between the measured count rate and the expected count rate must be less than 10%.
Alarm/Trip Setpoints for RIA-45:
The alarm/trip setpoints for these instruments are calculated in accordance with NRC approved methods in the Offsite Dose Calculation Manual (ODCM) to assure that the alarm/trip will occur prior to exceeding applicable dose limits in Selected Licensee Commitment (SLC) 16.11.2. As described below, the alarm setpoints are established based on a fraction of the reporting limit per release path and are therefore inherently conservative.
"Alert" alarm setpoints are set to alarm if 3% of reporting limit of 10 CFR 50.73 (a) (2) (viii) (A)
(20 times EC limit) for noble gases is exceeded based upon Xe-1 33 as the major noble gas contributor. The reporting limit is divided as follows: 3% of the limit for each of the three Unit Vents, 0.5% for the Rad Waste Facility Vent (4RIA-45) and 0.5% for the Interim Radwaste Facility Vent (RIA-53). When the alert alarm setpoint is exceeded, a "Process Radiation Monitor High" statalarm and a computer alarm for high radiation is received.
"High" alarm setpoints are set to alarm if 30% of instantaneous dose rate limit of SLC 16.11.2 is exceeded based upon Xe-133 as the major noble gas contributor. The instantaneous dose limit is divided as follows: 30% of the limit for each of the three Unit Vents, 5% for the Rad Waste Facility Vent (4RIA-45) and 5% for the Interim Radwaste Facility Vent (RIA-53). When the high alarm setpoint is exceeded, Reactor Building purge valves close, the Reactor Building purge exhaust trips, and the mini-purge also trips.
RAI 6
SR 3.3.19.1 and SR 3.3.20.1 (both are functional tests) are missing from 2.A "Non-Calibration Changes" of Attachment 6 of LAR. Provide the detailed evaluation results of those two SRs.
Duke Energy Response to RAI 6 NRC initially advised Duke Energy by electronic mail on September 14, 2010 that SR 3.3.19.1 and SR 3.3.20.1 (both channel functional tests) were missing from the 2.A non-calibration changes evaluation in Attachment 6 of the LAR. Duke Energy determined and advised the NRC by electronic mail that these two SRs should have been grouped with the other EPSL SRs (SRs 3.3.17.1, 3.3.18.1, 3.3.21.1, and 3.3.23.1) and addressed jointly. The NRC requested that Duke Energy provide the missing information associated with these two SRs when responding to other RAI questions that were forthcoming.
The detailed evaluation for these two channel functional tests is provided below:
Detailed GL 91-04 Evaluation Results TS 3.3.19 Emergency Power Switching Logic (EPSL) 230 KV Switchyard Degraded Grid Voltage Protection (DGVP)
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 7 SR 3.3.19.1 Perform CHANNEL FUNCTIONAL TEST.
TS 3.3.20 Emergency Power Switching Logic (EPSL) CT-5 Degraded Grid Voltage Protection (DGVP)
SR 3.3.20.1 Perform CHANNEL FUNCTIONAL TEST.
The surveillance test interval of these SRs is being increased from once every 18 months to once every 24 months, for a maximum interval of 30 months including the 25% grace period.
All of the actuation instrumentation and logic, controls, monitoring capabilities, and protection systems, are designed to meet applicable reliability, redundancy, single failure, and qualification standards and regulations as described in the ONS Updated Final Safety Analysis Report (UFSAR). As such, these functions are designed to be highly reliable. This is acknowledged in the August 2, 1993 NRC Safety Evaluation Report relating to extension of the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3 surveillance intervals from 18 to 24 months:
"Industry reliability studies for boiling water reactors (BWRs), prepared by the BWR Owners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic systems, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis. Since the probability of a relay or contact failure is small relative to the probability of mechanical component failure, increasing the Logic System Functional Test interval represents no significant change in the overall safety system unavailability."
A review of the applicable ONS surveillance history demonstrated that the instrumentation for these functions had no failures of the Technical Specification (TS) functions that would have been detected solely by the periodic performance of the above SRs.
As such, the impact, if any, on system availability is minimal from the proposed change to a 24-month testing frequency. Based on the history of system performance, the impact of this change on safety, if any, is small.
Revision to Initial LAR Submittal During review of the NRC RAI Duke Energy discovered that an instrument drift analysis to support the extension of TS SR 3.3.8.3 Post Accident Monitoring (PAM) function 4 RCS Pressure (Wide Range) from a frequency of 18 months to 24 months had not been completed.
There are two different Reactor Coolant System (RCS) Wide Range (WR) pressure applications. One set of transmitters monitor RCS WR pressure and provide input to the Engineered Safeguards Protective System (ESPS) and another set of transmitters monitor RCS WR Pressure for PAM. Instrument drift calculation OSC-9752 was performed for the ESPS transmitters. The PAM pressure transmitters are calibrated in the RVLIS Instrument Calibration.
During performance of the drift calculations to support the LAR, calibration data from historical performances of these procedures was used to perform a drift analysis of the RCS Hot Leg Level and RV Head Level applications, however the calibration data for the PAM pressure transmitters was over looked resulting in no drift analysis being completed for the pressure function. Subsequently, during the LAR development, the drift calculation performed for the
Enclosure - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 8 ESPS RCS WR pressure transmitters was inadvertently credited for completion of the RCS WR Pressure PAM function as well.
In addition to TS 3.3.8.3 PAM function 4, the RCS WR Pressure transmitters also provide input to the RVLIS for density compensation of the level signals during certain conditions and the Subcooling Monitor which are functions 3, 5 and 17 of SR 3.3.8.3. Drift analyses were completed for the primary instrumentation inputs to functions 3, 5 and 17; RCS Hot Leg Level, Reactor Vessel Head Level and RCS THOT temperature. However, due to the omission of the drift analysis for the PAM pressure'transmitters, potential impacts on these functions due to projected drift of these instrument loops for 24 month cycles were not evaluated. In addition, the PAM pressure transmitters provide an input to the Anticipated Transient Without Scram (ATWS) circuitry; however there are no 18 month TS calibration SRs in the LAR for ATWS.
In addition to evaluating the historical drift associated with current 18-month calibrations, extension of 18 month TS SRs were also supported by completion of a failure history review as noted in the LAR. The failure review was completed as required for the RCS WR Pressure PAM instrumentation.
The drift analysis (OSC 10180) for the PAM WR pressure transmitters has been completed and of OSC 9719 (Attachment 7 to LAR) has been revised to list the calculation for this function.
Enclosure, Attachment 1 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 1 AFAL Drift Evaluation Flow Chart Section 7.5 Determine ALle based on accuracy. 18 month drift and M&TE errors from applicable setpoint uncertaintv calculation.
aluate other options. These include t, are not limited to, revise design sis, revise setpoint, revise calibration ocedure, replace instrumentation retain current 18 month rveillance requirement.
Determine AL3 0 based on accuracy, 30 month drift and M&TE errors from applicable setpoint uncertainty calculation.
Issue PIP Corrective Action to evaluate instrument uncertainties Issue PIP Corrective Action to evaluate loop setting tolerance Section 7.6.3 issue rir torrecnve Action to evaluate Channel Functional Test acceptance criteria Issue PIP Corrective Action to evaluate Channel Check acceptance criteria Evaluation of the extended cycle Analyzed Drift is complete.
Enclosure, Attachment 1 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 2 The 18 month Analyzed Drift (or just Analyzed Drift) is determined in Sections 1.0 through 7.3 of the As-Found/As-Left (AFAL) Drift Analyses. The 18 month Analyzed Drift is statistically derived from the AFAL drift data. The extended cycle Analyzed Drift (ADE) is based on the 18 month Analyzed Drift and the determination of moderate or strong time dependency in Section 7.4 of the AFAL Drift Analyses.
The Analyzed Drift and extended cycle Analyzed Drift are evaluated with respect to current and extended calibration intervals in Section 7.5 and 7.6.
Section 7.5 The 18 month acceptable limit (AL18) is used to compare the instrument uncertainties to As-Found/As-Left (AFAL) data (i.e., loop past performance). If the AFAL data exceeds the acceptable limit at an unacceptable rate for an 18 month calibration interval, the loop cannot be expected to meet its performance requirements for extended fuel cycles. To be considered for calibration interval extension, 95% of all AFAL data (except that which represents failed instruments) should be less than or equal to the 18 month acceptable limit. The 18 month acceptable limit contains those uncertainty terms that are present in the AFAL Drift data. Typically, this includes accuracy, drift and Measurement and Test Equipment (M&TE) uncertainties.
Section 7.5 is intended to cover Issue 1 of Enclosure 2 to NRC Generic Letter 91-04.
Section 7.6.1 The 30-month acceptable limit (AL30) is the same as the 18 month acceptable limit except that it is based on the 30-month drift specification. If the extended cycle Analyzed Drift (ADE) is less than or equal to the 30-month acceptable limit, then the uncertainty calculation is conservative for use with extended fuel cycles.
This is because the instrument uncertainties resulted in a more conservative error term than one based on the extended cycle Analyzed Drift. If the extended cycle Analyzed Drift (ADE) is greater than the 30-month acceptable limit, then the uncertainty calculation may not be conservative for use with extended fuel cycles. This is because the instrument uncertainties resulted in a less conservative error term than one based on the extended cycle Analyzed Drift.
Section 7.6.1 is intended to cover Issues 4 and 5 of Enclosure 2 to NRC Generic Letter 91-04.
Enclosure, Attachment 1 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 3 Section 7.6.2 If the Analyzed Drift (ADE) is greater than or equal to the instrument loop setting tolerance (CTE), then the extended cycle Analyzed Drift cannot be exceeded without the loop setting tolerance being exceeded and the loop recalibrated. This eliminates the possibility of the loop drifting outside the extended cycle Analyzed Drift without identification or correction. If the extended cycle Analyzed Drift (ADE) is less than the instrument loop setting tolerance (CTE), then the instrument loop could drift outside the extended cycle Analyzed Drift without the loop being recalibrated. To reduce the likelihood of this possibility, the loop setting tolerance should be evaluated and reduced if necessary.
Section 7.6.2 is intended to support Issue 6 of Enclosure 2 to NRC Generic Letter 91-04.
Section 7.6.3 Where applicable, a Channel Functional Test (CFT) acceptance criteria based on the extended cycle Analyzed Drift (ADE) is determined and compared to the current CFT acceptance criteria. If the extended cycle CFT acceptance criteria are significantly larger than the current CFT acceptance criteria, the current criteria may be too restrictive. If the extended cycle CFT acceptance criteria are significantly smaller than the current CFT acceptance criteria, the current criteria may be too permissive. The determination of the appropriate extended cycle CFT acceptance criteria will be made by the engineer responsible for that instrument loop.
Section 7.6.3 is intended to support Issue 6 of Enclosure 2 to NRC Generic Letter 91-04.
Section 7.6.4 Where applicable, a Channel Check acceptance criteria based on the extended cycle Analyzed Drift (ADE) is determined and compared to the current Channel Check acceptance criteria. If the extended cycle Channel Check acceptance criteria are significantly larger than the current Channel Check acceptance criteria, the current criteria may be too restrictive for extended fuel cycles. If the extended cycle Channel Check acceptance criteria are significantly smaller than the current Channel Check acceptance criteria, the current criteria may be too permissive. The determination of the appropriate Channel Check acceptance criteria will be made by the engineer responsible for that instrument loop.
Section 7.6.3 is intended to support Issue 6 of Enclosure 2 to NRC Generic Letter 91-04.
Enclosure, Attachment 2 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 1 SR 3.3.1.7 Table 3.3.1-1 Reactor Protective System Instrumentation Uncertainty Calc AFAL Analysis AL3o ADE Cal Procedures OSC-9852,
- 1)
- a. Nuclear Overpower - High Setpoint
- b. Nuclear Overpower -Low Setpoint
- 2)
RCS High Outlet Temperature
- 3)
RCS High Pressure
- 4)
RCS Low Pressure
- 5)
RCS Variable Low Pressure
- 6)
Reactor Building High Pressure
- 7)
Reactor Coolant Pump to Power
- 8)
Nuclear Overpower Flux/Flow Imbalance
- 9)
Main Turbine Trip (Hydraulic Fluid Pressure) not applicable OSC-9852, Section 6.2 not applicable OSC-9852, Section 6.3 not applicable OSC-9771
+ 9.6 psi
+/- 11.1 psi random; +/- 2.6 psi bias OSC-4048 IP/OAI03051001 M, N, 0, P OSC-9771
+ 9.6 psi
+/- 11.1 psi random; +/- 2.6 psi bias OSC-4048 IP/O/A/0305/001 M, N, 0, P OSC-9771
+ 9.6 psi
+/- 11.1 psi random; +/- 2.6 psi bias OSC-4048 IP/O/A/0305/001 M, N, 0, P OSC-9819
+ 0.24 psi
+ 0.27 psi OSC-3446 IPIO/A/03051005 A, B, C, D OSC-9852, Section 6.4 not applicable OSC-3416, OSC-9793
+/- 0.93% span
+/- 0.89% span random; -0.07% span bias OSC-8857 IP/1,2,3/A/0305/001 I, J, K, L IP/O/A/0305/009, 010, 011, OSC-9792
+/- 17.0 psi
+ 27.5 psi OSC-3395 012 SIP/O/A/0305/009, 010, 011, OSC-9792
+/- 1.7 psi
+ 6.5 psi OSC-3395 012 10) 11)
Loss of Main Feedwater Pumps (Hydraulic Oil Pressure)
Shutdown Bypass RCS High Pressure OSC-9771
+/- 9.6 psi
+/- 11.1 psi random; +/- 2.6 psi bias OSC-4048 IP/O/A/0305/001 M, N, 0, P SR 3.3.5.4 Table 3.3.5-1 Engineered Safeguards Protective System Analog Instrumentation
- 1)
Reactor Coolant System Pressure - Low OSC-9752
+/- 0.87% span
+/- 0.54% span
- 2)
Reactor Coolant System Pressure - Low Low OSC-9752
+/- 0.87% span
+/- 0.54% span
- 3)
Reactor Building (RB) Pressure - High OSC-9720
+/- 1.74% span
+/- 0.54% span
- 4)
Reactor Building Pressure - High High OSC-9809
+/- 0.267 psi j
0.291 psi random; +/- 0.027 psi bias OSC-3446 IP/O/A10310/003D, 4D, 5D
Enclosure, Attachment 2 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 2 SR 3.3.8.3 Table 3.3.8-1 Post Accident Monitorina Instrumentation Uncertainty Calc AFAL Analvsis ALin ADE Cal Procedures AIDE OSC-9852, Section 6.5 1) 2)
3)
Wide Range Neutron Flux RCS Hot Leg Temperature RCS Hot Leg Level not aoolicable
- 4)
RCS Pressure (Wide Range)
- 5)
Reactor Vessel Head Level
- 6)
Containment Sump Water Level (Wide Range)
- 7)
Containment Pressure (Wide Range)
- 8)
Containment Isolation Valve Position
- 9)
Containment Area Radiation (High Range)
- 10)
Not Used
- 11)
Pressurizer Level
- 12)
Steam Generator Water Level
- 13)
Steam Generator Pressure
- 14)
Borated Water Storage Tank Level
- 15)
Upper Surge Tank Level
- 16)
Core Exit Temperature
- 17)
Subcooling Monitor (degrees subcooled)
- 18)
- 19)
LPI System Flow
- 20)
Not used
- 21)
Emergency Feedwater Flow
- 22)
Low Pressure Service Water Flow to LPI Coolers OSC-9791
+/- 1.01% span
+/- 0.49% span OSC-9825
+/- 2.74% span t 3.92% span OSC-4310 IP/1,2,3/A/0200/042 OSC-3862",
OSC-10180
+/- 47.1 psi t 51.2 psi OSC-5123*
IP/1,2,3/A10200/042 OSC-9852, Section 6.6, OSC-9825
+/- 2.74% span
+/- 3.92% span OSC-4310 IP/1,2,3/A/0200/042 OSC-9852, Section 6.7 not applicable NA (12 Month SR Frequency) not applicable OSC-9852, Section 6.8 not applicable
+/-24.3 % of OSC-9804 reading
+/- 15.03% of reading Not Used Not Used
+/- 8.3 inches (The System Engineer considered OSC-9776
+/- 8.2 inches this difference negligible.)
OSC-9781
+/- 4.4 inches
+/- 6.9 inches OSC-4478*
IP/O/A/0275/019 A, B OSC-9777
+/- 15.7 psig
+/- 9.0 psig OSC-9754
+/- 13.6 inches
+/- 18.1 inches OSC-3189*
IP/O/A/0203/001 A OSC-9741
+/- 1.98% span
+/- 2.23% span OSC-2248 IP/O/A/0275/010 L OSC-9746
+/- 0.33% span
+/- 0.26% span random; +/- 0.03% span bias See Items 2 & 4.
OSC-9732
+/- 1.77% span**
+/- 1.74% span (at 187.5 gpm)
OSC-9733
+/- 1.28% span**
+/- 1.59% span (at 187.5 gpm)
OSC-2533 IP/O/A/0202/001 D
+ 3.31%/-3.50%
OSC-9841 span**
+/- 4.05% span (at 400 gpm)
OSC-3566 IP/O/A/0203/001 C Not Used Not Used OSC-9786
+/- 3.01% span**
+/- 2.70% span (at 400 gpm)
NA (12 Month SR Freouencv) not aoplicable
- Revision of these calculations have not been approved but will be completed prior to implementation as is stated in the LAR.
The 30 Month Drift Acceptable Limit (AL) and extended cycle Analyzed Drift (ADE) shown are from the most conservative flow calibration point.
Enclosure, Attachment 2 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 3 Uncertainty Calc AFAL Analysis AL30 ADE Cal Procedures SR 3.3.10.2 Wide Range Neutron Flux OSC-9852, Section not applicable 6.5 SR 3.3.11.3 Automatic Feedwater Isolation OSC-9777 11.2 psig 11.1 psig System (AFIS) Instrumentation Low Pressure Service Water SR 3.3.28.2 (LPSW) Standby Pump Auto-Start OSC-9823
+/- 4.1 psig
+/- 7.0 psig OSC-7693 IP/1-2,3/A/0250/001 B Circuitry
Enclosure, Attachment 3 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 1 SR 3.3.1.7 Table 3.3.1-1 Reactor Protective System Instrumentation Analytical Limit(1)
Total Loop Uncertainty(
2)
Allowable ValueP)
Nominal Trip Margini')
SetpointPl)
Setting Tolerance(
6.7)
References
- 1)
- a. Nuclear Overpower - High Setpoint
- b. Nuclear Overpower - Low Setpoint
= 112% FP
+/- 5.0% FP 5 105.5% RTP(10) 0.1875%
= 104.75% FP
+/- 0.5625% FP IP/1,2,3/1A0305/003 A, B, C, D; OSC-7237 2) 3)
4) 5)
6) 7)
8) 9)
10) 11)
RCS High Outlet Temperature RCS High Pressure RCS Low Pressure RCS Variable Low Pressure Reactor Building High Pressure Reactor Coolant Pump to Power Nuclear Overpower Flux/Flow Imbalance Main Turbine Trip (Hydraulic Fluid Pressure)
Loss of Main Feedwater Pumps (Hydraulic Oil Pressure)
Shutdown Bypass RCS High Pressure T/S AV is the only requirement.
No uncertainties applied.
= 4% FP
+/- 0.5% FP CP/i,2,3/A/0305/003 A, B, I
____________C, D; OSC-7237 IP/1,2,3/A/0305/003 A, B,
-620 *F
+ 1.08 *F 5 618 *F 0.3 *F
= 617 *F
+ 0.7 *F C, D; OSC-4048, OSC-2729 See response to RAI, Question 1.
See response to RAI, Question 1.
See response to RAI, Question 1.
See response to RAI, Question 1.
R 2% RTP with <2 0.125%
= 1.5% FP with
- 2 IP/A/1,2,3/0305/003 A, B, RCP's either on or off.
+5.15% FP
+0.3759% FP CP/
pumps operating RTP pumps operating C, D; OSC-7237 See response to RAI, Question 1.
See response to RAI, Question 1.
See response to RAI, Question 1.
See response to RAI, Question 1.
SR 3.3.5.4 Table 3.3.5-1 Engineered Safeguards Protective System Analog Analytical Limit'1)
Total Loop Allowable Value)
Margin)')
Nominal Trip Setting References Instrumentation Uncertainty(2)
WSetpoinW)
Tolerance(6.7)
- 1)
Reactor Coolant System Pressure - Low=
1400 psig
+ 198.1 psi /- 120.4 psi
> 1590 psig 2.5 psi
= 1600 psig
+/- 7.5 psi IP/O/A/0310/014A, B, C; 1asOSC-2759
- 2)
Reactor Coolant System Pressure - Low 200
+ 2066 psi / 1204 psi 500 425 psi 550 psig
+/- 75 IP/O/A/0310/014A, B, C; Low
=
psg
. p psig s.psi OSC-2759
- 3)
Reactor Building (RB) Pressure-High
=9 psig 0.6 psi (Barton); +/- 0.4 IP/O/A/0310/014A, B, C; psi (Rsmt) 4 psig 0.91 psi 3 psig
+ 0.09 p OSC-2495
- 4)
Reactor Building Pressure - High High See response to RAI, Question 1.
Enclosure, Attachment 3 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 2 SR 3.3.8.3 Table 3.3.8-1 Post Accident Monitoring Instrumentation Example Process Limit(8)
Total Loop Uncertainty t 2)
Example SetpointlDecision Point(9)
Setting References Tolerance(6R7)
- 1)
Wide Range Neutron Flux Reactor is not shutdown when power is between 0.1% and 10% FP.
+/- 4.97% span (log scale);
(i.e., 0.31% FP to 3.25%
FP)
- 2)
RCS Hot Leg Temperature
- 3)
RCS Hot Leg Level
- 4)
RCS Pressure (Wide Range)
- 5)
Reactor Vessel Head Level
- 6)
Containment Sump Water Level (Wide Range)
- 7)
Containment Pressure (Wide Range)
- 8)
Containment Isolation Valve Position
- 9)
Containment Area Radiation (High Range)
- 10)
Not Used
- 11)
Pressurizer Level
- 12)
Steam Generator Water Level
- 13)
Steam Generator Pressure
- 14)
Borated Water Storage Tank Level
- 15)
Upper Surge Tank Level
- 16)
- 17)
Subcooling Monitor (degrees subcooled)
- 18)
- 19)
LPI System Flow
- 20)
Not used
- 21)
Emergency Feedwater Flow Low Pressure Service Water Flow to LPI
- 22)
Coolers None requiring uncertainties.
+ 10.10 *F See response to RAI, Question 1.
See response to RAI, Question 1.
See response to RAI, Question 1.
IP/0OA102031001 H: OSC-No longer used in EOP's
+ 8.9 inches / -18.4 inches NA
+ 0.4 ft-H20 I28OSC-220 2578, OSC-2820 NA (12 month SR Frequency)
PT/I,2,3/A10202/01 2, Open/Closed NA (limit switches)
Open/Closed NA PT/1,2,3/A10160/003 Within +100% / -50% of reading.
32.8% of reading None requiring 18.6% of reading IP/O/A/0361/004; OSC-uncertainties
+_18.6% ofreading_
2904, OSC-2820 Not used 375 inches (indicates IP/O/A10200/052 A; OSC-400 inches = water solid PZR
+ 24 inches water solid PZR during i
3.0 in-H1O20 normal conditions) 4263, OSC-2820 See response to RAI, Question 1.
770 psig (to prevent FDW isolation 27.3 psi (normal);
700 psig (manually IP/O/A10270/013; OSC-during normal cooldown)
-2.4 psi +/- 102.84 psi deactive AFIS)
+/- 9.0 psi 4295, OSC-2820 (harsh) deactive AFIS)4295 OSC-282 See response to RAI, Question 1.
See response to RAI, Question 1.
= 15 *F
+3.69°F +/- 11.26oF (five 700°F (indicates ICC IP/O/A/0200/041 D; OSC-TC average) conditions) 3862, OSC-2820 WR Pressure & HL RTD 150°F subcooled maintain> 100 °F from NOT curve uncertainties built into (natural circulation NA 3P/I,2,3A1A200/042; OSC-subcooled margin curves.
cooldown) 3862, OSC-2820 525 gpm (conservative HPI pump
++
22.4 gpm 22.8 gpm 475 gpm, indicated
+/- 5.6 gpm at 375 IP/O/AI0202/001 D; OSC-runout flow)
I__
.4___ /-_2.8gp_45_pm indiIad I
gpm 1 4083, OSC-2820 See response to RAI, Question 1.
See response to RAI, Question 1.
Not used SA assumes 200 gpm (to each SG 1 300 gpm (to each SG 3
IP/O/A/0275/006 A & B; on LSCM indication)
+/- 90.3 gpm on LSCM indication)
+/- 30 gpm OSC-3221, OSC-2820 NA (12 month SR Frequency)
Enclosure, Attachment 3 - Duke Energy Response to NRC Request for Additional Information February 11,2011 Page 3 Example Process Limit1')
Total Loop Example Setting Uncertainty(2)
Functional Limit(g)
Tolerance(67)
References SR 3.3.10.2 Wide Range Neutron Flux See PAM Table 3.3.8-1, Item 1.
Automatic Feedwater SA assumes 520 psig & 3.3
+ 18.2 psi and 0.028 550 psig and 2.97 IP/01/A02701013; OSC-SR 3.3.11.3 Isolation System (AFIS) psUsec psi/sec ps/sec 6.0 psig 4295, OSC-2820 Instrumentation Low Pressure Service Water SR 3.3.28.2 Standby Pump Auto.Start See response to RAI, Question 1.
Circuitry Notes:
- 1)
An Analytical Limit (AL) is defined by EDM-102 as "Limit of a measured or calculated variable established by the Safety Analyses (SA) to ensure that a Safety Limit is not exceeded".
- 2)
Total Loop Uncertainty (TLU).
- 3)
Allowable Value (AV) as specified in Table 3.3.1-1 (RPS) or Table 3.3.5-1 (ESPS) of ONS Technical Specifications
- 4)
Margin = AV - NTSP - CTE for increasing trip. Margin = NTSP -AV -CTE for decreasing trip.
- 5)
Nominal Trips Setpoint (NTSP), which is the value actually calibrated to in the instrument procedures (IP's).
- 6)
Device or rack string setting tolerance (CTE). In this case, the rack setting tolerance consists only of the bistable.
- 7)
The setting tolerance (CTE) is equal to the As-Found Tolerance (AFT). At Oconee, it is also equal to the As-Left Tolerance (ALT).
- 8)
Instrument loop functions do not have Analytical Limits. An example process limit is shown instead.
- 9)
Instrument loop functions do not have Allowable Values. An example setpointldecision point is shown instead.
- 10)