ML063320468
| ML063320468 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/03/2006 |
| From: | Peifer M Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:6331-05 | |
| Download: ML063320468 (37) | |
Text
Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, MI 49106 POWER6 AElcom A unit of American Electric Power November 3, 2006 AEP:NRC:6331-05 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001
SUBJECT:
Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Docket Nos.: 50-315 and 50-316 Application for Amendment to Revise Unit 1 and Unit 2 Technical Specifications to Reflect Replacement of Existing Reactor Coolant System Resistance Temperature
- Detectorsand Bypass Piping with Detectors Mounted in the Primary Loop Piping
Reference:
Letter from P. S. Tam, U. S. Nuclear Regulatory Commission, to M. K. Nazar, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1) - Issuance of Amendment Regarding Elimination of the Resistance Temperature Detector (RTD) Bypass Loop (TAC No. MD2106)," dated October 6, 2006.
Dear Sir or Madam:
Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP)
Unit 1 and Unit 2, proposes to amend the Technical Specifications (TS) to reflect a plant modification that will replace the Reactor Coolant System resistance temperature detectors (RTDs) and bypass piping with fast response thermowell detectors mounted directly in the primary loop piping. The specific TS requirements affected include the notes in the Unit 2 TS Surveillance Requirement (SR) for channel calibration of the overtemperature differential temperature (OTAT) and overpower differential temperature (OPAT) Reactor Trip System functions. The changes to the notes are needed to allow the unit to return to operation following the plant modification. The proposed change also affects the Unit 1 and Unit 2 TS Allowable Values for OTAT and OPAT Reactor Trip System functions. These changes are requested so that the TS Allowable Values will reflect the increased margin resulting from the plant modification. The associated TS Bases will be changed in accordance with the CNP Bases Control Program.
Changes to the notes in the Unit 1 TS SR for channel calibration of the OTAT and OPAT Reactor Trip System functions have already been approved by the referenced letter. The Unit 1 RTD bypass plant modification is being performed during the current Unit 1 refueling outage. New Unit 1 TS Allowable Values for OTAT and OPAT were not included in the requested amendment because the A001f
U. S. Nuclear Regulatory Commission AEP:NRC: 6331-05 Page 2 existing Allowable Values remain bounding for the new fast response thermowell RTD system, and because the amendment was needed on an expedited basis. However, the new Unit 1 TS OTAT and OPAT Allowable Values proposed by this letter will reflect the increased margin resulting from the plant modification. The changes to the Unit 2 TS OTAT and OPAT SR notes and the Unit 2 TS OTAT and OPAT Allowable Values will support the RTD bypass plant modification in Unit 2 during its Fall 2007 refueling outage. The new Unit 1 and Unit 2 OTAT and OPAT Allowable Values were determined using methodology previously approved by the Nuclear Regulatory Commission (NRC). to this-letter provides an affirmation affidavit pertaining to the proposed amendment. provides1 a detailed description and safety analysis to support the proposed amendment, including an evaluaion of significant hazards considerations pursuant to 10 CFR 50.92(c), and an environmental assessment. Enclosure 3 provides a description of the previously approved methodology used to. determine the new Unit 1 and Unit 2 TS Allowable Values for OTAT and OPAT. Attachme'nts IA and 1B provide the affected TS pages marked to show changes for Unit 1 and Unit 2, respe6t-iy1i: :Attachments 2A and 2B provide the affected TS pages with the proposed changes incorporated for Unit 1 and Unit 2 respectively.
I&M requests NRC approval of the proposed amendment by September 8, 2007. I&M requests implementation of the proposed amendment be required prior to Unit 2 entry into Mode 2 during the unit's Fall 2007 outage.
Copies of this letter and its enclosures and attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality in accordance with the requirements of 10 CFR 50.91.
This letter contains no new regulatory commitments. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.
JRW/jen
U. S. Nuclear Regulatory Commission AEP:NRC:6331-05 Page 3
Enclosures:
- 1. Affirmation.
- 2. Application for Amendment.
- 3. Previously Approved Methodology Used to Determine New Overtemperature Differential Temperature and Overpower Differential Temperature Allowable Values Attachments:
1A. Unit 1 Technical Specification Pages Marked to Show Proposed Change lB. Unit 2 Technical Specification Pages Marked to Show Proposed Change 2A. Unit 1 Proposed Technical Specification Pages 2B. Unit 2 Proposed Technical Specification Pages c: J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne J. T. King - MPSC MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam - NRC Washington, DC
Enclosure I to AEP:NRC:6331-05 AFFIRMATION I, Mark A. Peifer, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Con-inussion on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.
Indiana Michigan Power: Company MarkA. Peifevr Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS 0 DaY I- A OF JOVrva-_bQ.r- ,2006 Notay ublid)
REGAN D.WENDZEL My Commission Expires -Notary Public, Berrien County, MI
- My Commisslon Expires Jan. 21; 200P
Enclosure 2 to AEP:NRC:6331-05 APPLICATION FOR AMENDMENT
1.0 DESCRIPTION
References for this attachment are identified on Page 12 and Page 13.
Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP)
Unit 1 and Unit 2, proposes to amend the Technical Specifications (TS) to reflect a plant modification to replace the Reactor Coolant System (RCS) resistance temperature detectors (RTDs) and bypass piping with fast response therinowell detectors mounted in the primary loop piping. The specific TS requirements affected include the notes in the Unit 2 TS Surveillance Requirement (SR) for channel calibration of the overtemperature differential temperature (OTAT) and overpower differential temperature (OPAT) Reactor Trip System functions. The changes to the notes are needed to allow the unit to return to operation following the plant modification. The proposed change also affects the Unit 1 and Unit 2 TS Allowable Values for OTAT and OPAT Reactor Trip System functions. These changes are requested so that the TS Allowable Values will reflect the increased margin resulting from the plant modification. The new Unit, 1 and Unit 2 OTAT and OPAT Allowable Values were determined using methodology previously approved by the Nuclear Regulatory Commission (NRC).
Changes to the notes in the Unit 1 TS SR for channel calibration of the OTAT and OPAT Reactor Trip System functions have already been approved by Reference 1. The Unit 1 RTD bypass plant modification is being performed during the current Unit 1 refueling outage. New Unit 1 TS Allowable Values for OTAT and OPAT were not included in the requested amendment because the existing Allowable Values remain bounding for the new fast response thermowell RTD system, and because the amendment was needed on an expedited basis. However, the new Unit 1 TS OTAT and OPAT Allowable Values proposed by this letter will reflect the increased margin resulting from the plant modification. The changes to the Unit 2 TS OTAT and OPAT SR notes and the Unit 2 TS OTAT and OPAT Allowable Values will support the RTD bypass plant modification in Unit 2 during its Fall 2007 refueling outage.
2.0 PROPOSED CHANGE
Existing Note 1 in Unit 2 Reactor Trip System TS SR 3.3.1.15 requires that an OTAT or OPAT channel calibration include verification of Reactor Coolant System RTD bypass loop flow rate.
This note will be deleted. The note number will be removed from existing Note 2 of Unit 2 TS SR 3.3.1.15, since it will be the sole remaining note for this SR.
Existing Note 1 and Note 2 in Unit 1 and Unit 2 TS Table 3.3.1-1 specify the Allowable Values for the OTAT and OPAT Reactor Trip System functions as less than or equal to (<) a percent of differential temperature (AT) span. These Allowable Values will be changed as follows:
to AEP:NRC:6331-05 Page 2 Existing Allowable Value - Proposed Allowable Value -
Percent of AT Span Percent of AT Span Unit 1 OTAT function <0.008 <0.050 Unit 1 OPAT function <0.037 <0.132 Unit 2 OTAT function <0.012 <0.077 Unit 2 OPAT function <0.038 <0.140 Attachments IA and 1B provide the affected TS pages marked to show changes for Unit 1 and Unit 2, respectively. Attachments 2A and 2B provide the affected TS pages with the proposed changes incorporated for Unit 1 and Unit 2 respectively.
3.0 BACKGROUND
Existing Design The following provides a description of the Unit 1 design existing prior to the RTD bypass modification that is being performed during the current refueling outage, and the Unit 2 design existing prior to the RTD bypass modification which will be performed during the Fall 2007 refueling outage.
CNP Unit 1 and Unit 2 have four reactor coolant loops. In the existing design, an RTD bypass manifold system is used to obtain representative hot leg and cold leg temperatures in each reactor coolant loop. Each loop has separate hot leg and cold leg bypass inlet piping connections and manifolds. A representative hot leg temperature is obtained by mixing flow from three scoop connections. These scoops extend into the flow stream (at locations 120 degrees apart in the cross-sectional plane) on each reactor coolant hot leg. Each scoop has five flow holes which sample the hot leg coolant. Flow for the cold leg bypass manifold is obtained downstream of the reactor coolant pump (RCP) discharge. The hot and cold bypass manifold piping for each loop join to form a common discharge line. The combined flow discharges to the suction side of the RCP. The existing RTD bypass manifold system consists of approximately 500 feet of reactor coolant pressure boundary piping (not including instrument tubing), 44 valves, approximately 40 pipe hangers (including nine snubbers), eight sets of flanges, and eight RTD manifolds.
In the existing design, the RTDs extend directly (without thermowells) into the bypass manifold fluid flow. This minimizes the response time of the RTDs. The RTD outputs are used to calculate the AT and average reactor coolant loop temperature (Tavg) signals that are used by the Reactor Trip System and by control systems. The loop AT and/or loop Tavg signals are used in the OTAT and OPAT functions in the Reactor Trip System; the Steam Line Isolation on High Steam Flow Coincident with Low-Low Tavg and P-12 Low-Low Tavg interlock functions in the Engineered Safety Features Actuation System; and in verifying RCS total flow. In addition, the to AEP:NRC:6331-05 Page 3 RTD outputs are used for alarms and indication, rod control, turbine runback, pressurizer level, and other control systems.
The existing RTD bypass manifold system design was developed to resolve concerns with temperature streaming (temperature gradients) within the RCS hot legs. The temperature streaming i the hot leg piping was a result of incomplete mixing of coolant leaving various regions of the reactor core at different temperatures. The bypass manifold system compensates for the temperature streaming by sampling the primary coolant through scoop tubes and mixing the primary coolant within the bypass manifold to develop an average RCS hot leg temperature for the loop. The bypass manifold system also limits the velocity of the coolant flow to the RTDs, and allows RTD replacement without the need to drain the RCS loops.
The coolant velocity through the RTD bypass piping system is low relative to the RCS loop velocity. As a result, the bypass piping, valves, and manifolds become collection points for activated corrosion products. These components tend to become radiological hot spots, which significantly increase the general area radiation levels. Due to their proximity to the RCP and steam generator related work locations, the RTD manifolds are significant contributors to personnel radiation dose during unit outages. To reduce this dose contribution, temporary lead shielding is installed during each outage. Installation and removal of this shielding is labor intensive, negatively impacts other work activities requiring material transport through the containment hatch, and involves radiation exposure to personnel installing the shielding. I&M estimates that the RTD bypass piping system contributes approximately 40 percent (30 person-rem) of the overall dose each refueling outage. Additional concerns associated with the existing system include the large number of components requiring maintenance, repair, surveillance, and testing. Review of industry operating experience has identified the RTD bypass manifold valves as sources of RCS leakage events. Although the existing RTD bypass manifold system has been adequately performing its intended function, elimination of the system reduces personnel radiation exposure, outage costs, and maintenance concerns.
Modification The following provides a description of the RTD bypass modification which is being performed during the current Unit 1 refueling outage, and will be performed during the Fall 2007 Unit 2 refueling outage.
I&M is eliminating the RTD bypass manifold system from all four RCS loops. All piping and valves associated with this system will be removed. Associated RTD bypass system pipe supports will also be removed. This is a standard industry modification which has been implemented successfully by other Westinghouse plants.
The three hot leg scoops will be modified to accept new thermowells, which will contain new, fast-response RTDs (Model N9004E, manufactured by Weed Instruments Incorporated). A hole will be drilled through the end of each scoop to facilitate flow past the RTD. Water will enter to AEP:NRC:6331-05 Page 4 through the existing openings, flow past the RTD, and exit through the new hole at the end of the scoop. The existing cold leg RTD bypass piping nozzle will also be modified to accept an RTD thermowell. The cold leg RTD thermiowell will be inserted directly in the coolant flow path.
The Reactor Trip System will be modified to calculate an average hot leg temperature for each loop, using the three new RTD signals. The electronically-averaged temperature will function similarly to the single temperature input provided by the existing manifold RTD. The new fast-response thermowell RTD system will provide the same degree of independence and redundancy as the existing system.
4.0 TECHNICAL ANALYSIS
System Functions The proposed change will delete an existing Unit 2 TS note that requires verification of the RTD bypass loop flow rate as part of the OTAT and OPAT channel calibrations, and will increase the Unit 1 and Unit 2 TS Allowable Values for the OTAT and OPAT Reactor Trip System functions.
The OTAT Reactor Trip System function provides primary protection against departure-from-nucleate-boiling (DNB) during transients in Westinghouse pressurized water reactors. The measured AT is used as an indication of reactor power and is compared to the OTAT setpoint, which varies depending upon the measured Tavg, pressurizer pressure, and the reactor axial flux difference signal. If the measured AT exceeds the OTAT setpoint in more than one loop, a reactor trip signal is generated. The OPAT Reactor Trip System function is designed to protect against a high fuel rod power density and-thus preclude fuel centerline melting. The measured AT is used as an indication of reactor power and is compared to the OPAT setpoint, which varies depending upon the indicated Tavg. If the measured AT exceeds the OPAT setpoint in more than one loop, a reactor trip signal is generated.
Response Time The discussions of response time considerations in this section apply to Unit 2. I&M provided a similar discussion for Unit 1 in support of the amendment approved by Reference 1.
Replacement of the existing RTD bypass system with the new fast-response thermowell RTD system will affect the RCS temperature measurement response characteristics that are currently modeled as part of the OTAT and OPAT reactor trips credited in certain non-loss of coolant accident (LOCA) analyses. As shown in Unit 2 Table 14.1.0-4 of the Updated Final Safety Analysis Report (UFSAR), the OTAT and OPAT Reactor Trip System functions have a total time delay of 8 seconds assumed in the analyses. The following table shows how this 8-second assumption is maintained by the existing RTD bypass system and how it will be maintained by the new fast-response thermowell RTD systemn to AEP:NRC:6331-05 Page 5 Response Time Parameters for RCS Temperature Measurement Component Existing RTD Bypass New Fast-Response System Thermowell RTD System (seconds) (seconds)
RTD bypass piping transport and thermal 4 N/A lag.
RTD response time 2 4 Electronics signal processing, reactor trip 2. 2'..
signal, trip breaker opening, and rod cluster control assembly gripper release.
Total Response Time 8 Less than or I_ equal to 8 The OTAT and OPAT reactor trip model used in the non-LOCA analyses includes a 6-second first order lag time for the temperature sensor response plus a 2-second delay for the: electronic time response, for a total time of 8 seconds.from the time that the setpoint is reached to the loss of stationary gripper coil voltage, i.e., when the rod cluster control assemblies (RCCAs) are free to fall. As shown in the above table, the 6-second first order lag includes the RTD response time, and the bypass pipe coolant transport and thermal heatup time. The 2-second electronic delay accommodates electronic signal processing, reactor trip signal, trip breaker opening, and RCCA gripper release.
Due to the new fast response thermowell RTD system, the individual components that comprise the OTAT and OPAT reactor trip response time assumed in the non-LOCA analyses will be altered. Although the individual component times will be different, the total OTAT and OPAT reactor trip response time assumption of 8 seconds will be met. As described below, evaluations were performed to assess the impact that a change in the OTAT and OPAT component response times would have on non-LOCA accidents analyzed in the UFSAR that rely on OTAT or OPAT trips for reactor protection.
Uncontrolled RCCA Withdrawal at Power (UFSAR Unit 2 Section 14.1.2)
An uncontrolled RCCA bank withdrawal at power (RWAP) event can occur due to an improper operator action or a malfunction of the rod control system, and will result in an increase in the core heat flux due to the positive reactivity addition. The event would be terminated by either a high neutron flux or OTAT reactor trip function. The event is classified as a Condition II event, i.e., an incident of moderate frequency, as defined by the American National Standard ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of stationary Pressurized Water Reactor Plants." The event is analyzed to demonstrate that the DNB design basis is satisfied.
to AEP:NRC:6331-05 Page 6 A spectrum of reactivity insertion rates from several different power levels (100 percent (%),
60%, and 10% power) was considered for both beginning and end-of-life conditions. The cases that result in a reactor trip from the high neutron flux setpoint are unaffected by the change in the OTAT response time components. For lower reactivity insertion rates, the OTAT reactor trip function provides the primary protection, as indicated by Unit 2 Figures 14.1.2-7B through 14.1.2-9B, in the UFSAR. A CNP-specific sensitivity analysis was performed in which the limiting DNB ratio (DNBR) cases from the existing UFSAR RWAP analyses were run, assuming various combinations of lags and delays with a total response time of 8 seconds.
Based on the results of the existing UFSAR RWAP analyses and the sensitivity analysis, it was shown that there is no significant effect on the calculated minimum DNBR. In all cases, the minimum DNBR remained above the safety analysis DNBR limit. Therefore, the DNB design basis will be met and the conclusions in UFSAR Unit 2 Section 14.1.2 will remain valid.
Chemical and Volume Control System Malfunction (UFSAR Unit 2 Section 14.1.5)
The chemical and volume control system malfunction (boron dilution) analysis is performed to demonstrate that sufficient time is available following initiation of the event to allow an operator to determine the cause of the inadvertent dilution and take corrective action before'shutdown margin is lost. The event is classified as a Condition II event by ANSI N18.2-1973. This evenm is bounded by another Condition II event, the RWAP event, with respect to Condition II criteria, such as ensuring that the DNB design basis. is satisfied, and maintaining peak primary and secondary pressures less than 110 percent of design.
A change in OTAT individual component response time for the boron dilution event presented in the UFSAR would only potentially affect the case analyzed at full power with manual rod control. For this case, the operator action time is measured from the time of reactor trip (on OTAT) until a loss of the plant shutdown margin. Since a boron dilution transient at full power with manual rod control results in a reactivity insertion essentially equivalent to an RWAP event, a conservative time for reactor trip was selected from the times calculated for the full power RWAP analysis. Based on sensitivity runs generated for the RWAP event, where various combinations of RTD lag and delay response times were modeled, the time of rod motion would be delayed by a maximum of 3 seconds in the full power cases.
The acceptance criterion in the CNP accident analysis for a boron dilution event at full power with manual rod control is that a minimum of 15 minutes be available for operator action. The interval from the time of reactor trip on OTAT to loss of shutdown margin calculated for the UFSAR licensing basis boron dilution event at full power with manual rod control is 44 minutes for Unit 2. Thus, significant margin is available. A 3-second increase in OTAT reactor trip response does not affect the results of the licensing basis analyses. Therefore, the operator action time criterion continues to be met, and the conclusions in UFSAR Unit 2 Section 14.1.5 will remain valid.
to AEP:NRC:6331-05 Page 7 Loss of External Electrical Load (UFSAR Unit 2 Section 14.1.8)
A loss of external electrical load can result from an abnormal variation in the network frequency, a trip of the turbine, or the spurious closure of the turbine stop or control valves or steamline isolation valves. If a failure of the steam dump valve system also occurs, the sudden reduction in steam flow will result in an increase in the pressure and temperature in the steam generators.
Heat transfer will be reduced, causing the reactor coolant temperature and pressure to rise. The loss-of-load event is analyzed to confirm that the pressurizer and steam generator safety valves are adequately sized to prevent over-pressurization of the RCS and steam generators, and to ensure that the DNB design basis is satisfied. The event is classified as a Condition II event by ANSI N18.2-1973.
The high pressurizer pressure, low steam generator level, and OTAT reactor trip functions provide protection for a loss-of-load/turbine trip event. Four cases were analyzed for this event.
Two limiting cases were analyzed for peak RCS pressure concerns. These two cases assumed no pressurizer pressure control for both beginning and end-of-life reactivity feedback conditions, and a reactor trip results from the high pressurizer pressure Reactor Trip System function.
Therefore, these cases are not affected by the OTAT response time modeling.
The other two cases were analyzed to demonstrate that the DNB design basis is satisfied for limiting loss-of-load/turbine trip events. These cases assumed pressurizer pressure control for both beginning and end of life reactivity feedback conditions. In these two cases, the beginning of life (minimum) feedback case is terminated by the OTAT reactor trip function. For this case, explicit sensitivities were generated for CNP, with modeling of various combinations of lag and delays. The results of these sensitivity studies indicated there would be no significant effect on the calculated minimum DNBR. Therefore, the minimum DNBR remains above the safety analysis limit, the DNB design basis is met, and the conclusions in UFSAR Unit 2 Section 14.1.8 will remain valid.
Steam Line Isolation on High Steam Flow Coincident with Low-Low Tav_ and P-12 Low-Low T,,v_.Interlock Engineered Safety Features Actuation System Functions As indicated in TS Table 3.3.2-1 (Functions 4.e and 8.c) and UFSAR Table 7.2-7, there are no response time requirements for the Steam Line Isolation on High Steam Flow Coincident with Low-Low Tavg function or the P-12 Low-Low Tavg Interlock Engineered Safety Features Actuation System functions.
Response Time Testing The discussion of response time testing in this section applies to Unit 2. I&M provided an identical discussion for Unit 1 in support of the amendment approved by Reference 1.
to AEP:NRC:6331-05 Page 8 The RTD manufacturer will perform time response testing of each RTD and thermowell prior to installation at CNP. The RTDs and thermowells must exhibit a response time bounded by the values shown in the preceding table. In addition, response timie testing of the RTDs will be performed in-situ in accordance with TS SR 3.3.1.19. This testing will demonstrate that the RTDs can satisfy the response time requirement when installed in the plant.
Instrument Uncertainty Considerations OTAT and OPAT The description of instrument uncertainty considerations for Allowable Values for OTAT and OPAT in the following paragraph applies to both Unit 1 and Unit 2. I&M provided a description of instrument uncertainty considerations for Unit 1 OTAT and OPAT in support of the amendment approved by Reference 1.. However, the Unit 1 instrument uncertainty considerations were addressed only to the extent needed to support maintaining the existing TS Allowable Values. The following discussion supports the proposed new TS Allowable Values for both Unit 1 and Unit 2.
OTAT and OPAT instrument uncertainty calculations have been performed for the new fast-response thermowell RTD system in Unit 1 and Unit 2. The' uncertainty calculations include a measurement term to address the effects of hot leg temperature streaming. Temperature streaming will exist in the hot leg due to incomplete mixing of coolant leaving various regions of the reactor core. The use of three flow scoops located at 120 degree increments along the circumference of the hot leg loop pipe reduces the streaming effects. The effects of cold leg streaming are not included in the calculation because it is considered in the safety analysis margin. The results of the instrument uncertainty calculations were used to determine the proposed new Unit 1 and Unit 2 TS Allowable Values for the OTAT and OPAT functions based on use of the new fast response thermowell RTD system. The proposed new Allowable Values were determined in accordance with the same methodology used to determine Allowable Values during the conversion of the CNP TS to the Improved Standard TS format of NUREG 1431.
This methodology was described on Pages 590-1 through 590-5 of 666 of an NRC website established to document resolution of issues during the conversion of the CNP TS to the structure of NUREG 1431. These pages are provided in Enclosure 3 to this letter. As detailed in the cover page for Enclosure 3, the website contents were submitted (Reference 2) to the NRC in support of the Improved Technical Specification conversion. The conversion was approved by the NRC (Reference 3) based, in part, on the information provided on the website and the staff s review and approval of the methodology used to determine Allowable Values.
The proposed new OTAT and OPAT Allowable Values determined using this methodology are identified above in Section 2.0, "Proposed Change," and shown on the TS pages provided as Attachments IA, IB, 2A, and 2B to this letter.
to AEP:NRC:6331-05 Page 9 Low-Low Tv,,, Setpoint and RCS Total Flow Rate Analysis The discussions of instrument uncertainty considerations for TS Allowable Values for the Low-Low Tavg setpoint and for the RCS total flow rate analysis in the following paragraphs apply to Unit 2. I&M evaluated the potential effect of new instrument uncertainties on the Unit 1 TS allowable value for the Low-Low TYavg setpoint and on the Unit 1 RCS total flow rate analysis in support of the Unit 1 amendment approved by Reference 1.
Unit 2 Engineered Safety Features Actuation System instrumentation TS Table 3.3.2-1, Functions 4.e and 8.c, specify the Allowable Values for the Low-Low Tavg setpoint. I&M calculations have confirmed that the existing Unit 2 Low-Low T,,vg TS Allowable Values will bound the instrument uncertainty of the new fast-response thermowell RTD system.
Unit 2 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling Limnits TS SRs 3.4.1.3 and 3.4.1.4 requhe periodic verification that RCS total flow rate is greater than or equal to (>) 366,400 gallons per minute (gpm). RCS flow measurement uncertainty is dependent, in part, on the accuracy of hot and cold leg temperature measurement. I&M's engineering evaluation has determined that the existing Unit 2 TS RCS total flow rate limit of > 366,400 gpm will bound the instrument uncertainty impact of the new fast response thermowell RTD system.
RTD Element Failure The discussion of an RTD element failure in this section applies to Unit 2. I&M provided an identical discussion for Unit 1 in support of the amendment approved by Reference 1.
As with the existing RTD bypass system, the failure of an RTD would be identified using existing control board alarms and indicators following installation of the new fast response thermowell RTD system. These alarms and indicators include Tvg deviation alarms, AT deviation alarms, Tavg -Tref deviation alarms, and shiftly rounds, which verify all required Tavg and AT indications. If a deviation alarm for a channel is received, or if a channel check during operator rounds reveals a deviation in one or more channels, the condition is evaluated. If the condition is determined to be caused by a failed RTD, the following actions can be taken.
In the existing RTD bypass system, the hot and cold leg RTD manifolds each contain an active single element RTD and a spare single element RTD. The spare RTD can be connected in the event of a failure of an operating RTD. In the new fast response thermowell RTD system, each RTD contains a second element that can be connected to the circuit in the event of an operating element failure. Switchover to the spare RTD element can be performed at the appropriate terminal blocks in a junction box in the reactor cable tunnel, which is located outside the containment and is accessible during reactor operation at power.
to AEP:NRC:6331-05 Page 10 RCS Pressure Boundary Codes The discussion of RCS pressure boundary codes in this section applies to Unit 2. I&M provided a simnilar discussion for Unit 1 in support of the amendment approved by Reference 1.
The piping analyses for the RTD bypass removal modification will use the governing code for the CNP RCS piping, ANSI B31.1, 1967 Edition, or later editions of ASME Code Section III with reconciliation to ANSI B31.1, 1967 Edition. The RCS is an Inservice Inspection Class 1 system governed by ASME Code Section XI, 1989 Edition, for repair and replacement of pressure retaining components and their supports. The new welds will be inspected in accordance with Section XI requirements. The RCS pressure boundary will be leak tested at normal operating temperature and pressure per ASME Section XI, Code Case N-416-1.
5.0 REGULATORY SAFETY ANALYSIS No Significant Hazards Consideration Indiana Michigan Power Company (I&M) has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
I Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
Response: No The resistance temperature detectors (RTD) bypass system is the hardware associated with Reactor Coolant System instrumentation having control, indication, and protection functions.
The RTD bypass system is not considered a precursor to any previously analyzed accident.
The system is relied upon to mitigate the consequences of some accidents. The new system replacing the RTD bypass system will perform the same control, indication, and protection functions, and, similarly, will not be considered a precursor to any accident. The capability of the system to mitigate the consequences of the previously analyzed accidents will not be significantly affected. Therefore, replacement of the existing RTD bypass system with the new system will not increase the probability of occurrence of an accident, and will not increase consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
to AEP:NRC:6331-05 Page I11 The replacement of the existing RTD bypass with the new system would not create new failure modes, and the replacement system is not an initiator of any new or different kind of accident. The proposed deletion of the note in Technical Specifications (TS) Surveillance Requirement 3.3.1.15, and proposed changes to Allowable Values in TS Table 3.3.1-1 do not affect the interaction of the replacement system with any system whose failure or malflunction can initiate an accident. Therefore, the proposed change does not create the possibility of a new of different kind of accident from any previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No Margins ,of safety are established in the design of components, the configuration of components to meet certain perforimance parameters, and in the models and associated assumptions used to analyze the system's performance. The replacement system will continue to perform the same temperature detection function to the same level of reliability as defined in the Donald C. Cook Nuclear Plant Updated Final Safety Analysis Report.
Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.
In summary, based upon the above evaluation, I&M has concluded that the proposed change involves no significant hazards consideration under. the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.
Applicable Regulatory Requirements/Criteria 10 CFR 50.36 requires that each license authorizing operation of a production or utilization facility include TS. The TS are required to include SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that the facility operation will be within safety limits, and that the limiting conditions for operation will be met. This amendment deletes an SR that will become obsolete following the implementation of a plant modification to delete the RTD bypass manifold system. This amendment also changes Allowable Values associated with that SR, based on the effects of the plant modification. I&M has determined that no other TS are affected.
In conclusion, based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the NRC's regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
to AEP:NRC:6331-05 Page 12
6.0 ENVIRONMENTAL CONSIDERATION
I&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, or would change an inspection or SR. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
7.0 PRECEDENTS In addition to the CNP Unit 1 amendment approved by Reference 1, precedent amendments have been approved for at least 15 other nuclear power plants. Three other plants that more recently received amendments solely addressing removal of the RTD bypass were the Byron and Braidwood Nuclear Power Stations (Reference 4) and the North Anna Power Station (Reference 5). The proposed CNP amendment differs from the precedent amendments for the three other plants in that CNP has converted its TS to the NUREG 1431 Improved Standard Technical Specifications.
8.0 REFERENCES
- 1. Letter from P. S. Tam, U. S. Nuclear Regulatory Commission, to M. K. Nazar, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit 1 (DCCNP-1) - Issuance of Amendment Regarding Elimination of the Resistance Temperature Detector (RTD) Bypass Loop (TAC No. MD2106)," dated October 6, 2006.
- 2. Letter from M. K. Nazar, I&M, to U. S. NRC Document Control Desk, "Copy of Nuclear Regulatory Commission and Donald C. Cook Nuclear Plant Improved Technical Specifications Conversion Website (TAC Nos. MC2629 and MC2360)," dated April 15, 2005, AEP:NRC:5901-02, ML051150349.
- 3. Letter from J. Donohew, NRC, to M. K. Nazar, I&M, "D. C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendment for the Conversion to the Improved Technical Specifications with Beyond Scope Issues (TAC Nos. MC2629, MC2630, MC2653 through MC2687, MC2690 through MC2695, MC3152 through MC3157, and MC3432 through MC3453),"
dated June 1, 2005, ML050620034.
to AEP:NRC:6331-05 Page 13
- 4. Letter from R. R. Assa, NRC, to D. L. Farrar, Commonwealth Edison Company, "Issuance of Amendments - Byron and Braidwood Stations (TAC Nos. M91667, M91668, M91669, and M91670)," dated September 5, 1995, ML020870191.
- 5. Letter from L. B. Engle, NRC, to W. L. Stewart, Virginia Electric and Power Company, "North Anna Units 1 and 2 - Issuance of Amendments Re: Elimination of Resistance Temperature Detectors and Substitution of Thermowells (TAC Nos. M82838 and M82839),"
dated April 22, 1992, ML013480129.
Enclosure 3 to AEP:NRC:6331-05 PREVIOUSLY APPROVED METHODOLOGY USED TO DETERMINE NEW OVERTEMPERATURE DIFFERENTIAL TEMPERATURE AND OVERPOWER DIFFERENTIAL TEMPERATURE ALLOWABLE VALUES This enclosure provides pages 590-1 through 590-5 of 666 from a Nuclear Regulatory Conirission (NRC) website established to document resolution of issues during the conversion of the Donald C. Cook Nuclear Plant (CNP) Technical Specifications (TS) to the structure of NUREG 1431. These pages were provided in response to an NRC request made during that conversion process, and describe the methodology that was used to determine TS Allowable Values as part of the conversion. The NRC request and I&M's response are documented by Reference 2 of Enclosure 2 to this letter. The NRC approved this methodology as documented in Paragraphs G. 1.2.a, G. 1.2.b, and G.3.2 of the Safety Evaluation for the conversion of the CNP TS to Improved Technical Specifications, as documented in Reference 3 of Enclosure 2 to this letter.
This methodology was also used to determine the new Unit I and Unit 2 TS overtemperature differential temperature and overpower differential temperature Allowable Values proposed by this amendment request, based on use of a new fast-response thermowell resistance temperature detectors system.
IMPROVED TECHNICAL SPECIFICATION LICENSING DATABASE The following is a description of the current D. C. Cook Nuclear Plant (CNP) setpoint methodology, as discussed with the NRC Staff during the NRC Public Meeting held on September 14, 2004. From this description, and as agreed to by the NRC Staff during the meeting, Indiana Michigan Power Company (I&M) concludes that the current calculations performed to support the revised Allowable Values (AVs) provided in the ITS submittal that support increasing CHANNEL CALIBRATION for the affected instrumentation to 24 months are in accordance with NRC-approved methodologies consistent with the CNP licensing basis, and are therefore acceptable.
1.1. NTSP Evaluation Per WCAP-12741, the setpoint is established based on the following formula:
Nominal Trip Setpoint (NTSP) = Safety Analysis Limit (SAL) + Channel Statistical Allowance (CSA) + Margin CSA = [PMA + PEA + (SCA+SMTE+SD)2 + SPE2 + STE2 +
(RCA+RMTE+RCSA+RD) 2 + RTE 2]1/ 2 + EA Where:
- CSA Channel Statistical Allowance
- PMA Process Measurement Accuracy
" PEA Primary Element Accuracy
" SCA Sensor Calibration Accuracy
- SMTE Sensor Measurement and Test Equipment Accuracy
" SD Sensor Drift
- SPE Sensor Pressure Effects
- STE Sensor Temperature Effects
" RCA Rack Calibration Accuracy
- RMTE Rack Measurement and Test Equipment Accuracy
- RCSA Rack Comparator Setting Accuracy
" RD Rack Drift
- RTE Rack Temperature Effects
" EA Environmental Allowance CSA contains all identified errors for the process, instruments and environments identified for a given trip function at the time when the function is assumed to be required for protection or mitigation. The individual terms identified above may contain sub-terms that are not specifically identified. As an example, the SMTE term would contain the errors for input and output Measurement and Test Equipment (M&TE) and would also contain the errors associated with the standards and procedures used for calibration of the M&TE.
page 590-1 of 666 2/16/2005
IMPROVED TECHNICAL SPECIFICATION LICENSING DATABASE 1.2. AV Evaluation The AV is then established as the Nominal Trip Setpoint plus the lesser of the two trigger values, T1 or T2 (T 3 is the same as T1 for multiple sensor loop applications).
AV = NTSP + (smaller of T1 or T2)
The T1 term is defined as (Reference WCAP-12741):
T1 =[RCA + RMTE +RCSA + RD]
The T 2 term is defined as (Reference WCAP-12741):
T2 = TA - [(A+S 2 )1"2 + EA]
Where:
TA = NTSP - SAL A = [PEA 2 + PME 2 + SPE 2 + STE 2 + RTE 2]
S = [SCA +SMTE +SDR]
When the trigger values are evaluated to determine the most limiting or conservative Allowable Value, the following logic is used:
- 1. If T1 > T2:T 2 determines the Allowable Value (NTSP + T 2 )
- 2. If T1 <T 2: T, determines the Allowable Value (NTSP + TI)
A graphical depiction of the relationships between the terms used in the CNP setpoint methodology is as follows:
page 590-2 of 666 2/16/2005
IMPROVED TECHNICAL SPECIFICATION LICENSING DATABASE I Safety Analysis Limit (SAL)
CSA = TA = [PMA + PEA +
(RCA+RMTE+RCSA+RD) 2 +
RTE ]1/2 + EA ICSA TA (RCA+RMTE+RCSA+RD)
]
Setpoint Nominal Trip SetpointI Nominal page 590-3 of 666 2/16/2005
IMPROVED TECHNICAL SPECIFICATION LICENSING DATABASE The calculation of TI:
Is an arithmetic combination of rack uncertainties and assumes:
- The as-left condition was at the maximum allowed by the calibration procedure;
- The M&TE uncertainty was at the maximum allowed; and
- A process loop found within this value is operating within the drift tolerance This scenario for calculation of T1 is not considered a nominal condition, but an extreme "allowed" condition. The actual errors are not expected to all move in the same direction with the maximum possible magnitude. However, WCAP-12741 does not exclude the possibilities that the as-left setting is at the maximum allowed value, the M&TE has been calibrated with all error left at the maximum allowed value, and the unit will not trip until the maximum value is reached.
T, will always be greater than T2, with the possible exception when additional margin is added into the calculation of T1 . As noted in the graphic representation above, if the Nominal Trip Setpoint is not equal to the Calculated Trip Setpoint, then T1 could be more conservative than the T2 calculated value. In all cases, the Allowable Value selected by the setpoint methodology is the more conservative value of T1 and T2. Therefore, since T2 protects the Analytical Limit, any value more conservative would also protect the Analytical Limit. In addition, the use of a more conservative T1 value would identify instrument performance degradation prior to reaching the T2 value.
The results of an AV calculated based on the NTSP and T2 are mathematically equal to the AV results provided in ISA-$67.04.02 using Allowable Value Method 2.
Using there are three error groupings:
- 1. CSA, which includes all errors as follows:
CSA = TA = [PMA 2 + PEA 2 + (SCA+SMTE+SD) 2 + SPE 2 + STE 2 +
(RCA+RMTE+RCSA+RD)2 + RTE ]1/2 + EA
- 2. CHANNEL OPERATIONAL TEST (COT) errors, which include the errors associated with the performance of a COT on the applicable instrumentation as follows:
COT = (RCA+RMTE+RCSA+RD)
- 3. nCOT errors, which include all of the errors not associated with the performance of a COT on the applicable instrumentation as follows:
nCOT = [PMA2 + PEA2 + (SCA+SMTE+SD) 2 + SPE 2 + STE 2 + RTE 2] 1/2 + EA page 590-4 of 666 2/16/2005
IMPROVED TECHNICAL SPECIFICATION LICENSING DATABASE ISA-$67.04.02 Method 2 establishes the NTSP as follows:
NTSP = SAL - CSA ISA-S67.04.02 Method 2 establishes the AV as follows:
AV = SAL - nCOT The WCAP-12741 STEPIT methodology establishes the NTSP as follows:
NTSP = SAL - CSA The WCAP-12741 STEPIT methodology establishes the AV as follows:
AV = NTSP + T 2 (used for this discussion since it matches the ISA method)
By breaking the WCAP-12741 STEPIT methodology down into steps, it is apparent that the WCAP-12741 STEPIT methodology using T2 provides the same results as ISA-$67.04.02 Method 2 as follows:
NTSP = SAL - CSA AV = NTSP + T 2 T2 = TA (NTSP-SAL) - nCOT Substituting values results in the following:
AV = (SAL - CSA) + (CSA - nCOT)
Canceling the CSA terms results in the following:
AV = SAL - nCOT (equivalent to ISA-$67.04.02 Method 2)
In conclusion, the setpoint methodology described here is identical to the methodology in WCAP-12741, "Bases Document for Westinghouse Setpoint Methodology for Protection Systems," approved in the NRC SER for Technical Specification Amendments 175/160 for CNP Units 1 and 2, respectively. In addition, the NRC Staff review of the setpoint methodology during the September 14, 2004, Public Meeting has found that it is acceptable for use in determining new Allowable Values at CNP.
page 590-5 of 666 2/16/2005
Attactmient 1A to AEP:NRC:6331-05 UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGE 3.3.1-15 3.3.1-16
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function ...................... Allowable Value shall not exceed the following nominal Trip Setpoint by more than O8O% of AT span.
AT<Ao K1-K2(I ATýATO{KI{K +-C2 S) [T-T']+K3 (P-P')-fl(AI)}
2 (I+'zClS)
Where: AT is measured RCS AT,°F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec-'.
T is the measured RCS average temperature,°F.
T' is the nominal Tavg at RTP, < [*]OF.
P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, > [*] psig K, < [*] K2 > [*-]/F K3 > [*]/psig T1 > ['1sec tr 2 [*]sec fl(Al) = [*] {[*] + (qt - qb)} when qt - qb < [*]% RTP 0% of RTP when -[*]% RTP < qt - qb < [*]% RTP
-[*] {(qt - qb) - [*]} when qt - qb > [*]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 1 3.3.1-15 Amendment No. 287
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)
Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than O-.GM..1$O%of AT span.
AT <ATo{K 4 -K 5 T3-ST-K 6 IT -T"]-f 2 (AI)}
Where: AT is measured RCS AT, 0 F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec-'.
T is the measured RCS average temperature,°F.
T" is the nominal Tavg at RTP, :< [*]OF.
K4 < [*] K5 > [*]/OF for increasing Tavg K6 > [*]/OF when T > T"
[*]/OF for decreasing Tavg [*]/OF when T:5 T" a [*] sec f2(AI) = [*]
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 1 3.3.1-16 Amendment No. 287
Attachment 1B to AEP:NRC:6331-05 UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW PROPOSED CHANGE 3.3.1-9 3.3.1-15 3.3.1-16
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.11 -------------------- NOTES----------------
- 1. For Function 4, not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is below the P-1 0 interlock.
- 2. For Function 5, not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after THERMAL POWER is below the P-6 interlock.
Perform COT. 184 days SR 3.3.1.12 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.13 Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.14 --------------------- NOTE ----------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.15 -------------------- NOTES ----------------
- 1. This Sur*illaGne shall include verifiGation Of Reactor Coobn~t System resistanc-e temperature detecto~r bypass loopflow' rate.
- 2. Normalization of the AT is not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after THERMAL POWER is > 98% RTP.
Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.16 Perform COT. 24 months SR 3.3.1.17 Perform TADOT. 24 months Cook Nuclear Plant Unit 2 3.3.1-9 Amendment No. 269
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.01-26'OZ7% of AT span.
AAo(1+S) [T T']K 3 (P-P')-f AT*ATO{K 1 -K 2~ 1r 2S)(A)
Where: AT is measured RCS AT,°F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec'.
T is the measured RCS average temperature,°F.
T' is the nominal Tavg at RTP, _ [*]OF.
P is the measured pressurizer pressure, psig P. is the nominal RCS operating pressure, _ [*1 psig K1, [*] K2 > [*]/!F K3 > [*]/psig tC, [*]sec T2 < [*] sec fl(AI) = [*] {[*] + (qt - qb)} when qt - qb [*]% RTP 0% of RTP when -[*]% RTP < qt - qb [*]% RTP
-[*] {(qt - qb) - [*]} when qt - qb> [*]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 2 3.3.1-15 Amendment No. 269
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)
Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.038 A4O% of AT span.
AT *_ATo K4 -K5 3 T-K 6 [T ST"]- f 2(A1)}
0L 4 I++T 3S Where: AT is measured RCS AT,°F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec-'.
T is the measured RCS average temperature,°F.
T" is the nominal Tavg at RTP, < [*]OF.
K4 < [*] K5 > [*]/OF for increasing Tavg K6 > [*]/IF when T > T"
[*]/OF for decreasing TVg [*]/OF when T:5 T" T3 [*] sec f2 (AI) = [*]
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 2 3.3.1-16 Amendment No. 269
Attachment 2A to AEP:NRC:6331-05 UNIT 1 PROPOSED TECHNICAL SPECIFICATION PAGES 3.3.1-15 3.3.1-16
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.050% of AT span.
AT!A{K (1+CI S) [T-T']+K3 (P-P')-,f 1 (AI)
AT*ATo{K1-K 2 (1+-r12 S)
Where: AT is measured RCS AT,0 F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec1.
T is the measured RCS average temperature,°F.
T' is the nominal Tavg at RTP, < [*]OF.
P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, > [*] psig K1, [*] K2 > [*]/OF K3 > [*]/psig T, Ž [*] sec T2 :5 [*] sec fl(AI) = [*] {[*] + (qt - qb)} when qt - qb < -[*]% RTP 0% of RTP when -[*]% RTP < qt - qb [*% RTP
-[*] {(qt - q )- [*]} when qt - qb > [*]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 1 3.3.1-15 Amendment No. 2-87,
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)
Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.132% of AT span.
AT <ATo{K 4 -K 5 I3S T-K 6 [T -T"]-f 2 (AI)}
Where: AT is measured RCS AT,0 F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec-1.
T is the measured RCS average temperature,°F.
T" is the nominal Tavg at RTP, _ [*]OF.
K4 < [*] K5 > [*]/OF for increasing Tavg K6 > [*]/OF when T > T"
[*]/OF for decreasing Tavg [*]/oF when T < T" T3 [*] sec f2(AI) [*]
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 1 3.3.1-16 Amendment No. 2-97-,
Attachment 2B to AEP:NRC:6331-05 UNIT 2 PROPOSED TECHNICAL SPECIFICATION PAGES 3.3.1-9 3.3.1-15 3.3.1-16
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY i
SR 3.3.1.11 -------------------- NOTES----------------
- 1. For Function 4, not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is below the P-10 interlock.
- 2. For Function 5, not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after THERMAL POWER is below the P-6 interlock.
Perform COT. 184 days SR 3.3.1.12 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.13 Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.14 --------------------- NOTE ----------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.15 -------------------- NOTE ----------------
Normalization of the AT is not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after THERMAL POWER is
Ž 98% RTP.
Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.16 Perform COT. 24 months SR 3.3.1.17 Perform TADOT. 24 months Cook Nuclear Plant Unit 2 3.3.1-9 Amendment No. 2-&g,
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)
Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.077% of AT span.
{AO K (1+71S) [T-T']+K3 (P-P')-f(AI)
AT LTT (1+-C2S)
G Where: AT is measured RCS AT, F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec1 .
T is the measured RCS average temperature,0 F.
T"is the nominal Tavg at RTP, < [*]OF.
P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, > [*] psig K1 -<[*] K2 ->[*]/OF K3 > [*]/psig C,> [*]sec T25 [*] sec fl(Al) = [*] {[*] + (qt - qb)} when qt - qb < -[*]% RTP 0% of RTP when -[*]% RTP < qt- qb < [*]% RTP
-[*] {(qt - qb) - [*]} when qt" qb > [*]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 2 3.3.1-15 Amendment No. 2-&9,
RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)
Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal Trip Setpoint by more than 0.140% of AT span.
AT <ATo K4-K5IT3S T-K 6 [T -T"]-f 2 (AI)}
Where: AT is measured RCS AT,°F.
ATo is the indicated AT at RTP,°F.
S is the Laplace transform operator, sec-1.
T is the measured RCS average temperature,0 F.
T" is the nominal Tavg at RTP, < [*]OF.
K4 < [*] K5 > [*]/OF for increasing Tavg K6 > [*]/OF when T > T"
[*]/OF for decreasing Tavg [*]/OF when T _<T" T3 > [*] sec f2 (AI) = [*1
- These values denoted with [*] are specified in the COLR.
Cook Nuclear Plant Unit 2 3.3.1-16 Amendment No. 2-99,