NLS2010088, 10 CFR 50.59(d)(2) Summary Report of Evaluations Performed from 08/01/2008 to 07/31/2010
| ML102860134 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 10/07/2010 |
| From: | Vanderkamp D Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2010088 | |
| Download: ML102860134 (14) | |
Text
H Nebraska Public Power District Always there when you need us NLS2010088 50.59(d)(2)
October 7, 2010 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
10 CFR 50.59(d)(2) Summary Report Cooper Nuclear Station, Docket No. 50-298, DPR-46
Dear Sir or Madam:
The purpose of this letter is for the Nebraska Public Power District to provide the summary report of evaluations that have been performed for Cooper Nuclear Station, in accordance with the requirements of 10 CFR 50.59(d)(2). This report covers the time period from August 1, 2008, to July 31, 2010. Summaries of applicable facility changes are discussed in Attachment 1.
Summaries of other changes are discussed in Attachment 2. There were no changes to procedures implemented during this reporting period under the provisions of 10 CFR 50.59.
Should you have any questions concerning this matter, please contact me at (402) 825-2904.
Sincerely,
!David W. Van Der"]
Licensing Manager
/dm Attachments cc: Regional Administrator, w/attachments NPG Distribution, w/o attachments USNRC - Region IV Cooper Project Manager, w/attachments CNS Records, w/attachments USNRC - NRR Project Directorate IV- 1 Senior Resident Inspector, w/attachments USNRC - CNS COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 wwV.rnppd.Com
NLS2010088 Page 1 of 10 ATTACHMENT 1 FACILITY CHANGES Change Evaluation Document (CED) 6013140 (Including Change Notices 1 through 13)
(Evaluation 2008-009, Revision 0)
Title:
Service Air Compressor Replacement
==
Description:==
This modification replaced the station air compressors which supply air to the Plant Air system, including the Service Air and Instrument Air systems. This modification included the replacement of the reciprocation air compressor with new compressors and the upgrade of the controls. An alternate air source connection has also been provided to allow the connection of a portable air compressor to supply the Plant Air System. The replacement was necessary due to the old compressors nearing the end of the service life, exhibiting decreasing reliability, and requiring frequent overhauls.
The Service Air compressors were previously controlled by a pneumatic control system. This pneumatic control system was replaced by a digital control system.
The replacement compressors are maintained in the Compressor Control Mode (CCM) Local. Because of this, the Multiple Compressor Control feature (running all three compressors via one unit) was not evaluated.
10 CFR 50.59 Evaluation: Updated Safety Analysis Report (USAR)Section XIV and Appendix G do not identify the loss of Plant Air system as an accident initiator nor is plant air required to mitigate the consequences of an accident. The air compressors do not directly contribute to and are not required for timely response to accident prevention or mitigation. In the CCM Local mode of operation, the air compressors are running independent of each other which will prevent a "single point of vulnerability". The human system interface failure modes of the new system do not introduce/increase any radiological nor malfunctions of systems, structures, or components (SSCs) important to safety. Installation and digital control of the air compressors does not create any new events that can initiate accidents that are of a different type than those evaluated in the USAR. The system-level failure modes identified during the performance of the failure mode and effects analysis (FMEA) were no different for the new digital control than the pneumatic control. The Service Air compressors do not have an affect on Design Basis functions for any systems in supporting or determining the integrity of the fission product barriers. Based on the responses, prior Nuclear Regulatory Commission (NRC) approval was not required.
NLS2010088 Page 2 of 10 CED 6016559 (Including Change Notices 1 through 7)
(Evaluation 2008-014, Revision 0)
Title:
Digital Electro-Hydraulic (DEH) Control System and Main Turbine Trip Block Replacement
==
Description:==
This modification replaced the existing DEH turbine control system and main turbine trip block with Triconex hardware. The modification improves the availability and reliability of the DEH Control System and turbine trip functions in response to spare part shortages, obsolescent equipment, aging, excessive system maintenance requirements, and single point control system failures that have resulted in plant trips or power reductions and delays in plant startups from refueling outages and downpower activities.
The modification involved installation of a new cabinet (Trip Tricon system) on the turbine floor outside the shield wall and modification of the existing control cabinet located in the computer room (Control Tricon system). The modification replaces aging obsolete analog/digital components, and old mechanical trip devices with new digital Triconex hardware. The new system integrates operator control functions, and system information on a flat panel displays serving as the operator communication to the control and trip systems as Human Machine Interface (HMI). The digital upgrade provides operator enhancements and improvements in turbine control and overspeed protection and resolves system reliability and single point failure vulnerabilities. The modification installed new Triconex digital controllers for turbine speed/load and throttle header pressure control, including overspeed protection and protective functions. -Two HMI's have been installed in the main control room for operator interface, which replaced the DEH control panels.
10 CFR 50.59 Evaluation: This evaluation has determined that the proposed replacement of the main turbine speed/load and throttle header pressure control hardware (DEH) including the main turbine mechanical trip block devices and replacement of the mechanical overspeed trip device with integrated Tricon controllers will improve the availability and reliability of the turbine protection and throttle pressure control functions. The evaluation concludes that the upgrade will have no effect on any accidents or malfunctions previously evaluated in the USAR. It also concludes that the upgrade effort will have no potential for the creation of an event of a type not previously evaluated in the USAR, nor will the modification result in the introduction of new failure modes not previously evaluated in the USAR.
NLS2010088 Page 3 of 10 The FMEA for the replacement DEH Control and Turbine Trip systems demonstrates that no new accidents, accident initiators, or modes of operation are created; the frequency of occurrence of an accident or likelihood of an equipment malfunction is not increased; and that the sequences associated with the failure of the replacement control and trip systems are the same as the replaced DEH control system and main turbine trip block assembly. Additionally, neither DEH control system nor main turbine trip functions are credited with mitigating the consequences of an accident. Therefore the frequency of occurrence and consequence of an accident previously evaluated is not increased; and the likelihood or consequence of a malfunction of an SSC important to safety is not increased.
The FMEA for the replacement DEH Control and Turbine Trip systems also demonstrates that no new accidents of a different type, or malfunctions of an SSC important to safety with a different result are created, as no new modes of operation, accident initiators, accident consequences or equipment failures are created.
The replacement DEH Control and Turbine Trip systems do not interface or affect reactor coolant or fission product barriers, or change design criteria used in the determination of fission product barrier integrity. Therefore the design basis limits for fission product barriers are not affected.
Based on the above evaluation, prior NRC approval was not required to perform this modification.
CED 6024785 (Including Change Notices 1 through 6)
(Evaluation 2008-010, Revision 0)
Title:
Augmented Off-gas (AOG) System Controls Upgrade
==
Description:==
This modification installed new Siemens-Moore 353 type digital controllers in place of the obsolete Honeywell Vutronik analog controllers at various locations throughout both trains (A & B) of the AOG system, along with the addition of two new controllers. The new controllers have been configured to perform the same functions as the existing controllers in addition to integrating some additional functions.
The new controllers assumed signal processing functions previously performed by the stand alone device such as square rooters. The power input to the replacement controllers utilizes 120 VAC instead of the existing 24 VDC to separate the control functions from the indication-only functions and minimize power supply single point failures.
The new controllers are digital as opposed to the analog controllers. This is considered a different way of controlling the AOG system and thus required evaluation under 50.59.
NLS2010088 Page 4 of 10 10 CFR 50.59 Evaluation: There are no accidents as described by the USAR that are directly or indirectly impacted by this modification. The AOG system is not discussed as an accident initiator, either in normal operation or in response to any accident described in the USAR, through its design functions or its failure to perform those design functions.
The modification to the AOG system does not introduce the possibility of a change in the consequences of an accident because there are no failure modes resulting in a different effect from the current configuration. No new failure modes are introduced, nor will the modifications to the AOG system affect any other SSC important to safety.
In conclusion, the modification of the AOG system, as described above and as part of this evaluation, cannot cause an accident, introduce the possibility of a change in the consequences of an accident, introduce new failure modes due to their failure, nor introduce any new accident or scenario not already bounded by the safety analysis. Therefore, prior NRC approval is not required for the implementation of this modification.
CED 6026860 (Including Change Notices 1 through 10)
(Evaluation 2009-016, Revision 0)
Title:
Main Turbine Lube Oil Conditioner System Upgrade
==
Description:==
This modification replaced the turbine lube oil conditioner system with a new skid mounted turbine lube oil conditioner that provides improved particulate and water filtration capabilities. This modification also replaced the CUNO brand Hydrogen Seal Oil Filters with a new duplex design that provides improved filtration capabilities and allows on-line filter changes. Procedures were updated to utilize the Dirty Oil Storage Tank and transfer in combination with the new turbine lube oil conditioner.
A 50.59 evaluation was necessary due to a change in how the design functions are performed or controlled for the Turbine Lube Oil Conditioner system.
Specifically, the changes which required further evaluation were the digital controls (PLC) and the trip function of the conditioner from the DEH system.
10 CFR 50.59 Evaluation: Use of the digital upgrade and trip signal will not change, prevent or degrade the effectiveness of actions described or assumed in any accident discussed in the USAR. The Turbine Lube Oil Conditioner does not play a role in mitigating the radiological consequences of any accidents. The PLC and any new failures within the control logic remain bounding by the failure of the Turbine Lube Oil Conditioner itself as no new accidents would be created.
NLS2010088 Page 5 of 10 The Turbine Lube Oil Conditioner and DEH systems do not directly interface with the fuel cladding, the reactor coolant system pressure boundary or containment. No USAR described calculated values are affected or changed nor were any elements of analysis methodology changed.
Therefore, NRC approval was not required prior to implementing this change.
CED 6030461 (Including Change Notices 1 through 3)
(Evaluation 2010-013, Revision 0)
Title:
Diesel Generator Fuel Oil Day Tank Piping
==
Description:==
Plugging of the Diesel Generator (DG) day tank float admission valves has caused past reliability problems and low flows to the day tank. This modification improved the reliability of the Diesel Generator Diesel Oil (DGDO) system by reducing the potential for flow blockage in the day tank level controls, flow measurement instrumentation and attendant piping. The change removed the float admission valve DGDO-FOV-FLTV10 and DGDO-FOV-FLTV1 1, on each diesel fuel oil day tank fill line. The day tank connections where the float admission valves have been attached have been capped. New piping discharges directly to the fuel safety solenoid inlet valve at each day tank. The pipe spool and the upstream flow indicating transmitter for each fuel oil day tank fill line has been removed and replaced with an orifice plate and pressure transmitter. Instrument power for the new flow indicator is connected to a permanent source of 120 VAC.
Gasket materials for the spool flanges have been selected for compatibility with diesel fuel, to preclude historical problems associated with elastomer degradation and debris plugging in the downstream float admission valves. Piping integrity is assured by use of maximum pump flow capability in lieu of using total flow and tank inventory change.
Due to the removal of the float valves, which were automatic control elements, the mechanical shutoff function of the float valve was also deleted. Solenoid valves DGDO-SOV-SSV5028 (5029) are the final electro-mechanical control elements and the shutoff pressure boundary. This portion of the modification activity required further evaluation under 10 CFR 50.59.
10 CFR 50.59 Evaluation: This modification re-routed the DG Day Tank Fuel Oil Transfer system piping, upgraded the instrumentation, and deleted the day tank float admission valves (FAVs), which were automatic control elements and which also performed a backup, mechanical shutoff function to the pump stop switches. The day tank fuel safety solenoid inlet valves replaced the FAVs as the automatic control elements. These changes do not adversely affect the design functions of the diesel fuel oil transfer system. There is no new failure mechanism introduced by these activities. The modifications identified by the configuration change document have been examined relative to the design requirements, design basis and design
NLS2010088 Page 6 of 10 functions. Reliability of the DG day tank fuel transfer system is increased, because components susceptible to plugging are being removed. This system supports operation of the DGs, which are not accident initiators; hence, there is no increase in accident frequency of occurrence. No assumptions previously made in evaluating the radiological consequences of an accident described in the USAR are altered as a result of the modifications. There is no increase in the radiological consequences of a malfunction of the day tank fuel transfer system.
The modifications do not create any new events that can initiate accidents of a different type than those evaluated in the USAR. The modifications do not result in a design basis limit for a fission product barrier (DBLFPB) as described in the USAR being exceeded or altered. No USAR described calculated values are affected or changed, nor were any elements of the analytical methodology changed. Therefore, NRC approval was not required prior to implementing this change.
Engineering Evaluation (EE)08-026, Revision 0 (Evaluation 2008-015, Revision 0)
Title:
Re-configure DGDO-V- 19 from Open to Closed
==
Description:==
This activity changed the position of the DG day tanks 1 and 2 inlet tie valve, DGDO-V- 19, from a normally "open" configuration to a normally "closed" configuration. This was pursued as a result of failure of the float valve and the solenoid valve on the day tank to isolate flow from the opposite division transfer pump. This condition was documented in the Cooper Nuclear Station corrective action program. The failure of both valves could have resulted in a loss of fuel oil inventory. Changing the valve to a closed position separated the system into two independent divisions. Each division would be capable of meeting the design function for the transfer system and would not be affected by failures in the other train. The change of valve position enhances system operation and increases the overall reliability of the system.
10 CFR 50.59 Evaluation: The change in position is not an accident initiator and would not cause an increase in the frequency of an accident. Nor does this change affect or increase the consequences that would result from a malfunction or postulated accident. The fuel oil transfer system has no impact on fission product barriers and no analysis was altered as a result of this evaluation. Changing of the valve position enhances the reliability of the system and can be initiated without prior NRC approval.
NLS2010088 Page 7 of 10 EE 09-007, Revision 0 (Evaluation 2009-020, Revision 0)
Title:
Refueling Mast Modification Evaluation
==
Description:==
The Independent Spent Fuel Storage Installation design utilizes a cask whose length is longer than that used in the original station design. To allow for the extra height required to clear the top of the longer cask, the spare refueling mast was sent to AREVA for modification, where it was shortened.
The EE assessed use of the shortened mast to temporarily replace the normal mast during cask loading and unloading operations, and to restore the normal mast when cask loading operations have been completed.
10 CFR 50.59 Evaluation: The applicable accident is the fuel handling accident (FHA) over the spent fuel pool. The activity involved changes in the length of the cask loading mast and maximum lift height, in support of cask loading and unloading operations. The mast was modified in conformance with station design requirements. Changes in the mast or lift height is not an accident initiator. The accident frequency is not changed. The increase in lift height has no impact on the likelihood of dropping a fuel bundle, nor does it threaten criticality or rack, slab, and liner structural integrity beyond the FHA design basis accident (DBA). Therefore, there is no increase in the likelihood of equipment failure, nor is there is a possible increase in consequences of an accident or equipment malfunction. The activity does not introduce any new failure modes. The activity does not introduce possibility of a new or different type accident, nor does it introduce the possibility for a malfunction of an SSC with a different result.
The activity does not affect any system, performance, or response parameter; therefore, it can not result in a DBLFPB being exceeded or even altered. The activity is consistent with the methodology described in the USAR and in NRC acceptance criteria from Amendment 227; therefore, it does not depart from a method of evaluation described in the USAR. NRC approval was not required prior to implementing this change.
Temporary Configuration Control (TCC) 4686902, Revision 0 (Evaluation 2009-017, Revision 1)
Title:
DEH Bypass Valve Solenoid Bypass Jumper
==
Description:==
This TCC provided the operator with a manual switch to ensure the existing system design function of the bypass valves fast permissive below 25% reactor thermal power is enforced to allow bypass valves to control reactor pressure using DEH pressure controllers as designed. To accomplish this, the TCC installed an electrical switch across pins 30 and 34 on the O1K4L relay card in DEH cabinet 01 to jumper out the fast open logic and allow the MG-6 (K2) relay to stay
NLS2010088 Page 8 of 10 energized during turbine startup. This was a defense in depth action to avoid plant transients due to DEH malfunctions as a result of age related failures and single failure vulnerability.
The proposed jumper and switch have two possible failure modes: open circuit failure and closed circuit failure.
Open circuit failure could be postulated as a switch failure (fail to close or spuriously open) and a jumper failure (broken wire or loss of termination). An open circuit failure of the new switch would leave the DEH system in its originally designed condition. Unless there was a failure of another piece of equipment, the system would function as designed (same as not having the new switch installed).
Closed circuit failure could be postulated as a switch failure (fail to open or spuriously close). This would cause the Operator to be informed via annunciator window and would defeat the fast open permissive which is the same failure mode as the existing logic.
10 CFR 50.59 Evaluation: Since the proposed switch/jumper is defense in depth only and does not introduce new failure mode effects beyond those previously evaluated in the USAR, the proposed activity does not result in an increase in the frequency of occurrence of an accident previously evaluated in the USAR. The new jumper being installed will not initiate any new malfunctions. The DEH system is neither safety-related nor does this activity adversely affect systems important to safety. Therefore, there will be no increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the USAR.
The DEH system is not used to mitigate the consequences of any accidents and the new jumper will not initiate any new accidents. This modification will not impair or prevent emergency core cooling systems from mitigating the consequences of any design basis accident. Therefore, this activity does not increase the consequences of occurrence of an accident previously evaluated in the USAR. Failure or malfunction of the switch/jumper will not prevent or affect the ability of safety-related systems or systems important to safety to respond to the accidents described in the USAR. Therefore, consequences of a malfunction of an SSC important to safety previously evaluated in the USAR will not be increased.
Based on the failure modes analyzed, the potential switch/jumper malfunctions do not create any new DEH responses beyond those already bounded by USAR transient analysis. Therefore, the possibility of an unanalyzed malfunction of an SSC important to safety or a new type of accident is not created. As described in the USAR Chapter XIV transient analysis, no malfunction of the DEH system can cause a transient sufficient to damage the fuel barrier or exceed the nuclear
NLS2010088 Page 9 of 10 system pressure limits as required by the safety design basis. Therefore, the TCC does not result in the design basis limits for fission product barriers being exceeded or altered. The TCC does not result in a departure from a method of evaluation described in the USAR in establishing the design bases or in the safety analysis. Therefore, NRC approval was not required prior to implementing this TCC.
TCC 4699557, Revision 0 (Including Temporary Change Notice 1)
(Evaluation 2009-019, Revision 0)
Title:
DG Mechanical Overspeed Cable Removal
==
Description:==
A TCC has been implemented to install gagging (locking) devices on the DG1 and DG2 air inlet butterfly valves in order to maintain them in the open position.
The TCC also removed the operating cables which connect the actuator to the butterfly valve trip levers. The cables were susceptible to binding which not only affected proper operation of the butterfly valves but also affected the pneumatic fuel rack shutoff of the mechanical overspeed trip. This TCC was required to improve DG reliability by eliminating this failure mechanism associated with the cables.
10 CFR 50.59 Evaluation: Since the DGs are not an initiator of any of the abnormal operating transients or postulated accidents described in the USAR, this temporary configuration change does not increase the frequency of occurrence of an accident previously evaluated in the USAR. With the TCC installed, the DGs retained the safety shutdown features and emergency operation/functions as specified in the USAR.
Implementation of the TCC does not result in a more than minimal increase in the likelihood of a malfunction of an SSC important to safety previously evaluated in the USAR.
This modification does not adversely affect the ability of the DGs to provide emergency power to systems used for accident mitigation, and therefore does not increase the consequences of an accident previously evaluated in the USAR. The temporary configuration change will not change the consequences of a failure of a DG on any other safety system as stated in the USAR and does not increase the consequence of a malfunction of an SSC important to safety.
Changes to the DGs cannot create the possibility of an accident of a different type (i.e., the DGs are a mitigation system) and as such there is no increase in the possibility of an accident of a different type than any previously evaluated in the USAR. The USAR assumes that only one onsite DG is available during the entire DBA loss of coolant accident, therefore the possibility of a malfunction of an SSC important to safety with a different result than any previously evaluated in the USAR will not be created.
NLS2010088 Page 10 of 10 The changes incorporated by this TCC do not affect the accident analysis for the release of radioactive material and does not affect any of the radioactive material barriers. The proposed activity does not result in design basis limit for fission product barrier as described in the USAR being exceeded or altered. This TCC does not result in a departure from a method of evaluation described in the USAR used in establishing the design bases or in the safety analysis.
The conclusion reached by this safety evaluation is that locking open the DG air inlet butterfly valves and removing the operating cables per TCC 4699557 did not require NRC approval prior to implementation.
NLS2010088 Page 1 of 2 ATTACHMENT 2 OTHER CHANGES License Basis Document Change Request 2007-018 (Evaluation 2008-006, Revision 0)
Title:
Change to Technical Requirements Manual (TRM) 3.7.1 and TRM Bases 3.7.1, Updated Safety Analysis Report (USAR) Section 11-4.2, and Procedure 5.1FLOOD
==
Description:==
The USAR, TRM 3.7.1, and Procedure 5.1FLOOD were revised to allow continued plant operation until within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of the time when river level is forecast to reach 902 feet elevation. This change was necessary to prevent unnecessary power reductions or plant shutdown due to potential inaccuracies in river level forecasts. The standard period allowed is 12 and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> in the Technical Specifications and TRM to reach MODE 3 and MODE 4 with the plant operating in MODE 1, and provides adequate time to reach cold shutdown without significantly challenging plant safety or the operating staff.
10 CFR 50.59 Evaluation: As described in the USAR, Service Water (SW) and Circulating Water (CW) pumps provide cooling to required equipment. The Ultimate Heat Sink (UHS) provides a source to dissipate the heat absorbed. The intake structure provides support and protection to the SW and CW pumps. The ability of the SW system and UHS to provide adequate cooling is assumed in the safety analyses. Per the USAR and TRM (prior to revision), the site flooding procedure was initiated at river elevation 895 feet and plant shutdown is initiated when floodwaters reach 902 feet or forecast to reach 902 feet (as from the 10,000 year flood or an upstream dam failure).
The critical event in the "sequence" is a failure of a large upstream dam (either Oahe or Fort Randall), and would take at least three days to reach Cooper Nuclear Station (CNS). Additionally, the USAR describes the combination of a flood with wind and the potential wave action in addition to the resultant water level, and a combined flood and dam failure. The potential wave action extends to the impact on the power block. These analyses conclude that equipment important to safety will be adequately protected against flooding given a design basis flood with concurrent wind action.
The levee surrounding CNS is at elevation 902 feet, above which floodwaters can begin to affect operation of the switchyard, although plant structures and equipment would not be affected until level is above elevation 903 feet, which is nominal plant grade elevation. Prior to the change, the USAR and TRM required commencement of a plant shutdown upon receipt of a forecast that river level will exceed 902 feet elevation. Changing the timing of a plant shutdown did not
NLS2010088 Page 2 of 2 impact the response of the plant structures and systems to a flooding event. The 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> limit is an adequate time to implement a plant shutdown to MODE 4 prior to river level going above elevation 902 feet. The timing of the installation of flood protection barriers in the flooding procedure has not been changed.
High river level is not an initiator of an accident, and required equipment will remain functional until MODE 4 is reached, thus this change does not affect the frequency of occurrence of an accident, nor the likelihood of malfunction of equipment required to mitigate an accident. Since the plant will be placed in MODE 4 prior to the river reaching elevation 902 feet, the consequences of an accident is not increased. All required equipment is well above elevation 902 feet; thus this change does not change the likelihood of a malfunction of required equipment.
Since river level will remain below elevation 902 feet until MODE 4 is reached, the possibility of a different type accident is not created. Similarly, the possibility of a malfunction with a different result is not created, as all required equipment is located well above elevation 902 feet.
The fission product barriers discussed in the Resource Manual are fuel clad, reactor coolant pressure boundary, and containment. The limits for these barriers will not be affected, as the ultimate heat sink is not adversely challenged by this change, since required equipment will remain available until the plant is in MODE 4.
Lastly, this change is consistent with and does not revise any of the methods of evaluation used to establish design bases or used in the safety analyses. The change is consistent with the analysis in the USAR which established the limit for river elevation (902 feet) which requires a plant shutdown, and adequate time is allowed to place the plant in MODE 4 upon notification of these conditions.
This evaluation concludes that implementation of the above change to allow operation of the plant to continue until within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of water level being predicted to go above 902 feet does not adversely affect the ability of the plant to survive a projected design basis flood as described in the USAR. Therefore, Nuclear Regulatory Commission approval was not required prior to implementing this change.
4 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS©O Correspondence Number: NLS2010088 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None N/A N/A 4 .1-4 +
4 +
4 4 4 4 +
4 +
I PROCEDURE 0.42 REVISION 26 PAGE 19 OF 26 1