ML061020058
| ML061020058 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/03/2006 |
| From: | Jamil D Duke Energy Carolinas, Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML061020058 (8) | |
Text
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IDu;re D.M. JAMIL Vice President PIPowere A Duke Energy Company Duke Power Catawba Nuclear Sta'ion 4800 Concord Rd. / CNOI VP York, SC 29745-9635 803 831 4251 803 831 3221 fax April 3, 2006 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
SUBJECT:
Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC Catawba Nuclear Station, Unit 1 and Unit 2 Docket Numbers 50-413 and 50-414 2005 10 CFR 50.59 Report Attached please find a report containing a brief description of changes, test, and experiments, including a summary of the safety evaluation for each, for Catawba Nuclear Station, Units 1 and 2 for the year 2005. This report is submitted pursuant the provisions of 10 CFR 50.59(d)(2) and 10 CFR 50.4.
Questions regarding this report should be directed to Kay Nicholson at 8C3.831.3237.
Sincerely, D. M. Jamil Attachments
.S-I www. dukepower. corn
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Nuclear Regulatory Commission April 3, 2006 Page 2 xc:
W. D. Travers, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Sam Nunn Atlanta Federal Center 23T85 61 Fcrsyth St., SW Atlanta, GA 30303 J. F. Stang, Jr. (Addressee only)
U. S. Nuclear Regulatory Commission Mail Stop 0-8 H4A Washington, DC 20555-0001 E. F. Guthrie NRC Senior Resident Inspector (CNS)
CNOlNC
9 I i U. S. Nuclear Regulatory Commission April 3, 2006 Page I cf 6 Type: Miscellaneous Items Unit: 0
Title:
Revision of Physics Tests Exception Bases (B 3.1.8) to clarify use of ADRC on power range channel Description A clarification is being added to Technical Specification Bases B 3.1.8, Surveillance Requirement (SR) 3.1.8.1 identifying that the reactivity computer (ADRC) is connected to the signal output of one of the four power range detectors. The associated power range channel (drawer) is placed in the trip condition. This action is necessary to satisfy the requirements of UFSAR 14.3, Startup Physics Test Program. A power range detector, with its upper and lower chambers, provides the most core coverage and is utilized to provide the flux input to the ADRC. Since the detector signal is provided to the computer, it can not supply the drawer (channel). Thus, the channel is removed from service and placed in the trip condition (as allowed by Technical Specification Evaluation The proposed activity is a clarification of the Technical Specification Bases 3.1.8 as it relates to SR 3.1.8.1. The condition being clarified is the removal of one Power Range channel from service to allow the output of the associated Power Range detector to be connected to the ADRC.
UFSAR Section 14.3, Startup Physic Test Program, identifies that the ADRC is connected to an external signal which is proportional to the core neutron flux. A Power Range detector, with its upper and lower changes, provided the most core coverage and is utilized to provide the flux input to the ADRC. By using the output of one Power Range detector for the ADRC input, the associated Power Range channel (drawer) becomes unable to perform its trip function. Therefore it is placed in the Trip condition.
Technical Specification 3.3.1 allows one Power Range channel to be placed in Trip condition with power limited to less than or equal to 75 percent.
The conclusion of this evaluation is that the proposed activity does not require approval of the NRC prior to implementation.
a L4 U. S. Nuclear Regulatory Commission April 3, 2006 Page 2 of 6 Type: Nuclear Station Modification Unit: 2
Title:
C2C14 Reload Safety Description In order to ensure the C2C14 reload core needs no prior NRC review and approval, a 10 CFR 50.59 evaluation is performed in accordance with approved administrative procedure, NSD 209. This evaluation determines if a license amendment request (LAR) is required as a result of changes in the physics parameters predicted for this reload.
UFSAR changes identified as a part of C2C14 are evaluated herein and included as part of the reload calculation file CNC-1552.08-00-0350. Station modifications, changes in test, or changes in procedures during the refueling outage must be addressed in separate evaluations.
Evaluation A 10 CFR 50.59 evaluation is performed for the Catawba Nuclear Station Unit 2, Cycle 14 (C2C14) core reload in calculation file CNC-1552.08-00-0350. The impact of any other plant changes made concurrent with the refueling outage is not addressed in this evaluation.
The C2C14 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Nuclear Engineering Division workplace procedure NE-102, Workplace Procedure for Nuclear Fuel Management, and the C2C14 Reload Safety Evaluation, confirm the UFSAR Chapter 15 accident analyses remain bounding with respect to the C2C14 safety analysis reactor physics parameters. The safety analysis reactor physics parameters method is described in topical report DPC-NE-3001 -PA.
The C2C 14 core reload is similar to past cycle core designs, with a design generated using NRC-approved methods. The C2C14 Core Operating Limits Report (COLR) is prepared in accordance with Technical Specification 5.6.5 and submitted to the NRC in accordance with 10 CFR 50.4. Additionally, applicable sections of Technical Specifications and the UFSAR have been reviewed. No changes to Technical Specifications are necessary for C2C14. UFSAR changes for C2C14 are evaluated and documented as part of the C2C14 Reload Safety Evaluation. The UFSAR changes are necessary to reflect changes made to the COLR for the heat flux hot channel factor.
U. S. Nuclear Regulatory Commission April 3, 2006 Page 3 cf 6 Type: Procedure Change Unit: 0
Title:
Procedure Change AP/O/A/5500/045, Plant Fire, Revision 0 Description AP/0/A/5500/045, Plant Fire, is being created to address a problem investigation report (PIP C03-3258). For areas that credit the SSF for safe shutdown capability, a partial transfer to SSF control shall be initiated when an active fire is detected in said area.
This partial transfer to the SSF is done by placing EMXS on its alternate power.
Complete swapping to the SSF would then be directed by AP/I (2)/A/5500/017, Loss of Control Room.
Furthermore, this new procedure will direct the operators to close the suction of the train related ND pump from the FWST to prevent loss of the FWST inventory from a spurious opening of the train associated containment sump isolation valve. This action is required for a fire in a given fire area which would affect one train of ND.
Evaluation This evaluation concluded that neither UFSAR changes nor Technical Specification changes are required for the activities of swapping EMXS to alternate power given a fire in the SSF fire area or for the closing of the suction valves of the ND pump on the train associated with a fire. Both of these activities are done to ensure that the plant maintains the ability to perform a safe shutdown using A train, B train, or the SSF.
This 10 CFR 50.59 was revised on November 14,2005 in order to incorporate comments from the Nuclear Safety Review Board (NSRB). The NSRB requested that question 4 be revised to state what the required time frame was for completing the task of swapping EMXS and why there is reasonable confidence that this task can be achieved within the given time frame.
U. S. Nuclear Regulatory Commission April 3,2006 Page 4 (of 6 Type: Temporary Design Change Unit: 2
Title:
Temporary Design Change CD1200384, Disable Lower Containment Ventilation Unit 2D Low-Speed Operation Description The motor lower containment ventilation unit 2D (LCVD-2D) has developed a ground on the low-speed windings. However, the high-speed windings are good and the ground appears to have no affect on high-speed operation. The temporary design change disables the low-speed controls for LCVU-2D since a replacement motor is not readily available, and allows operation of LCVU-2D in high speed if needed. This provides additional flexibility in maintaining containment temperatures. The temporary design change will be removed when the motor is replaced.
Evaluation Temporary Design Change CD200384 disables the low-speed controls for Containment Ventilation unit 2D (LCVU-2D), which is part of the Containment Ventilation (VV)
System. The low-speed controls will be disabled due to an electrical ground on the low-speed windings. Disabling the low-speed controls allows operation of LCVD-2D in high speed without tripping the associated breaker when cycling from OFF to HIGH. This provides additional flexibility in maintaining containment temperature until the motor is replaced. The other three LCVD fans can still operate in either low or high speed.
Normally, three fans operating in low speed maintain Lower Containment within the required Technical Specification temperature limits.
Per the UFSAR 9.4.6, the VV System "is designed to maintain acceptable temperature limits within the confines of the reactor building upper and lower compartments to ensure proper operation of equipment and controls during normal plant operation and normal shutdown. The Containment Lower Compartment Ventilation Subsystem is designed to maintain a maximum temperature of 120 degrees F in the lower compartment during normal plant operation. The Containment Ventilation System is energized from the Blackout Power System upon loss of offsite power. The Containment Ventilation System is not considered an engineered safety feature and no credit has been taken for the operation of any subsystem or component in analyzing the consequences of any accident." Catawba Technical Specification Limiting Condition of Operation (LCO) for Containment Air Temperature states, "Containment average air temperature shall be greater than or equal to 100 degrees F and less than or equal to 120 degrees F for the containment lower compartment" during Modes I through 4. The associated Bases states, "The containment average air temperature is limited, during normal operation, to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). This LCO ensures that initial conditions assumed in the analysis of containment response to a DBA are not violated during unit operations."
The VV System is not an accident or malfunction initiator. Several layers of electrical protection exist to isolate the fault such that the fault does not affect other equipment or systems. Applicable accidents/events related to this design change are a Loss-of-Offsite Power (LOOP), a train-related Blackout (B/O), or a Station Blackout (SBO).
With low-speed operation disabled for LCVU-2D, a LOOP event will result in up to three LCVUs starting on low speed, which is the same as during normal station operation.
Furthermore, the VV System is not credited in the safety analysis in mitigating the consequences of a LOOP. For an SBO, no fans operate, and previous analyses for an SBO reveal that Containment temperature remains below the peak temperature limit.
The worst-case scenario of an SBO bounds the consequences of a train-related B/O since a B/O has some fans operating compared to none during an 580. Operating with LCVU-2D in high-speed could result in a slight increase in Lower Containment temperature. This increase, however, remains within the 120 degrees F Technical
U. S. Nuclear Regulatory Commission April 3,2006 Page 5 (if 6 Specification limit, or remains within the additional margin assumed if the EQ service life was base on a temperature lower the 120 degrees F limit. Therefore, the frequency of any accidents or the likelihood of SSC malfunctions evaluated in the UFSAR does not increase by the temporary design change. Furthermore, the consequences of an accident or a malfunction of an SSC previously evaluated in the UFSAR is not increased.
Since the VV System is not an accident initiator and the existing fault is bounded by existing analyses, and since EQ Service Life effects on Containment SSCs is not impacted, the temporary design change does not create the possibility for an accident of a different type than previously evaluated in the UFSAR. The only failure mode associated with the faulty LCVU-2D low-speed winding is the propagation of this failure onto the high-speed windings, and electrical protection exists to isolate the fault should this occur. Furthermore, the LCVU fans and the VV System are not credited in operating during Design Basis Events. Therefore, the temporary design change does not create the possibility for a malfunction of an SSC important to safety with a different result than those previously evaluated in the UFSAR.
An SBO is the worst-case scenario from a VV System perspective on Containment temperature since no LCVU fans are operating. Containment temperature remains below the calculated peak temperature during such as event. Therefore, the integrity of Containment is not affected. In addition, maintaining temperature with the Technical Specification limits during normal operation ensures acceptable response during others Design Basis Events. Therefore, the temporary design change does not exceed or alter a design basis limit for a fission product barrier.
This temporary design changes does not involve a method of evaluation as described in the UFSAR. Therefore, the temporary design change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases, or used in the safety analyses.
The presentation in UFSAR Table 8-6 might indicate that the temporary design change may be an adverse effect to an UFSAR design function by defeating low speed operation of LCVU-2D.
However, the remaining high speed capability and the lack of reliance on the LCVUs for mitigation of Design Basis Accidents allows for the conclusion of "NO LAR" required. No UFSAR changes will be submitted since this is a temporary change and will be restored to as described in the UFSAR. Furthermore, no Tech Spec or SLC changes are required.
4, U. S. Nuclear Regulatory Commission April 3,2006 Page 6 (if 6 Type: UFSAR Revision Unit: I
Title:
UFSAR Chapter 15.4.1 and 15.4.2 Analyses Update Description In order to ensure that UFSAR Chapter 15 changes need no prior NRC review and approval, a 10 CFR 50.59 evaluation is performed in accordance with approved administrative procedure, NSD 209. The UFSAR chapter 15.4.1 and 15.4.2 accident analyses have been updated to incorporate a revision to the way the moderator temperature coefficient (MTC) is treated and, for the Chapter 15.4.1 analysis, a reduction in reactor coolant system (RCS) flow rate to 388,000 gpm, a change in full power average full temperature and a change to the treatment of the re-insertion of the withdrawn rod worth. This evaluation determines if a license amendment request (LAR) is required as a result of UFSAR changes required following the reanalysis of Chapter 15.4.1 and 15.4.2 accident analyses. Revision of the UFSAR Chapter 15.4.1 and 15.4.2 analysis text and figures related to reanalysis are covered in this evaluation.
Evaluation A 10 CFR 50.59 evaluation has been performed for updated text, tables, and figures for the Catawba Nuclear Station UFSAR Chapters 15.4.1 and 15.4.2. These changes are required following an update to the Unit I analyses.
This 10 CFR 50.59 evaluation supports the update to the Catawba UFSAR Chapter 15 analyses for both the Uncontrolled Bank Withdrawal at Power (Chapter 15.4.2) and Uncontrolled Bank Withdrawal at Zero Power (Chapter 15.4.1) accidents with a revision to the way the moderator temperature coefficient (MTC) is treated and, for the Chapter 15.4.1 analysis, a reduction in RCS flow rate to 388,000 gpm, a change in the full power average fuel temperature, and a change in the treatment of the re-insertion of the withdrawn rod worth. The reanalysis are performed similar to past analyses using methods previously reviewed and approved by the NRC.
It has been concluded in the Catawba 10 CFR 50.59 evaluation that no prior NRC approval is necessary.