ML050970285
| ML050970285 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/31/2005 |
| From: | Jamil D Document Control Desk |
| To: | Duke Power Co |
| References | |
| Download: ML050970285 (34) | |
Text
Duke D.M. JAMIL rdwPowere Vice President A Duke Energy Company Duke Power Catawba Nuclear Station 4800 Concord Rd. / CNO1 VP York, SC 29745-9635 803 831 4251 803 831 3221 fax March 31, 2005 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
SUBJECT:
Duke Energy Corporation Catawba Nuclear Station, Unit 1 and Unit 2 Docket Numbers 50-413 and 50-414 2004 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 2004. This report is being 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 803.831.3237.
Sincerely, D. M. Jamil Attachments
- 6W?1 www. dukepower. corn
Nuclear Regulatory Commission March 31, 2005 Page 2 xc:
W. D. Travers, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Sam Nunn Atlanta Federal Center 23T85 61 Forsyth St., SW Atlanta, GA 30303 S. E. Peters (Addressee only)
U. S. Nuclear Regulatory Commission Mail Stop 0-8 G12 Washington, DC 20555-0001 E. F. Guthrie NRC Senior Resident Inspector (CNS)
CNO1NC
Nuclear Regulatory Commission March 31, 2005 Page 3 bxc:
K. E. Nicholson CNO1RC ELL ECO50 File CN:801.01 CN04MD CNS Date File CNO1SA bxc w/o attachments:
L. A. Keller CN0lRC C. J. Thomas MG01RC B. G. Davenport ON03RC R. L. Gill ECO50 Catawba Owners NCMPA-1, SREC, RMPA, NCEMC
U. S. Nuclear Regulatory Commission Page I of 31 Type: Design Change Unit: I
Title:
Design Change CD100147 - VQ Alternate Release Method and OP/I/A/6450/017 Enclosure 4.9 Restricted Change Number 56A Description A temporary modification will be installed within the Unit I VQ System (Train A) to enable containment air releases without the aid of an exhaust fan (compressor). The air release will be performed using differential pressure between the containment and unit vent. The modification will remove or disconnect the VQ exhaust fan (IVQ XF A) and check valve (IVQ 19) from the discharge piping flow path. The VQ filter unit will be connected to the downstream piping using wire supported neoprene/fiberglass hosing to complete the flow path from the containment to the unit vent. The hose will be connected to flanges installed on the system piping using hose clamps. This type of flexible hose connection is currently used to connect the suction and discharge piping to the exhaust fans. Before the Train IA exhaust fan and check valve are removed, the Train IB will be separated from the Train IA by closing IVQ8 and IVQ9. This will ensure that air does not pass through the opposite train while the alternate release method is implemented. The Train I B will remain available for an air release if desired or the alternate method is unsuccessful. If Train IB is needed to perform an air release, IVQ8 and IVQ9 will be opened and IVQ6 and IVQI will be closed. A restricted change will be made to OP/1/A/6450/17 to perform the air release without the Train IA exhaust fan. The VQ alternate air release will be monitored as usual using I EMF 39(L) or IEMP 36(L). The Gaseous Waste Release (GWR) volume will be conservatively estimated using the flow totalizer or calculated using 350 scfm, which is the same methodology applied when the flow totalizer in the normal air release flow path is inoperable. The maximum air release flow rate using the alternate release method will be much lower due to the removal of the exhaust fan. Therefore the 350 scfm used in the GWR calculations will be conservative. Each alternate air release will be implemented until the containment pressure stabilizes. The containment pressure will stabilize when the air inleakage is equivalent to the air release flow rate. After the containment pressure stabilizes, the containment isolation valves, IVQ2A and IVQ3B will be closed to comply with the technical specifications. Based upon engineering judgement, each release will be performed for less than one shift. Containment air releases performed at the McGuire Nuclear Station (MNS) vary from two to eight hours depending upon the position of the unit vent control air flow control valve. The alternate air release method will be repeated for up to 30 days so that Engineering can obtain air release data and monitor the effectiveness of the method. After the air release data is obtained, the temporary modification will be removed and the Unit I VQ System returned to its normal configuration. If this alternate air release method proves to be effective, CNS plans to submit a license amendment to enable continuous air releases.
Evaluation The Containment Air Release and Addition (VQ) System is utilized during normal plant operation to maintain containment pressure within the technical specification limits.
All VQ System equipment, except the containment isolation valves, penetrations, and associated containment isolation valve instrumentation is non-safety related. The ability of the VQ System containment isolation valves to automatically close upon a containment ventilation isolation signal will not be affected by this modification or implementation of this restricted procedure change.
Technical Specification 3.6.3 allows opening of the VQ System containment isolation valves for pressure control, ALARA, air quality considerations, or surveillances. There is not any specific time limit in the technical specifications regarding how long the isolation valves may be open for pressure control. However, there is an implied restriction that the valves should be closed after each containment air release. Since the
U. S. Nuclear Regulatory Commission Page 2 of 31 alternate air releases will be performed without an exhaust fan, the time that the valves will be open for each release will increase. This will not violate any technical specification requirements.
This temporary modification and restricted procedure change alters the way Unit I containment air release will be performed to reduce containment pressure as described in UFSAR Section 9.5.10. This temporary alternate air release or pressure control method will be implemented to verify that it is feasible and therefore will not be described in the UFSAR. After each air release, the containment isolation valves will be closed to comply with Technical Specification 3.6.3 and ensure that the containment pressure boundary is maintained during normal plant operation. This alternate air release method is similar to the method utilized at the MNS except the MNSdoes not close both containment isolation valves after each release. The ability of the VQ System containment isolation valves to perform their safety related design basis fission product barrier function will not be affected by this alternate air release method.
Radiological sampling and monitoring of releases will be the same as normal releases.
The methodology used to calculate the normal does rates or total doses will not be affected by this procedure. Therefore, there will not be any impact to the 10 CFR 20, 10 CFR 50, Appendix 1, and 40 CFR 190 licensing bases. Tornado protection measures will be controlled by the normal means using RP/0/A15000/007. Based upon the above, no technical specification or UFSAR changes are necessary.
Type: Minor Modification Unit: I
Title:
Minor Modification CNCE-I 1024 - Remove straightening section from the auxiliary building ventilation (VA) system air flow monitor IABUX-AFMD-I Description This modification will remove the straightening section from the Auxiliary Building Ventilation (VA) System air flow monitor IABUX-AFMD-I (IVAFE5240) located in the unfiltered exhaust ductwork.
Evaluation This modification will remove the straightening section, honeycomb, from the VA system air flow monitor IABUX-AFMD-1. The honeycomb has small openings which straighten the air, and these openings are getting blocked with dirt and debris. After removal of the straightening section the air flow monitor will continue to perform its design function of monitoring air flow rate. After removal of the straightening section the seismic integrity of the air tlOw monitor will be maintained as a ducting enclosure for the passage of air during both normal and accident conditions. The VA filtered exhaust will be performance tested after this modification to ensure that it continues to meet air flow requirements of Technical Specification 5.5.11 VFTP. No new failure modes have been identified with implementation of this modification. There are no Unreviewed Safety Questions associated with this proposed modification. No Technical Specification or UFSAR changes are required.
U.S. Nuclear Regulatory Commission Page 3 of 31 Type: Minor Modification Unit: 2
Title:
Minor Modification CNCE-61427 - Modify D/G 2A lube oil supply tubing to the turbocharger and install locking device on valves 2LD149 and 2LD150 Description The activity associated with this evaluation is Minor Modification CE-61427. This modification replaces the one-quarter inch lube oil supply tubing to the left and right banks of the D/G 2A Turbocharger with one inch tubing. Valves 2LD149 (Left Bank) and 2LD 150 (Right Bank) will be added as throttle valves so the supply pressure can be adjusted, as required, to maintain the appropriate setting. In addition, a locking device will be added to valves 2LD149 and 2LD150 to prevent tampering and mispositioning of the set throttled position.
The larger tubing will result in a smaller pressure drop and therefore will provide more lube oil pressure at the turbochargers. The addition of the throttle valves will provide the capability to adjust the lube oil pressure to turbochargers to achieve the proper pressure.
Evaluation The new valves (2LD 149 and 2LD 150) are one inch full port ball valves. These new valves are stainless steel QA-I Class B with design conditions of 150 psig at 100 degrees F. and will meet the design temperature and pressure requirements of the interfacing system (100 psig and 200 degrees F.). These valves are suitable for a low pressure throttling applications. The new one inch lube oil supply tubing will also meet the design requirements. The new ball valves and tubing are suitable for use in the LD System. The addition of the weight of the new components has been evaluated for impact to the stress analysis and new support/restraints have been designed. Adding the new ball valves and one inch tubing will not affect the design function of the LD System during any phase of operation. Increasing the size of the turbocharger lube oil supply tubing from three-quarter inch to one inch combined with the addition of the ball valves for throttling will continue to allow the oil pressure to be set properly. Throttling of the ball valves to maintain the correct oil pressure will ensure the lube oil supply pressure to the turbocharger is adequate. The handle assembly for the 2LD149 and 2LD150 application of Item Number DMN-904 will be modified to allow installation of a locking device. This locking device is shown on CNM 1205.28-0075.002. The removal of the majority of the valve handle and addition of a locking plate and hardware will result in a small weight and center of gravity change for the valve. These valve configuration changes are considered minor, and their affect on the structural integrity and vibrational frequency of the valve is considered negligible. Thus the seismic report for this valve will not be affected. The locking device will be used to lock the valve at a specific throttle position and will not affect the pressure boundary integrity of the valve or the ability of the valve to function properly. QA-l Condition materials will be used to fabricate the locking device. The valves will be throttled and set at the proper position by an approved maintenance procedure. During D/G operation, turbocharger lube oil supply pressure is monitored by the Operations Group.
Also, turbocharger lube oil supply pressure is trended by System Engineering for any fluctuations in pressure. The affected portions of the LD System will continue to function as described in the UFSAR and the LD System design basis specification.
Modification CE-61427 does not involve an Unreviewed Safety Question. No changes to the Technical Specifications or UFSAR are required.
U. S. Nuclear Regulatory Commission Page 4 of 31 Type: Minor Modification Unit: I
Title:
Minor Modification CNCE-62336, Install surge protection on the FWST level instrument loops Description Refueling Water Storage Tank (FWST) level transmitters have a history of failures due to lightning strikes (Reference Root Cause in Problem Investigation Process [PIP]
C03-456G). As a corrective action in the referenced PIP, Modification CNCE-62336 will install a surge protector in the 4-20mA signal loop at both the transmitter and at the process cabinet to protect the equipment against damage caused by voltage surges due to lightning strikes.
Four separate channels of tank level measurement are provided by I FWLT5000, I FWLT50 I 0, IFWLT5120, and I FWLT5130 for Channels 1, 2, 3, and 4 respectively.
Each channel provides an input to the 7300 Process Instrumentation and Control System. Level indications for Channels I through 4 are provided by indicators IFWP5000, IFWP5010, IFWP5120, and IFWP5130, located on main control board IMC9. Channels I and 4 are recorded on chart recorders INSCR5390 and INSCR5040, respectively, also located on main control board IMC9. Each channel also provides an analog input to the OAC and provides inputs to the Solid State Protection System (SSPS). The SSPS uses the inputs in a 2/4 logic scheme to generate the automatic control signals to open Containment Sump Recirculation suction valves INI184B and INI1 85A and to close I FW27A and 1FW55B for swapover from Injection to Recirculation upon reaching the Lo Level setpoint.
The following setpoints are associated with the FWST level indications:
High Level - Overflow Alarm - 97.9% (493.88 inches*)
Make Up Level - Make Up Alarm Level - 95.3 % (481.26 inches*')
Lo Level Automatic ND Pump Swapover - 36.6% (197.15 inches*)
Lo Lo Level - NS Pumps must be Secured - 10.6% (71.07 inches*)
- AII inches are with respect to the bottom of tank.
The level transmitters are located in terminal boxes in the annular region formed between the FWST and the missile wall. At this location, a rupture of the tank would submerse the terminal boxes. The transmitters are in the EQ program for a harsh environment due to the potential for submergence, which they are qualified for.
The Innovative Technology (IT) OEM-D48-6C transient voltage surge suppressors (TVSS) being installed by this modification have been qualified for use in this safety related application under the commercial grade program (see CGD-3006-03-18-0001).
This commercial grade documentation includes: I) the Lightning Surge Testing Report (PQGITOEMD48) which documents the ability of the surge suppressors to protect the equipment from surges; 2) the EQ qualification along with an evaluation of the qualification of the device for submergence; 3) the seismic qualification; 4) documentation of testing which showed acceptability of loop response time with added surge Suppressors; and 5) documentation of testing to show that the surge suppressors do not change the FWST level, total loop uncertainty.
The Lightning Surge Testing (PQGITEMD48) performed on the surge suppressors found that when a voltage surge was applied to the 4-2OmA signal line, the input to the loops NAL card would spike low, the output of the NAL card would blip high, but because the time duration was so fast the bi-stable card would not change state. This demonstrated that when the surge suppressor was functioning to protect the equipment against the lightning surges, it did not cause an inadvertent actuation signal to occur.
U.S. Nuclear Regulatory Commission Page 5 of 31 As a result of this modification, actual response time testing will be reinstated for the FWST level transmitter and surge suppressors to verify the total response time for the transmitter and surge suppressors is less than the allocated time for the transmitter, since the surge suppressors were not addressed in the WCAP reports (WCAP-14036-P-A Rev. I and WCAP-13632-P-A, Rev. 2).
In order to facilitate installation of the surge suppressors at the FWST level transmitters, the modification will make changes to the wire termination specification (DPS-1390.01-00-0003), to allow splicing of the surge suppressor leads in the FWST Level transmitters using Thomas & Betts (T&B) nylon insulated splice connector, catalog number RBB25.
The applicable UFSAR sections for the FWST level monitoring instrumentation are 6.3.2.8, 6.3.5.4, 7.6.5 and Table 6-93. The applicable Technical Specifications are 3.3.2-1 7(a) and (b), 3.3.3-119, and 3.5.4. Neither the Technical Specifications nor the UFSAR documentation will need to be changed as a result of this modification.
Evaluation Modification CNCE-62336 adds transient Voltage surge suppression devices in the 4-2OmA FWST level channel instrument loops, to protect these loops from damage due to lightning. The addition of the surge suppressors will not change the function of the FWST level monitoring instrumentation. The evaluation concluded that none of the criterion of IOCFR50.59(c) (2) are met by the proposed activity. Therefore, prior NRC approval is not required. Furthermore, no Technical Specification or UFSAR document revisions are necessary.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 6 of 31 Type: Miscellaneous Item Unit: 0
Title:
Change exercising frequency for RF389B and RF447B from quarterly to cold shutdown
Description:
This evaluation is to provide documentation of the impracticality of the quarterly tests and confirms the change may be implemented without prior NRC approval.
Currently the RF CIVs are exercised at a quarterly interval, which is the preferred testing frequency if plant conditions allow such testing. The fire protection header in containment is maintained dry, and is only filled with water when called upon for fire suppression. Quarterly stroking of the CIVs (MOVs) introduces water into the containment header which has to be drained. Additional valves arc closed to facilitate this testing; however, the piping configuration typically results in water migrating into the containment header. This necessitates a containment entry at power which exposes personnel to higher levels of radiation when draining the system after the quarterly test.
Additionally, the introduction of water into the containment header accelerates corrosion due to draining/depressurization, produces additional wastewater for processing and time pressure to complete the test within the one hour action statement. This is impractical and is justified for deferral to cold shutdown in accordance NUREG-1482.
According to the ASME Section Xl OM-10 Code and NUREG-1482 (section 2.4.5),
valves should be tested at a quarterly frequency if possible. If this is not possible, valve testing can be deferred to cold shutdown. If this is also not possible, the testing can be deferred to refueling. The basis for deferring a test is the determination that the test is impractical at the required frequency. The definition of "impractical" is not given in the Code orNUREG-1482; however, several examples are given. Potential radiation exposure and accelerated corrosion due draining and depressurizing of the piping are examples of impracticality. There is justification to move these tests to cold shutdown, but not refueling.
The purpose of Code required testing is to power operational readiness. Changing the testing frequency to cold shutdown docs not sacrifice component reliability or challenge operational readiness. These valves arc normally closed (their safety related position).
They are only opened upon verification of a fire in containment and would only be used when the unit is shutdown. Therefore, they only need to function during shutdown periods and can be manually opened if needed, since an accident is not assumed concurrent with a fire.
For consistency, the MNS IST Program was reviewed to determine the frequency for similar valves at that site. MNS RF CIVs are stroked on a cold shutdown frequency.
Evaluation: Given the impracticality of testing these valves at a quarterly frequency, the frequency will be moved to cold shutdown. Prior NRC approval is not required when deferring testing to cold shutdown or refueling, if the tests are impractical (ref. NUREG-1482 section 2.4.5). Under the guidelines for the OM-10 Code and NUREG-1482, moving the frequency from quarterly to cold shutdown isjustified.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 7 of 31 Type: Miscellaneous Item Unit: 2
Title:
Compensatory Action for PIP C04-0114 Operable but Degraded Evaluation - NS HX 2A Degraded Baffle Plate Conditions
Description:
PIP C04-0114 identifies degraded conditions associated with the Containment Spray Heat Exchanger 2A. Replacement of this heat exchanger is currently planned for 2EOC13. An operability analysis has been performed to evaluate the structural integrity of the heat exchanger under all design basis loadings. The supporting engineering analysis is based on limits for particular parameters. These parameters require control and compensatory actions related to limits on RN system flow to the NSHX 2A have been established. In addition, seismic accelerations limits which trigger operator actor actions have been reduced specifically for the NSHX 2A. The compensatory action assures that the operability of the NSHX 2A, the containment spray system, and the containment is maintained. This compensatory action will support the assumptions included in the technical evaluation associated with the Operable but Degraded (OBD) evaluation of PIP C04-0114. The issue is maintaining the structural integrity of the NSHX 2A tubing. The verification of assumptions used in the operability evaluation is necessary to assure compliance with Technical Specification 3.6.6, Containment Spray System.
Evaluation: In order to maintain the structural integrity of the NSHX 2A, compensatory actions have been established to limit RN system flow to the NSHX 2A. This restriction is necessary to limit imposed forces on the heat exchanger tubing from dynamic vortex shedding of the secondary side flow. In addition, seismic accelerations have been limited to any detectable earthquake levels as a conservative measure to ensure an inspection opportunity is available. The heat exchanger is qualified for all operability limits and will function under all design basis accidents and design events provided the flow limits are satisfied. The heat exchanger remains available and operable up to and including accelerations associated with a safe shutdown earthquake (SSE). The limit associated with seismic accelerations is simply a conservative measure for continued assurance of the heat exchanger structural integrity. The issue is maintaining the structural integrity of the NSHX tubing. The verification of assumptions used in the operability evaluation is necessary to assure compliance with Technical Specification 3.6.6, Containment Spray System.
The compensatory action does not affect any control systems or modify any SSCs. No physical changes are involved. No malfunctions of equipment important to safety can be created through the imposition of this compensatory action. The fission product barrier of Containment is assured through continued operability of the NSHX 2A by control of RN system now to the heat exchanger to known and previously evaluated parameters. The method of evaluation regarding the heat exchanger performance or structural integrity presented in the UFSAR are unaffected by this Compensatory Action.
No Technical Specification, UFSAR, or SLC changes are required. Prior NRC review and approval is not required.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 8 ot31 Type: Miscellaneous Item Unit: 0
Title:
Evaluate spring only PORV stroke time requirement
Description:
This evaluation will determine if the spring-only fail-safe operation is required.
Evaluation: The spring function of the PORV is never relied upon to close the PORV by itself. Other motive forces are available to provide that function. MCE recommendation is to delete the spring only valve stroke timing tests. McGuire Nuclear Station completed a similar evaluation in 2002, with the same conclusion. It would be prudent to continue the spring only functional tests as this verifies freedom of movement. The existing Periodic Test Procedures (PT/I (2)/A/42001023A) will be revised to remove the stroke time acceptance criteria, and add steps to confirm the PORVs move from full open to full close under spring force only.
Fail-safe actuation implies spring only ability to close. In all of the PORV closing scenarios, there arc no situations where the spring is acting alone to close these valves. In every case, the spring supplements air, nitrogen and NC system pressure when closing the valve.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 9 of 31 Type: Miscellaneous Item Unit: 0
Title:
Operable but Degraded evaluation associated with the non-conservative RN pit level transfer setpoint in Technical Specification 3.3.2, Table 3.3.2-1, Function 10, Nuclear Service Water Suction Transfer - Low Pit Level
Description:
The purpose of this evaluation is to determine if a License Amendment Request (LAR) is required for Revision 244 Change B to OPIO/N64001006C, Nuclear Service Water (RN) System Temporary Modification CNTM-O 158, RN Pit B Swap Set Point Temporary Modification CNTM-0159, IRN-4B removed from the RN system The Operable-but-Degraded evaluation associated with the non-conservative RN Pit level transfer set point in Technical Specification 3.3.2, Table 3.3.2-1 Function 10, Nuclear Service Water Suction Transfer - Low Pit Level.
Evaluation: IRN-4B, which provides an assured suction flow path from the SNSWP to the B Train RN pumps is being maintained in the open position with power removed or is removed from the'system, and the lake isolation valves I RN-5A and I RN-6B are being maintained closed with power removed. This is due to internal problems with the valve. This has necessitated aligning the suction and discharge RN flow path to the SNSWP. This is an unacceptable long term solution since the SNSWP will heat up over time. Technical Specifications place an upper limit on the SNSWP temperature of 91.5 degrees F. To reduce the heat up to the SNSWP, the normal RN system operation is being altered to allow the A Train of RN to take suction from Lake Wylie and discharge to Lake Wylie.
The suction of B Train of RN will be aligned to the SNSWP while the discharge is aligned to Lake Wylie. This condition is acceptable as long as B Train of RN is not running. Gravity draining of the SNSWP will not occur since the Pit B lake isolation valves IRN-5A and I RN-6B are maintained closed with power removed. Also, due to the design of the RN pumps, it is impossible for a 26 foot column of water to be maintained in the pump casing and siphon from the pump pit up through the pump discharge isolation valve and check valve, into the RN system, and ultimately to the lake.
The pump discharge isolation valve and check valve are closed when the pump is off.
Some events (Ss signal, Loss of Offsite Power, and Loss of Lake Wylie) automatically start the standby RN pumps. This start of the B Train RN pumps will cause the SNSWP to be pumped to Lake Wylie. To prevent the draining of the SNSWP below Technical Specification (TS) limits (571 feet elevation), the B Train RN low pit level auto swap will be adjusted (Temporary Modification CNTM-O 158) from a current nominal value of 557.5 feet to 571.5 feet. This alternate RN system alignment and the supporting modification and procedures are temporary and will be in place until valve IRN-4B is repaired and reinstalled, prior to or during the next unit 2 refueling outage (2EOC13). At that time the temporary modification will be removed, and the RN system will be returned to its normal configuration and alignment.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 10 of 31 Type: Miscellaneous Items Unit: 0
Title:
Compensatory Action associated with the removal of the hatches in the Unit 2 UHI building roof
Description:
The purpose of this evaluation is to document why no compensatory actions are needed for the removal of the unit 2 UHI Building roof hatches. These hatches will be removed to facilitate the renovation of the UHI Building for the MOX Fuel Project. There are several modifications associated with the renovation work including CN-28001, CN-28002, CN-28008, CNCE28 102, CNCE-28103 and CNCE-28012. The hatches will be removed (opened) under Work Order #98549483.
This evaluation is applicable before and during the UHI renovation work. After the renovation work is completed, another evaluation will need to be performed if it is desired to pull the UHI hatch plugs.
Evaluation: The UHI system at CNS was an early design feature that was determined to be unnecessary. The system was abandoned and never used. The building that housed the unit 2 UHI system contains five components that will be in-service during the renovation process. Roof hatches typically provide protection for equipment and components located underneath them. With the hatches removed, this protection will be unavailable, and thus, these in-service components must be evaluated for damage or loss of function during the time the roof hatches are removed. The components under evaluation and their significant functions are:
(I) A 2 inch UHI Makeup Line (associated penetration is M405) (maintains annulus integrity and VE system operability)
(2) Two 12 inch UHI Injection Lines (associated penetrations are M406 and M407)
(maintains containment integrity and VE system operability)
(3) A 2 inch unused capped pipe (associated penetration is M454) (maintains containment integrity and VE system operability)
(4) A 2 inch WL Loop Seal (maintains annulus integrity and a flow path for the UHI sump)
(5) A 2 inch FW Makeup Line for the WL Loop Seal J(provides makeup as needed to maintain the WL loop seal and maintains FWST inventory)
It should also be noted that there is an in-service VI line in the UHI Building. This line will be isolated during renovation using valve 2VICCI3. With valve 2VICC13 closed, there will be less than 3.5 feet of I inch VI line in the UHI Building that will be significant. That portion of the VI line is located near a wall and is behind existing equipment. It is thus protected and further evaluation is not needed.
The UHI roof hatches serve the following functions: HVAC Control, Tornado Missile Protection, Turbine Missile Protection, and Environmental Protection. With the hatch plugs removed, each of these functions is affected. The following explanations provide the justification for why no adverse impact is created to important plant equipment by removing the unit 2 UHI roof hatches. Justification is also be provided as to why no tornado pressure protection concerns are created by the removal of the hatch plugs, and why no fire barrier or security barrier protection is needed.
HVAC Control The HVAC for this room was originally designed to provide basic radiological and temperature control. Since the UHI Building was abandoned, the radiological function
U. S. Nuclear Regulatory Commission April 1, 2005 Page 11 of 31 has not been a concern. In the past, the temperature control function was installed to prevent freezing of water filled lines in the building. During renovation of the building, the only water-filled lines of concern will be the WL and FW lines. These lines contain borated water and are located low in the building. The building will be heated from either duct mounted heaters or unit heaters. This is adequate to prevent any freezing problems with the WL or FW lines. Thus, HV AC control is acceptable with the hatch plugs removed and no specific compensatory measures are needed for this design function.
Tornado Missile The dominant sources of tornado generated missiles are expected to originate from structures and objects located at ground level not from the roof area. Since the hatches are located on the Auxiliary Building roof, well above grade level, smaller lighter weight objects will have a higher potential to be lifted high enough to reach the hatch area on the roof. Larger and heavier missiles, which have a greater damage capacity, tend to have a low trajectory and not strike targets significantly higher than their original height. It should also be noted that tornado generated debris would not cause significant damage to the components of concern by simply falling through an open hatch and striking components below. Valves and piping in these systems are robust enough to withstand minor impacts and they are seismically supported. For any significant damage to occur, a tornado missile would need to be propelled through the open hatch at a steep angle.
UFSAR Table 3-15 shows that each design basis tornado missile has a horizontal velocity component that is greater than its vertical velocity component. Thus, the possibility of a steep angle tornado missile is not credible. Additionally, the target area of the components in question is relatively small and the components are located well below the roof hatch openings. Thus, based on the size and location of the target areas and on the steep angle a damaging missile would have to travel, damage from tornado generated missiles is not considered credible. No compensatory measures are necessary to protect against the loss of tornado protection caused by removing the roof hatches.
Turbine Missile Per UFSAR section 3.5.1.3.2 and Figure 3-5, the roof over the UHI Building is outside of the critical 25 degree angle and; therefore, high-energy, low trajectory missiles are not applicable. However, high trajectory missile probability must be evaluated. Per UFSAR section 3.5.1.3.4, an event having a probability of causing unacceptable damage of the order of E-7 per reactor year at the plant is not considered significant. For the two-unit Catawba Nuclear Station, an event having a probability of occurrence of the order of2E-7 is sufficient to fulfill this criterion. Table 3-12 shows lifetime probability of hitting the UHI Building (4284 square feet) is 2.6E-4. As shown in Table 1, the cross-sectional or target area of the in-service components is less than 20 square feet. Factoring in the small, exposed target area of the components reduces the strike probability to approximately:
Probability = (2.6E-4)140 years x (20 sq ft/4284 sq ft) = 3E-8/year. Thus, no specific compensatory measures are needed to protect against turbine missiles with the roof hatches removed.
TABLE I -TARGET AREA (I) A 2 inch UHI Makeup Line (penetration M405) - 59 inch/12 x 2 inch/12 = 0.8 sq ft (2) Two 12 inch UHI Injection Lines (penetrations M406 and M407) - (4 feet + 3 feet) x 12 inch/12 = 7 sq ft (3) A 2 inch unused capped pipe (penetration M454) - 21 inch/12 x 2 inch/12 = 0.3 sq ft (4) A 2 inch WL Loop Seal - 50 inch/12 x 2 inch/12 = 0.7 sq ft (5) A 2 inch FW Makeup Line for the WL Loop Seal - 57 feet x 2 inch/12 = 9.5 sq ft Environmental Protection
U. S. Nuclear Regulatory Commission April 1, 2005 Page 12 of 31 The hatches for the UHI Building protect interior components from the environment. The need to protect against freezing conditions inside the room has been previously discussed.
In addition to freezing concerns, rain, sleet and snow should be kept out of the building.
This will prevent potential flooding problems as well as ensuring that Chemistry does not have to treat and release significant quantities of rain water. However, rain, sleet and snow will not cause any of the in-service components from performing their intended safety functions. Thus, from a practical standpoint, a metal cover (scuttle) will be placed over the hatch openings when the hatch plugs are removed but this is not a requirement for any design function. No compensatory measures are required for environmental protection concerns.
Tornado Pressure The hatch plugs do not serve as a tornado pressure barrier. Although they were originally intended to be such, the ventilation system design was flawed so these are not a pressure boundary. Not having a tornado pressure boundary for the building is acceptable since the building is isolated from other structures, and since components within the building are not vulnerable to damage from tornado induced pressures. Thus, no specific compensatory measures are needed to address tornado pressure.
Fire Barrier The exterior of the UHI Building is not considered a Fire Barrier and thus the hatches are not fire barriers. Thus, no compensatory measures are needed to address fire barrier issues.
Security Barrier The removal of the roof hatch creates a large entry into the UHI Building that bypasses normal security controls. However, based on the fact that there are no vital areas below the hatches and access to the open hatches is within the protected area of the plant, no compensatory measures are needed to address security issues while the hatches are open.
Based on the above discussion, it is acceptable to remove the roof hatch plugs above the unit 2 UHI Building without implementing any compensatory actions. This evaluation is applicable before and during the UHI renovation work. After the renovation work is completed, another evaluation will need to be performed if it is desired to pull the hatch plugs.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 13 of 31 Type: Miscellaneous Items Unit: I
Title:
Compensatory Action for PIP C03-6488 Operable but Degraded Evaluation NS HX IA Degraded Baffle Plate Conditions
Description:
PIP C03-6488 identifies degraded conditions associated with the Containment Spray Heat Exchanger IA. Replacement of this heat exchanger is currently planned for IEOC 15. An operability analysis has been performed to evaluate the structural integrity of the heat exchanger under all design basis loadings. The supporting engineering analysis is based on limits for particular parameters. These parameters require control and thus a compensatory action related to limits on RN system flow to the NSHX IA has been established. This compensatory action assures the operability of the NSHX IA, the Containment Spray System, and the containment is maintained. This compensatory action will support the assumptions included in the technical evaluation associated with the Operable but Degraded (OBD) Evaluation of PIP C03-6488. The issue is maintaining the structural integrity of the NSHX IA tubing. The verification of assumptions used in the operability evaluation is necessary to assure Compliance with Technical Specification 3.6.6, Containment Spray System.
NSD-203 and Site Directive 3.1.18 applies to the OBD associated PIP C03-6488 including the requirement to review the specific compensatory actions under 50.59.
Evaluation: In order to maintain the structural integrity of the NSHX IA, compensatory actions have been established to limit RN system flow to the NSHX IA.This restriction is necessary to limit imposed forces on the heat exchanger tubing from dynamic vortex shedding of the secondary side flow. The heat exchanger is qualified for all operability limits and will function under all design basis accidents and design events provided the flow limit is satisfied. The heat exchanger remains available and operable up to and including acceleration associated with an operating basis earthquake (OBE) or safe shutdown earthquake (SSE). The issue is maintaining the structural integrity of the NSHX IA tubing. The verification of assumptions used in the operability evaluation is necessary to assure compliance with Technical Specification 3.6.6, Containment Spray System, The compensatory action does not affect any control systems or modify any SSCs. No physical changes are involved. No malfunctions of equipment important to safety can be created through the imposition of this compensatory action. The fission product barrier of containment is assured through continued operability of the NSHX I A by control of RN system flow to the heat exchanger to known and previously evaluated parameters. The method of evaluation regarding the heat exchanger performance or structural integrity
.presented in the UFSAR are unaffected by this compensatory action.
No Technical Specification, UFSAR, or SLC changes are required. No License Amendment is necessary; prior NRC review is not required.
U. S. Nuclear Regulatory Commission i April 1, 2005 i Page 14 of 31 I
. i Type: Miscellaneous Items Unit: 0
Title:
PIP C04-01445 - Actions for 1(2)EMF53A/B to go from OPERABLE to OBDINCI (Operable But Degraded/Nonconforming)
Description:
PIP C04-1445 identifies that there is an issue with the containment high range radiation monitors at both McGuire and Catawba. This is due to a thermal effect on the coaxial cabling that applies the current signal that is converted to radiation indications in the control roo. Specifically, the effect is Thermally Induced Current (TIC) which is caused by a temperature differential between the cent conductor and the outer shield of coaxial cable.
As the external temperature (Te) of the cable increases above that of the center conductor (Tc), a small DC current (3E-10 amps) is generated on the conductor. As the differential temperature is increased, the current increases. This results in the EMF53 indicating higher than actual. Once containment spray and ice condenser functions actuate to mitigate the steam in containment, the external temperature will begin to drop (Tc < Te) which results in a negative current flow on the center conductor. EMF53 indication is then less than actual (if the negative current is sufficient, there may be a loss of signal at the detector).
Evaluation: The proposed activity is the incorporation of required corrective actions to change the Containment High Range Radiation Monitors (HRRM), l(2)EMF53A/B from INOPERABLE to OPERABLE BUT DEGRADED during the onset of a Loss of Coolant Accident (LOCA) and/or High Energy Line Break (HELD). These actions are to revise the requisite procedures and advise the appropriate Operations, Emergency Coordinator, Emergency Operations Facility Director, Accident and Dose Assessment personnel that the Thermally Induced Current phenomenon may cause the indication on the HRRMs to be invalid for up to nineteen (19) minutes following the onset of the event.
All FSAR Chapter 15 Design Basis line break accidents inside containment, of all sizes, from the smallest to the largest (both primary and secondary system breaks included), and the containment temperature profiles identified in UFSAR Chapter 6 were evaluated. The profiles of Chapter 6, specifically Figures 6.8 and 6.20, represented the longest event duration and highest temperatures. Other line breaks were found to deposit less energy or equivalent energy over shorter duration, thus leading to a faster dissipation of the TIC effect. Figures 6.8 and 6.20 (all cases) were analyzed via the EPRI TIC software for determination and quantification of adverse affect(s) on the EMF 53A/B licensing basis function. The analysis determined that the limiting condition is bounded by the data represented by UFSAR Figure 6.8. Based on this analysis, a maximum period of instability in the EMF output was determined to be 19 minutes. As a result, this is the time frame that has been used in establishing the associated compensatory actions.
The evaluation was performed to determine if waiting approximately nineteen (19) minutes following the onset of the event would adversely impact SSC performance, accident scenarios, fission product barriers, or the ability of operators to mitigate the consequences of the accident, thereby protecting the health and safety of the public.
It is concluded that these corrective actions, as proposed, do not represent said adverse impact, do not require a change to the Technical Specifications, do not require a change to the licensing basis, do not require an UFSAR change to implement, and do not require NRC approval prior to implementation.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 15 of 31 Type: Nuclear Station Modification Unit: 2
Title:
Nuclear Station Modification CN-21407/0 - Install leak-off line in containment for each ND line
Description:
Nuclear Station Modification NSM CN-21407/0 will install a manual vent line, capable of being placed in a continuous venting mode, on the ND discharge headers in containment. The purpose of the vent lines is to maintain the pressure upstream of the secondary NI check valves as low as possible in order to maintain high seating forces (D/P) on the secondary pressure boundary NI check valves. In addition, the vent lines have a secondary function of continuous removal of non-condensable gases that may come out of solution in the lower pressure ND header. Vent lines will be added to the A train and B train headers as shown on flow diagram CN-2562-1.3 as modified per CN-21407/0.
Evaluation: The modification will install necessary Class B and Class E stainless steel piping and tubing on the existing Class B vent valve 2NI-416 and 2NI-418 for the A Train NI header. The new piping/tubing will be the appropriate design temperature (650 degrees F) and pressure (2500 psia) to be compatible with the existing ECCS piping. A flow restricting orifice will be installed in each line to limit ECCS out flow to an acceptable value should a Class E pipe break occur. The lost flow out of this flow path has been accounted for in the ECCS flow balance tests. A Class B check valve will be included in each line to eliminate direct communication with the containment atmosphere and maintain the design requirements of the applicable GDC-55. A Class E needle valve will be located in the Class E line at the valve station to allow continuous venting. The venting effluent will be routed to the containment floor and equipment sump IA.
The NI System is nuclear safety related and the modification up to and including the flow restricting orifice is QA-I Penetrations M307 and M336 are type B5 as shown on UFSAR Figure 6-113. UFSAR Table 6-77 is revised to reflect the addition of the new Class B Containment Penetration Check Valves 2NI-495 (M307) and 2NI-501 (M336) and associated vents 2NI-496 (M307) and 21-502 (M336) and drains 2NI-494 (M307) and 2NI-500 (M336).
The effluent input to the WL sump has no effect on the ability to perform reactor coolant system leakage surveillances pursuant Technical Specification 3.4.13 (Reactor Coolant System Operation Leakage). Also, Technical Specification 3.4.15 (Reactor Coolant System Leakage Detection Instrumentation) is not adversely affected. Input into this sump will increase background leakage but staying below the alarm setpoint assures no problem will surveillances.
The added Class B (QA-1) piping and valves are the same class and design conditions as the interfacing piping. A pip break in the NI system is not more likely to occur since the same design conditions are invoked. Therefore, there is no increase in frequency of occurrence of an accident evaluated in the UFSAR. CN-21407/0 does not depart from design, fabrication, construction, testing and performance standards applicable to the NI system and these containment penetrations. Thus, there is no more than a minimal increase in the likelihood of occurrence of a malfunction of equipment of an SSC important to safety evaluated in the UFSAR. All of the performance capabilities of the NI system are unchanged. Thus, the accident mitigation functions of the plant remain unchanged. No safety analysis assumptions are affected. Since the modified design will still safety related and single failure proof, the response to any accidents will be unaffected. No fission product barriers are degraded since the QA-I Type C leak rate
U. S. Nuclear Regulatory Commission April 1, 2005 Page 16 of3l tested check valve prevents communication of the containment atmosphere to the piping penetrating containment. GDC-55 requirements are maintained in the design features of CN-2 1407/0. No initial conditions of any UFSAR evaluated accidents are impacted.
Thus the consequences of accidents previously evaluated in the UFSAR are not increased. Additionally, no design basis limits for a fission product barrier described in the UFSAR are being degraded or altered. No analytical methods need to be changed in order to demonstrate that this facility continues to meet any required design basis. No methods of evaluation are impacted by this modification. ECCS flow testing is unaffected. Measuring and detecting reactor coolant system leakage is not adversely impacted. No methods of analysis related to the core or dose consequences are affected.
CN-21407/0 does not require prior NRC approval. This evaluation does not involve an Unreviewed Safety Question. No changes to the Technical Specifications changes are required. UFSAR changes are required to Table 6-77.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 17 of 31 Type: Nuclear Station Modification Unit: 2
Title:
Nuclear Station Modification NSD CN-21432/01 NS System Heat Exchanger 2B Replacement
Description:
Containment Spray (NS) System heat exchangers (HX) are degrading due to raw water corrosion. The current HX design has a carbon steel shell and 304 stainless steel tubes, with raw water (RN) on the shell side. This design is undesirable because of raw water attack on carbon steel, difficulty in cleaning the shell side, and MIC attack on the stainless steel tubes. To adequately address these concerns, the new HX will have a 304 stainless steel shell, Titanium tubes, and AL6XN channel ends. The tube sheet will be stainless steel (AL6XN) with mechanically bonded Titanium cladding, so the Titanium tubes can be rolled and welded to the tube sheet. Raw water (RN) will now be on the tube side. The NS water will now be on the shell side. For ease in cleaning, the new HX design will not have U-tubes as the current design does, and the new HX will have access ports at each end. Currently, the piping to and from the HX is carbon steel, which has degraded due to exposure to raw water. New six percent Molybdenum nickel al l oy piping (AL6XN) will be installed on a section of piping within the NS HX room contained within (but not completely up to) the HX isolation valves 2RN-225B and 2RN-229B, which are normally closed.
Evaluation: Containment Spray (NS) System heat exchangers (HX) are degrading due to raw water corrosion. The current HX design has a carbon steel shell and 304 stainless steel tubes, with raw water (RN) on the shell side. This design is undesirable because of raw water attack on carbon steel, difficulty in cleaning the shell side, and MIC attack on the stainless steel tubes. To adequately address these concerns, the new HX will have a 304 stainless steel shell, Titanium tubes, and AL6XN channel ends. The tube sheet will be stainless steel (AL6XN) with mechanically bonded Titanium cladding, so the Titanium tubes can be rolled and welded to the tube sheet. Raw water (RN) will now be on the tube side. The NS water will now be on the shell side. For ease in cleaning, the new HX design will not have V-tubes as the current design does, 'and the new HX will have access ports at each end. Currently, the piping to and from the HX is carbon steel, which has degraded due to exposure to raw water. New 6 percent Molybdenum nickel alloy piping (AL6XN) will be installed on a section of piping within the NS HX room contained within (but not completely up to) the HX isolation valves 2RN-225B and 2RN-229B, which are normally closed.
The replacement NS HX 2B has some different performance characteristics. The heat transfer capability has been increased by 25 percent resulting in a need to increase the design temperature from 150 deg F to 160 deg F from the HX discharge to where the 18 inch RN header ties into the 42 inch RN common header for NS HX 2B and KC HX 2B.
The pressure drop across the NS 2B HX is also reduced due to the two-pass straight-tube design compared to the V-tube design and swapping the NS flow from the tube side to the shell side. The lower pressure drop results in a slightly higher NS flow and corresponding impact on RWST depletion during the injection phase of ECCS operation. Per the UFSAR 6.3.2.8 change in the Final Scope Document (FSD), this has been evaluated to be acceptable with respect to RWST vortexing avoidance. Section 6.3.2.8 of the UFSAR claims "The Low Level setpoint provides a volume above the no vortex level to account for the maximum RWST outflow during switchover with the most limiting single failure (approximately 116,292 gallons) plus an allowance for instrument error" (approximately 11,315 gallons). No setpoint changes have been made per CN-21432/01. The volume available in the RWST at the time of securing the ECCS pumps decreased 2.1 percent per the UFSAR change provided in the FSD, as a result of CN-21432/01. Based on the
U. S. Nuclear Regulatory Commission April 1, 2005 Page 18 of 31 information provided in the UFSAR 6.3.2.8 and the UFSAR change provided in the FSD, the minimum volume to prevent vortexing increased by 7/10 of 1 percent or 0.007.
Consideration of both of these changes reveals a reduction in margin to vortexing of approximately 6 percent. This change is compatible with the "no more than minimal increase standard" discussed in NEI-96-07 Rev. 1.
The increased NS heat transfer characteristics (flow, area) have been evaluated for impact on the Containment Minimum Backpressure Analysis presented in the UFSAR 6.2.1.5.
There is no adverse effect.
The NS heat exchangers are not accident initiators. The NS heat exchangers perform the accident mitigation function of containment heat removal capability for the Ice Condenser Reactor Containment. The containment sump performance has been evaluated with respect to vortexing, since an increased flow demand occurs as a result of the higher NS flow (lower tube bundle delta-P). The large margin that remains (18601 gallons) is judged to be compatible with "no more than a minimal increase" in sump malfunction potential.
This modification will not degrade the accident response of any QA-I SSCs (NS HXs, RN) which perform accident mitigation functions. No new malfunctions are created since all SSCs remain within their applicable design criteria. Specifically, the heat transfer provided by the new NS HX meets or exceeds its design requirement as defined in the UFSAR. The replacement NS Heat Exchangers, piping, valves and supports will perform their UFSAR defined design functions within applicable acceptance criteria. The NS System will remove heat from the Containment by transferring heat from the NS System to the RN System via the replacement NS heat exchangers. No adverse changes to any fission product barriers are being directly imposed by this modification. Therefore, no DBLFPBs as described in the UFSAR are being exceeded or altered. No analytical methods need to be changed in order to demonstrate that this facility continues to meet any required design basis. No methods of evaluation are impacted by this modification.
NSM CN-2 1432/0 does not require prior NRC approval. This evaluation does not involve an Unreviewed Safety Question. No changes to the Technical Specifications changes are required. UFSAR changes are required to 6.2.2.2, Table 6-71, Table 6-1, Table 34, and section 6.3.2.8.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 19 of 31 Type: Nuclear Station Modification Unit: I
Title:
Nuclear Station Modification NSM CN-l 1407/0 - Install leak-off line in containment for each ND line
Description:
Nuclear Station Modification NSM CN-1 1407/0 will install a manual vent line, capable of being placed in a continuous venting mode, on the ND discharge headers in containment. The purpose of the vent lines is to maintain the pressure upstream of the secondary NI check valves as low as possible in order to maintain high seating forces (DIP) on the secondary pressure boundary NI check valves. In addition, the vent lines have a secondary function of continuous removal of non-condensable gases that may come out of solution in the lower pressure ND header. Vent lines will be added to the A train and B train headers as shown on flow diagram CN-1562-1.3 as modified per CN-11407/0.
Evaluation: The modification will install necessary Class B and Class E stainless steel piping and tubing on the existing Class B vent valve INI-416 (Train B NI header) and a new valve, INI-499 for the A Train NI header. The new piping/tubing will be the appropriate design temperature and pressure to be compatible with the existing ECCS piping. A flow restricting orifice will be installed in each line to limit ECCS out flow to an acceptable value should a Class E pipe break occur. The lost flow out of this flow path has been accounted for in the ECCS flow balance tests. A Class B check valve will be included in each line to eliminate direct communication with the containment atmosphere and maintain the design requirements of the applicable GDC-55. A Class E needle valve will be located in the Class B line at the valve station to allow continuous venting. The venting effluent will be routed to the containment floor and equipment sump IA.
The NI System is nuclear safety related and the modification up to and including the flow restricting orifice is QA-I Penetrations M307 and M336 are type B5 as shown on UFSAR Figure 6-113. UFSAR Table 6-77 is revised to reflect the addition of the new Class B Containment Penetration Check Valves INI-495 (M307) and INI-501 (M336) and associated vents INI-494 (M307) and INI-500 (M336) and drains INI-496 and INI-502 (M336).
The effluent input to the WL sump has no effect on the ability to perform reactor coolant system leakage surveillances pursuant Technical Specification 3.4.13 (Reactor Coolant System Operation Leakage). Also, Technical Specification 3.4.15 (Reactor Coolant System Leakage Detection Instrumentation) is not adversely affected. Input into this sump will increase background leakage but staying below the alarm setpoint assures no problem will surveillances.
The added Class B (QA-1) piping and valves are the same class and design conditions as the interfacing piping. A pip break in the NI system is not more likely to occur since the same design conditions are invoked. Therefore, there is no increase in frequency of occurrence of an accident evaluated in the UFSAR. Cn-1 1407/0 does not depart from design, fabrication, construction, testing and performance standards applicable to the NI system and these containment penetrations. Thus, there is no more than a minimal increase in the likelihood of occurrence of a malfunction of equipment of an SSC important to safety evaluated in the UFSAR. All of the performance capabilities of the NI system are unchanged. Thus, the accident mitigation functions of the plant remain unchanged. No safety analysis assumptions are affected. Since the modified design will still safety related and single failure proof, the response to any accidents will be unaffected. No fission product barriers are degraded since the QA-I Type C leak rate
U. S. Nuclear Regulatory Commission April 1, 2005 Page 20 of 31 tested check valve prevents communication of the containment atmosphere to the piping penetrating containment. GDC-55 requirements are maintained in the design features of CN-1 1407/0. No initial conditions of any UFSAR evaluated accidents are impacted.
Thus the consequences of accidents previously evaluated in the UFSAR are not increased. Additionally, no design basis limits for a fission product barrier described in the UFSAR are being degraded or altered. No analytical methods need to be changed in order to demonstrate that this facility continues to meet any required design basis. No methods of evaluation are impacted by this modification. ECCS flow testing is unaffected. Measuring and detecting reactor coolant system leakage is not adversely impacted. No methods of analysis related to the core or dose consequences are affected.
CN-1 1407/0 does not require prior NRC approval. This evaluation does not involve an Unreviewed Safety Question. No changes to the Technical Specifications changes are required. UFSAR changes are required to Table 6-77.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 21 of 31 Type: Nuclear Station Modification Unit: I
Title:
Nuclear Station Modification NSM CN-I 1438/0 - Replace Hydrogen Igniters in EHM System
Description:
This nuclear station modification replaces the existing Champion hydrogen igniter glow plugs with a different manufacturer, Tayco, glow coil. The replacement glow coil is needed because of unavailability of the Champion glow plug. The Tayco glow coil has been accepted by the NRC as a replacement. The Tayco glow coil draws more current than the existing Champion glow plug. Consequently, some cabling and breakers will need to be upgraded. The containment electrical penetrations are adequate without modification for the increased current draw of the new glow coils. Catawba's 600VAC Essential Auxiliary Power System, including the effect on DIG loading can accommodate the additional current demand and is provided via the DIG from load group 13. The glow coils' function is not safety related. The power is provided from load group 13 originating from IEMXB and IEMXI. These loads receive a permissive and may be started manually if desired.
The new coils are larger and will need a larger mounting box which will be QA-4 mounted. The new coils utilize a spray shield. A review was performed to determine the licensing basis of Catawba with respect requirements of the replacement plugs. Licensing correspondence in reference 5 was reviewed. It was determined that the Tayco glow coil, previously accepted by the NRC at other utilities, along with the associated spray shield met the requirements for Catawba. A Duke Energy Corporation letter to the NRC dated June 29, 2004, communicates to the NRC for information purposes, that the replacement igniters (plugs to coils) are being changed from the Champion plugs to the Tayco coils.
Since the component manufacturer is being changed it wvas deemed appropriate to update licensing correspondence to accurately reflect the manufacturer of the igniters from Champion to Tayco.
The Hydrogen Igniters are not accident initiators. The igniters are designed to ensure a controlled burning of hydrogen in the unlikely event that excessive quantities of hydrogen are generated as a result of a postulated degraded core accident. The HIS is designed to promote the combustion of hydrogen in a manner such that containment integrity is maintained. As such, the igniters are accident mitigation equipment.
The HIS is not Nuclear Safety Related (QA-1) but supported by electrical power and components that are QA- 1. The electrical penetrations have been evaluated as capable of passing the increased current draw of the new igniters. The Class I E emergency D/G has been evaluated as capable of supplying the increased current and remains within its design capabilities. None of these components have had changes imposed on them that deviate from the applicable design criteria. The location (coverage) of the igniters is not changing. The temperature achieved by the igniters (1700 deg F) is not changing; specified in Technical Specification 3.6.9. The reliability of the igniters with respect to spray shield effectiveness is not degraded. No new malfunctions are created since all SSCs remain within their applicable design criteria. No changes to any fission product barriers are being directly imposed by this modification. No methods of evaluation are impacted by this modification. UFSAR section 6.2.5.7, Table 8-6 and Technical Specification BASES section SR 3.6.9.3 requires changes.
Evaluation: Nuclear Station Modification NSM CN-1 1438/0 does not require prior NRC approval.
This evaluation does not involve an Unreviewed Safety Question. No changes to any fission product barriers are being directly imposed by this modification. No methods of
U. S. Nuclear Regulatory Commission April 1, 2005 Page 22 of 31 evaluation are impacted by this modification. UFSAR section 6.2.5.7, Table 8-6 and Technical Specification BASES section SR 3.6.9.3 requires changes.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 23 of 31 Type: Nuclear Station Modification Unit: 2
Title:
Nuclear Station Modification NSM CN-21446/01 - NS Heat Exchanger 2A Replacement
Description:
Containment Spray (NS) System heat exchangers (HX) are degrading due to raw water corrosion. The current HX design has a carbon steel shell and 304 stainless steel tubes, with raw water (RN) on the shell side. This design is undesirable because of raw water attack on carbon steel, difficulty in cleaning the shell side, and MIC attack on the stainless steel tubes. To adequately address these concerns, the new HX will have a 304 stainless steel shell, Titanium tubes, and AL6XN channel ends. The tube sheet will be stainless steel (AL6XN) with mechanically bonded Titanium cladding, so the Titanium tubes can be rolled and welded to the tube sheet. Raw water (RN) will now be on the tube side. The NS water will now be on the shell side. For ease in cleaning, the new HX design will not have U-tubes as the current design does, and the new HX will have access ports at each end. Currently, the piping to and from the HX is carbon steel, which has degraded due to exposure to raw water. New six percent Molybdenum nickel alloy piping (AL6XN) will be installed on a section of piping within the NS HX room contained within (but not completely up to) the HX isolation valves 2RN-144A and 2RN-148A, which are normally closed.
Evaluation: Containment Spray (NS) System heat exchangers (HX) are degrading due to raw water corrosion. The current HX design has a carbon steel shell and 304 stainless steel tubes, with raw water (RN) on the shell side. This design is undesirable because of raw water attack on carbon steel, difficulty in cleaning the shell side, and MIC attack on the stainless steel tubes. To adequately address these concerns, the new HX will have a 304 stainless steel shell, Titanium tubes, and AL6XN channel ends. The tube sheet will be stainless steel (AL6XN) with mechanically bonded Titanium cladding, so the Titanium tubes can be rolled and welded to the tube sheet. Raw water (RN) will now be on the tube side. The NS water will now be on the shell side. For ease in cleaning, the new HX design will not have V-tubes as the current design does, and the new HX will have access ports at each end. Currently, the piping to and from the HX is carbon steel, which has degraded due to exposure to raw water. New 6 percent Molybdenum nickel alloy piping (AL6XN) will be installed on a section of piping within the NS HX room contained within (but not completely up to) the HX isolation valves 2RN-144A and 2RN-148A, which are normally closed.
The replacement NS HX 2A has some different performance characteristics. The heat transfer capability has been increased by 25 percent resulting in a need to increase the RN System design temperature from 150 deg F to 160 deg F from the HX discharge to where the 18 inch RN header ties into the 42 inch RN common header for NS HX 2A and KC HX 2A. The pressure drop across the NS 2A HX is also reduced due to the two-pass straight-tube design compared to the V-tube design and swapping the NS flow from the tube side to the shell side. The lower pressure drop results in a slightly higher NS flow and corresponding impact on RWST depletion during the injection phase of ECCS operation. Per the UFSAR 6.3.2.8 change in the Final Scope Document (FSD), this has been evaluated to be acceptable with respect to RWST vortexing avoidance. Section 6.3.2.8 of the UFSAR claims "The Low Level setpoint provides a volume above the no vortex level to account for the maximum RWST outflow during switchover with the most limiting single failure (approximately 116,292 gallons) plus an allowance for instrument error" (approximately 11,315 gallons). No setpoint changes have been made per CN-21446/01. The volume available in the RWST at the time of securing the ECCS pumps decreased 2.1 percent per the UFSAR change provided in the FSD, as a result of CN-21446/01. Based on the information provided in the UFSAR 6.3.2.8 and the UFSAR
U. S. Nuclear Regulatory Commission April 1, 2005 Page 24 of 31 change provided in the FSD, the minimum volume to prevent vortexing increased by 7/10 of 1 percent or 0.007. Consideration of both of these changes reveals a reduction in margin to vortexing of approximately 6 percent. This change is compatible with the "no more than minimal increase standard" discussed in NEI-96-07 Rev. 1.
The increased NS heat transfer characteristics (flow, area) have been evaluated for impact on the Containment Minimum Backpressure Analysis presented in the UFSAR 6.2.1.5.
There is no adverse effect.
The NS heat exchangers are not accident initiators. The NS heat exchangers perform the accident mitigation function of containment heat removal capability for the Ice Condenser Reactor Containment. The containment sump performance has been evaluated with respect to vortexing, since an increased flow demand occurs as a result of the higher NS flow (lower tube bundle delta-P). The large margin that remains (18,601 gallons) is judged to be compatible with "no more than a minimal increase" in sump malfunction potential. This modification will not degrade the accident response of any QA-I SSCs (NS HXs, RN ) which perform accident mitigation functions. No new malfunctions are created since all SSCs remain within their applicable design criteria. Specifically, the heat transfer provided by the new NS HX meets or exceeds its design requirement as defined in the UFSAR. The replacement NS Heat Exchangers, piping, valves and supports will perform their UFSAR defined design functions within applicable acceptance criteria. The NS System will remove heat from the Containment by transferring heat from the NS System to the RN System via the replacement NS heat exchangers. No adverse changes to any fission product barriers are being directly imposed by this modification. Therefore, no DBLFPBs as described in the UFSAR are being exceeded or altered. No analytical methods need to be changed in order to demonstrate that this facility continues to meet any required design basis. No methods of evaluation are impacted by this modification.
All of the 50.59 evaluation criteria are satisfied with negative answers such that prior NRC approval is not required and a License Amendment Request is not necessary.
UFSAR changes are required to section 6.2.2.2, Table 6-71, Table 6-1, Table 34, and section 6.3.2.8. No Technical Specification or SLC changes are required.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 25 of 31 Type: Procedure Change Unit: 0
Title:
Procedure Change OP/0/A/6400/006C - Nuclear Service Water System, Revision 244 Change B
Description:
The purpose of this evaluation is to determine if a License Amendment Request (LAR) is required for:
Revision 244 Change B to OP/ON6400/006C, Nuclear Service Water (RN) System Temporary Modification CNTM-0158, RN Pit B Swap Set Point Temporary Modification CNTM-0159, IRN-4B removed from the RN system The Operable-but-Degraded evaluation associated with the non-conservative RN Pit level transfer set point in Technical Specification 3.3.2, Table 3.3.2-1 Function 10, Nuclear Service Water Suction Transfer - Low Pit Level.
Evaluation: I RN44B, which provides an assured suction flow path from the SNSWP to the B Train RN pumps is being maintained in the open position with power removed or is removed from the system, and the lake isolation valves I RN-5A and I RN-6B are being maintained closed with power removed. This is due to internal problems with the valve. This has necessitated aligning the suction and discharge RN flow path to the SNSWP. This is an unacceptable long term solution since the SNSWP will heat up over time. Technical Specifications place an upper limit on the SNSWP temperature of 91.5 degrees F. To reduce the heat up to the SNSWP, the normal RN system operation is being altered to allow the A Train of RN to take suction from Lake Wylie and discharge to Lake Wylie.
The suction of B Train of RN will be aligned to the SNSWP while the discharge is aligned to Lake Wylie. This condition is acceptable as long as B Train of RN is not running. Gravity draining of the SNSWP will not occur since the Pit B lake isolation valves I RN-5A and 1 RN-6B are maintained closed with power removed. Also, due to the design of the RN pumps, it is impossible for a 26 foot column of water to be maintained in the pump casing and siphon from the pump pit up through the pump discharge isolation valve and check valve, into the RN system, and ultimately to the lake.
The pump discharge isolation valve and check valve are closed when the pump is off.
Some events (Ss signal, Loss of Offsite Power, and Loss of Lake Wylie) automatically start the standby RN pumps. This start of the B Train RN pumps will cause the SNSWP to be pumped to Lake Wylie. To prevent the draining of the SNSWP below Technical Specification (TS) limits (571 feet elevation), the B Train RN low pit level auto swap will be adjusted (Temporary Modification CNTM-0158) from a current nominal value of 557.5 feet to 571.5 feet. This alternate RN system alignment and the supporting modification and procedures are temporary and will be in place until valve IRN14B is repaired and reinstalled, prior to or during the next unit 2 refueling outage (2EOC 13). At that time the temporary modification will be removed, and the RN system will be returned to its normal configuration and alignment.
The procedure change, temporary modification, and the Operable but Degraded evaluation do not significantly increase any risk and can be implemented without prior NRC approval. There are no Technical Specification or SAR documents that will need to be revised.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 26 of 31 Type: Procedure Change Unit: 0
Title:
Procedure Change OP/l/A/6400/006F - RN Inlet to NS HXl IA Flush, Enclosure 4.23, and RN Inlet to NS HX I B Flush, Enclosure 4.24, Revision 44
Description:
This evaluation is performed to determine if a License Amendment Request (LAR) is required for Enclosure 4.23, RN Inlet to NS HX IA Flush" and Enclosure 4.24, RN Inlet to NS HX l B Flush of procedure OPIO/AJ6400/006F, rev 44. The procedure enclosures were developed to flush any residual clams or corrosion products from the RN supply piping to NS HX IA and NS HX 1B, The flush of the RN supply line to NS IA HX will use an 8 inch diameter manual valve, ]RN-F79. This valve is connected to 12 inch polyethylene piping, which is routed to the RN shared return header via 10 inch valve IRN-F88 and 12 inch valve IRN-F83. Valve IRN-F83 will be used to throttle flush flow to less than 6000 gpm.
The flush for the NS HX IB supply piping is identical, but the flush will be through 8 inch valve IRN-F80 and into 12 inch polyethylene piping which is connected to common NS HX polyethylene flush piping, and through valves IRN-F88 and IRN-F83 into the RN shared return header. The flushing of these RN piping sections wvill require passing enough flow so that operability of essential header components may not be maintained.
For the flush of NS HX IA supply piping, this could adversely affect the RN flow rate to KC HX IA,CA assured makeup train IA,KD HX IA,NS HX I A and YC Chiller A. For the flush of NS HX lB supply piping, this could adversely affect the RN flow rate to KC HX I B, CA assured makeup train I B, 'KD HX I B, NS HX I B and YC Chiller B.
There will be a compensatory measure on I RN-F83 so that the valve is closed in case of a Safety Injection on either unit, a Loss of OffMite Power on either unit, a fire in the Auxiliary Building, or Flush header leakage/rupture resulting in Auxiliary Building flooding that presents a challenge to sump capacity. These procedure enclosures also secure the flush if any of the operators stationed for the flush need to leave their dedicated location for any reason. Either flush is secured upon a Safety Injection on either unit, a Loss of Off site Power on either unit, a fire in the Auxiliary Building or Flush header leakage/rupture resulting in Auxiliary Building flooding that presents a challenge to sump capacity by manually closing valve IRN-F83 within 10 minutes.
Evaluation: Enclosure 4.23 RN Inlet to NS HX IA Flush has been developed for procedure OP/O/
A/6400/006F rev 44. The procedure enclosure was developed to flush any residual clams or corrosion products from the RN supply header to NS HX IA. Two different flushes are to be performed. One through an 8 inch diameter manual valve, IRN-F79, which flushes the RN supply line to NS HX IA. Enclosure 4.24 RN Inlet to NS HX 1B Flush has been developed for to flush the RN supply line to NS HX lB via IRNF80. The flushing of these RN piping sections will require passing enough flow so that operability to the affected components may not be maintained.
This could adversely affect the RN flow rate to KC HX IA, CA assured makeup train IA, KD HX IA,NS HX IA and YC Chiller A. The flush will have to be secured for a Safety Injection on either unit, a Loss of Offsite Power on either unit, a fire in the Auxiliary Building, or flush header leakage/rupture resulting in Auxiliary Building flooding that presents a challenge to sump capacity. These procedure enclosures also secure the flush if any of the operators stationed for the flush need to leave their dedicated location for any reason.
Enclosure 4.23, RN Inlet to NS HX IA Flush, is discussed. The evaluation also applies to
U. S. Nuclear Regulatory Commission April 1, 2005 Page 27 of 31 Enclosure 4.24, RN Inlet to NS HX IB Flush, but the flush will be accomplished via IRN-F80 versus IRN-F79, and the discussion of affected Train A components also applies to Train B components.
The procedure enclosures and the guidance they contain do not significantly increase any risk and can be used without prior NRC approval. There are no Technical Specification or SAR documents that will need to be revised due to these procedure enclosures.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 28 of 31 Type: Temporary Modification Unit: 2
Title:
Temporary Modification CNTM-00 153 - Temporary support of the lower backup screen in the 2C cooling tower outlet basin
Description:
The purpose of this temporary modification is to document the backup screen stuck in the 2C Cooling Tower Outlet Basin and to document the temporary support configuration used to support the backup screen. During the cleaning of the 2C Cooling Tower primary outlet screens the lower backup screen was found misaligned in the screen guides. The screen guides consist of a 5 inch by 5 inch by three-eighth inch angle on the down stream side and 5 inch by 5 inch by three-eighth inch lube steel on.the upstream side. During the removal of the lower backup screen it was discovered that the screen was no longer between the screen guides on the southeast side. The guide angle was bent and wedged between the flanges of the screen channel. The top of the screen could be raised to the water level and no further. The screen is temporarily supported from the existing platformn.
Evaluation: The stuck backup screen is located in the 2C Cooling Tower Basin. The cooling tower basin is part of the Condenser Circulating Water System (RC). The RC System supplies cooling water to the main and feed water pump turbine condensers to condense the turbine exhaust steam. The rejected heat from the condensers is dissipated to the ambient surroundings by the cooling lowers. The RC System is not assigned a safety class as it is not required for a safe reactor shutdown. The cooling towers are located such that their structural failure due to a seismic event, a tornado, or any other natural phenomenon could not damage any safety related structure, system, or component.
The screen stuck in the 2C cooling lower basin will not affect the function of the RC system in its current condition since it is down stream of the primary screens and would not degrade. However, te stuck screen will prevent the primary screens from being cleaned normally. This could lead to a reduction of RC flow as the primary screens degrade. This loss of flow would reduce the Main and Feedwater Pump Condensers ability to maintain vacuum. The loss of condenser vacuum would cause a Turbine Trip and a Reactor Trip depending on reactor power. The buried portion of the RC piping is also a non-safety backup supply to the Auxiliary Feedwater System (CA) during SSF operations.
The current method of cleaning the primary screens is not described in the UFSAR, only the function of the RC system and cooling towers. This temporary modification will be in place until the upcoming unit 2 outage (2EOC 13). If the primary screens require cleaning before the upcoming outage, the described function of the 2C cooling tower could be affected. The primary cooling tower screens would degrade slowly causing the basin level in the 2C cooling tower to increase and level in the 2A and 2B basins to decrease.
The cooling tower level is monitored from the control room by annunciators (Panel 2AD-8, Annunciators B/4 and B/5. Per the annunciator response, model work orders to clean the outlet screens are issued if the high level is determined not to be from excessive makeup. The screen would be cleaned to the extent possible using an alternate method.
If the flow/level could not be restored to normal, operator action using existing procedures would be taken to reduce power and remove the affected tower from service before condenser vacuum would be affected. This would be a normal response to a degraded screen that could not be cleaned for any reason. The degraded screens would not affect the volume of water available for the CA system since it would not affect the volume of water in the buried piping.
The flow from the outlet basin passes into a 9 foot diameter pipe which is flared to 11
U. S. Nuclear Regulatory Commission April 1, 2005 Page 29 of 31 foot 8.375 inch diameter in the basin. Based on the size of the bottom screen (12 foot 6 inch x 12 foot 7 inch), the screen could not pass through the opening of the outlet piping causing damage to outlet piping components or the condenser. Since the condition of the screen guide on the southeast side could not be determined, the screen was temporarily supported from the existing platform to prevent the possibility of blocking the outlet pipe.
The proposed activity does not affect the current licensing basis as described in the UFSAR for the RC system or the cooling towers. This evaluation does not involve an Unreviewed Safety Question. No UFSAR or Technical Specification changes are required.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 30 of 31 Type: Temporary Modification Unit: 0
Title:
Temporary Modification CNTM-0158 - RN Pit Swap Setpoint
Description:
The purpose of this evaluation is to determine if a License Amendment Request (LAR) is required for:
Revision 244 Change B to OP/O/N6400/006C, Nuclear Service Water (RN) System Temporary Modification CNTM-O 158, RN Pit B Swap Set Point Temporary Modification CNTM-0159, 1RN-4B removed from the RN system The Operable-but-Degraded evaluation associated with the non-conservative RN Pit level transfer set point in Technical Specification 3.3.2, Table 3.3.2-1 Function 10, Nuclear Service Water Suction Transfer - Low Pit Level Evaluation: IRN-4B, which provides an assured suction flow path from the SNSWP to the B Train RN pumps is being maintained in the open position with power removed or is removed from the system, and the lake isolation valves I RN-5A and 1RN-6B are being maintained closed with power removed. This is due to internal problems with the valve. This has necessitated aligning the suction and discharge RN flow path to the SNSWP. This is an unacceptable long term solution since the SNSWP will heat up over time. Technical Specifications place an upper limit on the SNSWP temperature of 91.5 degrees F. To reduce the heat up to the SNSWP, the normal RN system operation is being altered to allow the A Train of RN to take suction from Lake Wylie and discharge to Lake Wylie.
The suction of B Train of RN will be aligned to the SNSWP while the discharge is aligned to Lake Wylie. This condition is acceptable as long as B Train of RN is not running. Gravity draining of the SNSWP will not occur since the Pit B lake isolation valves I RN-5A and I RN-6B are maintained closed with power removed. Also, due to the design of the RN pumps, it is impossible for a 26 foot column of water to be maintained in the pump casing and siphon from the pump pit up through the pump discharge isolation valve and check valve, into the RN system, and ultimately to the lake.
The pump discharge isolation valve and check valve are closed when the pump is off.
Some events (Ss signal, Loss of Offsite Power, and Loss of Lake Wylie) automatically start the standby RN pumps. This start of the B Train RN pumps will cause the SNSWP to be pumped to Lake Wylie. To prevent the draining of the SNSWP below Technical Specification (TS) limits (571 feet elevation), the B Train RN low pit level auto swap will be adjusted (Temporary Modification CNTM-0 158) from a current nominal value of 557.5 feet to 571.5 feet. This alternate RN system alignment and the supporting modification and procedures are temporary and will be in place until valve I RN34B is repaired and reinstalled, prior to or during the next unit 2 refueling outage (2EOC 13). At that time the temporary modification will be removed, and the RN system will be returned to its normal configuration and alignment.
The procedure change, temporary modification, and the Operable but Degraded evaluation do not significantly increase any risk and can be implemented without prior NRC approval. There are no Technical Specification or SAR documents that will need to be revised.
U. S. Nuclear Regulatory Commission April 1, 2005 Page 31 of 31 Type: Temporary Modification Unit: 0
Title:
Temporary Modification CNTM-0159- IRN4B Removed from the RN System
Description:
The purpose of this evaluation is to determine if a License Amendment Request (LAR) is required for:
Revision 244 Change B to OP/O/N6400/006C, Nuclear Service Water (RN) System Temporary Modification CNTM-0 158, RN Pit B Swap Set Point Temporary Modification CNTM-0159, IRN-4B removed from the RN system The Operable-but-Degraded evaluation associated with the non-conservative RN Pit level transfer set point in Technical Specification 3.3.2, Table 3.3.2-1 Function 10, Nuclear Service Water Suction Transfer - Low Pit Level.
Evaluation: 1RN-4B, which provides an assured suction flow path from the SNSWP to the B Train RN pumps is being maintained in the open position with power removed or is removed from the system, and the lake isolation valves IRN-SA and IRN-6B are being maintained closed with power removed. This is due to internal problems with the valve. This has necessitated aligning the suction and discharge RN flow path to the SNSWP. This is an unacceptable long term solution since the SNSWP will heat up over time. Technical Specifications place an upper limit on the SNSWP temperature of 91.5 degrees F. To reduce the heat up to the SNSWP, the normal RN system operation is being altered to allow the A Train of RN to take suction from Lake Wylie and discharge to Lake Wylie.
The suction of B Train of RN will be aligned to the SNSWP while the discharge is aligned to Lake Wylie. This condition is acceptable as long as B Train of RN is not running. Gravity draining of the SNSWP will not occur since the Pit B lake isolation valves I RN-5A and I RN-6B are maintained closed with power removed. Also, due to the design of the RN pumps, it is impossible for a 26 foot column of water to be maintained in the pump casing and siphon from the pump pit up through the pump discharge isolation valve and check valve, into the RN system, and ultimately to the lake.
The pump discharge isolation valve and check valve are closed when the pump is off.
Some events (Ss signal, Loss of Offsite Power, and Loss of Lake Wylie) automatically start the standby RN pumps. This start of the B Train RN pumps will cause the SNSWP to be pumped to Lake Wylie. To prevent the draining of the SNSWP below Technical Specification (TS) limits (571 feet elevation), the B Train RN low pit level auto swap will be adjusted (Temporary Modification CNTM-0158) from a current nominal value of 557.5 feet to 571.5 feet. This alternate RN system alignment and the supporting modification and procedures are temporary and will be in place until valve I RN-4B is repaired and reinstalled, prior to or during the next unit 2 refueling outage (2EOC13). At that time the temporary modification will be removed, and the RN system will be returned to its normal configuration and alignment.