ML090210728
| ML090210728 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 01/21/2009 |
| From: | Brandon M Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML090210728 (13) | |
Text
January 21, 2009 10 CFR 50.59(d)(2)
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket No. 50-390 Tennessee Valley Authority )
WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - 10 CFR 50.59, CHANGES, TESTS, AND EXPERIMENTS AND COMMITMENT
SUMMARY
REPORTS, SUPPLEMENTAL REPORT This letter supersedes TVAs letter, Watts Bar Nuclear Plant (WBN) Unit 1 - 10 CFR 50.59, Changes, Tests, and Experiments and Commitment Summary Reports, dated September 24, 2008. This letter includes two additional 50.59 safety evaluations that should have been included in the September report. These omissions were identified in a recent self-assessment and have been entered for resolution into TVAs corrective action program as Problem Evaluation Report 160317. Additional clarifications were also made regarding the conclusions of two evaluations.
Pursuant to 10 CFR 50.59(d)(2), this letter provides the complete Summary Report of the implemented changes, tests, and experiments in which evaluations were performed in accordance with 10 CFR 50.59(c). The enclosure provides a summary of the evaluations associated with Updated Final Safety Analysis Report Amendment 7 (provided separately) and includes other evaluations implemented during the period from March 17, 2007 to June 30, 2008.
During this reporting period, there were no previous commitments that TVA has evaluated and revised using administrative controls that incorporate the Nuclear Energy Institutes (NEI) 99-04 Guideline For Managing NRC Commitments.
U.S. Nuclear Regulatory Commission Page 2 January 21, 2009 There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please call me at (423) 365-1824.
Sincerely, Original signed by M. K. Brandon Manager, Site Licensing and Industry Affairs Enclosure cc: See Page 3
U.S. Nuclear Regulatory Commission Page 3 January 21, 2009 Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303
U.S. Nuclear Regulatory Commission Page 3 January 21, 2009 MKB:
Enclosure cc (Enclosure):
NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 U.S. Nuclear Regulatory Commission Mr. John G. Lamb, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS O-8 H1A Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 G. Arent, EQB 1B-WBN G. A. Boerschig, MOB 2R-WBN W. R. Campbell, LP 3R-C W. T. Coutu, LP 3R-C D. E. Jernigan, LP 3R-C M. T. McFadden, ADM 1Q-WBN G. W. Mauldin, EQB 2A-WBN L. E. Nicholson, LP 3R-C M. A. Purcell, LP 4K-C NSRB Support, LP 4K-C (including Advisors)
A. J. Scales, MOB 2M-WBN J. E. Semelsberger, EQB 2W-WBN M. D. Skaggs, ADM 1V-WBN J. F. Tortora, Jr., ADM 1B-WBN E. J. Vigluicci, WT 6A-K K. W. Whittenburg, SP 2B-C Sequoyah Licensing Files, OPS 4C-SQN EDMS, WT 3B-K M:\SUBMIT\Cycle 8 50.59 Report Supplement.pdf
ENCLOSURE WATTS BAR NUCLEAR PLANT UNIT 1 10 CFR 50.59
SUMMARY
REPORT
- 1. SA-SE Number: WBPLMN-07-008-1
- 2. SA-SE Number: WBPLMN-05-026-0
- 3. SA-SE Number: WBPLEE-07-022-1
- 4. SA-SE Number: WBPLMN-07-014-0
- 5. SA-SE Number: WBPLMN-08-006-0
- 6. SA-SE Number: WO 07-814239-000
- 7. SA-SE Number: WBPLMN-08-004-0
- 8. SA-SE Number: WBPLMN-08-005-0 E-1 of 9
SA-SE Number: WBPLMN-07-008-1 Implementation Date: 03/29/2007 Document Type: Affected Documents:
Title:
Temporary Temporary Alteration 1 Post-Accident Safety Function of Alteration Change 0002-065, R1 Emergency Gas Treatment Form (TACF) FSAR Change Package System (EGTS) 1915 TS Bases change package 2007-02 Description and Safety Assessments:
The following issue was documented in TVAs corrective action program as a problem evaluation report (PER):
Constant Operator monitoring will be required if, following an accident, the containment isolation phase A (CIA) signal is reset prior to annulus differential pressure decreasing to the setpoint needed for the EGTS pressure control loops isolation valves to open. EGTS will not perform its safety function without these valves opened as required. This was identified by Operators during simulator exercises and has been identified as a Priority 1 Operator Work Around.
The EGTS consists of two separate control loops. The hand switches of the control loops can be placed in the A-AUTO or A-AUTO STANDBY positions. The annulus pressure setpoint for the A-AUTO STANDBY position is -0.812 inches water gauge (wg). If a large break Loss of Coolant Accident (LOCA) occurs, the annulus pressure will increase from -5 inches wg to approximately -1 inch wg due to effects from the LOCA. The annulus pressure will continue to increase to the -0.812 inches wg setpoint since neither the annulus vacuum control system nor EGTS is operating to reduce the annulus pressure; i.e. the annulus vacuum control system is turned off by a safety injection system signal initiated subsequent to a LOCA. When the -0.812 inches wg EGTS setpoint is reached, both EGTS control loops will operate to filter the airborne vapors and particulates in the annulus. The off site and main control room doses will remain less than the 10 CFR 50, Appendix A, GDC 19 control room limits and the 10 CFR 100 offsite dose limits.
If a small break LOCA occurs, the annulus pressurization effects will not be as pronounced as with a large break LOCA. In this event, the annulus pressure may remain less than the -0.812 inches wg setpoint until the CIA signal is reset by operations. When the CIA signal is reset, the EGTS control loops in the A-AUTO STANDBY position are disabled. In this case, no EGTS train will operate.
Consequently, the EGTS will not function as described in the Updated Final Safety Analysis Report (UFSAR) and the EGTS safety functions will not be performed.
The primary safety function of the EGTS is to maintain the offsite and Main Control Room (MCR) doses less than the allowable 10 CFR 100 and 10 CFR 50, Appendix A General Design Criteria 19 limits during a LOCA. This safety function is accomplished during a LOCA by (1) keeping the annulus pressure below the outside atmospheric pressure and (2) removing airborne particulates and vapors that may contain radioactive nuclides from air drawn from the annulus and exhausted to the atmosphere. These safety functions will be accomplished with implementation of the TACF corrective actions assuming a credible failure of one train of EGTS exhausting to the outside environment for the duration of a LOCA accident.
These changes will increase the reliability because active components in the auxiliary relay circuit that are currently required to open the valves will be eliminated. No failure assumptions are changed, and no accident frequencies are increased. Specific changes addressed in this TACF do not create the possibility of a malfunction with different than currently evaluated results. Additionally, no fission product barriers are affected.
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SA-SE Number: WBPLMN-05-026-0 Implementation Date: 06/25/2007 Document Type: Affected Documents:
Title:
Engineering EDC 51619-A Main Steam Valve Vault Airflow Document Change Rates (EDC)
Description and Safety Assessments:
The airflow rates of the exhaust fans serving the Main Steam Valve Vaults (MSVVs) are outside of the
+/-10% requirement stipulated by General Specification G-37. The fans are designed to maintain an adequate temperature environment for the main steam safety related valves, but are non-safety related. Design Change Notice (DCN) M18106-A added ductwork to the suction side of these fans, which caused the fan static pressures to be considerably increased. Because of this, air flow rates were significantly reduced. During the periodic tests, the fan speeds were increased to meet the design airflow rates.
Problem Evaluation Report (PER) 03-006901-000 revealed that the fan speeds were higher than those permitted in Technical Instruction-5.002, Flow Testing of Ventilation Systems. Further investigation revealed that the fan speed limits provided in Table 1 of the Technical Instruction may exceed the structural classification limits of the MSVV exhaust fan housings. This was confirmed through a review of the contract documentation and discussions with the manufacturer. It was advised to limit the fan speed to 535 rpm.
Temperature data was collected for a limited time period within the WBN MSVVs to justify the acceptability of the reduced total flow rates. In the north MSVV, temperature values remained unchanged, so the reduced exhaust fans test flow rates are considered acceptable. In the south MSVV, hot spot locations were generally greater than current values. These changes were updated via Drawing Change Authorization-51619-002. The hot spot changes do not affect the Minimum, Normal Average, Normal Maximum and/or Abnormal Maximum temperatures. Quality of life of components was re-evaluated based on the new hot spot temperatures. The changes required were as follows: 1) Four cables in binder WBNEQ-CABL-049 reduced their qualified life span from 34.5 years to 25.2 years. 2) The qualified life for the o-ring seals within the motor operated valves was changed from 35.39 years to 30.47 years. 3) Solenoid valve life span was reduced from 30.49 years to 27.32 years. 4) EQ Binder WBNEQ-IZS-003 was revised from 9.17 years to 8.17 years, EQ Binder WBNEQ-IZS-007 was revised from 7.94 years to 7.06 years, and the qualified life for certain limit switches listed in EQ Binder WBNEQ-IZS-005 was revised from 7.76 years to 7.05 years and from 9.89 years to 7.67 years.
The proposed changes do not introduce the possibility of a change in the frequency of an accident and no new failure modes are introduced. Operation of the MSVV exhaust fans within the acceptable limits based on the manufacturers specifications inherently reduces the possibility of their malfunction, and therefore this change will decrease the likelihood of a malfunction. Based on the analyses, the Heating Ventilation Air Conditioning system will still perform its intended design basis functions. No changes are made that would increase the consequences of any existing postulated malfunction, and the reduction in air flow rates does not have the potential to create a new accident.
No fission product barriers are unduly challenged due to this change.
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SA-SE Number: WBPLEE-07-022-1 Implementation Date: 10/30/2007 Document Type: Affected Documents:
Title:
Design Change DCN 52233-A Tornado Watch/Warning Notice (DCN)
Final Safety Analysis Report (FSAR) Change Pkg. 1919 Description and Safety Assessments:
During the Component Design Basis Inspection (CDBI), the NRC identified a potential problem with the low temperature switches associated with the Diesel Generator (DG) room exhaust fans and the requirement that the fans be energized and running during a tornado watch/warning condition. Since the low temperature switches have a re-set setpoint value of 73°F and a setpoint of 68°F, the potential exists that attempts to start the fans and open the associated dampers may be unsuccessful, or that the fans may start initially, but later stop and close the dampers when they are required to be open. If the temperature switches have cooled down below their setpoint value of 68° F, the switches will prevent manual start of the fans until the re-set setpoint value of 73°F is reached. Problem Evaluation Report (PER) 120005 was initiated to document the issue and establish the required corrective actions. As part of the immediate actions for the PER, Abnormal Operating Instruction (AOI) 8, Tornado Watch or Warning, was revised to require installation of a jumper wire in each of the DG room exhaust fan Motor Control Center panels during a tornado watch. This was done in accordance with a Temporary Alteration (TA) created by a technical procedure per Site Programs and Procedures-9.5, Temporary Alterations. DCN 52233-A supersedes this TA by installing tornado bypass hand switches for each of the exhaust fans as explained below.
The design change is being issued to install a new tornado bypass hand switch in the control circuit for each of the diesel generator room exhaust fans in order to bypass the low temperature switch function during a tornado watch or warning in accordance with AOI-8. During normal operating conditions, the new hand switch will be placed in the NORMAL position allowing the low temperature switch to control their respective fan, as necessary, in order to maintain the diesel generator room exhaust temperature between 50°F and 120°F. In response to the requirements being added to Section 5.6 of System Description N3-30DB-4002 and captured in AOI-8, the new tornado bypass hand switch will be placed in the BYPASS position to allow the DG room exhaust fans to start and cause the associated inlet and exhaust dampers to open at temperatures less than the 68°F setpoint during tornado conditions.
If the outside temperature is expected to drop below 50°F during a tornado warning, then a start of each DG is required to be performed to ensure DG functionality. Monitoring the outside air temperature is required during tornado warnings to assure the DGs are started prior to a drop in temperature below 50°F. When tornado conditions are cancelled, the exhaust fans are required to be turned off and the hand switches will be returned to NORMAL promptly, and independently verified in accordance with AOI-8. Based on the evaluation of historical data, it is very unlikely that a tornado will occur with a temperature concurrent or lower than 50°F. Therefore, this change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated FSAR.
The change supports successful completion of AOI-8 actions to open each DG room exhaust fan supply and dampers and start each exhaust fan. The change does not result in new accidents or malfunctions, and it does not increase the frequency or consequences of accidents or malfunctions evaluated in the UFSAR. Additionally, no fission product barriers are challenged by the change.
E-4 of 9
SA-SE Number: WBPLMN-07-014-0 Implementation Date: 03/13/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52220-A ABI/CVI Intertie Notice (DCN)
Description and Safety Assessments:
DCN 52220-A modifies the Train A and Train B electrical circuits for high radiation in the refueling area logic bus and the Train A and Train B Solid State Protection System (SSPS) input for high radiation in the containment purge air exhaust, which initiates Containment Vent Isolation (CVI). The modification to these circuits allows WBN to maintain the Auxiliary Building Secondary Containment Enclosure (ABSCE) in the event of a high radiation signal in either the refueling area or the containment purge exhaust during refueling operations while the containment and/or annulus is open to the auxiliary building ABSCE spaces. This change will permit, but not require, operation of containment purge when moving irradiated fuel in the Auxiliary Building with containment and/or annulus hatches, personnel hatches, or penetrations open to the ABSCE spaces.
This change does not allow the equipment hatch to be open during fuel movement inside containment since Technical Specification 3.9.4, Refueling Operations, Containment Penetrations, requires the penetration to be closed during this time.
The electrical circuit modifications are implemented in two Auxiliary Relay Racks. One hand switch for each train will be added to the appropriate rack to allow the circuits to be swapped from a normal mode to a refueling mode. During normal operation, the circuits will perform the same logic as before the modification. Prior to starting refueling operations, the hand switch will be placed into the refueling mode, and it will be returned to the normal mode prior to entering Mode 4 from Mode 5.
During the refueling mode, the ABSCE isolation valves will be closed by receiving either the current high radiation signal from the spent fuel pool accident radiation monitors or upon a CVI signal. The Containment Ventilation Isolation Valves that are currently closed by a CVI signal will also be closed upon receiving a high radiation signal from the spent fuel pool accident radiation monitors. Isolation of these valves will establish the ABSCE boundary, prevent the potential for back flow from the Shield Building exhaust vent, and ensure that radioactive releases due to fuel handling accidents in containment will be processed by the Auxiliary Building Gas Treatment System (ABGTS) if they migrate into the Auxiliary Building.
The response time requirement for the spent fuel pool radiation monitors is not affected by this modification. A new response time requirement for the containment purge monitors during refueling operations is required. The new response time requirements do not affect the calculated off site or main control room dose analyses for the fuel handling accident.
This change complies with the safety and functional requirements specified in the applicable design bases documents and does not adversely affect the performance of any safety related equipment.
The proposed modifications do not increase the frequency or likelihood of accidents or malfunctions, increase the consequences of an accident or malfunction beyond the ten percent allowed by 10CFR50.59, or create a new type of accident. The design bases for fission product barriers will not be altered or exceeded and no new methods of evaluation were used in evaluating the proposed modifications.
E-5 of 9
SA-SE Number: WBPLMN-08-006-0 Implementation Date: 04/24/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52307-A Lower Compartment Coolers Notice (DCN) Blank Off Plate Installation Description and Safety Assessments:
The design function of the Lower Compartment Coolers (LCCs), along with the Control Rod Drive Mechanism (CRDM) coolers, is to maintain acceptable temperature within the reactor building lower compartment areas for the protection of equipment and control during normal reactor operation and shutdown. Any combination of three LCCs, or any combination of two LCCs and two CRDM coolers are required for safe shutdown per 10CFR50 Appendix R to prevent containment temperatures from exceeding established Equipment Qualification profiles used in the 10CFR50.49 program.
The LCC 1D-B upper outboard coil leaks at or near the header/tube interface and cannot be repaired in place. This design change will blank off the supply and return of the Essential Raw Cooling Water (ERCW) to this cooling coil, resulting in seven of the eight coils remaining functional. Blank off plates of stainless steel that are capable of withstanding the ERCW system design pressure will be installed at the supply and return flanges associated with the leaking coil. The change will isolate ERCW flow to the offending coil only and will not adversely impact continued flow to the remaining seven coils.
The piping system will remain seismically qualified with the blanking plates installed.
If the water leak in the offending coil is allowed to continue, an increase in the relative humidity levels inside lower containment is expected. This condition could mask water leaks inside lower containment. Failure of the blank off plates is less likely to occur than the failure of the existing coil components, and the installation of the plates eliminates the potential for the coil to rupture. This change will prevent one of the coils from performing its cooling function. Calculations have been revised to note that the capacity reduction of 1/8 on the cooler will not affect the current requirements.
Any combination of three LCCs or two LCCs and two CRDM coolers remains acceptable and does not affect Technical Specification requirements.
Proper operation of the LCC cooling coils is not a safety related function and is not required for the mitigation of any Updated Final Safety Analysis Report (UFSAR) Chapter 6 or 15 accidents. The blanking plate will be designed to meet the design conditions of the ERCW system. The design basis functions of the LCC 1D-B is maintained and is not adversely impacted by this proposed design change. Therefore, this change does not result in any new accidents or malfunctions, and does not result in increased frequency or consequences of accidents or malfunctions evaluated in the UFSAR.
In addition, no fission product barriers are challenged by this change. Safety related functions of the LCCs are not affected by this proposed modification.
E-6 of 9
SA-SE Number: WO 07-814239-000 Implementation Date: 05/29/2008 Document Type: Affected Documents:
Title:
Maintenance Work Work Order 07-814239-000 Fuel Cleaning Order (WO) Areva Procedure FS-235 Description and Safety Assessments:
During Cycle 7 operation, the WBN Unit 1 core exhibited symptoms of Crud Induced Power Shift (CIPS). The root cause of this is boiling near the upper surfaces of high-powered fuel rods, leading to the depositing of crud in these locations. The Ultrasonic Fuel Cleaning System (UFC) was developed by Dominion Engineering, Inc. with Electric Power Research Institute (EPRI) to remove the corrosion products from Pressurized Water Reactor (PWR) fuel using ultrasonic energy. Laboratory and field experience has shown an average crud removal rate of greater than 80%. All UFC operations will be performed underwater in the Unit 2 side of the transfer canal at WBN. Lifting and handling of the UFC system and its support stand will be performed in accordance with NUREG 0612 to mitigate the potential for damage to stored fuel.
The existing potential for pellet-clad interaction resulting from agitation of fuel pellet fragments that were cracked prior to cleaning has been evaluated and determined to be very unlikely. Normal administrative and procedural controls will be employed to ensure fuel is not damaged. Because of this evaluation and administrative control, the risk of damaged fuel or an increased risk of a fuel handling accident is negligible. Additionally, the cleaning apparatus is not located in an area where it can fall on any safety-related equipment besides the transfer canal liner.
The increased risk of crud particle release into the Unit 2 transfer canal is mitigated by the current operating procedures that require the fuel to be secured safely and cleaning suspended until pool visibility is restored. The potential to release dissolved radioactive gasses is minimized by design.
The average flow velocity in the cleaning apparatus is sufficiently high to entrain all gas bubbles smaller than 3 mm in diameter. These bubbles are mixed vigorously in the connection hose between the cleaner and the filter skid. This mixing action causes most of the gasses to re-dissolve into the bulk fluid. Any that are not re-dissolved have been shown to not cause a significant radiological hazard.
The possibility of a fuel assembly becoming stuck in the cleaning fixture is very remote. It is protected against by the small amount of force required to straighten a 1 inch bow in a fuel assembly, which can be provided by the sidewalls of the cleaning fixture. Also, the fuel can be indefinitely cooled in the cleaning apparatus by natural convection, as previously discussed.
The Spent Fuel Cooling and Cleaning System is a safety related system required to mitigate the consequences of Design Basis Events. During operation of the UFC system, fuel assemblies are placed in the fuel cleaning chambers and bombarded with high frequency pressure oscillations. This adds negligible heat loads to the system and will not prevent the spent fuel cooling system from performing its intended design function. Additionally, the UFC system is designed such that natural circulation will cool a fuel assembly in the UFC chamber in the event of a station blackout.
All fuel handling will be performed within the bounds of Updated Final Safety Analysis Report (UFSAR) procedures and interlocks. This change does not affect the consequences of an accident or malfunction because it does not affect the dose results of the accident analyses reported in the UFSAR. Because of the normal administrative and procedural controls in place, the increase in probability of a fuel handling accident is negligible. Potential accidents that have not been previously evaluated are not introduced by this change. No fission product barriers are affected by this activity.
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SA-SE Number: WBPLMN 08-004-0 Implementation Date: 03/27/2008 Document Type: Affected Documents:
Title:
Design Change DCN 52211-A Relocation of Radiation Monitor Notice (DCN) Safety Analysis Report for Transfer Canal Fuel Handling (SAR) Change Package Accident Detection 1927 Description and Safety Assessments:
A fuel handling accident (FHA) in the fuel transfer canal may not be detected by the radiation monitors located on the wall of the spent fuel pool below the 757 floor elevation. Consequently, the Auxiliary Building Gas Treatment System (ABGTS) would not be automatically established. As a result of this condition, calculation WBNTSR-009 was revised. The revised calculation determined that such an accident can be mitigated if spent fuel pool accident radiation monitor 0-RE-90-103-B is relocated above the refueling floor elevation and the exhaust vents located around the fuel transfer canal perimeter are isolated during movement of irradiated fuel in that area. Air intake must be isolated to ensure the activity released by a FHA in the transfer canal is not removed and exhausted by the ventilation system prior to radiation monitor detection.
This DCN is only relocating the radiation monitor, not changing its design functions. The mounting method for the new location has been evaluated and has been determined to meet seismic requirements.
Additionally, this DCN includes the isolation of air flow from the transfer canal by closing either a fire damper or a balancing damper during irradiated fuel movement in the transfer canal. Alternate ventilation requirements are required to ensure the fuel handling area ventilation system remains balanced to support both normal and post-accident conditions. The DCN requires opening one or two selected fuel handling area exhaust ductwork access panels and adding a sheet metal plate to restrict air flow through one access opening as needed. The access panel(s) will be secured in place using heavy gauge wire, and stainless steel screens will be installed to prevent debris from entering the ductwork. Static pressure will be measured upstream of the access panel both before and after isolation of flow from the transfer canal. This will verify required normal or post accident air flows are maintained.
Relocation of the spent fuel pool radiation monitor will not affect the safety related operation of the monitor. The proposed changes to the fuel handling area ventilation system as described above will not adversely affect establishment of the Auxiliary Building Secondary Containment Enclosure or operation of the ABGTS during FHA or Loss of Coolant Accident (LOCA) conditions. Thus, the proposed changes do not result in an increase in frequency of occurrence of accidents or likelihood of occurrence of malfunctions evaluated in the Updated Final Safety Analysis Report (UFSAR), and do not result in an increase in the consequences of an accident or malfunction. The proposed changes do not result in an accident of a different type or a malfunction with a different result. Further, the changes do not affect fission product barriers and the method of evaluation described in the UFSAR.
E-8 of 9
SA-SE Number: WBPLMN 08-005-0 Implementation Date: 04/14/2008 Document Type: Affected Documents:
Title:
Engineering Design Change Notice Number 2 Heater Drain Bypass to Document Change (DCN) 52270-A Condenser Level Control Valves (EDC)
Description and Safety Assessments:
When the Number 2 Heater Drain Bypass to Condenser level control valves (LCVs) are opened during high turbine load conditions, violent water hammer occurs. This design change of documentation only allows for the bypass LCVs to be opened only during low power operations to protect against severe water hammer incidences.
The three bypass to condenser lines from the No. 2 heater were not part of original plant design. A condition was identified where the No. 2 heaters did not drain to the No. 3 Heater Drain Tank (HDT) during rapid load reductions at low power operating conditions. To account for this, the bypass lines were installed. The LCVs are normally in the closed position and were designed to automatically open if the heater water level reaches a prescribed high setpoint. This design change will still allow the LCVs to be opened at low power levels. The change will require the bypass lines to be isolated during normal high turbine load operation to prevent water hammer.
If maintenance is needed on the main drain valves, the main drain path bypass valves can be used to maintain No. 2 heater water level. In the event the normal drain path malfunctions during high power operation, the bypass to condenser LCVs will no longer open automatically, and a heater string isolation could occur. It was determined that the reduced feedwater temperature caused by a heater string isolation will not compromise plant safety.
This design change has the potential to increase the frequency of loss of normal feedwater if all three strings were to isolate. Isolation of main feedwater from the steam generators (SGs) would require Operations to start the Auxiliary Feedwater (AFW) system manually, or AFW will automatically start on a low low SG level. The Loss of Normal Feedwater is categorized as a fault of moderate frequency, implying once a year occurrence. Historical operating experience at WBN shows that heater drain failures do not typically occur at a frequency greater than once per year. Further, no new failure modes are introduced that could increase the frequency of occurrence of an isolation of all three intermediate pressure heaters and no new components are added with this design change.
The design basis accident analyses evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected, and the system is not credited for accident mitigation. No equipment required for safe operation or shutdown is changed by this design. No fission product barriers are challenged by this change.
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