ML11215A198
| ML11215A198 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/02/2011 |
| From: | Gillespie T Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LAR-10-03 | |
| Download: ML11215A198 (11) | |
Text
Duke T. PRESTON GILLESPIE, JR.
Ernergy Vice President Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Hwy.
Seneca, SC 29672 August 2, 2011 864-873-4478 864-873-4208 fax T.Gillespie@duke-energy.com U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001
Subject:
Duke Energy Carolinas, LLC Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Response to Request for Additional Information Regarding License Amendment Request for Approval to Operate a Reverse Osmosis System during Unit Operation License Amendment Request (LAR) No. 2010-03, Supplement 3 On November 15, 2010, Duke Energy Carolinas, LLC (Duke Energy) submitted a License Amendment Request (LAR) to request approval to operate a Reverse Osmosis System to remove silica from the Borated Water Storage Tanks and Spent Fuel Pools during Unit Operation at ONS. Duke Energy supplemented the LAR by letter dated February 18, 2011, in response to a request for additional information (RAI) transmitted electronically on December 20, 2010. By electronic mail dated March 16, 2011, and April 24, 2011, NRC requested Duke Energy supplement the LAR with further information. Duke provided the requested information by letter dated May 12, 2011. By electronic mail dated July 11, 2012, NRC requested Duke Energy to supplement the LAR with additional information. Enclosure 1 provides this additional information. Revisions to the significant hazards consideration were necessary as a result of the changes to the LAR made by the LAR supplements. The revised significant hazards consideration is provided in Enclosure 2. There are no Regulatory Commitments made by this LAR supplement.
Inquiries on this proposed amendment request should be directed to Boyd Shingleton of the ONS Regulatory Compliance Group at (864) 873-4716.
I declare under penalty of perjury that the foregoing is true and correct. Executed on August 2, 2011.
Sincerely; T. Preston Gillespie, Jr., Vice President Oconee Nuclear Station
Enclosures:
- 1. Duke Energy Response to NRC Request for Additional Information
- 2. Revised Significant Hazards Consideration www. duke-energy. corn
Nuclear Regulatory Commission August 2, 2011 Page 2 bc w/enclosures:
Mr. Victor McCree, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. John Stang, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, D. C. 20555 Mr. Andy Sabisch Senior Resident Inspector Oconee Nuclear Site Ms. Susan E. Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.
Columbia, SC 29201
ENCLOSURE I DUKE ENERGY RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 1 NRC RAI I How were the 33 minutes required for the credited manual actions calculated? (were there scenarios run with individual operators to determine this time or is this time calculated according to when actions need to take place within your accident scenario resolution/procedures); what type of task analysis was used if multiple scenarios were run with individuals?
Duke Energy Response to RAI I The time allowed for operator action to isolate the Reverse Osmosis (RO) system to preclude the intake of post-LOCA (Loss of Coolant Accident) fluids is based on the earliest calculated time following a LBLOCA (Large Break Loss of Coolant Accident) to deplete the Borated Water Storage Tank (BWST) to the point where operator action is taken to transfer Emergency Core Cooling System (ECCS) pump suction to the Reactor Building (RB) sump. The minimum time to reach a BWST level setpoint for switchover is calculated as 33 minutes. The calculated time is based on maximum ECCS flows taking suction off the BWST, assumes atmospheric Reactor Coolant System (RCS) and RB pressure, and credits operator action to secure the High Pressure Injection (HPI) pumps according to procedural guidance at a BWST level of 15 feet.
NRC RAI 2 What role do your tech specs (relating to your low level on BWST) play in the manual actions credited.
Duke Energy Response to RAI 2 The BWST level corresponds to the required volume in Technical Specification (TS) 3.5.4. The BWST level in feet is shown in the TS 3.5.4 Bases. This BWST level corresponds to the minimum initial BWST level assumed in the design basis LOCA analyses. Together with the assumed pump flow rates, the TS BWST level is used to determine the minimum time allowed for the manual operator action to isolate the RO system.
NRC RAI 3 Have you done any operator reliability assessments on this manual actions and what would happen if the operator were unable to complete the action?
Duke Energy Response to RAI 3 Duke Energy performed a time study to validate the ability of the operator to reliably isolate the valve in the time required. See responses to RAI 4 & 5 below.
If the operator is unable to complete the action, radioactive fluid may be transported from the RB sump to several locations in the Auxiliary Building (AB) creating an external radiation hazard and an unanalyzed release pathway. This is dependent upon the amount of back leakage through valves HP-1 01, HP-1 02, LP-42, LP-29 and LP-30. The maximum hypothetical accident at ONS assumes there is 5 gpm back leakage into the BWST.
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 2 NRC RAI 4 What type of environmental assessment have you performed to determine the acceptability of the conditions surrounding the operator actions? (for example see below)
- a. Occupational Radiation Exposure
- b. Temperature
- c. Distance
- d. Time necessary to get to remote location (as opposed to inside the control room) where the action is to take place
- e. Physicality required during the accident scenario (any special garments or apparatus that would need to be used during completing the action)
Duke Energy Response to RAI 4.a Occupational radiation exposure will be within limits for the manual action in the spent fuel cooler room, since there will be no radiation source in the RO piping prior to sump recirculation. There is no radioactive sump effluent prior to 33 minutes post-accident.
Duke Energy Response to RAI 4.b The temperature for the spent fuel cooler room, where the isolation valve is located, is calculated to rise less than 50 F, from an initial temperature of 95 0 F, in the first hour after evaluated accidents.
Therefore an operator can access this room post LOCA.
Duke Energy Response to RAI 4.c The RO isolation valve is located in Spent Fuel (SF) Cooler Room, approximately 397 feet from the back of the Control Room 1 .
Duke Energy Response to RAI 4.d Approximately 3 minutes Duke Energy Response to RAI 4.e No special physical attributes or apparatus are required to complete this task. The operator will be equipped with a headlamp or flashlight to use if needed. This task requires an operator to walk approximately 133 feet from the back of the Unit 1 control room to the Unit 2 Stairwell then descend approximately 60 steps down from 5th floor to 2nd floor then walk approximately 204 feet to the SF-181 in the SF Cooler Room. SF-181 is not currently installed, however, it is a 2 inch quarter turn ball valve that will be located approximately 6 feet off the floor1 .
1 Unit 1 and 2 share a common RO isolation valve (SF-181). This valve is slightly further from the Unit 1 & 2 Control Room (CR) than the RO isolation valve for Unit 3 (3SF-181) is from the Unit 3 CR. Therefore, the time study for SF-1 81 was used as the bounding case.
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 3 NRC RAI 5 What is the actual completion time associated with completing the action (given that the available time is 33 minutes)
Duke Energy Response to RAI 5 Approximately 17 minutes from event initiation. Duke Energy performed a time study for isolating the RO System at the seismic boundary valves, SF-181 (Unit 1 & 2) or 3SF-181 (Unit 3). SF-181, which is slightly further from the Unit 1 & 2 Control Room (CR) than 3SF-181 is from the Unit 3 CR, is the bounding case. The timeline for this isolation is as follows:
Note: Time in min:second format T=0 Large Break LOCA occurs T=2:00 Emergency Operating Procedure (EOP) has been entered. Immediate Manual Actions (IMAs) and Symptom Check is complete. Reactor Operator (RO) begins EOP Enclosure 5.1 (ES Actuation). SRO performs notification requirement on Parallel Action Page of EOP which sends plant page for Abnormal Procedure (AP)/EOP Nuclear Equipment Operator (NEO) to report to affected unit and all other NEOs report to Unit 1 & 2 Control Room.
T=10:00 RO dispatches AP/EOP NEO to close SF-1 81. Step to close SF-1 81 will be placed immediately prior to starting Control Room (CR) Outside Air Booster Fans. Data from time validation performed 6/26/11, showed 9.68 minutes to start CR Outside Air Booster Fans in a large LOCA scenario from T=0.
T=10:00 - 12:55 Travel time from back of Unit 1 Control Room to SF-181 walked down by an SRO in normal walking pace on July 16, 2011.
Distance:
-133 feet from U1 CR to U2 Stairwell
-60 steps down from 5th floor to 2nd floor
-204 feet from 2nd floor stairs to SF-181 located in U1-2 SF Cooler Room.
Total distance - 397 feet T=12:55 - 13:25 Conservatively assume a 30 second stroke time. SF-1 81 is not currently installed, however, it is a 2 inch quarter turn ball valve located approximately 6 feet off the floor.
T=13:25 SF-181 closed.
Since this action is not currently in a procedure, and only validated with one individual/team, an additional 25% margin is being applied as a conservative measure.
Time for closing SF-1 81 = 13:25 x 25% = 16.8 minutes
ENCLOSURE 2 REVISED SIGNIFICANT HAZARDS CONSIDERATION
LAR No. 2010-03, Supplement 3 - Revised Significant Hazards Consideration August 2, 2011 Page 1 Significant Hazards Consideration Duke Energy Carolinas, LLC (Duke Energy), has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in Title 10, Code of Federal Regulations, Part 50, Section 92 (10 CFR 50.92), "Issuance of amendment," as discussed below:
- 1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
No. The proposed change requests Nuclear Regulatory Commission (NRC) approval of design features and controls that will be used to ensure that periodic limited operation of a Reverse Osmosis (RO) System during Unit operation does not significantly impact the Borated Water Storage Tank (BWST) or Spent Fuel Pool (SFP) function or other plant equipment. Duke Energy evaluated the effect of potential failures, identified precautionary measures that must be taken before and during RO System operation, and required operator actions to protect affected structures, systems, and components (SSCs) important to safety.
The new high energy piping and non seismic piping being installed for the RO System is non QA-1 and is postulated to fail and cause an Auxiliary Building flood. Duke Energy determined that adequate time is available to isolate the flood source (BWST or SFP) prior to affecting SSCs important to safety.
The existing Auxiliary Building Flood evaluation postulates a single break in the non-seismic piping occurring in a seismic event. The addition of the RO System will not increase the frequency of a seismic event. This event does not consider the amount of non-seismic piping that is currently in the Auxiliary Building. The new piping is not more likely to fail as compared to the existing non-seismic piping. The existing postulated source of the pipe break in the Auxiliary Building is due to the piping not being seismically designed. The new RO System piping is considered a potential source of a single pipe break for the same reason. Since the accident itself is defined as the failure of non-seismic pipe, the new non-seismic piping does not increase the frequency of occurrence of an Auxiliary Building flood. The mitigation of an Auxiliary Building flood due to non seismic piping failure is by manual operator action. The same mitigation technique is used for the high energy line break.
The RO System takes suction from the top of the SFP to protect SFP inventory. Plant procedures will prohibit the use of the RO System during the time period directly after an outage that requires the Unit 1 & 2 SFP level to be maintained higher than the Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.11 level requirement. The higher level is required to support TS LCO 3.10.1 requirements for Standby Shutdown Facility (SSF) Reactor Coolant (RC) Makeup System operability (due to the additional decay heat from the recently offloaded spent fuel). Plant procedures will also specify the siphon be broken during this time period so the SFP water above the RO suction point cannot be siphoned off if the RO piping breaks. The proposed change does not impact the fuel assemblies, the movement of fuel, or the
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 2 movement of fuel shipping casks. The SFP boron concentration, level, and temperature limits will not be outside of required parameters due to restrictions/requirements on the system's operation. In addition, Duke Energy will prohibit fuel movement and cask handling activities during operation of the RO System, so a FHA will not occur while the RO System is in operation. This restriction will remain in place until such time Duke Energy completes calculations that demonstrate the transport of SFP water with radioactivity concentrations resulting from the FHA do not prevent the access required to shutdown the RO system when required.
The BWST is used for mitigation of Steam Generator Tube Rupture (SGTR), Main Steam Line Break (MSLB) and Loss of Coolant Accidents (LOCAs). The SGTR and MSLB are bounded by the SBLOCA analyses with respect to the performance requirements for the HPI System. In the normal mode of Unit operation, the BWST is not an accident initiator. The SFP is assumed to maintain acceptable criticality margin for all abnormal and accident conditions including Fuel Handling Accidents (FHAs) and cask drop accidents. Both the BWST and SFP are specified by TS requirements to have minimum levels/volumes and boron concentrations. The BWST also has TS requirements for temperature. Prior to RO operation, procedures will require that minimum required initial boron concentration, and initial level/volume be adjusted and that the RO System be operated for a specified maximum time period before readjusting volume and boron concentration prior to another RO session to ensure that the TS specified boron concentration and level/volume limits for both the SFP and the BWST are not exceeded during RO System operation. Thus, the design functions of the BWST and the SFP will continue to be met during RO System operation.
An Auxiliary Building flood due to a non-seismic RO System pipe break does not increase the consequences of the flood since the new non-seismic pipe break is bounded by the Auxiliary Building flood caused by existing non-seismic pipe breaks.
Although the RO System will return water with lower boron concentration, procedural controls will ensure that the TS boron concentration level does not go below the limit.
Thus, no adverse effects from decreased boron concentration levels will occur.
The BWST and SFP will still have TS required boron concentration and level/volume, Since the design basis LOCA analysis for Oconee assumes 5 gpm back-leakage from the Reactor Building sump to the BWST, the Emergency Operating Procedure will require the RO System be isolated prior to switchover to the recirculation phase.
Therefore, the mitigation of a LOCA or FHA does not result in an increase in dose consequence.
Therefore, installation and operation of the RO System during Unit operation does not significantly increase the probability or consequences of any accident previously evaluated.
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 3
- 2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
No. The RO System adds non-seismic piping in the Auxiliary Building. However, the break of a single non-seismic pipe in the Auxiliary Building has already been postulated as an event in the licensing basis. The RO System also does not create the possibility of a seismic event concurrent with a LOCA since a seismic event is a natural phenomena event. The RO System does not adversely affect the Reactor Coolant System pressure boundary. The suction to the RO System, when using the system for BWST purification, contains a normally closed manual seismic boundary valve so the seismic design criteria is met for separation of seismic/non-seismic piping boundaries.
Duke Energy also evaluated potential releases of radioactive liquid to the environment due to RO System piping failures. Design features and administrative controls preclude release of radioactive materials outside the Auxiliary Building or by different release pathways. Releases inside the Auxiliary Building are bounded by existing analyses.
The SFP suction line is designed such that the SFP water level will not go below TS required levels, thus the fuel assemblies will have the TS required water level over them. Procedural controls will restrict the use of the RO System and require breaking vacuum on the SFP suction line when the SSF conditions require the SFP level be raised to support SSF RC Makeup System operability. Thus, the SFP water level will not be reduced below required water levels for these conditions. RO System operating restrictions will prevent reducing the SFP boron concentration below TS limits. /
Therefore, operation of the RO System during Unit operation will not create the possibility of a new or different kind of accident from any kind of accident previously evaluated.
- 3) Does the proposed amendment involve a significant reduction in a margin of safety?
No. The RO System adds non-seismic piping in the Auxiliary Building. Duke Energy evaluated the impact of RO System operation on SSCs important to safety and determined that procedural controls will ensure that TS limits for SFP and BWST volume, temperature and boron concentration will continue to be met during RO operation. For the BWST, these controls will ensure the TS minimum BWST boron concentration and level are available to mitigate the consequences of a small break LOCA or a large break LOCA. Procedural controls will also ensure that post LOCA fluids from the reactor building sump (due to back-leakage) cannot be circulated by the RO System. For the SFP, procedural controls will ensure that the assumptions of the fuel handling and cask drop accident analyses are preserved. Additionally, the failure of non seismic RO System piping will not significantly impact SSCs important to safety.
The BWST level may drop below the TS required level due to a rupture of the non seismic piping during a seismic event. However, due to the low probability of a
LAR No. 2010-03, Supplement 3 - Duke Energy Response to NRC RAI August 2, 2011 Page 4 seismic event coupled with the relatively short period of time the RO System will be aligned to the BWST, the possibility of dropping below the TS required level does not involve a significant reduction in the margin of safety. In addition, Oconee's licensing basis does not assume a design basis event occurs simultaneously with a seismic event. The proposed change does not significantly impact the condition or performance of SSCs relied upon for accident mitigation. This change does not alter the existing TS allowable values or analytical limits. The existing operating margin between Unit conditions and actual Unit setpoints is not significantly reduced due to these changes. The assumptions and results in any safety analyses are not impacted.
Therefore, operation of the RO System during Unit operation does not involve a significant reduction in a margin of safety.
Duke Energy has concluded based on the above, that there are no significant hazards considerations involved in this amendment request.