ML101830409

CFR 50.59 Annual Report

ML101830409
Person / Time
Site:	Oconee
Issue date:	06/30/2010
From:	Baxter D Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML101830409 (11)

Text

'Duke DAVE BAXTER Vice President SEnergy Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Highway June 30, 2010 Seneca, SC 29672 864-873-4460 864-873-4208 fax dave. baxter@duke-energy.corn U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington. DC 20555

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2, and 3 Docket Nos. 50-269, 50-270, 50-287 10 CFR 50.59 Annual Report Attached are descriptions of Oconee facility changes, tests, and experiments which were completed subject to the provisions of 10 CFR 50.59 between January 1, 2009 and December 31, 2009. This report is submitted pursuant to the requirement of 10CFR 50.59 (d)(2).

If there are any questions, please contact Corey Gray at (864)-873-6325.

Sincerely, Dave Baxter Vice President Oconee Nuclear Station Attachment www. duke-energy corn

U. S. Nuclear Regulatory Commission June 30, 2010 Page 2 xc: Mr. Luis Reyes Regional Administrator, Region I1 U. S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Ga 30303-1257 Mr. Andrew Sabisch Senior NRC Resident Inspector Oconee Nuclear Site Mr. John Stang Senior Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555

Attachment U. S. Nuclear Regulatory Commission Page 1 of 9 Type: UFSAR Change 09-84 (AR#00285558)

Title:

Update to UFSAR Table 15-16. This activity evaluated herein is the addition of a direct radiation or shine constituent to the post Maximum Hypothetical Accident (MHA) control room radiation dose.

==

Description:==

The Alternative Source Terms (AST) analysis of the MHA at Oconee Nuclear Station has been revised. One of the changes in the revision was the addition of a constituent for direct radiation from external sources into the control room radiation dose for this accident. This change was screened as a change to the method for the AST analysis of the MHA at Oconee. Adding this constituent into the post MHA control room radiation dose clearly is conservative compared to not adding this constituent into the post MHA control room radiation dose. For this reason, this change to the AST analysis of the MHA is not a departure from a method of evaluation described in the UFSAR and used in the safety analysis and so does not require NRC approval before use.

Type: Nuclear Station Modification (AR#00220509)

(Omitted 2008 50.59 evaluation see PIP 0-10-5233 for details)

Title:

01C25 Reload Core Design

==

Description:==

This activity installs the core designed for Oconee Nuclear Station Unit 1 Cycle 25. The 01C25 Reload Design Safety Analysis Review (REDSAR),

performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the 01C25 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted 01C25 safety analysis physics parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3005-PA.

The 01C25 core reload is similar to past cycle core designs, with a design generated using NRC approved methods. The 01C25 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5.

Additionally, applicable Technical Specifications and the UFSAR have been reviewed and no changes are required for the operation of O1C25. This 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for 01C25 operation.

Attachment U. S. Nuclear Regulatory Commission Page 2 of 9 Type: Nuclear Station Modification (AR#00285548)

Title:

01C26 Reload Core Design

==

Description:==

This activity installs the core designed for Oconee Nuclear Station Unit 1 Cycle 26. The 01C26 Reload Design Safety Analysis Review (REDSAR),

performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the 01C26 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted 01C26 safety analysis physics parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in topical report DPC-NE-3005-PA.

The 01 C26 core reload is similar to past cycle core designs, with a design generated using NRC approved methods. The 01 C26 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5.

Additionally, applicable Technical Specifications and the UFSAR have been reviewed and no changes are required for the operation of O1C26. This 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for OLC26 operation.

Type: UFSAR Change 09-25 (AR#00264292)

Title:

fUpdate to UFSAR 15.1 - Replacement of ORIGEN-2 SCALE

==

Description:==

An update to Section 15.1 of the Oconee Nuclear Station UFSAR has been prepared. This update reports that isotopic activities used in the analyses of radiological consequences of design basis accidents are calculated with the modular computer code suite SCALE in place of the computer code ORIGEN-2.

Also with this update, the term "End-of-cycle" is stricken from UFSAR 15.1.10 to indicate that isotopic activities are calculated not just at end of cycle but over a range of time in cycle. This latter part of the update reflects current practice and so is only a "virtual change."

Both parts of this update were screened as changes to a method of evaluation described in the UFSAR used in the safety analysis. In at least six safety evaluations, the NRC has endorsed the use of SCALE for the calculation of isotopic activities. Striking the term "End-of-cycle" as noted above reflects expansion in the set of values taken for an input parameter and so is inherently conservative. Neither change is a departure in a method of evaluation described in the UFSAR used in establishing the design basis or in the safety analysis.

UFSAR 15.1 may be updated as noted above without prior NRC approval.

Attachment U. S. Nuclear Regulatory Commission Page 3 of 9 Type: UFSAR Change 09-25 (AR#00263355)

Type: Updates to UFSAR 15.1, 15.11, 15.15, & associated Tables. The UFSAR updates reported revised radiation doses for the Fuel Handling Accidents (FHA) and Maximum Hypothetical Accident (MHA). The activity under evaluation consists of changes to certain input parameters used to calculate isotopic activities for a fuel assembly (FHAs) and reactor core (MHA).

==

Description:==

A set of updates to UFSAR 15.1, 15.11, and 15.15 and Tables 15-1, 15-15, and 15-16. These updates report revised Alternative Source Terms (AST) analyses of the fuel handling accidents (FHAs) involving one fuel assembly and several fuel assemblies (a fuel cask drop) and the loss of coolant accident (LOCA). The updates also report revised isotopic activities for a fuel assembly and the reactor core which were used in the revised AST analyses. The changes prompting the updates and evaluated here are changes to values taken for certain input parameters as follows:

1) Expansion of the range for fuel assembly averaged fuel enrichment.

2) Expansion of the range for reactor core averaged fuel enrichment.

3) Increase in the upper limit for feed batch size.

4) Decrease in the batch averaged burnup limits.

These changes are made only to accommodate current practices in the design of the nuclear fuel used at Oconee. No changes are made to the design of Oconee nuclear fuel or the representation of that design in the Oconee UFSAR. No modifications are made to any Oconee system, structure or component (SSC) in association with the input changes. No new SSC is added in association with the input changes. The radiation doses for the FHAs and LOCA do not correspond to more than a minimalincrease in consequences of an accident or malfunction of an SSC important to safety currently evaluated in the UFSAR. The input changes and revised calculations may be retained and the associated updates inserted in the Oconee UFSAR without prior approval from the NRC.

Type: UFSAR Change 09-25 (AR#00242281).

Title:

Updates to ONS UFSAR 15.11, 15.15, & Associated Tables. These portions of ONS UFSAR 15 are updated to report revised values for post accident radiation doses. The post accident control room radiation doses were revised account for an increase in the rate of unfiltered inleakage to the Unit 1 and 2 control rooms with the Control Room Ventilation System booster fans off. This activity is evaluated herein.

==

Description:==

A set of updates to Oconee UFSAR Sections 15.1, 15.11, 15.15, and certain associated tables have been prepared. The updates report revised Alternative Source Terms (AST) analyses of the loss of coolant accident (LOCA), fuel handling accident (FHA) involving one fuel assembly, and a FHA involving

Attachment U. S. Nuclear Regulatory Commission Page 4 of 9 multiple fuel assemblies. This effort was made in part to account for an increase in the measured rate of unfiltered inleakage into the control room (of Units 1 and

2) with the Control Room Area Ventilation System (CRAVS) booster fans off.

This specific change is evaluated here.

It is the increased rate of unfiltered inleakage to the control room (of Units 1 and

2) with the CRAVS booster fans off that was evaluated. All other measured values of control room unfiltered inleakage showed decreases from the corresponding values measured in the first test and therefore below the corresponding values assumed in current license basis analyses.

The evaluation covered not only the impact on post accident radiation doses in the control room. It also accounted for the potential impact on a toxic gas release and smoke intrusion and on the potential for overheating the control room and the controls inside it. The post accident radiation doses were found to not constitute more than a minimal increase in consequences of an accident or malfunction of a structure, system, or component (SSC) important to safety.

Loss of the control room following a poisonous gas release or intrusion of smoke is not made more likely with this change. The inleakage remains less than the rate of inflow of hot air assumed in the design of the CRAVS air conditioning equipment at Oconee; the increase does not create the possibility of malfunction of instrumentation and controls in the control room due otherwise to overheating.

No changes in methodology were associated with any supporting analysis.

The rate of unfiltered inleakage to the Oconee control rooms may remain at their measured values without prior approval of the NRC.

Type: Calculation (AR#00282528)

Title:

DPC-NE-3005,Rev. 3b Methodology Report Revision

==

Description:==

DPC-NE-3005 is included in the UFSAR transient analyses as Reference 1 in Section 15.1.14. The following changes to DPC-NE-3005-PA, Rev. 2 were evaluated to determine if they represent a departure from a method of evaluation as described in the UFSAR used in establishing the safety analyses.

1) Change how the main feedwater system and turbine control are modeled in the rod withdrawal accident.

2) Change the time-in-life assumption for the physics parameters in the dropped control rod accident.

3) Add dropped rod worth to the list of physics parameters that must be checked each reload for the dropped control rod accident.

The changes grouped in 1) change the Reference 1 methodology from requiring a sensitivity study be performed on whether automatic or manual control is conservative in the rod withdrawal at power accident (UFSAR 15.3) to just stating that manual control is conservative as verified in the analysis of record (AOR).

The changes grouped in 2) change the requirement of performing the dropped

Attachment U. S. Nuclear Regulatory Commission Page 5 of 9 rod accident (UFSAR 15.7) using both beginning of cycle (BOC) and end of cycle (EOC) physics parameters to just using BOG physics parameters. Using BOC physics parameters is conservative as verified in the AOR. The changes grouped in 3) add dropped rod worth to the list of physics parameters that must be checked each reload for the dropped rod accident (UFSAR 15.7). This check has always been performed as part of the reload process. The associated text change to recognize this check will have no impact on the results of the dropped rod accident. Since this evaluation is the result of an affirmative answer to screen question #3 (Evaluation Methodology), only evaluation question #8 is addressed in this evaluation per Section 4.2.1.3 of NEI 96-07, Revision 1.

After making the above changes, the results are either more conservative than the results generated using the current methodology or the same as the results using the current methodology. Therefore, there is no departure from a method of evaluation as described in the UFSAR used in establishing the safety analyses and the changes can be made to DPC-NE-3005-PA, Revision 2.

Type: Nuclear Station Modification (AR#00268521)

Title:

03C25 Reload Core Design

==

Description:==

This activity installs the core designed for Oconee Nuclear Station Unit 3 Cycle 25, which is the first Oconee Unit 3 core design to employ Mark-B-HTP fuel, and the second core design at Oconee (the first being 02C24). Mark-B-HTP fuel and the methods necessary to evaluate have been granted NRC approval as published in the DPC-NE-2015-PA methodology report. The 03C25 Reload Design Safety Analysis Review (REDSAR), performed in accordance with Engineering Directives Manual EDM-501, "Engineering Change Program for Nuclear Fuel", and the 03C25 Reload Safety Evaluation confirm the UFSAR accident analyses remain bounding with respect to predicted 03C25 safety analysis physics parameters (SAPP), and fuel thermal and mechanical performance limits. The SAPP method is described in the NRC approved DPC-NE-3005-PA methodology report.

Except for the Mark-B-HTP fuel design change, the 03C25 core reload is similar to past cycle core designs, with a design generated using NRC approved methods. The O3C25 Core Operating Limits Report (COLR) was prepared in accordance with Technical Specification 5.6.5. Additionally, applicable Technical Specifications have been reviewed and no additional changes are required for the operation of 03C25. The 10CFR50.59 evaluation concluded that no prior NRC approval is necessary for 03C25 operation.

Attachment U. S. Nuclear Regulatory Commission Page 6 of 9 Type: Nuclear Station Modification NSM ON-33098 (AR#00238541)

Title:

NSM ON-33098 Phase 2- Upper Surge Tank (UST) Inventory Protection (10 CFR 50.59 Revision 0)

==

Description:==

The implementation of NSM on-33098 was broken into two phases. Phase 1 included the structural steel work to the Upper Surge Tank (UST) platform and the UST Dome Tank and a 10 CFR 50.59 screening was performed for that phase. The remainder of the NSM is to be implemented under Phase 2. This 10 CFR 50.59 summary only addresses Phase 2.

Phase 2 of NSM ON-33098 will make some modifications to eliminate single active failures associated with the Upper Surge Tank (UST). These modifications involve both "active type" and "passive type" isolation. "Active type" isolation requires a signal to be sent to a valve to close to assure flowpath isolation. "Passive type" isolation uses a check valve to prevent reverse flow.

This modification is to add four air-operated valves (AOV) that automatically close when the UST level drops below 7.5 feet. New valves 3C-903 and 904 are to isolate flow to the hotwell and to the Auxiliary Boiler Feedwater (FDW) pump.

A bypass valve, 3C-912, is provided around 3C-903 and 904. Valves 3C-906 and 907 isolate flow to the Powdex Backwash Pump. A pressure switch will be added in the suction line of the Polishing Deminerizer Backwash pump that will trip it on low suction pressure.

The modification is also to add a new Condensate Recirculation path to the UST.

This path will allow flow from the Condensate Booster Pump suction line to the UST Riser. A manual throttle valve, 3C-899, and flow indication (locally and on the Operator Aid Computer) are provided. This new Condensate Recirculation path should provide significant operational flexibility during unit start-ups.

Other changes by this modification include:

• upgrade of the hotwell level control system, including replacement of valve 3C-192, and level transmitters LT-17 and LT-19

• removal of electric motor operated valves 3C-152 and 153

• upgrade of the UST level transmitters to reduce instrument uncertainty 0 upgrade of the Emergency Feedwater (EFW) Pump recirculation path to the UST Dome Tank to Class F, QA Condition 1; including seismic qualification of the Dome Tank.

Expansion joints are to be installed in Class G piping connections to the UST Dome tank.

Attachment U. S. Nuclear Regulatory Commission Page 7 of 9 There are currently three somewhat independent paths of water from the UST to the condenser. The modification will combine the supplies to these paths with a common header. With the existing design for normal operating conditions, one of the AOVs can fail closed (e.g., loss of air to valve operator) and this failure will not necessarily prevent another of the AOVs from opening (or remaining open) to supply water from the UST to the condenser hotwell ifthere is a low hotwell level.

In the new design, a failure of one of the new AOVs could prevent water from transferring to the hotwell from the UST through this pathway. The safety function of the valves is to close on low UST level. During certain accidents/events, the valves in both the current and proposed new design will receive a close signal on low UST level. In the new design, the failing closed of one of the new AOVs (3C-903 or 3C-904) can stop the flow of water from the UST to the condenser. The UFSAR has described the three separate pathways from the UST to the condenser hotwell, but the context of this wording is in relation to the pathways being automatically isolated on a low UST level. In the unlikely event that one of the new AOVs does inadvertently close, manual actions could be taken to open the bypass around the new valves. If this action was not taken and the level dropped significantly, a trip of the unit could occur.

The loss of air to the new AOVs could occur due to a loss of the non-safety related Instrument Air System or due to a loss of air locally at the valve (e.g., loss of the safety related power supply to the solenoid valve on one of the new valves.

A loss of the Instrument Air System would result in the same effect in both the existing and new design. The effect would be that the flowpath(s) from the USTs to the hotwell would be isolated due to fail closed AOVs. If the air supply is lost to one of the new valves (3C-903 or 3C-904), then all three flowpaths would be isolated. But, there are manual actions that could be taken to bypass the failed closed AOV. If the hotwell level is not replenished over time, a trip of the unit could occur. Thus, there is a possible increase in the potential for a turbine/reactor trip if makeup to the hotwell is not able to be achieved. But there are other means for a reactor/turbine trip to occur. UFSAR Section 15.8 provides a number of means for a turbine trip. The potential cause for a turbine trip is described as including a generator trip, low condenser vacuum, loss of turbine lubrication oil, turbine thrust bearing failure, turbine overspeed, main feedwater pump trip, high steam generator level, or a reactor trip. The loss of hotwell level, if low enough, could cause the main feedwater pumps to lose suction pressure and ultimately trip. The trip of the main feedwater pumps are listed in the UFSAR section described above. This small potential for a localized loss of air to one of the new valves is considered to be a negligible increase to the overall turbine trip potential and thus is not a "more than a minimal increase" in the frequency of occurrence of an accident previously evaluated in the UFSAR.

The two AOVs (3C-903 and 3C-904) in the common header from the USTs to the hotwell will isolate the UST if a low UST level is detected or if air is lost to the

Attachment U. S. Nuclear Regulatory Commission Page 8 of 9 valves. These two AOVs are to be used as the QA-1 Class F boundary so that the UST tank contents will be isolated even in the event of a single failure. The potential for "more than a minimal increase" in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR was investigated with respect to the UST's makeup going into a common header before going to the three separate pathways. Equipment important to safety affected by this modification includes the UST (assured source of EFW),

condenser hotwell (one of the potential long term sources of EFW), and EFW System (provides feedwater in the event of a loss of main feedwater). There are currently three somewhat independent paths of water from the UST to the condenser. The modification will combine the supplies to these paths with a common header. Although flow from the UST to the condenser hotwell is not required to be designed to withstand a single failure, the potential for "more than a minimal increase" in likelihood of occurrence of a malfunction of an SSC important to safety needs to be considered. With the existing design for normal operating conditions, one of the AOVs can fail closed (e.g., loss of air to valve operator) and this failure will not necessarily prevent another of the AOVs from opening (or remaining open) to supply water from the UST to the condenser hotwell if there is a low hotwell level. In the new design, a failure of one of the new AOVs could prevent water from transferring to the hotwell from the UST through this pathway. The safety function of the valves is to close on low UST level. During certain accidents/events, the valves in both the current and proposed new design will receive a close signal on low UST level. In the new design, the failing closed of one of the new AOVs (3C-903, 3C-904) can stop the flow of water from the UST to the condenser. The UFSAR describes the three separate pathways from the UST to the condenser hotwell, but the context of this wording is in relation to the pathways being automatically isolated on a low UST level. The "important to safety" aspect of the condenser hotwell is its function of supplying an EFW supply of water after the UST source has been exhausted.

The flowpath from the condenser to the EFW pumps is not adversely affected with the new valves since that flowpath is not used when supplying the EFW pumps via the hotwell. The "important to safety function" of the existing AOVs and the new AOVs in the new design is considered to be their closure on low UST level. This function is enhanced in the new design. Supplying the hotwell from the UST is considered more of an operational issue versus an "equipment important to safety" issue. Thus, the use of a common header with two AOVs in series is not considered to cause a "more than minimal" increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

The EFW System is used to mitigate accidents involving the loss of main feedwater. The modification will not change the design function of the EFW supply sources as evaluated in the UFSAR. Thus, in an accident involving loss

Attachment U. S. Nuclear Regulatory Commission Page 9 of 9 of main feedwater, the EFW System will still be able to mitigate the event as currently described in the UFSAR. There is no adverse effect on containment integrity and no new release paths are created. The design is such that all valves, piping, components, and circuitry which are required to assure the UST is not prematurely depleted are QA-1 and seismically qualified.

The UST will be designed to provide a source of water to the EFW System even in the event of a single failure. The hotwell backup source is not designed to provide the additional EFW water supply in the event of a single failure. The flowpath from the UST to the hotwell is not required to be designed to withstand a single failure for the function of allowing water to flow. This path is designed such that a single failure does not allow UST flow to be depleted to the Hotwell.

Thus, the EFW function is not adversely affected with respect to mitigating loss of feedwater scenarios previously evaluated in the UFSAR.

If air is lost to either new valve 3C-906 or 3C-907, UST supply to the polishing demineralizer backwash pumps could be lost. The polishing demineralizers are used for normal plant operation and they do not serve a safety function nor are they designed to the single failure criterion.

If air is lost to new valves30-903 or 3X-904, the Unit 3 supply to the Aux Boiler could be lost. The Aux Boiler is not required to be designed so that it is available following a single failure.

The 10 CFR 50.59 criteria for requiring prior NRC approval were not met. No technical specification or SLC changes are required. UFSAR Section 3.2.2 is to be updated to reflect Phase 2 of this NSM. In addition, UFSAR Sections 10.1, 10.4.1.5, 10.4.7.2.3, and 10.4.7.3.2.3, and UFSAR Figure 10-6 were inadvertently revised in the December 31, 2007 update. These sections and figure were revised prior to this NSM's implementation and thus do not need to be revised now. The inadvertent updating was documented in the Duke corrective action program.