ML13168A579

Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports - January 1, 2011 Through December 31, 2012

ML13168A579
Person / Time
Site:	Oyster Creek
Issue date:	05/24/2013
From:	Stathes G Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13168A579 (11)

Text

RA-13-052 May 24,2013 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station

• ReneweaFacilifV opemtin?fUCenseNo: DPR:1e NBC Pocket No. 50-219 10 CFR 50.59 10CFR 72.48

Subject:

Bienniat 10 CFB 50.59 and 10 CFB 72.48 Change Summary Reports -

January 1, 2011 through December 31, 2012 Enclosed are the Oyster Creek Nuclear Generating Station 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports for regulatory commitments changed during the period of January 1, 2011 through December 31,2012.

There are no regulatory commitments contained in this submittal.

Please contact Jeff ChrisJey at (609) 971-4469 if any further Information or assistance is n~ed.

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Garey l. Stathes ~r Vice President Oyster Creek Nuclear Generating Station Enclosure cc:

Administrator, USNAC Region I USNRC Senior Project Manager, Oyster Creek USNRC Senior Resident Inspector, Oyster Creek

Enclosure - RA-11-005 Page 1 of 10 Exelon Generation Company, LLC Oyster Creek Nuclear Generating Station Enclosure to RA-13-052 Docket No. 50-219 xx-xx 2011 - 2012 Biennial 10 CFR 50.59 and 10 CFR 72.48 Change Summary Reports These summary reports are issued pursuant to reporting requirements for Oyster Creek Nuclear Generating Station (OCNGS). These reports address tests, experiments, and changes to the facility and procedures as they are described in the Final Safety Analysis Report for the OCNGS station and the Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System (TN-61BT Spent Fuel Cask at OCNGS).

These reports summarize the four tests, experiments, and changes that were implemented between January 1, 2011 and December 31, 2012 under 10 CFR 50.59.

There were no tests, experiments, or changes implemented by OCNGS under 10 CFR 72.48.

Enclosure - RA-11-005 Page 2 of 10 Evaluation Number:

OC-2011-E-001 Rev 0 PORC Review Meeting No. (Date):

2011-02 (12/05/2011)

Activity/Document No.:

Modification / ECR OC 11-00379

Title:

Multiple Spurious Operations (MSO) Bus ID Appendix R (APPR) Permissive Switch Installation Description of Activity:

ECR OC 11-00379 installs a second permissive switch, hereafter referred to as the Appendix R (APPR) permissive switch, in series with the existing 69 permissive switch in the closing circuit of 4160V SWGR ID Unit Dl, Main Breaker ID. The APPR permissive switch is located on the 4160V SWGR ID Unit DI door.

Reason for Activity:

The design change addresses a concern identified in MSO Scenario 5g, which is addressed in IR 01059439. Specifically, two scenarios were identified that could potentially occur for the 1D main breaker to re-close once Emergency Diesel Generator EDG-2 is providing power to 4160V SWGR 1D. In the first scenario, Breaker 1D would re-close with the Main Generator still connected and coasting down (varying frequency).

This places a large load on the diesel for the dynamic breaking load of the Main Generator coast down through the Auxiliary Transformer, Bus 1B, Bus 1D, and EDG-2.

Protective relaying then causes EDG-2 to trip. In the second scenario, Breaker 1D could pick non-safety related loads powered by 4160V SWGR 1B, which in turn could overload EDG-2. For certain fires, the plant relies on EDG-2 to accomplish safe shutdown.

Effect of Activity:

Implementing the design change prevents the inadvertent reclosing of the 1D Breaker should it trip during an Appendix R / MSO fire or for any other reason. The APPR permissive switch also prevents the resynchronization of 4160V Buses 1B and ID. In order to close Breaker ID or to resynchronize 4160V Bus 1B to 4160V Bus 1D, the APPR permissive switch must be closed. The APPR permissive switch is maintained in the 'open' position during normal plant operation. Operation of the APPR permissive switch is performed locally at Breaker ID.

By preventing Breaker 1D from closing spuriously during certain fire scenarios, prevents the possibility of overloading EDG-2 and ensuring that EDG-2 is available for safe shutdown of Oyster Creek.

This change does not address a fire in the 4160V D SWGR Room. Operation of the ID breaker, which is located in the 4160V D SWGR Room, is not required to address a fire in 4160V D SWGR Room. Per the Oyster Creek post fire safe shutdown program, other systems powered from offsite power are available to achieve hot or cold shutdown.

Offsite power feeds the 4160V Buses 1A and lB, which are in a different fire area than 4160 Bus 1 D. If offsite power is not available, then EDG-1 is available to provide power to 4160V Bus IC. 4160V Bus 1C also is in a different fire area than 4160V Bus 1D.

Enclosure - RA-11-005 Page 3 of 10 Implementing the design change per ECR OC 11-00379 does not have any impact on any plant design basis or safety analysis as described in the UFSAR.

Summary of Conclusion for the Activitys 50.59 Review:

Implementing the design change per ECR OC 11-00379 does involve a change to a procedure that adversely affects how a UFSAR described SSC design function is operated or controlled. Therefore, implementing ECR OC 11-00379 does require a 50.59 Evaluation.

The 50.59 Review has determined that implementing the design change does not result in more than a minimal increase in the frequency of an occurrence of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Implementing the design change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Implementing the design change does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR, nor does it create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.

Implementing the design change does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

Implementing the design change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses for Oyster Creek.

Therefore, NRC approval is not required prior to implementing the design change per ECR OC 11-00379.

Enclosure - RA-11-005 Page 4 of 10 Evaluation Number:

OC-2011-E-002 Rev 2 PORC Review Meeting No. (Date):

2012-01 (11/15/2012)

Activity/Document No.:

Modification /ECR OC 11-00368

Title:

Core Spray System Multiple Spurious Operations (MSO) Modifications MSO Scenarios 5k & 10b Description of Activity:

The design change per ECR OC 11-00368 replaces the Core Spray System 1 & 2 suction valve control switches. It also modifies the control logic, including the respective control switches, to provide a shorting function across the closing coil for the Core Spray System 2 parallel isolation valves and all four Core Spray pump suction valves.

The design change also electrically rearranges the torque and limit switch protection of both Core Spray System 2 parallel valves and all four Core Spray pump suction valves.

This is done to prevent a hot short in the open circuit from bypassing the torque and limit switches, resulting in potential damage to the valve internals that could prevent local manual operation of the valves.

It should be noted that the Core Spray System 1 parallel valves are not being modified.

But to address human factor concerns, the operation of the parallel valves for both Core Spray Systems is to be the same. Procedures regarding the operation of the Core Spray System parallel valves are revised such that the operator has to hold the switch in a maintained position to move the valve to the fully closed position, whereas before the operator did not have to hold the switch in a maintained position.

Reason for Activity:

The design change per ECR OC 11-00368 addresses a concern identified in MSO Scenario 5k and 10b, which is addressed in IR 01059021. Specifically, it identifies fire-induced spurious operations scenarios regarding the potential for spurious operation of several motor operated valves (MOVs), including the Core Spray System 2 parallel isolation valves and the Core Spray Pump suction valves.

In addition, NRC IN 92-18 addressed those motor-operator valves that could open or close spuriously and sustain damage due to torque and limit switch alignment in the valve circuit. With regard to the Core Spray suction valves, which normally are open, the shorting function prevents spurious closing of the valves during an Appendix R / MSO fire. With regard to the parallel isolation valves, the shorting function prevents a hot short from driving the valves, which normally are closed, from being driven further in the closed direction during an Appendix R / MSO fire. Driving the valves further in the closed direction could damage the valve internals and prevent local manual operation of the valves. Installing the shorting function on the Core Spray System 2 parallel valves also requires removing the seal-in circuit for the Core Spray System 2 parallel valves for the closing circuit only.

Enclosure - RA-11-005 Page 5 of 10 Effect of Activity:

Implementing the design change per ECR OC 11-00368 ensures the Core Spray System 2 parallel isolation valves and the Core Spray System 1 and 2 pump suction valves are not damaged during MSO Scenarios 5k and 10b. It also brings Oyster Creek into compliance with NRC IN 92-18 with regard to these same valves.

Implementing the design change also changes how Operators close the Core Spray System 1 and 2 parallel valves in that the Operator now has to hold the respective control switches in a close position in order to close the valves.

Implementing the design change per ECR OC 11-00368 does not have any impact on any plant design basis or safety analysis as described in the UFSAR.

Summary of Conclusion for the Activitys 50.59 Review:

Implementing the design change per ECR OC 11-00368 does involve a change to a procedure that adversely affects how a UFSAR described SSC design function is operated or controlled. Therefore, implementing the design change does require a 50.59 Evaluation. Revision 2 of the 50.59 evaluation supersedes previous revisions.

Implementing the design change does not result in more than a minimal increase in the frequency of an occurrence of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Implementing the design change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Implementing the design change does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR, nor does it create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.

Implementing the design change does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

Implementing the design change does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses for Oyster Creek.

Therefore, NRC approval is not required prior to implementing the design change per ECR OC 11-00368.

Enclosure - RA-11-005 Page 6 of 10 Evaluation Number:

OC-2012-E-001 Rev 0 PORC Review Meeting No. (Date):

2012-14 (10/17/2012)

Activity/Document No.:

Use of TRACGO4P Version 4.2.69.0 / ECR OC 11-00491

Title:

Application of TRACGO4P Version 4.2.69.0 for APRM Stability Protection Settings Determination at OCNGS Description of Activity:

This activity addresses the use of the General Electric - Hitachi (GEH) advanced, multi-purpose NSSS thermal-hydraulic transient code TRACGO4P, Version 4.2.69.0, for the purpose of determining the BWROG Long-Term Stability Solution 'Option II' APRM Stability Protection settings for Oyster Creek Nuclear Generating Station (OCNGS).

APRM Stability Protection settings are evaluated each operating cycle as part of the standard reload licensing process performed in accordance with General Electric's Standard Application for Reactor Fuel (GESTAR II) methodology (Reference 6). The APRM Stability Protection settings are identified in the Core Operating Limits Report (COLR). Version 4.2.69.0 of TRACGO4P is an upgraded version of the NRC approved TRACGO2A program originally developed and licensed to determine APRM Stability Protection settings.

Version 4.2.69.0 of TRACGO4P has not been generically approved by the NRC for APRM Stability Protection settings. The APRM Stability Protection methodology is described in the UFSAR and cited in Technical Specification 6.9.1 by reference.

Therefore, use of TRACGO4P Version 4.2.69.0 constitutes a change in methodology requiring evaluation in accordance with 10CFR 50.59.

The TRACGO2A version of the TRACG thermal-hydraulic code was approved by the NRC and used in the preparation of NEDO-32465-A (Reference 5) during the original design and licensing of the BWROG Long-Term Stability Solutions, including Option 11.

In 2006 the TRACG code was upgraded to TRACGO4 to support coupling with an improved kinetics model resulting from GE's transition to the PANAC11 version of the 3-cliniensional core simulator program PANACEA (References 4). This earlier software upgrade was evaluated under 10CFR 50.59, as documented in 50.59 Evaluation 0C-2006-E-0003.

In 2009 GE implemented a PC-based version of the TR9 CG04 program, TRACGO4P, Version 4.2.57.11 (Reference 3).

In 2010 GE implemented further changes to the PC-based version of the TRACGO4 program, TRACGO4P, Version 4.2.60.3 (Reference 2). These earlier software upgrades were evaluated under 10CFR 50.59, as documented in 50.59 Evaluation 0C-2010-E-0001.

Enclosure - RA-11-005 Page 7 of 10 This 50.59 Evaluation has been prepared to support; upgrading TRACGO4P to Version 4.2.69.0. Version 4.2.69.0 implements fixes to several programming deficiencies, and includes all previous changes implemented in TRACG. This 50.59 evaluation necessarily addresses all software changes implemented subsequent to the version of the program reviewed and approved by the NRC, TRACGO2A.

Reason for Activity:

The TRACGO4P code has recently been revised by the vendor, GE-Hitachi (GEH), to address a number of programming issues identified since its initial release.

The use of TRACGO4P version 4.2.69.0 for Oyster Creek Nuclear Generating Station DIVOM analysis when determining APRM Stability Protection settings constitutes a change in methodology. The upgraded version of the code was developed under the GEH NRC-approved Quality Assurance Program. However, since TRACGO4P Version 4.2.69.0 has not been reviewed and approved by NRC for DIVOM analysis, and GEH is not a license holder, the change needs to be evaluated under 10CFR 50.59 by Exelon.

Effect of Activity:

Oyster Creek has implemented BWROG Stability Long-Term Solution Option II as stated in section 4.3.2.7.2 of the UFSAR. The Option II solution provides fuel cladding integrity safety limit protection by automatically detecting and suppressing reactor instability. The Option 11 solution implements modified flow-biased APRM Scram settings. The modified APRM Scram settings (APRM Stability Protection settings) are established to initiate a reactor scram before power oscillations caused by thermal-hydraulic instability challenge the integrity of the fuel.

The requirement for APRM Stability Protection settings is specified in OCNGS Technical Specification 2.3.A.1 and 2.3.B. The actual APRM Stability Protection settings are documented in the COLR.

The TRACG thermal-hydraulic code is used to develop a conservative relationship between the change in fuel bundle critical power ratio (CPR) and the hot bundle oscillation magnitude. This conservative relationship is used to determine the Delta CPR Over Initial MCPR Verses Oscillation Magnitude (DIVOM) curve. The DIVOM curve, in conjunction with the initial minimum critical power ratio (IMCPR) and the hot bundle oscillation magnitude, is used by GEH to determine the APRM Stability Protection settings.

The slope of the DIVOM curve represents the thermal-hydraulic responsiveness of the fuel to a given oscillation magnitude. Thus, a steeper slope is more conservative than a flatter slope (NEDO-32465-A). Benchmarking of the NRC-approved TRACGO2A code and the TRACGO4P Version 4.2.69.0 code has determined that the DIVOM slope developed using TRACGO4P generates a slightly more conservative (steeper) DIVOM slope. Therefore TRACGO4P Version 4.2.69.0 can be applied for OCNGS DIVOM analysis when determining APRM Stability Protection settings without prior NRC approval.

Enclosure - RA-11-005 Page 8 of 10 Summary of Conclusion for the Activitys 50.59 Review:

The use of TRACGO4P version 4.2.69.0 for OCNGS DIVOM analysis when determining APRM Stability Protection settings constitutes a change in methodology that is addressed by this 50.59 Review.

As discussed in the 50.59 Evaluation, GEH benchmarking analyses confirm that TRACGO4P produces DIVOM curves that are slightly conservative (more limiting) than those produced by TRACGO2A. Therefore, the use of TRACGO4P Version 4.2.69.0 does not constitute a departure from a method of evaluation described in the UFSAR and TRACGO4P can be used to support the determination of APRM Stability Protection settings without prior NRC approval.

The version of TRACG is below the level of detail discussed in the UFSAR and Technical Specifications, therefore a change to the UFSAR and Technical Specification BASES is not necessary. This 50.59 Review is being processed as part of the OCNGS Reload Fuel Design Change Package per ECR OC 11-00491.

Evaluation Number:

OC-2012-E-002 Rev 1 PORC Review Meeting No. (Date):

2012-17 (10/26/2012)

Activity/Document No.:

Procedure Revision / 305

Title:

Shutdown Cooling System Operation Description of Activity:

This activity revises Procedure 305, Shutdown Cooling (SDC) System Operation, to include a revised procedurally controlled Temporary Configuration Change (TCC), which extends the installation during system operation of a jumper that disables the low suction pressure and the high suction temperature trips from one pump to all three SDC pumps.

Reason for Activity:

SDC pump low suction pressure trip switches have a range of 0-12 psig and are always over-ranged, and frequently damaged during the system startup (since SDC system is started at - 120 psig) or during power operation due to valve seat leakage. That condition resulted in numerous spurious pump trips. Similarly, pump trips can be experienced due to spurious operation of the high suction temperature trip switches that could result in inadvertent loss of a SDC pump.

The procedurally controlled TCC designed to jumper/disable the low suction pressure and the high suction temperature trips for all three SDC pumps will prevent pump trips due to spurious operation or failure of these switches.

Enclosure - RA-11-005 Page 9 of 10 Effect of Activity:

The enhanced system operating criteria (i.e., flow rate limitations for various plant conditions) included in the Procedure 305 revision 110 were designed to prevent SDC pump cavitation due to low suction pressure, as well as prevent reactor thermal stratification during SDC operation, while having no impact on system design basis or safety analyses described in the UFSAR. The calculation results were benchmarked against actual data collected since the SDC pump startup in 1R24 and determined to be conservative by a significant margin.

The TCC will disable the low suction pressure and the high suction temperature trips on all three SDC pumps in order to prevent pump trips due to spurious operation or failure of these switches. These interlocks are described in the UFSAR Section 5.4.7.2 as follows: 'The pumps are provided with interlocks which prevent operation unless the suction pressure exceeds 4 psig, the suction temperature is below 350°F, and the suction line isolation valve (V-17-19) is open. This assures that suction pressure is greater than the minimum NPSH and protects the system against inadvertent operation at excessive temperature". SDC Pump NPSH calculation C-1302-214-E310-047 demonstrated that during pump operation within the procedurally mandated criteria, the pump suction pressure remains above the low suction pressure trip setpoint for all considered reactor conditions (i.e., pressure, temperature, and level) and prescribed pump flow rates.

Disabling of the pump suction high temperature (i.e., 350 °F) trip switches will minimize the risk of spurious pump trips due to switch failure, while the pumps remain protected against high temperature by the interlock with the system inlet isolation valve (V-17-19) logic. Temperatures above 350°F in any recirculation loop will prevent V-17-19 from opening or will close it if it is open. When this valve closes, all running SDC pumps will trip. While in cold shutdown, the high temperature closure of V-17-19 is disabled. In this mode, with the reactor vented, the reactor water temperature cannot exceed saturation temperature (i.e., 212 °F) and thus remains below 350°F.

Based on the above, disabling the low suction pressure and the high suction temperature trips for all three SDC pumps does not increase the pump risk of operating under low suction pressure or high suction temperature conditions, while being protected against potential trips due to spurious operation or failure of these switches.

Summary of Conclusion for the Activitys 50.59 Review:

Revision 1 of the 50.59 evaluation supersedes previous revisions.

Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps is a change that adversely affects how any UFSAR described design function is operated or controlled. Therefore, this procedure revision requires a 50.59 Evaluation.

Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps does not result in more than a minimal increase in the frequency of an occurrence of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Enclosure - RA-11-005 Page 10 of 10 Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR, nor does it result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR, nor does it create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR.

Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps does not result in a design basis limit for a fission product barrier as described in the UFSAR being altered.

Revising Procedure 305 to include a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses for Oyster Creek.

Therefore, NRC approval is not required prior to issuing the Procedure 305 revision including a procedurally controlled TCC that disables the low suction pressure and the high suction temperature trips for all three SDC pumps.

The proposed activity does not require a License Amendment.