BSEP 10-0097, Report of 10 CFR 50.59 Evaluations and Commitment Changes

Report of 10 CFR 50.59 Evaluations and Commitment Changes

ML102430165
Person / Time
Site:	Brunswick
Issue date:	08/18/2010
From:	Mentel P Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 10-0097
Download: ML102430165 (14)

Text

Progress Pr G Energy En.ergy 10 CFR 50.59(d)(2)

AUG 1.8 2010 Serial: BSEP 10-0097 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Report of 10 CFR 50.59 Evaluations and Commitment Changes Ladies and Gentlemen:

In accordance with 10 CFR 50.59(d)(2), Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., is providing a report summarizing the 10 CFR 50.59 evaluations of changes, tests, and experiments implemented during the period from August 1, 2008, to July 31, 2010. This report is provided in Enclosure 1. In addition, a summary of commitment changes for the same period, made in accordance with CP&L's commitment management program (i.e., REG-NGGC-0 110, "Regulatory Commitments"), is provided in Enclosure 2.

No regulatory commitments are contained in this submittal. Please refer any questions regarding this submittal to Ms. Annette Pope, Supervisor - Licensing/RegulatorPrograms, at (910) 457-2184.

Sincerely, Phyllis N. Mentel Manager - Support Services Brunswick Steam Electric Plant MAT/mat

Enclosures:

1. Summary of Changes, Tests, and Experiments

2. Regulatory Commitment Change Summary Report Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant PO Box 10429 Southport, NC28461

Document Control Desk BSEP 10-0097 / Page 2 cc (with Enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510

BSEP 10-0097 Enclosure 1 Page 1 of 10 Summary of Changes, Tests, and Experiments Table of Contents Activity Title Pare Main Turbine Backup Pressure Regulator Out of Service 2 Unit 2 Design Modification to the Under-Vessel Carousel Platform 3 Control of Instrument Air Standby Compressors 4 Temporary Modification to Disable Compressor 5 SRM/IRM Spiking Criteria 6 Unit 2 Turbine Building HVAC Operation 7 Eliminate Removal of Reactor Protection System Shorting Links 8 Unit 1 Variable Frequency Drives 9 Reactor Building HVAC Pressure Controller 10

BSEP 10-0097 Enclosure 1 Page 2 of 10

Title:

Main Turbine Backup Pressure Regulator Out of Service Evaluation Identification Number: Action Request 301377 Brief

Description:

This procedure change revised the permissible range of power operation when the main turbine backup pressure regulator is out-of-service (PROOS) from when power levels are below, 23%

rated thermal power (RTP) or above 90% RTP to when power levels are below 50% RTP or above 90% RTP.

Summary of 10 CFR 50.59 Evaluation:

The scope of change affected the Pressure Regulator Failure - Closed (PRFC) event. The PRFC event credits the backup pressure regulator to mitigate the transient response. If the backup pressure regulator is out-of-service, normal pressure control will not be restored during the event, and the transient response becomes more severe as core power and system pressure rise.

However, raising the lower limit of power operation from 23% RTP to 50% RTP continues to ensure that the PRFC event remains bounded by the Power Load Unbalance Out-of-Service (PLUOOS) event presently analyzed for Brunswick operations below 50% RTP. As such, the PROOS event remains a non-limiting event in comparison to the PLUOOS event that already establishes the fuel operating limits.

Plant

References:

001-01.08, Control of Equipment And System Status, Revision 84

BSEP 10-0097 Enclosure 1 Page 3 of 10

Title:

Unit 2 Design Modification to the Under-Vessel Carousel Platform Evaluation Identification Number: Action Request 306643 Brief

Description:

The scope of change involves a modification to the under-vessel carousel platform. The modification to the carousel platform involves adding new platform extensions to both sides of the existing platform to fill the circular area inside the reactor pedestal. The new platform areas will be constructed of aluminum structural framing and the new grating will be aluminum bar grating. This modification was previously installed on Unit 1.

Summary of 10 CFR 50.59 Evaluation:

This change involved a modification to the under-vessel carousel platform where new platform areas, constructed of aluminum structural framing and grating, were added to both sides of the existing platform to fill the circular area inside the reactor pedestal. The addition of the aluminum structural members for the modification warranted evaluation because of the potential for hydrogen production during a Design Basis Loss-of-Coolant Accident (DBA LOCA). The amount of hydrogen that could potentially be produced as a result of this modification is of such a small magnitude, that it does not adversely impact the plants ability to monitor or control hydrogen in the event of a DBA LOCA.

Plant

References:

EC 66861

BSEP 10-0097 Enclosure 1 Page 4 of 10

Title:

Control of Instrument Air Standby Compressors Evaluation Identification Number: Action Request 316950 Brief

Description:

The instrument air standby compressors are being controlled OFF as an interim measure to prevent operation of this equipment until they can be permanently removed from service.

Summary of 10 CFR 50.59 Evaluation:

This change is an interim measure to control the instrument air standby air compressors OFF prior to being permanently removed from service. This change eliminated a small portion (i.e.,

30 scfm) of the service/instrument air but has been more than compensated for by the historical modifications which have brought the total system capacity up from approximately 1000 scfm to more than 2080 scfm and added the pneumatic nitrogen system (PNS) and nitrogen backup (BUN) subsystems.

Plant

References:

1-EC-05-145 2-EC-05-257

BSEP 10-0097 Enclosure 1 Page 5 of 10

Title:

Temporary Modification to Disable Compressor Evaluation Identification Number: Action Request 320909 Brief

Description:

This change was used to temporarily disable the 2A compressor of the Control Building Air Cooled Condensing Unit ID (i.e., 1-VA-1D-CU-CB). This action disabled the control circuit of the unit.

Summary of 10 CFR 50.59 Evaluation:

The change of the control circuit to isolate the 2A compressor in the 1-VA-1D-CU-CB reduced the total number of compressors available in the 1-VA-1D-CU-CB that can operate from four to three. This configuration resulted in a reduction in cooling capacity that affected the control room air conditioning system. However, the remaining three compressors were available to start and provide up to a nominal 30 tons of cooling. This provided an excess cooling capacity beyond that which is required for accident and normal conditions.

Plant

References:

EC 72525

BSEP 10-0097 Enclosure 1 Page 6 of 10

Title:

SRM/IRM Spiking Criteria Evaluation Identification Number: Action Request 321750 Brief

Description:

This activity provided a change to plant procedures to assist in making operability decisions and initial troubleshooting when Source Range Monitor (SRM) or Intermediate Range Monitor (IRM) spiking is occurring.

Summary of 10 CFR 50.59 Evaluation:

The procedure revisions provided a consistent, controlled method for Operations use to evaluate the impact of SRM or TRM spiking observed during plant operation while maintaining the current design basis of the SRM and IRM instrumentation. The spiking criteria provide objective measures of when the affected instrument should be considered inoperable. No other equipment other than the SRM and IRM instruments are affected by this change. Since spiking is conservative for the trip function, this change does not affect the SRM and IRM trip functions.

For monitoring, appropriate guidance is provided to ensure operators can properly monitor core flux and that in conditions where operators cannot monitor the flux appropriately, the instrument is declared inoperable and appropriate Limiting Conditions for Operation (LCOs) are observed.

Plant

References:

1OP-09, Neutron Monitoring System Operating Procedure, Revision 20 20P-09, Neutron Monitoring System Operating Procedure, Revision 27

BSEP 10-0097 Enclosure 1 Page 7 of 10

Title:

Unit 2 Turbine Building HVAC Operation Evaluation Identification Number: Action Request 320882 Brief

Description:

During the B219R1 refueling outage, the Unit 2 turbine building HVAC (TB HVAC) system was operated with only the exhaust fans operating to provide building ventilation during the replacement of supply fan dampers and the turbine building HVAC supply fan plenum.

Summary of 10 CFR 50.59 Evaluation:

During the operation of the Unit 2 TB HVAC without supply fans, the make-up air to the turbine building was not filtered and was not routed via the air duct system as normally operated.

During this time, Unit 2 was shut down and most equipment within the turbine building was not in operation. As such, the air flow pattern within the turbine building was not important since most equipment was not in use and the corresponding heat loads were greatly reduced. In addition, lack of filtration of the supply air was a concern because it is bounded by normal operating unfiltered inleakage when both units are operated in the recirculation mode.

Plant

References:

2SP-09-202, U2 TB HVAC OTV Operation Without Supply Fans, Revision 0

BSEP 10-0097 Enclosure 1 Page 8 of 10

Title:

Eliminate Removal of Reactor Protection System Shorting Links Evaluation Identification Number: Action Request 317090 Brief

Description:

This change allowed for control rod exercising and stroking of rods prior to tensioning of the head without requiring removal of the Reactor Protection System (RPS) shorting links, provided core verification and subsequent strongest rod out verification have been performed and the one-rod out refuel interlocks have been demonstrated to be operable.

Summary of 10 CFR 50.59 Evaluation:

Keeping the shorting links installed in the Reactor Protection System (RPS) scram circuit and not removing them in Mode 5 (i.e., Refuel) is acceptable because: (1) removal of the RPS circuitry shorting links is not assumed in any design-basis accident (DBA); (2) the SRM non-coincident high-flux full-scram logic is not credited in any DBA; (3) the IRM one-out-of-two-taken-twice full-scram logic provides the credited protection with respect to safety analysis; (4) the SRM and IRM control rod withdrawal block interlocks are not affected by the removal or installation of the RPS circuitry shorting links; and (5) refueling interlocks and shutdown margin requirements ensure that the reactor is maintained in a subcritical condition in Mode 5.

Plant

References:

UFSAR Change Package 09FSAR-001

BSEP 10-0097 Enclosure 1 Page 9 of 10

Title:

Unit 1 Variable Frequency Drives Evaluation Identification Number: Action Request 355253 Brief

Description:

The scope of this change replaced the existing recirculation pump motor-generators and related controls with Variable Frequency Drives (VFDs)

Summary of 10 CFR 50.59 Evaluation:

This change improved the reliability of the Reactor Recirculation System. The installed VFDs are provided with appropriate safety features, such as redundant controllers and backup protective devices, to prevent the possibility of their acting outside the design basis of the original motor-generators sets. The design parameters for the equipment installed as part of this activity have been deliberately chosen to closely match those of the equipment it is replacing.

The operating characteristics are essentially unchanged. A transient analysis has been performed to demonstrate that implementation of the VFD will not significantly impact fuel performance under accident or transient conditions.

Plant

References:

EC 68547 UFSAR Change Package 09FSAR-030

BSEP 10-0097 Enclosure 1 Page 10 of 10

Title:

Reactor Building HVAC Pressure Controller Evaluation Identification Number: Action Request 385540 Brief

Description:

This change provided Reactor Building (RB) HVAC System procedural guidance for lowering the 1-VA-PDC-1507 pressure controller set point from approximately -0.35 inches WC to -0.125 to -0.20 inches WC when secondary containment is not required.

Summary of 10 CFR 50.59 Evaluation:

During outages secondary containment requirements may be relaxed and contingency plans are required to make secondary containment functional. 1SP- 10-101 is implemented only when secondary containment is not required.

Plant

References:

1SP-10-101, Unit 1 RB HVAC Operation With The Setpoint For Rx Bldg Exhaust Fans Vortex Dampers Press Differential Controller Set Less Negative Than 0.25 Inches WC, Revision 0

BSEP 10-0097 Enclosure 2 Page 1 of 2 Regulatory Commitment Change Summary Report Table of Contents Activity Title Page Engineered Safety Feature (ESF) Status Boards 2

BSEP 10-0097 Enclosure 2 Page 2 of 2

Title:

Engineered Safety Feature (ESF) Status Boards Ori$inatin$ Document:

Letter from E. E. Utley (CP&L) to NRC, "Regulatory Operations Bulletin No. 73-6," dated December 13, 1974.

Original Commitment:

The December 13, 1974, letter stated, in part:

Operating Guideline 4, "Engineered Safety Feature Operability" describes procedures to be used by shift personnel in utilizing the emergency safety feature status board. This procedure assures that the current status of emergency safety feature equipment is always available to shift personnel. In addition, it is required that the shift foreman enter the status of plant equipment in the shift foreman's log at the completion of each shift.

Revised Commitment:

ESF status boards are no longer used and have been removed from the control room.

Basis:

Since original licensing of Brunswick, additional regulatory requirements have obviated the need for the ESF status boards. 10 CFR 50.65 (i.e., the Maintenance Rule) ensures that work is scheduled so that maintenance is performed in a manner that enhances the reliability and availability of systems and components that is commensurate with safety. In order to accomplish this, the plant site maintains a high level of knowledge of the operability of ESF systems.

Compliance with 10 CFR 50.65 has more than offset any benefit gained from the ESF Status Board.

Additionally, since 1974, Brunswick has implemented flow chart based Emergency Operation Procedures (EOPs). These symptom-based EOPs assist operators in accident and transient response and provide detailed guidance on confirmation of system response and taking manual actions, as necessary, if a system does not operate as required. As with the Maintenance Rule, this tends to obviate any benefit of the ESF status boards.

Finally, use of the ESF status board is not discussed in any plant procedure which impacts the way in which a UFSAR-described SSC design function is performed or controlled. There are no operating procedures or annunciator response procedures which require Operators to check the ESF status board to determine the status of ESF systems. Additionally, Operators are not trained to use the ESF status board in either initial or requalification training. Since no process or training is provided for the use of the ESF status board, its function is not required for the safe operation of the Units.

Based on the above discussion, the commitment for an ESF status board is no longer needed to ensure the safe operation of the units.