ML081710131

Submittal of Report of Changes Pursuant to CFR 50.59

ML081710131
Person / Time
Site:	Harris
Issue date:	06/12/2008
From:	Corlett D Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081710131 (8)

Text

Progress Energy SERIAL: HNP-08-050 10 CFR 50.59(d)(2)

JUN 1 2 2008 U. S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Ladies and Gentlemen:

In accordance with 10 CFR 50.59(d)(2), Carolina Power & Light Company (doing business as Progress Energy Carolinas, Inc.) submits the attached report for the Harris Nuclear Plant (HNP). The report provides a brief description of changes to the facility and a summary of the evaluations required per 10 CFR 50.59 for those items, regardless of implementation status, between September 16, 2006, and February 29, 2008.

This letter also informs the NRC that there have been no unreported changes in commitments made during the period from September 16, 2006, through February 29, 2008.

This letter contains no new regulatory commitments. Please contact me if you have any questions regarding this submittal at (919) 362-3137.

Sincerely, D. H. Corlett Supervisor, Licensing/Regulatory Programs Harris Nuclear Plant DHC/wrk

Attachment:

1. Report of Changes Pursuant to 10 CFR 50.59 c: Mr. P. B. O'Bryan, NRC Sr. Resident Inspector Ms. M. G. Vaaler, NRC Project Manager Mr. L. A. Reyes, NRC Regional Administrator, Region II Progress Energy Carolinas, Inc.

Harris Nuclear Plant P. 0. Box 165 S4f7 New Hill, NC 27562

Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number/ Description of Change Evaluation Summary Implementing Document 00195126 The minimum required incore There were 38 operating incore detector thimbles at the EC 64213, Rev. 0 detector thimbles were start of Cycle 14. However, the Movable Incore Detector reduced from 38 to 25 for System (MIDS) experienced several hardware problems.

Cycle 14. Engineering Therefore, the minimum number of detector thimbles was Change (EC) 64213 made the reduced from 38 to 25 for the duration of Cycle 14. This supporting changes for this reduction of active detector thimbles increased the thimble reduction to the Core calculated error band for the peaking factors. Therefore, Operating Limits Report the peaking factor limits were decreased and the minimum (COLR) and Plant Program number of operable detector thimbles per quadrant was Procedure PLP-106, increased from two to three as compensatory measures.

"Technical Specification This ensured that the peaking factors assumed as initial Equipment List Program and conditions in the safety analysis were not exceeded. This Core Operating Limits activity does not increase the frequency, likelihood of Report," Attachment 9, in occurrence or consequences of an accident or a accordance with step 9.5.2 of malfunction of structures, systems, and components (SSC) the Nuclear Generation important to safety more than minimally, does not create a Group Common Documents possibility for an accident of a different type or a NFP-NGGC-0018, "Core malfunction with a different result, does not result in a Operating Limits Report design basis limit being exceeded or altered, and does not Generation for HNP, RNP depart from a method of evaluation.

and CR3."

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number I Description of Change Evaluation Summary Implementing Document 00209005 The silica concentration This activity resulted in an increase to the silica PLP-715, Rev. 4 guidance for the reactor concentration limit in the Reactor Coolant System (RCS) coolant was changed to the from 2000 ppb to 3000 ppb. Silica concentration in the fuel manufacturer RCS is a diagnostic parameter, and allowance is being recommendations in the Plant made in PLP-715 for a concentration of 3000 ppb to be Program Procedure PLP-715, consistent with Fuel Vendor and EPRI Guideline "System Chemistry Strategic allowances for operation above 3000 ppb silica during the Plan," Rev. 4. first 30 days of power operation. AREVA and the EPRI Guidelines identify that zeolite forming elements required in combination with silica are controlled in the make-up sources to prevent zeolite formation. The procedural controlled chemistry target value for silica of 2000 ppb will remain to provide additional evaluation of the effect of silica and potential influence on the RCS system. The increase in silica concentration from 2000 ppb to 3000 ppb has been reviewed and the increase has been found not to increase the probability of zeolite induced fuel failures. The EPRI Primary Water Guidelines have identified that plant experience with operation at 2000 to 3000 ppb for the first two months of operations for several cycles has had no effect on fuel cladding corrosion. The limits that are in place for the cation species (Calcium, Magnesium, and Aluminum) have not changed in this revision, and as a result, the incidence of zeolite formation (which may result in fuel clad damage) is not increased. Therefore, it is determined that the increase in silica concentration does not pose an increase in the probability of fuel cladding damage or an increase in radiological consequences. This activity does not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number/ Description of Change Evaluation Summary Implementing Document 00256415 The silica concentration This activity resulted in an increase to the silica CRC-001, Rev. 47 guidance for the reactor concentration limit in the RCS from 2000 ppb to 3000 ppb.

coolant was changed to the Silica concentration in the RCS is a diagnostic parameter, fuel manufacturer and allowance is being made in CRC-001 for a recommendations in the concentration of 3000 ppb to be consistent with the Fuel Chemistry and Vendor and EPRI Guideline allowances for operation Radiochemistry Procedure above 3000 ppb silica during the first 30 days of power CRC-001, "HNP operation. AREVA and the EPRI Guidelines identify that Environmental and Chemistry zeolite forming elements required in combination with silica Sampling and Analysis are controlled in the make-up sources to prevent zeolite Program," Rev. 47. formation. The only credible accident associated with an increase in silica concentration from 2000 ppb to 3000 ppb is fuel failure as a result of increased fuel clad corrosion.

Operating with the higher silica concentration has been reviewed and experienced by several plants and this increase has been found not to increase the probability of zeolite induced fuel failures. The limits that are in place for the cation species (Calcium, Magnesium, and Aluminum) have not changed in this revision, and as a result, the incidence of zeolite formation (which may result in fuel clad damage) is not increased. These conditions have been documented in the EPRI Primary Water Guidelines as well as the fuel vendor's recommendations. The EPRI Primary Water Guidelines have identified that plant experience with operation at 2000 to 3000 ppb for the first two months of operations for several cycles has had no effect on fuel cladding corrosion. Therefore, it is determined that the increase in silica concentration does not pose an increase in the probability of fuel cladding damage or an increase in radiological consequences. This activity does not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number/ Description of Change Evaluation Summary Implementing Document 00211239 A new method of backflushing During Refueling Outage 13 the use of the Chemical and OP-120.02.39, Rev. 22 the Reactor Coolant Filter Volume Control System (CVCS) Demineralizers was lost was addedto Section 8.8 of when the Nitrogen Inlet Valve (1NI-119) for the Reactor Operating Procedure OP- Coolant Filter backwash system failed in the closed 120.02.39, "Fuel Handling position. This procedure change will allow the CVCS Building and Reactor Auxiliary Demineralizers to remain in service while the Reactor Building Filter Backflush," Coolant Filter is being backflushed to protect the Volume Rev. 22. This method will be Control Tank from corrosion products. The Reactor used when the plant is in acid Coolant Filter is a 25 micron filter. The Reactor Coolant reducing conditions during the Pump (RCP) seals are protected by the Seal Water cooldown cycle after Injection Filters which contain 5 micron filters. Therefore, shutdown for RCS cleanup, the Seal Water Injection Filters will protect the RCP when the Reactor Coolant Filter is being backflushed. This change is expected to reduce total dose for outage personnel. This activity does not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

00216674 EC 64030 was the design The Cycle 15 fuel design and neutronic design made EC 64030, Rev. 0 documentation for Cycle 15 changes to the core operating limits that affected the local that supported the fuel power peaking factors and the Axial Flux Difference (AFD) loading pattern and other which were documented in the COLR. The safety analyses plant configuration changes support Cycle 15 operation at a nominal core power level of including acceptance test 2900 MWt for up to 527 effective full-power days (EFPD).

criteria, Final Safety Analysis The analyses results including the Large Break Loss of Report (FSAR) changes, Coolant Accident (LBLOCA) and Non-Loss of Coolant procedure changes, key Accidents (NON-LOCA) Safety Analyses of Cycle 15 core performance parameters, and design satisfied the requirements and acceptance criteria cycle specific operational defined in the FSAR. This activity does not increase the data. frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number I Description of Change Evaluation Summary Implementing Document 00221554 EC 64893 modified the Boron The removal of the internals of 3BR-260 disabled the EC 64893, Rev. 0 Recycle System (BRS) piping and backflow restriction of the valve. This change allows removed the internals of valve 3BR- the transfer of boric acid from the RMTs to the BAT 260 to facilitate the transfer of pre- without using the temporary hoses that were batched concentrated boric acid required before the change. This simplifies solution from the Recycle Monitor transferring boric acid to the BAT and reduces Tanks (RMT) to the Boric Acid possible operational errors. The operator actions Storage Tank (BAT). It also added required to perform the new acid transfer evolution a normally open isolation valve to are within the normal skills of operators, for example, the BRS system piping. manipulation of valves. The operator actions will not replace any automatic plant functions. These activities will be procedurally controlled. These activities do not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, do not create a possibility for an accident of a different type or a malfunction with a different result, do not result in a design basis limit being exceeded or altered, and do not depart from a method of evaluation.

00236905 EC 67112 revised two analyses, The Movable Incore Detector System (MIDS) was EC 67112, Rev. 0 Large Break Loss of Coolant experiencing hardware problems during Cycle 14.

Accident (LBLOCA) and Axial Flux Therefore, LBLOCA and AFD analyses were Difference (AFD) that provide performed for safely operating with less than 38 flux additional steady state FQ margin for traces. The AFD operating region was reduced, the flux maps to be used after 15,000 Peak Clad Temperature limit was reduced and MWd/MTU for Cycle 14. several other changes were made to provide additional steady state FQ margin. This activity does not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number/I Description of Change Evaluation Summary Implementing Document 00243150 EC 67430 relocated the Emergency New containment recirculation sump screens were EC 67430, Rev. 0 Core Cooling System (ECCS) installed to resolve NRC Generic Letter 2004-02, recirculation sump level instruments "Potential Impact of Debris Blockage on Emergency LE- 1CT-716ASA and LE-01 CT- Recirculation During Design Basis Accidents at 716BSB from the upstream side of Press urized-Water Reactors." This required the the screens to the downstream side relocation of the ECCS recirculation sump level of the strainers, due to physical instruments. The instrumentation was designed with interferences with the new sump perforated plate to protect the operation of each strainers that were installed by EC instrument's float from debris-laden water. The 64377 (discussed below in Log instrument ranges and setpoints are the same as the Number 00246460). The old original design. These level instruments are still location of the instruments allowed used for the detection of an accident and for monitoring of the strainers to accomplishment/verification of accident mitigation.

determine if flow was being Only the monitoring for flow restriction function of restricted, which cannot be these instruments was eliminated, since the new accomplished by the new design. strainers were designed for operation in debris laden The new strainers are designed for water. The FSAR and emergency operating operation in debris-laden water and procedures were reviewed and revised to assure eliminate flow restriction problems. that all regulatory criteria were met and the Therefore, monitoring for flow descriptions were up to date. This activity does not restrictions is not a concern with the increase the frequency, likelihood of occurrence or new design. consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

00246460 EC 64377 replaced the containment The new strainers were installed to resolve NRC EC 64377, Rev. 1 recirculation sump screens to Generic Letter 2004-02. The new design utilized resolve NRC Generic Letter 2004- approved industry analytical methods that were 02, "Potential Impact of Debris consistent with current regulatory guidance to Blockage on Emergency demonstrate adequate Net Positive Suction Head Recirculation During Design Basis (NPSH) with up to 50% blockage and resistance to Accidents at Pressurized-Water air entraining vortices. This was a change in the Reactors." The new strainers were ESAR described methodology for the old design that designed to improve long-term used scale model testing coupled with analysis to Residual Heat Removal and demonstrate that under the worst case design basis Containment Spray sump conditions, adequate NPSH is always available to all recirculation capabilities with debris- ECCS pumps. This activity does not increase the laden post-accident water. frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and

________________________does

_______________ not depart from a method of evaluation.

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Attachment 1 to SERIAL: HNP-08-050 SHEARON HARRIS NUCLEAR POWER PLANT REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 Log Number Description of Change Evaluation Summary Implementing Document 00256968 EC 54865, Rev. '0 identified HNP implemented a safe shutdown re-validation EC 54865, Rev. 0 postulated fires that presented program to validate aspects of the fire safe design concerns that were not shutdown analysis, identify deficiencies, and addressed in the original fire determine corrective actions. In the safe shutdown protection design of the safe re-validation, additional components were shutdown components. Therefore, incorporated into the analysis to evaluate the impact compensatory measures that of postulated spurious operations on the capability of require additional manual actions safe shutdown systems to perform their safe were established in accordance with shutdown functions. New circuit analysis criteria the approved HNP fire protection have added failure modes not evaluated in the program (FPP-013) to correct the previous analysis. The result is that new manual design concerns. These actions must be credited in several plant fire areas in compensatory measures were order to achieve safe shutdown. The placed in effect until EC 54865, Rev. addition/revision of new manual actions has resulted 1 can be developed and approved to in the need to add or revise post fire plant operating make the physical changes required procedures. EC 54865, Rev 1 will make any to correct the design deficiencies. required physical plant changes and update the post EC 54865, Rev. 0 also described fire plant operating procedures. Compensatory the revisions to calculations, measures, in accordance with the approved HNP fire analyses, procedures, and other protection program (FPP-013), have been plant documents as necessary to established for each proposed operator manual implement these compensatory action. These compensatory measures were measures and satisfy the concerns established as a condition of turnover of the new of the Safe Shutdown Analysis safe shutdown program. The compensatory (SSA). No physical plant changes measures have been evaluated and verified as were authorized with EC 54865, feasible. This feasibility evaluation was performed Rev. 0. using licensed or certified Senior Reactor Operators and qualified Safe Shutdown Auxiliary Operators.

The safe shutdown re-validation demonstrated that safe shutdown can be achieved and maintained using compensatory measures. This was accomplished by achieving each of the individual safe shutdown functions of RCS inventory and pressure control, decay heat removal, required support functions, and monitoring critical plant parameters as specified in applicable NRC guidance documents. The safe shutdown re-validation did not affect the frequency of postulated fires, since it does not change the combustible inventory or fire loading of any plant fire areas. No new ignition sources or new activities that may contribute to the occurrence of fires were introduced. No physical changes to plant equipment were made. This activity does not increase the frequency, likelihood of occurrence or consequences of an accident or a malfunction of an SSC important to safety more than minimally, does not create a possibility for an accident of a different type or a malfunction with a different result, does not result in a design basis limit being exceeded or altered, and does not depart from a method of evaluation.

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