ML063170207

Month 10CFR50.59 Evaluation Summary Report for the Period July 1, 2004 Through June 30, 2006

ML063170207
Person / Time
Site:	Limerick
Issue date:	11/10/2006
From:	Degregorio R Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML063170207 (11)

Text

10CFR50.59(d)(2)

November 10, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

24-Month 10CFR50.59 Evaluation Summary Report For the Period July 1, 2004 through June 30, 2006 Attached is the 24-Month 10CFR50.59 Evaluation Summary Report for Limerick Units 1 and 2 for the period of July 1, 2004 through June 30, 2006, forwarded pursuant of 10CFR50.59(d)(2). The report includes brief descriptions of any changes, tests and experiments, including a summary of the evaluation of each. Four plant changes were implemented using 10CFR50.59 Evaluations during this 24-month period. The summaries of these changes are included in this report.

There are no commitments contained in this letter.

If you have any questions or require additional information, please do not hesitate to contact us.

Sincerely, Original signed by Chris Mudrick for Ron J. DeGregorio Vice President - Limerick Exelon Generation Company, LLC

Attachment:

Limerick Generating Station 24-Month 10CFR50.59 Evaluation Summary Report, July 1, 2004 through June 30, 2006 cc: S. J. Collins, Administrator Region I, US NRC S. L. Hansell, USNRC Senior Resident Inspector, LGS

10 CFR 50.59 Evaluation 24-Month Summary Report Limerick Generating Station 2006 Note: This report summarizes 10 CFR 50.59 Evaluations that were approved between July 1, 2004 and June 30, 2006.

Evaluation number: LG2003E003 Rev.1 50.59 Reviewer approval date: 4/16/04 PORC number: 04-035 WA 04-003 PORC approval date: 9/16/04 Implementing document: ECR 01-01233 Rev.1 Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [x] [ ]

Complete on: [ ] [ ] [x] [ ]

Title:

Evaluation of GE SIL 636 Impact - Unconservative Reactor Decay Heat Values Description of Activity:

Note: LG2003E003 Rev.1 incorporates NSRB suggested revised wording in Evaluation questions #3 and #4 to eliminate the use of the words estimated and expected to more positively describe the conclusions of the evaluations performed in support of the SIL 636 changes.

General Electric issued SIL 636 Revision 1 advising BWR owners that decay heat curves based on the ANSI/ANS-5.1-1979 Decay Heat Standard as used by GE may be non-conservative. The evaluation of the impact of SIL 636 is documented in ECR 01-01233.

In addition, the value of 8.0 BTU/hr-ft-F for thermal conductivity of AL-6XN tube material used in the Unit 1 RHR heat exchangers is non-conservative and is reduced to 6.8 BTU/hr-ft-F.

Reason for Activity:

Many of the design basis analyses performed by General Electric that are based on decay heats from ANS-5.1-1979 did not include the effects of actinides and activation products.

Previous evaluations concluded that the impacts from these sources were negligible.

However, current evaluations conclude that the contribution of actinides and activation products, collectively, may have a non-negligible impact on the total decay heat for long-term calculations. Analyzed events potentially impacted are the suppression pool response for LOCA, ECCS pump NPSH, shutdown cooling, alternate shutdown cooling, station blackout, fire safe shutdown, and ATWS. In addition, including 2 uncertainty for plant events further increases the values for decay heat to be used in the containment analyses.

The thermal conductivity of 8.0 BTU/hr-ft-F for AL-6XN tube material used in the Unit 1 RHR heat exchanger analyses is applicable to a metal temperature of 212F. However, typical average metal temperature in the heat exchanger tubes is approximately 150F. At this temperature, a value of 6.8 BTU/hr-ft-F for thermal conductivity is appropriate.

Effect of Activity:

Including the actinide and activation product decay heats per SIL 636 does not measurably increase the decay heat for the time period from shutdown to about 1000 seconds. However, for time periods greater than 1000 seconds, the decay heat (including the 2 uncertainty) could increase by as much as 10%. The impact of the decay heat increase is offset by existing conservatisms used in the affected plant analyses and maintaining the RHR heat exchangers in a cleaner operating condition. For example, most of the analyses performed in support of the power rerate project used a reactor power of 110% of original licensed power when only 105% was required to support the rerate licensed power increase. The increase in the decay heat rates and added uncertainty utilizes the margin inherent in the 5% power differential and also requires a nominal increase in RHR heat exchanger performance to maintain peak suppression pool temperatures within those determined for the power rerate project, thereby, assuring that design and licensing limits for containment response for the affected events are not exceeded.

Reducing the thermal conductivity from 8.0 to 6.8 BTU/hr-ft-F for AL-6XN material reduces the heat transfer capability of the Unit 1 RHR heat exchangers. This impact of this change is managed by allowing a smaller fouling factor and/or lower number of plugged tubes to satisfy heat transfer requirements.

Summary of Conclusion for the Activitys 50.59 Review:

The increase in decay heat rates results in an element of an analysis that is considered an adverse change and requires a 50.59 evaluation. The available margin (power level) from the power rerate analyses and requirements to maintain the RHR heat transfer capability assure that containment design parameters are not exceeded. Therefore, prior NRC approval is not required.

The reduction in thermal conductivity for AL-6XN tube material is also a change to the facility. Since the impact on the heat exchanger is minimal (less than 4%), and the Unit 1 RHR heat exchangers can still perform their design function, prior NRC approval is not required.

Evaluation number: LG2005E001 Rev.0 50.59 Reviewer approval date: 1/7/05 PORC number: 05-001 PORC approval date: 1/12/05 Implementing document: ECR 04-00620 Rev.0 Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [x] [ ]

Complete on: [ ] [ ] [x] [ ]

Title:

D12 EDG Jacket Water Coolant System Permanent Sight Glass Installation Description of Activity:

This activity involves the permanent installation of a sight glass in the LGS Emergency Diesel Generator (EDG) jacket water coolant system. This sight glass facilitates visual monitoring for detection of entrapped exhaust gases in the jacket water cooling system that if left uncorrected can cause a loss of EDG safety function due to jacket water cooling system pump gas binding.

As described in UFSAR Chapter 3, the EDG jacket water cooling system piping and components are classified as an ASME Section III, Class 3 but contains several non-ASME components that had additional quality assurance requirements imposed to make these non-ASME components, ASME equivalent, to meet the intent of the ASME Code and Regulatory Guide 1.26.

A 50.59 Evaluation was performed for this activity because the sight glass is a non-ASME Section III, Class 3 component that is being added to a section of ASME Section III, Class 3 piping. This sight glass is being procured, tested, and inspected in the same manner as the other non-ASME EDG jacket water cooling system components to make it ASME equivalent, which is an acceptable method as described in the UFSAR to meet the intent of the ASME Code and Regulatory Guide 1.26.

Although ECR 04-00620 is specifically applicable to the D12 EDG, this 50.59 Evaluation is applicable to all 8 LGS EDGS.

Reason for Activity:

Following the PBAPS dual unit SCRAM with a Loss of Offsite Power in August 2003, their E2 EDG tripped off after one hour of operation on low jacket water cooling system pressure. It was later found that the cause of the E2 trip was the entrapment of exhaust gases from cylinder adapter leakage in the jacket water cooling system. In response to this issue, PBAPS installed a sight glass on each of their EDGS to enable monitoring of the jacket water cooling system for the presence of entrapped exhaust gases.

In response to the PBAPS issue, it was decided to install a sight glass on each EDG at LGS for the purpose of monitoring for entrapped exhaust gases.

Effect of Activity:

The UFSAR described design function of the EDG cooling water system is to maintain the temperature of the diesel engine within a safe operating range under all load conditions and to maintain the engine coolant preheated during standby conditions to improve engine starting reliability. The installation of the sight glass maintains this design function because the sight glass is being procured, tested, and inspected to the highest available quality standards and provides visual monitoring for the detection of entrapped exhaust gases in the jacket water cooling system during routine surveillance testing. Therefore, based upon all available technical and licensing information, it is acceptable to install this sight glass because the EDG, and other SSCS will continue to operate safely and within analyzed Technical Specifications and UFSAR limits.

Summary of Conclusion for the Activitys 50.59 Review:

Installation of the sight glass in the EDG jacket water cooling system and the operation of the EDG with this sight glass installed is acceptable because the 50.59 Evaluation has concluded that the design, function, the method of performing functions of the EDGs, and plant response to design basis accidents and transients are maintained as described in the UFSAR and this activity. Therefore, this activity:

1. Does not require a License Amendment

2. Is compliant with NRC regulations

3. Is consistent with the Defense-In-Depth philosophy

4. Maintains sufficient safety margins

Evaluation number: LG2005E002 Rev.0 50.59 Reviewer approval date: 2/3/05 PORC number: 05-002 PORC approval date: 2/15/05 Implementing document: ECR 04-00672 Rev.0 S97.0.M Rev.17 M-041-201 Rev.0 Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [x] [ ]

Complete on: [ ] [ ] [x] [ ]

Title:

New Reactor Cavity Work Platform (AREVA Platform) and Associated Procedures Description of Activity:

This activity is the temporary installation and use by personnel of a new Reactor Cavity Work Platform (RCWP) for outage reactor internals maintenance activities, including the permanent installation of structural attachments to Reactor Enclosure structural members surrounding the reactor cavity to facilitate installation of the RCWP. The RCWP has been designed by Framatome-AREVA as a non-nuclear safety related, seismic category IIA temporary platform to be used for in-vessel maintenance/inspection activities during outages at the Limerick and Peach Bottom Stations. This 10 CFR 50.59 Review is for Limerick Units 1 & 2 only.

The RCWP will be fully assembled and then placed in the outage unit reactor cavity above the open Reactor Pressure Vessel (RPV). It will rest slightly submerged into the Reactor Cavity water supported by the refuel floor operating deck and a temporary "Y" shaped beam spanning the opening between the Reactor Cavity and Steam Dryer/Separator storage pool. The RCWP is an octagonal shaped work platform with four (4) personnel work baskets, including a jib crane, supported by eight radial legs, two of which extend onto the "Y" shaped temporary beam and six that extend onto the refuel floor operating deck. There is an opening in the shape to allow the transport of irradiated components between the reactor cavity and spent fuel pool. A rigging beam is used during the handling of the RCWP to span this opening temporarily. The RCWP inner edge extends just slightly beyond the inside diameter of the RPV and the RCWP jib crane extends several inches over the periphery of the core. Personnel will use the RCWP while refueling and in-vessel maintenance activities are in progress.

This 10 CFR 50.59 Review is applicable to the ECRs listed above and all the controlled document changes associated with these ECRs. (Note: Even if not listed above, a procedure change associated with these ECRs identified subsequently may be covered by this 50.59 Review without revising the 50.59 Review.)

Reason for Activity:

The RCWP is intended to reduce the time needed to complete refueling outage reactor vessel maintenance and refueling activities by performing more activities simultaneously.

Effect of Activity:

Reactor refueling and servicing procedures will be changed to address the RCWP.

Controls will be implemented to limit the radiation dose to personnel on the RCWP as low as reasonably achievable (ALARA). Engineering evaluations have determined that the RCWP can be installed, used and removed safely and effectively. Changes will be made to the UFSAR descriptions of reactor refueling and servicing equipment and operations, and the Fuel Handling Accident (FHA) analysis. The Technical Specifications Bases will be changed to add the basis for the Surveillance Requirements for the uptravel stops of the Refueling Platform fuel grapple assembly (active fuel) and Refueling Platform hoists (control rods). Administrative barriers and procedural controls will be in place to assure the fuel grapple assembly of the Refueling Platform does not collide with the RCWP jib crane during fuel handling in the reactor cavity. The permanently attached structural elements will have no impact on the performance, operation or control of structures, systems, or components (SSC) or analyses of SSC functions.

Summary of Conclusion for the Activitys 50.59 Review:

The UFSAR, the NRC SER for Limerick Units 1 & 2 Operating Licenses (NUREG-0991

& Supplements), Regulatory Guides 1.13 and 1.25, the associated General Design Criteria and other documentation were reviewed. The RCWP has been designed and constructed such that it does not fundamentally alter the existing means of controlling or performing any design functions, and it does not change any man-machine interface for performing design functions. The RCWP jib crane, however, does introduce one unwanted or previously unreviewed system or material interaction - a potential collision between the fuel grapple assembly and the RCWP jib crane over the core when handling fuel. A complete screening was performed and no other changes of this activity require a 10 CFR 50.59 Evaluation. This one adverse change of the activity has been evaluated under 10 CFR 50.59 Criteria (c)(2)(i-vii), and determined not to require prior NRC approval because the RCWP jib crane does not more than minimally increase the likelihood of a Fuel Handling System malfunction, cannot increase the consequences of an FHA or Fuel Handling System malfunction as previously evaluated in the UFSAR, and does not create a new type of accident or different malfunction result than previously evaluated in the UFSAR. Margin exists between the number of fuel pins predicted to fail during the design basis FHA using the current NRC-approved methodology in GESTAR II (NEDE-24011-P-A) and the current calculated number of fuel pins used to establish the bounding dose value in the UFSAR for the design basis FHA. Therefore, the RCWP jib crane could cause additional fuel pin failures during an FHA and still be bounded by

the dose value in the UFSAR because of this available margin. Compensatory measures will be implemented, as appropriate, to minimize the likelihood of a collision with the RCWP jib crane.

The RCWP will not affect the conditions, assumptions or methodology of the fuel handling accidents, will not affect shutdown cooling or coolant circulation, will not violate any design or licensing requirements/commitments, will not create any new failure modes for the fuel handling system and will not cause excessive radiation exposure to personnel. Design functions will be fulfilled and analyses of SSC will not be adversely affected by the RCWP. Existing controls for the handling of Heavy Loads are sufficient to address the RCWP, the supporting "Y" beam and the rigging beam. No license amendment or Tech Spec change is required.

Evaluation number: LG2005E003 Rev.0 50.59 Reviewer approval date: 2/17/05 PORC number: 05-005 PORC approval date: 2/23/05 Implementing document: ECR 03-00583 Rev.1 Unit 1 Unit 2 Units 1 & 2 Common Unit applicability: [ ] [ ] [x] [ ]

Complete on: [ ] [ ] [x] [ ]

Title:

Installation of Permanent Drywell Lead Blanket Shielding Description of Activity:

This activity installs permanent lead wool blanket shielding in the drywell for LGS Units 1 and 2 at several locations. A DCP-type ECR is provided for each location or group of locations where discreet Shielding Request Forms (SRFs) have been generated. This 50.59 addresses all locations inside the drywell where permanent blankets have been requested to be installed. Also included is storage inside the drywell during power operation of blankets temporarily installed during reactor shutdown periods.

Reason for Activity:

Currently, temporary shielding is installed at the beginning of a refueling outage to reduce personnel radiation exposure during maintenance or inspection activities in the area of local hotspots, and is subsequently removed prior to the plant's return to power.

These hotspots are associated with the Reactor Recirculation, the Reactor Water Cleanup System, and RHR shutdown cooling piping (Unit 1 only). Installation and removal of the temporary shielding and its supporting structures is time consuming and involves repeated radiation exposures which may be averted by using a permanent support and shielding arrangement. The location and thickness of the proposed permanent shielding is based on past ALARA experience during outages using temporary shielding. For some of the locations, it is not practical to permanently install the shielding, but where possible, long term dose savings could be achieved by the permanent installation of support systems and blankets for shielding. Where permanent shielding installation is not practical, shielding blankets temporarily installed during outages will be stored in selected locations within the drywell to minimize the personnel dose during shielding installation.

Effect of Activity:

The addition of lead wool blankets and their supporting systems introduces materials into the drywell environment that were not assumed to be present during prior analyses. The effects of these material additions on the following parameters or assessments of the drywell have been investigated and are discussed in the screening/evaluation.

Fire Hazard from blankets and supporting structures Seismic support of the lead wool blankets Seismic Capacity of Supporting Structures Seismic Separation LOCA Jet Impingement Loads/ECCS Suction Strainers Missile Hazard from LOCA Jet Impingement Hydrogen Generation Containment Net Free Volume Drywell Temperature Effects Containment sprays Summary of Conclusion for the Activitys 50.59 Review:

A 50.59 Screening Form was completed and four of five questions were answered "No".

As a result, a 50.59 Evaluation was performed to address the significance of the reduction in drywell free volume created by the shielding addition. The 50.59 Screening determined that there is no adverse impact on the functions of other SSC, including the drywell sprays, drywell HVAC, hydrogen generation, recirc system, RWCU system and RHR shutdown cooling piping. The shield blankets are qualified to be placed in the containment environment and are installed and supported in accordance with Seismic I or IIA requirements, as applicable. It has been determined that the addition of the shielding does have an adverse impact on the containment due to the post LOCA peak pressure increase or combustible gas concentration that result from the reduction in drywell free volume. The 50.59 Evaluation concludes that the impact of the drywell free volume change is less than minimal and that a Tech Spec change and license amendment are not required. Therefore, approval from the NRC is not required for installation and/or storage of the specified shielding in the drywell.