ML090760977
| ML090760977 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/11/2009 |
| From: | Flannigan R Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA 09-0054 | |
| Download: ML090760977 (12) | |
Text
WSLF CREEK I NUCLEAR OPERATING CORPORATION Richard D. Flannigan Manager Regulatory Affairs March 11, 2009 RA 09-0054 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Docket No. 50-482: Wolf Creek Generating Station Biennial 50.59 Evaluation Report Dear Gentlemen; This letter transmits the Biennial 50.59 Evaluation Report for Wolf Creek Generating Station (WCGS), which is being submitted pursuant to 10 CFR 50.59(d)(2). The attachment provides the WCGS Biennial 50.59 Evaluation Report including a summary of the evaluation results.
This report covers the period from January 1, 2007, to December 31, 2008, and contains a summary of 50.59 evaluations performed during this period that were approved by the WCGS onsite review committee.
This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4117, or Diane Hooper at (620) 364-4041.
Sincerely, Richard D. Flannigan RDF/rlt Attachment cc: E. E. Collins (NRC), w/a V. G. Gaddy (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET
Attachment to RA 09-0054 Page 1 of 11 WOLF CREEK NUCLEAR OPERATING CORPORATION Wolf Creek Generating Station Docket No.: 50-482 Renewed Facility Operating License No.: NPF-42 BIENNIAL 50.59 EVALUATION REPORT Report No.: 21 Reporting Period: January 1, 2007 through December 31, 2008
Attachment to RA 09-0054 Page 2 of 11
SUMMARY
This report provides a brief description of changes, test, and experiments performed at Wolf Creek Generating Station (WCGS) and evaluated pursuant to 10 CFR 50.59(c)(1).
This report includes summaries of the associated 50.59 evaluations that were reviewed and found to be acceptable by the Plant Safety Review Committee (PSRC) for the period beginning January 1, 2007 and ending December 31, 2008.
This report is submitted in accordance with the requirements of 10 CFR 50.59(d)(2).
On the basis of these evaluation of changes:
" There is less than a minimal increase in the frequency of occurrence of an accident previously evaluated in the Updated Final Safety Analysis Report (USAR).
- There is less than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the USAR.
- There is less than a minimal increase in the consequences of an accident previously evaluated in the USAR.
" There is less than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the USAR.
" There is no possibility for an accident of a different type than any previously evaluated in the USAR being created.
- There is no possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the USAR being created.
" There is no result in a design basis limit for a fission product barrier as described in the USAR being exceeded or altered.
There is no result in a departure from a method of evaluation described in the USAR used in establishing the design bases or in the safety analyses.
Therefore, all items contained within this report have been determined not to require a license amendment.
Attachment to RA 09-0054 Page 3 of 11 Evaluation Number: 59 2008-0001 Revision: 0
Title:
Evaluation of Voids in the Emergency Core Cooling System Suction PiDingl Activity
Description:
A configuration change, consisting of a total of approximately 10 cubic feet (at atmospheric pressure) of non-condensable gas, was identified at various locations in the suction and discharge piping of the Emergency Core Cooling System (ECCS) pump systems at Wolf Creek. This 50.59 evaluation analyzes the effect of the voids on the suction piping.
The configuration change for this magnitude of non-condensable gas presence is based upon a spectrum of typical voids, but this 50.59 evaluation is not limited to specific void locations. The location(s) of the non-condensable gas voiding would only allow introduction of the voids into the RCS during ECCS sump recirculation. However, this evaluation also considered the potential impact if the voiding was available to occur immediately, i.e., during ECCS injection mode.
50.59 Evaluation:
This system configuration change, consisting of the presence of a total of approximately 10 cubic feet (at atmospheric pressure) of non-condensable gas at various locations in the suction and discharge piping of the ECCS pump systems, has been evaluated.
Updated Safety Analysis Report transients affected include Loss of Coolant Accident (LOCA), Steam Line Break, Steam Generator Tube Rupture and rod ejection.
An evaluation pertinent to the effects of non-condensable gas voids of this magnitude on the spectrum of Large Break LOCA and Small Break LOCA transients indicated negligible effects. Also, an examination of the various non-LOCA transients affected was conducted. The capability of the ECCS providing borated water to the RCS would not be adversely affected and there would be no impact on the consequences of the Design Basis Accidents analyzed.
Attachment to RA 09-0054 Page 4 of 11 Evaluation Number: 59 2008-0002 Revision: 0
Title:
Evaluation of Voids in the Emergency Core Cooling System Dischargle Piping Activity
Description:
A configuration change, consisting of a total of approximately 10 cubic feet (at atmospheric pressure) of non-condensable gas was identified at various locations in the suction and discharge piping of the Emergency Core Cooling System (EGOS) pump systems at Wolf Creek. This included approximately one cubic feet in the hot leg injection piping of the Reactor Coolant System (RCS). This 50.59 evaluation analyzes the effect of the voids on the discharge piping.
The configuration change for this magnitude of non-condensable gas presence is based upon a spectrum of typical voids, but this 50.59 evaluation is not limited to specific void locations. The location(s) of the non-condensable gas voiding would only allow introduction of voids into the RCS during ECCS sump recirculation. However, this evaluation also considered the potential impact if the voiding was available to occur immediately, i.e., during ECCS injection mode.
50.59 Evaluation:
This system configuration change, consisting of the presence of a total of approximately 10 cubic feet (at atmospheric pressure) of non-condensable gas at various locations in the suction and discharge piping of the ECCS pump systems, has been evaluated. An evaluation of the pertinent effects of non-condensable gas voids of this magnitude on the spectrum of Large Break Loss of Coolant Accident and Small Break Loss of Coolant Accident transients, and an examination of the pressure surge analyses, indicated negligible effects. The safety function of the ECCS during hot leg recirculation and cold leg injection would be met.
Attachment to RA 09-0054 Page 5 of 11 Evaluation Number: 59 2008-0003 Revision: 0
Title:
Use of Dedicated Operator for the 'A' Safety Iniection Room Cooler Replacement Activity
Description:
A configuration change is proposed to substitute a manual action, performed by a dedicated operator, to locally start the containment spray (CS) pump 'A' room cooler, for an automatic action in the event that Safety Injection (SI) is initiated. This change is needed during the period that the SI pump 'A' room cooler is replaced at power. The dedicated operator action to activate the CS pump 'A' room cooler serves the purpose of providing backup cooling for the adjacent SI pump room. The operation of the CS pump room cooler is essential to assure the sustainability of the SI pump operation due to the potential time delay of the automatic CS actuation. For instance, certain design basis accidents such as Small Break Loss Of Coolant Accidents (SBLOCA) and main steamline breaks may not release sufficient mass and energy to the containment in a short time period following event initiation and consequently the CS actuation setpoint of 27 psig containment pressure may not be reached until a few minutes later. The CS pump room cooler provides adequate cooling capacity to cool both the CS pump room and the SI pump room.
50.59 Evaluation:
This system configuration change to substitute a temporary manual (operator) action, to locally start the CS pump 'A' room cooler, for an automatic action in the event that SI is initiated has been evaluated.
USAR transients affected include LOCA, Steam Line Break, Steam Generator Tube Rupture and rod ejection. Based upon the evaluation pertinent to the manual action substitution for automatic action on the spectrum of Large Break LOCA and SBLOCA transients, which indicated negligible effects, and also an examination of the various non-LOCA transients affected, the capability of the SI providing borated water to the Reactor Coolant System would not be adversely affected.
There would be no impact on the consequences of the Design Basis Accidents analyzed.
Attachment to RA 09-0054 Page 6 of 11 Evaluation Number: 59 2008-0004 Revision: 0
Title:
Main Steam Feedwater Instrumentation System Controls Replacement Activity
Description:
A design change to replace the existing electro-hydraulic actuated Main Steam Isolation Valves (MSIVs) and Main Feedwater Isolation Valves (MFIVs) with system medium actuated MSIVs and MFIVs is being made. The proposed replacement MSIVs and MFIVs are designed to utilize the system fluid (main steam or feedwater) as the motive force to open and close. The new actuators are simple pistons with the valve stem attached to both the discs and piston.
The valve actuation (open or close) is accomplished through positioning a series of six electric solenoid pilot valves to either direct the system fluid to the Upper Piston Chamber (UPC) and/or the Lower Piston Chamber (LPC), or vent either or both piston chambers. The six solenoid pilot valves are divided into two trains that are independently powered and controlled.
The subject of this 50.59 evaluation is a change in the failure mode of the MSIVs, from the existing fail-as-is to fail-closed, on loss of power to the pilot valves of either of the two closure trains. The replacement MSIVs and MFIVs are configured to fail-closed on loss of power to either of the two closure trains. Therefore, failure of a single power source will cause the replacement MSIVs to close, instead of fail-as-is.
50.59 Evaluation:
This change in component failure mode for the MSIVs from fail-as-is to fail-closed on loss of power to either of the two closure trains has been evaluated. Updated Safety Analysis Report transients affected include inadvertent closure of the main steam isolation valves. Based upon the evaluation, implementation of this proposed change will have a minimal to negligible impact on the overall internal events core damage frequency (CDF). The estimated impact on the CDF of a 0.07% increase is considered bounding, but remains insignificant. Since there is minimal increase in the transient event frequency, the current analyses remain bounding.
Attachment to RA 09-0054 Page 7 of 11 Evaluation Number: 59 2008-0004 Revision: 1
Title:
Main Steam Feedwater Instrumentation System Controls Replacement Activity
Description:
A design change to replace the existing electro-hydraulic actuated Main Steam Isolation Valves (MSIVs) and Main Feedwater Isolation Valves (MFIVs) with system medium actuated MSIVs and MFIVs is being made.
The proposed replacement MSIVs and MFIVs are designed to utilize the system fluid (main steam or feedwater) as the motive force to open and close. The new actuators are simple pistons with the valve stem attached to both the discs and piston.
The valve actuation (open or close) is accomplished through positioning a series of six electric solenoid pilot valves to either direct the system fluid to the Upper Piston Chamber (UPC) and/or the Lower Piston Chamber (LPC), or vent either or both piston chambers. The six solenoid pilot valves are divided into two trains that are independently powered and controlled.
The subject of this 50.59 evaluation is a change in the failure mode of the MSIVs, from the existing fail-as-is to fail-closed, on loss of power to the pilot valves of either of the two closure trains. The replacement MSIVs and MFIVs are configured to fail-closed on loss of power to either of the two closure trains. Therefore, failure of a single power source will cause the replacement MSIVs to close, instead of fail-as-is.
Revision 1 was initiated to add additional discussion about the MSIV failure being a Category II event.
50.59 Evaluation:
The inadvertent closure of the MSIVs is a USAR accident whose frequency of occurrence is credibly affected by a change in the MSIV failure mode from fail-as-is to fail-closed on loss of either of the two electrical trains providing power to the MSIVs.
Inadvertent closure of an MSIV is a Category II event, which is an incident of moderate frequency (could occur during a calendar year).
As described in the evaluation, implementation of this proposed change will have minimal to negligible impact on the overall internal events core damage frequency (CDF), with the estimated impact on CDF of a 0.07% increase (approximately 2E-08 delta CDF) being considered bounding, but remaining insignificant.
The combined initiating event frequency for loss of either of the two trains providing power to the MSIVs, which would cause MSIV closure, is 5.270E-03/yr.
This is an addition of one Category II event (inadvertent MSIV closure) every 190 years (approximately), which is a less than minimal increase in frequency.
Attachment to RA 09-0054 Page 8of 11 Evaluation Number: 59 2008-0005 Revision: 0
Title:
Polar Crane Modifications Activity
Description:
Specifications for the Containment Building Polar Crane are revised to incorporate the design and analysis requirements for polar crane load capacity analysis and load drop analysis. Previous revisions of these specifications did not address these requirements in detail. The items considered are (a) Polar crane load lifting capacity increase from 200 Tons to 260 Tons and (b) Qualified life change from 40 to 60 years.
50.59 Evaluation:
A design basis calculation, dated March 1989, analyzed 269 Tons as the load lifting capacity of the polar crane under Safe Shutdown Earthquake (SSE) conditions. This calculation was based on the SNUPPS Response Spectra. It was anchored at 0.25g SSE zpa value at 4% critical damping.
This meets all licensing commitments and is conservative.
Even though specification M-063 section 6.2.1 had 200 Tons as the lifting capacity under SSE conditions, appendices B and C had 260 Tons as the lifting capacity. The analyzed design information was not translated into these specifications.
Changing qualified life from 40 to 60 years is made to be consistent with the Life Extension project. The polar crane is a critical component and is under the Wolf Creek Generating Station preventive maintenance Program.
This component is not an American Society of Mechanical Engineers (ASME) Class I component and a slight increase in operating cycles of the polar crane will not cause any changes to the principal stresses.
Attachment to RA 09-0054 Page 9 of 11 Evaluation Number: 59 2008-0006 Revision: 0
Title:
WCGS Simplified Head Assembly Drop Analysis Activity
Description:
A calculation was performed to analyze the affects of a "Heavy Load" drop of the Simplified Head Assembly (SHA) during refueling outages.
This reanalysis was performed because of changes to the Reactor Vessel (RV) stiffness values and the Simplified Head Assembly (SHA) drop heights. The SHA drop heights have increased from 28 ft to 32 ft as a result of an increase to "Reactor Head Stand" height. The drop load has been reduced from the previously analyzed value of 375,000 lb to 357,000 lb.
This analysis is required because the polar crane was not procured as a "Single Failure Proof" crane. The analysis needs to meet the technical intent of (a) NRC Regulatory Issue Summary 2005-25, "Clarification of NRC Guidelines for Control of Heavy Loads,"
(b) NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" and (c) NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants."
The design basis calculation considered a weight of 375,000 lb and stiffness values from Westinghouse WCAP-9198 revision 0. The RV stiffness values were changed in WCAP-9198 revision 1. The "Lift" height considered in that analysis is 28.0 ft.
50.59 Evaluation:
This calculation analyzed the increased SHA drop height with the new SHA load of 357,000 lb. The drop heights have changed due to the height increase of the "Reactor Head Stand" modification. The increased drop height will not increase the frequency of a reactor head drop event; it will only affect the potential consequences of such an event.
The results show that the principle stresses at the RV nozzle is 81,000 psi compared with the allowable of 84,000 psi with the increased SHA drop height. The RV support system deflections are 1.0 inch for a 32 ft fall and are below the allowable 2.20 inch calculated in the design basis calculation. The maximum displacement of the RV will not have an effect on the ability of the reactor coolant loop piping and essential auxiliary piping to circulate borated water to the core and remove residual heat.
The core cooling capability and the fuel cladding integrity are maintained and there is no increase in radiological consequences.
The Critical Buckling load for the drive rods is based on reactor vessel design information provided in WCAP-9198 revision 0. Revision 1 of this document changed the "L" dimension from 148.75 inch to 138 in.
The method of calculating Critical Buckling load changed to include an effective length factor value of 0.80. The change in the Critical Buckling load value will not adversely affect the fuel assembly, and fuel cladding integrity will be maintained.
Attachment to RA 09-0054 Page 10 of 11 Evaluation Number: 59 2008-0007 Revision: 0
Title:
Polar Crane Load Capacity Reanalysis Activity
Description:
A calculation was performed to analyze the Polar Crane load carrying capacity under a safe shutdown earthquake (SSE).
This reanalysis was performed with boundary conditions that allowed two diagonally opposite snubbers to be removed and the other two snubbers retracted 0.75 inch during all plant modes. Alternately, all four snubbers could be retracted 0.75 inch without removal. Currently the snubbers are set at 0.25 inch from the main girder, to transfer the lateral forces from the polar crane to the main girder during an SSE.
A detailed stress analysis is performed under this calculation to verify all the components of the polar crane to be within the current licensing basis without the two lateral snubbers or with a 0.75 inch gap at the snubber locations. The end brackets, liner plate and rail stresses are also checked.
This analysis conforms to the requirements specified in specification M-063 revision 11 and M-900 revision 3. This analysis is required because the polar crane was not procured as a "Single Failure Proof" crane. The analysis will meet the technical intent of (a) NRC Regulatory Issue Summary 2005-25, "Clarification of NRC Guidelines for Control of Heavy Loads," (b)
NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" and (c) NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants."
50.59 Evalutation:
The reanalysis performed under this calculation, used boundary conditions that allowed the polar crane to move in the axial direction of the polar crane (Snubber direction) to see if the seismic deflections would exceed 0.75 inch.
The 0.75 inch gap is the allowable gap between the polar crane wheels and the rail.
If this deflection is exceeded the polar crane would derail and become a missile. The maximum SSE deflection found per this analysis is 0.364 inch.
Therefore, the polar crane will not derail and become a missile. This analysis meets all licensing commitments.
Attachment to RA 09-0054 Page 11 of 11 Evaluation Number: 59 2008-0008 Revision: 0
Title:
Use of Dedicated Operator for 'B' Safety Iniection Room Cooler Replacement Activity
Description:
A configuration change is proposed to substitute a manual action, performed by a dedicated operator, to locally start the containment spray (CS) pump 'B' room cooler, from an automatic action in the event that Safety Injection (SI) is initiated. This change is needed for the period of the SI pump 'B' room cooler replacement. The dedicated operator action to activate the CS pump 'B' room cooler serves the purpose of providing backup cooling for the adjacent SI pump room. The operation of the CS pump room cooler is essential to assure the sustainability of the SI pump operation due to the potential time delay of the automatic CS actuation. For instance, certain design basis accidents, such as small break Loss Of Coolant Accidents (SBLOCA) and main steamline breaks, may not release sufficient mass and energy to the containment in a short time period following event initiation and consequently the CS actuation setpoint of 27 psig containment pressure may not be reached until a few minutes later. The CS pump room cooler provides adequate cooling capacity to cool both the CS pump room and the SI pump room.
50.59 Evaluation:
This system configuration change to substitute a temporary manual (operator) action, to locally start the CS pump 'B' room cooler, from an automatic action in the event that SI is initiated has been evaluated. USAR transients affected include LOCA, Steam Line Break, Steam Generator Tube Rupture and rod ejection. Based upon the evaluation pertinent to the manual action substitution for automatic action on the spectrum of large break LOCA and SBLOCA transients, which indicated negligible effects, and also an examination of the various non-LOCA transients affected, the capability of the SI providing borated water to the reactor coolant system would not be adversely affected.
There would be no impact on the consequences of the-design basis accidents analyzed.