ML073440385
| ML073440385 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/04/2008 |
| From: | Markley M NRC/NRR/ADRO/DORL/LPLIV |
| To: | Bannister D Omaha Public Power District |
| Wang A, NRR/DORL/LPL4, 301-415-1445 | |
| Shared Package | |
| ML073440430 | List: |
| References | |
| TAC MD6808 | |
| Download: ML073440385 (16) | |
Text
February 4, 2008 Mr. David J. Bannister Site Director Omaha Public Power District Fort Calhoun Station FC-2-4 Post Office Box 550 Fort Calhoun, NE 68023-0550
SUBJECT:
FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:
ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE (TSTF) 445-A, REVISION 1 (TAC NO. MD6808)
Dear Mr. Bannister:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 252 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1.
The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 21, 2007.
The amendment will revise the TS safety limit (SL) requirements related to the use of a non-cycle specific peak linear heat rate SL of 22 kilowatt hours per foot (kW/ft) to fuel centerline melt. The proposed change is consistent with the objective of the TS Task Force (TSTF) 445-A, Revision 1, which determined that the current SL does not clearly conform to paragraph 50.36(d)(1)(ii)(A) of Title 10 of the Code of Federal Regulations. TS Section 1.1, Safety Limits - Reactor Core, is revised to incorporate the TSTF-445-A, Revision 1, peak fuel centerline temperature criteria. Also, TS 1.2, Safety Limits - Reactor Coolant System Pressure, is revised to incorporate the SL violation action which is currently delineated in administrative control TS 5.7.1. TS Section 1.3, Limiting Safety System Settings [LSSS], is being relocated to the currently unused TS Section 2.13 to be more consistent with the content of the Combustion Engineering Standard TS (i.e., the LSSS will be located in the Limiting Conditions for Operation (LCO) section of the FCS TS which is similar to the LCO/Surveillance Requirements Section 3.0 of the TS).
The administrative control in TS 5.7.1, Safety Limit Violation, is deleted. Also, administrative control TS 5.9.5, Core Operating Limits Report (COLR), item a., is revised to add TS 2.13, "RPS [Reactor Protective System] Limiting Safety System Settings," Table 2-11, Items 6, 8, and 9, to the list of items that shall be documented in the COLR. The TS Table of Contents (TOC) is also updated to reflect the deletion and subsequent renumbering of Section 1.3 and Table 1-1 to TS 2.13 and Table 2-11, respectively. The TOC is also updated to delineate the new TS subsections 1.1.1 and 1.1.2, provide the revised titles for TS 1.0, 1.1, 1.2, and 2.13, and to reflect TS 5.7.1 as Not used."
D. J. Bannister A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RA/
Michael T. Markley, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285
Enclosures:
- 1. Amendment No. 252 to DPR-40
- 2. Safety Evaluation cc w/encls: See next page
Pkg ML073440430, Amdt./License ML073440385, TS Pgs ML073450566
(*) SE input memo
(**) See previous concurrence NRR-058 OFFICE DORL/LPL4/PM DORL/LPL4/LA DIRS/ITSB/BC OGC DORL/LPL4/BC NAME AWang, MTM for JBurkhardt**
TKobetz**
MLoftus**
THiltz DATE 1/4/08 1/3/08 1/16/08 1/28/08 2/4/08
Ft. Calhoun Station, Unit 1 Updated 11/26/07 cc:
Winston & Strawn ATTN: James R. Curtiss, Esq.
1700 K Street, N.W.
Washington, DC 20006-3817 Chairman Washington County Board of Supervisors P.O. Box 466 Blair, NE 68008 Mr. John Hanna, Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 310 Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. Joe L. McManis Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
P.O. Box 550 Fort Calhoun, NE 68023-0550 Ms. Melanie Rasmussen Radiation Control Program Officer Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319
OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 252 Renewed License No. DPR-40
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Omaha Public Power District (the licensee), dated September 21, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 252, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented prior to startup from the 2008 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications Date of Issuance: February 4, 2008
ATTACHMENT TO LICENSE AMENDMENT NO. 252 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
License Page REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT TOC Page 1 TOC Page 1 TOC Page 2 TOC Page 2 TOC Page 3 TOC Page 3 TOC Page 4 TOC Page 4 TOC Page 5 TOC Page 5 TOC Page 6 TOC Page 6 TOC Page 7 TOC Page 8 1.0 - Page 1 1.0 - Page 1 1.0 - Page 2 1.0 - Page 2 1.0 - Page 3 1.0 - Page 3 1.0 - Page 4 1.0 - Page 4 1.0 - Page 5 1.0 - Page 5 1.0 - Page 6 1.0 - Page 7 1.0 - Page 8 1.0 - Page 9 1.0 - Page 10 1.0 - Page 11 1.0 - Page 12 2.13 - Page 1 2.13 - Page 1 2.13 - Page 2 2.13 - Page 2 2.13 - Page 3 2.13 - Page 3 2.13 - Page 4 2.13 - Page 4 2.13 - Page 5 2.13 - Page 5 5.0 - Page 5 5.0 - Page 5 5.0 - Page 8 5.0 - Page 8
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 252 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285
1.0 INTRODUCTION
By application dated September 21, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072670206), Omaha Public Power District (OPPD, the licensee) requested changes to the Technical Specifications (Appendix A to Renewed Facility Operating License No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS).
The proposed amendment would revise the Technical Specifications (TS) safety limit (SL) requirements related to the use of a non-cycle specific peak linear heat rate (PLHR) SL of 22 kilowatts per foot (kW/ft) to fuel centerline melt (FCM). The proposed change is consistent with the objective of the TS Task Force (TSTF) 445-A, Revision 1, which determined that the current SL does not clearly conform to paragraph 50.36(d)(1)(ii)(A) of Title 10 of the Code of Federal Regulations (10 CFR). TS Section 1.1, Safety Limits - Reactor Core, is revised to incorporate the TSTF-445-A, Revision 1, peak fuel centerline temperature (PFCT) criteria. Also, TS 1.2, Safety Limits - Reactor Coolant System Pressure, is revised to incorporate the SL violation action which is currently delineated in administrative control TS 5.7.1. TS Section 1.3, Limiting Safety System Settings [LSSS], is being relocated to the currently unused TS Section 2.13 to be more consistent with the content of the Combustion Engineering (CE) STS (i.e., the LSSS will be located in the Limiting Conditions for Operation (LCO) section of the FCS TS which is similar to the LCO/Surveillance Requirements (SR) Section 3.0 of the STS).
The administrative control in TS 5.7.1, Safety Limit Violation, is deleted. Also, administrative control TS 5.9.5, Core Operating Limits Report (COLR), item a., is revised to add TS 2.13, "RPS [Reactor Protective System] Limiting Safety System Settings," Table 2-11, Items 6, 8, and 9, to the list of items that shall be documented in the COLR. The TS Table of Contents (TOC) is also updated to reflect the deletion and subsequent renumbering of Section 1.3 and Table 1-1 to TS 2.13 and Table 2-11, respectively. The TOC is also updated to delineate the new TS subsections 1.1.1 and 1.1.2, provide the revised titles for TS 1.0, 1.1, 1.2, and 2.13, and to reflect TS 5.7.1 as Not used.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations, Section 50.36 requires that each licensee authorizing operation of a production or utilization facility include technical specifications. In particular, one type of technical specification is a limiting safety system setting. 10 CFR 50.36(d)(1)(ii)(A) states that the LSSS are settings for automatic protective devices related to those variables having safety significant functions.
The proposed change will replace the PLHR SL with a PFCT SL. In accordance with 10 CFR 50.36(d)(1)(ii)(A), LSSS must be chosen such that automatic action will prevent a SL from being exceeded. This assessment is applicable during steady state operations and Anticipated Operational Occurrences (AOOs). The NRC staff recognized that the PLHR SL of 22 kW/ft would be exceeded for an AOO. Therefore, conformance with 10 CFR 50.36 was not being clearly demonstrated. This change is being undertaken to conform with 10 CFR 50.36(d)(1)(ii)(A), which requires that LSSSs prevent an SL from being exceeded during normal operations and AOOs.
In addition, FCS was licensed for construction prior to May 21, 1971, and is committed to the draft General Design Criteria (GDC). These draft GDC are contained in Appendix G of the FCS updated safety analysis report (USAR). The preliminary design criterion which relates to this license amendment request LAR is GDC 14, Core Protective Systems.
The proposed FCM-based PLHR SL will be consistent with the FCS USAR Appendix G, GDC 14, Core Protective Systems, which states the following:
Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.
3.0 BACKGROUND
During review of the Waterford Steam Electric Station, Unit 3 (Waterford), 10 CFR Part 50, Appendix K, Margin Recovery Power Uprate request, the U.S. Nuclear Regulatory Commission (NRC) staff recognized that the PLHR SL of 21 kW/ft would be exceeded for an AOO. In accordance with 10 CFR 50.36(d)(1)(ii)(A), LSSSs must be chosen such that automatic action will prevent an SL from being exceeded. This assessment is applicable during steady state operations and AOOs. Therefore, conformance with 10 CFR 50.36 was not being clearly demonstrated. A similar condition exists for other plants within the CE fleet.
At Waterford, the current steady state limit of 21 kW/ft is exceeded during two AOOs. However, the corresponding PFCT does not exceed the melting point during these events. The affected AOOs are the Control Element Assembly (CEA) Withdrawal Events from both Subcritical and at Low Power Startup conditions. The analysis for these events results in the 21 kW/ft limit being exceeded, although this had been previously reviewed and found to be acceptable by the NRC staff for at least two plants.
In preparation for the core reload analysis to be performed by OPPD and AREVA, and through review of industry operating experience, it was identified that Reference 8.9 was available to justify changing the present FCS safety limit (SL) of 22 kW/ft to that of the more appropriate
FCM temperature. Use of 22 kW/ft could be exceeded for selected AOOs without having FCM occur. Therefore, as noted above, conformance with 10 CFR 50.36 is not clearly demonstrated for FCS.
At FCS, the current steady state PLHR SL is 22 kW/ft. While the 22 kW/ft may be exceeded momentarily during some AOOs for FCS (such as the CEA Drop, Excess Load, and Loss of Feedwater Heating, and also potentially for the CEA Withdrawal), the PFCT does not significantly challenge or exceed the melting point during these events.
3.1 TECHNICAL ANALYSIS
The intent of the PLHR SL is to prevent the fuel centerline temperature (FCT) from exceeding the melting point, which conservatively assures that there will be no breach in cladding integrity.
The current 22 KW/ft limit was historically chosen as a conservative limit at which the fuel can operate without causing the FCT to exceed the melting point and is a parameter that can be monitored directly by the operators in the Control Room.
For the AOOs identified above, calculations have shown that FCT remains below the melt temperature at linear heat rates of 22 kW/ft. While the AOO analyses show that the PLHR may exceed 22 kW/ft, the FCT does not exceed the melt temperature, thereby fully satisfying the intent of the SL.
Per Appendix G of the FCS USAR, Responses to 70 Criteria, Criterion 14, Core Protective Systems," the acceptance criteria for normal operation and AOOs is that the acceptable fuel design limits shall not be exceeded. The specified acceptable fuel design limit (SAFDL) of interest, in this case, is the PFCT limit. Although FCS was not licensed under the Standard Review Plan (SRP), this SAFDL is discussed in detail in SRP Section 4.2 which states:
(II)(A)(2)(e) "Overheating of Fuel Pellets: It has also been traditional practice to assume that failure will occur if centerline melting takes place. For normal operation and anticipated operational occurrences, centerline melting is not permitted The centerline melting criterion was established to assure that axial or radial relocation of molten fuel would neither allow molten fuel to come into contact with the cladding nor produce local hot spots. The assumption that centerline melting results in fuel failure is conservative."
Therefore, a more representative SL than the current generic PLHR would be one that is based upon the PFCT. A PFCT SL would address both normal operation and AOOs. In addition to the FCS USAR Appendix G, Criterion 14, a PFCT SL would also be consistent with the STS (as discussed in Reference 8.8), the SRP, 10 CFR 50.36, and the FCS licensing basis.
The melting point of the fuel is dependent on fuel burnup and the amount and type of burnable poison used in the fuel. For AREVA fuel, the design melting point of unirradated fuel containing no burnable poison is 5081 degrees Fahrenheit (°F) (Reference 8.14). The melting point is adjusted downward from this temperature depending on the amount of burnup and amount and type of burnable poison in the fuel. The adjustment for burnup of 58 °F/10,000 megawatt days per metric ton unit (MWD/MTU) is consistent with STS as discussed in Section 4.2 of the licensees submittal, and is the same as reported in References 8.7, 8.8, and 8.9. The burnable poison adjustments are determined in accordance with Reference 8.12 for fuels containing
gadolinium. The specific formula for adjustment to the burnable poison is considered to be proprietary information and, therefore, can not be included in this application. The mode of applicability and actions required, if the proposed FCM-based PLHR SL is exceeded, would be the same as they are for the current PLHR SL; the only difference is the use of a cycle-specific PLHR corresponding to FCM which will be updated every fuel cycle and included in the COLR versus use of a generic PLHR which does not correspond to FCM.
Therefore, a PFCT SL of less than 5081 °F decreasing by 58 °F per 10,000 MWD/MTU adjusting for burnup and burnable absorbers per Reference 8.14 is more appropriate, from a verbatim compliance perspective, than the current PLHR-based SL. The proposed FCM-based PLHR SL will:
address both normal operations and AOOs, be consistent with FCS USAR Appendix G, Responses to 70 Criteria, be consistent with SAFDLs, be consistent with SRP acceptance criteria, be consistent with the FCS current licensing basis, be determined using NRC-approved methodologies, and clearly conform to 10 CFR 50.36(d)(1)(ii)(A).
3.1.1 TSTF-445-A, Revision 1 TS Changes Although the licensee is not licensed under the SRP, the change to establish the PFCT as the SL is consistent with the SRP for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The PFCT is the basis for protecting the fuel and is consistent with the analogous TS wording for Westinghouse and Babcock & Wilcox designed plants. In addition, the licensee has also proposed to make several editorial and administrative changes to the TS to more clearly align with the CE STS 2.0.
3.1.2 TS Section 1.1 The PFCT SL has been added to current TS Section 1.1. The wording for this is verbatim to the language in TSTF-445-A. The verbiage from the TSTF-445-A, Revision 1, has been added as TS Section 1.1.1(b). Therefore, as discussed in Section 3.2, the NRC staff concludes that this change is acceptable and is consistent with the TSTF-445-A, Revision 1, and the STS.
3.1.3 Administrative and Editorial Changes As noted above, the licensee has made several administrative and editorial changes to more align its current custom TS with the CE STS.
3.1.4 TS Section 1.0 The licensee has proposed to revise the title of TS Section 1.0, Safety Limits and Limiting Safety System Settings to Safety Limits, which is consistent with Section 2.0 title of the STS.
The NRC has reviewed this change and concludes that it is editorial in nature and, therefore, is acceptable.
3.1.5 TS Section 1.1 The licensee has proposed to revise the title for TS Section 1.1, Safety Limits - Reactor Core, to Safety Limits (SLs) to be consistent with STS Section 2.1, SLs. In addition, current TS Section 1.1 has been modified to more align with the CE STS by renaming current TS Section 1.1 to Section 1.1.1, Reactor Core SLs and denoting the only current specification as (a). As noted in Section 3.3.1, the new PFCT SL has been added as TS 1.1.1(b). No other changes were made to the current TS.
In addition, current TS 1.2, Safety Limit, Reactor Coolant System Pressure, has been relocated in its entirety and unchanged to new TS Section 1.1.2, Reactor Coolant System Pressure SL.
This was done to more align with the content of STS 2.1.2, Reactor Coolant System Pressure SL.
The NRC has reviewed these changes and concludes that they are editorial in nature and, therefore, are acceptable.
3.1.6 Deletion of TS 5.7.1 and the Addition of New TS Section 1.2 The licensee has proposed to delete TS 5.7.1 and replace it with a new TS Section 1.2, Safety Limit Violations. The new TS Section 1.2 will meet the intent of TS 5.7.1. In addition, the new TS 1.2 will be consistent with the content of STS 2.2, Safety Limit Violations, as it defines SLs for two categories (1.2.1, Reactor Core and 1.2.2, Reactor Coolant Pressure) and provides additional limitations for the reactor coolant pressure SL that did not exist in TS 5.7.1. A discussion of the new TS 1.2 is provided below:
3.1.6.1 Reactor Core SLs The Reactor Core SLs only apply in MODES 1 and 2 (STS) because these are the only modes in which the reactor is critical. In MODES 3, 4, 5, and 6, applicability is not required as the reactor is not generating significant thermal power.
For the licensee, Reactor Core SLs would only apply to MODES 1 and 2, because these are the only modes in which the reactor is critical and in operating MODES 3, 4, and 5, applicability is not required since the reactor is not generating significant thermal power.
Therefore, going to MODE 3, or HOT SHUTDOWN, places the unit in a MODE in which the SL is not required. The allowed completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operations where this SL is not applicable and reduces the probability of fuel damage.
3.1.6.2 Reactor Coolant System (RCS) Pressure SL If the RCS pressure SL is violated when the reactor is in MODES 1 or 2, the requirement for STS is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
For the licensee, this would be HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, which provides the operator time to complete the necessary actions to reduce RCS pressure by terminating the cause of the pressure increase, removing mass or energy from the RCS or a combination of these actions, and to establish MODE 3 (HOT SHUTDOWN) conditions. This is consistent with the existing TS requirement to be in HOT SHUTDOWN (MODE 3) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if an SL is violated.
The licensee has added a new requirement for the RCS pressure SL TS 1.2.2 to be consistent with the STS 2.2. If the RCS pressure SL is exceeded while in operating MODES 3, 4, or 5, RCS pressure must be reduced to less than the SL within 5 minutes, with no mode change, as this would require temperature reduction. This 5-minute requirement does not currently exist in the licensees TS; therefore, it was added in the proposed change to be consistent with the content of the TSTF-445-A, Revision 1. This change is more conservative than the current TS, is consistent with the STS 2.2, and, therefore, is acceptable.
The new TS 1.2 does deviate from the STS in that it requires the plant to be placed in HOT SHUTDOWN if an SL is violated rather than MODE 3. The HOT SHUTDOWN requirement is consistent with the current TS 5.7.1. HOT SHUTDOWN for the licensee is where the average temperature (Tavg) of the reactor coolant is greater than or equal to (>) 515 °F and the reactor is subcritical. Operating MODE 3 or Hot Standby for the STS is where the Tavg of the reactor coolant is 350 °F and subcritical. Operating MODE 4 or HOT SHUTDOWN, for STS is where the reactor coolant Tavg is between 350 °F and 200 °F. The licensee does not currently have TS-defined operating modes for these temperature ranges. Operating MODE 4 as defined for the licensee is COLD SHUTDOWN which is Tcold less than or equal to () 210 °F. Therefore, the definition of Operating MODE 3 for STS is most similar to the licensees definition of HOT SHUTDOWN. As the conditions from the existing TS SL violation requirement is maintained, and the licensees most similar mode to MODE 3 of the STS is HOT SHUTDOWN, the NRC staff finds this change acceptable.
3.1.7 TS Section 1.3 The licensee has relocated TS Section 1.3, Limiting Safety System Settings, RPS, to the currently unused TS Section 2.13 to be more closely aligned with the CE STS in that the LSSS will be located in the LCO Section 2.0 of the FCS TS which is similar to the LCO/SR Section 3.0 of the STS. The Bases for this section has been revised to incorporate TSTF-445-A, Revision 1.
TS Bases Section 1.3(8), "Axial Power Distribution," states that the RPS trip setpoint will ensure that a PLHR of 22 kW/ft (deposited in the fuel) will not be exceeded as a result of axial power maldistributions. TS Bases Section 1.3(9), "Steam Generator Differential Pressure," states that this reactor protective system trip setpoint will ensure that a peak linear heat rate of 22 kW/ft will not be exceeded as a result of the loss of load to one steam generator. The proposed change replaces 22 kW/ft (deposited in the fuel) in TS Bases 1.3(8) and TS Bases 1.3(9) with a fuel centerline temperature greater than the safety limit corresponding to Fuel Centerline Melt (FCM),
as determined each fuel cycle and contained in the COLR." The general numbering scheme for TS Section 1.3 is revised to coordinate with TS Section 2.13, accordingly.
TS Section 1.3 has been relocated in its entirety and unchanged to TS Section 2.13. The NRC staff has reviewed this change, concluded it is editorial in nature, and, therefore, is acceptable.
3.1.8 TS Section 5.7.1 The administrative control TS Section 5.7.1, Safety Limit Violation, is being deleted, and will be delineated as Not used. As discussed in Section 3.4.3, the SL violation will be incorporated into new TS Section 1.2 to meet the intent of the TSTF-445-A, Revision 1, and the STS content conversion. The NRC staff has reviewed this change, concludes that the intent of the TS 5.7.1 is maintained in new TS 1.2, is consistent with the STS, and, therefore, is acceptable.
3.1.9 Figure 1.1 The licensee has proposed to add the title,1.0 SAFETY LIMITS, 1.1 Safety Limits (SLs)
(continued), to Figure 1.1. No other changes were made to Figure 1.1. The NRC staff has reviewed this change and agrees that it is editorial in nature and therefore acceptable.
3.1.10 Table of Contents (TOC)
The Table of Contents (TOC) is also being updated to reflect the deletion of TS Sections 1.3 and 5.7, as well as the re-use of TS Sections 1.2 and 2.13. The TOC will also be changed to reflect the new section titles for Sections 1.0, 1.1, 1.1.1, and 1.1.2 and the numbering scheme for Table 1-1, "RPS Limited Safety System Settings," which is changed from Table 1-1 to Table 2-11. The NRC staff has reviewed these changes and concludes that they are editorial in nature and, therefore, are acceptable.
3.1.11 TS 5.9.5, Core Operating Limits Report (COLR)"
The administrative controls in TS 5.9.5, "Core Operating Limits Report (COLR), is being revised to remove the existing TS Bases section 1.3(4), Thermal Margin/Low Pressure Trip, and Section 1.3(8), Axial Power Distributions," from item a and to add TS 2.13, Table 2-11, Items 6, 8, and 9, to the list of items that shall be documented in the COLR. Items 6 and 8 in Table 2-11 correlate to 1.3(4) and 1.3(8), respectively, and the NRC staff concludes this is an administrative change and, therefore, is acceptable. In addition, the licensee added item 9, Steam Generator Differential Pressure, as the basis for this item also was changed as a result of the new PFCT SL and, therefore, should also be documented in the COLR.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on November 6, 2007 (72 FR 62690). Accordingly, the
amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3)the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. As a result, the proposed change conforms with 10 CFR 50.36(d)(1)(ii)(A) and is approved.
7.0 PRECEDENT The proposed change is consistent with the objective of TSTF-445-A, Revision 1 (Reference 8.9). The licensee proposes to implement the change through the use of the NRC-approved AREVA reload analysis methodology, EMF-1961(P)(A), in lieu of the NRC-approved CE-Westinghouse methodology, CENPD-382-P-A. Precedence for use of the AREVA (formerly Siemens Power Corporation) methodology was set for similar changes at Millstone Unit No. 2 in References 8.2, 8.3, and 8.4. NRC staff approval of the AREVA methodology at FCS is delineated in Reference 8.1 as discussed below.
Additional similar changes are found in References 8.5, 8.6, and 8.10 for Arkansas Nuclear One, Unit 2, Waterford, and San Onofre Nuclear Generating Station, Units 2 and 3, respectively.
8.0 REFERENCES
8.1.
Letter from NRC (L. R. Wharton) to OPPD (S. K. Gambhir), dated March 14, 2001 (NRC-01-0019) [FCS Technical Specification Amendment No. 196].
8.2.
Letter B18185 from Northeast Nuclear Energy Company (R. P. Necci) to NRC (Document Control Desk), Millstone Nuclear Power Station, Unit No. 2; License Amendment Request-Unreviewed Safety Question; Proposed Revision to Final Safety Analysis Report; Fuel Centerline Melt Peak Linear Heat Rate Limit (PLAR-2-00-2), dated July 31, 2000.
8.3.
Letter B18294 from Northeast Nuclear Energy Company (R. P. Necci) to NRC (Document Control Desk), Millstone Nuclear Power Station, Unit No. 2; Response to a Request for Additional Information Regarding Proposed Revision to Final Safety Analysis Report; Fuel Centerline Melt Peak Linear Heat Rate Limit (PLAR-2-00-2), dated July 31, 2000.
8.4.
Letter from NRC (D. S. Collins) to Northeast Nuclear Energy Company (R.G. Lizotte),
Millstone Nuclear Power Station, Unit No. 2 - Issuance of Amendment Re: Fuel Centerline Melt Linear Heat Rate Limit (TAC NO. MA9646), dated March 29, 2001
[Technical Specification Amendment No. 255].
8.5.
Letter from NRC (T. W. Alexion) to Entergy Operations ANO (C. W. Anderson),
Arkansas Nuclear One, Unit 2 - Issuance of Amendment Re: Peak Fuel Centerline Temperature (TAC MB3935), dated March 4, 2002 [Technical Specification Amendment No. 238] (ADAMS Accession No. ML020730211).
8.6.
Letter from NRC (N. Kalyanam) to (J. E. Venable) Entergy Operations, Inc., Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re: Revision to Peak Linear Heat Rate Safety Limit (TAC MB3926), dated March 5, 2002 [Technical Specification Amendment No. 181] (ADAMS Accession No. ML020640587).
8.7.
Letter from NRC (W. D. Beckner) to NEI (A. Pietrangelo), dated December 23, 2002 (ADAMS Accession No. ML023570417).
8.8.
Letter from NRC (W. D. Beckner) to NEI (A. Pietrangelo), dated March 18, 2003 (ADAMS Accession No. ML030770832).
8.9 TSTF-445-A, Rev. 1, Revision to Peak Linear Heat Rate Safety Limit, August 4, 2003.
8.10. Letter from NRC (B. M. Pham) to SCE (H. B. Ray), San Onofre Nuclear Generating Station, Units 2 And 3 - Issuance of Amendments on Peak Fuel Centerline Temperature Safety Limit (TAC NOS. MC0801 and MC0802), dated June 10, 2004 [Technical Specification Amendment Nos. 192 and 183] (ADAMS Accession No. ML041680581).
8.11. EMF-1961(P)(A), Revision 0, Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors, July 2000.
8.12. XN-NF-79-56(P)(A), Revision 1, Supplement 1, Gadolinia Fuel Properties of LWR Fuel Safety Evaluation, November 1981.
8.13. XN-NF-85-92(P)(A), Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results, November 1986.
8.14. XN-NF-82-06(P)(A), Revision 1, Supplements 2, 4 and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, October 1986.
Principal Contributor: A. Wang Date: February 4, 2008