NRC-05-0011, Proposed License Amendment Request to Revise the Reactor Coolant System Pressure and Temperature Limit Curves in Technical Specification 3.4.10
| ML050830480 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 03/17/2005 |
| From: | O'Connor W DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-05-0011 | |
| Download: ML050830480 (45) | |
Text
W\\illiam T. O'Connor, Jr.
Vice Presideint, Nuclear Generation Fermi 2 6400 North Dixie 1Hy., Nlewort, Michigan 48166 Tel: 734-586-5201 Fax: 734-5864172 DTE Energy-Proprietary Information Enclosed
,=;,7 10 CFR 50.90 March 17, 2005 NRC-05-0011 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001
Reference:
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
Subject:
Proposed License Amendment Request to Revise the Reactor Coolant System Pressure and Temperature Limit Curves in Technical Specification 3.4.10 Pursuant to 10 CFR 50.90, Detroit Edison hereby proposes to amend the Fermi 2 Plant Operating License, Appendix A, Technical Specifications (TS) to revise TS 3.4.10, "Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits."
Specifically, this proposed amendment would replace the P/T curves for Hydrostatic Pressure Test, Non-Nuclear Heatup and Cooldown, and Nuclear (Core Critical)
Limits illustrated in TS Figure 3.4.10-1 with six recalculated separate curves for 24 and 32 Effective Full Power Years (EFPY) of reactor operation.
The new P/T curves are based on the 1998 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code including the 2000 Addenda. This edition of the Code has been approved for use in both 10 CFR 50.55a and Regulatory Guide (RG) 1.147. Additionally, the new curves are based on updated Reactor Pressure Vessel (RPV) neutron fluence calculations performed in accordance with an NRC-approved methodology consistent with RG 1.190.
The NRC has previously approved similar license amendment requests for the Clinton Power Station (ADAMS Accession Number ML003765368), River Bend Station (ML012280403), Perry Nuclear Power Plant (ML030700189), Browns Ferry Nuclear Plant, Units 2 and 3 (ML040480013) and Hope Creek Generating Station (ML042050079).
ApI
USNRC NRC-05-0011 Page 2 provides an evaluation of the proposed license amendment, including an analysis of the issue of significant hazards consideration using the standards of 10 CFR 50.92. Detroit Edison has concluded that the change proposed in this submittal does not result in a significant hazards consideration. Enclosure 2 provides marked up pages of the existing TS to show the proposed changes. Enclosure 3 provides a clean version of the affected TS pages with the proposed changes incorporated. provides a copy of marked up TS Bases pages affected by this change for information only.
The General Electric (GE) Report Number NEDC-33133P provided in Enclosure 5 contains detailed information regarding the development of the proposed P/T curves.
Some of the information contained in this report is considered GE proprietary and should be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4) and 10 CFR 2.390(a)(4). An affidavit attesting to this fact is provided in Enclosure 6. A non-proprietary version of the GE Report is provided in Enclosure 7.
Detroit Edison has reviewed the proposed change against the criteria of 10 CFR 51.22 and has concluded that it meets the criteria provided in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement or an Environmental Assessment.
Detroit Edison requests NRC approval of this proposed license amendment by January 3, 2006, with an implementation period of within 60 days following NRC approval. The requested approval date is based on the plan for applying the revised PIT curves during the upcoming eleventh refueling outage scheduled to start in March 2006.
No commitments are being made in this letter.
Should you have any questions or require additional information, please contact Mr.
Norman K. Peterson of my staff at (734) 586-4258.
Sincerely,
USNRC NRC-05-0011 Page 3 List of
Enclosures:
- 1) Evaluation of the Proposed License Amendment Request
- 2) Marked-Up TS Pages
- 3) Clean TS Pages
- 4) Marked-Up TS Bases Pages
- 5) GE Report NEDC-33133P, Revision 0 (Proprietary)
- 6) Affidavit in Accordance with 10 CFR 9.17(a)(4) & 10 CFR 2.390(a)(4)
- 7) Non-Proprietary Version of GE Report, NEDC-33133 cc:
E. R. Duncan (w/o Enclosure 5)
N.K.Ray NRC Resident Office (w/o Enclosure 5)
Regional Administrator, Region III (w/o Enclosure 5)
Supervisor, Electric Operators, Michigan Public Service Commission (w/o Enclosure 5)
USNRC NRC-05-001 1 Page 4 I, WILLIAM T. O'CONNOR, JR., do hereby affirm that the foregoing statements are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.
WILLI JR.
Vice President - Nuclear Generation On this 1 7 day of __M a__
, 2005 before me personally appeared William T. O'Connor, Jr., being first duly sworn and says that he executed the foregoing as his free act and deed.
Notary Public KAREN M. REED V!otaryPubfic, Monroe Couaty, Ml My Commission Explres 09/022005
ENCLOSURE 1 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT-SYSTEM PRESSURE & TEMPERATURE LIMITS EVALUATION OF THE PROPOSED LICENSE AMENDMENT REQUEST to NRC-05-0011 Page 1 Evaluation of the Proposed License Amendment Request
Subject:
Revision of the Pressure and Temperature Limit Curves 1.0 Description Detroit Edison is requesting NRC approval of this proposed revision to the Fermi 2 Technical Specification (TS) Number 3.4.10, "Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits," to replace the P/T curves for Hydrostatic Pressure Test, Non-Nuclear Heatup and Cooldown, and Nuclear (Core Critical) Limits illustrated in TS combination Figure 3.4.10-1 with six recalculated separate curves. The proposed figures provide separate composite curves for each one of the three operational categories, with the Reactor Pressure Vessel (RPV) Bottom Head limit curve also provided on the Hydrostatic Pressure Test and Non-Nuclear Heatup and Cooldown figures. Additionally, two sets of curves have been calculated, the first one assumes reactor cumulative operating time not exceeding 24 Effective Full Power Years (EFPY) and the second set assumes operating time not exceeding 32 EFPY.
The new P/T curves have been calculated based on the 1998 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, including the 2000 Addenda. For the RPV beltline region, the effect of irradiation embrittlement has been incorporated in the evaluation based on updated neutron fluence calculations. The fluence calculations are performed in accordance with General Electric (GE) licensed methodology complying with Regulatory Guide (RG) 1.190 (Reference 1).
2.0 Proposed Change The proposed change replaces TS Figure 3.4.10-1 with six Figures, 3.4.10-1 through 3.4.10-6.
The current TS Figure 3.4.10-1 is a combination figure with three P/T limit curves shown on the same figure. The current curves provide composite limits applicable to all RPV regions for three categories of operations: Hydrostatic Pressure Test [Curve A], Non-Nuclear Heatup and Cooldown [Curve B], and Nuclear (Core Critical) Operation [Curve C]. On the same figure, three other curves labeled A', B' and C' delineate limits for the RPV core beltline region. The effects of irradiation embrittlement are factored into the beltline limit curves. A note on the figure indicates that all curves are valid for 32 EFPY of operation and that the beltline curves are not limiting. Limiting Condition of Operation (LCO) 3.4.10 requires maintaining RCS pressure, RCS temperature and RCS heatup and cooldown rates within limits at all times. Two surveillances (SR 3.4.10.1 and 3.4.10.2) also include timed verification requirements of operating within the P/T curve limits.
The revised fluence calculations (Reference 2) used in the development of the new P/T curves are based on a GE licensed methodology complying with RG 1.190. In Reference 3, the NRC staff found Detroit Edison's commitment to update neutron fluence calculations for Fermi 2 RPV utilizing NRC-approved methodologies consistent with the guidance in RG 1.190 by December 31, 2005 acceptable. The NRC staff stated that the current RPV fluence calculations were to NRC-05-0011 Page 2 expected to remain conservative with respect to the actual accumulated RPV neutron fluence through the December 31, 2005 operating date. The updated fluence calculations used in the development of the revised P/T curves were performed in support of Extended Power Uprate (EPU) evaluation. Fermi 2 RPV flux for the first ten cycles of operation was determined utilizing a representative cycle with a thermal power limit of 3430 Megawatt Thermal (MWt).
The total energy generated by the Fermi 2 reactor for cycles 1 through 10 (through November 2004) is 15,089 Gigawatt-Days (GWd) which is equivalent to 12.04 EFPY at a 3430 MWt licensed thermal power. The fluence calculations conservatively assumed the reactor would operate at 3952 MWt (the planned EPU power) for the remaining cycles until the end of the 40-year License in 2025. It is assumed that the total energy generated at that time would be equivalent to 32 EFPY. This is conservative because the implementation of EPU at Fermi 2 has been delayed and the current licensed power is still limited to 3430 MWt.
The proposed change would replace the combination figure with separate P/T limit figures for each one of the three categories of operation. The new curves also provide composite limits for all RPV regions including core beltline region. RPV bottom head individual limit curves are superimposed on curves A and B. In addition, two sets of curves are calculated; one for 32 EFPY which represents the end of the current 40-year plant license and the other one is for 24 EFPY which has been selected as an intermediate point between the current EFPY and 32 EFPY.
3.0
Background
10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50.
Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.
10 CFR 50, Appendix G requires the P/T limits for an operating plant to be at least as conservative as those that would be generated if the methods of Appendix G to Section XI of the ASME B&PV Code were applied. Appendix G to Section XI of the 1998 Edition of the Code including the 2000 Addenda has been endorsed in 10 CFR 50.55a. This edition incorporates the provisions of ASME Code Cases N-588 and N-640. The methodology of Appendix G to Section XI of the Code postulates the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum applied stress. For materials in the beltline and upper and lower head regions of the RPV, the maximum flaw size is postulated to have a depth that is equal to one-fourth of the thickness and a length equal to 1.5 times the thickness. The basic parameter of the linear elastic fracture mechanics methodology of Appendix G to Section XI of the ASME B&PV Code is the stress intensity factor K1, which is a function of the stress state and flaw configuration. The methodology requires a safety factor of 2.0 on stress intensities resulting to NRC-05-0011 Page 3 from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 for hydrostatic testing curves.
The methodology in Appendix G to the ASME B&PV Code also requires that the reference temperature and Upper Shelf Energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel. The Adjusted Reference Temperature (ART or adjusted RTNDT) is defined as the sum of the initial (unirradiated) nil ductility transition reference temperature (initial RTNDT), the mean value of the adjustment in reference temperature caused by irradiation (ARTNDT), and a margin (M) term.
The adjusted RTNDT is a product of a chemistry factor and a fluence factor. The chemistry factor is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2 (Reference 4), or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Revision 2, or surveillance data.
The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the fluence, and the calculational procedures. RG 1.99, Revision 2, describes the methodology to be used in calculating the margin term.
The current Fermi 2 PIT curves were approved in License Amendment No. 87 (Reference 5).
This Amendment revised previously approved P/T curves (in Amendment No. 77) to account for the effects of power uprate from 3293 to 3430 MWt. Fermi 2 licensed thermal power was 3293 MWt for the first three cycles of operations and 3430 MWt thereafter. The P/T curves approved in License Amendment No. 77 (Reference 6) were developed in accordance with RG 1.99, Revision 2, in response to NRC Generic Letter (GL) 88-11 (Reference 7). In the Safety Evaluation for Amendment No. 77, the NRC staff concluded that the proposed curves are valid for 32 EFPY because they conform to the requirements of Appendices G and H of 10 CFR 50 and satisfy the requirements in GL 88-11.
In License Amendment No. 152 (Reference 3), the NRC approved a change in the RPV material surveillance program to incorporate the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) into the Fermi 2 licensing basis. In this Amendment, the NRC acknowledged that the current Fermi 2 RPV fluence calculations were expected to remain conservative with respect to the actual accumulated RPV neutron fluence through December 31, 2005. The Amendment Safety Evaluation concluded that Detroit Edison's commitment in Reference 8 to update neutron fluence calculations for the Fermi 2 RPV utilizing NRC-approved methodologies consistent with the guidance in RG 1.190 by December 31, 2005 was acceptable.
The updated fluence calculations used in the development of the new P/T curves proposed in this license amendment request fulfil the commitment made in Reference 8. The fluence calculations to NRC-05-0011 Page 4 (Reference 2) are performed in accordance with GE's licensed methodology (Reference 9) complying with RG 1.190.
4.0 Technical Analysis Revised pressure and temperature curves were developed for hydrostatic pressure test, core not critical, and core critical conditions. A report describing the inputs, methodology and results for the revised curves is provided in Enclosure 5. Two sets of curves were developed; the first set is applicable for 24 EFPY representing an intermediate point between the current EFPY and end of license; the second set is applicable for 32 EFPY representing the end of the current 40-year plant license.
The new P/T curves were developed using the 1998 Edition through the 2000 Addenda of the ASME B&PV Code,Section XI, Appendix G and 10 CFR 50, Appendix C The 1998 edition of the Code incorporates ASME Code Cases N-588 and N-640.
ASME Code Case N-640 permits application of the lower bound static initiation fracture toughness value equation (KIC equation) as the basis for establishing the P/T curves in lieu of the lower bound crack arrest fracture toughness value equation (KIA equation). The KIA equation is based on conditions needed to arrest a dynamically propagating crack and is the method invoked by Appendix G to Section XI of the 1989 ASME Code. Use of the Kic equation in determining the lower bound fracture toughness in the development of the P/T operating limit curves is more accurate than the KLA equation because the rate of loading during a heatup or cooldown is slow and is more representative of a static condition than a dynamic condition. RG 1.147, Revision 13 (Reference 10) states that Code Case N-640 is acceptable to the NRC for application in licensees'Section XI inservice inspection programs.
ASME Code Case N-588 permits the postulation of a circumferentially oriented flaw (in lieu of an axially-oriented flaw) for the evaluation of the circumferential welds in RPV P/T limit curves.
RG 1.147 also states that Code Case N-588 is acceptable to the NRC for application in licensees'Section XI inservice inspection programs.
Neutron Fluence Calculations Neutron irradiation of the RPV by fast neutrons (with energies greater than 1 MeV) causes a reduction in material ductility and creates structural embrittlement at higher operating temperatures. The Fermi 2 RPV neutron fluence calculations were updated using the NRC-approved General Electric proprietary methodology documented in GE's Licensing Topical Report NEDC-32983P-A (Reference 9). In Reference 11, the NRC accepted the GE methodology described in NEDC-32983P-A. GE's methodology is consistent with the guidance in RG 1.190 for neutron flux calculations and is based on a two-dimensional discrete ordinates computer code.
to NRC-05-0011 Page 5 The updated fluence calculations are based on the actual energy generated in the first ten cycles of operation (equivalent to 12.04 EFPY) at a thermal power of 3430 MWt and at 3952 MWt, thereafter. Fluence was calculated at two projected energy levels; 24 EFPY and 32 EFPY. The calculated fluence at 32 EFPY bounds operation through the end of the 40-year operating license.
The calculated peak fast neutron flux at the RPV inside surface is provided below:
Thermal Power Flux (n/cm 2-s) 3430 MWt 8.73E8 3952 MWt 1.01E9 The calculated peak fast neutron fluence for 24 and 32 EFPY are provided below:
Parameter 24 EFPY Fluence 32 EFPY Fluence (n/cm2)
(n/cm2)
RPV ID Peak Fluence 7.13E17 9.68E17 Peak 1/4 T Fluence 4.94E17 6.70E17 Initial Reference Temperature The data and methodology used to determine the initial nil ductility transition reference temperature (RTNDT) for Fermi 2 RPV material are detailed in Section 4.1 of Enclosure 5. The initial RTNDT is the reference temperature for the unirradiated material. The original Certified Material Test Reports (CMTRs) for Fermi 2 were used to obtain the initial reference temperature values. The initial RTNDT for all materials remain unchanged from values previously reported with one exception for the lower-intermediate shell plate heat C4564-1. The initial RTNDT of -12 degrees Fahrenheit previously reported for this heat has been recalculated and determined to be -
10 degrees Fahrenheit.
Adiusted Reference Temperature for Beltline Material The adjusted reference temperature calculations, methodology and results for 24 and 32 EFPY are included in Section 4.2 of Enclosure 5. Adjusted reference temperature is the reference temperature of the beltline material after including irradiation shift and a margin term.
The value of ART is a function of RPV 1/4 T fluence and material chemistry. Regulatory Guide 1.99, Revision 2, provides acceptable methods for calculating ARTs based on fluence values at 1/4 T depth. The peak inside diameter RPV surface fluence and 1/4 T fluence are tabulated above for both the 24 EFPY and 32 EFPY generated energy levels.
Detroit Edison participates in the BWRVIP Integrated Surveillance Program and complies with the requirements of Appendix H to 10 CFR 50. The ISP data applicable to Fermi 2 RPV beltline materials has been utilized in the evaluation and development of the new P/T limit curves.
to NRC-05-001 1 Page 6 Evaluation of Upper Shelf Energy The calculations for the reduction in Upper Shelf Energy (USE) of the beltline materials are included in Appendix F of Enclosure 5. Based on the results of these calculations, the USE values for Fermi 2 reactor vessel beltline materials remain within the limits of RG 1.99, Revision 2, and 10 CFR 50, Appendix Q for up to 32 EFPY of operation. Furthermore, the USE values in the transverse direction for beltline base metals and along the beltline welds remain above 50 ft-lb, as required by 10 CFR 50, Appendix G.
5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.92, Detroit Edison has made a determination that the proposed amendment involves no significant hazards consideration. The proposed revision to the Reactor Coolant System (RCS) Pressure and Temperature (PIT) limit curves in Technical Specification (TS) 3.4.10 does not involve a significant hazards consideration for the following reasons:
- 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The revised P/T curves are based on the 1998 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, including the 2000 Addenda. This edition of the Code has been approved for use in both 10 CFR 50.55a and Regulatory Guide (RG) 1.147. The revised curves are also based on updated fluence calculations performed utilizing NRC-approved methodology consistent with RG 1.190 for calculating Reactor Pressure Vessel (RPV) neutron fluence. Revised fluence calculations are applicable for 24 and for 32 Effective Full Power Years (EFPY).
The 32 EFPY represents a conservative exposure level at the end of the current 40-year plant operating license. The proposed change incorporates adjustment of the reference temperature for all beltline material to account for irradiation effects and provide a comparable level of protection as previously evaluated and approved. The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR 50 Appendix G using the guidance contained in RG 1.99, Revision 2, to provide operating limits for up to 32 EFPY.
There are no changes being made to the RCS pressure boundary or to RCS material, design or construction standards. The proposed P/T curves define limits that continue to ensure the prevention of nonductile failure of the RCS pressure boundary. The revision of the P/T curves does not alter any assumptions previously made in the radiological consequence evaluations since the integrity of the RCS pressure boundary is unaffected. Therefore, the proposed changes will not significantly increase the probability or consequences of an accident previously evaluated.
to NRC-05-001 1 Page 7
- 2.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The revised PIT curves are based on a later edition and addenda of the ASME Code that incorporates current industry standards for the curves. The revised curves are also based on an RPV fluence that has been recalculated in accordance with the methodology of RG 1.190. The proposed change does not involve a modification to plant structures, systems or components. There is no effect on the function of any plant system, and no newly introduced system interactions. The proposed change does not create new failure modes or cause any systems, structures or components to be operated beyond their design bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3.
The proposed change does not involve a significant reduction in the margin of safety.
The proposed P/T curves define the limits of operation to prevent nonductile failure of the RPV upper vessel, bottom head and beltline region. The new curves conform to the guidance contained in RG 1. 190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," and RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," and maintain the safety margins specified in 10 CFR 50 Appendix G. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Detroit Edison has determined that the proposed license amendment does not involve a significant hazards consideration.
5.2 Applicable Regulatory Requirements 10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50.
Appendix G is the regulatory basis for Pressure and Temperature (P/T) curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Upper Shelf Energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel. Regulatory Guide (RG) 1.190 (Reference 1) contains the NRC staff guidance on determining neutron fluence. RG 1.99 (Reference 4) describes general procedures acceptable to the NRC for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.
I I
I to NRC-05-0011 Page 8 Based on the considerations discussed above, it is concluded that, (1) there is a reasonable assurance that the health and safety of the public will not be endangered by operating in the proposed manner, (2) activities will be conducted in compliance with NRC regulations, and (3) the approval and issuance of this proposed amendment will not be inimical to the common defense and security of the health and safety of the public.
6.0 Environmental Considerations Detroit Edison has reviewed the proposed change against the criteria of 10 CFR 51.22 for environmental considerations. The proposed change does not involve a significant hazards consideration, nor does it significantly change the types or significantly increase the amounts of effluents that may be released offsite. The proposed change does not significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, Detroit Edison concludes that the proposed change meets the criteria provided in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement or an Environmental Assessment.
7.0 References
- 1)
Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Revision 0, March 2001
- 2)
GE Nuclear Energy Report, "DTE Energy Fermi-2 Energy Center, Neutron Flux Evaluation," GE-NE-0000-0031-6254, Revision 1, February 2005 (Proprietary)
- 3)
NRC Letter to Detroit Edison (Amendment No. 152), "Fermi 2 - Issuance of Amendment Re: Implementation of the BWRVIP RPV ISP to Address the Requirements of Appendix H to 10 CFR 50 (TAC No. MB5840)," dated January 30, 2003
- 4)
Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"
Revision 2, May 1988
- 5)
NRC Letter to Detroit Edison, "Fermi Amendment No. 87 to Facility Operating License No. NPF-43 (TAC No. M82102) - Power Uprate from 3293 to 3430 MWt,"
dated September 9, 1992
- 6)
NRC Letter to Detroit Edison, "Fermi Amendment No. 77 to Facility Operating License No. NPF43 (TAC No. 81232) - Revision of P/T Curves in Accordance with RG 1.99, Revision 2," dated December 27, 1991
- 7)
NRC Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations," dated July 12, 1988 to NRC-05-0011 Page 9
- 8)
Detroit Edison Letter to NRC, "Response to NRC Request for Additional Information Regarding the Proposed License Amendment for Participation in the BWRVIP ISP Plan for RPV Material Surveillance," NRC-02-0069, dated October 23,2002
- 9)
GE Report, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation," NEDC-32983P-A, December 2001
- 10) Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 13, January 2004
- 11) Letter from NRC to GE, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No.
MA9891)," MFN 01-050, dated September 14, 2001
ENCLOSURE 2 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT SYSTEM PRESSURE & TEMPERATURE LIMITS MARKED-UP TS PAGES Affected Pages:
3.4-24 3.4-25 3.4-27 3.4-28 (to be replaced with 6 pages)
RCS P/T Limits 3.4.10 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. --------- NOTE-........
C.1 Initiate action to Immediately.
Required Action C.2 restore parameter(s) shall be completed if to within limits.
this Condition is entered.
AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LCO not met in other operation.
or 3 than MODES 1. 2.
and 3.
SURVEILLANCE-REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 NOTE--------------------
Only required to be performed as applicable during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
Verify:r
- b. RCS heatup and cooldown rates are limited to:
- 1. < 1000F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period: and
- 2. < 200F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
30 minutes I ___________________________
(continued)
FERMI - UNIT 2 3.4-24 Amendment No. 134
RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.10.2 Verify RCS pressure and RCS temperature are Ohce within within the criticality limits specified in 15 minutes prior to control rod withdrawal for 3-othe purpose of r'fwfwes 3.4.io-3 or 3.A.1o-(0, as 4pR;c4-t) achieving criticality SR 3.4.10.3 ------------
NOTE-------------------
Only required to be met in MODES 1. 2. 3.
and 4 during recirculation pump startup.
Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) steam space coolant prior to each temperature is 5 1450F.
startup of a recirculation pump SR 3.4.10.4 NOTE--------------------
Only required to be met in MODES 1. 2. 3.
and 4 during recirculation pump startup.
Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is 5 500F.
startup of a recirculation pump (continued)
FERMI - UNIT 2 3.4-25 Amendment No. 134
RCS P/T Limits 3.4.10 ppVpI i
^Fn RpTITM;TM-Mr frnntininri)
- ~
flA.
- t~&..
8~~IU *~
- =--
1* I
- LJ'.t U.
SURVEILLANCE FREQUENCY SR 3.4.10.7 NOTE-Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor-vessel flange @
head flange temperatures are L -e F when the reactor vessel head bolt studs are under tension.
30 minutes SR 3.4.10.8 NOTE-------------------
Not required to be performed until 30 minutes after RCS temperature < 800F in MODE 4.
Verify reactor vessel flange and head 30 minutes flange temperatures are F
V7_.
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 1000F in MODE 4.
Verify reactor vessel flan and head 12 hDurs flange temperatures are F
I FERMI - UNIT 2 3.4-27 Amendment No. 134
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TCALF 714F2 EFPY CURVES A', L. C-NOT LM,47hG* ?FORWMATlON OfLY
_____CL.RVES A. B. C ARE VALb FOR 32 EFPY OF OPSRATION
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-F I GURE 3.-4 D-1I IIMUM REACTOR PRESSURE VESSEL METIAL-TEMPERATURE'VS. REACTOR VESSEL PRESSURE FERMI - -UNIT 2
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1400 1300 1200 1100 ln C 1000
- 0.
900 0
I--J U
800 cz:o 700 as E
600 500 lii UJ 0:
g?
400 300 INITIAL RTndt VALUES ARE F FOR BELTLINE, 25'F FOR UPPER VESSEL, AND 30'F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (OF) 24 119 HEATUPICOOLDOWN RATE OF COOLANT
< 20FIHR 200 100 UPPER VESSEL AND BELTUNE LIMITS BOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE
('F) 200 Figure 3.4.10-1 System Pressure Test P/T Curve [Curve A] for up to 24 EFPY
1400 1300 1200 1100 I-en
] 1000
'U o
700 it
'U a:
600 2
j 500
'U 8
400
'U 0.
300 200 100 INITIAL RTndt VALUES ARE
-44'F FOR BELTUNE.
25F FOR UPPER VESSEL, AND 44.6-F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (F) 24 106 HEATUPICOOLDOWN RATE OF COOLANT 1 I0OF/HR UPPER VESSEL AND BELTUNE UMITS
...... BOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE
('F) 200 Figure 3.4.10-2 Core Not Critical P/T Curve [Curve B] for up to 24 EFPY
1400 1300 1200 1100 la 0
cL 0
1000 L
00 2 800 o
700 I.-
V tuw 6400 Z
I.-
=
500 LU 30i 12400 w
ix 300 I
-MII-I I I 740 I
Mirimurn Cribcality_
_00
-1 _ITemperature 72-F A'[-.
I INITIAL RTndt VALUES ARE
-441F FOR BELTLINE.
25'F FOR UPPER VESSEL.
AND 30'F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT ('F) 24 106 HEATUP/COOLDOWN RATE OF COOLANT
, I00FIHR 200 100 0
-BELTLINE AND NON-BELTLINE LIMI TS 0
25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.10-3 Core Critical P/T Curve [Curve C] for up to 24 EFPY
1400 1300 1200 1100 L
900 I.-
-U 8
800 o
700 F-t 600 Z
S-J 500
'U b
400 a:
300 200 100 0
~ f l
-I I
I I
I I
I 4-II 1-I I
IA I
IE I O
-III 7 2
-M -I
-I INITIAL RTndt VALUES ARE F FOR BELTUNE.
25-F FOR UPPER VESSEL, AND 30'F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (F) 32 126 HEATUP/COOLDOWN RATE OF COOLANT
' 20FIHR
-UPPER VESSEL AND BELTUNE UMITS
...... BOTTOM HEAD CURVE S -
a
S -
i -
S -
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(.F)
Figure 3.4.104 System Pressure Test P/T Curve [Curve A] for up to 32 EFPY
1400 1300 1200 1100 5P2 a
1000 LU IL 900 0
-j'U gg 00
'U a:
o 700 t
'U 2 600 2
400
'U 300 200 100 0
INITIAL RTndt VALUES ARE
.44-F FOR BELTUNE.
25'F FOR UPPER VESSEL, AND 44.6-F FOR BOTTOM HEAD BELTUNE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (-F) 32 121 HEATUP/COOLDOWN RATE OF COOLANT I 100IF/HR
-LUPPER VESSEL AND BELTUNE UMITS
...... BOTTOM HEAD CURVE 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
(.F)
Figure 3.4.10-5 Core Not Critical P/T Curve [Curve B] for up to 32 EFPY
1400-INITIAL RTndt VALUES ARE 1300
- F FOR BELTLINE, 25F FOR UPPER VESSEL.
1200-
-AND 30F FOR BOTTOM HEAD 1100 1000 -
lBELTLINECURVE 000 7ADJUSTED AS SHOWN:
EFPY SHIFT ('F)
-9w 32 121 So 8_HEATUPICOOLDOWN RATE OF COOLANT
< 100F/HR o
700-
= 600-.
zt-3 00 M_
r LU 400 nU 300
200 20/
-BELTLINE AND
_NON-BELTLINE 100 Minimum Citca LIMITS Temperature 72F 0
_I 0
25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Figure 3.4.10-6 Core Critical PtIT Curve [Curve C] for up to 32 EFPY
ENCLOSURE 3 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT SYSTEM PRESSURE & TEMPERATURE LIMITS CLEAN TS PAGES New Pages:
3.4-24 3.4-25 3.4-27 3.4-28 3.4-28a 3.4-28b 3.4-28c 3.4-28d 3.4-28e
RCS P/T Limits 3.4.10 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. --------- NOTE---------
C.1 Initiate action to Immediately Required Action C.2 restore parameter(s) shall be completed if to within limits.
this Condition is entered.
AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2
- LCO not met in other operation.
or 3 than MODES 1, 2, and 3.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1
NOTE--------------------
Only required to be performed as applicable during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.
Verify:
30 minutes
- a. RCS pressure and RCS temperature are to the right of the applicable limits specified in Figures 3.4.10-1 through 3.4.10-6; and
- b. RCS heatup and cooldown rates are limited to:
- 1.
- 1001F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period; and
- 2.
- 201F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves..
(continued)
FERMI - UNIT 2 3.4-24 Amendment No. Ad
RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.10.2 Verify RCS pressure and RCS temperature are Once within within the criticality limits specified in 15 minutes Figures 3.4.10-3 or 3.4.10-6, as prior to applicable.
control rod withdrawal for the purpose of achieving criticality SR 3.4.10.3
NOTE-------------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
Verify the difference between the bottom Once within head coolant temperature and the reactor 15 minutes pressure vessel (RPV) steam space coolant prior to each temperature is
- 1450F.
startup of a recirculation pump SR 3.4.10.4
NOTE--------------------
Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.
Verify the difference between the reactor Once within coolant temperature in the recirculation 15 minutes loop to be started and the RPV coolant prior to each temperature is < 500F.
startup of a recirculation pump (continued)
FERMI - UNIT 2 3.4-25 Amendment No. A,1$
RCS P/T Limits 3.4.10 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.10.7 NOTE--------------------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor vessel flange and head flange temperatures are 2 720F when the reactor yessel head bolt studs are under tension.
30 minutes SR 3.4.10.8
NOTE.-------------------
Not required to be performed until 30 minutes after RCS temperature
- 800F in MODE 4.
Verify reactor vessel flange and head 30 minutes flange temperatures are 2 721F.
NOTE-------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature
- 1001F in MODE 4.
Verify reactor vessel flange and head 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> flange temperatures are 2 720F.
l FERMI - UNIT 2 3.4-27 Amendment No. a X
1400 1300 1200 Is 1100 an 0
< 1000 w
C.
° 900
-j uL 800 Lu 0I-700 0
Lu z
600 X 500 ui it u, 400 Lu 300 200 100 I
INITIAL RTndt VALUES ARE
-440F FOR BELTLINE, 25¶F FOR UPPER VESSEL, AND 30¶F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (fF) 24 119 HEATUP/COOLDOWN RATE OF COOLANT
< 20FIHR UPPER VESSEL AND BELTLINE LIMITS
BOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE
( 0F)
SYSTEM PRESSURE TEST FIGURE 3.4.10-1 P/T CURVE (CURVE A) FOR UP TO 24 EFPY FERMI - UNIT 2 3.4-28 Amendment No. OX,
RCS P/T Limits 3.4.10 1400 1300 1200 Is 1100 U) 0.
0.S 1000 w
CLo 900
-jw U) us 800 W
0-700 us tL 300 2 600 1
500 w
U)
'm 400 w
a-300 200 100 INITIAL RTndt VALUES ARE
-44°F FOR BELTLINE, 25¶F FOR UPPER VESSEL, AND 44.6°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:
EFPY SHIFT (0F) 24 106 HEATUP/COOLDOWN RATE OF COOLANT
< 100F/HR UPPER VESSEL AND BELTLINE LIMITS
BOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
FIGURE 3.4.10-2 CORE NOT CRITICAL P/T CURVE (CURVE B) FOR UP TO 24 EFPY FERMI - UNIT 2 3.4-28a Amendment No. A,4,
RCS P/T Limits 3.4.10 1400 1300 1200 la 1100 In
< 1000 Lu a-o 900
-j U)
LO 800 u
0u7 700 U
Lu Z
600
=i 500 U,co 400 CL 300 INITIAL RTndt VALUES ARE
-440F FOR BELTLINE, 25°F FOR UPPER
- VESSEL, AND 30'F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 24 106 HEATUP/COOLDOWN RATE OF COOLANT
< 1001F/HR 200 100 BELTLINE AND NON-BELTLINE LIMITS 0
0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
( 0F)
FIGURE 3.4.10-3 CORE CRITICAL P/T CURVE (CURVE C) FOR UP TO 24 EFPY FERMI - UNIT 2 3.4-28b Amendment No. OaX,
RCS P/T Limits 3.4.10 1400 1300 1200 i, 1100
.0 0
< 1000 muw o 900 IL
-j cu 800 w
0° 700 w
z 600 I--
X 500 w
cn, 400 w
c0 300 200 100 INITIAL RTndt VALUES ARE
-440F FOR BELTLINE, 25°F FOR UPPER VESSEL, AND 30@F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN EFPY SHIFT (°F) 32 126 HEATUP/COOLDOWN RATE OF COOLANT
<20*F/HR UPPER VESSEL AND BELTLINE LIMITS
.... gBOTTOM HEAD CURVE 0
0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
SYSTEM PRESSURE TEST FIGURE 3.4.10-4 P/T CURVE (CURVE A) FOR UP TO 32 EFPY FERMI - UNIT 2 -
3.4-28c Amendment No. OX.
RCS P/T Limits 3.4.10 1400 1300 1200 la 1100 CL a.
<. 1000 u
(Lo 900 a)gn 800
°-
700 Lu z 600 500 a,
a, 400 U]
Lu W-300 200 100 1
0 0
25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (0F)
FIGURE 3.4.10-5 CORE NOT CRITICAL P/T CURVE (CURVE B) FOR UP TO 32 EFPY FERMI - UNIT 2 3.4-28d Amendment No. O,4
RCS P/T Limits 3.4.10 1400 1300 1200 m 1100 0.
< 1000 us ELo 900
-J CO 800
(-
700 u
z 600
-i 500 Fw cix uc 400 us a.
300 200 100 0
INITIAL RTndt VALUES ARE
-44@F FOR BELTLINE, 25¶F FOR UPPER
- VESSEL, AND 30°F FOR BOTTOM HEAD BELTLINE CURVE ADJUSTED AS SHOWN:
EFPY SHIFT (°F) 32 121 HEATUP/COOLDOWN RATE OF COOLANT
< 100°F/HR BELTLINE AND NON-BELTLINE LIMITS 0
25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE
( 0F)
FIGURE 3.4.10-6 CORE CRITICAL P/T CURVE (CURVE C) FOR UP TO 32 EFPY FERMI - UNIT 2 3.4-28e Amendment No. OXg,
ENCLOSURE 4 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT SYSTEM PRESSURE & TEMPERATURE LIMITS MARKED-UP TS BASES PAGES (For Information Only)
Affected pages:
B 3.4.10-1 B 3.4.10-2 B 3.4.10-3 B 3.4.10-6 B 3.4.10-9
RCS P/T Limits B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
containirA/T limit curves for hydrostatic or leak testfirng Curve.A); for heatup by non-nuclear means, 3..4./°-I/
cooldown following a nuclear shutdown and low power physics 9
vg4 X t.4o-tests (Curve B); and for operations with a critical core c
other than low power physics tests (Curve C). Other related P/T limits are provided in the SRs.
)
II Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational 7Z guidance during heatup or cooldown maneuvering, when i7c-,Its~ 5 cpressure and temperature indications are monitored and Af~ f c G compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the 1'
d i
c component most subject to brittle failure.
Therefore, the
_,,_l C. e.v_!v____"____*_s__Ia 24 EffAev(e' FYUz Powt' Beoor (upy),
a14 4i4 >j4x Jet-LIU limits apply mainly to-the vessel.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code,Section III, I. S --f f )'
6 w.
lor-Ip 1 q.1 Lse R" c
. I Appendix G (Ref. 2).
opt-Nic-44, o k1i 4 0-Youv 1z CP_#%4 FERMI - UNIT 2 8 3.4. 10-1I Revision O
RCS P/T Limits B 3.4.10 BASES BACKGROUND (continued)
The actual shift in the RTNDT of the vessel material will be established periodically in accordance with Appendix H of 10 CFR 50 (Ref. 3). Appendix H requires a vessel material surveillance program; as such, Fermi 2 is participating in the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP)(Ref. 4). Results of the ISP evaluation may require the operating P/T limit curves to be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.
""' B,~p Rp'v
- 6im L'tJ IctAvVa-% tre-
-I rshe
- ovinb, ke.'J C,\\yve-4deve The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.,
The criticality limits include the Reference 1 requirement that they be at least 40'F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident.
In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
ASME'Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
4,14ephiCf ryer,,f o
V FERMI UNIT 2 B 3.4.10-2 Revision 17 FERMI - UNIT 2 B 3.4.10-2 Revision 17
RCS P/T Limits B 3.4.10 BASES APPLICABLE SAFETY ANALYSES The P/T limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a
___J: __
ZL_
oh=___
S
_ _zL1 _L
-C-imits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The elements of this LCO are:
- a. RCS pressure, temperature, and heatup or cooldown rate are within limits during RCS heatup, cooldown, and inservice leak and hydrostatic testing;
- b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) steam space coolant is within limit during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel meets limit during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow with an idle recirculation loop;
- d. RCS pressure and temperature are within limits, prior to achieving criticality; criticality and
- e. The reactor vessel flange and the head flange temperatures are within limits when the reactor vessel head bolting studs are under tension.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
FM T FERMI - UNIT 2 B 3.4.10 -3 Revision 0
RCS P/T Limits B 3.4.10 BASES ACTIONS (continued)
C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 2000F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.
ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.
Condition C is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits.
Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that operation is within the limits of is required every 30 minutes when RCS
"+s 3.4.ro.-I pressure and temperature conditions are undergoing planned d
3changes. This Frequency is considered reasonable in view of
-7'o&2..A 3.4.wo, the control room indication available to monitor RCS status.
Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
FERMI - UNIT 2 B 3.4.10 - 6 Revision 0
RCS P/T Limits B 3.4.10 BASES REFERENCES
- 2. ASME. Boiler and Pressure Vessel Code.Section III.
Appendix G.
- 3. 10 CFR 50. Appendix H.
- 4. BWRVIP-86-A, October 2002
- 5. Regulatory Guide 1.99. Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
I I
FERMI - UNIT 2 B 3.4.10-9 Revision 17
ENCLOSURE 6 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT SYSTEM PRESSURE & TEMPERATURE LIMITS GE'S AFFIDAVIT IN ACCORDANCE W'ITIH 10 CFR 9.17(a)(4) & 10 CFR 2.390(a)(4)
General Electric Company AFFIDAVIT I, George B. Stramback, state as follows:
(1) I am Manager, Regulatory Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.
(2) The information sought to be withheld is contained in the GE proprietary report NEDC-33133P, Pressure-Temperature Curves for DTE Energy Fermi Unit 2, Class III (GE Proprietary Information), dated February 2005. The proprietary information is delineated by a double underline inside double square brackets. Figures and large equation objects are identified with double square brackets before and after the object.
In each case, the superscript notation 3 ) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.
(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission.
975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA.
704F2dl280 (DC Cir. 1983).
(4) Some examples of categories of information which fit into the definition of proprietary information are:
- a.
Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;
- b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
- c.
Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, resulting in potential products to General Electric; GBS-05-1-Af Ferni 2 P-T curves NEDC-33133P.doc Affidavit Page I
- d.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a., and (4)b, above.
(5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.
(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.
(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.
(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed methods and processes, which GE has developed and applied to pressure-temperature curves for the BWR over a number of years. The development of the BWR pressure-temperature curves was achieved at a significant cost, on the order of 3/4 million dollars, to GE.
The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities.
The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes GBS-05-1-Af Fermi 2 P-T curves NEDC-33133P.doc Affidavit Page 2
(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities.
The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.
The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.
The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to GE would be lost if the information were disclosed to the public.
Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.
I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.
Executed on this /e day of i
2005.
Geole B. Sramback General Electric Company GBS-05-1-AF DTE Fluence Report 0031-6254-RI.doc Affidavit Page 3