NLS2004005, License Amendment Request to Revise Technical Specification 3.4.9 Pressure Temperature (P/T) Curves Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3
| ML040340749 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/29/2004 |
| From: | Edington R Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2004005 | |
| Download: ML040340749 (96) | |
Text
Nebraska Public Power District Always there when you need us 50.90 NLS2004005 January 29, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
License Amendment Request to Revise Technical Specification 3.4.9 Pressure Temperature (P/T) Curves figures 3.4.9-1, 3.4.9-2, and 3.4.9-3.
Cooper Nuclear Station, Docket 50-298, DPR-46 The purpose of this letter is for the Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 in accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). The proposed amendment would revise TS section 3.4.9 Pressure Temperature curves 3.4.9-1, 3.4.9-2 and 3.4.9-3 for Heatup/Cooldown-Core not Critical, Pressure Test and Heatup/Cooldown-Core Critical conditions. The curves are being revised to accommodate Reactor Operation to 32 Effective Full Power Years. The curves were developed in accordance with the 1995 Edition, 1996 Addenda, American Society of Mechanical Engineers (ASME)
Boiler Pressure Vessel Code Section Xl Appendix G, 10 CFR 50 Appendix G, and ASME Section Xl Code Case N-640. Regulatory Guide 1.147, Revision 13, identifies Code Case N-640 as an NRC acceptable Section Xl code case.
NPPD requests NRC approval of the proposed TS change and issue of the requested license amendment by September 30, 2004. Once approved, the amendment will be implemented within 60 days.
Attachment I provides a description of the TS change, the basis for the amendment, the no significant hazards consideration evaluation pursuant to 10 CFR 50.91(a)(1), and the environmental impact evaluation pursuant to 10 CFR 51.22. Attachment 2 provides the proposed changes to the current CNS TS and Bases (provided for information) on marked up pages. provides the revised TS and Bases (provided for information) pages in final typed format. Enclosure I provides the calculations and methodology used by Structural Integrity Associates, Inc. to create the revised CNS Pressure Temperature curves.
This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 202 issued December 5, 2003, have been COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 17 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2004005 Page 2 of 3 incorporated into this request. NPPD has concluded that the proposed change does not involve a significant hazards consideration and that it satisfies the categorical exclusion criterion of 10 CFR 51.22(c)(9). This request is submitted under oath pursuant to 1 0 CFR 50.30(b).
By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91 (b)(l). Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b)(1).
Should you have any questions concerning this matter, please contact Mr. Paul Fleming at (402) 825-2774.
Sin.
Ael R ndall K. Edington Vice President - Nuclear and Chief Nuclear Officer
/rar Attachments and Enclosure cc:
Regional Administrator w/ attachments and enclosure USNRC - Region IV Senior Project Manager w/ attachments and enclosure USNRC - NRR Project Directorate IV-I Senior Resident Inspector %v/ attachments and enclosure USNRC Nebraska Health and Human Services w/ attachments and enclosure Department of Regulation and Licensure NPG Distribution w/o attachments and enclosure Records w/ attachments and enclosure
NLS2004005 Page 3 of 3 Affidavit STATE OF NEBRASKA )
)
NEMAHA COUNTY
)
Randall K. Edington, being first duly swom, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this correspondence on behalf of Nebraska Public Power District; and that the statements contained herein are true to the best of his knowledge and belief.
L/S/. Xf49
\\-Randall K. Edingto:)
Subscribed in my presence and sworn to before me this G.29 day of A n
, 2004.
1i/7/An I
I GENERAL NOTARY-State of Nebraska L)rt,
/2?Lvj2Az7hI WA WM.
WERNER S
lMy Commn.
Exp.
Oct 26, 206 NOTARY PUBLIC
NLS2004005 Attachment I Page I of 11 NPPD's Evaluation 1.0 Introduction 2.0 Description of Proposed Amendment 3.0
Background
4.0 Technical Analysis 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration (NSIIC) 5.2 Regulatory Requirements and Guidance 6.0 Environmental Consideration 7.0 References
NLS2004005 Attachment I Page 2 of II LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.4.9 PRESSURE TEMPERATURE (PIT) CURVES FIGURES 3.4.9-1, 3.4.9-2, AND 3.4.9-3.
Cooper Nuclear Station, NRC Docket 50-298, DPR46 Revised Pages 3.4-23 3.4-24 3.4-25 B 3.4-44 B 3.4-46 B 3.4-49 B 3.4-52 1.0 Introduction This letter is a request to amend Operating License DPR-46 for Cooper Nuclear Station (CNS).
The current Technical Specification (TS) 3.4.9 Pressure Temperature (P/T) Curves in figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 provide for plant operation through 21 Effective Full Power Years (EFPY). The proposed change will revise the P/T curves for plant operation through 32 EFPY. Section 3.4.9 of the TS define, through Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3, the pressure temperature boundaries within which the plant must operate to ensure adequate margin against brittle fracture of the Reactor Coolant System. The curves were developed in accordance with the 1995 Edition, 1996 Addenda of the American Society of Mechanical Engineers (ASME) Section Xl Appendix G, 10 CFR 50 Appendix G, and ASME Section XI Code Case N-640. The primary effect of the proposed revision is continued plant operation from 21 EFPY (expected by December of 2004) to 32 EFPY, and to allow required reactor vessel hydrostatic and leak tests to be performed at a significantly lower temperature.
2.0 Description of Proposed Amendment This proposed change will revise TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3, to allow plant operation from 21 EFPY through 32 EFPY. The proposed amendment includes a full set of updated P/T curves for pressure test, core not critical and core critical conditions. The three regions of the reactor pressure vessel that are evaluated are the beltline region, the bottom head region, and the feedwater nozzle/upper vessel region.
These regions bound all other regions with respect to brittle fracture. The P/T curve limits are developed according to Appendix G of ASME Boiler and Pressure Vessel Code, Section Xl. ASME Section XI Code Case N-640 was used in the development of the P/T curves. There are tvo lower bound fracture toughness curves available in Section
NLS2004005 Attachment I Page 3 of 11 XI: Kia, which is a lower bound on all static, dynamic and arrest fracture toughness, and Kic, which is a lower bound on static fracture toughness only. ASME Section XI Code Case N-640 changes the fracture toughness curve used for development of P/T limit curves from Kia to Kic. The P/T curves based on the Kic fracture toughness limits enhance industrial safety by expanding the P/T window low-temperature operating region. Potential benefits from this proposed TS change include reduced challenges to personnel safety as personnel can perform their duties at lower ambient temperatures, reduced dose to inspectors due to increased inspection effectiveness at the lower ambient temperatures and reduced critical path time associated with hydrostatic and leak testing during refueling outages.
The TS Bases for TS 3.4.9 will be revised as appropriate to provide supporting and clarifying information for revision of the P/T curves.
3.0
Background
The current P/T curves were developed based on the results from the first and second vessel material surveillance. The Adjusted Reference Temperature (ART) values were developed for the reactor pressure vessel materials in accordance with Regulatory Guide (RG) 1.99 Revision 2. The most limiting beltline material is the Lower Longitudinal Weld with an ART value of 127.6 degrees Fahrenheit ('F).
The P/T curves for the CNS beltline, bottom head, and upper vessel regions have been updated. Heatup/Cooldown curves were developed for 32 EFPY. The curves revised were developed in accordance with the requirements of 10 CFR 50 Appendix G. The methods in the 1995 Edition, 1996 Addenda of ASME Code,Section XI, Appendix G were used with ASME Section XI Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves." Regulatory Guide (RG) 1.147, Revision 13, identifies ASME Section XI Code Case N-640 as a Nuclear Regulatory Commission (NRC) acceptable ASME Section XI code case. ASME Section XI Code Case N-640 was also used in the development of the P/T curves approved by the NRC staff for Susquehanna Units I and 2. The new P/T curves for CNS and Susquehanna were developed by Structural Integrity Associates.
4.0 Technical Analysis CURVE A (DEVELOPMENT OF UPDATED PRESSURE TEST P/T CURVES)
Updated Safety Analysis Report, Section IV-2.6.2, "Brittle Fracture Consideration" provides the reference temperature of nil-ductility temperature (RTNDT) estimates for the CNS reactor pressure vessel (RPV) materials in accordance with RG 1.99, Revision 2.
NLS2004005 Page 4 of II Three regions were evaluated: (1) the beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. The approach used for calculating the pressure test P/T curves for each of these regions is summarized in the following subsections.
Beltline Region The temperature at the assumed flaw tip, Tl/4t (i.e., 1/4 thickness into the vessel wall), is equal to the fluid temperature, as the pressure test condition neglects any thermal effects.
The assumed temperature that is read on the plant's gauge has an uncertainty of +/- 5'F.
A temperature 5 F less than that assumed is used for calculating the allowable pressure.
This method produces a lower allowable pressure, making it conservative.
Bottom Head Region The bottom head region calculations are the same as for the beltline region, except that the equation for pressure stress in a spherical shell is substituted for the cylindrical pressure stress equation and a stress concentration factor is used to account for the Control Rod Drive (CRD) penetrations.
Feedwater Nozzle/Upper Vessel Region The feedwater nozzle is selected to represent non-beltline/upper vessel components for fracture toughness analyses because the stress conditions at this location are the most severe experienced in the vessel. In addition to the more severe pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow into hotter vessel coolant.
The methodology used for the feedwater nozzle is contained in Welding Research Council Bulletin 175, "PVRC [Pressure Vessel Research Committee] Recommendations on Toughness Requirements for Ferritic Materials."
Two additional requirements were used to define the lower portion of the upper vessel P/T curve. These limits are established by the discontinuity regions of the vessel (i.e.,
flanges), and are specified in Table 1 of 10 CFR 50, Appendix G. The requirements are:
If the calculated pressure, P, is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDT of the limiting flange material + 90'F. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 20F.
If the calculated pressure, P, is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is typically greater than or equal to the RTNDT of the limiting flange material.
NLS2004005 Attachment I Page5 of l General Electric recommends the application of an additional 60'F margin to the RTNDT value as a standard for the Boiling Water Reactor (BWR) industry for non-ductile failure protection. For the CNS flange material, this minimum would be 80'F (i.e., 20 + 60'F).
The gauge's uncertainty of +/- 5.F is not applied here since the included margin encompasses instrument uncertainty. Use of the minimum temperature of 80 degrees is consistent with previous NRC approved P/T curve TS amendments 120 and 155 as well as section 2.6.3.5 of the Updated Safety Analysis Report.
CURVE B (I]EATUP/COOLDOWNTN, CORE NOT CRITICAL)
There are three regions that were evaluated: (I) the beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. The methodology used to calculate the heatup/cooldown P/T curves for each of these regions is summarized in the following subsections:
Beltline Region The temperature at the assumed flaw tip, Tl/4t (i.e., 1/4t into the vessel wall), was conservatively assumed to be zero and the metal temperature was assumed equivalent to the fluid temperature. The CNS temperature gauge has an uncertainty of +/- 5F. A temperature 5F less than the initially assumed fluid temperature was used to calculate the allowable pressure, producing a lower allowable pressure, and therefore making the calculation conservative.
Bottom Head Region The bottom head region calculations are the same as for the beltline region, except that the equation for pressure stress in a spherical shell is substituted for the cylindrical pressure stress equation and a stress intensity factor is used to account for the CRD penetrations.
Feedwater Nozzle/Upper Vessel Region The feedwater nozzle was selected to represent non-beltline/upper vessel components for fracture toughness analyses because the stress conditions at this location are the most severe that the vessel experiences. In addition to the more severe pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences thermal cycling due to relatively cold feedwater flow into hotter vessel coolant.
Two additional requirements were used to define the lower portion of the upper vessel P/T curve. These limits are established by the discontinuity regions of the vessel (i.e.,
flanges), and are specified in Table I of 10 CFR 50, Appendix G. The requirements are:
NLS2004005 Attachment I Page 6 of II If the calculated pressure, P, is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDT of the limiting flange material + 120'F. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 20'F.
If the calculated pressure, P, is less than or equal to 20% of the pre-service hydro test pressure, the same requirements apply as were applicable for Curve A.
CURVE C (HEATUP/COOLDOWN, CORE CRITICAL)
Curve C, the core critical operation curve, is generated from the requirements of 10 CFR 50 Appendix G. Table I of Appendix G requires that core critical P/T limits be 40'F above any Curve A (Updated Pressure Test) or B (Heatup/Cooldown, Core not Critical) limits wvhen pressure exceeds 20% of the pre-service system hydrostatic test pressure.
Curve B is more limiting than Curve A, so limiting Curve C values are equal to Curve B plus 40'F for pressures above 312 psig.
Table I of Appendix G also dictates that, for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60'F) at pressures below 312 psig (20% of pre-service hydrostatic pressure test). This requirement makes the minimum criticality temperature 80F for CNS, based on an ART value of 20F for the flange region. In addition, above 312 psig, the Curve C temperature must be at least the greater of ART of the closure flange region + 160'F, or the temperature required for the hydrostatic pressure test (Curve A at 1,100 psig). Therefore, this requirement causes a temperature shift in Curve C at 312 psig.
4.1 Precedent A similar TS change for P/T curves was approved by the NRC staff for Susquehanna Units 1 and 2 (Amendment No. 200/Amendment No. 174, dated February 7, 2002). The P/T curves for Susquehanna were also developed by Structural Integrity Associates using the requirements of 10 CFR 50 Appendix G, ASME Section Xl, Appendix G, and ASME Section XI Code Case N-640. During NRC review of the Susquehanna amendment request, supplemental information was provided by Susquehanna to facilitate the performance of independent fluence calculations.
Amendment 201 to the CNS Operating License, issued October 31, 2003, implemented the Boiling Water Reactor Vessel and Internals Reactor Pressure Vessel Integrated Surveillance Program (ISP). As part of the ISP, CNS vessel surveillance capsules are evaluated using fluence calculations that conform with Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.
Additionally, CNS committed in Amendment 201 to a recalculation of fluences for previously pulled surveillance capsules in conformance with RG 1.190.
NLS2004005 Attachment I Page 7 of I1 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazard posed by issuance of an amendment. Nebraska Public Power District (NPPD) has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92 (c).
This proposed change will revise Technical Specification (TS) 3.4.9 Pressure Temperature (P/T) curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3, to allow plant operation from 21 Effective Full Power Years (EFPY) through 32 EFPY. The proposed change includes a full set of updated P/T curves for pressure test, core not critical and core critical conditions. The three regions of the reactor pressure vessel that are evaluated are the beltline region, the bottom head region, and the feedwater nozzle/upper vessel region. These regions bound all other regions with respect to brittle fracture.
The following evaluation supports a finding of"no significant hazards consideration" associated with this proposed change.
5.1.1 Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed revisions to the Cooper Nuclear Station (CNS) P/T curves are based on the recommendations in Regulatory Guide (RG) 1.99, Revision 2, and are therefore in accordance with the latest Nuclear Regulatory Commission (NRC) guidance. The evaluation for the P/T curves for 32 EFPY was performed using the approved methodologies of 10 CFR 50, Appendix G. The curves generated from these methods provide guidance to ensure that the PIT limits will not be exceeded during any phase of reactor operation. Accordingly, the proposed revision to the CNS P/T curves is based on an NRC accepted means of ensuring protection against brittle reactor vessel fracture, and compliance with 10 CFR 50 Appendix G. Therefore, this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Based on the above, NPPD concludes that the proposed TS change to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not significantly increase the probability or consequences of an accident previously evaluated.
NLS2004005 Attachment I Page 8 of I 5.1.2 Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change updates existing P/T operating limits to correspond to the current NRC guidance. The proposed TS change provides more operating flexibility in the P/T curves for in-service leakage and hydrostatic pressure testing, non-nuclear heatup and cooldown, and criticality, with the benefits primarily in the area of pressure test being performed at a lower temperature. The proposed change does not involve a physical change to the plant, add any new equipment or any new mode of operation. These changes demonstrate compliance with the brittle fracture requirements of 10 CFR 50 Appendix G, and therefore do not create the possibility for a new or different kind of accident from any accident previously evaluated.
Based on the above, NPPD concludes that the proposed TS change to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not create the possibility of a new or different kind of accident from any accident previously evaluated.
5.1.3 Do the proposed changes involve a significant reduction in the margin of safety?
The proposed change to the CNS P/T curves does not create a significant reduction in the margin of safety. The proposed change revises the existing CNS P/T curves to be consistent with recommendations of RG 1.99, Revision 2, the current NRC guidance given to ensure compliance with 10 CFR 50 Appendix G.
For P/T curve development ASME Section Xl Code Case N-640 uses the Kic fracture toughness curve as the lower bound for fracture toughness. P/T curves based on the Kic fracture toughness limits enhance industrial safety by expanding the P/T window in the low-temperature operating region. The potential benefits are a reduction in the duration of the pressure test and, associated increase in personnel safety, while conducting inspections in primary containment.
Therefore, operational flexibility is gained while maintaining an adequate margin of safety to Reactor Pressure Vessel brittle fracture. As stated above the development of the P/T curves to 32 EFPY was performed per the guidelines of 10 CFR 50 Appendix G, and thus, the margin of safety is not significantly reduced as the result of the proposed TS change.
Based on the above, NPPD concludes that the proposed TS change to TS 3.4.9 P/T curves, figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 does not involve a significant reduction in the margin of safety.
NLS2004005 Attachment I Page 9 of II In conclusion, NPPD has determined that the proposed amendment involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Regulatory Requirements and Guidance 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention of Lightwvater Nuclear Power Reactors for Normal Operation," provides the requirements that the pressure and temperature limits as wvell as the associated vessel surveillance program are consistent with 10 CFR 50 Appendix G, "Fracture Toughness Requirements," and 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
10 CFR 50 Appendix G and Appendix H also describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in establishing P/T curves. 10 CFR 50 Appendix G specifies the fracture toughness and testing requirements for reactor vessel material in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G. 10 CFR 50, Appendix G also requires the prediction of the effects of neutron irradiation on the vessel embrittlement by calculating the ART and Charpy upper shelf energy. Generic Letter 88-1 1, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations,"
requires that the methods in Regulatory Guide 1.99, Revision 2, be used to predict the effect of neutron irradiation on the reactor vessel material. Appendix H of 10 CFR 50 requires the establishment of a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
The heatup and cooldown process for CNS is controlled by TS 3.4.9 P/T curves, which are developed based on fracture mechanics analysis. These limits are developed according to Appendix G of ASME Boiler and Pressure Vessel Code,Section XI, 10 CFR 50, Appendix G and ASME Section XI Code Case N-640.
There are two lower bound fracture toughness curves available in Section Xl: Kia, which is a lower bound on all static, dynamic and arrest fracture toughness, and Kic, which is a lower bound on static fracture toughness only. ASME Section Xl Code Case N-640 changes the fracture toughness curve used for development of P/T limit curves from Kia to Kic. RG 1.147, Revision 13, identifies ASME Section XI Code Case N-640 as an NRC acceptable ASME Section Xl code case.
In conclusion, based on the considerations discussed above, (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.
NLS2004005 Attachment I Page 10ofl 6.0 Environmental Consideration 10 CFR 51.22(b) allows that an environmental assessment (EA) or an environmental impact statement (EIS) is not required for any action included in the list of categorical exclusions in 10 CFR 51.22(c). 10 CFR 51.22(c)(9) identifies an amendment to an operating license which changes a requirement with respect to installation or use of a facility component located within the restricted area, or which changes an inspection or a surveillance requirement, as a categorical exclusion if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amount of any effluents that may be released off-site, or (3) result in an increase in individual or cumulative occupational radiation exposure.
NPPD has reviewed the proposed license amendment and concludes that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the proposed license changes. The basis for this determination is as follows:
I.
The proposed license amendment does not involve significant hazards as described previously in the No Significant Hazards Consideration Evaluation.
- 2.
This proposed change does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released off-site.
The proposed license amendment does not introduce any new equipment, nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. NPPD has concluded that there will not be a significant increase in the types or amounts of any effluents that may be released off-site and these changes do not involve irreversible environmental consequences beyond those already associated with normal operation.
- 3.
This change does not adversely affect plant systems or operation and therefore, does not significantly increase individual or cumulative occupational radiation exposure beyond that already associated with normal operation.
7.0 References
- 1.
Structural Integrity Associates Calculation No. COOP-05Q-301, Development of Updated Pressure Test (Curve A) P-T Curves.
- 2.
Structural Integrity Associates Calculation No. COOP-05Q-302, Development of Updated Heatup/Cooldown (Curves B &C) P-T Curves.
NLS2004005 Attachment I Page 11 of 11
- 3.
ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternate Reference Fracture Toughness for Development of P-T Limit Curves, "Section XI, Division 1, Approved February 26, 1999
- 4.
USAR, Section IV-2.6.2, Brittle Fracture Consideration
- 5.
Letter NSD930270, from G. R. Horn, NPPD, to USNRC, Submittal of Reactor Vessel Surveillance Test results, February 25,1993.
- 6.
1995 Edition, 1996 Addenda, ASME Boiler and Pressure Vessel Code Section Xi, Appendix G, "Fracture Toughness Criteria for Protection Against Failure"
NLS2004005 Page 1 of II ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS AND ASSOCIATED BASES REVISIONS MARKUP FORMAT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Listing of Revised Pages TS Pages 3.4-23 3.4-24 3.4-25 TS Bases Pages B 3.4-44 B 3.4-46 B 3.4-49 B 3.4-52 Note: Bases are provided for information. Following approval of the proposed TS change, Bases changes will be implemented in accordance with TS 5.5.10, Technical Specification (TS) Bases Control Program.
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RCS P/T Limits 3.4.9 a
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Temperature/Pressure Limits for Criticality Cooper 3.4-25 Amendment No. 178
RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,600 1,500 1,400 1,300 1,200 a.
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Pressure/Temperature Limits for Criticality Cooper 3.4-25 Amendment No. XXX
RCS P/T Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.9 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.
This Specification contains PIT limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, criticality, and data for the maximum rate of change of reactor coolant temperature.
Each P/T limit curve defines an acceptable region for normal operation.
The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and 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 component most subject to brittle failure. Therefore, the LCO 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, abnormal operational transients, and system hydrostatic tests.
It mandates the use of the ASME Code, Section 1I1, Appendix G (Ref. 2).
The NRC has also approved the use of alternate fracture toughness curves for establishing these limits (Ref. 10).
The actual shift in the RTNOT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with BWRVIP-86-A (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be
- adjusted, (continued)
Cooper B 3.4-44 Cooper B 3.4-44
~~~~~~~~~~~~~~II4 1
RCS P/T Limits B 3.4.9 BASES APPLICABLE P/T limits are not derived from any DBA, there are no acceptance limits SAFETY ANALYSES related to the P/T limits. Rather, the P/T limits are acceptance limits (continued) themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 8).
LCO The elements of this LCO are:
- a.
RCS pressure and temperature (Beltline. Bottom Head, and Unper Vessel) are within the applicable limits of Figure 3.4.9-1 and Figure 3.4.9-2, and heatup or cooldown rates are < 100'F when averaged over a one hour period during RCS heatup, cooldown, and inservice leak and hydrostatic testing (The Adjusted Reference Temperature (ART) beltline region must be determined from the appropriate beltline curve (1 3, 1v, or 21 Effcctive Full rower Years 1
IEFPY)
Figure 3.4.9-2,dependng on the current accumulated nu~mber fEFPY);
- b.
The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is <
145 0F during recirculation pump startup;
- c.
The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is < 50'F during recirculation pump startup;
- d.
RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior to achieving criticality; and
- e.
The reactor vessel flange and the head flange temperatures are >
80*F when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
(continued)
I I
I I
I I
Cooper B 3.4-46 Revision 0 1
RCS P/T Limits RCS P/T Limits B 3.4.9 BASES ACTIONS B.1 and B.2 (continued) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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 > 212'F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code, Section Xl, Appendix E (Ref. 7), 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 C.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 reactor pressure vessel integrity.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within RCS pressure, RCS temperature, and RCS heatup and cooldown rate limits by monitoring the bottom head drain, recirculation loop temperatures, and RPV metal temperatures (Beltline, Bottom Head, and Upper Vessel) is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered (continued)
Cooper B 3.4-49 Revision O I
RCS P/T Limits RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.5. SR 3.4.9.6. and SR 3.4.9.7 (continued)
REQUIREMENTS The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
SR 3.4.9.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs.
SR 3.4.9.6 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature < 900F in MODE 4.
SR 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 100'F in MODE 4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be within the specified limits.
REFERENCES
- 1.
- 2.
ASME, Boiler and Pressure Vessel Code, Section 1I1, Appendix G.
- 3.
BWRVIP-86-A, October 2002.
- 4.
- 5.
Regulatory Guide 1.99, Revision 2, May 1988.
- 6.
USAR, Section IV-2.6.
- 7.
ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
- 8.
- 9.
USAR, Appendix G.
- 10.
ASME Xl Code Case N-640 Cooper B 3.4-52 C1126/03 I
NLS2004005 Page l of8 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS AND ASSOCIATED BASES REVISIONS FINAL TYPED FORMAT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR46 Listing of Revised Pages TS Pages 3.4-23 3.4-24 3.4-25 TS Bases Pages B 3.4-44 B 3.4-46 B 3.4-49 B 3.4-52 Note: Bases are provided for information. Following approval of the proposed TS change, Bases changes will be implemented in accordance with TS 5.5.10, Technical Specification (TS) Bases Control Program.
RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,600 1,500 1,400 i
I I
I L
II I
_ 1,300 u0 0.
CL 0
1,200 I
1,100 0.
0 I--
iJ 1,000 u)
U)LU 900 o
800 z
700 600 D
500 U)
- 9.
4nn I-
-If OF '
I ! I 2~~
I I I -I 1
I I 807, 448 psig
_-I I
I
,I I
I 7 01 140-F, 541 psig 1407, 313 psiig
. I I
Safe r
Operating Region I
I I
I I
I I
300 200 100 -
0-I r
_~=r =i 80F, 341 psig l z
lT~ j 80F, 313 psig
-i--I Boltup
==
8F
~ Beltline Bottom Head Upper Vessel 9
i 1
11 1
1 0
50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Figure 3.4.9-1 (page 1 of 1)
Pressure/Temperature Limits for Non-Nuclear lcatup or Cooldown Following Nuclcar Shutdown Cooper 3.4-23 Amendment No. XXX
RCS P/T Limits 3.4.9 Cooper Pressure Test Curve (Curve A), 32 EFPY 1,600 1,500 1,400 1,300
=
a.
0 4
KU rx n
0 I-fit KU U) 0, KU a-1,200 1,100 1,000 900 800 700 600 500 400
,11=
123°F, 753 psi, I
801F, 597 psi-,
I __
8 F,555 psig I £ I
i i
I I
I i
I 11 !
i i
Safe Operating r
t _
Region
-Beittine
- 'Bottom H-lead
-Upper Vessel I
I 1
I 2 r 18 00 20 4
300 200 100 0
l o tup
>80OF t_-
123°F, 269 psig I_
0 20 40 60 80 100 120 140 160 MINIMUM REACTOR VESSEL METAL TEMPERATURE: (Ff)
Figure 3.4.9-2 (page 1 of 1)
Pressure/Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests Cooper 3.4-24 Amendment No. XXX
RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,600 1,500 1,400 1,300
-1 I I I
I I
I I
f-I I
i I
i I
i I
I I
I f
i I
I I
I I
I I
I I
I i
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
i I
I I
I I
I I
I I
I I
I I
I I
.2 1,200 In o< 1100 0-0 1,000
-j all W
900 rr O
800 I.-
It 700 z
I--
E 600
=
500 C,,
a-oo ax 400o
~-
Curve C
-E Safe Operating Region 300 -
200 100 0-
_1 80'
=
Minimum
-I Criticality with Normal Water
=
Level 80-F F
,.4
.9 (g
r I i
I 21, 313 psi g
'F, 313 psig 0
50 100 150 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F) 250 300 Figurc 3.4.9-3 (p~age I of 1)
Pressure/Temperature Limits for Criticality 3.4-25 Cooper Amendment No. XXX
RCS P/T Limits B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.9 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.
This Specification contains P/T limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, criticality, and data for the maximum rate of change of reactor coolant temperature.
Each P/T limit curve defines an acceptable region for normal operation.
The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and 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 component most subject to brittle failure. Therefore, the LCO 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, abnormal operational transients, and system hydrostatic tests.
It mandates the use of the ASME Code, Section 1I1, Appendix G (Ref. 2).
The NRC has also approved the use of alternate fracture toughness curves for establishing these limits (Ref. 10).
The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with BWRVIP-86-A (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be
- adjusted, (continued)
Cooper B 3.4-44 Co e B 3 1
e63 I
RCS P/T Limits B 3.4.9 BASES APPLICABLE P/T limits are not derived from any DBA, there are no acceptance limits SAFETY ANALYSES related to the P/T limits. Rather, the P/T limits are acceptance limits (continued) themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 8).
LCO The elements of this LCO are:
- a.
RCS pressure and temperature (Beltline, Bottom Head, and Upper Vessel) are within the applicable limits of Figure 3.4.9-1 and Figure 3.4.9-2, and heatup or cooldown rates are < 1000F when averaged over a one hour period during RCS heatup, cooldown, and inservice leak and hydrostatic testing (The Adjusted Reference Temperature (ART) beltline region must be determined from Figure 3.4.9-2;
- b.
The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is <
1450F during recirculation pump startup;
- c.
The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is < 500F during recirculation pump startup;
- d.
RCS pressure and temperature are within the criticality limits specified in Figure 3.4.9-3, prior to achieving criticality; and
- e.
The reactor vessel flange and the head flange temperatures are >
80'F when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
(continued)
I Cooper B 3.4-46 Revison-
RCS P/T Limits B 3.4.9 BASES ACTIONS B.1 and B.2 (continued) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
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 > 212'F. 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. 7), 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 C.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 reactor pressure vessel integrity.
SURVEILLANCE SR 3.4.9.1 REQUIREMENTS Verification that operation is within RCS pressure, RCS temperature, and RCS heatup and cooldown rate limits by monitoring the bottom head drain, recirculation loop temperatures, and RPV metal temperatures (Beltline, Bottom Head, and Upper Vessel) is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered (continued)
Cooper B 3.4-49 Revision O I
RCS P/T Limits RCS PIT Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.5, SR 3.4.9.6. and SR 3.4.9.7 (continued)
REQUIREMENTS The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
SR 3.4.9.5 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs.
SR 3.4.9.6 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature < 900F in MODE 4.
SR 3.4.9.7 is modified by a Note that requires the Surveillance to be initiated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 100'F in MODE 4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be within the specified limits.
REFERENCES
- 1.
- 2.
ASME, Boiler and Pressure Vessel Code, Section 1II, Appendix G.
- 3.
BWRVIP-86-A, October 2002.
- 4.
- 5.
Regulatory Guide 1.99, Revision 2, May 1988.
- 6.
USAR, Section IV-2.6.
- 7.
ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
- 8.
- 9.
USAR, Appendix G.
- 10.
ASME Xl Code Case N-640 Cooper B 3.4-52 Cooper B
3.4-52 ~~~~~~~~~
~~~~~11126/
1
NLS2004005 Enclosure I Page I of 62 ENCLOSURE I REVISED PRESSURE TEMPERATURE CURVES METHODOLOGY AND SUPPORTING CALCULATIONS PROVIDED FOR COOPER NUCLEAR STATION BY STRUCTURAL INTEGRITY ASSOCIATES, INC.
COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR46
Structural Integrity Associates, Inc.
6595 S. Dayton Street Suite 3000 Greenwood Wfage. CO 80111-6145 Phone:
303-792-0077 Fax:
303-792-2158 wwshbuctntcon dimpusestrudintcom October 15, 2003 CRL-03-004 SIR-03-074, Rev. 1 Mr. Ken Thomas Nebraska Public Power District Cooper Nuclear Station P. O. Box 98 Brownville, NE 68321
Subject:
Updated Pressure-Temperature Curves for Cooper
Reference:
Cooper Purchase Order No. 4500000438 dated 03/06/2003.
Dear Ken:
The attachment to this letter documents the updated set of pressure-temperature (P-T) curves developed for the Cooper Nuclear Station, in accordance with Si's Quality Assurance Program. This work was performed in accordance with the referenced contract, and includes a full set of updated P-T curves (i.e.,
pressure lest, core not critical, and core critical conditions) for 32 effective full power years (EFPY).
The curves wvere developed in accordance with 1995 Edition, 1996 Addenda ASME Code Section Xl Appendix G, U.S. I OCFR 50 Appendix G, and ASME Code Case N-640.
The inputs, methodology, and results for this effort are summarized in the attachment. The calculations for this work (COOP-05Q-301 and -302) are also attached.
Please don't hesitate to call me if you have any questions.
Prepared By:
L Reviewed By:
Carl R. Li nrptoj q'~y e
Engineer E
cer Approved By:
_4 it GaryL. Stevens, P. E.
Senior Associate crl Attachments cc:
COOP-05Q-401 Ausin.TX Chsdrte, NC 512-533-9191 704-573-1369 M. Slonmon. CT Sunrse.FL Rocky tMD 860-599-6050 954-572 2902 301-231-7746 San Jose, CA UnWontown.04H 408-978-8200 330899753
ATT'ACHMENT Updated P-T Curves for Cooper 1.0 Introduction This attachment documents the updated set of pressure-temperature (P-T) curves developed for the Cooper Nuclear Station (CNS). This work includes a full set of updated P-T curves (i.e.,
pressure test, core not critical, and core critical conditions) for 32 effective full power years (EFPY). The curves were developed using the methodology specified in ASME Code Case N-640 [4], as well as the 1995 Edition, 1996 Addenda ASME Code Section XI Appendix G [3],
1 OCFR50 Appendix G [I ], and WRC-1 75 [7]. The improvement realized from the Code Case methodology is as much as 60'F, and is primarily obtained from using the critical fracture toughness, KIC, in accordance with Code Case N-640.
2.0 RTNDT Values Adjusted reference temperature (ART) values were obtained from Reference [5] for the Cooper reactor pressure vessel (RPV) materials in accordance with NRC Regulatory Guide 1.99, Revision 2 19]. Table I shows the results of the calculations for 32 EFPY. The most limiting beltline material is the Lower Intermediate Longitudinal Weld.
2 V
Strvctural InlegrilyAssociates, Inc.
Attachment to SIR-03-074, Rev. I/CRL-03-004
Table 1: Cooper RPVMaterial ART32 EFPY Calculations Chenmistry Adjustments For 1142 Component Part No.
Heat Initial RT m Chemistry Factor ARTo0 Margin Terms ART No.
rF)
Cu (wt i)
Ni (wt V.)
VF)
(F) a, F) a, (1F)
(F)
Beltine Plates' G2803-1 C2274-1 14.0 0.20 0 68 153.00 87.5 17.0 0.0 118.6 Lower She1 G2803-2 C2307-1 0.0 0.21 0-73 162.80 93.1 17.0 0.0 110.1 Plates G2803-3 C2274-2
-8.0 0.20 0.68 153.00 87.5 17.0 0.0 96.5 Lower-G2801-7 C2407-1
-10.0 0.13 0.65 92.25 64.6 17.0 0.0 71.6 Intermediate Sbel' G2802-1 C233 1-2 10.0 0.17 0.58 125.30 87.7 17.0 0.0 114.7 Plates G2802-2 C2307-2
-20.0 0.21 0.73 162.80 113.9 17.0 0.0 110.9 Beltine Welds' Lower~tudn I
2-233 12420
-50.0 0.22 1.02 234.50 133.0 28.0 0.0 111.0 Lower-Intermediate 1-233 27204
-50.0 0.19 0.97 215.65 149.6 28.0 0.0 127.6 Longitudinal Lower to Lower-1-240 21935
-50.0 0.20 0.69 175.30 121.6 280 0.0 99.6 Non-Beltine Regions' Closure Flange 20.0 20.0 Region Bottom Head 28.0 Region 28.0 28.0 Fluence Wall Thickn ss (in.)
Peak Fluence Attenuation @
Peak Fhrence @
Fluence Factor, FF Location Fun 114t at ID for 32 1142 e
..4, 1t fo/cr 32e EFPY -
EFPY EOL 114 ef (n~crn2)
Lower-lnt Shel 5.375 1.344 1.57E418 0.724 1.14E+18 0.443 Lower Shell 6 375 1.594 1.10E418 0.682 7.52E.17 0.362 Note 1: Beftline material input per Table 7-2 of Reference 15)
Note 2: Non-betline material input per Reference 15].
Surveillance Ad nt Pi;tes s
3 V3 Structural Integrity Associates, Inc.
Attachment to SIR-03-074, Rev. ]/CR1-03-004
3.0 P-T Curve Methodology The P-T curve methodology is based on the requirements of References [I] through [4] and [7].
Thie supporting calculations for the curves are contained in References [ 1 0] and [ I I ]. There are three regions of the reactor pressure vessel (RPV) that are evaluated: (1) the beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. These regions bound all other regions with respect to brittle fracture.
The approach used for the beltline and bottom head (all curves), and upper vessel (Curve A only) includes the following steps:
- a.
Assume a fluid temperature, T. The temperature of the metal at the assumed flaw tip, T1/4t (i.e., 1/4t into the vessel wall) is conservatively assumed equal to the fluid temperature. The assumed temperature must also account for an instrument uncertainty of 5PF [ 1 5].
- b.
Calculate the allowable stress intensity factor, KIc, based on T1 1/4 using the relationship from Code Case N-640 [4], as follows:
K
= 20.734 c1 02(TIAIARTND71] + 33.2 (eqn. from Ref. [2])
where: T1 /41
= metal temperature at assumed flaw tip (-F)
ARTNDT = adjusted reference temperature for location under consideration and desired EFPY (TF)
Kic
= allowable stress intensity factor (ksi'inch)
C.
Calculate the thermal stress intensity factor, Krr for the beltline and bottom head regions, or from finite element results for the feedwater nozzle/upper vessel region.
- d.
Calculate the allowable pressure stress intensity factor, Kwn, using the following relationship:
Kip = (Kic-Krr)/SF where:
Kip = allowable pressure stress intensity factor (ksNinch)
SF
= safety factor
=
1.5 for pressure test conditions (Curve A)
=
2.0 for heatup/cooldown conditions (Curves B and C)
- e.
Compute the allowable pressure, P, from the allowable pressure stress intensity factor, Kip.
Attachment to SIR-03-074, Rev. 1/CRI.-03-004 4
Structural Integrity Associates, Inc.
- f.
Subtract any applicable adjustments for pressure from P. The beltline and bottom head include a pressure adjustment of 19.9 psig to account for the static pressure bead of a full vessel. An instrument error of 24.0 psig (2% of 1,200 psig) was applied [15].
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.
The approach used for the upper vessel (Curves B & C) includes the following steps:
- a.
Assume a fluid pressure, P. The pressure includes an instrument uncertainty of 24.0 psig [15].
- b.
Calculate the thermal stress intensity factor, Krr, based on finite element stresses.
The feedwater nozzle stresses were obtained from the finite element analysis results contained in Reference [8].
The highest linearized (membrane and membrane + bending) thermal stresses for the entire design basis transients were selected to encompass all expected operating conditions.
u3.
= 43.975 ksi @ 575°}; for SA-508 Cl. 2 [8, 12]
Calculate t"2. The resulting Mm value is obtained from G-2214.1[3].
Kim is calculated from the equation in Paragraph G-2214.1 [3]:
Kim = Mm* 0;sm Ksb is calculated from the equation in Paragraph G-2214.2 [3]:
Kib = (2/3) Mm* usb The total Krr is therefore:
Krr R*SF*(Klm+ Klb) where:
R
=
correction factor, calculated to consider the nonlinear effects in the plastic region based on the assumptions and recommendations of WRC Bulletin 175 [7].
[cry, - apm + ((aowt
- ay,) / 30)] / ((total - upm)
SF
=
Safety Factor for Krr
=
1.3 (conservatively used based on the recommendation in WRC-175 [7])
Attachment to SIR-03-074, Rev. 1/CRL-03-004 Structural Integrity Assocates, Inc.
- c.
Compute the allowable pressure stress intensity factor, Kip, is as follows:
KU,p = F7(a /r. 4na a["
where:
r
=
actual inner radius of nozzle rc
=
nozzle comer radius 114]
rn
=
apparent radius of nozzle = r1 + 0.29r, t'
=
nozzle corner thickness a
=
crack depth (inches) 1/4t' F(arn) nozzle stress factor, from Figure A5-1 of [7]
Kip
=
allowable pressure stress intensity factor (ksi-4nch).
TPM
=
primary membrane stress, PRlt (primary bending stresses are conservatively treated as membrane stresses, so 0 pb = 0)
- d.
Calculate the allowable stress intensity factor, KIc, using the following relationship for a heatup/cooldown P-T curve:
2.0 thus:
K1c =2.OKI + KIT
- c.
Calculate the temperature, T1141, using the relationship from Code Case N-640
[4], as follows:
K= 20.734 cJ0-2(TB42-MTM)) + 33.2 (cqn. from Ref. [2])
where: Ti14t
= metal temperature at assumed flaw tip ('F),
assumed equal to T, the temperature at the inner vessel wall ARTNDT = adjusted reference temperature for location under consideration and desired EFPY ('F)
Klc
= allowable stress intensity factor (ksivinch) thus:
T,4g =50*LNI Kc -33
+AR7NDT L20.734 Attachment to SIR-03-074, Rev. /CRL.03.004 6C Structural Integfrity Associates, Inc.
- f.
The curve was generated by scaling the stresses used to determine the pressure and thermal stress intensity factors. The primary stresses were scaled based on pressure, while the secondary stresses were scaled based on temperature difference.
- g.
Repeat steps (a) through (I) for other pressures to generate a series of P-T points.
The following additional requirements were used to define the P-T curves. These limits are established in Reference [I]:
For Pressure Test Conditions (Curve A):
If the calculated pressure, P, is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDr of the limiting flange material + 90TF. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 200F [5].
If the calculated pressure, P, is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is typically greater than or equal to the RTNDT of the limiting flange material. GE typically applies an additional 60TF margin to the RTNjrW value and has been a standard recommendation for the BWR industry for non-ductile failure protection.
For the Cooper flange material, this minimum would be 80TF (i.e., 20 + 60TF). Since the 60TF margin is only a recommendation, the minimum temperature for Cooper was set to 80TF to be consistent with past work as well as adequately encompass instrument uncertainty. The gauge's uncertainty of +/- 5T is not applied here since the included margin just described adequately encompasses instrument uncertainty.
For Core Nol Critical Conditions (Curve B):
If the calculated pressure, P, is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDT of the limiting flange material + ] 20TF. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 20TF [5].
If the calculated pressure, P, is less than or equal to 20% of the pre-serviec hydro test pressure, the minimum temperature is typically greater than or equal to the RTNDT of the limiting flange material. GE typically applies an additional 60TF margin to the RTNDT value and has been a standard recommendation for the BWR industry for non-ductile failure protection. For the Cooper flange material, this minimum would be 80TF (i.e., 20 + 600F). Since the 60TF margin is only a recommendation, the minimum temperature for Cooper was set to 80TF to be consistent with past work as well as adequately encompass instrument uncertainty. The gauge's uncertainty of +/- 50F is not applied here since the included margin just described adequately encompasses instrument uncertainty.
Attachment to SIR-03-074, Rev. IICRL-03-004 Structural Integrity Associates, Inc.
For Core Critical Conditions (Curve C):
Per the requirements of Table I of Reference 11], the core critical P-T limits must be 40'F above any Pressure Test or Core Not Critical curve limits. Core Not Critical conditions are more limiting than Pressure Test conditions, so Core Critical conditions are equal to Core Not Critical conditions plus 40'F.
Another requirement of Table 1 of Reference [1] (or actually an allowance for the BWR), concerns minimum temperature for initial criticality in a startup.
Given that water level is normal, BWRs are allowed initial criticality at the closure flange region temperature (ART + 60'F) if the pressure is below 20% of the pre-service hydro test pressure.
Also per Table I of Reference [1], at pressures above 20% of the pre-service hydro test pressure, the Core Critical curve temperature must be at least that required for the pressure test (Pressure Test Curve at 1,100 psig). As a result of this requirement, the Core Critical curve must have a step at a pressure equal to 20% of the pre-service hydro pressure to the temperature required by the Pressure Test curve at 1,100 psig, or Curve 13 + 40'F, whichever is greater.
After accounting for instrument uncertainties, the resulting pressure and temperature series constitutes the P-T curve. The P-T curve relates the minimum required fluid temperature to the reactor pressure.
5.0 P-T Curves Tabulated values for the P-T curves are shown in Tables 2 through 8. The resulting P-T curves arc shown in Figures I through 3. Note that the upper vessel (non-beltline) curve is limiting for core not critical conditions for 32 EFPY.
Attachment to SIR-03-074, Rev.
8/CRL,03-004 Structural Integrity Associates, Inc.
6.0 References
- 1.
U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," 1-1-98 Edition.
- 2.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components Nonmandatory Appendix A, "Analysis of Flaws,"
- 3.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G. "Fracture Toughness Criteria for Protection Against Failure," l996-Ediin. /?9
/CAG-Z A't/?/dŽ
- 4.
ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," Section Xl, Division 1, Approved February 26, 1999.
- 5.
GE Document No. GE-NE-523-159-1292 (DRF B13-01662), "Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," Revision 0, February 1993, S1 File No. COOP-05Q-202.
- 6.
Not Used.
- 7.
WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," PVRC Ad Hloc Group on Toughness Requirements, Welding Research Council, August 1972.
- 8.
CBI Stress Report DC22A7245, Revision 0, "Feedwater Nozzle Modification Cooper RPV," 3/20/80, SI File NPPD-13Q-205.
- 9.
USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 3054), May 1988.
- 10.
S1 Calculation No. COOP-05Q-301, Revision 1, "Development of Updated Pressure Test (Curve A) P-T Curves," October 2003.
- 11.
SI Calculation No. COOP-05Q-302, Revision 1, "Development of Updated Ieatup/Cooldown (Curves B & C) P-T Curves," October 2003.
- 12.
ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, Appendices, 1989 Edition.
AthntSR
/R00 9
V Structural Integrity Associates, Inc.
Aittachmnent to SIR-03-074, Rev. I/CRL-03.004
- 13.
Combustion Engineering Drawing No. E-232-230, Revision 3, "General Arrangement Elevation for: General Electric Co. APED 218" I.D. BWR," SI File No. NPPD-06Q-208.
- 14.
SI Calculation NPPD-13Q-302, Revision 0, "Feedwater Nozzle Stress Analysis."
- 15.
NPPD Memo DED 2003-005, Alan Able to Ken Thomas, dated August 14,2003, "Instrument Uncertainty Associated With Technical Specification 3.4.9," S1 File No.
COOP-05Q-203.
Attachment to STR-03-074, Rev. I/CRL-03-004 10 Strctural Integrity Associates, Inc.
Table 2 Tabulated Values for Beitline Pressure Test Curve (Curve A) for 32 EFPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
Inputs:
Plant Cooper.
Component =
Belttimae:
Vessel thickness, t =
5.875
,.inches. so t=
2.424 nch Vessel Radius. R =
.110.375 inches ART=
127.6 i'F 32 EFPY--
Cooldown Rate, CR 0
°F/hr Kn=
O ksi-inch"'
AT 1 41 =
0! 0
'F (no thermal for pressure test)
Safety Factor 1 50 (for pressure test)
Temperature Adjustment
-5 0
F Height of Water for a Full Vessel =
551175-;H inches Pressure Adjustment =,
19
%., psig (hydrostatic pressure for a full vessel at 70'F)
Pressure Adjustment = '
24 0 +
psig (Instrument Uncertainty)
Hydro Test Pressure = 7'!d,
i psig Flange RT1 DT =
,. '20 F
Gauge Fluid Temperature 80F) 80.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 200.0 114t Temperature (IF) 80.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 Kc (ksilinch 1 "2) 41.20 40.44 40.74 41.04 41.36 41.70 42.04 42.41 42.78 43.17 43.58 44.00 45.14 46.39 47.78 49.32 51.01 52.88 54.95 57.24 59.77 62.56 65.65 69.07 72.84 77.01 81.61 86.70 92.33 98.55 105.42 113.02 Kip (ksi'inch1n) 27.47 26.96 27.16 27.36 27.58 27.80 28.03 28.27 28.52 28.78 29.05 29.33 30.09 30.93 31.85 32.88 34.01 35.26 36.64 38.16 39.85 41.71 43.77 46.04 48.56 51.34 54.41 57.80 61.55 65.70 70.28 75.35 Calculated Pressure P
(psig) 0 641 645 650 655 661 666 672 678 684 690 697 715 735 757 781 808 838 871 907 947 991 1040 1094 1154 1220 1293 1374 1463 1561 1670 1790 Temperature for P-T Curve (fr) 80.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 200.0 Adjusted Pressure for P-T Curve (psig) 0 597 601 606 611 617 622 628 634 640 646 653 671 691 713 737 764 794 827 863 903 947 996 1,050 1,110 1.176 1.249 1.330 1.419 1,517 1,626 1.746 11 Sfructural integrity Associates, Inc.
Attachment to SIR-03-074, Rev. 1/CRL-03-004
- Table 3 Tabulated Values for Bottom lead Pressurc Test Curve (Curve A)
Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
InDuts:
Plant =
Cooper.
Component = Bottom Head- (Penetrations Portion)
Vessel thickness. t =
':3 inches, so 4t=
1.785 4inch Vessel Radius. R =
110.37S inches ART =
28.0 F==All=>
.: AIEFPYs.
Safety Factor =
1.50 Safety Factor =
3.00.
Bottom Head Penetrations M,, =
1.85 Temperature Adjustment =
5.0
- F (Instrument Uncertainly)
Height of Water for a Full Vessel 551.75 inches Pressure Adjustment =
19 9 psig (hydrostatic pressure for a full vessel at 70-F)
Pressure Adjustment =
24.0 psig (Instrument Uncertainty)
Unit Pressure =
1,563
, psig Flange RTNOt =
20 0 F
Gauge Fluid Temperature VFI 1/14t Tem perature (F)
Calculated Pressure P
(osia)
Kic Kip (ksi inch2
&)
(lksi-inch"2)
Tem perature for P-T Curve (F)_
Adjusted Pressure for P-T Curve (osiq) 80.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 118.0 120.0 122.0 124.0 126.0 1t28.0 130.0 132.0 134.0 136.0 138.0 140.0 142.0 144.0 146.0 148.0 80.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 97.0 99.0 101.0 103.0 105.0 107.0 109.0 111.0 113.0 115.0 117.0 119.0 121.0 123.0 125.0 127.0 129.0 131.0 133.0 135.0 137.0 139.0 141.0 143.0 91.86 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.4 3 106.30 109.28 112.38 115.62 118.98 122.48 126.12 129.92 133.86 137.97 142.25 146.70 151.33 156.15 161.17 166.39 171.83 177.48 183.37 189.50 195.88 202.52 209.43 216.62 224.10 231.90 240.00 61.24 57.52 58.96 60.47 62.03 63.66 65.35 67.12 68.95 70.86 72.85 74.92 77.08 79.32 81.65 84.08 86.61 89.24 91.98 94.83 97.80 100.89 104.10 107.44 110.93 114.55 118.32 122.25 126.33 130.59 135.01 139.62 144.41 149.40 154.60 160.00 0
599 614 629 646 662 680 698 718 737 758 780 802 825 850 875.
901 929 957 987 1018 1050 1083 1118 1154 1192 1231 1272 1315 1359 1405 1453 1503 1555 1609 1665 80 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 142 144 146 148 0
555 570 585 602 619 636 655 674 694 714 736 758 782 806 831 857 885 913 943 974 1.006 1.039 1.074 1,110 1.148 1.187 1.228 1.271 1.315 1.361 1.409 1.4 59 1.511 1,565 1.621 12 V
Structural Integrity Associates, Inc.
Attachment to SIR-03-074, Rev. I /CRL-03-004
Table 4 Tabulated Values for Feedwater Nozzle/Upper Vessel Region Pressure Test Curve (Curve A)
Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
Inputs:
Plant =
Cooper
. 1Y.;:....-...
Component =.Upper Vessel (based on FW nozzle)
ART = U,,20.0: -..; F======>
All EFPYs Vessel thickness, t =
I
.5.875 inches, so 't =
2.424 Vessel Radius, R =
1O375 ; inches Nozzle comer thickness, t' = Y7108K'<
inches, approximate F(a/m) =
- 1.
nozzle stress factor Crack Depth, a
=
inches winch Safety Factor =
Temperature Adjustment =
Height of Water for a Full Vessel =
Pressure Adjustment =
Pressure Adjustment =
Unit Pressure =
Flange RTND1 =
- 'F (not applied)
I inches psig (hydrostatic pressure for a full vessel at 70'F) psig (Instrument Uncertainty) psig
°F
-: 20.0 Gauge Fluid Temperature (0F) 80.0 80.0 118.0 118.0 123.0 128.0 133.0 138.0 114t Temperature (8F) 80.0 80.0 118.0 118.0 123.0 128.0 133.0 138.0 KOc (ksi'inchs'2) 102.04 102.04 180.40 180.40 195.88 200.00 200.00 200.00 K~p (ksi*inch 12) 68.03 68.03 120.26 120.26 130.59 133.33 133.33 133.33 Calculated Pressure P
(psig) 0 313 313 1693 1839 1877 1877 1877 Temperature for P-T Curve (8F) 80.0 80.0 123.0 123.0 128.0 133.0 138.0 143.0 Adjusted Pressure for P-T Curve (psig) 0 269 269 1649 1795 1833 1833 1833 13 V
Structural Integrity Associates, Inc.
Attachment to SIR-03-074, Rev. l/CR1-03-004
Table 5 Tabulated Values for Beltline Core Not Crntical Curve (Curve B) for 32 EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)
Inuts:
Plant =,:Cooper Component = - Botiiie6..-
Vessel thickness. t =
- i 5.875.S:
Inches, so Jt =
2.424 Jinch Vessel Radius. R 110 375.: inches ART 1=t27 6 F
.. 32 EFPY:
Cooldown Rate. CR 100 F/hr
=,.
7,97' ksi'inch"'
AT,,4, =
o.00 F = Conservatively assumed zero Safety Factor 2.00 M
=
Temperature Adjustment
'F (Instrument Uncertainty) not applied Height of Water for a Full Vessel i551.75
! inches Pressure Adjustment 1 9;'
psig (hydrostatic pressure for a full vessel at 70°F)
Pressure Adjustment = '
- 24O, psig (Instrument Uncertainty)
Hydro Test Pressure = t.,,;,,S63 psig Fbnge RTND =;
20F0 A4'F Gauge Fluid 114t Temperature Temperature (F)
('F) 80.0 75.0 80.0 75.0 82.0 77.0 87.0 82.0 92.0 87.0 97.0 92.0 102.0 97.0 107.0 102.0 112.0 107.0 117.0 112.0 122.0 117.0 127.0 122.0 132.0 127.0 137.0 132.0 142.0 137.0 147.0 142.0 152.0 147.0 157.0 152.0 162.0 157.0 167.0 162.0 172.0 167.0 177.0 172.0 182.0 177.0 187.0 182.0 192.0 187.0 197.0 192.0 202.0 197.0 207.0 202.0 212.0 207.0 217.0 212.0 222.0 217.0 K,
(kslilnch 12) 40.44 40.44 40.74 41.53 42.41 43.37 44.44 45.63 46.93 48.38 49.97 51.74 53.69 55.84 58.22 60.85 63.76 66.98 70.53 74.46 78.79 83.59 88.89 94.75 101.22 108.37 116.28 125.01 134.67 145.34 157.14 YIP (ksirlnch"r 2 )
16.23 16.23 16.38 16.78 17.22 17.70 18.24 18.83 19.48 20.20 21.00 21.88 22.86 23.93 25.12 26.44 27.89 29.50 31.28 33.24 35.41 37.81 40.46 43.39 46.62 50.20 54.15 58.52 63.35 68.69 74.58 Calculated Pressure P
(p519) 0 385 388 398 408 420 432 446 462 479 498 519 542 568 596 627 662 700 742 788 840 897 959 1029 1106 1190 1284 1388 1502 1629 1769 Temperature for P-T Curve (8F) 80.0 80.0 82.0 87.0 92.0 97.0 102.0 107.0 112.0 117.0 122.0 127.0 132.0 137.0 142.0 147.0 152.0 157.0 162.0 167.0 172.0 177.0 182.0 187.0 192.0 197.0 202.0 207.0 212.0 217.0 222.0 Adjusted Pressure for P-T Curve (psls) 0 341 345 354 364 376 389 403 418 435 454 475 498 524 552 583 618 656 698 744 796 853 916 985 1.062 1.147 1.240 1.344 1,458 1,585 1.725 14 V
Structural Integrity Associates, Inc.
Attachment to SIR-03-074, Rev. I/CRL-03-004
Table 6 Tabulated Values for Bottom Head Vessel Rep-ion Core Not Critical Curve (Curve B)
Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)
Inputs:
Plant =
.- Cooper*.: +
Component = Bottom Head' (Penetrations Portion)
Vessel thickness, t = .
3.1875> inches, so Jt =
1.785 4inch Vessel Radius, R =
110.375 inches ART =
28.0..
F======>
All EFPYs Safety Factor =
2.00 Stress Concentration Factor =
.3.00
- i'.- Bottom Head Penetrations Cooldown Rate, CR =/100.°F/hr KrT 1 7 3
ksiainch1 '
Temperature Adjustment 0
m
- F, Instrument Uncertainty Height of Water for a Full Vessel 551 75
'. inches (normal full vessel water level)
Pressure Adjustment 19 9 psig (full vessel at 70F)
Pressure Adjustment =
24.0 psig (Instrument Uncertainty)
Unit Pressure =
20.0
'F Gauge Temperature T
0.F) 80.0 80.0 82.0 86.0 90.0 94.0 98.0 102.0 106.0 110.0 114.0 118.0 122.0 126.0 130.0 134.0 138.0 142.0 146.0 150.0 154.0 158.0 Adjusted Temperature 75.0 75.0 77.0 81.0 85.0 89.0 93.0 97.0 101.0 105.0 109.0 113.0 117.0 121.0 125.0 129.0 133.0 137.0 141.0 145.0 149.0 153.0 Kic (ksilinch' 2) 86.28 86.28 88.44 93.05 98.03 103.43 109.28 115.62 122.48 129.92 137.97 146.70 156.15 166.39 177.48 189.50 202.52 216.62 231.90 248.44 266.37 285.79 Kip (ksi'inch' r) 42.27 42.27 43.36 45.66 48.15 50.85 53.78 56.94 60.38 64.09 68.12 72.48 77.21 82.33 87.88 93.89 100.39 107.45 115.08 123.36 132,32 142.03 Calculated Pressure P
(psig) 0 492 505 532 561 592 626 663 703 746 793 844 899 959 1023 1093 1169 1251 1340 1437 1541 1654 Temperature for P-T Curve (8F) 80 80 82 86 90 94 98 102 106 110 114 118 122 126 130 134 138 142 146 150 154 158 Adjusted Pressure for P-T Curve (Psig) 0 448 461 488 517 548 582 619 659 702 749 800 855 915 979 1049 1125 1207 1296 1393 1497 1610 15 V
Structural Integity Associates, Inc.
Attachment to SIR-03-074, Rev. I/CRL-03-004
Table 7 Tabulated Values for Feedwater NozzlwfUpper Vcssel Core Not Critical Curve (Curve B)
Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)
Inputs:
Plant Cooier Component =
.Uppor Vessel RTIjOr 200.:;.
- 'F Cy
-21.31 6
ksi for a pressure of 1.038 pslg Op 0.00 ksi for a pressure of 1.038 psig aCie.
5.89S5 ksi for a temperature of 55t-F O=
18.41 ksi for a temperature of 551'F Or=
.43.975 ksi D 575'F Nozzle comer thickness, r
- 7 108&,.
- inches approximate Jt =
a (1/4 t)
I 17 F(a/r.)
I.t
- r Krr Safety Factor 130 Temperature Adjustment =
- -'i 5 0 '
-F Pressure Adjustment =
24.0.'.
'1/4 psig Hydro Test Pressure 563' psig Flange RTNDa =
20.0
'F 2.666 4inch Pressure Saturation P
Temperature (ps1g)
(F) 0 212.1 200.0 387.9 210.0 391.8 220.0 395.6 230.0 399.3 240.0 402.8 250.0 406.2 260.0 409.6 270.0 412.8 331.0 430.9 335.8 432.2 336.0 432.3 336.5 432.4 336.8 432.5 396.8 447.8 456.8 461.4 516.8 473.7 576.8 485.0 640.0 496.0 670.0 500.9 730.0 510.2 734.0 510.8 794.0 519.6 854.0 527.8 914.0 535.6 974.0 543.1 1034.0 550.2 1094.0 556.9 1154.0 563.4 1214.0 569.7 1274.0 575.7 1334.0 581.5 1394.0 587.1 1454.0 592.5 1514.0 597.7 1574.0 602.8 1634.0 607.7 Ka (ksi'inch'2) 19.6 39.7 40.1 40.6 41.0 41.4 41.8 42.2 42.5 44.6 44.8 44.8 44.8 44.8 46.5 48.1 49.5 50.8 52.0 52.6 53.7 53.7 54.7 55.7 56.6 57.4 54.7 51.9 49.0 46.2 43.3 40.5 37.7 34.8 32.0 29.2 26.3 K,.
(ksl'inch" 2 )
0.0 15.5 16.3 17.1 17.9 18.6 19.4 20.2 21.0 25.7 26.1 26.1 26.1 26.1 30.8 35.5 40.1 44.8 49.7 52.0 56.7 57.0 61.6 66.3 71.0 75.6 80.3 84.9 89.6 94.2 98.9 103.6 108.2 112.9 117.5 122.2 126.8 (ksi'inch 1 *)
19.6 70.8 72.8 74.7 76.7 78.7 80.6 82.5 84.5 96.0 96.9 96.9 97.0 97.1 108.1 119.0 129.7 140.3 151.4 156.6 167.0 167.7 178.0 188.3 198.5 208.6 215.2 221.7 228.2 234.7 241.2 247.6 254.1 260.6 267.1 273.5 280.0 Calculated Temperature T
49.7 52.3 54.7 57.1 59.3 61.4 63.3 65.3 75.4 76.1 76.1 76.2 76.3 84.2 91.0 96.9 102.1 107.0 109.2 113.2 113.5 117.2 120.6 123.8 126.8 128.6 130.4 132.1 133.7 135.3 136.8 138.3 139 7 141.1 142.5 143.8 Adjusted Temperature for P-T Curve
('F) 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 142 143 145 146 148 149 Adjusted Pressure for P-T Curve (psig) 0 176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 616 646 706 710 770 830 890 950 1010 1070 1130 1190 1250 1310 1370 1430 1490 1550 1610 16 Structural Integrity Associates, Inc.
Attachment to SIR-03.074, Rcv. 1/CRL-03-004
Table 8 Tabulated Values for Core Critical Curve (Curve C) for 32 EFPY Pressure-Temperature Curve Calculation (Core Critical = Curve C)
Inputs:
Plant=- Cooper EFPY=
.32 Curve A Leak Test Temperature =
- 45.0.o
- F (at 1t,100 psig)
Hydro Test Pressure =
20.0
'F Beltine Curve B Curve B Temperature Pressure Bottom Head Upper Vessel Composite (
Curve B Curve B Curve B Curve B Minimum Temperature Pressure Temperature Pressure Temperature Curve B Composite Curve C Minimum Pressure Temperature Pressure (8F) 80 80 82 87 92 97 102 107 112 117 122 127 132 137 142 147 152 157 162 167 172 177 182 187 192 197 202 207 212 217 222 i2gL 0
341 345 354 364 376 389 403 418 435 454 475 498 524 552 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 (8F) 80 80 82 86 90 94 98 102 106 110 114 118 122 126 130 134 138 142 146 150 154 158 Irsig) 448 461 488 517 548 582 619 659 702 749 800 855 915 979 1049 1125 1207 1296 1393 1497 1610 80F) 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 142 143 145 146 148 149 (psig) 0 176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 616 646 706 710 770 830 890 950 1010 1070 1130 1190 1250 1310 1370 1430 1490 1550 1610 (8F) 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 147 152 157 162 167 172 177 182 187 192 197 202 207 212 217 222 (psA 0
176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 (Ff) 80 80 80 80 80 80 80 80 80 80 80 80 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 (psig) 0 176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 17 Strlidural Infegrity Associates, Inc.
Attachment to SIR-03-074, Rev. I/CRL-03-004
Figure I Pressure Test P-T Curve (Curve A) for 32 EFPY Cooper Pressure Test Curve (Curve A), 32 EFPY 1,600 1,500 1,400 1,300 il l
iI
! II I1ic=4-S
-a 1,200
-a n 1,100 o
1,000
--j lii CO 900 LU Q:
0 800 cc 700 I-600
-j us So Cm LU Q-400 1-1-
300 200 100 0
Boltup=
80 0 F I
I
- : z :
- : : t =
Beltline
- Bottom Head Upper Vessel 1-1 i
0 20 40 60 80 100 120 140 160 180 200 220 240 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Attachment to STR-03-074, Rev. I/CRL-03-0(4 18 Structural Integrity Associates, Inc.
Figure 2 Core Not Critical Curve (Curve B) for 32 EFPY Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,600 1,500 1,400 1,300 1,200 ar 1,100 0
1.000
-I cos 900 Lu cz o
800 F-E 700 00
=
500 U) 400 300 200 100 0
I-1 1 4-1 1 1--V-i t+/-+/-----
-4A- -
-f---+/-A 1---+---*-i-4--1+++-+FF-f---1-44---I-*-+
.---- i I
t----- -----t--- -
a- -u
- --- -f
~Beltline 1_
I I
I 1 Boln~tin Bottom Head Upper Vessel
-l I
I 0
50 100 I50 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F) 19 C
Structural Integrity Associates, Inc.
Attachment to SIR-03-074, Rev. I/CRL-03-004
Figure 3 Core Critical Curve (Curve C) for 32 EFPY Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,600 1,500 1,400 1,300 c
1,200 1,100 o 1,000 co Xn 900 tu o
800 M
700 E
600 tu g
500 en v) cr 400 300 200 100 0
- Composil i
i Curve C e B I
-1.-
I -- I -1 EiEiEiEl I I I Minimum Criticality with Normal Water Level 80°F i_ i I
iI 0
50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)
Attachment to SIR-03-074, Rev. I/CRL-03-004 20 V
Structural Integrity Associates, Inc.
FILE No.: COOP-05Q-302 l
q2 STRUCT URL~
CALCULATION INTEGRITY PACKAGE PROJECT No.: COOP-05Q Associates, Inc.
PROJECT NAN1ME: Develop Revised Pressure-Temperature Curves at Cooper Nuclear Station CLIENT: Nebraska Public Powver District (NPPD)
CALCULATI ON l ITLE: Development of Updated Heatup/Cooldown (Curves B & C) P-T Curves Project Mgr.
Preparer(s) &
Document Affected Revision Description Approval Checker(s)
Revision PSages Signature &
Signatures &
Date Date A
1-21 Draft Issue G. L. Stevens C. R. Limpus 06/05/03 06/03/03 Al -A4 In Computer B. P. Templeton Files 06/03/03
° 1-21 Initial Issue G. L. Stevens C. R. Limpus Al -A4 08/11/03 08/11/03 B. P. Templeton In 08/11/03 Computer Files 1
1-21 Revised to correct closure flange G. L. Stevens C. R. Limpus RTNDT and use ART values of record.
A Pj C
Al -A4 6/
In 10/15/03 0o /., 5lo-3 Computer B./P.
3 io/
5/
FilesB.P epto Page I of 21 F2001R1
Tal)lc of Contents
1.0 INTRODUCTION
/ OBJECTIVE...........................................................
3 2.0 CURVE DEVELOPMENT...........................................................
3 2.1 CURVE B (HJEATUP/COOLDOWN. CORE NOT CRITICAL)..............................................
3 2.1.1 Beltlinc Region.............................................................
3 2.1.2 Bottom Head Region............................................................
7 2.1.3 F-eedwater Nlozzle/Upper V4essel Region............................................................
10 2.2 CURVE C (HEATUP/COOLDONVN, CORE CRITICAL).
17
3.0 REFERENCES
20 APPENDIX A...........................................................
I CURVE FIT OF SATURATION TEMPERATURE VS. PRESSURE.................................................... I List of Tables Table 1: Beltline Curve B for 32 EFPY............................................................
6 Table 2: Bottom Head Curve B for All EFPY............................................................
9 Table 3: Upper Vessel Curve B for All EFPY...........................................................
14 Table 4: Curve C for 32 EFPY...........................................................
1 8 List of Figures Figure 1: 1-leatup/Cooldown, Core Not Critical P-T Curve (Curve B) for 32 EFPY............................. 16 Figure 2: Heatup/Cooldown. Core Critical P-T (Curve C) for 32 EFPY............................................... 1 9 Revision A
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1.0 INTRODUCTION
/ OBJECTIVE This calculation updates the Cooper Nuclear Station (CNS) pressure-temperature (P-T) curves for the beldine, bottom head, and upper vessel regions. Ileatup/cooldoxnv curves are developed for 32 effective full power years (EFPY). All curves are developed using methods in the 1995 Edition, 1996 Addenda of ASME Code.Section XI, Appendix G [3]. incorporating the methods specified in ASME Code Case N-640 [41.
2.0 CURVE DElVEILOPMENT In this section, the methodology for calculating the heatup/cooldown P-T curves is detailed. This methodology describes the equations used by the EXCEL spreadsheet (CURVE-BCrI.XLS) developed for the heatup/cooldown curves for core not critical and core critical operation, also known as Curves B and C, respectively.
2.1 CURVE B (IIEATUIP/COOlI)OWNN, CORE, NOT CRITICAL)
There are three regions that are evaluated: (I) tile beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. The methodology used to calculate the licatup/cooldown P-T curves for each of these regions is summarized in the following subsections:
2.1.1 Belflilne Region
- a.
Assume a fluid temperature, T. The temperature at the assumed flaw tip, T1/4t (i.e., 1/4t into the vessel wall), is conservatively assumed to be zero and the metal temperature is assumed equivalent to the fluid temperature. The plant's temperature gauge has an uncertainty of +/-
5F [17]. A temperature 5' less than the initially assumned fluid temperature is used to calculate the allowable pressure, producing a lower allowable pressure, and therefore making the calculation conservative.
- b.
Calculate the allowable stress intensity factor, using KIc as per Code Case N-640 [4], and T114t as follows:
KXC = 20.734 eI002(Tit 4c44RT)1 + 33.2 (eqn. from Ref. 12])
wvhere: T1/4,
=
metal temperature at assumed flaw tip (fF)
= adjusted reference temperature for location under consideration and desired EFPY (fF)
Kic
= allowable stress intensity factor (ksi'inch)
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Updated ART values for all reactor pressure vessel materials (plates, forgings, and welds) are provided in Table I of Reference [12]. The limiting beltline material is the Lower-Intermediate Longitudinal Weld, having the following ART value:
- c.
Calculate the thermal stress intensity 'factor, KIT, for a cooldown as follows, from lj2214.3 of Reference [3]:
KIT = 0.953 x 10-3 CR t25 where:
KIT
=
CR
=
=
thermal stress intensity factor (ksi'Iinch) 0.953 for postulated axial or circ inside surface defect of a cooldown (0.753 for postulated axial or circ outside surface defect of a heatup) heatup/cooldown rate (0F/hr) 100 [5]
vessel wall thickness excluding clad (inches) 5.875 [8]
P-T curves for the heatup and cooldown conditions apply at a given EFPY for both 1/4t and 3/41 locations, thereby evaluating stresses at the respective flaw locations. For conservative simplification, the thermal gradient stress at the 1/14 location is assumed tensile for both heatup and cooldown, -which results in applying the maximum tensile stress at the 1/4t location. It is conservative for three reasons:
- 1) The maximum stress is used regardless of flaw location,
- 2) Irradiation effects cause the allowable toughness, KIc, at 1/41 to be less than that at 3/4t for a given metal temperature, and
- 3) No credit was taken for the lower stress for heatup per the equation above.
Note that the BWR metal temperature is always limited by steam saturation conditions during operation.
- d.
Calculate the allowable pressure stress intensity factor, Kjp, using the following relationship for a heatup/cooldowvn P-T curve 13]:
=KC - KIT SF where:
Kip
=
allowvable pressure stress intensity factor (ksiNlinch)
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SF 2
- e.
Compute the allowable pressure, P. The relationship for the pressure, P, to the allowable pressure stress intensity factor, Kjp, is as follows (where bending stresses are conservatively excluded and all stresses are treated as membrane) 13]:
Kip = MmGm where:
am membrane stress due to pressure (ksi) = PR/t P
pressure (ksi)
R vessel radius (inches)
=
110.375 inches from 18]
t vessel wvall thickness (inches)
=
5.875 inches from [8]
Mm membrane stress correction factor
=
0.926'It for 4t = 2.424 2.24 Thuls.
P = Kipt
- f.
Apply any applicable adjustments for temperature and/or pressure to T and P. respectively.
The temperature adjustment is 50F as discussed above per Reference [17]. The pressure adjustment is 19.9 psig to account for the hydrostatic pressure of at normal vessel water level (wvater height = 551.75" [1] at room temperature, p = 62.4 lb/ft3, so AP =
(62.4*551.75)/1728), and minus 24.0 psig (2% of 1,200 psig) [17] to account for the gauge's uncertainty. Subtracting the gauge's uncertainty produces a lower allowable pressure, making it the most conservative method.
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.
The resulting pressure and temperature series constitutes the P-T curve. The P-T curve relates the minimum required fluid temperature to the reactor pressure. The resulting P-T curves for the beltline region are generated from Table 1.
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Table 1: Beltline Curve 1B for 32 EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)
Inpu ts:
Plant =,
73 Component =
Vessel thickness. t =,,,
inches. so -4t =
2.424 Nfinch Vessel Radius. R
'.110 375,'j; inches ART F=====>
32 EFP, S
Cooldown Rate. CR =
'F/hr Krr =
7j4, ksi-inch"'
AT 114, =
OO
°F Conservatively assumed zero Safety Factor -Y2O.
0 Temperature Adjustment =
'5,0, i
F (Instrument Uncertainty) not applied Height of Water for a Full Vessel = jtZ5 5 i47 5, inches Pressure Adjustment psig (hydrostatic pressure for a full vessel at 70°F)
Pressure Adjustment =.
psig (Instrument Uncertainty)
Hydro Test Pressure psig Flange RTNDT =.
F Gauge Calculated Adjusted Fluid 1/4t Pressure Temperature Pressure for Temperature Temperature K~c Kgp P
(°F)
(ksilInch"2 )
(ksi-inch 1 12 )
(Psig)
(°F)
(psig) 80.0 75.0 40.44 16.23 0
80.0 0
80.0 75.0 40.44 16.23 385 80.0 341 82.0 77.0 40.74 16.38 388 82.0 345 87.0 82.0 41.53 16.78 398 87.0 354 92.0 87.0 42.41 17.22 408 92.0 364 97.0 92.0 43.37 17.70 420 97.0 376 102.0 97.0 44.44 18.24 432 102.0 389 107.0 102.0 45.63 18.83 446 107.0 403 112.0 107.0 46.93 19.48 462 112.0 418 117.0 112.0 48.38 20.20 479 117.0 435 122.0 117.0 49.97 21.00 498 122.0 454 127.0 122.0 51.74 21.88 519 127.0 475 132.0 127.0 53.69 22.86 542 132.0 498 137.0 132.0 55.84 23.93 568 137.0 524 142.0 137.0 58.22 25.12 596 142.0 552 147.0 142.0 60.85 26.44 627 147.0 583 152.0 147.0 63.76 27.89 662 152.0 618 157.0 152.0 66.98 29.50 700 157.0 656 162.0 157.0 70.53 31.28 742 162.0 698 167.0 162.0 74.46 33.24 788 167.0 744 172.0 167.0 78.79 35.41 840 172.0 796 177.0 172.0 83.59 37.81 897 177.0 853 182.0 177.0 88.89 40.46 959 182.0 916 187.0 182.0 94.75 43.39 1029 187.0 985 192.0 187.0 101.22 46.62 1106 192.0 1,062 197.0 192.0 108.37 50.20 1190 197.0 1,147 202.0 197.0 116.28 54.15 1284 202.0 1.240 207.0 202.0 125.01 58.52 1388 207.0 1,344 212.0 207.0 134.67 63.35 1502 212.0 1.458 217.0 212.0 145.34 68.69 1629 217.0 1.585 222.0 217.0 157.14 74.58 1769 222.0 1.725
2.1.2 Bottoi Head Regioni The bottom head region calculations are the same as for the beitline region, except that the equation for pressure stress in a spherical shell is substituted for the cylindrical pressure stress equation and a stress intensity factor is used to account for the CRD penetrations. Thus:
- a.
Same as Step (a) for the beltline region.
- b.
Same as Step (b) for the beltline region, except that ART values for the lower head region are used. From Table I of Reference 112], the limiting bottom head material is the bottom head torus plates, having bounding RTNDT value of 280F. The bottom head region is not affected by fluence, so this value is also valid for ART.
- c.
Same as Step (c) for the beltline region.
- d.
Same as Step (d) for the beltline region.
- e.
Compute the allowable pressure, P. The relationship for the pressure. P. to the allowable pressure stress intensity factor, Kip. is as follows [3]:
KIp = MmOm + Mbab where:
(TM membrane stress due to pressure (ksi) = 3PR/(2t) assuming a stress concentration factor of 3 for the bottom head penetrations, a standard of GE RPVs.
P
=
pressure (ksi)
R
=
vessel radius (inches)
=
110.375 inches from 18]
t
=
vessel wall thickness (inches)
=
3.1875 inches from [ 1 3]
Mm
=
membrane stress correction factor
=
0.9264t for t = 2.424
=
2.24 C71b
=
bending stress due to pressure
=
0 for a thin valled vessel
- Thus, P
2K1pt 2.3RMm Revision A
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- f.
Same as Step (O) for the beltline region.
- g.
Repeat steps (a) through (1) for other temperatures to generate a series of P-T points.
The resulting P-T curve for the bottom head region is tabulated in Table 2.
A 0
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Taiblc 2: Bottom Ilead Curve B for All EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)
Inpujts:
Plant C " -
Component = Bottom,Head: (Pen Vessel thickness,
=
3 -875; i nchi Vessel Radius, R =
ii I 75'.inchi ART =
-R 8.
F Safety Factor =
-.)2.00.
Stress Concentration Factor =
TenprtreAjsm Botki Cooldown Rate, CR = <;lO*O ZrfP 'F/hi M
=
KIT 7
- 3.-
Iksitii Temperature Adjustment =..
';1~
F. Ir Height of Water for a Full Vessel =
- 5.
J5, inch(
Pressure Adjustment 9.9 j* psig Pressure Adjustment -
psig Unit Pressure - g,1,563 Ž psig Flange RTt.T i
F netrations Portion) es, so qt =
1.785 as
=====>
'._iiEf g
..2.-t 4inch lm Head Penetrations I
jichliz Istrument Uncertainty es (normal full vessel water level)
(full vessel at 70'F)
(Instrument Uncertainty)
Gauge Temperature T
(F)
Adjusted Tem perature (F)
Calculated Pressure P
(psin)
K,c (ksi-inch"11)
Kip (ksi-inch"'
Temperature for P-T Curve
('F)
Adjusted Pressure for P-T Curve (psig) 80.0 80.0 82.0 86.0 90.0 94.0 98.0 102.0 106.0 110.0 114.0 118.0 122.0 126.0 130.0 134.0 138.0 142.0 146.0 150.0 154.0 158.0 75.0 75.0 77.0 81.0 85.0 89.0 93.0 97.0 101.0 105.0 109.0 113.0 117.0 121.0 125.0 129.0 133.0 137.0 141.0 145.0 149.0 153.0 86.28 86.28 88.44 93.05 98.03 103.43 109.28 115.62 122.48 129.92 137.97 146.70 156.15 166.39 177.48 189.50 202.52 216.62 231.90 248.44 266.37 285.79 42.27 42.27 43.36 45.66 48.15 50.85 53.78 56.94 60.38 64.09 68.12 72.48 77.21 82.33 87.88 93.89 100.39 107.45 115.08 123.36 132.32 142.03 0
492 505 532 561 592 626 663 703 746 793 844 899 959 1023 1093 1169 1251 1340 1437 1541 1654 80 80 82 86 90 94 98 102 106 110 114 118 122 126 130 134 138 142 146 150 154 158 0
448 461 488 517 548 582 619 659 702 749 800 855 915 979 1049 1125 1207 1296 1393 1497 1610
2.1.3 Feedt,'aier ANozzle/Uppjwer Vessel Region The feedwater nozzle is selected to represent non-beltline/upper vessel components for fracture toughness analyses because the stress conditions at this location are the most severe that the vessel experiences. In addition to the more severe pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences thermal cycling due to relatively cold feedwater flow in hotter vessel coolant.
The methodology used for the feedwater nozzle is contained in WNIRC-175 [7]. Thus:
- a.
Assume a fluid pressure, P. The pressure used is the plant's gauge pressure, which has an uncertainty of +/- 24.0 psig (2% of 1,200 psig) [17]. A pressure 24.0 psig more than that assumed is used for calculating the temperature, producing a higher calculated temperature, therefore making the calculation conservative.
- b.
Calculate the thermal stress intensity factor, KIT, based on finite element stresses. The feedwater nozzle stresses were obtained from the finite element analysis results contained in Reference [8]. The maximum primary + secondary stress intensity ranges for (membrane and membrane + bending) thermal stresses were obtained for the entire design basis transients to encompass Normal and Upset Conditions. These stresses are shown below for the path defined by Elements 490 and 552 (nozzle bore blend radius location) in Reference [8, Figure S-I]:
Thermal (secondary) membrane + bending stress, (Usm+ Gsb):
The transient maximum thermal plus mechanical stress ranges were reported in [8] for elements 490 to 552. The total mechanical (membrane + bending) stresses are small in comparison to the thermal (membrane + bending) stresses and will remain included.
(Osm + Gsb) for 552 = 24.3 ksi per Table S-4 (Osm + GOb) for 490 = 21.5 ksi per Table S-4 The maximum of the elements above is used to bound this analysis.
Thermal (secondary) membrane stress, Osm:
asm for 552 = 4.288 - 0.0 ksi per Transient 15 [p. S-73, 8] - Transient I (Zeroload) asm for 490 = 4.680 - (-1.1215) = 5.895 ksi per Transient 17 [p. S-79] - 19 [p. S-84]
Note that the above stresses include pressure stress. Thus, the limiting condition is:
Maximum secondary bending stress. aCb = 24.3 - 5.895 = 18.405 ksi Secondary membrane stress, 0 sm 5.895 ksi Revision A
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t = 7.108 inches 19]
oay = 43.975 ksi @ 575°F for SA-508 Class 2 [14]
The value of Mm from G-2214.1 [3] is based on a thickness of 7.108 inches, so t"n = 2.666 and the resulting M.m value is 2.47.
Kim is calculated from the equation in Paragraph G-2214.1 [3]:
Kim = Mm* 0sm K1b is calculated from the equation in Paragraph G-2214.2 [3]:
KIb = (2/3) Mm* Gsb The total KIT is therefore:
KIT = R*SF*(Km+ Kib) where: R =
correction factor, calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WNIRC Bulletin 175 17]
=
bays - Gpm + ((Uaotal
- (y,) / 30)] / (aotali.
pm)
SF =
safety factor for KIT
=
1.3 (conservatively used based on the recommendation in WRC-175 [7])
- c.
Compute the allowable pressure stress intensity factor, Kjp, as follows:
K 1P =F(a/r.)VHa [rm where:
ri
=
actual inner radius of nozzle = 6.6875 inches 18]
7c nozzle comer radius = 5.0 inches [8]
- r.
apparent radius of nozzle = ri + 0.29r, = 8.1375 to
=
nozzle comer thickness
=
7.108 inches [9]
a
=
crack depth (inches)
=
1/4 t'= 1.777 inches F (a,r.)=
nozzle stress factor
=
1.6 for alrn = 0.22 from Figure A5-1 of [7]
Kip
=
allowable pressure stress intensity factor (ksi4inch).
UPm primary membrane stress. PR/t (primary bending stresses are conservatively treated as membrane stresses, so aryp = 0)
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R I
=
vessel radius (inches) = 110.375 inches [8]
=
vessel wall thickness (inches) = 5.875 inches [8]
- d.
Calculate the allowable stress intensity factor, KIc, using the following relationship for a heattup/cooldown P-T curve:
K Kic -KIT 2.0 thus:
K1C = 2.0OK11
+ IKIT
- c.
Calculate the temperature T1141 based on the allowable stress intensity factor, KIc as per Code Case N-640 [4], as follows:
K=
20.734 eJOo2(Td,-sAR)1 + 33.2 (eqn. from Ref. [2])
where: TI/41
=
metal temperature at assumed flaw tip (0F),
assumed equal to T. the temperature at the inner vessel wvall ART
=
adjusted reference temperature for location under consideration and desired E-FPY (0F)
Kic
= allowable stress intensity factor (ksi1inch) thus:
T = S0 *LN[KIC 33.2 + ART
-t4
[~20.734J
- f.
The curve was generated by scaling the stresses used to determine K1; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 407F water injected into the hot reactor vessel nozzle.
The normal pressure is 1038 psig [1 5] (power uprate) and the hot reactor vessel temperature is 551OF [15]. Since the reactor vessel temperature follows tle saturation temperature curve [App. A], the secondary stresses are scaled by (Ts3fur;aion - 40)/
(551 - 40).
- g.
Repeat steps (a) through (f) for other pressures to generate a series of P-T points.
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Two additional requirements were used to define the lower portion of the upper vessel P-T curve.
These limits are established by the discontinuity regions of the vessel (i.e., flanges), and are specified in Table I of Reference [10]:
If the calculated pressure, P. is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDT of the limiting flange material + 1200F. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 20'F [5].
If the calculated pressure, P, is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is typically greater than or equal to the RTNDT of the limiting flange material. GE typically applies an additional 60'F margin to the RTNDT value and has been a standard recommendation for the BWR industry for non-ductile failure protection. For the Cooper flange material, this minimum would be 807F (i.e., 20 +
60'F). Since the 607F margin is only a recommendation, the minimum temperature for Cooper was set to 80'F to be consistent with past work as well as adequately encompass instrument uncertainty. The gauge's uncertainty of +/- 5T is not applied here since the included margin just described adequately encompasses instrument uncertainty.
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Table 3: Upper Vessel Ctrve B for All EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve 8)
Inputs:
Plant =
Component =
RTNor =
Op.'. =
opb =
on, =
O.b Ory*
Nozzle corner thickness. t =
a (1/4 t')
F(alr,) =
KT Safety Factor =
Temperature Adjustment =
Pressure Adjustment Hydro Test Pressure =
Pressure Saturation P
Temperature KI (Psiq)
(-F)
(ksilinch"2) ksi for a pressure of 1.038 psig ksi for a pressure of 1,038 psig ksi for a temperature of 551'F ksi for a temperature of 551'F ksi @? 575'F inches. approximate 4t' =
2.666 'Jinch Calculated Temperature K,
Kv,,W.-bl T
(ksi-inch"7) lksilinch 112)
('F)
Adjusted Temperature for P-T Curve
(.F)
Adjusted Pressure for P-T Curve (psiq) 0 200.0 210.0 220.0 230.0 240.0 250.0 260.0 270.0 331.0 335.8 336.0 336.5 336.8 396.8 456.8 516.8 576.8 640.0 670.0 730.0 734.0 794.0 854.0 914.0 974.0 1034.0 1094.0 1154.0 1214.0 1274.0 1334.0 1394.0 1454.0 1514.0 1574.0 1634.0 212.1 387.9 391.8 395.6 399.3 402.8 406.2 409.6 412.8 430.9 432.2 432.3 432.4 432.5 447.8 461.4 473.7 485.0 496.0 500.9 510.2 510.8 519.6 527.8 535.6 5.43.1 550.2 556.9 563.4 569.7 575.7 581.5 587.1 592.5 597.7 602.8 607.7 19.6 39.7 40.1 40.6 41.0 41.4 41.8 42.2 42.5 44.6 44.8 44.8 44.8 44.8 46.5 48.1 49.5 50.8 52.0 52.6 53.7 53.7 54.7 55.7 56.6 57.4 54.7 51.9 49.0 46.2 43.3 40.5 37.7 34.8 32.0 29 2 26.3 0.0 15.5 16.3 17.1 17.9 18.6 19.4 20.2 21.0 25.7 26.1 26.1 26.1 26.1 30.8 35.5 40.1 44.8 49.7 52.0 56.7 57.0 61.6 66.3 71.0 75.6 80.3 84.9 89.6 94.2 98.9 103.6 108.2 112.9 117.5 122.2 126.8 19.6 70.8 72.8 74.7 76.7 78.7 80.6 82.5 84.5 96.0 96.9 96.9 97.0 97.1 108.1 119.0 129.7 140.3 151.4 156.6 167.0 167.7 178.0 188.3 198.5 208.6 215.2 221.7 228.2 234.7 241.2 247.6 254.1 260.6 267.1 273.5 280.0 49.7 52.3 54.7 57.1 59.3 61.4 63.3 65.3 75.4 76.1 76.1 76.2 76.3 84 2 91.0 96.9 102.1 107.0 109.2 113.2 113.5 117.2 120.6 123.8 126.8 128.6 130.4 132.1 133.7 135.3 136.8 138.3 139.7 141.1 142.5 143.8 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 142 143 145 146 148 149 0
176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 616 646 706 710 770 830 890 950 1010 1070 1130 1190 1250 1310 1370 1430 1490 1550 1610 A
1 0 1
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The resulting P-T curves are shown in Figure I for 32 EFPY. Note that only the beltline curve differs due to irradiation effects.
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Figure 1: lleatup/Cooldown, Core Not Critical P-T Curve (Curve 1B) for 32 EFPY Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,600
-f I
1,500 1,400 1,300
'a 1,200 2 1,100 6
1,000 U
900 800 0
700 z
n 600 500 400 300 200 I.
100 -
0 0
SO 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)
A 0
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2.2 CURVE C (IIEATUP/COOLDOWN, CORE CRITICAL)
Curve C, the core critical operation curve, is generated from the requirements of IOCFR50 Appendix G
[10]. Table I of Reference [10] requires that core critical P-T limits be 40F above any Curve A or B limits when pressure exceeds 20% of the pre-service system Hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40'F for pressures above 312 psig.
Table I of Reference [10] indicates that. for a BWR -with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 600F) at pressures below 312 psig (20% of pre-service hydro). This requirement makes the minimum criticality temperature 80'F for Cooper, based on an ART value of 207F for the flange region. In addition, above 312 psig, the Curve C temperature must be at least the greater of ART of the closure region + 1600F, or the temperature required for the hydrostatic pressure test (Curve A at I,100 psig).
Therefore, this requirement causes a temperature shift in Curve C at 312 psig.
The resulting Curve C P-T curve is tabulated in Table 5 for 32 EFPY. Tihe curve is graphically shown in Figure 2.
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Tablc 4: Cure C for 32 EFPY Pressure-Temperature Curve Calculation (Core Critical = Curve C)
Inputsr Plant =.
EFPY = I Curve A Leak Test Temperature = -
Hydro Test Pressure = 1 Flange RTNDT = :
Beltline Curve B Curve B Temperature Pressure Bottom Head Upper Vessel Composite C Curve B Curve B Curve B Curve B Minimum Temperature Pressure Temperature Pressure Temperature Curve B Composite Curve C Minimum Pressure Temperature Pressure F) 80 80 82 87 92 97 102 107 112 117 122 127 132 137 142 147 152 157 162 167 172 177 182 187 192 197 202 207 212 217 222 (psiq)
(F)
(psig)
('F)
(F)
(asiq)
I(F) ftsic?)
0 341 345 354 364 376 389 403 418 435 454 475 498 524 552 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 80 80 82 86 90 94 98 102 106 110 114 118 122 126 130 134 138 142 146 150 154 158 0
448 461 488 517 548 582 619 659 702 749 800 855 915 979 1049 1125 1207 1296 1393 1497 1610 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 140 142 143 145 146 148 149 0
176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 616 646 706 710 770 830 890 950 1010 1070 1130 1190 1250 1310 1370 1430 1490 1550 1610 80 80 80 80 80 80 80 80 80 80 80 80 80 140 140 140 140 140 147 152 157 162 167 172 177 182 187 192 197 202 207 212 217 222 0
176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 80 80 80 80 80 80 80 80 80 80 80 80 88 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 220 0
176 186 196 206 216 226 236 246 307 312 312 313 313 373 433 493 553 583 618 656 698 744 796 853 916 985 1062 1147 1240 1344 1458 1585 1725 Revision A
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Figure 2: lleatup/Cooldown, Core Critical P-T (Curve C) for 32 EFPY Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,600 -
1,500 -
1,400 1,300 -
B 1,200-1,100 1,000 0
900 800-700 z
5 600 4 500 400 i+ H1 :
1 h+/-H -
- Composite B
I
Curve C l l
- --- 300 -
200 100 -
0 0
50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F) 300 Revision A
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3.0 REFERENCES
- 1.
GE Drawing No. 729E479-B, Revision 0, "Reactor Primary SYS. WTS. & Vols.," Sheet I of 3, SI File No. NPPD-05Q-208.
- 2.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix A, "Analysis of Flaws," 1995 Edition, 1996 Addenda.
- 3.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1995 Edition, 1996 Addenda.
- 4.
ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves," Section Xl, Division 1, Approved February 26, 1999.
- 5.
GE Document No. GE-NE-523-159-1292 (DRF B 13-01662), "Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," Revision 0, February 1993, SI File No. COOP-05Q-202.
- 6.
Not Used.
- 7.
WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," PVRC Ad Hloc Group on Toughness Requirements, Welding Research Council, August 1972.
- 8.
CB1 Stress Report DC22A7245, Revision 0, "Feedwater Nozzle Modification Cooper RPV,"
3/20/80, SI File NPPD-I 3Q-205.
- 9.
SI Calculation NPPD-1 3Q-30], Revision 0, "Feedwater Nozzle Finite Element Model."
- 10.
U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," 1-1-98 Edition.
- 11.
USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 305-4), May 1988.
File No.
COOP-05Q-302
- 12.
SI Calculation No. COOP-05Q-301, Revision 1 A, "Development of Updated Pressure Test (Curve A) P-T Curves," October 2003.
- 13.
Combustion Engineering Drawing No. E-232-230, Revision 3, "General Arrangement Elevation for: General Electric Co. APED 218" I.D. BWR," SI File No. NPPD-06Q-208.
- 14.
ASME Boiler and Pressure Vessel Code, Section 111, Rules for Constnuction of Nuclear Power Plant Components, Division 1, Appendices, 1989 Edition.
- 15.
SI Calculation NPPD-13Q-302, Revision 0, "Feedwater Nozzle Stress Analysis."
- 16.
C-E Power Systems, Steam Tables, Reprinted from the 1967 ASME Steam Tables, 1967.
- 17.
NPPD Memo DED 2003-005, Alan Able to Ken Thomas, dated August 14, 2003, "Instrument Uncertainty Associated With Technical Specification 3.4.9," SI File No. COOP-05Q-203.
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APPENDIX A CURVE FIT OF SATURATION TEMPERATURE VS. PRESSURE Revision A
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The following curve fit relationship wvas derived for saturation temperature as a function of absolute pressure for water using the data in Reference [16]:
TSAT = 1 19.3* (0.7987)
- PSAT PSAT = RPVPRESS + 14.696 (OF)
(psia) wvhere:
PSAT TSAT
=
RPVPRESS
=
calculated saturation pressure (psia) calculated saturation temperature ( 0F) measured reactor pressure (psig) lThe input data, calculated data, and a plot showing the adequacy of this relationship is provided below.
Curve Fit for Saturated Steam Conditions 000 UM.
I-C an t X0.
I II II i
i I
I I
A SleamTable i
I-Cum Fh #1 I i
I lt 4 0
a 5W0 urn ISM Saluraflon Pressuie (psia) 21M 2.5W Revision A
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Curve Fit for Saturated Steam
Reference:
Steam Table data obtained from "Steam Tables, Properties of Saturated and Superheated Steam," CE Power Systems, 7th Printing.
Curve Fit: Tsat = 119.3(0.7 9 8 7)(1iPsal). Psat0 2198 Pressure Temperature Curve Fit Psat Tsat Tsat Difference Error (psia)
("F)
("F)
(OF)
(%)
14.696 15 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 280 290 300 350 400 212.00 213.03 227.96 250.34 267.25 281.02 292.71 302.93 312.04 320.28 327.82 334.79 341.27 347.33 353.04 358.43 363.55 368.42 373.08 377.53 381.80 385.91 389.88 393.70 397.39 400.97 404.44 407.80 411.07 414.25 417.35 431.73 444.60 212.10 213.13 227.89 250.07 266.89 280.62 292.32 302.55 311.69 319.96 327.54 334.54 341.06 347.16 352.91 358.34 363.49 368.40 373.08 377.57 381.87 386.01 390.00 393.84 397.56 401.16 404.65 408.03 411.32 414.51 417.62 432.06 444.97 0.10 0.10
-0.07
-0.27
-0.36
-0.40
-0.39
-0.38
-0.35
-0.32
-0.28
-0.25
-0.21
-0.17
-0.13
-0.09
-0.06
-0.02 0.00 0.04 0.07 0.10 0.12 0.14 0.17 0.19 0.21 0.23 0.25 0.26 0.27 0.33 0.37 0.05%
0.05%
-0.03%
-0.11%
-0.13%
-0.14%
-0.13%
-0.13%
-0.11%
-0.10%
-0.09%
-0.07%
-0.06%
-0.05%
-0.04%
-0.03%
-0.02%
-0.01%
0.00%
0.01%
0.02%
0.03%
0.03%
0.04%
0.04%
0.05%
0.05%
0.06%
0.06%
0.06%
0.07%
0.08%
0.08%
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450 500 550 600 650 700 750 800 850 900 950 1000 1050 1100 1150 1200 1250 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 1900 1950 2000 2100 2200 2300 2400 2500 456.28 467.01 476.94 486.20 494.89 503.08 510.84 518.21 525.24 531.95 538.39 544.58 550.53 556.28 561.82 567.19 572.38 577.42 582.32 587.07 591.70 596.20 600.59 604.87 609.05 613.13 617.12 621.02 624.83 628.56 632.22 635.80 642.76 649.45 655.89 662.11 668.11 456.67 467.39 477.30 486.54 495.19 503.33 511.03 518.34 525.31 531.95 538.32 544.43 550.31 555.97 561.43 566.71 571.83 576.78 581.59 586.26 590.80 595.22 599.53 603.73 607.83 611.84 615.75 619.58 623.32 626.99 630.58 634.10 640.94 647.53 653.89 660.04 665.99 Maximum =
Minimum =
Average =
Std. Deviation 0.39 0.38 0.36 0.34 0.30 0.25 0.19 0.13 0.07 0.00
-0.07
-0.15
-0.22
-0.31
-0.39
-0.48
-0.55
-0.64
-0.73
-0.81
-0.90
-0.98
-1.06
-1.14
-1.22
-1.29
-1.37
-1.44
-1.51
-1.57
-1.64
-1.70
-1.82
-1.92
-2.00
-2.07
-2.12 0.39
-2.12
-0.41 0.71 0.08%
0.08%
0.08%
0.07%
0.06%
0.05%
0.04%
0.03%
0.01%
0.00%
-0.01%
-0.03%
-0.04%
-0.06%
-0.07%
-0.08%
-0. 1 0%
-0.11%
-0.13%
-0.14%
-0.15%
-0.16%
-0.18%
-0.19%
-0.20%
-0.21%
-0.22%
-0.23%
-0.24%
-0.25%
-0.26%
-0.27%
-0.28%
-0.30%
-0.30%
-0.31%
-0.32%
0.08%
-0.32%
-0.07%
0.12%
A I
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CALCULATION FILE No.: COOP-05Q-301 STRUCTURAL CALCULATION INTEGRITY PACKAGE PR1OJECT No.: COOP-05Q Associates, Inc.
PROJEICT NAME: Develop Revised Pressure-Temperature Curves at Cooper Nuclear Station CLIENT: Nebraska Public Power District (NPPD)
CALCULATION TITLE: Development of Updated Pressure Test (Curve A) P-T Curves Project Mgr.
Preparer(s) &
Document Affected Approval Checker(s)
Revision Pages Revision Description Signature &
Signatures &
I)ate Date A
1-15 Draft Issue G. L. Stevens C. R. Limpus 06/05/03 06/03/03 In B. P. Templeton Computer 06/03/03 Files 0
1-21 Initial Issue G. L. Stevens C. R. Limpus 08/11/03 08/11/03 Computer B. P. Templeton Computer 08/11/03 Files 1-16 Revised to correct closure flange G. L. Stevens C. R. Limpus RTNDT and use ART values of record.
o Computer B. P. Tmlt Files B
p Page I of 16 F2001 RI
Table of Contents 1.0 2.0 3.0 3.1 3.2 3.3
4.0 INTRODUCTION
/ OBJECTIVE................................................
3 RTNDT DETERMINATION................................................
3 CURVE DEVELOPMENT...............................................
S5 Beltline Region................................................
5 Bottom Head Region................................................
8 Feedwater Nozzle/Upper Vessel Region................................................
1.1 REFERENCES
15 List of Tables Table 1: Cooper RPV Material ART 32 EIFPY Calculations.....................................................
4 Table 2: Beltline Curve A for 32 EFPY.....................................................
7 Table 3: Bottom Head Curne A for All EFPY....................................................
10 Table 4: Feedwater Nozzle/Upper Vessel Region Curve A for All EFPY............................................ 13 List of Figures Figure 1: Pressure Test P-T Curve (Curve A) for 32 EFPY.1.
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1.0 INTROI)UC TION / O1JECTIVE This calculation updates the Cooper Nuclear Station (CNS) pressure-temperature (P-T) curves for the beltline, bottom head, limiting flange and non-beltline locations (feedwater nozzle/upper vessel). The pressure test curves are developed for 32 effective fill power years (EPI'Y). P-T curves are developed using methods in the 1995 Edition, 1996 Addenda of ASME Code,Section XI, Appendix G [3],
incorporating the methods specified in ASME Code Case N-640 14].
2.0 RTNDT DIETERiMINATION Reference f5] provides RTNDT estimates for the Cooper reactor pressure vessel (RPV) materials in accordance with Regulatory Guide 1.99, Revision 2 (RG 1.99) 111]. An EXCEL spreadsheet (RTndt(Coop)rl.xls) was created to perform the RTNDT calculations for 32 EFPY. Table 7-2 of Reference [5] provides the calculated Adjusted Reference Temperature (ART) values of record produced by General Electric (GE) for Cooper Nuclear Station. A benchmark was performed to reproduce the GE ART values using the same methodology (Section 7.6.1) documented by GE [5].
Slight differences in the calculated ARTNDT and subsequently ART values were encountered for a few non-limiting materials due to round-off. The GE values for the limiting materials were successfully reproduced and will be used since they are the docketed numbers of CNS. Table I shows the RTNDT calculations for 32 EFPY.
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Tal)me 1: Cooper RPV M\\laterial ART 32 EFIlY Calculations Chemistry Chemistry Adjustments For 114t Component Part No.
Heat Initial RTw, Factor ARTNDt Margin Terms ART No.
I-F)
Cu (wt %)
Ni (wt %)
(F)
(F) a, (F) oc ('F)
(F)
Beltline Plates' G2803-1 C2274-1 14.0 0 20 0.68 153.00 87.5 17.0 0.0 118.5 Lower Shell G2803-2 C2307-1 0 0 0.21 0.73 162.80 93.1 17.0 0.0 110.1 Plates G2803-3 C2274-2
-8.0 0.20 0.68 153.00 87.5 17.0 0.0 96.5 Lower-G2801-7 C2407-1
-10.0 0.13 0.65 92.25 64 6 17.0 0.0 71 6 Intermediate Shell G2802-1 C2331.2 10.0 0.17 0.58 125.30 87.7 17.0 0.0 114.7 Plates G2802-2 C2307.2
-20.0 0 21 0.73 162.80 113.9 17.0 0 0 1109 Bettline Welds' Lower 2-233 12420
-50.0 0.22 1.02 234 50 133 0 28.0 0 0 111.0 Longitudinal Lower-Intermediate 1-233 27204
-50.0 0.19 0.97 215.65 149.6 28.0 0.0 127.6 Longitudinal LowermtoaLowert 1-240 21935
-500 0.20 0.69 17530 121.6 26.0 00 996 Non-Beltline Regions2 Closure Flange 20.0 20.0 Region Bottorn Head 28 0 28.0 R egion Fluence Wall lhickness fin.)
Peak Fluence Attenuation @
Peak Fluence @
Fluence Factor, FF Surveillance Location Full 114t at ID for 32 1/4t = el.2 1/4t for 32 FFPY 1
p26.010"0 Adjustment EFPY EOL I_____
(nlcm2)
Plates Welds Lower-Int. Shell 5.375 1.344 1.57E+18 0.724 1.14E*18 0443 1 2 57 1.57 Lower Shell 6.375 1.594 1 10E+18 0.682 7.52E#17 0362 Note 1: BelIline material input per Table 7-2 of Reference 15J Note 2: Non-beltline material input per Reference [S).-.
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3.0 CURVE )EVElLOPMElNT This section explains the methodology for calculating the pressure test P-T curves. This methodology describes the equations used by the EXCEL spreadsheet (CURVE-Ari.XlS) developed for the pressure test curves, also known as Curve A.
Three regions are evaluated: (1) the beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. The approach used for calculating the pressure test P-T curves for each of these regions is summarized in the following subsections.
3.1 Beltline Rcgion
- a.
Assume a fluid temperature, T. The temperature at the assumed flaw tip, T]/4z (i.e., 1/4t into the vessel wall), is equal to the fluid temperature, as the pressure test condition neglects any thermal effects (¶G-2215 [3]). The assumed temperature that is read on the plant's gauge has an uncertainty of +/- 5T [1 3]. A temperature 5T less than that assumed is used for calculating the allowable pressure. This method will produce a lower allowable pressure, making it conservative.
- b.
Calculate the allowable stress intensity factor, using KIc as per Code Case N-640 14],
and T,/41 as follows:
Ki,-
20.734 e10.2(T,,,,-ART)
+ 33.2 (eqn. from Ref. [2])
where: T,141
= metal temperature at assumed flaw tip (fF)
= adjusted reference temperature for location under consideration and desired EFPY (0f)
Kic
= allowable stress intensity factor (ksiz/inch)
RTNDT values for all reactor pressure vessel materials (plates, forgings, and welds) are provided in Table 1. The limiting beltline material is the Lower Longitudinal Weld, having the following ART value:
=
127.60F at 32 EFPY
- c.
Calculate the thermal stress intensity factor, KIT. Thermal stresses are neglected for pressure test conditions, so KIT = 0°
- d.
Calculate the allowable pressure stress intensity factor. Kip, using the followving relationship for a pressure test P-T curve [3]:
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K - = Krr KIc SF SF where:
Kip
=
allowable pressure stress intensity factor (ksi4inch)
SF
=
1.5
- c.
Compute the allowable pressure, P. The relationship for the pressure, P, to the allowable pressure stress intensity factor, K1r, is (where bending stresses are conservatively excluded and all stresses are treated as membrane) [3]:
KrP = MmCm wvhere:
lm membrane stress due to pressure (ksi) = PR/t P
=
pressure (ksi)
R
=
vessel radius (inches)
=
110.375 inches from [8]
t
=
vessel wvall thickness (inches)
=
5.875 inches from [8]
Mm
=
membrane stress correction factor
=
0.926'It for 'It = 2.424
=
2.24 RMm
- f.
Apply any applicable adjustments for temperature and/or pressure to T and P, respectively. The temperature adjustment was assumed to be 5F as discussed above.
The pressure adjustment is 19.9 psig to account for the hydrostatic pressure of at normal vessel wvater level (water height = 551.75" [1] at room temperature, p = 62.4 lb/ft3, so AP = (62.4*551.75)/1728), and minus 24.0 psig (2% of 1,200 psig) [13] to account for the gauge's uncertainty. Subtracting the gauge's uncertainty produces a lower allowable pressure, making it the most conservative method.
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.
The resulting pressure and temperature series constitutes the P-T curve. The P-T curve relates the minimum required fluid temperature to the reactor pressure. The resulting P-T curves for the beltline region are generated from Table 2.
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Trzblc 2: Belfline Curve A for 32 EFIPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
Inputs:
Plant =
'oopErT Component ='
BeItliri Vessel thickness, t =
i5875 4finches. so -It 2.424 4inch Vessel Radius, R = '?'1D 3759inches ART =
"F fl
=
Cooldown Rate, CR
'Fhr KIT -
o ksilinch"'
AT 1 41 =%jO.O J<,. 'F (no thermal for pressure test)
Safety Factor 1
' E (for pressure test)
Temperature Adjustment =. ?
F Height of Water for a Full Vessel 551 75g inches Pressure Adjustment psig (hydrostatic pressure for a full vessel at 70-F)
Pressure Adjustment -
psig (Instrument Uncertainty)
Hydro Test Pressure psig Flange RINOT =F
.J Gauge Fluid Temperature (F) 1/4t Temperature (1F)
Calculated Pressure P
(Psin)
Kic tksi-inch"'2)
Kip ksi-inch1 1 2 )
Temperature for P-T Curve (F)
Adjusted Pressure for P-T Curve (pslQ)
A A
B0.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0
.140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 200.0 80.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 41.20 40.44 40.74 41.04 41.36 41.70 42.04 42.41 42.78 43.17 43.58 44.00 45.14 46.39 47.78 49.32 51.01 52.88 54.95 57.24 59.77 62.56 65.65 69.07 72.84 77.01 81.61 86.70 92.33 98.55 105.42 113.02 27.47 26.96 27.16 27.36 27.58 27.80 28.03 28.27 28.52 28.78 29.05 29.33 30.09 30.93 31.85 32.88 34.01 35.26 36.64 38.16 39.85 41.71 43.77 46.04 48.56 51.34 54.41 57.80 61.55 65.70 70.28 75.35 0
641 645 650 655 661 666 672 678 684 690 697 715 735 757 781 808 838 871 907 947 991 1040 1094 1154 1220 1293 1374 1463 1561 1670 1790 80.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 175.0 180.0 185.0 190.0 195.0 200.0 0
597 601 606 611 617 622 628 634 640 646 653 671 691 713 737 764 794 827 863 903 947 996 1.050 1.110 1.176 1.249 1,330 1,419 1.517 1.626 1.746 Revision A
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3.2 Bottomn Iea(l Region The bottom head region calculations are the same as for the beltline region, except that the equation for pressure stress in a spherical shell is substituted for the cylindrical pressure stress equation and a stress concentration factor is used to account for the CRD penetrations. Thus:
- a.
Same as Step (a) for the beltline region.
- b.
Same as Step (b) for the beltline region, except that ART values for the lower head region are used. Per Reference [5], the limiting bottom head material is the bottom head torus plates, having bounding RTNDT value of 280F. The bottom head region is not affected by fluence, so this value is also valid for ART.
- c.
Same as Step (c) for the beltline region.
- d.
Same as Step (d) for the beltline region.
- e.
Compute the allowable pressure, P. The relationship for the pressure, P, to the allowable pressure stress intensity factor, Kip, is as follows:
Ki o= M (.m + MI bob where:
am membrane stress due to pressure (ksi) = 3PR/(2t) assuming a stress concentration factor of 3.0 for the bottom head penetrations. The value of 3.0 wvas arrived at based on SI experience with similar BOWR RPV parameters.
P
=
pressure (ksi)
R vessel radius (inches)
]1 l0.375 inches from [8], assumed from beltline region t
=
vessel wvall thickness (inches) 3.l875 inches [12]
Mm
=
membrane stress correction factor
=
2.24 for 4t = 2.424 (fib bending stress due to pressure
=
0 for a thin valled vessel 2Kt ThusP
= 2KIpt 2.3 RMm AI 0
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- f.
Same as Step (f) for the beltline region.
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.
The resulting P-T curve for the bottom head region is tabulated in Table 3.
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Table 3: Bottom Head Cunre A for All EFPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
Inputs:
Plant =-
0 i
Component =Botto Head,(Penetrations Portion)
Vessel thickness. I =, -
w-3 88tinches, so -t =
1.785
- linch Vessel Radius. R = ;-A;,.1iO.37,5:t., inches ART =
- F======>
Safety Factor = $'1 50.-
Safely Factor =
3 OP Bottom Head Penetrations Temperalure Adjustment =
5 F (Instrument Uncertainly)
H eight of Water for a Full Vessel = :
... i inches Pressure Adjustm ent = j',2: 9'9.,^:we psig (hydrostatic pressure for a full vessel at 70*F)
Pressure Adjustment =
psig (Instrument Uncertainty)
Unit Pressure = x71.563'g,,,
psig Flange RTNDI =
F Gauge Fluid Temperature
(.F )
114t Tem perature
(-F)
Kic (k s i inch' tZ)
Kip (k s iinc h"'2 )
Calculated Pressure P
(psig)
Tem perature for P-T Curve
(.F)
Adjusted Pressure for P-T Curve (psig) 80.0 80.0 82.0 84.0 86.0 88.0 90.0 92.0 94.0 96.0 98.0 100.0 102.0 104.0 106.0 108.0 110.0 112.0 114.0 116.0 1 18.0 120.0 122.0 124.0 126.0 128.0 130.0 132.0 134.0 136.0 138.0 140.0 142.0 144.0 146.0 148.0 80.0 75.0 77.0 79.0 81.0 83.0 85.0 87.0 89.0 91.0 93.0 95.0 97.0 99.0 1 01.0 103.0 1 05.0 107.0 1 09.0 111.0 113.0 115.0 117.0 119.0 121.0 123.0 125.0 127.0 1 29.0 131.0 133.0 135.0 137.0 139.0 141.0 143.0 91.86 86.28 88.44 90.70 93.05 95.49 98.03 100.68 103.43 106.30 109.28 112.38 115.62 118.98 122.48 126.12 129.92 133.86 137.97 142.25 146.70 151.33 156.15 161.17 166.39 171.83 177.48 183.37 189.50 195.88 202.52 209.43 216.62 224.10 231.90 240.00 61.24 57.52 58.96 60.47 62.03 63.66 65.35 67.12 68.95 70.86 72.85 74.92 77.08 79.32 81.65 84.08 86.61 89.24 91.98 94.83 97.80 100.89 104.10 107.44 110.93 114.55 118.32 122.25 126.33 130.59 135.01 139.62 144.41 149.40 1 54.60 160.00 0
599 614 629 646 662 680 698 718 737 758 780 802 825 850 875 901 929 957 987 1018 1050 1083 1118 t 154 1192 1231 1 272 1315 1359 1405 1453 1 503 1 555 1609 1665 80 80 82 84 86 88 90 92 94 96 98 100 102 104 106 108 110 112 114 116 118 120 122 124 126 128 130 132 134 136 138 140 142 144 146 148 0
555 570 585 602 619 636 655 674 694 714 736 758 782 806 831 857 885 913 943 974 1.006 1,039 1.074 1.110 1.148 1.187 1.228 1.271 1.315 1.361 1.409 1.459 1,511 1.565 1.621 Revision A
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3.3 Fec(lwiater Nozzle/Upper Vessel Region The feedwater nozzle is selected to represent non-beltline/upper vessel components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the more severe pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.
The methodology used for the feedwater nozzle is contained in WVRC-175 [7]. Thus:
- a.
Same as Step (a) for the beitline region.
- b.
Same as Step (b) for the beltline region, except that ART values for the feedwater nozzle/upper vessel region are used. From Reference [5], the closure flange region limiting RTNDT value is 207F. The feedwater nozzle/upper vessel region is not affected by fluence, so this value is valid for ART.
- c.
Same as Step (c) for the beltline region.
- d.
Same as Step (d) for the beltline region.
- e.
Compute the allowable pressure, P. The relationship for the pressure. P, to the allowable pressure stress intensity factor, Kp.. is as follows [7]:
Ki F(a/r )RjRUaP t
where:
r
=
actual inner radius of nozzle = 6.6875 inches [8]
rc
=
nozzle comer radius = 5.0 inches [8]
rn
=
apparent radius of nozzle = ri + 0.29r= 8.1375 nozzle comer thickness, approximate
=
7.108 inches [9]
a
=
crack depth (inches)
=
1/4 t'= 1.777 inches F(a,rn) =
nozzle stress factor
=
1.6 for a/rn = 0.22 from Figure A5-I of [7]
P
=
pressure (ksi)
R vessel radius (inches) = 110.375 inches [8]
t
=
vessel wall thickness (inches) = 5.875 inches [8]
Kip allottable pressure stress intensity factor (ksiinch)
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Thus.
P=
KUA F(a / r,,)R7n=,i
- f.
Same as Step (f) for the beltline region.
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.
Two additional requirements were used to define the lower portion of the upper vessel P-T curve.
These limits are established by the discontinuity regions of the vessel (i.e., flanges), and are specified in Table I of Reference [10]:
If the calculated pressure, P, is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than RTNDT of the limiting flange material + 90'F. The pre-service hydro test pressure was 1,563 psig, and the limiting flange material has an RTNDT of 20'F [5].
If the calculated pressure, P. is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is typically greater than or equal to the RTNDT of the limiting flange material. GE typically applies an additional 60°F margin to the RTNDT value and has been a standard recommendation for the BWVR industry for non-ductile failure protection. For the Cooper flange material, this minimum would be 80F (i.e., 20 +
60'F). Since the 60'F margin is only a recommendation. the minimum temperature for Cooper was set to 80'F to be consistent with past work as well as adequately encompass instrument uncertainty. The gauge's uncertainty of +/- 5F is not applied here since the included margin just described adequately encompasses instrument uncertainty.
The resulting P-T curve for the feedwater nozzle/upper vessel region is tabulated in Table 4.
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Table 4: Feedwivater Nozzle/Upper Vessel Region Curve A for All EFPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A)
InpoIs:
Plant =
Component = Udp.VseI (based on FW nozzle)
ART = sa id-i AF Vessel thickness, t =
.875 inches, so 4t=
2.424
'linch Vessel Radius, R =
10.375';.. inches Nozzle corner thickness, t' =
7.108,;
inches, approximate F(alrn) =-..
1 6
- -. nozzle stress factor Crack Depth, a =
f77 i inches Safety Factor 1.50 Temperature Adjustment =!
0F (not applied)
Height of Water for a Full Vessel =
55.7Sfi inches Pressure Adjustment =.
B y
psig (hydrostatic pressure for a full vessel at 700F)
Pressure Adjustment psig (Instrument Uncertainty)
Unit Pressure =
<563<:; psig Flange RTNDT 20.0;,-F Gauge Fluid Temperature
( 0F) 114t Temperature
( 0F)
Calculated Pressure p
(psig)
(ksilinch12)
(ksi-inch112)
Temperature for P-T Curve
('F)
Adjusted Pressure for P-T Curve (psig) 80.0 80.0 118.0 118.0 123.0 128.0 133.0 138.0 80.0 80.0 118.0 118.0 123.0 128.0 133.0 138.0 102.04 102.04 180.40 180.40 195.88 200.00 200.00 200.00 68.03 68.03 120.26 120.26 130.59 133.33 133.33 133.33 0
313 313 1693 1839 1877 1877 1877 80.0 80.0 123.0 123.0 128.0 133.0 138.0 143.0 2
269 269 1649 1795 1833 1833 1833 The resulting P-T curves are shown in Figure I for 32 EFPY. Note that only the beltline curve differs in the figure due to irradiation effects.
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Figure 1: Pressure Test P-T Curvc (Curve A) for 32 EFPY Cooper Pressure Test Curve (Curve A), 32 EFPY 1,600 1,500 1,400 1,300 i 1,200 i
1,100 1,000 0
900 800 E
700 z
5 600 400 I
I
___q
" _+_f__i -
i I
i I
i -l-A -
i
.- i _4---
-- i1 I I
I
-I--4 4
300-Boltup=
80 0 F 44i iz4:4~
I 1 1 1 1.
Beltline
- Bottom Head
-Upper Vessel 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF) 0 I
08/1 1/03 ] CR1. 10/15/03 OPT 08/11 1/03 I BPT 10/15/03
4.0 RE FERENCES
- 1. GE Drawing No. 729E479-B, Revision 0, "Reactor Primary SYS. WTS. & Vols.." Sheet I of 3, SI File No. NPPD-05Q-208.
- 2. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix A, "Analysis of Flaws," 1995 Edition, 1996 Addenda.
- 3. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Powver Plant Components, Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1995 Edition, 1996 Addenda.
- 4. ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,"Section XI, Division 1, Approved February 26, 1999.
- 5. GE Document No. GE-NE-523-159-1292 (DRF B I3-01662), "Cooper Nuclear Station Vessel Surveillance Materials Testing and Fracture Toughness Analysis," Revision 0, February 1993, SI File No. COOP-05Q-202.
- 6. Not Used.
- 7. WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials."
PVCR Ad Hoc Group on Toughness Requirements, Welding Research Council, August 1972.
3/20/80, SI File NPPD-I 3Q-205.
- 10. U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," 1-1-98 Edition.
I1. USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 305-4), May 1988.
- 12. Combustion Engineering Drawing No. E-232-230, Revision 3, "General Arrangement Elevation for: General Electric Co. APED 218" 1.D. BWR," SI File No. NPPD-06Q-208.
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COOP-05Q-301 Page 15 of 16
- 13. NPPD Memo DED 2003-005, Alan Able to Ken Thomas, dated August 14, 2003, "Instrument Uncertainty Associated With Technical Specification 3.4.9," SI File No. COOP-05Q-203.
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COOP-05Q-301 Page 16 of 16
I ATTACHMENT 3 LIST OF NRC COMMITMENTS©l Correspondence No: NLS2004005 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing
& Regulatory Affairs Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE None 1*
1-1-
t 4-4-
+
+
4-4-
PROCEDURE 0.42 1
REVISION 14 l
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