NLS2011015, License Amendment Request to Revise Technical Specification Pressure/Temperature Limit Curves and Surveillance Requirements
| ML11272A057 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/22/2011 |
| From: | O'Grady B Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2011015 | |
| Download: ML11272A057 (108) | |
Text
N Nebraska Public Power District Always there when you need us 50.90 NLS2011015 September 22, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
License Amendment Request to revise Technical Specification Pressure/
Temperature Limit Curves and Surveillance Requirements Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46
Dear Sir or Madam:
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 revises the curves in TS 3.4.9, "RCS Pressure and Temperature (P/T) Limits," to replace the 28 Effective Full Power Years (EFPY) restriction in TS Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 and revises minimum temperature in Surveillance Requirement (SR) 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7. The proposed amendment includes a set of updated P/T curves for pressure test, core not critical, and core critical conditions for 32 EFPY based on a fluence evaluation performed using Nuclear Regulatory Commission (NRC) approved fluence methodology. The new curves show a shift of minimum operating temperature which allows the bolt-up and minimum temperatures specified for SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 to be changed from 80'F to 70 0 F.
NPPD requests approval of the proposed amendment by September 21, 2012 which allows one year for NRC review. Once approved, the amendment will be implemented within 60 days in support of the next refueling outage. provides a description of the TS changes, the basis for the amendment, the no significant hazards consideration evaluation pursuant to 10 CFR 50.91 (a)(1), and the environmental consideration pursuant to 10 CFR 51.22. Attachment 2 provides the proposed changes to the current CNS TS in marked up format. Attachment 3 provides the final typed TS pages to be issued with the amendment. Attachment 4 provides conforming changes to the TS Bases for NRC information. CNS Calculation NEDC 07-048, "Revised Vessel Pressure Temperature Curves" is enclosed for information.
No formal licensee commitments are being made for this proposed amendment.
COOPER NUCLEAR STATION A P.O. Box 98 / Brovwnville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 wvw.nppd.com
NLS2011015 Page 2 of 2 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 238 issued July 27, 2011 have been incorporated into this request.
By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91(b)(1). Copies are also being provided to the NRC Region IV office and the CNS Senior Resident Inspector in accordance with 10 CFR 50.4(b)(1).
Should you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on , -
((D te)
Sincerely, Brian J. ry Vice President-Nuclear and Chief Nuclear Officer
/em Attachments Enclosure cc: Regional Administrator w/attachments & w/o enclosure USNRC - Region IV Senior Resident Inspector w/attachments & w/o enclosure USNRC - CNS Nebraska Health and Human Services w/ attachments & w/o enclosure Department of Regulation and Licensure Cooper Project Manager w/attachments & enclosure USNRC - NRR Project Directorate IV-1 NPG Distribution w/o attachments & w/o enclosure CNS Records w/attachments & enclosure
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© 4 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS©'
Correspondence Number: NLS2011015 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 Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None PROCEDURE 0.42 REVISION 27 PAGE 18 OF 25
NLS2011015 Page 1 of 11 Attachment 1 License Amendment Request to revise Technical Specification Pressure/Temperature Limit Curves and Surveillance Requirements Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46 1.0 Summary Description 2.0 Detailed Description 3.0 Technical Evaluation 3.1 Current Licensing Basis 3.2 Background 3.3 Technical Analysis 4.0 Regulatory Safety Analysis 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion 5.0 Environmental Consideration 6.0 References
NLS2011015 Page 2 of 11 1.0
SUMMARY
DESCRIPTION This is a request to amend the Cooper Nuclear Station (CNS) Facility Operating License DPR-46 Technical Specifications (TS) for Reactor Coolant System (RCS) pressure and temperature limits. The proposed changes revise the curves in TS 3.4.9, "RCS Pressure and Temperature (P/T) Limits," to replace the 28 Effective Full Power Years (EFPY) restriction in TS Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3. It also revises the minimum temperatures in Surveillance Requirement (SR) 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7. This change includes a set of updated P/T curves for pressure test, core not critical, and core critical conditions for 32 EFPY based on a fluence evaluation performed using Nuclear Regulatory Commission (NRC) approved fluence methodology. The new curves show a shift of minimum operating temperature which allows the bolt-up and minimum temperatures specified for SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 to be changed from 80'F to 70 0F.
This change is being proposed, because it resolves the limitation on the current P/T Limit Curves from Amendment 231 in which the NRC authorized their use through 28 EFPY.
The revised fluence includes the 1.62% Appendix K power uprate previously approved and 24-Month Cycle considerations.
CNS requests approval of the proposed amendment by September 21, 2012 which allows one year for NRC review. Once approved, the amendment will be implemented within 60 days in support of the next refueling outage.
2.0 DETAILED DESCRIPTION The following revisions are proposed to TS Section 3.4.9.
2.1 The requirement in SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 to verify reactor vessel flange and head flange temperatures are greater than 80'F will be changed from 80°F to 70OF to correspond to the revised minimum operating temperatures in P/T Limit Curves A and B. Notes in SR 3.4.9.6 and SR 3.4.9.7 for Mode 4 are also changed to reduce by 10°F the points at which the surveillances are required to be performed in order to maintain the same margins from the minimum operating parameters.
2.2 The proposed amendment will revise TS Figures 3.4.9-1, "Pressure/Temperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear Shutdown," 3.4.9-2, "Pressure/Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests," and 3.4.9-3, "Pressure/Temperature Limits for Criticality." This change replaces the 28 EFPY curves with new curves based on 32 EFPY, Appendix K power uprate, and a fluence evaluation performed by TransWare Enterprises Inc. and accepted by CNS using the NRC-approved Radiation Analysis Modeling Application (RAMA) fluence methodology.
NLS2011015 Page 3 of 11 2.3 Conforming TS Bases changes are provided in Attachment 4 for NRC information.
These Bases revisions will be made as an implementing action pursuant to TS 5.5.10, TS Bases Control Program, following issuance of the amendment.
3.0 TECHNICAL EVALUATION
3.1 Current Licensing Bases The reactor pressure vessel (RPV) is a vertical cylindrical pressure vessel with hemispherical heads of welded construction. The RPV is designed and fabricated for a useful life of 40 years based upon the specified design and operating conditions. It is designed, fabricated, inspected, tested, and stamped in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III (1965 Edition and January 1966 Addenda), its interpretations, and applicable requirements for Class A Vessels as defined therein.
RCS components are designed to withstand effects of cyclic loads due to system pressure and temperature changes introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. The P/T limit curves apply to the RPV, since it is the RCS component most subject to brittle failure, and is bounding over other components that comprise the reactor coolant pressure boundary (RCPB). TS 3.4.9 establishes operating limits that provide a margin to brittle failure of the RPV and piping of the RCPB.
The actual shift in the Reference Temperature of Nil-Ductility Transition (RTNDT) of the vessel material is established periodically by removing and evaluating the irradiated RPV material specimens, in accordance with Boiling Water Reactor Vessel and Internals Project (BWRVIP) document BWRVIP-86-A and Appendix H of 10 CFR 50. The operating P/T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.
The P/T limits are not derived from Design Basis Accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the RPV, a condition that is unanalyzed.
Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
NLS2011015 Page 4 of 11 3.2 Background Amendment 201 to the CNS Operating License, issued October 31, 2003, implemented the BWRVIP Reactor Pressure Vessel Integrated Surveillance Program (ISP). As part of the ISP, CNS vessel surveillance capsules are evaluated using fluence calculations that conform to RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."
The neutron fluence calculational methodology employs the RAMA fluence methodology to evaluate neutron flux through the core, vessel internals, and vessel geometry. The code uses the BUGLE-96 cross-section library to calculate the neutron transport and to determine the reaction-specific measured activities. The neutron transport calculation is 3-dimensional, rather than a synthesis of two 2-dimensional calculations used in the finite differences method. The neutron source is determined based on core power density and region-wise power distribution. The RAMA source accounts for the exposure dependence of the core neutron source and allows for a detailed pin power description of the source distribution. The physical geometry is represented without approximation. The RAMA code is applied in the reactor beltline region defined by the top and bottom planes of the active fuel and the inside surface of the biological shield.
The current CNS Heatup/Cooldown curves were, in fact, developed by Nebraska Public Power District (NPPD) for 32 EFPY in accordance with the requirements of 10 CFR 50 Appendix G. The methods in the 2001 Edition, 2003 Addenda of ASME Code,Section XI, Appendix G, were used. However, in License Amendment 231 the NRC authorized use of the current P/T curves only through 28 EFPY, pending recalculation of the fluence for the Appendix K power uprate and consequent revalidation of the curves.
Hence, this request seeks to replace the 28 EFPY restriction after recalculating the curves for the Appendix K uprate using the RAMA methodology.
3.3 Technical Analysis 10 CFR Part 50 Appendix G invoked by 10 CFR 50.60, specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the RCPB, including RPV, during any condition of normal plant operation, including anticipated operational occurrences and system hydrostatic tests. Evaluations to demonstrate continued compliance require information regarding irradiated RPV material properties and the neutron fluence level of the RPV.
NLS2011015 Attachment 1 Page 5 of 11 The RAMA methodology for neutron fluence has been benchmarked using experimental and numerical problems specified in RG 1.190. The results of the benchmark cases are documented in the Electric Power Research Institute report entitled "RAMA Fluence Methodology Benchmark Manual Evaluation of RG 1.190 Benchmark Problems." CNS RPV neutron fluence was calculated using RAMA methodology. Projected neutron fluence values are presented for two points in time:
the end of cycle 23, and 32 EFPY. It was assumed in projecting the 32 EFPY fluences that cycle 23 was an equilibrium cycle and representative of future operating cycles including consideration of 24-Month Cycles. Thus, the incremental fluence change determined for cycle 23 was assumed to be constant which allows the fluence between cycle 23 and 32 EFPY to be estimated using linear interpolation for the purpose of evaluating the P/T curves. The calculation meets applicable industry standards and provides a reasonable estimate of the RPV beltline region neutron fluence. The calculation uses a methodology approved by the NRC.
The CNS Updated Safety Analysis Report (USAR) Section IV-2.7.2 describes the RPV Material Surveillance Test Program, which is used to periodically revalidate and update the P/T curves. The revised P/T curves were developed based on the results from the first and second vessel material surveillance capsules. Adjusted reference temperature (ART) values were developed for the RPV materials in accordance with RG 1.99 Revision 2, based on the current fluence data. The fluence was recalculated to incorporate the Appendix K power uprate and the renewal of the license for an additional 20 years. The most limiting beltline material is the Lower Longitudinal Weld with an ART value of 103.5 'F. The most limiting upper vessel material is at the feedwater nozzles.
The changes to the P/T limits provide margin to prevent brittle type fracture of the RPV. Three regions of the RPV were evaluated, the bottom head region, the beltline region, and the upper vessel region.
Revising the TS P/T limit curves and associated SRs as proposed does not affect assumptions in USAR accident analyses.
In summary, the proposed change is technically sound and continues to maintain the same level of safety as the current licensing basis.
4.0 REGULATORY SAFETY ANALYSIS 4.1 Applicable Regulatory Requirements/Criteria Construction of CNS predated the 1971 issuance of 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants." Appendix F, "Conformance to AEC Proposed General Design Criteria," of the CNS USAR discusses that CNS is designed to conform to the proposed general design criteria (GDC) published in the July 11, 1967, Federal Register, except where commitments were made to specific
NLS2011015 Page 6 of 11 1971 GDC. It notes that the Atomic Energy Commission accepted CNS conformance with these proposed GDC.
The following is a discussion of the applicable regulations and the Draft GDC from USAR Appendix F, along with a discussion of continued conformance.
10 CFR 50.36, Technical Specifications 10 CFR 50.3 6(b) requires that each license authorizing operation of a utilization facility to include TS. 10 CFR 50.36(c)(2) identifies Limiting Conditions for Operation (LCOs) as one of the categories to be included in TS. 10 CFR 50.36(c)(2)(ii) states:
"A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:
(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
(B) Criterion2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(C) Criterion3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
(D) Criterion4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety."
10 CFR 50.36(c)(3) identifies SRs as one of the categories to be included in TS.
10 CFR 50.36(c)(3) states:
"Surveillance Requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."
NLS2011015 Page 7 of 11 The revised P/T curves, based on RAMA fluence methodology and conformance with ASME code, will continue to ensure the RCS is able to perform its function. The changes to SR 3.4.9.5, SR 3.4.9.6 and SR 3.4.9.7 continue to assure necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met. Therefore, CNS continues to meet this regulation with the proposed changes to Figures 3.4.9-1, 3.4.9-2 and 3.4.9-3.
10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation 10 CFR 50.60 requires that the pressure and temperature limits as well as the associated vessel surveillance program be consistent with 10 CFR 50 Appendix G, "Fracture Toughness Requirements," and 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements." Since the proposed changes to the P/T curves have been developed consistent with Appendixes G and H, CNS meets this requirement.
10 CFR 50 Appendix G, Fracture Toughness Requirements 10 CFR 50 Appendix G specifies the fracture toughness and testing requirements for reactors vessel material in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G. 10 CFR 50 Appendix G requires prediction of the effects of neutron irradiation on the vessel embrittlement by calculating the ART and Charpy upper-shelf energy. Generic Letter 88-11, "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. RG 1.190 is the standard for calculating fluence, and NRC has accepted RAMA methodology as meeting that standard. RG 1.99 methods use that fluence to calculate the shift in ART of the RPV internals. Since RAMA methodology satisfies RG 1.190, CNS meets the Appendix G requirement.
10 CFR 50 Appendix H, Reactor Vessel Material Surveillance Program Requirements Appendix H of 10 CFR 50 requires the establishment of a surveillance program to periodically withdraw surveillance capsules from the reactor vessel. USAR Section IV-2.7.2 describes the CNS RPV Material Surveillance Test Program, which is used to periodically revalidate and update the P/T curves. Since the revised P/T curves were developed based on the results from the first and second vessel material surveillance capsules, CNS continues to meet these requirements and utilize the results as intended.
NLS2011015 Page 8 of 11 4.2 Precedent This requested amendment would be similar to Duane Arnold Amendment 253 sent by letter from Darl S. Hood, NRC, to Mark A. Peifer, Nuclear Management Company, LLC., dated August 25, 2003, "Duane Arnold Energy Center - Issuance of Amendment Re: Pressure and Temperature Limit Curves (TAC NO. MB8750),"
although NPPD used a different NRC-approved method for calculating neutron fluence.
4.3 No Significant Hazards Consideration 10 CFR 50.91(a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of no significant hazard posed by issuance of the amendment. Nebraska Public Power District (NPPD) has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92(c). The following is the evaluation required by 10 CFR 50.91(a)(1).
NPPD is requesting an amendment to the Cooper Nuclear Station Technical Specifications (TS). The proposed amendment revises the curves in TS 3.4.9, "RCS Pressure and Temperature (P/T) Limits," to replace the 28 Effective Full Power Years (EFPY) restriction in TS Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 and revises minimum temperature in Surveillance Requirement (SR) 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7.
The proposed amendment includes a set of updated P/T curves for pressure test, core not critical, and core critical conditions for 32 EFPY based on a fluence evaluation performed using Nuclear Regulatory Commission (NRC) approved fluence methodology. The new curves show a shift of minimum operating temperature which allows the bolt-up and minimum temperatures specified for SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 to be changed from 80'F to 70 0 F.
- 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The P/T limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed by American Society of Mechanical Engineers (ASME)
Code Section XI, 10 CFR 50 Appendix G and H, and associated guidance documents such as Regulatory Guide 1.99 Revision 2, as restrictions on normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary. Thus, they ensure that an accident precursor is not likely. Hence, they are included in the TS as satisfying Criterion 2 of 10 CFR 50.36(c)(2)(ii). The revision of the numerical value of these limits, i.e., new curves, using an NRC-approved methodology, does not change the existing regulatory requirements, upon which the curves are
NLS2011015 Page 9 of 11 based. Thus, this revision will not increase the probability of any accident previously evaluated.
The proposed change does not alter the design assumptions, conditions, or configuration of the facility or the manner in which the facility is operated or maintained. The proposed changes will not affect any other System, Structure or Component designed for the mitigation of previously analyzed events. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Thus, the proposed revision of the existing numerical values with the updated figures for the Reactor Coolant System (RCS) P/T limits, which are based upon an NRC-approved methodology for calculating the neutron fluence on the Reactor Pressure Vessel (RPV) and new bolt-up limit, will not increase the consequences of any previously evaluated accident.
Therefore, this proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the processes governing normal plant operation. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. NPPD is only requesting to revise the existing numerical values and update the TS figures for the RCS P/T limits based upon an NRC-approved methodology for calculating the neutron fluence on the RPV, and to reflect a new bolt-up limit. The curves continue to be based upon ASME Code.
Therefore, the proposed change does not create the possibility for a new or different kind of accident from any accident previously evaluated.
- 3. Do the proposed changes involve a significant reduction in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which Safety Limits, Limiting Safety System Settings or Limiting Conditions for Operation are determined.
The setpoints at which protective actions are initiated are not altered by the proposed changes. Sufficient equipment remains available to actuate upon
NLS2011015 Page 10 of 1I demand for the purpose of mitigating an analyzed event. NPPD is only requesting to revise the existing numerical values and update the TS figures for the RCS P/T limits based upon an NRC-approved methodology for calculating the neutron fluence, Radiation Analysis Modeling Application. The new curves also reflect a new bolt-up limit. No changes to the other Limiting Conditions for Operation or SRs of TS 3.4.9 are proposed.
10 CFR 50, Appendix G specifies fracture toughness requirements to provide adequate margins of safety during operation over the service lifetime. The values of adjusted reference temperature and upper-shelf energy will remain within the limits of Regulatory Guide 1.99 Revision 2 and Appendix G of 10 CFR 50 for at least 32 EFPY of operation. The safety analysis supporting this change continues to satisfy the ASME Code, 10 CFR 50 Appendixes G and H requirements, and associated guidance documents such as Regulatory Guide 1.99 Revision 2. Thus, the proposed changes will not significantly reduce any margin of safety that currently exists.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the responses to the above questions, NPPD concludes that the proposed amendment presents 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.
4.4 Conclusion 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.
5.0 ENVIRONMENTAL CONSIDERATION
10 CFR 51.22 provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment or environmental impact statement. 10 CFR 51.22(c)(9) identifies an amendment to an operating license for a reactor which changes an inspection or a surveillance requirement as a categorical exclusion provided that 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 a significant increase in individual or cumulative occupational radiation exposure.
NLS2011015 Page 11 of 11 CNS review has determined that the proposed amendment, which would change a TS SR, does not involve (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluent that might be released offsite, or (3) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0 REFERENCES
6.1 ASME Boiler and Pressure Vessel Code,Section III (1965 Edition and January 1966 Addenda) 6.2 BWRVIP- 115, "RAMA Fluence Methodology Benchmark Manual Evaluation of Regulatory Guide 1.190 Benchmark Problems" 6.3 Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" 6.4 CNS License Amendment 231 dated June 30, 2008 6.5 CNS License Amendment 201 dated October 31, 2003 6.6 NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 6.7 NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988 6.8 NRC Safety Evaluation of Proprietary EPRI Reports, "BWR Vessel and Internals Project, RAMA Fluence Methodology Manual (BWRVIP-1 14)," "RAMA Fluence Methodology Benchmark Manual-Evaluation of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-1 15)," "RAMA Fluence Methodology - Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-1 17)," "RAMA Fluence Methodology Procedures Manual (BWRVIP- 121)," and "Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001-R-001)" (TAC No. MB9765) 6.9 CNS USAR Section IV-2.5.1 6.10 CNS USAR Section IV-2.7.2 6.11 ASME Section XI, Appendix G
NLS2011015 Page 1 of 8 Attachment 2 Proposed Technical Specification (TS) Revisions (Markup)
Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised TS Pages 3.4-22 Insert 1 3.4-23 Insert 2 3.4-24 Insert 3 3.4-25
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.9.5 -------------------NOTE --------------------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor vessel flange and head 30 minutes flange temperatures are > "80F r-80 SR 3.4.9.6 ------------------- NOTE ------------
Not required to be performed until 30 minutes after RCS temperature < 90*F in MODE 4.
Verify reactor vessel flange and head 30 minutes flange temperatures are > 89"F.
.J SR 3.4.9.7 -------------------NOTE -------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature _S1"'F in MODE 4.
Verify reactor vessel flange an-- head 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> flange temperatures are > 80"F.
"/I-Cooper 3.4-22 Amendment No. 4&
Insert 1 Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,300
-S1,000
- 900 L
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500 JF I V, NRpqtwegion 200 0
O 400 00 100 110 In IM180 MM M M AM M 3n W MMIIIUM REACTOR VESSEL METAL TEMPERATURE (OF)
eplace Figure ith Insert 1 RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 28 EFPY 1,600 1,500 -
1 z:1
-~H{ ~$~z~p
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200 Bottom Head 100 >80*1F 7 - tpper Vessel 100 150 200 250 300
/0 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)
Figure 3.4.9-1 (fpe-4of 44-)
Pressure/Temperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear Shutdown Valid Through 2--EFPY Anien 3.4-23 Amendment
Insert 2 Cooper Pressure Test Curve (Curve A), 32 EFPY 1,31W 1,100, - -.
IVO. "4. pi ploV 1600 - -- -
3i. Safe S T~Operating
- QO * ! Region "'
4000 I I I 1-
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-0" - - : J - -i . .. .........
1=4--
IN 0
0 2D0 4060 80 100 120 140 160 180 200 2202 40 260 280 300 320 340 MtINUM REACTOR VESSEL METAL TEMPERATUfRE M¶
eplace Figure RCS P/T Limits ith Insert 2 3.4.9 Cooper Pressure Test Curve (Curve A), 28 EFPY 1,6040 1.504 1,404 0 ........ ..
1,304
- ~ ~~~~~~~~7
.- ...... 4
- 1,20i 0 ~~... ... -. . . ....
zus 1,100 3 . ~. .................
1,004 90 0 - . . .-.-. ,... .-. . ........ -..----
0 80
- ....... L Opermin w 0 w 70 0: ~ . . .- - . .1J 23 F,753 rzg k S 60
.J w
Ixg 0 809' 597rig .- ----. --...
50 lu 40
- Beklkie 30 - ,--..--.
. -...-.... ...........- -- Bottom Head 20 0 ~ .. r. 123'P, 269ptig .. .
- UpperVessel 10(
0>80-F ....... -.......-...- ---..-.-...-.
20 40 60 80 100 120 140 160 180 200 220 240 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
/0 Figure 3.4.9-2 (page 1of4 Pressure/Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests Valid Through.28 EFPY I ZýýD Amendment 233 3.4-24 Amendment
Insert 3 Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,300 1,100 1,2000_+,
r 1700 z Lin
&*500 prtn 100 +
400 0 24012 I 1 F I -T. 140 I 1 40 20 406 0101010101020 2 4 6 8 0 24 MIMMU RECO VESSEL META TEPEAUR ..
- M(, RECO VE E
Replace Figure RCS P/T Limits ith Insert 3 3.4.9 Cooper Heatup/Cooldown Core Critical Curve 1,600 (Curve C), 28 EFPY
/ I 1,500 1,400 1,300 I... ...... ... .
. 1,200 1,100
-I-o11000
-J U, 900 800 Lii 700 Ie z L - Curve C 2
-J w
ui E
600 500 400
.T.TT.T... .......Y "! -
gOF,13pi ULlI17fdtiI
-I
-17.
I -- 4.- -- - -- - ý. .....
7' Region
-7 1
300 Minimum Criticality ~lV~-.Z22.t2Z2 ::Z:rcZz~.
with/
200 Normal I ter .. . T ,
el 100
>80_F I--I ý-_: I--i:
ýý4-1_--_ 4, tz ý::1:
0 V nmmm w m m,=*m* n 0 50 100 150 200 250 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
Figure 3.4.9-3 (page I-f,)
Pressure/Temperature Limits for Criticality Valid Through..2. EFPY Cooper
-A=ý 3.4-25 Amendment
NLS2011015 Page 1 of 5 Attachment 3 Proposed Technical Specification (TS) Revisions (Re-Typed)
Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised TS Pages 3.4-22 3.4-23 3.4-24 3.4-25
RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.9.5 -------- NOTE -.--------.........-------
Only required to be performed when tensioning the reactor vessel head bolting studs.
Verify reactor vessel flange and head flange 30 minutes temperatures are > 70 0 F.
SR 3.4.9.6 -------- NOTE ---------........------.--
Not required to be performed until 30 minutes after RCS temperature < 80°F in MODE 4.
Verify reactor vessel flange and head flange 30 minutes temperatures are > 700 F.
SR 3.4.9.7 -------- NOTE ---------..--------------..
Not-required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature < 90°F in MODE 4.
Verify reactor vessel flange and head flange 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> temperatures are > 70'F.
Cooper 3.4-22 Amendment __
RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,300 1,200 I t I U 1,100 919919 1199 199i119E9 9191119 99 i91911 I 1t,000 ' I
.9 U3 IJ t 0800 700 z
I,-
Safe Operating 1 -1_ Region S500 aL - t I I II I I 400 ! [ lI I '------Be*Idh* Region m JI I I I Bolt-up - 7W'.*M WV.i . ..
200 TeW: .. .
> 70 F - -: - . .. ... . .. . . ...... .. . .. .
100 I1I :1 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MIIIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Figure 3.4.9-1 Pressure/Temperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear Shutdown Valid Through 32 EFPY Cooper 3.4-23 Amendment No.
RCS P/T Limits 3.4.9 Cooper Pressure Test Curve (Curve A), 32 EFPY 1,300 1,200 1,100 1,000 900 800 700 IL 600 500 400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.9-2 Pressure/Temperature Limits for Inservice Hydrostatic and Inservice Leakage Tests Valid Through 32 EFPY Cooper 3.4-24 Amendment No.
RCS P/T Limits 3.4.9 Cooper Heatup/Cooldown, Core Critical Curve (Curve C), 32 EFPY 1,300 1,200 1,100
', i I 1,000
_J900 I80IF 67 P8-9
'U to fa w
>800 0
z
,' _-1p*1
.E600 p-Safe eratng w5O gio X tI "7-Z n 400 I"-I " --
Minimum Core I I I 300 Critical t7f. 313 ýP"1 Temperature:
> 80'F 200 100 I Orr, in pug 0 .Y-T-Y-T-0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 3.4.9-3 Pressure/Temperature Limits for Criticality Valid Through 32 EFPY Cooper 3.4-25 Amendment No.
NLS2011015 Page 1 of 4 Attachment 4 Proposed Technical Specification (TS) Bases Revisions (Information Only)
Cooper Nuclear Station, Docket No. 50-298, DPR-46 Revised TS Bases Pages B 3.4-46 B 3.4-51 B 3.4-52
RCS P/T Limits B 3.4.9 BASES APPLICABLE SAFETY ANALYSES (continued)
P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) (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 < 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 Figure 3.4.9-2. During RCS heatup and cooldown operation (i.e., not critical and not performing inservice leak or hydrostatic testing) verify RCS pressure and temperature are within the applicable limits specified in Figure 3.4.9-1. During RCS inservice leak and hydrostatic testing verify RCS pressure and temperature are within the applicable limits specified in Figure 3.4.9-2;
- b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is <
145°F 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
90OF m 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.
Cooper B 3.4-46 04/11/06
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)
REQUIREMENTS Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.3 is to compare the bottom head drain temperature to the RPV steam dome saturation temperature.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4 during a recirculation pump startup since this is when the stresses occur. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.
SR 3.4.9.5, SR 3.4.9.6, and SR 3.4.9.7 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
80 The flange temperatures must be veri ed t- above the limits within 30 minutes before and while tensioni g the vessel head bolting studs to ensure that once the head is tensiord the limits are satisfied. When in MODE 4 with RCS temperature < 00 °F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits.
When in MODE 4 with RCS temperature < 4.Q°F, monitoring of the flange temperature is required every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within the specified limits.\ 90 (continued)
Cooper B 3.4-51 Revision 0
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.
80 SR 3.4.9.5 is modified by a Note that requires the urveillance to be performed only when tensioning the reactor vess head bolting studs.
SR 3.4.9.6 is modified by a Note that requires t Surveillance to be initiated 30 minutes after RCS temperature < OF 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 <4-00°F in MODE 4. The Notes contained in these SRs are necessary to spech,,when the reactor vessel flange and head flange temperatures are requiredt be within the specified limits.
REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
- 3. BWRVIP-86-A, October 2002.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. USAR, Section IV-2.6.
- 7. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
- 9. USAR, Appendix G.
- 10. ASME XI Code Case N-640.
Cooper B 3.4-52 08/11/04
NLS2011015 Enclosure Enclosure Calculation NEDC 07-048, Revision 5C1, Revised Pressure Temperature Curves (76 pages)
Page 1 of 7
Title:
Revised Pressure Temperature Curves Calculation Number: NEDC 07-048 Rev. 5C1
_CED/EE Number: EE1 0-028 Revision 2 System/Structure: NB Setpoint Change/Part Eval Number N/A Component: RPV Discipline: Civil Classification: [X] Essential; [ ] Non-Essential SQAP Requirements Met? [ ] Yes; [X] N/A Proprietary Information Included? [ ]Yes; [X] No
==
Description:==
The purpose of this calculation is to review Structural Integrity (SI) calculation 1100445.303, which develops the vessel pressure temperature (P-T) curves for the betline for 32 and 54 effective full power years (EFPY). The calculation includes the 2% power uprate, which was implemented after RE24 and the proposed changed to a 24 month fuel cycle.,This calculation also reviews and accepts Structural Integrity (SI) calculation 1100445.304, which evaluates the Core Differential Pressure Nozzle.
Revision I changed the minimum temperature in Table 3 and Figure 1 to 70*F since the moderator temperature for the shutdown margin calculation is 687F. Also revised reference 2 to the latest version.
Revision 2 revises curves A and B to add the lower head and upper vessel (all EFPY) regions to show that the beltline curves are limiting.
Revision 3 corrects the graph of Curve A for the FW Nozzle and Curve C. Ref. CR-2010-2176.
Revision 4 changed the implementing EE number and noted the revisions to Design Inputs 3 and 4. Changed Design Input 2 to NEDC 07-032. These changes did not affect the calculation. None of the previous revisions were docketed with the NRC.
Revision 5 corrects Curve B for the lower head region (Ref CR-2010-4548). Also corrected a typo in a TS reference listed in the first paragraph under methodology.
Revision 5C1 adjusts the curves for the new fluences calculated for a 24 month fuel cycle. It also includes SIA Calculation 1100445.304 Revision 0 for the Core Differential Pressure Nozzle.
Conclusions and Recommendations:
The calculation to revise the P-T curves to 32 EFPY and 54 EFPY is consistent with the design basis of the plant. The P-T curves were developed inaccordance with the methods in the 2001 Edition, 2003 Addenda of ASME Section X1, Appendix G.
The ART and USE were calculated in accordance with RG 1.99. The curves generated from these methods ensure that the P-T limits will not be exceeded during any phase of reactor operation. These curves are suitable for inclusion in the Technical Specifications, USAR and other affected documents following NRC approval.
Structural 7/15/1011 Integrity Associates 501 3 r ý,V as NIA f 3K.B. Thomas Mark E. Unruh Mark E. Unruh Jerry Horn 8/16/10 8/16/10 8/18/10 43 3 3 K.
K.B.B.Thomas Thomas Kirby Woods Kirby Woods Kirby Woods Kirby Woods Willia/Gree William Green 5/27/2010 3 3 K.B. Thomas Kirby Woods ~4/29/2010Kirby Woods William Green 2 3 K.B. Thomas Kirby Woods Kirby Woods Willia Gree 1 3 K.B. Thomas Kirby Woods Kirby Woods/Jerry Horn William Green 1/15/2010 0 3 K.B. Thomas & Ronald L.Yantz William Green Kirby Woods Ronald L. Yantz 7/30/2009 Rev.
N er Status Prepared By/Date Reviewed By/Date IDVed By/Date Approved By/Date NumberI Status Codes
- 1. Active 4. Superseded or Deleted 7. PRA/PSA
- 2. Information Only 5. OD/OE Support Only
- 3. Pending 6. Maintenance Activity Support Only
NEBRASKA PUBLIC POWER DISTRICT PAGE: 2 OF 7.
NEDC: 07-048.
DESIGN CALCULATIONS SHEET REV. NUMBER: 5C1 DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REV. PENDING CHANGES NO. DESIGN INPUTS NO. TO DESIGN INPUTS ASME Code,Section XI, Rules for 1 Inservice Inspection of Nuclear Power N/A 2001 Edition, 2003 Addenda Plant Components, Appendix G 2 NEDC 07-032, Vessel Fluence 2 None 3 NEDC 07-045, ART 1C1 None 4 NEDC 07-046, USE 2 None 5 BWRVIP-74-A 0 None Memo DED 2003-005, Alan Able to Ken 6 Thomas, dated August 14, 2003, 0 None "Instrument Uncertainty Associated With Technical Specification 3.4.9 7 CE Dwg No. E-232-230 3 None 8 GE Dwg No. 729E479-B, Sheet 1 of 3 2 None MISC-PENG-ER-012, Cooper Reactor 9 Pressure Vessel, Plates, Forgings, 0 None Welds and Cladding 10 OP 2.2.47, HVAC Reactor Building 42 None 11 CE Stress Report, CENC-1 150 N/A April 1971 12 CE Dwg No. E-232-243 7 None GE-NE-523-159-1292, Vessel 13 Surveillance Materials Testing and N/A February 1993 Fracture Toughness Analysis Report 14 NEDC99-020 1 ICI
NEBRASKA PUBLIC POWER DISTRICT PAGE: 3 OF 7__
NEDC: 07-048.
DESIGN CALCULATIONS SHEET REV. NUMBER: 5C__1 DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REV.
NO. AFFECTED DOCUMENTS NUMBER 1 TS 3.4.9 and Bases 07/16/08 2 6.MISC.502, ASME Class I System Leakage Test 35 3 6.MISC.504, ASME Class 1 Hydrostatic Test 12 6.RCS.601, Technical Specification Monitoring of RCS Heatup/Cooldown 18 Rate 5 PMIS 07, PMIS HYDRO display N/A 6 OP 2.1.1, Startup Procedure 159 7 USAR, Section IV2.6.3 8/07/08 8 USAR, Section IV2.6.2 8/07/08 4 4 4
4 4 4 -4 4
.1.
-4 .1.
+ .1.
4 4 4 +
+
.1. .1.
NEBRASKA PUBLIC POWER DISTRICT PAGE: 4 OF 7.
NEDC: 07-048 DESIGN CALCULATIONS SHEET REV. NUMBER: 5C1 The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations. 2 SCREENING QUESTIONS YES NO UNCERTAIN
- 1. Does it involve the addition, deletion, or manipulation of a component or [] [x] []
components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?
- 2. Could it impact system operating parameters (e.g., temperatures, flowrates, I I[X] [I pressures, voltage, or fluid chemistry)?
- 3. Does it impact equipment operation or response such as valve closure I] I(X] [I time?
- 4. Does it involve assumptions or necessitate changes to the sequencing of [I [x] [I operational steps?
- 5. Does it transfer an electrical load to a different circuit, or impact when electrical loads are added to or removed from the system during an event?
[] x] []
- 6. Does it influence fuse, breaker, or relay coordination? I [XI [I
- 7. Does it have the potential to affect the analyzed conditions of the I [XI []
environment for any part of the Reactor Building, Containment, or Control Room?
- 8. Does it affect TS/TS Bases, USAR, or other Licensing Basis documents? [I [X] [I
- 9. Does it affect a Dry Fuel Storage (CoC) or associated Technical Specification, Dry Fuel Storage UFSAR, or CNS 1CCFR72.212 Report? [] [x] []
- 10. Does it constitute any change or addition to, or removal from, the H [XI IH Independent Spent Fuel Storage Installation (ISFSI) facility or spent fuel storage cask design, or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished?
- 11. Does it affect DCDs? II XI [I 12.. Does it have the potential to affect procedures in any way not already [I [X] [I]
mentioned (refer to review checklists in Procedure EDP-06)? If so, identify:
If all answers are NO, then additional review or assistance is not required.
If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2.
- Supervisor Quality Review for Change Evaluation Documents (CEDs) and Engineering Evaluations (EEs)
Rating Item N@)4 3 2 1 a. The configuration issue is clearly stated, accurate and complete, succinctly W2=
1describing the problem.
N)4 3 2 1 b. The design bases for the issue are defined and referenced and the stated bases that bound the issue are evident.
N(94 3 2 1 c. The need for the change is established and adequately supported.
N64 3 2 1 d. Assumptions, thought processes, methodology and facts are complete, accurate and thoroughly documented.
5 4321 e. The appropriate support for product development has been supplied from other departments.
NA4 3 2 1 f. The change is clearly defined.
N@ 4 3 2 1 g. Nuclear safety impact of the change is evaluated.
54321 h. The change resolves the original issue and does not present new, unresolved issues.
NC 4 3 2 1 i. Affected processes, procedures, programs are identified to ensure effective change.
54321 j. Testing and acceptance requirements are identified to address all functional requirements.
1N44 3 2 1 k. Appropriate level of review has been identified on the CED/EE. This is evidenced throughout the product and in the level and breadth of review specified.
Nb4 3 2 1 1. The CEDIEE is well organized, easy to read and navigate, sequence is logical and facilitates understanding for both short- and long-term historical purposes.
No 4 3 2 1 m. If the CED/EE is prepared under a contract, the preparer has met the terms and conditions of the contract and the product is acceptable.
N*)4 3 2 1 n. The reviewer(s) has demonstrated a thoroughness reflective of station expectations for professionalism and quality (i.e., ensured editorial and technical error's have been captured and the review comments were constructive and productive).
95 432 1 o. If the CED/EE is reviewed under a contract, the reviewer has met the terms and conditions of the contract and the product is acceptable.
Item Reviewed: /3/b,D 6 o 7.- 0 Ye* 5C/ Overall Rating:;
Reviewer: ~ ~
Date: " '
(Attach comments as necessary.) Send copy to QA when CEDJEE is prepared or reviewed under contract.
EDP-12 Revision 10 Page 5 of 9
NEBRASKA PUBLIC POWER DISTRICT PAGE: 5 OF 7 NEDC: 07-048.
DESIGN CALCULATIONS SHEET REV. NUMBER: 5C1 PURPOSE:
The purpose of this calculation is to review Structural Integrity (SI) calculation 1100445.303, which develops the vessel pressure temperature (P-T) curves for the beltline for 32 and 54 effective full power years (EFPY). The calculation includes the 2% power uprate, which was implemented after RE24 and the proposed changed to a 24 month fuel cycle.
This calculation also reviews and accepts Structural Integrity (SI) calculation 1100445.304, which evaluates the Core Differential Pressure Nozzle. The evaluation shows that this nozzle is bounded by the lower head and, therefore, does not affect the P-T curves.
ASSUMPTIONS:
Assumptions are identified within the body of the attached SI calculation according to context and use.
Obvious assumptions are identified as follows:
The calculation includes the 2% power uprate that was implemented after RE24. The adjusted reference temperature (ART) values were determined in accordance with Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (DI 3). The limiting beltline material has an ART value of 103.2°F and 123.5°F for 32 and 54 EFPY, respectively. Non-beltline regions are not subjected to fluence in excess of 10E+17 n/cm 2; therefore, the RTNDT values are valid substitutions for corresponding ART values. These values are considered to be the most appropriate for computation of updated P-T curves.
Vessel Dimensions and Fluid Properties The inner radius of the cylindrical portion of the reactor pressure vessel (RPV) is 110.375 inches. The vessel shell base metal varies in thickness from 5.375 to 6.375 inches, with 5.375 the predominant base metal thickness in the beltline region. Therefore, 5.375 inches is the smallest thickness (i.e.,
bounding input for P-T calculations).
The normal water level in the RPV is 551.75 inches. The maximum potential water level in the RPV during the pressure test is 825.1875 inches.
The wall thickness in the lower head varies from 3.1875 inches, where it connects to the lower shell, to 6.8125 inches. The thicker section provides the reinforcement for the CRD penetrations.
The Feedwater nozzle penetration in the upper vessel is considered the limiting location for stress in this region. The nozzle inner diameter is 11.969 inches with an outside diameter of 23.0625 inches. The nozzle corner radius, both inside and outside, is 5% in. and the thickness at the corner is also 5/ in.
The Heatup and Cooldown temperature rate-of-change assumed for the vessel metal for normal operation is 100°F/hr.
Pressure and Instrument Uncertainty The pre-service system hydrostatic test pressure was 1,563 psig, which corresponds to 1.25 times the design pressure of 1,250 psig. This is consistent with previous calculations. The following are the results of the instrument uncertainty for CNS pressure and temperature measurements:
Reactor Vessel Metal Temperature is bounded by +/- 5°F Reactor Vessel Pressure is bounded by +/- 2%
Assumptions were reviewed and found to be acceptable.
NEBRASKA PUBLIC POWER DISTRICT PAGE: 6 OF 7.
NEDC: 07-048 DESIGN CALCULATIONS SHEET REV. NUMBER: 5C1 METHODOLOGY:
The limits in this report were derived from the NRC-approved methods listed in TS 3.4.9, using the Structural Integrity Associates Topical Report SIR-05-044, "Pressure Temperature Report Methodology for Boiling Water Reactors". This methodology has been reviewed and accepted by the NRC (Ref 3).
The methodology is described within the body of the attached SI calculations.
Minimum Boltup Temperature For conditions where the core is not critical (Curves A and B), the minimum boltup temperature is equal to the material RTNDT of the limiting region affected by boltup stresses per Table I of 10CFR50, Appendix G. However, the temperature controller for the refueling floor is set for a minimum of 65°F.
Per Technical Specification, the moderator temperature for the shutdown margin calculation is 68°F.
Thus, the minimum boltup temperature should be 68°F or the material RTNDT, whichever is higher. The flange RTNDT is equal to 20°F; therefore, the minimum boltup temperature of 70'F is acceptable.
Instrumentation Nozzles The instrument nozzles, N16A/B, are the only nozzles in the extended beltline region. The instrument nozzles are located in RPV Region B, which has a maximum temperature of 522°F during the shutdown transient. The instrument nozzle material is SB-166. The N16B nozzle is the limiting instrument nozzle in the extended beltline region with ART values for 32 and 54 EFPY of 52.4°F and 70.8°F, respectively. Therefore a composite beltline curve was developed.
CONCLUSION:
These calculations of the P-T curves to 32 EFPY and 54 EFPY are consistent with the design basis of the plant. The P-T curves were developed in accordance with the methods in the 2001 Edition, 2003 Addenda of ASME Section Xl. The ART and USE were calculated in accordance with RG 1.99. The curves generated from these methods ensure that the P-T limits will not be exceeded during any phase of reactor operation. These curves are suitable for inclusion in the Technical Specifications, USAR and other affected documents following NRC approval.
NEBRASKA PUBLIC POWER DISTRICT PAGE: 7 OF 7 NEDC: 07-048 DESIGN CALCULATIONS SHEET REV. NUMBER: 5C1
REFERENCES:
- 1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
- 2. Structural Integrity Associates Topical Report SIR-05-044, "Pressure Temperature Report Methodology For Boiling Water Reactors"
- 3. U.S.NRC Final Safety Evaluation For The Boiling Water Reactor Owners' Group (BWROG)
Structural Integrity Associates Topical Report (TR) SIR-05-044, "Pressure Temperature Report Methodology For Boiling Water Reactors", February 6. 2007
- 4. BWRVIP-135 Revision 2: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations"
- 5. Technical Specification Section 1.1, Definition of Shutdown Margin ATTACHMENTS:
A. Structural Integrity Calculation 1100445.304, Revision 0, " Revised P-T Curve Calculation", is attached to this calculation in its entirety.
B. Structural Integrity Associates Calculation No. 1100445.304, Revision 0, "Core Differential Pressure Nozzle Finite Element Model and Stress Analysis." is attached to this calculation in its entirety.
V StructuralIntegrityAssociates, Inc. File No.:No.:
Project 1100445.303 1100445 CALCULATION PACKAGE Quality Program: M Nuclear [] Commercial PROJECT NAME:
Cooper P-T Curve Revision CONTRACT NO.:
4200001742 CLIENT: PLANT:
Nebraska Public Power District Cooper Nuclear Station CALCULATION TITLE:
Revised P-T Curve Calculation Project Manager Preparer(s) & Checker(s)
Document Affected Revision Description Approval Revision Pages Signature & Date Signatures & Date 01 - 40 Initial Issue A-i -A-4 B-1 -B-13 Eric Houston Raoul Gnagne EJH 08/05/11 LRG 08/05/11 Vikram Marthandam VM 08/05/11 Page 1 of 40 F0306-O1RI
dUMa hiftgdtyAuobss Table of Contents
1.0 INTRODUCTION
.................................................................................................... 4 2.0 METHODOLOGY .................................................................................................. 4 3.0 ASSUMPTIONS / DESIGN INPUTS ................................................................... 10 4.0 CALCULATIONS ................................................................................................. 13 4.1 Pressure Test (Curve A) ............................................................................ 13 4.2 Normal Operation - Core Not Critical (Curve B) ....................................... 14 4.3 Normal Operation - Core Critical (Curve C) .............................................. 14
5.0 CONCLUSION
S .................................................................................................... 15
6.0 REFERENCES
...................................................................................................... 16 APPENDIX A: P - T CURVE INPUT LISTING .......................................................... A-1 APPENDIX B: BOUNDING BELTLINE SUPPORTING ANALYSIS ....................... B-1 File No.: 1100445.303 Page 2 of 40 Revision: 0 F0306-01RY
~jjsiuciiwhftdgrly AssUciMA/awekc" List of Tables Table 1: CNS Polynomial Coefficients for Feedwater Nozzle Stress Intensity Distributions ........... 18 Table 2: CNS Beltline Region, Curve A, for 32 EFPY ................................................................. 19 Table 3: CNS Beltline Region, Curve A, for 54 EFPY ................................................................. 20 Table 4: CNS Bottom Head Region, Curve A, for All EFPY ....................................................... 21 Table 5: CNS, Upper Vessel Region, Curve A, for All EFPY ........................................................... 22 Table 6: CNS, Beltline Region, Curve B, for 32 EFPY .............................................................. 23 Table 7: CNS, Beltline Region, Curve B, for 54 EFPY .............................................................. 24 Table 8: CNS Bottom Head, Curve B for All EFPY ................................................................... 25 Table 9: CNS Bottom Head-CDP Nozzle, Curve B for All EFPY .............................................. 26 Table 10: CNS Upper Vessel, Curve B for All EFPY ................................................................... 27 Table 11: CNS Curve C for 32 EFPY .......................................................................................... 28 Table 12: CNS Curve C for 54 EFPY .......................................................................................... 29 List of Figures Figure 1: Feedwater Nozzle Path Stress Distribution ................................................................... 30 Figure 2: CNS (Hydrostatic Pressure and Leak Test) P-T Curve A for 32 EFPY ......................... 31 Figure 3: CNS (Hydrostatic Pressure and Leak Test) P-T Curve A for 54 EFPY ......................... 32 Figure 4: CNS (Hydrostatic Pressure and Leak Test) Composite P-T Curve A for 32 EFPY ..... 33 Figure 5: CNS (Hydrostatic Pressure and Leak Test) Composite P-T Curve A for 54 EFPY ..... 34 Figure 6: CNS P-T Curve B (Normal Operation - Core Not Critical) for 32 EFPY ..................... 35 Figure 7: CNS P-T Curve B (Normal Operation - Core Not Critical) for 54 EFPY ..................... 36 Figure 8: CNS (Normal Operation - Core Not Critical) Composite P-T Curve B for 32 EFPY ........ 37 Figure 9: CNS (Normal Operation - Core Not Critical) Composite P-T Curve B for 54 EFPY ........ 38 Figure 10: CNS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY .......................... 39 Figure 11: CNS P-T Curve C (Normal Operation - Core Critical) for 54 EFPY .......................... 40 File No.: 1100445.303 Page 3 of 40 Revision: 0 F0306-01RE
1.0 INTRODUCTION
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 P-T curves are developed for 32 and 54 effective full power years (EFPY) of operation, and are developed using the methodology of the 2001 Edition through 2003 Addendum of the ASME Code,Section XI, Appendix G [1] and 10CFR50 Appendix G [2]. This calculation has been developed in accordance with the Boiling Water Reactor Owner's Group (BWROG) Licensing Topical Reports (LTRs), "Pressure Temperature Limits Report Methodology for Boiling Water Reactors" [3] and "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations" [4].
2.0 METHODOLOGY A full set of P-T curves are computed, including the following plant conditions: Pressure Test (Curve A), Normal Operation - Core Not Critical (Curve B), and Normal Operation - Core Critical (Curve C).
The curves are consolidated into three evaluation regions of the reactor pressure vessel (RPV): (1) the beltline, (2) the bottom head, and (3) the feedwater nozzle / upper vessel. The beltline region, which is 7 typically the most limiting region, is the region adjacent to the core where the fluence exceeds 1.0 x 101 n/cm2 [3].
The methodology for calculating P-T curves described below is taken from Reference [3] unless specified otherwise. Additional guidance regarding analysis of water level instrument nozzles is taken from Reference [4]. The P-T curves are calculated by means of an iterative procedure, in which the following steps are completed:
Step 1: A fluid temperature, T, is assumed. The P-T curves are calculated considering a postulated flaw that has extended V4 of the way through the vessel wall. According to Reference [3],
the temperature at the assumed flaw tip, T1/4 , may be treated as equal to the coolant temperature.
Step 2: The static fracture toughness factor, KIc, is computed using the following equation:
K1, = 20.734 . e°02(T-ART) + 33.2 (1) where: Kit = the lower bound static fracture toughness (ksi4in).
T = the metal temperature at the tip of the postulated Y4 through-wall flaw (0F). Note that the coolant temperature is typically used, as described above.
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~~Sfuo~aJIfudlyAnaocatas. /meP ART = the Adjusted Reference Temperature (ART) for the limiting material in the RPV region under consideration (°F).
Step 3: The allowable stress intensity factor due to pressure, Kip, is calculated as:
KP ,-K 1 , (2)
SF where: Kip = the allowable stress intensity factor due to membrane (pressure) stress (ksiqin).
KI, =the lower bound static fracture toughness factor calculated in Equation 1 (ksiin).
Kit = the thermal stress intensity factor (ksiqin) from through wall thermal gradients.
SF = the ASME Code recommended safety factor, based on the reactor condition.
Note: For hydrostatic and leak test conditions (i.e., P-T Curve A), the SF = 1.5. For normal operation, both non-critical and critical reactor (i.e., P-T Curves B and C), the SF = 2.0.
When calculating values for Curve A, the thermal stress intensity factor is neglected (Kit =
0), since the hydrostatic leak test is performed at or near isothermal conditions (typically, the rate of temperature change is 25°F/hr or less).
For Curve B and Curve C calculations, Kit is computed in different ways based on the evaluated region. For the beltline (with the exception of nozzles) and bottom head regions, Kit is determined using the following equation:
Kit = 0.953 x 10 3 . CR t2 (3) where: CR = the cooldown rate of the vessel (°F/hr).
t = the RPV wall thickness, per region (in).
For the feedwater nozzle/upper vessel region Kit is obtained from the stress distribution output of a finite element model (FEM). A thermal transient finite element analysis (FEA) is performed, and a polynomial curve-fit is applied to the through-wall stress distribution at each time point. The subsequent method to evaluate Kit is:
K-~H076C 2a a' 4a' KI, =t-G 0.706CO, + - .0.537Ct, +---0.448C 2, + 0.393C3, (4) where: a = 1/4through-wall postulated flaw depth, a = 1/4 t (in).
t = thickness of the cross-section through the limiting nozzle inner blend radius corner (in).
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Cj miai"n ltnbgri Assocalu, /oc The thermal stress polynomial coefficients are based on the assumed polynomial form ofo-(x) = C, + C1 . x + C 2
- x 2 + C3 - x3 . In this equation, "x" represents the radial distance in inches from the inside surface to any point on the crack front.
For the water level instrument nozzle configurations (which are present in the extended beltline), the nozzle assembly consists of an insert attached to the RPV with a partial penetration weld. The nozzle material is not ferritic and does not need to be specifically evaluated. However, the effect of the penetration on the adjacent shell must be considered.
Reference [4, Equation 8-2] provides the following for generic calculation of the thermal stress intensity factor due to a thermal ramp transient:
K1 _ra.mp = 874,844[a (t, +t,)] -20.715 (5) where: KI-rap = the generic Kit (thermal stress intensity factor from through wall thermal gradients) for instrument nozzle subjected to 100 0 F/hr thermal ramp transient (ksiq1in).
a = the instrument nozzle material thermal expansion coefficient at the highest thermal ramp temperature (in/in/0 F).
tv = the vessel thickness (in).
tn = the nozzle thickness (in).
Step 4: The allowable internal pressure of the RPV is calculated differently for each evaluation region. For the beltline region (with the exception of nozzles), the allowable pressure is determined as follows:
P1111. = Kf (6)
Mm. R, where: Pa11ow = the allowable RPV internal pressure (psig).
Kip = the allowable stress intensity factor due to membrane (pressure) stress, as defined in Equation 2 (ksi/in).
t = the RPV wall thickness, per region (in).
Mm = the membrane correction factor for an inside surface axial flaw:
Mm = 1.85 for qt < 2 Mm = 0.926 4t for 2 _<t < 3.464 Mm = 3.21 for 4t > 3.464.
Ri = the inner radius of the RPV, per region (in).
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For the bottom head region, the allowable pressure is calculated with the following equation:
= 2. -KpIP.-t Pfo. 'aiw= (7)
SCF .Mm," R, where: SCF = conservative stress concentration factor to account for bottom head penetration discontinuities; SCF = 3.0 per Reference [3].
Paliow, Kip, t, Mm and Ri are defined in the footnotes of Equation 6.
For the feedwater nozzle / upper vessel region, the allowable pressure is determined from a ratio of the allowable and applied stress intensity factors. The applied factor can be determined from a FEM that outputs the stresses due to the internal pressure on the nozzle /
RPV. The methodology for this approach is as follows:
K p -P f I, (8)
Kipap,,(
where: Pref = RPV internal pressure at which the FEA stress coefficients (Equation 9) are valid (psi).
Kip-app = the applied pressure stress intensity factor (ksWin).
Pajow and Kip are defined in the footnotes of Equation 6.
The applied pressure stress intensity factor is determined using a polynomial curve-fit approximation for the through-wall pressure stress distribution from a FEA, similar to the methodology of Equation 4:
2a a 4a'
= N'70p+
KIK~p 0 6 -.," 0.537Cip +-_.
2 0.448C 2 p.+ 37r . 0.393C3p (9) where: a = 1/4 through-wall postulated flaw depth, a = 4 t (in).
t = thickness of the cross-section through the limiting nozzle inner blend radius corner (in).
For the instrument nozzle, the nozzle material is not ferritic and does not need to be specifically evaluated. However, the effect of the penetration on the adjacent shell must be considered. The allowable pressure is determined from the ratio of the allowable and applied stress intensity factors given in Equation 8. The applied stress intensity factor is calculated generically as follows [4, Equation 8-1]:
K, Pressu.e = 2.9045 LR t +t"1 4.434 (10) where: KIvpresu* = generic Kip-app for instrument nozzle(ksi*in).
R = RPV nominal radius (in).
t, and t, are described in the footnotes of Equation 5.
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The core differential pressure (CDP) nozzle is located in the thinner portion of the bottom head. The methodology for analyzing the bottom head, specifically Equation 7, proves to be overly restrictive for the CDP nozzle. The effect of the penetration on the bottom head may be accounted for by determining an applied pressure stress intensity factor. Using a polynomial curve-fit approximation for the through-wall pressure stress distribution form a FEA, the CDP nozzle applied stress intensity factor is calculated by [4, Equation 2-1]:
-(0.723C +0.551C, + + 0.408C (a)
KIP-.PP 'f1X*, + 0.462C 2 2 3 ;r) where: K=p-app plant specific Kip-app for CDP nozzle (ksi'iin).
a = 1/4 through-wall postulated flaw depth, a = 1/44 t (in).
t = thickness of the cross-section through the limiting nozzle inner blend radius comer (in).
Co, C 1, pressure stress polynomial coefficients, obtained from C 2, C 3 curve-fit of the extracted stresses from FEA.
Step 5: Steps 1 through 4 are repeated in order to generate a series of P-T points; the fluid temperature is incremented with each repetition. Calculations proceed in this iterative manner until 1,300 psig. This value bounds expected pressures.
Step 6: The following minimum temperature requirements apply to the feedwater nozzle / upper vessel region according to Table 1 of IOCFR50, Appendix G [2]:
" If the pressure is greater than 20% of the pre-service hydro-test pressure, the temperature must be greater than the RTNDT of the limiting flange material plus a temperature adjustment. For Curve A calculations, the temperature adjustment is 90'F; for Curve B, the temperature adjustment is 120', and for Curve C the temperature adjustment is the highest value between the minimum permissible temperature for the inservice system hydrostatic pressure test and the sum of the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 160'F.
" If the pressure is less than or equal to 20% of the pre-service hydro-test pressure, the minimum temperature must be greater than or equal to the RTNDT of the limiting flange material. For Curve A and B calculations, the minimum temperature is the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload; for Curve C calculations, the minimum temperature is the highest value between the minimum permissible temperature for the inservice system hydrostatic pressure test and the sum of the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 40°F.
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fiigqly Step 7: The final P-T limits are calculated using the following equations:
TP-T =T+UT (12)
PP-T Pa,,o,v - PH - UP (13) where: TP-T The allowable coolant (metal) temperature (°F).
UT The coolant temperature instrument uncertainty (0 F).
PP-T = The allowable reactor pressure (psig).
PH4 The pressure head to account for the water in the RPV (psig).
Can be calculated from the following expression: P,, = p" Ah.
p Water density at ambient temperature (lb/in3).
Ah Elevation of full height water level in RPV (in).
Up The pressure instrument uncertainty (psig).
These additional pressure and temperature limits are not applicable to the IOCFR50 Appendix G [2] limits described in Step 6.
The P-T Curves for hydrostatic leak test (Curve A) and normal operation - core not critical (Curve B) may be computed by following Steps 1 through 7. Table 1 of Reference [2] requires that core critical P-T limits be 40OF above any Curve A or Curve B limits at all pressures. Therefore, values for Curve C, the core-critical operating curve, are generated from the requirements of 10CFR50 Appendix G [2] and the Curve A and Curve B limits. 10CFR50 Appendix G [2] also stipulates that, above 20% of the pre-service system hydrostatic test pressure the Curve C temperatures must be either the reference temperature (RTNDT) of the closure flange region plus 160 0 F, or the temperature required for the hydrostatic pressure test, whichever is greater.
For P-T Curves A and B, the initial fluid temperature assumed in Step I is typically taken at the bolt-up temperature of the closure flange minus coolant temperature instrument uncertainty. According to Reference [2], the minimum bolt-up temperature is equal to the limiting material RTNrD of the regions affected by bolt-up stresses. Consistent with Reference [3], the minimum bolt-up temperature shall not be lower than 60'F. Thus, the minimum bolt-up temperature shall be 60'F, the material RTNDT, or other plant specific limit identified by the plant owner, whichever is higher.
For P-T Curve C, when the reactor is critical, the initial fluid temperature is equal to the calculated minimum core critical temperature in the reactor region. Table 1 of Reference [2] indicates that, for a BWR with normal operating water levels, the minimum core critical temperature at the closure flange region is equal to the reference temperature (RTNDT) at the flange region plus 60'F. Thus, the minimum core critical temperature shall be the limiting closure flange region material RTND-+6 0 °F or other plant specific limit identified by the plant owner, whichever is higher.
Nozzles in the beltline introduce stress concentration effects and have the potential to be more limiting than the generic beltline P-T curves. Instrument nozzles in the beltline are analyzed generically in accordance with Reference [4]. There are no additional nozzles in the extended beltline [5]. Nozzles or File No.: 1100445.303 Page 9 of 40 Revision: 0 F0306-01RI:
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discontinuities outside the beltline are considered to be bounded by the upper vessel / feedwater nozzle or bottom head region P-T curves.
3.0 ASSUMPTIONS / DESIGN INPUTS The design inputs and assumptions used to develop the CNS P-T curves are discussed below. Design inputs and assumptions are summarized in the input listings in Appendix A.
The adjusted reference temperature (ART) values in the CNS beltline region are obtained for 32 and 54 EFPY from Reference [5]. Note that the height of the beltline increases in direct proportion with EFPY; this change in the beltline region from initial startup to end of life is referred to as the extended beltline.
The ART value calculations are performed in accordance with Nuclear Regulatory Commission (NRC)
Regulatory Guide 1.99, Revision 2 (RG1.99) [6]. Based on Tables 1 and 2 of Reference [5], the limiting beltline material is the Lower/Intermediate shell plate, which has an ART value of 105.8°F for 32 EFPY and 131.2°F for 54 EFPY.
Non-beltline regions are not subjected to the effects of fluence; therefore, reference temperature (RTNTJT) values are valid substitutions for corresponding ART values. RTNDT values for non-beltline regions are obtained from Reference [7].
The upper bound for the calculated static fracture toughness (Kin) is assumed to be 200 ksi'lin. This limit is assumed based on earlier versions of the ASME Code and is the limit of applicability for linear elastic fracture mechanics, rather than a material property limit.
The inner radius of the RPV at the beitline region is 110.375 inches [8]. The vessel shell thickness is taken as 5.375 inches at the beltline region from the same source. Dimensions for the bottom head radius and the thicknesses are obtained from Reference [8]. The bottom head radius is 110.5 inches and the thickness is 3.188 inches in the thin portion and 6.813 inches in the thick portion.
The GE design pressure is defined in Reference [9] as 1,250 psig. Typically, the pre-service system hydrostatic test pressure is taken as 1.25 times the design pressure, resulting in a value of 1,563 psig.
The instrument uncertainties for both temperature and pressure are given in Reference [10] as follow:
- Reactor Vessel Metal Temperature is bounded by +/- 50F.
" Reactor Vessel Pressure is bounded by +/- 25 psig.
The full vessel height in the RPV is 831.75 inches, as shown in Reference [11]. The normal operating temperature in the RPV is 547°F [9]. However, the water density is conservatively taken at a lower temperature. Thus, the static pressure adjustment due to the pressure head of the water in the RPV is conservatively calculated as 30 psi for all evaluation regions and all temperatures using a water density of 62.4 lbmrft3 . The maximum cool-down rate of the vessel is 100°F/hr per Reference [9].
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According to Section 2.8 of Reference [3], the minimum bolt-up temperature for the RPV shall not be lower than 60'F. Since the RTNDT values for all regions highly stressed by bolt preload are all less than 60°F (in this case, that of the closure flange region: 20'F [7]), the initial assumed fluid temperature in the iterative P-T curve calculation process should be set equal to 60'F minus coolant temperature uncertainty (5°F in this case 110]). However, the minimum containment temperature is 70'F, which bounds the shutdown margin analysis [12]. Therefore, as specified in Section 2.0, the minimum bolt-up and minimum criticality temperature shall not be less than 70'F. A temperature increment of 2°F between subsequent iterations is assumed.
The 70'F initial temperature does not include the additional 60'F add-on margin for Curves A and B that was previously applied. This additional conservatism was required in pre-1971 ASMiE Section III Code, but is no longer required in ASME Section XI, Appendix G [1] or 10CFR50, Appendix G [2].
When the LTR [3] was developed, SI consciously recognized the additional 60'F margin and chose to exclude it, as it is not technically required.
Vessel nozzles are generally incorporated into P-T curve calculations using stress distributions from FEAs and applying them to geometry specific fracture mechanics models. The feedwater nozzle (upper vessel region), instrument (N16) nozzle (extended beltline region), and core differential pressure (CDP) nozzle require this type of analysis due to bounding transients they experience, limiting ART (RTNDT outside beltline) values, and/or stress concentration effects. The core differential pressure CDP nozzle (bottom head region) is analyzed because it is the limiting discontinuity in the thin portion of the bottom head. FEA is performed in Reference [131 for the CDP nozzle.
The feedwater nozzle is the bounding component in the upper vessel because it is a stress concentrator (essentially a hole in a plate) and because it typically experiences more severe thermal transients compared to the rest of the upper vessel region. A two-dimensional finite element model (FEM) of the feedwater nozzle is created as described in Section 2.0 of Reference [14]. The stress distribution acting normal to the postulated 'A thickness crack (or hoop stress distribution) due to a 1,000 psig unit pressure is obtained along a limiting path in the nozzle-to-RPV blend radius [14]. Pressure stress coefficients are obtained from Table 2 of Reference [14] and used to calculate the applied pressure stress intensity factor using Equation 9.
The hoop stress distribution in the feedwater nozzle is also obtained along the same path for a thermal down -shock of 450'F [14]. Stress coefficients are calculated for all time steps in Table 3 of Reference
[14] and used to calculate a thermal stress intensity factor, K1 t, due to the 450'F thermal shock using Equation 4. The maximum Kit for all time steps is used in the evaluation. Because operation is along the saturation curve, the limiting Kit is scaled to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant, which could result in a near zero Kit. Therefore, a minimum K1 t is calculated based on the shutdown transient; scaling of the upper vessel / feedwater nozzle Kit based on the available temperature difference is not allowed below this minimum Kit. The feedwater nozzle shutdown transient is analyzed with the hoop stress distribution given along the same limiting path in Reference [15]. The analysis in Reference [15] provides results for a stress free reference temperature of 70'F as well as 550'F. The File No.: 1100445.303 Page 11 of 40 Revision: 0 F0306-01RI:
choice of stress free reference temperature affects the magnitude of the differential thermal expansion stresses in the component. Both analyses are curve fit with a third order polynomial for all time points, and a thermal stress intensity factor is calculated using Equation 4. The maximum Kit for all time steps considering both stress free reference temperatures is used as the minimum K11 for the upper vessel /
feedwater nozzle. The limiting path defines the nozzle comer thickness to be 5.75 inches [15] and the postulated flaw location at 1/4t to be 1.44 inches.
Vessel nozzles in the beltline are generally incorporated in a similar manner as the feedwater nozzle.
CNS has one set of instrument nozzles in the RPV beltline where the fluence exceeds 1.0 x 1017 n/cm2
[5]. These nozzles introduce stress concentration effects to the beltline plates and must be specifically analyzed. The instrument nozzles, N16A/B, are the only nozzles in the extended beltline region [5].
The instrument nozzles are located in the RPV Region B, which has a maximum temperature of 5220 F during the shutdown transient [9]. The instrument nozzle material is SB-166 [5]. The N16B nozzle is the limiting instrument nozzle in the extended beltline region with ART values for 32 and 54 EFPY of 52.4°F and 70.8 0 F, respectively [5]. The K1 t and KIj-app values are determined generically in accordance with Reference [4], as described in Section 2.0 with the following inputs:
a = 7.6x10-6 (in/in/IF) [17, Table TE-4].
R = 109 (in) [8].
tv = 5.375 (in) [8].
tn = 0.28 (in) [16].
In order to incorporate the N16 nozzle curves into the beltline, a composite beltline curve is developed which bounds each of the two curves (Beltline and NI 6 nozzle).. This composite beltline curve is used to describe the pressure and temperature limits for the extended beltline region.
The CNS bottom head exhibits a variation in thickness for different sections. It is observed that a nozzle exists in the thinner section of the bottom head. Although the nozzle is not ferritic and does not specifically require evaluation, the stress concentration effects of the penetration must be accounted for.
Per Reference [3] a nozzle specific evaluation is performed to ensure that the CDP nozzle is not limiting for any part of the bottom head curve. Initially, only Curve B is analyzed because it utilizes a higher safety factor and also incorporates the effects of through wall thermal stresses. If any portion of the CDP nozzle proves to be limiting for the bottom head, a composite bottom head curve will be created for Curve A, Curve B, and Curve C.
For the CDP nozzle a unit pressure FEA is performed in Reference [13] and a third order polynomial curve-fit is applied to the through-wall stress distribution [13, Table 1]. The Kip value is caluculated using Equation 11. The Kit value for the CDP nozzle is calculated using Equation 3.
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4.0 CALCULATIONS The P-T curves in this calculation were developed using an Excel spreadsheet, which is independently verified for use on a project-specific basis in accordance with SI's Nuclear QA program.
For the feedwater nozzle shutdown thermal transient analysis, the stress free temperature of 70'F is the bounding case. The Kit value is calculated for all time steps with the bounding Kit values shown in the plot of the polynomial curve fit (Figure 1) for time = 6792 seconds.
The polynomial stress coefficients due to pressure, thermal shock, and thermal ramp are given in Table 1 and are applied to Equations 4 and 9 to calculate the stress intensity factors shown in Table 1. The resulting applied pressure stress intensity factor, Kip-app, is 38.9 ksi*Iin, the thermal stress intensity factor due to thermal down shock is 63.5 ksihin, and the thermal stress intensity factor due to thermal ramp is 11.1 ksi/in.
For the generic analysis of the instrument nozzles, the resulting applied pressure stress intensity factor, Kipapp, is 70.6 ksi /in and the thermal stress intensity factor due to thermal ramp is 17.1 ksi/in.
Supporting extended beltline calculations for pressure test (Curve A) and Normal Operation - Core Not Critical (Curve B) are shown in Appendix B.
For the analysis of the core differential pressure nozzle, the resulting applied pressure stress intensity factor, Klp-app is 35.2 ksilin and the thermal stress intensity factor due to the 100*Fihr cooldown rate is calculated using Equation 3 as 1.7 ksi*in.
4.1 Pressure Test (Curve A)
The minimum bolt-up temperature of 70'F minus instrument uncertainty (5'F) is applied to all regions as the initial temperature in the iterative calculation process. The static fracture toughness (KIc) is calculated for all regions using Equation 1. The resulting value of Kie, along with a safety factor of 1.5 is used in Equation 2 to calculate the pressure stress intensity factor (Kip). The allowable RPV pressure is calculated for the beltline, bottom head and upper vessel regions using Equations 6, 7, and 8, as appropriate. For the feedwater nozzle / upper vessel region, the additional constraints specified in Step 6 of Section 2.0 are applied. Final P-T limits for temperature and pressure are obtained from Equations 12 and 13, respectively.
The data resulting from each P-T curve calculation is tabulated. Values for the beltline region at 32 and 54 EFPY are provided in Table 2 and Table 3, respectively. Additionally, more detailed data for the composite beltline are provided in Appendix B. Data for the bottom head region is listed in Table 4, and data for the feedwater nozzle / upper vessel region is presented in Table 5. The data for each region is graphed, and the resulting P-T curves for 32 and 54 EFPY are provided in Figure 2 and Figure 3, File No.: 1100445.303 Page 13 of 40 Revision: 0 F0306-01RI:
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respectively. Additionally, a composite Curve A for the beltline, instrument nozzle, bottom head, and upper vessel regions is graphed and the resulting P-T curves for 32 and 54 EFPY are provided in Figure 4 and Figure 5.
4.2 Normal Operation - Core Not Critical (Curve B)
The minimum bolt-up temperature of 70'F minus coolant temperature instrument uncertainty (5°F) is applied to all regions as the initial temperature in the iterative calculation process. The static fracture toughness (Kic) is calculated for all regions using Equation 1. The thermal stress intensity factor (Kit) is calculated for the beltline plate and bottom head regions using Equation 3, for the feedwater nozzle using Equation 4, and for the instrument (NI 6) nozzle using Equation 5.
The resulting values of K1 c and Kit, along with a safety factor of 2.0, are used in Equation 2 to calculate the pressure stress intensity factor (Kip). The allowable RPV pressure is calculated for the beltline, bottom head, and upper vessel regions using Equations 6, 7, and 8, as appropriate. For the feedwater nozzle / upper vessel region, the additional constraints specified in Step 6 of Section 2.0 are applied.
Final P-T limits for temperature and pressure are obtained from Equations 12 and 13, respectively.
The data resulting from each P-T curve calculation is tabulated. Values for the beltline region at 32 and 54 EFPY are given in Table 6 and Table 7. Data for the bottom head region is listed in Table 8 and data for the bottom head region represented by the CDP nozzle is listed in Table 9. Comparison of these two tables show that data from Table 8 is bounding, therefore the CDP nozzle in the bottom head region is not included as part of the composite curve for Curve A, Curve B, or Curve C. Data for the feedwater nozzle / upper vessel region is presented in Table 10. The data for each region is graphed, and the resulting P-T curves for 32 and 54 EFPY are provided in Figure 6 and Figure 7, respectively.
Additionally, a composite Curve B for the beltline, instrument nozzle, bottom head, and upper vessel regions is graphed and the resulting P-T curves for 32 and 54 EFPY are provided in Figure 8 and Figure 9.
4.3 Normal Operation - Core Critical (Curve C)
The pressure and temperature values for Curve C are calculated in a similar manner as Curve B, with several exceptions. The initial evaluation temperature is calculated as the limiting upper vessel RTNDT that is highly stressed by the bolt preload (in this case, that of the closure flange region: 20'F per Section 3.0) plus 607F, resulting in a minimum critical temperature of 800F. When the pressure exceeds 20% of the pre-service system hydrostatic test pressure (20% of 1,563 psig = 313 psig), the P-T limits are specified as 40°F higher than the Curve B values. The minimum temperature above the 20% of the pre-service system hydrostatic test pressure is always greater than the reference temperature (RTNDT) of the closure region plus 160°F, or is taken as the minimum temperature required for the hydrostatic pressure test. The final Curve C values are taken as the absolute maximum between the regions of the beltline, the bottom head, and the upper vessel.
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~Sfrua lauIk~dly Assoctato, 10 Tabulated overall values of Curve C are provided at 32 and 54 EFPY in Table I I and Table 12, respectively. The corresponding P-T curve plots for 32 and 54 EFPY are given in Figure 10 and Figure 11, respectively.
5.0 CONCLUSION
S P-T curves are developed for CNS using the methodology in Section 2.0 and the design inputs and assumptions defined in Section 3.0. A full set of P-T curves are developed at 32 EFPY and 54 EFPY, for the following plant conditions: Pressure Test (Curve A), Normal Operation - Core Not Critical (Curve B), and Normal Operation - Core Critical (Curve C). Calculations are performed for the beltline, bottom head, feedwater nozzle / upper vessel regions, and the instrument nozzles (NI 6A/B).
Tabulated pressure and temperature values are provided for all regions and EFPY levels in Table 2 through Table 12. The accompanying P-T curve plots are provided in Figure 2 through Figure 11.
File No.: 1100445.303 Page 15 of 40 Revision: 0 F0306-O1RI:
Cj amnrd" hiftgrfiyAsscates, IWO
6.0 REFERENCES
- 1. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Appendix G, "Analysis of Flaws," 2001 Edition including the 2003 Addenda.
- 2. U. S. Code of Federal Regulations, Title 10, Energy, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix G, "Fracture Toughness Requirements," (60 FR 65474, Dec. 19, 1995; 73 FR 5723, Jan. 2008).
- 3. Structural Integrity Associates Report No. SIR-05-044, Revision 1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," June 2011, SI File No. GE-1OQ-401.
- 4. Structural Integrity Associates Report No. 0900876.401, Revision 0, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations."
- 5. Cooper Nuclear Station Calculation No. NEDC07-045, Structural Integrity Associates Calculation No. 1100445.301, Revision 1, "ARTNDT and ART Evaluation."
- 6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
- 7. 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, SI File No. COOP-05Q-202.
- 8. 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.
- 9. Cooper Nuclear Station Drawing Change Notice No. 08-1427, "RPV Thermal Cycles," SI File No. 1100445.203.
- 10. 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.
- 11. General Electric Drawing No. 729E479-B, Revision 0, "Reactor Primary SYS. WTS. & Vols.,"
Sheet I of 3, SI File No. NPPD-06Q-204.
- 12. Email Correspondence between Kenneth Thomas (NPPD) and Eric Houston (SI), Received on 5/18/2011, "RE: DRAFT P-T Curves," SI File No. 1100445.103.
File No.: 1100445.303 Page 16 of 40 Revision: 0 F0306-01RI:
VOW itagrI/y Assoc"t W
- 13. Structural Integrity Associates Calculation No. 1100445.304, Revision 0, "Core Differential Pressure Nozzle Finite Element Model and Stress Analysis."
- 14. NPPD File No. NEDC99-020, Structural Integrity Associates Calculation No. NPPD-13Q-302, Revision 1, "Feedwater Nozzle Stress Analysis."
- 15. Cooper Nuclear Station Calculation No. NEDC99-020, Structural Integrity Associates Calculation No. 1100445.302, Revision 0, "Finite Element Stress Analysis of Cooper RPV Feedwater Nozzle."
- 16. Combustion Engineering, Inc. Drawing No. E-232-242, Revision 7, "Nozzle Details For:
General Electric Corp. APED 218" I.D. BWR," SI File No. 1100445.204.
- 17. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section II, Part D - Properties, 2001 Edition with Addenda through 2003.
File No.: 1100445.303 Page 17 of 40 Revision: 0 F0306-01RI:
Table 1: CNS Polynomial Coefficients for Feedwater Nozzle Stress Intensity Distributions 1 30388 1 -7238.8 1 967.40 1 -82.863 1 138,904 1 S62433 -34071 51810.3 -339.94 I I 63,449 11322 1 -6463.6 931.8 -45.270 1 1 1,105o 1 File No.: 1100445.303 Page 18 of 40 Revision: 0 F0306-01R1:
f, A ssocd tv hWE Tab" leBely Table 2: CNS Beitline Region, Curve A, for 32 EFPY P-T Curve P-T Curve Temperature Pressure 70.00 0 70.00 50 70.00 100 70.00 150 70.00 200 70.00 250 70.00 300 70.00 312 70.00 313 70.00 350 70.00 400 70.00 450 70.00 500 77.37 550 85.28 600 92.11 650 100.07 700 109.08 750 116.71 800 123.34 850 129.19 900 134.42 950 139.16 1000 143.49 1050 147.47 1100 151.16 1150 154.60 1200 157.82 1250 160.83 1300 File No.: 1100445.303 Page 19 of 40 Revision: 0 F0306-OIRI
jSUc 1 09eIIY Associates, Inc."
Table 3: CNS Beitline Region, Curve A, for 54 EFPY P-T Curve P-T Curve Temperature Pressure 70.00 0 70.00 50 70.00 100 70.00 150 70.00 200 70.00 250 70.00 300 70.00 312 70.00 313 70.00 350 70.00 400 74.78 450 86.37 500 95.77 550 103.68 600 114.45 650 125.46 700 134.48 750 142.12 8oo 148.74 850 154.59 900 159.83 950 164.56 1000 168.89 1050 172.87 1100 176.56 1150 180.00 1200 183.21 1250 186.24 1300 File No.: 1100445.303 Page 20 of 40 Revision: 0 F0306-01RF
! Sbictjw ufgdty Assc*n..es, IaR.
Table 4: CNS Bottom Head Region, Curve A, for All EFPY Plant =
Component = (penetrations portion)
Bottom Head thickness, t = inches Bottom Head Radius, R = inches ART = *F ======> All EFPY Kit = (no thermal effects)
Safety Factor =
Stress Concentration Factor = (bottom head penetrations)
Mm=
-Temperature Adjustment 'F (applied after bolt-up, instrument uncertainty)
Height of Water for a Full Vessel = inches Pressure Adjustment psig (hydrostatic pressure head for a full vessel at 70°F)
Pressure Adjustment psig (instrument uncertainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature K1. Kim for P-T Curve P-T Curve (OF) (ksi*inch'12) (ks*i nch112 ) (OF) (psig) 65.0 76.66 51.10 70 0 65.0 76.66 51.10 70 814 67.0 78.43 52.29 72 834 69.0 80.28 53.52 74 855 71.0 82.20 54.80 76 877 73.0 84.20 56.13 78 900 75.0 86.28 57.52 80 923 77.0 88.44 58.96 82 948 79.0 90.70 60.47 84 973 81.0 93.05 62.03 86 1,000 83.0 95.49 63.66 88 1,028 85.0 98.03 65.35 90 1,056 87.0 100.68 67.12 92 1,086 89.0 103.43 68.95 94 1,118 91.0 106.30 70.86 96 1,150 93.0 109.28 72.85 98 1,184 95.0 112.38 74.92 100 1,219 97.0 115.62 77.08 102 1,256 99.0 118.98 79.32 104 1,294 File No.: 1100445.303 Page 21 of 40 Revision: 0 F0306-0IRE
VftrUO , hfitgyAsM, m Table 5: CNS, Upper Vessel Region, Curve A, for All EFPY Plant =
Component =
ART = °F ..... ==> All EFPY Vessel Radius, R = inches Nozzle comer thickness, t' = inches, approximate Kit = (no thermal effects)
Kip-applied ksi*inch1l 2 Crack Depth, a inches Safety Factor =
Temperature Adjustment = °F (applied after bolt-up, instrument uncertainty)
Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)
Pressure Adjustment = psig (instrument uncertainty)
Reference Pressure = psig (pressure at which the FEA stress coefficients are valid)
Unit Pressure psig (hydrostatic pressure)
Flange RTNDT *F ======> All EFPY Gauge P-T P-T Curve Fluid Curve IOCFR50 Temperature KIc Kip Temperature Adjustments
(*F) (ksi*inch 1 2 ) (ksi*inchl12 ) (OF) (psig) 65.0 84.20 56.13 70 0 65.0 84.20 56.13 70 313 67.0 86.28 57.52 110 313 69.0 88.44 58.96 110 1461 71.0 90.70 60.47 110 1499 73.0 93.05 62.03 110 1539 75.0 95.49 63.66 110 1581 77.0 98.03 65.35 110 1625 File No.: 1100445.303 Page 22 of 40 Revision: 0 F0306-01RI.
Table 6: CNS, Beitline Region, Curve B, for 32 EFPY P-T Curve P-T Curve Temperature Pressure 70.00 0 70.00 50 70.00 100 70.00 150 70.00 200 70.22 250 81.92 300 84.36 312 84.55 313 91.39 350 99.35 400 106.22 450 114.04 500 123.11 550 130.80 600 137.46 650 143.34 700 148.60 750 153.35 800 157.70 850 161.70 900 165.40 950 168.84 1000 172.07 1050 175.09 1100 177.96 1150 180.65 1200 183.21 1250 185.65 1300 File No.: 1100445.303 Page 23 of 40 Revision: 0 F0306-0IRI:
Ass7a:s Tait le CBg Table 7: CNS, Beitline Region, Curve B, for 54 EFPY P-T Curve P-T Curve Temperature Pressure 70.00 0 70.00 50 70.00 100 70.00 150 73.31 200 88.61 250 100.32 300 102.75 312 102.95 313 109.79 350 117.76 400 128.32 450 139.43 500 148.52 550 156.21 600 162.86 650 168.74 700 174.01 750 178.75 800 183.10 850 187.09 900 190.79 950 194.25 1000 197.47 1050 200.49 1100 203.35 1150 206.06 1200 208.61 1250 211.05 1300 File No.: 1100445.303 Page 24 of 40 Revision: 0 F0306-O1RI"
rIiugwd t Acates,/nic Table 8: CNS Bottom Head, Curve B for All EFPY Plant =
Component = (penetrations portion)
Bottom Head thickness, t = Inches Bottom Head Radius, R = Inches ART = 'F ======> All EFPY Kit = ksi*inch11 2 Safety Factor =
Stress Concentration Factor = (bottom head penetrations)
Mm =
Temperature Adjustment = 'F (applied after bolt-up, instrument uncertainty)
Height of Water for a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 70°F)
Pressure Adjustment = psig (instrument uncertainty)
Heat Up and Cool Down Rate = °F/Hr Gauge Adjusted Fluid Temperature Pressure for Temperature Kle Kim for P-T Curve P-T Curve (ksi*inch'11)
(°F) (ksi*lnchlI 2 ) (OF) (psig) 65.0 76.66 37.46 70 0 65.0 76.66 37.46 70 582 67.0 78.43 38.35 72 597 69.0 80.28 39.27 74 613 71.0 82.20 40.23 76 629 73.0 84.20 41.23 78 646 75.0 86.28 42.27 80 664 77.0 88.44 43.36 82 682 79.0 90.70 44.49 84 701 81.0 93.05 45.66 86 721 83.0 95.49 46.88 88 742 85.0 98.03 48.15 90 764 87.0 100.68 49.47 92 786 89.0 103.43 50.85 94 810 91.0 106.30 52.28 96 834 93.0 109.28 53.78 98 859 95.0 112.38 55.33 100 886 97.0 115.62 56.94 102 913 99.0 118.98 58.63 104 942 101.0 122.48 60.38 106 972 103.0 126.12 62.20 108 1,003 105.0 129.92 64.09 110 1,035 107.0 133.86 66.07 112 1,068 109.0 137.97 68.12 114 1,103 111.0 142.25 70.26 116 1,140 113.0 146.70 72.48 118 1,178 115.0 151.33 74.80 120 1,217 117.0 156.15 77.21 122 1,258 File No.: 1100445.303 Page 25 of 40 Revision: 0 F0306-0IRI:
Table 9: CNS Bottom Head-CDP Nozzle, Curve B for All EFPY Plant =
Component =
ART = 'F ======> ALL EFPY Heat up/Cool down Rate "F/hr Nominal Vessel Radius, R = nches Vessel Thickness, t, rnches Nozzle Thickness, tn rnches 11 2 Kit = ksi*inch 1 12 Kip-applied = siinch Height of Water for a Full Vessel = rnches Reference Pressure = )sig Pressure Adjustment = sig (hydrostatic pressure head for a full vessel at 70°F)
Pressure Adjustment = sig (instrument uncertainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature Kip for P-T Curve P-T Curve 2
(*F) (ksi*inch11 ) (ksi*inchl1 )2 (OF) (psig) 65.0 76.66 37.46 70.0 0 65.0 76.66 37.46 70.0 1,010 67.0 78.43 38..35 72.0 1,035 69.0 80.28 39.27 74.0 1,062 71.0 82.20 40.23 76.0 1,089 73.0 84.20 41.23 78.0 1,117 75.0 86.28 42.27 80.0 1,147 77.0 88.44 43.36 82.0 1,178 79.0 90.70 44.49 84.0 1,210 81.0 93.05 45.66 86.0 1,243 83.0 95.49 46.88 88.0 1,278 85.0 98.03 48.15 90.0 1,314 87.0 100.68 49.47 92.0 1,352 89.0 103.43 50.85 94.0 1,391 91.0 106.30 52.28 96.0 1,431 93.0 109.28 53.78 98.0 1,474 95.0 112.38 55.33 100.0 1,518 97.0 115.62 56.94 102.0 1,564 File No.: 1100445.303 Page 26 of 40 Revision: 0 F0306-011Rl'
Table 10: CNS Upper Vessel, Curve B for All EFPY Plant =
Component =
ART = -F ======> All EFPY Vessel Radius, R = inches Nozzle comer thickness, t = inches, approximate K., = ksi-inch"2 11 2 KI p-a pplied ksi-inch Crack Depth, a inches Safety Factor =
Temperature Adjustment °F (applied after bolt-up, instrument uncertainty)
Height of Water for a Full Vessel Inches Pressure Adjustment psig (hydrostatic pressure head for a full vessel at 70°F)
Pressure Adjustment" psig (instrument uncertainty)
Reference Pressure psig (pressure at which the FEA stress coefficients are valid)
Unit Pressure = psig (hydrostatic pressure)
Flange RTNTr = °F ======> All EFPY Gauge P-T P-T Fluid Curve Curve Temperature Kle Kip Temperature Pressure 11 2 112
('F) (ksi~inch ) (ksl*inch ) ("F) (psig) 65.0 84.20 10.37 70 0 65.0 84.20 19.46 70 313 67.0 86.28 20.51 140 313 69.0 88.44 19.26 140 440 71.0 90.70 20.06 140 461 73.0 93.05 20.91 140 482 75.0 95.49 21.80 140 505 77.0 98.03 22.73 140 529 79.0 100.68 23.71 140 554 81.0 103.43 24.74 140 581 83.0 106.30 25.82 140 609 85.0 109.28 26.95 140 638 87.0 112.38 28.15 140 668 89.0 115.62 29.40 140 701 91.0 118.98 30.71 140 734 93.0 122.48 32.09 140 770 95.0 126.12 33.54 140 807 97.0 129.92 35.04 140 846 99.0 133.86 36.63 140 887 101.0 137.97 38.30 140 929 103.0 142,25 40.04 140 974 105.0 146.70 41.87 140 1021 107.0 151.33 43.79 140 1071 109.0 156.15 45.80 140 1122 111.0 161.17 47.90 140 1176 113.0 166.39 50.10 140 1233 115.0 171.83 52.39 140 1292 File No.: 1100445.303 Page 27 of 40 Revision: 0 F0306-01RI:
Table 11: CNS Curve C for 32 EFPY Curve A Leak Test Temperature =
Curve A Pressure Plant ==
Unit Pressure = psig (hydrostatic pressure) psig Flange RTNDT = =F P-T Curve P-T Curve Temperature Pressure 80.00 0 80.00 50 80.00 100 80.00 150 94.91 200 110.21 250 121.92 300 124.36 312 180.00 313 180.00 350 180.00 400 180.00 450 180.00 500 180.00 550 180.00 600 180.00 650 183.34 700 188.60 750 193.35 800 197.70 850 201.70 900 205.40 950 208.84 1000 212.07 1050 215.09 1100 217.96 1150 220.65 1200 223.21 1250 225.65 1300 File No.: 1100445.303 Page 28 of 40 Revision: 0 F0306-0IRV
AssocCt Table 2:fCgCIry fW5 Table 12: CNS Curve C for 54 EFPY Curve A Leak Test Temperature = =F Curve A Pressure Plant == psig Unit Pressure = psig (hydrostatic pressure Flange RTNDT = oF P-T Curve P-T Curve Temperature Pressure 80.00 0 80.00 50 80.00 100 91.10 150 113.26 200 128.61 250 140.32 300 142.75 312 180.00 313 180.00 350 180.00 400 180.00 450 180.00 500 188.52 550 196.21 600 202.86 650 208.74 700 214.01 750 218.75 800 223.10 850 227.09 900 230.79 950 234.25 1000 237.47 1050 240.49 1100 243.35 1150 246.06 1200 248.61 1250 251.05 1300 File No.: 1100445.303 Page 29 of 40 Revision: 0 F0306-01RL
~§§Sfroiu nftdgi~y ASSOWNte, looP0 Hoop Stress due to Shutdown Transient 12000 i0000 First (Bounding) Thermal Load Case at Time = 6792 sec 10000 8000m [ Second Thermal Load Case at Time = 6792 sec 8O000 L 1/4 Thickness Location 6000
- 4000 BoundingThermal Load Case Stress Free ReferenceTemperature = 70'F y=-45.27x' + 931.75x2 -6463.6x + 11322 2000 Kt= 11.1 ksl-Vin, R'= 1 0
1 2 4 5
-2000 -1 ý !
- Second Thermal Load Case Stress Free Reference Temperature = 550"F
-4000 y =-45.141x0 + 937.34x 2 - 6507.2x + 10261 K1, = 9.47 ksi-Vi n, R2 = 1
-6000 I Depth Along Limiting Nozzle Path (In)
Figure 1: Feedwater Nozzle Path Stress Distribution File No.: 1100445.303 Page 30 of 40 Revision: 0 F0306-01RI.
~Sftd"IrdirItegtt AgSad&IS, /W9 1,300 1,200 1,100 1,000 900 LuI C0 LuJ 800 10 700 lu 600 50o 400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (CF)
Figure 2: CNS (Hydrostatic Pressure and Leak Test) P-T Curve A for 32 EFPY File No.: 1100445.303 Page 31 of 40 Revision: 0 F0306-0IRIE
- VtSfrucura IntWdY Auocaets, hW 1,300 1,200 1,100 1,000
.J 900 800 w
U 700 z Boo i-400 30 LI) 5o0
- 01. 2700 100 300 200 I0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (CF)
Figure 3: CNS (Hydrostatic Pressure and Leak Test) P-T Curve A for 54 EFPY File No.: 1100445.303 Page 32 of 40 Revision: 0 F0306-01RE
Assofeti Cn SP Ao uer Out o CNS Pressure Test(Composite Curve A), 32 EFPY 1,300 1,200 1,100 1,000 I
LU 900 V) co w
800 0, W I'l561 1.1 700 w
- 3 600
_o I-500 LU 400 70*F, 313p 19 I-Bolt-up Composke Curve A 300 Temp: 110"F, 313 psig 70 FI 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 4: CNS (Hydrostatic Pressure and Leak Test) Composite P-T Curve A for 32 EFPY File No.: 1100445.303 Page 33 of40 Revision: 0 F0306-0IR1:
!V ,wcgwhftqdiY A W
/soCAft CNS Pressure Test(Composite Curve A), 54 EFPY 1,300 1,200 1,100 1,000 a.
-_ 900 co w
800 02-700 z
600 a2 500 LI 400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)
Figure 5: CNS (Hydrostatic Pressure and Leak Test) Composite P-T Curve A for 54 EFPY File No.: 1100445.303 Page 34 of 40 Revision: 0 F0306-OIRL
!Vftn"bft*,AMddwhMe 1,300 1,200 1,100
- 1,0oo
-j co 900 (J
i-o 800 Uj C'
z 700 I-
- 600 m
to 500 w
'U 400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
Figure 6: CNS P-T Curve B (Normal Operation - Core Not Critical) for 32 EFPY File No.: 1100445.303 Page 35 of 40 Revision: 0 F0306-0IRL
!C~SNW fru ndly t~I Assot /W 1,300 1,200 1,100 1,000
,a
-j U) 900 U) uw 0 800 I-700 Uj 600 LI-w (L 500 400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)
Figure 7: CNS P-T Curve B (Normal Operation - Core Not Critical) for 54 EFPY File No.: 1100445.303 Page 36 of 40 Revision: 0 F0306-01RI'
- Ue~dUFW Mftgdyt Aasocits, .n.
CNS Normal Operation- Core Not Critical (Composite Curve B), 32 EFPY 1,300 1,200 1,100
_ 1,000 a.
-j 900 iul 70D D
o 600 I-6*00 IjJ (0
.L}
400 300 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)
Figure 8: CNS (Normal Operation - Core Not Critical) Composite P-T Curve B for 32 EFPY File No.: 1100445.303 Page 37 of4 Revision: 0 F0306-01RI:
jjsftwui IuftegY Assoiates,Ina?
CNS Normal Operation - Core Not Critical (Composite Curve B), 54 EFPY 1,300 1,200 1,100 1,000
.J L, 900 U,
o I-800 z, 700 I-i600 LU Vf) 0 500 Ue' a,
400 300 200 100 0
0 20 40 60 80 100 120 140 160 150 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (CF)
Figure 9: CNS (Normal Operation - Core Not Critical) Composite P-T Curve B for 54 EFPY File No.: 1100445.303 Page 38 of 40 Revision: 0 F0306-01oRI
~S/mudfw h*WEy A*ocatha, a."
CNS Normal Operation - Core Critical Composite Curve C, 32 EFPY 1,300 1,200 "j I T 1,100 1,000 - -
-900 w
w
>800 - -
0 I-
< 700 tSDIF, 6711psig r z ,
5600 0)500
~fl w
I-400
'" 12rF, 313 psig Minimum Core 300 Critical 18WF, M3 PWg Temperature:
- 80*F 200 r7 0-F, S163PSID 100 M I I "13 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)
Figure 10: CNS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY File No.: 1100445.303 Page 39 of 40 Revision: 0 F0306-OIRI:
jSfti*uW hrtifu* Assodates, An-CNS Normal Operation - Core Critical Composite Curve C, 54 EFPY 1,300 1,200 1,100
- 4,000
.9
'i 900 C,)
Ili 0
0r 800
> 0 1-.
w
. 700 z
00 600 w
o500 M2 180IF, 903 psig 0.
400 300 4:
i, !- !
Minimum Core 1 0 F13ps1ig T71 Critical i 200 LTemperature:&80°F 80,131 pl 100 0-0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)
Figure 11: CNS P-T Curve C (Normal Operation - Core Critical) for 54 EFPY File No.: 1100445.303 Page 40 of 40 Revision: 0 F0306-OIRl:
VajmrntWuOr htgrAsOWIS, W9IC.
APPENDIX A:
P - T CURVE INPUT LISTING File No.: 1100445.303 Page A- I of A-4 Revision: 0 F0306-O1RI
InaG E warInp 3 V2 u t u t Assadtsi 32 EFPY Input Listing:
Instrument Uncertainty
Reference:
Reactor Vessel Metal Temp 5 *F [101 Reactor Vessel Pressure 25 psig [10]
Geometry Vessel Radius 110.375 in. [8]
Vessel Shell thickness 5.375 in. [8]
Bottom Head Thickness 6.8125 in. [8]
Bottom Head Radius 110.5 in. [8]
Feedwater Nozzle Thickness 10.87 in. [15]
Bottom Head Thickness (CDP Nozzle) 3.1875 in. [8]
Core Differential Pressure Nozzle Thickness 0.281 in. [16]
ART/RTNDT 32 EFPY Umiting Beltine 105.8 'F [5]
Umiting Bottom Head 28 *F [7]
Umiting Upper Vessel (Feedwater) RTNDT 20 *F [7]
Flange Material (Bolt-up) RTNDT 20 *F [71 Limiting Instrument (N16) Nozzle 52.4 *F [5]
Safety Factor/Stress Concentration Factor Core Not Critical (Curve B) Core Critical (Curve C) 2 [31 Pressure (Curve A) t.5 [31 Lower Penetrations (SCF) 3 [3]
Limiting Instrument (N16) Nozzle (SCF)
During Pressure Test (near isothermal conditions) 0 ksivin [3]
Water Density 62.4 lblft3 Assumed Pressure 1250 psig [9]
Full Water Elevation (pressure head) 831.75 in [11]
Hydrostatic Test Pressure 1563 psig Calculated Static Head Pressure Adjustment 30.0 psig Calculated Assumed Temperature Bolt Up Temperature 70 °F [12]
Increment 2 °F Assumed Rate of Temp Change Heat Up and Cool Down Rate 100 °Flhour [8]
File No.: 1100445.303 Page A-2 of A-4 Revision: 0 F0306-01RI
54 EFPY Input Listing:
instrument Uncertainty
Reference:
Reactor Vessel Metal Temp 5 IF [10]
Reactor Vessel Pressure 25 psig [10]
Geometry Vessel Radius 110.375 in. [8]
Vessel Shell thickness 5.375 in. [8]
Bottom Head Thickness 6.8125 in. [8]
Bottom Head Radius 110.5 in. [8]
Feedwater Nozzle Thickness 10.87 in. [15]
Bottom Head Thickness (CDP Nozzle) 3.1875 in. [8]
Core Differential Pressure Nozzle Thickness 0.281 in. [16]
ART/RTNDT 54 EFPY Limiting Beltine 131.2 =F [5]
Limiting Bottom Head 28 °F [7]
Limiting Upper Vessel (Feedwater) RTNoT 20 °F [7]
Flange Material (Bolt-up) RTNoT 20 °F [7]
Limiting Instrument (N16) Nozzle 70.8 OF [5]
Safety Factor/Stress Concentration Factor Core Not Critical (Curse B) Core Critical (Curse C) 2 [3]
Pressure (Curse A) 1.5 [3]
Lower Penetrations (SCF) 3 (3]
Limiting Instrument (N16) Nozzle (SCF) -
Kt During Pressure Test (near isothermal conditions) 0 ksivin [3]
Water 3
Density 62.4 lb/ft Assumed Pressure 1250 psig [9]
Full Water Elevation (pressure head) 831.75 in [11]
Hydrostatic Test Pressure 1563 psig Calculated Static Head Pressure Adjustment 30.0 psig Calculated Assumed Temperature Bolt Up Temperature 70 'F [12]
Increment 2 °F Assumed Rate of Temp Change Heat Up and Cool Down Rate 100 °F/hour [8]
File No.: 1100445.303 Page A-3 of A-4 Revision: 0 F0306-0IRI
V 1Sbfurn! Ifutegrfy MOWN esnG.
Inputs and Outputs Listing ofN 16 Instrument Nozzle for 32 and 54 EFPY Vessel ID (Nominal) 218.0 [8]
Vessel thickness, t, (in) 5.375 [8]
Nominal Vessel Radius (in) 109.0 [8]
Nozzle OD (in) ____ [16]
Nozzle ID (in) 1.94 [16]
Nozzle Thickness, t, (in) 0.28 Calculated Thermal Expansion, a (in/in/°F) 7.6E-06 [17]
File No.: 1100445.303 Page A-4 of A-4 Revision: 0 F0306-OIRI
APPENDIX B:
BOUNDING BELTLINE SUPPORTING ANALYSIS File No.: 1100445.303 Page B-I of B-13 Revision: 0 F0306-OIRI
j -ajoi Iftfgty AssocGWa. Itnrc Table B-i: CNS, Beltline Region, Curve A, for 32 EFPY Plant =
Component =
Vessel thickness, t = theas Vessel Radius. R = ches ART = !=====> 32 FFPY KjT= nothermsi effects)
Safety Factor =
Mm=
Temperature AdPustment = F(applied after betx-up,Instrument uncertatnty)
Height of Water for a Full Vessel = ches Pressure Adjustment = sig (hydrostatic pressure head for a fullvessel at 70'F)
Pressure Adjustment rOg(instrument uncerlainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature K .5m
('F) (k'lnchh ) (Iklinch"') rF) (psig .
o,.u 42.37 28.25 70.0 0 65.0 42.37 28.25 70.0 588 67.0 42.74 28.50 72.0 591 69.0 43.13 28.75 74.0 597 71.0 43.54 29.02 760 603 73.0 43.96 29.31 78.0 610 75.0 44,40 29.60 80.0 616 77.0 44.88 29.90 82.0 623 79.0 45.33 30.22 84.0 630 81.0 45.83 30.55 86.0 638 8&0 46.34 30.89 88.0 646 85.0 48.88 31.25 90.0 654 87,O 47.44 31.62 92.0 662 89.0 48.02 32.01 94.0 671 91.0 48.62 32.41 96.0 680 93.0 49.25 32.83 98.0 690 95.0 49.91 33.27 100.0 700 97.0 50.59 33.73 182-0 710 99.0 51.30 34.20 104.0 721 101.0 52.04 34.69 1060 732 103.0 o28o 35.20 108.0 743 105.0 58360 35.74 110.0 758 107.0 54.44 36.29 112.0 768 109.0 55.30 36.87 114.0 781 111.0 56.21 37.47 116.0 795 113.0 57.15 38.10 118.0 809 115.0 58.12 38.75 120.0 824 117.0 59.14 39.43 1220 839 119.0 60.20 40.13 124.0 855 121.0 61.30 40.87 125.0 872 123.0 62.45 41.63 128.0 889 125.0 63.64 4243 130.0 907 127.0 64.88 43.26 132.0 928 129.0 68.18 44.12 134.0 948 131.0 67.52 45.01 136.0 966 133.0 68.92 45.95 138.0 987 135.0 70.38 46.92 140.0 1,009 137.0 71.90 47.93 1420 1.032 139.0 73.48 48.98 144.0 1,056 141.0 75.12 50.08 146.0 1,081 143.0 76.83 51.22 148.0 1,107 145.0 78.61 5241 150.0 1,134 147.0 80.47 53.64 1520 1,162 149.0 82.39 54.93 154.0 1,191 151.0 84.40 58.27 15&0 1,221 153.0 88.49 57.66 158.0 1,253 155.0 88.67 59.11 160.0 1,286 157.0 90.93 60.62 162.0 1,320 159.0 M0.29 6219 164.0 1,3586 161.0 95.74 63.83 166.0 1,393 163.0 98.29 65.53 168.0 1,431 165.0 100.95 67.30 170.0 1,472 167.0 103.71 69.14 172.0 1,513 169.0 10659 71.06 174.0 1,557 171.0 109.58 7206 176.0 1,602 File No.: 1100445.303 Page B-2 of B-13 Revision: 0 F0306-OIRI
S",UuInd IfgrAr AssoW.ps, hIn, Table B-2: CNS, Instrument (N16) Nozzle Beltline Region, Curve A, for 32 EFPY Plant =
Component m ART = "F ======> 32 EFPY Heat up/Cool down Rate Nominal Vessel Radius. R - ches Vessel Thickness, t, ches Nozzle Thickness, t. ches Themial Expansion Coefficlent, o lfirVIF 12 Kit = ksiinch "
Kip.epphld=
Height of Water for a Full Vessel = ches Pressure Adjustment = psig (hydrostatic pressure head for a full vessel at 701F)
Pressure Adjustment = psig (instrument uncertainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature KIp K1 forP-TCurve P-TCurve
('F) (kstiinch..) (ksi'inch") (JF) (psig) 65.0 59.88 39.92 70.0 0 650 59.88 39.92 70.0 510 67.0 60.96 40.64 72.0 520 69.0 62.10 41.40 74.0 531 71.0 63.28 42.18 76.0 542 73.0 64.50 43.00 78.0 554 75.0 65.78 43.85 80.0 566 77.0 67.11 44.74 82.0 578 79.0 68.50 45.66 84.0 591 81.0 69.94 46.62 86.0 605 83.0 71.44 47.62 88.0 619 85.0 73.00 48.66 90.0 634 87.0 74.62 49.75 92.0 649 89.0 76.31 50.87 94.0 665 91.0 78.07 52.05 96.0 682 93.0 79.90 53.27 98.0 699 95.0 81.81 54.54 100.0 717 97.0 83.79 55.86 102.0 736 99.0 85.86 57.24 104.0 755 101.0 88.00 58.67 106.0 775 103.0 90.24 60.16 108.0 797 105.0 92.57 61.71 110.0 818 107.0 94.99 63.33 112.0 841 109.0 97.51 65.01 114.0 865 111.0 100.14 66.78 116.0 890 113.0 102.87 68.58 118.0 916 115.0 105.71 70.48 120.0 943 117.0 108.67 72.45 122.0 970 1190 111.75 74.50 124.0 999 121.0 114.96 76.64 126.0 1,030 123.0 118.30 78.86 128.0 1,061 125.0 121.77 81.18 130.0 1,094 127.0 125.38 83.59 132.0 1,128 129.0 129.14 86.10 134.0 1,164 131.0 133.06 88.71 136.0 1,201 133.0 137.14 91.42 138.0 1,239 135.0 141.38 94.25 140.0 1,279 137.0 145.79 97.19 142,0 1.321 139.0 150.39 100.26 14.4.0 1.364 141.0 155.17 103.45 146.0 1,409 143.0 160.15 106.77 148.0 1.456 145.0 165.33 110.22 150,0 1,505 147.0 170.72 113.81 152.0 1,556 149.0 176.33 117.56 154.0 1,609 File No.: 1100445.303 Page B-3 of B-13 Revision: 0 F0306-OIRI
V sfmaturuIalgrl~scaa Table B-3: CNS, Beltline Region, Curve A, for 54 EFPY Vessel
... icknetiS, I = inches Vessel Redo*., R - iches ART T == " ,t 94 8FPy Vt- = . (tinothemal abuts)
Safety Fautor Oemperaltre Auslmer- 7i (Tappoed taerbolt-tp,tIrumnt unuetiairrly)
Height of Wale: for a Full Vesnal - rIches Pressure Adjustment - psig frydrstali pressure head for a fulltas=elat 70"F)
PressmrreAdjuslarmt a pig famrtura uncertainty)
Gang. Adjusted Fluid Temperature Presure IOr Temperature Ku l. a ferP-TCure P-TroNe f') (kal'ia~h")= (W-1i-h-) ('F) fP*ia) 65.0 38.72 25.81 70.0 0 65.0 39.72 26.81 70.0 5m0 67.0 38.94 25.90 7?.0 534 69.0 39.18 26.12 74.0 537 71.0 39.42 26.28 79.0 541 73.0 39.67 26.45 78.0 045 75.0 39.94 26,63 90.0 549 77.0 40.21 26.81 82.9 553 78.0 40.00 27.00 84.0 557 81.0 40.80 27.20 08.0 502 93.0 41.11 27.40 88.0 527 85.0 41.43 27.62 00.0 571 a7.0 41.77 27.04 92.0 5O 08.0 42.12 28.08 94.0 382 81.0 42.48 28.32 96.0 587 83.0 42-85 28.57 98.0 593 95.0 43.25 20.83 100.0 309 67.0 43.68 26.11 102.0 ms 08.0 44.09 26.36 104.0 612 101.0 44.53 29.69 106.0 618 103.0 45.00 30.00 108.0 625 105.0 45.49 30.32 110.0 633 107.0 45.98 30.65 112.0 640 109.0 46.50 31.69 114.0 648 111.0 47.04 31..30 116.0 656 113.0 47.61 31.74 118.0 665 115.0 48.20 32.12 126.0 674 117.0 48.81 32.54 122.0 883 1`19.0 49.44 3206 124.0 693 121.0 50.11 33.41 126.0 703 123.0 50.00 33287 128.0 713 125.0 51.52 34.34 130.0 724 127.0 52.26 34.84 132'0 732 129.0 53.04 35.36 134.0 747 131.0 63.90 35.80 136.0 759 133.0 04.69 30.46 138.0 772 135.0 55.57 37.05 140.0 789 137.0 56.48 37.66 142.0 798 139.0 57.43 38.29 144.0 814 141.0 53.42 38.95 ¶460.0 528 143.0 59.45 39.64 148.0 844 145.0 60,52 40.35 150.0 860 147.0 61,64 41.09 152.0 877 149.0 62.83 41.87 154.0 895 151.0 04.01 42.87 158.0 013 153.0 65.,27 43.51 158.0 932 155.0 68057 44.38 100.0 952 157.0 67,94 45.2.9 162.0 972 159.0 69035 46.24 144.0 994 161.0 70.83 47.22 166.0 1,010 1632. 72.36 48.24 168.0 1,039 165.0 73.90 49.31 170.D 1.063 167.0 70.63 50.42 172.0 1,065 169.0 T7.36 51.57 174.0 1.115 171.0 79.16 52.77 170.0 1.142 173.0 81.04 54.02 178.0 1,179 175.0 82.99 55.33 180,0 1,200 177.0 80.02 50.68 182.0 1,231 179.0 87.13 58.09 184.0 12'3 181.0 69.34 59.56 108.0 1,208 162.0 91.63 61.8 108.0 1,331 165.0 94.01 62.87 iao.o 1,367 187.0 96.49 64.33 192.0 1,404 189.0 09.08 66.05 194.0 1.443 191.0 101.76 07.84 196.0 1.484 193.0 104.56 69.71 198.0 1.526 190.0 107.40 71.65 2000 1570 File No.: 1100445.303 Page B-4 of B-13 Revision: 0 F0306-OIRI
tiS#Uc rJ9h Iutuu*iAsoda~s, nOP Table B-4: CNS, Instrument (N16 Nozzle Beltline Region, Curve A, for 54 EFPY Plant =
Component =
t ART - °F . .... > 54EFPY Heat uplCool down Rate - 'Fibr Nominal Vessel Radius, R = inches Vessel Thickness, t, inches Nozzle "hckness, t. inches ThermalExpansion Coeffcient, a inrs/.F K11= kwlillehw/0 Kl 11 = ksi'inch' Height ofWater fer a Full Vessel = inches Pressure Adjustment = psig (hydrostatic pressiue head for a furl esel at 70°F)
Pressure Adjustment = psig gnstrument uncertainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature K. Kp for P-T C urve P-T Curve J'Fl (ksI'nch'r) (ktlfinch,") (°F) (psig) 65.0 51.66 34.44 70.0 0 65.0 51.66 34.44 70.0 432 67.0 52.42 34.94 72.0 440 69.0 53.20 35.47 74.0 447 71.0 54.02 38.01 76.0 455 73.0 54.87 36.58 78.0 463 75.0 55.75 37.17 80.0 471 77.0 56.67 37.78 82.0 480 79.0 57.63 38.42 84.0 489 81.0 58.63 39.08 86.0 498 83.0 59.66 39.78 88.0 508 85.0 60.74 40.50 90.0 518 87.0 61.87 41.25 92.0 529 89.0 63.04 42.03 94.0 540 91.0 64.26 42.84 98.0 551 93.0 65.52 43.68 98.0 95.0 66.84 44,56 100,a 576 97.0 68.21 45.48 12.0 889 99.0 69.64 46.43 104.0 602 101.0 71.13 47.42 106.0 616 103.0 72.68 48.45 108.0 631 105,0 74.29 49.53 110.0 646 107.0 75.97 50.64 112.0 662 109.0 77.71 51.81 114.0 678 111.0 79.53 53.02 116.0 695 113.0 81.42 54.28 118.0 713 115.0 83.39 55.59 120.0 732 117.0 85.44 56.95 122.0 751 119.0 87.57 58.38 124.0 771 121.0 89.79 59.86 128.0 792 123.0 92.10 61.40 128.0 814 125.0 94.50 63.00 130.0 837 127.0 97.00 64.67 132.0 868 129.0 99.61 68.40 134.0 885 131.0 102.32 68.21 136.0 910 133.0 105.14 70.09 138.0 937 135.0 108.07 72.05 140.0 965 137.0 111.13 74.08 142.0 994 139.0 114.31 76.20 144.0 1,024 141.0 117.62 78.41 146.0 1,055 143.0 121.06 80.71 148.0 1,087 145.0 124.65 83.10 150.0 1,121 147.0 128.38 85.59 152.0 1.158 149.0 132.28. 88.18 154.0 1,193 151.0 136.31 90,87 156.0 1,231 153.0 140.62 93.68 158.0 1,271 155.0 144.90 96.60 160.0 1,312 157.0 149.45 99.64 162.0 1,355 159.0 154.20 102.80 164.0 1.400 161,0 159.14 106.09 166.0 1.447 163.0 164.28 109.52 168.0 1.495 185,0 189.62 113.08 170.0 1.548 187.0 175.19 116.80 172.0 1,598 File No.: 1100445.303 Page B-5 of B-13 Revision: 0 F0306-O1RI
Mqr MAssoOa InO."
Table B-5: CNS, Beltline Region, Curve B, for 32 EFPY Plant =
Component =
Vessel thickness,t = inches Vessel Radius, R = inches ART kF= > 32EFPY Safety Factor-Tenpersture Adjustment F (applted after bolt-up, Instrument uncerteinty)
Height ofWater lora FuLtVessel = inches Pmrssure Adtfusmnt me= r hiydrostaticpressure head for a full assel at 7AM)
Pressurs Adjlstmerlt Pig nascnent uncertainty)
Heal Up and Coa.on Rote = 1 0 'Ft-r Gauge Adjusaad Fluid Temperatures Presse for Temperature Kb s toeP-T Curve P-T COrne (IF) (k ntnch') "F F)i (patg) 65.0 42.37 17.99 70.0 0 65.0 42.37 17.09 70.0 353 67.0 42.74 16.16 72.0 357 69.0 43.13 18.37 74.0 362 71.0 43.04 18.58 70.0 366 73.0 43.96 18.79 7M.0 371 76.0 44.40 19.01 80.g 376 77.0 44.00 19.24 02. 361 70.0 .-. 33 16.47 84.0 387 81.0 45.83 16.72 66.0 392 03.0 48.34 19.98 60.0 396 06.0 46.88 20.25 90.0 404 07.0 47.44 20.53 02.0 411 69.0 48.02 20.82 94.0 417 91.0 48.62 21.12 66.0 424 93.0 40.25 21.43 98.0 431 06.0 40.91 21.76 10.0 439 97.3 50.59 22.10 102.0 446 99.0 51.30 22.46 104.0 404 101.0 52.04 22.63 166.0 463 103.0 52.60 23.21 106.0 471 105.0 53.60 23.61 110.0 481 107.0 54.44 24.03 112.0 436 Ou49 nasa £.. 0 a i,1.2.
I18u 111.0 36.21 24,91 116.0 510 113.0 57.15 25.38 116.0 521 115.0 58.12 25.87 123.0 532 117.0 59.14 25,38 1[2.0 543 119.0 60.20 26.01 124.0 555 121.0 61.30 27.46 126.0 560 123.0 02.45 26.03 128.0 Sa1 12S.0 03.64 28.63 130.0 594 127.0 04.80 29.25 132.0 506 129.0 66.10 29.90 134.0 623 131.0 57.52 30.57 136.6 638 133.0 88.92 S1.27 138.0 004 135.0 70.38 32.00 140.0 671 137.0 71.90 32.76 142.0 606 139.0 73.46 33.55 144.6 70.
141.0 70.12 34.37 146.0 725 143.0 76.83 35.22 148.0 744 145.0 78.61 36.11 150.0 764 147.0 80.47 37.04 10ZO 705 14900 82.39 38.01 154.0 807 151.0 84.40 39.51 158.0 830 153.0 86.49 40.06 158.0 B64 155.0 88.67 41.14 160.0 078 157.0 50.93 42.27 152.6 04 159.0 93.29 43.45 164.0 501 161.0 95.74 44.68 166.0 O5f 163.0 96.29 45.95 168.0a 97 165.0 150.06 47.28 170.6 1,017 167.0 103.71 48.66 172.6 1,049 109.0 16,069 50.10 174.0 1,081 171.0 106.58 51.60 176.6 1.115 173.0 112.70 .5316 178.0 1,151 175.0 115.95 54.78 162.9 1,180 177.0 119.32 36.47 102.0 1,226 179.0 122.84 68.23 184.0 1,256 1i1.o 132.50 00.36 186.0 1,307 163.0 130.30 61.96 1880. 1,350 1865.0 134.27 3.094 150.6 1,395 187.0 138.39 66.00 12.6 1.442 189.0 142.60 668.1 194.6 1,491 191.0 147.15 70,38 196 1,542 193.0 151.80 72.71 160.6 1,504 File No.: 1100445.303 Page B-6 of B-13 Revision: 0 F0306-OIRI
~jj~SbvahnhIufuudiv Auocletosi maP Table B-6: CNS, Instrument (N16) Nozzle Beltline Region, Curve B, for 32 EFPY Plant = w Component :
ART F =====> 32 EFPY Heat up/Cool down Rate .F/hr Nominal Vessel Radius, R = icheas Vessel Thickness, t, nches Nozzle Thickness, t. tcheas Thermal Expansion Coefficient. a /ir-F K1t= st-inch"'
2 Ktp.*pplled
= si-inch" Height of Water for a Full Vessel = ches Pressure Adjustment = sig (hydrostatic pressure head for a full essel at 70'F)
Pressure Adjustment = slg (instrument uncertainty)
Gauge Adjusted Fluid Temperature Pressure for Temperature K*. KP for P-T Curve P-T Curve 1
(°F) (ks*inch") (ksi'tnch "9) ('F) (palg) 65.0 59.88 21.39 70.0 0 65.0 59.88 21.39 70.0 248 67.0 60,96 21.94 72.0 255 69.0 62.10 22.50 74.0 263 71.0 63.28 23.09 76,0 272 73.0 64.50 23.71 78.0 281 75.0 65.78 24.35 80.0 290 77.0 67.11 25.01 82.0 299 79.0 68.50 25.70 84.0 309 81.0 69.94 26.42 86.0 319 83.0 71.44 27.17 88.0 330 85.0 7.00 27.95 90.0 341 87.0 74.62 28.77 92.0 352 89.0 76.31 29.61 94.0 364 91.0 78.07 30,49 96.0 377 93.0 79.90 31.41 98.0 389 95.0 81.81 32.36 100.0 403 97.0 83.79 33.35 102.0 417 99.0 85.86 34.38 104.0 432 101.0 88.00 35.46 106.0 447 103.0 90.24 36.58 108.0 483 105.0 92.57 37.74 110.0 479 107.0 94.99 38.95 112.0 496 109.0 97.51 40.21 114.0 514 111.0 100.14 41.52 116.0 533 113.0 10287 42.89 118.0 552 115.0 105.71 44.31 120.0 572 117.0 108.57 45.79 122.0 593 119.0 111.75 47.33 124.0 615 121.0 114.96 48.93 126.0 638 123.0 118.30 50.60 128.0 661 125.0 121.77 52.34 130.0 686 127.0 125.38 54.15 132.0 711 129.0 129.14 56.03 134.0 738 131.0 133.06 57.99 136.0 786 133.0 137.14 60.02 138.0 795 135.0 141,38 62.14 140.0 825 137.0 145.79 64.35 142.0 856 139.0 150.39 66,65 144.0 888 141.0 155.17 69.04 146.0 922 143.0 160.15 71.53 148.0 957 145.0 165.33 74.12 150.0 994 147.0 170.72 76.82 152.0 1,032 149.0 176.33 79.62 154.0 1,072 151.0 182.17 82.54 156.0 1.113 153.0 188.25 85.58 158.0 1,156 188.0 194.58 88.75 160.0 1,201 File No.: 1100445.303 Page B-7 of B-13 Revision: 0 F0306-01R1
jjSbvotwW Inhuurly Assactuos, ft@
Table B-7: CNS, Beltline Region, Curve B, for 54 EFPY Pil.n Vessel t.hokness, t = B htlle V.1.,! Radius. R - 0 hto ARTh 3 -F.--- O4 EFPY Salfety adclor 0 Temperature AQUstenof = 8 "F (appe nr bolt-up, Insttument unceattnlry)
Htight of wator- era Fua Vessel inches Pressure AttLttamo = ., polo ftyh MOst.tla pre-ss-e head for. *lt U sed at 7WPF)
Pressure AsjY-tmAst Lb M lo(ht etear oetoisty) pPag HI-t Up an Coal 0ows Rits - 6 "FJHr GOago Adj.nd Fluid Tempo rtatue Proo.-r. f.r Tetparoture it Ki far P-T Cars. P.T Curve rt1 (llrlaith-) Ot*sflahnch 1r*) tepeo) 05.0 38.72 10.17 70.0 0 85.0 38.72 10.17 70.0 312 67.0 30.04 10.28 72.0 314 6900 30.18 16.40 74.0 317 71.0 39.42 16.52 70.0 320 73.0 39,807 10.03 78.0 323 75.0 39.54 10.78 08.0 328 77.0 40.21 168.1 02.0 329 75.0 40.0 17.06 84.0 332 81.0 40.00 17.21 08.0 335 63.0 41.11 17.30 00.0 330 85.0 41.43 17.02 00.0 342 07.0 41.77 17,05 92.0 340 00.0 42.12 17,07 94.0 350 81.0 42.40 18.05 00.0 354 83.0 42.850 10.24 08.0 359 83.0 43.25 10.43 100.0 303 97.0 43.60 10.04 102.0 360 98.0 44.09 10.05 104.0 373 101.0 44.53 10.08 150N0 378 103.0 45.00 15.31 l8a.0 303 105.0 45.45 19.55 110.0 350 107.0 45.00 19.:0 112.0 304 111.0 47.:4 20.33 110.0 400 113.0 47.01 20.01 118.0 413 115.0 48.25 20.01 120.0 410 117.0 48.01 21.21 122.0 428 119.0 49.44 21.53 124.0 433 121.0 50.11 21.00 128.0 441 123.0 50.00 23.21 128.0 449 125.0 51.52 22.57 130.0 457 127.0 52.30 22.94 132.0 465 120.0 53.04 23.33 134.0 474 131.0 53.05 23.73 130.0 403 133.0 54.80 24.16 138.0 453 135.0 55.57 24.50 140.0 503 137.0 56.48 25.05 142.0 513 135.0 57.43 25.53 144.0 524 141.0 58.42 26.02 148.0 525 143.0 59.45 28.03 148.0 047 145.0 60.52 27.07 150.0 558 147.0 61.04 27.03 1520 572 140.0 62.80 20.21 154.0 35 151.0 64.01 28.81 156.0 599 102.0 65.27 29.44 158.0 613 152.0 65.57 30.10 160.0 6Z0 157.0 07.04 30.70 162.0 043 159.0 00.35 31.49 164.0 059 t10.a 70.03 32.22 100.0 76 163.0 73.30 32.00 165.0 683 1:5.0 73.98 33.79 170.0 711 107.0 75.03 34.02 172.0 730 16950 T7.38 35.45 174.0 75o 171.0 70.10 20.39 170.0 770 173.0 01.04 27.33 178.0 792 175.0 02.5 2 32.30 100.0 014 177.0 85.02 39.32 182.0 a37 179.0 07.13 40.38 104.0 861 101.0 80.34 41.40 10,60 000 183.0 901.63 42.02 10,0 5 12 185.0 94.01 45.01 100.0 039 187.0 0.49 45.05 192.0 907 189.0 "0.08 46.35 104.0 5ON 191.0 101.76 47.05 156.0 1,027 193.0 104.56 49.09 198.0 1.058 105.0 107.40 50.55 200.0 1,092 157.0 110.51 52,00 202.0 1.126 100.0 113.60 53.64 204.0 1.102 201.0 118.04 55.28 206.0 1.109 203.0 120.38 26.9 20t.0 1:230 205.0 123.92 09.77 210.0 1,270 207.0 127.62 00.82 212.0 1.320 209.0 131.48 62.55 214.0 1.364 211.0 135.49 64.55 216.0 1.400 213.0 130.00 06.04 218.0 1.457 215.0 144.01 00.01 335.0 1.508 217.0 148.53 71.07 222.0 1.557 File No.: 1100445.303 Page B-8 of B-13 Revision: 0 F0306-OIRI
W, d ASWOWA ka&
Table B-8: CNS, Instrument (N16) Nozzle Beltline Region, Curve B, for 54 EFPY Component =
ART- *F =.=c.> S4 EFPY Heatl upCeoI dse Rate Nominal Vessel RoI~us. R 0 Inches Vessel Thickness. 1 Inches Nozzle Thickness, t. inches ThemnalExpansion Coefliclan, e irk'siF Height of Water for e Ful Vessel Inches PressureAdjustment psig ("ydrdstatic plosuor head for a fulltesel at 7011F)
Pressure Adj..tment = psig Onlnst*rnt uncertainty)
Gauge Adjusted Fluid Temperaone Pressure fOW Temnpemrle Kle K or P-T Cume P-T Curve
( F) (kstnthn (kinch"), MF' (polg) 65.0 51.60 17.29 70.0 0 5g.0 1.66 17.29 70.0 10g 87,0 52.42 17.86 72.0 195 g9.0 53.20 18.06 74.0 201 71.0 54.02 18.40 78.0 208 73.0 54.87 1n,19 70.0 212 75.0 55,75 18.33 80.0 219 77.0 50.67 19.79 82.0 225 79.0 57.63 20.27 84.0 232 01.0 58.63 20.77 88.0 239 03.0 50.68 21.29 88.0 246 85.0 060.74 21.83 90.0 254 87.0 81,87 22.39 92.0 262 08.0 63.04 22.97 94.0 270 91.0 64'.26 23.58 96.0 279 03.0 53.52 24.22 00.0 288 0.0 98.84 24.88 100.0 297 87.0 58.21 25.88 102.0 307 90.0 09.84 26.28 104.0 217 101.0 71.13 27.02 100.0 327 103.0 72.68 27.79 100.0 338 105.0 74.29 28.60 M10.0 350 107.0 75397 29.44 112.0 62 109.0 77.71 30.31 114.0 374 111.0 79.53 31.22 110.0 387 113.0 01.42 32.16 110.0 400 115.0 93.38 33.15 120.0 414 117.0 85.44 34.17 122-0 420 11..0 87.57 35.24 124.0 444 121.0 00.70 36.35 125.0 459 123.90 02.1 37.00 12E.0 476 12500 94.50 38.70 130.0 493 127.0 97.00 39.96 132.0 511 120.0 99.61 41.26 134.0 529 131.0 102.32 42.61 136.0 548 133.0 105.14 44.02 138.0 080 130.0 100.07 45.49 140.0 009 137.0 111.13 47.02 142,0 610 130.0 114.31 48.61 144.0 633 141.0 117.02 50.20 146.0 656 143,0 12186 5`1.99 148.0 681 145.0 124.65 53.78 150.0 706 147.0 128.38 05.65 102.0 733 140.0 132.21 57.59 154.0 780 151.0 136.31 59.61 150.0 789 153.0 140.52 61.71 158.0 818 150.0 144.00 63.90 160.0 849 157.0 149045 60,10 162.5 082 108.0 164.20 68.50 164.0 910 161.0 109.14 71.02 100.0 950 153.0 164.28 73.59 1080. 987 100.5 169.62 70.27 170.0 1,024 107.0 175.10 79.05 172.0 1.064 105.0 190.99 81095 174.0 1M100 171.0 187.02 84.96 1760. 1,148 173.0 183.30 80.10 178.0 1,192 175.0 109.83 91.37 190.0 1,230 File No.: 1100445.3.03 Page B-9 of B-13 Revision: 0 F0306-01RI
~.Stmawr~uI *ftgrty AssOWus, hic"'
CNS Pressure Test (Curve A), 32 EFPY 1,300 1,200 1,100 1,000
_j 900 0
70"F,814pslg 800 LU 700 U) 600 70"F, 586 pslg 70"F, 510 psli 500 Beftline Region 400 - Beftline N16 Nozzle 70'F, 313 psig - -Bottom Head 0 1101F, 313 pslg
-- LýUperVessei 300 F'-u7TempFIF 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)
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300 ITemp:
70 F l: - - UpperVessel 200 100 0
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)
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on We ott, C NSIurW o a O p eats V
CNS Normal Operation - Core Not Critical (Curve B), 32 EFPY 1,300 T 1,200 1,100 S1,000
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0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (=F)
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,bVohw&uIEftur Assocaes, RIC° CNS Normal Operation - Core Not Critical (Curve B), 54 EFPY 1,300 1,2001 1i 1 1,100 S~ 1,000 CI,/
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0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)
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iStructuralIntegrity Associates, IncO File No.: 1100445.304 Project No.: 1100445 CALCULATION PACKAGE Quality Program: M Nuclear F] Commercial PROJECT NAME:
Cooper P-T Curve Revision CONTRACT NO.:
4200001742 CLIENT: PLANT:
Nebraska Public Power District Cooper Nuclear Station CALCULATION TITLE:
Core Differential Pressure Nozzle Finite Element Model and Stress Analysis Document Affected Project Manager Preparer(s) &
Revision Pages Revision Description Approval Checker(s)
Signature & Date Signatures & Date 01 -11 Initial Issue Eric J. Houston Tyler D. Novotny EricJ.Ho8/2/TDN 7/25/11 EJ-H 8/2/11 Craig Jenson CJ 7/29/11 Page 1 of 11 F0306-OIRI
Table of Contents 1.0 OBJECTIVE ........................................................................................................ 3 2.0 ASSUM PTIONS ....................................................................................................... 3 3.0 DESIGN INPUTS ..................................................................................................... 3 3.1.1 FiniteElement M odel (FEM)........................................................................ 3 3.1.2 Geometry andM aterialProperties............................................................... 3 3.1.3 M esh ........................................................................................................... 3 3.1.4 Boundary Conditions.................................................................................... 3 4.0 CALCULATIONS .................................................................................................... 4 4.1 Unit Pressure Stress Analysis ........................................................... I................ 4 4.2 Determining Critical Stress Path ................................................................ 4 5.0 RESULTS OF ANALYSIS ...................................................................................... 4
6.0 REFERENCES
......................................................................................................... 4 List of Tables Table 1: M aterial Properties ............................................................................................... 6 Table 2: Path I Mapped Stresses for the Unit Pressure Loading ....................................... 6 List of Figures Figure 1: Core Differential Pressure Nozzle Dimensions [2] .............................................. 7 Figure 2: Core Differential Pressure Nozzle FEM Mesh ................................................... 8 Figure 3: Boundary Conditions .......................................................................................... 9 Figure 4: Pressure and Cap Load Application .................................................................... 10 Figure 5. Stress Path Location ........................................................................................... 11 File No.: 1100445.304 Page 2 of 11 Revision: 0 F0306-01R1
1.0 OBJECTIVE The objective of this calculation is to perform a finite element stress analysis on Cooper Nuclear Station's core differential pressure nozzle to support a revision to the Pressure-Temperature curves.
2.0 ASSUMPTIONS The following assumptions are made for modeling simplicity:
- 1. Cladding is neglected in the analysis because no thermal transients in which differential thermal expansion would occur are being analyzed. By neglecting the cladding, the smaller cross section results in conservative stress results when the pressure load is applied. This approach is consistent with the intent of the ASME Code.
- 2. The weld between the nozzle insert and the vessel bottom head is assumed to be the same material as the nozzle and weld butter.
- 3. The weld pad on the outside surface of the vessel bottom head is assumed to have the same material properties as the vessel bottom head.
- 4. Density and Poisson's ratio are assumed temperature independent for all materials. In addition, typical values are assumed for these.
- 5. The material properties are taken from the vessel stress report, Reference [1]. The properties from the vessel stress report are at 325°F.
3.0 DESIGN INPUTS 3.1.1 Finite Element Model (FEM)
A 2-D axis-symmetric FEM of the core differential pressure nozzle is developed using the program ANSYS
[1]. The model includes the nozzle insert, J-groove weld, weld butter and sufficient portion of the attached RPV bottom head such that end effects do not influence the region of interest. Further details of model boundary conditions and the application of loads to the model are described in subsequent sections.
3.1.2 Geometry and MaterialProperties The FEM is constructed using the dimensions shown in Figure 1, per Reference [2], to create a 2-D FEM using the ANSYS [3] finite element software. Material types used in the FEM are provided in Table 1 with properties taken from the vessel stress report, Reference [1]. Note that Code Case 1339-2 [4] is applicable for the vessel wall, however the code case does not alter the material properties. The type of material modeled for each component is shown in Figure 2 as a different color.
3.1.3 Mesh Figure 2 shows the mesh defined for the core differential pressure nozzle. The FEM consists of four node PLANE42 structural elements. The FEM is saved as CoreDPgeom.INPin the supporting files.
3.1.4 Boundary Conditions The model is constructed as an axis-symmetric model. Nodes at each end of the nozzle are coupled in the nozzle axial direction to simulate the remaining parts of the nozzle insert and attached piping that are not modeled. Symmetry boundary conditions are applied to the cut plane of the bottom head to simulate the File No.: 1100445.304 Page 3 of II Revision: 0 F0306-OIRI
remaining portion of the bottom head that is not modeled. Figure 3 shows the boundary conditions applied to the model. Blue triangles indicate the symmetry condition while green triangles indicate a coupled condition.
4.0 CALCULATIONS 4.1 Unit Pressure Stress Analysis A 1,000 psi unit pressure stress analysis is performed using the input file Pressure.inp. Figure 4 shows the applied 1,000 psi internal pressure distribution and cap load applied to the FEM. A cap load is applied to the bottom face of the nozzle body using the equation:
P.R R2 _R Where Ri is the inner radius of the attached piping and RI is the outer radius of the attached piping. This cap load is given a negative sign in order to exert tension on the model. The ANSYS input file Pressure.inp contains the loading and boundary conditions for the pressure loading and is saved in the supporting files.
4.2 Determining Critical Stress Path One critical stress path is selected for the core differential pressure nozzle. The path used to extract results is shown in Figure 5 and is consistent with the guidance in Reference [5]. Mapped stress results in the hoop direction for the path are extracted. Because the nozzle is in the spherical section of the vessel (bottom head), no stress concentration factor is needed to correct the path stresses.
5.0 RESULTS OF ANALYSIS A unit pressure stress analyses is performed for the stress paths for the core differential pressure nozzle.
Table 2 provides the unit pressure stress analysis results along the limiting path.
Pathi.lin: Unit pressure stress analysis results for Path 1
6.0 CONCLUSION
The stress analysis has been performed for the core differential pressure nozzle and may be used as input to the revision to the Pressure-Temperature curve calculations.
7.0 REFERENCES
- 1. Combustion Engineering, CENCI 150, Part 1, Analytic Report for Consumers Reactor Vessel Cooper Station, SI File No. COOP-07Q-208.
- 2. Combustion Engineering Drawing E 232-242, Nozzle Details, SI File No. 1100445.204.
- 3. ANSYS Mechanical APDL and PrepPost, Release 12.1 x64, ANSYS, Inc., November 2009
- 4. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Code Case 1339-2, "Requirements For Plate," Approved June 24, 1966.
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- 5. Structural Integrity Associates Report No. 0900876.401, Revision 0, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations."
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Table 1: Material Properties Nozzle Body Vessel Wall J-groove Weld Weld Butter Weld Pad J-groove Weld B Properties A-53J-groveassumed Grae SB-166 [2] [1, Page 5] assumed to be SB-166 [2] identical to
[1, Pge 5] SB-166 ietclt A-533 Grade B Elastic Modulus 30.4E6 psi o 28.8E6 psi
- 30.4E6 psi t 30.4E6 psi "I 28.8E6 psi (2)
Density (3) 0.300 Ibm/in3 0.283 Ibm/in' 0.300 Ibm/in' 0.300 ibm/in' 0.283 Ibm/in' Poisson's Ratio 0.3 0.3 0.3 0.3 0.3 (3)
Notes: 1. Sheet 11 ofA-682 of Reference [1] for inconel at 3250 F.
- 2. Sheet 11 of A-682 of Reference [1] for low alloy at 325°F.
- 3. Assumed typical values.
Table 2: Path 1 Mapped Stresses for the Unit Pressure Loading Sz Distance (Hoop w/respect to Nozzle) 0.00000 26186 0.22512 31610 0.45025 30525 0.67537 28449 0.90050 26377 1.12560 23518 1.35070 22272 1.57590 21189 1.80100 20222 2.02610 19351 2.25120 18559 2.47640 17831 2.70150 17156 2.92660 16522 3.15170 15916 3.37690 15327 3.60200 14746 3.82710 14169 4.05220 13594 4.27740 12838 4.50250 12045 File No.: 1100445.304 Page 6 of 11 Revision: 0 F0306-OIRI
Figure 1: Core Differential Pressure Nozzle Dimensions [21 File No.: 1100445.304 Page 7 of 11 Revision: 0 F0306-ORI
1 AN I Is coope :ore DP Nozzle, Pressure Stress Figure 2: Core Differential Pressure Nozzle FEM Mesh File No.: 1100445.304 Page 8 of 11 Revision: 0 F0306-OIRI
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