ML023250201
| ML023250201 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/15/2002 |
| From: | Conway J Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NMPIL 1697, TAC MB6687, TAC MB6703 | |
| Download: ML023250201 (94) | |
Text
P.O. Box 63 Lycoming, New York 13093 0
Constellation Energy Group Nine Mile Point Nuclear Station November 15, 2002 NMP1L 1697 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
Nine Mile Point Unit 1 Docket No. 50-220 License Amendment Request Pursuant to 10 CFR 50.90: Revision of Reactor Pressure Vessel Pressure-Temperature Limits and Request for Exemption from Requirements of 10 CFR 50.60 TAC Nos. MB6687 and MB6703 Gentlemen:
Pursuant to 10 CFR 50.90, Nine Mile Point Nuclear Station, LLC, (NMPNS) hereby requests an amendment to Nine Mile Point Unit 1 (NMP1) Operating License DPR-63.
The proposed changes to the Technical Specifications (TSs) contained herein would revise the Reactor Coolant System (RCS) Pressure-Temperature (P-T) limit curves and associated limit tables specified in Section 3/4.2.2, "Minimum Reactor Vessel Temperature for Pressurization " Specifically, the proposed changes replace TS Figures 3.2.2.a through 3.2.2.e with new figures, deleting Figures 3.2.2.f and 3.2.2.g, and replace associated Tables 3.2.2.a through 3.2.2.e with new Tables, deleting Tables 3.2.2.f and 3.2.2.g. Specification 3.2.2.c is updated to eliminate the references to the deleted figures.
The Bases for TS 3/4.2.2 have been revised to reflect the proposed changes to the TSs The P-T limit curves and tabular listing of P-T limit values contained in the new figures and tables are based, in part, on an alternative methodology and will be valid for 28 Effective Full Power Years (EFPY). The estimated EFPY at the end of the current operating cycle is 21.63 EFPY.
The alternative methodology used to develop the new P-T limit curves and tables has been endorsed by the American Society of Mechanical Engineers (ASME), but has not yet received formal approval by the NRC for generic application. The alternative methodology uses the ASME Boiler and Pressure Vessel (B&PV) Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1," in calculating the new RCS P-T limits. The use of this alternative methodology requires an exemption from the current requirements of 10 CFR
Page 2 NMPIL 1697 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," pursuant to 10 CFR 50.60(b) and 10 CFR 50.12, "Specific exemptions." The exemption request is provided in Attachment 4. The NRC has granted similar exemptions and approved the associated TS changes for a number of other Boiling Water Reactor (BWR) plants, including: Pilgrim (ADAMS Accession Numbers ML010720448 and ML010790519), Brunswick Units 1 and 2 (ADAMS Accession Numbers ML012760157 and M1L012780286), and Susquehanna Units 1 and 2 (ADAMS Accession Numbers MLO013520568 and MLO13520605).
The procedures and methodology that were previously used to calculate the RCS P-T limit curves and tables for NMP 1 were revised to recalculate the curves based, in part, on the ASME N-640 Code Case. The neutron fluence values for the Reactor Pressure Vessel (RPV) are unchanged from those calculated for the current P-T limit curves and tables. Therefore, the new P-T limit curves and tables were developed using the ASME N-640 Code Case in conjunction with the current neutron fluence values.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that the changes involve no significant hazards considerations. In addition, the proposed exemption to 10 CFR 50.60 has been evaluated and determined to be acceptable pursuant to the provisions of 10 CFR 50.12.
NMPNS requests approval of this application and issuance of the TS amendment by March 15, 2003 with 60 days allowed for implementation. The amendment is needed for the Spring 2003 refueling outage (RFO17) in anticipation of commencing hydrostatic testing on March 27, 2003. This letter contains one (1) new commitment as defined in Section 5.3 of Attachment 1.
Pursuant to 10CFR50.91(b)(1), NMPNS has provided a copy of this license amendment request and the associated analyses regarding no significant hazards considerations to the appropriate state representative.
I declare under penalty of perjury that the foregoing is true and correct. Executed on November 15, 2002.
Sincerely, hnT. C-o"onway Vice President Nine Mile Point JTC/CDM/jm
Page 3 NMP1L 1697 Attachments:
- 1. Evaluation of Proposed Technical Specification Changes
- 2. Proposed Technical Specification Changes (Mark-up)
- 3. Technical Specification Bases Changes (Mark-up For Information Only)
- 5. Report No. SIR-02-129 cc:
Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)
Mr. John P. Spath, NYSERDA
ATTACHMENT 1 EVALUATION OF PROPOSED TECHNICAL SPECIFICATION CHANGES
Subject:
License Amendment Request Pursuant to 10 CFR 50.90: Revision of Reactor Pressure Vessel Pressure-Temperature Limits and Request for Exemption from Requirements of 10 CFR 50.60
1.0 DESCRIPTION
2.0 PROPOSED CHANGE
3.0 BACKGROUND
4.0 TECHNICAL ANALYSIS
5.0 REGULATORY SAFETY ANALYSIS
6.0 ENVIRONMENTAL CONSIDERATION
Page 1 of 11
1.0 DESCRIPTION
This letter is a request to amend Operating License DPR-63 for Nine Mile Point Unit 1 (NMP1).
The proposed changes would amend the Operating License to revise the Reactor Coolant System (RCS) Pressure-Temperature (P-T) limit curves and associated limit tables specified in Technical Specification (TS) Section 3/4.2.2, "Minimum Reactor Vessel Temperature for Pressurization." Specifically, the proposed changes replace TS Figures 3.2.2.a through 3.2.2.e with new figures, deleting Figures 3.2.2.f and 3.2.2.g, and replace associated Tables 3.2.2.a through 3.2.2.e with new Tables, deleting Tables 3.2.2.f and 3.2.2.g. Specification 3.2.2.c is updated to eliminate the references to the deleted figures.
The Bases for TS 3/4.2.2 have been revised to reflect the proposed changes to the TSs.
The proposed changes to the TSs and associated changes to the TS Bases are indicated in the mark-up pages provided in Attachments 2 and 3, respectively. The TS Bases changes are provided for information only and will be controlled by the TS Bases change control process.
The P-T limit curves and tabular listing of P-T limit values contained in the new figures and tables are based, in part, on an alternative methodology and will be valid for 28 Effective Full Power Years (EFPY). The estimated EFPY at the end of the current operating cycle is 21.63 EFPY.
The proposed P-T limit curves and tables have been developed using the alternative methodology permitted by American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Case N-640, "Alternate Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1." Code Case N 640 permits the use of an alternative fracture toughness curve (i.e., Kjc in lieu of Kia) for the development of P-T limit curves. The use of this alternative methodology represents an exception to the requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," and therefore, requires an exemption from the requirements. A formal Exemption Request is provided in Attachment 4.
A report summarizing the inputs, methodology, and results of the calculations used in the development of the proposed (new) P-T limit curves and tables is included in Attachment
- 5.
2.0 PROPOSED CHANGE
TS Figures 3.2.2.a and b and Tables 3.2.2.a and b are being replaced with new figures and tables to provide RCS P-T limits that continue to be valid for up to 28 EFPY and that are applicable to reactor vessel heatup and cooldown with the reactor core not critical.
Page 2 of 11
TS Figures 3.2.2.c and d and Tables 3.2.2.c and d are being replaced with new figures and tables to provide RCS P-T limits that continue to be valid for up to 28 EFPY and that are applicable to reactor vessel heatup and cooldown with the reactor core critical.
TS Figures 3.2.2.e, f, and g and Tables 3.2.2.e, f, and g are being replaced with a new single Figure 3.2.2.e and a new single Table 3.2.2.e to provide RCS P-T limits that are valid for up to 28 EFPY and that are applicable to hydrostatic and leak tests. Figures 3.2.2.f and g and Tables 3.2.2.f and g are being deleted since they have been superseded by the new single (28 EFPY) figure and new single (28 EFPY) table.
TS 3.2.2.c is being updated to eliminate the references to the deleted figures. This is a conformance change only.
The Bases for TS 3.2.2 and 4.2.2 are being revised to reflect the changes to the TSs.
3.0 BACKGROUND
In accordance with 10 CFR 50, Appendix A, General Design Criterion (GDC) 31, "Fracture prevention of reactor coolant pressure boundary," the reactor coolant pressure boundary is required to be designed with sufficient margin to assure that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a non-brittle manner. The GDC also requires consideration of the uncertainties in determining the effects of irradiation on material properties. These requirements are reiterated in 10 CFR 50.60. The requirements of 10 CFR 50.60 are described in 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
Fracture toughness and reactor vessel material surveillance program requirements as specified in 10 CFR 50, Appendices G and H, must be considered in establishing RCS P T limits. Appendix G specifies that the fracture toughness and testing requirements for reactor vessel material meet the requirements of the ASME B&PV Code and requires that the beltline material in the surveillance capsules be tested in accordance with the requirements of 10 CFR 50, Appendix H. Appendix G of 10 CFR 50 endorses ASME B&PV Code,Section XI, Appendix G, as providing a conservative method for developing reactor vessel P-T limits. In addition, Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations," requires that the methods described in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," be used to predict the effect of neutron irradiation on the Adjusted Reference Temperature (ART). The ART is defined as the sum of the initial nil-ductility transition reference temperature (RTNDT) of the material, the increase in RTNDT caused by neutron irradiation, and a margin to account for uncertainties in the prediction method.
The current P-T limit curves and tables were approved by the NRC on November 25, 1998 and issued as Amendment No. 164 to the NMP1 TSs. Approval of the current P-T limit curves and tables was based on the conformance of the limits to the requirements of Page 3 of 11
10 CFR 50, Appendix G, and Generic Letter 88-11. The current P-T limits satisfied Generic Letter 88-11 since the method used to calculate the ART conformed to Regulatory Guide 1.99, Revision 2.
4.0 TECHNICAL ANALYSIS
In September 2002, Nine Mile Point Nuclear Station, LLC, (NMPNS) contracted with Structural Integrity Associates (SIA) to recalculate the NMP1 P-T limit curves and tables.
The recalculated (proposed) P-T limit curves and tables are based, in part, on the fluence values calculated for the previously approved (current) P-T curves and tables. In addition, the recalculated (proposed) P-T limit curves and tables include improvements that have been made to the calculational methodology contained in Section XI, Appendix G, of the ASME B&PV Code. The proposed new P-T limit curves and tables are all valid for 28 EFPY.
The methodology improvements were the application of ASME B&PV Code Case N 640, which permits fracture toughness curve Kic, as found in ASME B&PV Code,Section XI, Appendix A, to be used in lieu of curve KIa of Section XI, Appendix G, for the development of P-T limit curves. The proposed (new) P-T limit curves and tables for NMP1 were, therefore, developed in accordance with 10 CFR 50, Appendix G, and the 1989 Edition of ASME B&PV Code,Section XI, Appendix G, as modified by the ASME N-640 Code Case. Use of the 1989 Edition of the ASME Code is acceptable based on 10 CFR 50, Appendix G, and 10 CFR 50.55a(b)(2). Application of the methodology improvements of ASME N-640 Code Case are further discussed in Section 4.1 below.
4.1 Application of ASME N-640 Code Case The proposed P-T limits were developed based on the methodology specified in Section XI, Appendix G, of the ASME B&PV Code, as modified by ASME B&PV Nuclear Code Case N-640. ASME Code Case N-640 permits the use of alternate material fracture toughness when developing minimum vessel temperatures. Specifically, fracture toughness K1e values as defined in ASME B&PV Code,Section X), Appendix A, Figure A-4200-1, were used in lieu of the Kka values defined in ASME B&PV Code,Section XI, Appendix G, Figure G-2210-1, for the development of the proposed (new) P-T limit curves and tables.
Use of the Kjc curve in determining the lower bound fracture toughness in the development of P-T limit curves is more technically correct than the Kia curve. The Kic curve models the slow heatup and cooldown processes that a Reactor Pressure Vessel (RPV) normally undergoes. These slow heatup and cooldown limits are enforced by NMP1 TS Sections 3.2.1 and 3/4.2.2. Specifications 3.2.1, 3.2.2.a, b, and c, and 4.2.2.a provide assurance that the heatup and cooldown rate limit of< 100' FAHR, as specified in Updated Final Safety Analysis Report (UFSAR) Section V-CA and Table V-1, is met.
Page 4 of 11
Use of this approach is justified by the initial conservatism of the KIa curve when it was incorporated into the ASME B&PV Code in 1974. This initial conservatism was necessary due to the limited knowledge of RPV material fracture toughness at the time.
Since that time, considerable knowledge has been gained regarding fracture toughness of RPV materials and their fracture response to applied loads. This increased knowledge has served to demonstrate that the fracture toughness provided by the KIa curve is well beyond the margin of safety required to protect against potential RPV failure, and the Kie fracture toughness curve provides an adequate margin of safety for such a failure.
The acceptability of, and technical basis for, the use of ASME Code Case N-640 is described in "Technical Basis for Revised P-T Limit Curve Methodology," by W. H.
Bamford (Westinghouse Electric), S. N. Malik (NRC), et. al. This methodology was presented at the 2000 ASME Pressure Vessels and Piping Conference. In general, the revised methodology removes excess conservatism in the current ASME,Section XI, Appendix G, approach. Performance of leak tests at artificially high temperatures could impact test personnel safety, challenge operators with maintaining a high temperature in a limited operating band, and decrease the availability of plant systems, including shutdown cooling, due to the longer RPV heatup and test time.
Notwithstanding that the use of the ASME N-640 Code Case changes the methodology used to calculate the proposed P-T limit curves and tables, the modified methodology continues to satisfy the guidance contained in the 1989 Edition of ASME B&PV Code,Section XI, Appendix G. Therefore, it follows that the proposed P-T limit curves and tables will also continue to satisfy the intent of the guidance contained in 10 CFR 50, Appendices G and H.
The NRC has found the application of the ASME N-640 Code Case acceptable. A number of nuclear facilities have previously requested the use of the N-640 Code Case and their applications have been approved by the NRC. [
Reference:
Pilgrim (ADAMS Accession Numbers ML010720448 and MLO 10790519), Brunswick Units 1 and 2 (ADAMS Accession Numbers ML012760157 and ML012780286), and Susquehanna Units 1 and 2 (ADAMS Accession Numbers ML013520568 and ML013520605)]. Also, note that the NRC is currently in the process of providing generic approval of ASME Code Case N-640 by including it in Revision 13 of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1 [
Reference:
Draft Regulatory Guide DG-1091 (66 FR 67335, 12/28/01)].
Based on the technical basis provided in "Technical Basis for Revised P-T Limit Curve Methodology," by W. H. Bamford (Westinghouse Electric), S. N. Malik (NRC), et. al.,
and continued compliance with 10 CFR 50, Appendices G and H, NMPNS has concluded that the proposed P-T limit curves and tables maintain an adequate margin of safety for brittle fracture.
Page 5 of 11
4.2 Fluence Calculations GDC 31 and 10 CFR 50, Appendix G, require the prediction of the effects of neutron irradiation on vessel embrittlement. In accordance with Generic Letter 88-11, the NRC requires the methods described in Regulatory Guide 1.99, Revision 2, to be used to predict these effects. The Regulatory Guide requires the ART to be calculated to account for the effects of neutron embrittlement. One of the key components used in the calculations of the ART is RPV neutron fluence.
Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," provides guidance for the calculation of RPV neutron fluence. The neutron fluence values calculated using the methodology described in Regulatory Guide 1.190 satisfy the requirements of 10 CFR 50, Appendix G, and Regulatory Guide 1.99, Revision 2. Accordingly, ART values calculated using these neutron fluence values would also satisfy 10 CFR 50, Appendix G, and Regulatory Guide 1.99, Revision 2, and thereby satisfy Generic Letter 88-11.
The current P-T limit curves and tables were developed in 1998 following the 1997 withdrawal and testing of the 210 degree surveillance capsule. The calculations supporting the current P-T limit curves and tables utilized RPV neutron fluence values calculated consistent with the methodology of Draft Regulatory Guide DG-1053, which was a previous draft for Regulatory Guide 1.190 and the latest guidance available at the time.
The current P-T limit curves and tables were approved by the NRC on November 25, 1998 and issued as Amendment No. 164 [
Reference:
TAC No. MA1413] to the NMP1 TSs. The current P-T limits are valid for up to 28 EFPY and satisfy Generic Letter 88-11 since the method used to calculate the ART is consistent with Regulatory Guide 1.99, Revision 2. Accordingly, the current P-T limit curves and tables and supporting RPV neutron fluence values satisfy the requirements of 10 CFR 50, Appendix G.
The RPV neutron fluence values used for the proposed (new) P-T limit curves and tables are unchanged from those previously calculated for the current P-T limit curves and tables. The ART for the limiting beltline material (Plate G-307-4/5) at 28 EFPY is unchanged and remains less than the 2000 F limit required by Regulatory Guide 1.99, Revision 2.
As discussed above, the RPV neutron fluence values for NMP1 were calculated consistent with Draft Regulatory Guide DG-1053, which was a draft for Regulatory Guide 1.190, and the most recent guidance for neutron fluence calculations available in 1998. Subsequently, in March 2001, the NRC issued Regulatory Guide 1.190. As a result, the RPV neutron fluence values for NMP 1 (previously calculated consistent with DG-1053) were verified to be consistent with the requirements and methodology, of Regulatory Guide 1.190. The Regulatory Position 1.4 uncertainty analyses and comparisons with benchmark measurements and calculational benchmark problems (as provided in NUREG/CR-6115) have been completed and the Position 1.4 methodology Page 6 of 11
qualification and uncertainty estimates have been satisfied. A summary of the results of the uncertainty analyses and benchmarking comparisons follows:
- 1. The Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel simulator benchmark was used due to its high-accuracy measurement results extending from inside a simulated thermal shield through to the outside of a simulated vessel. The calculational results in the PCA show a slight consistent bias (less than 10%) with respect to the measurements, but no significant change in bias was observed with change in irradiation position. This indicates that the transport methodology is calculating the flux attenuation outside the core region with high accuracy. The observed bias is consistent with that obtained by other synthesis calculations.
- 2. The calculational benchmark was a typical BWR geometry similar to those for NMP 1 and Nine Mile Point Unit 2 (NMP2). Comparisons were made between the results obtained using the calculational methodology for NMP 1 [
Reference:
Letter No.
NMP1L 1373, dated 10/22/98] and the results obtained from the calculational benchmark (NUREG/CR-6115). The results of these comparisons showed very good agreement. In the representative RPV surveillance capsule (located at 30 azimuth),
the average results were approximately 3% low. Within the RPV, the average results were approximately 2 to 3% high at the vessel inner radius (IR). All compared results were within +/-10%.
- 3. Additional comparisons were made with surveillance capsule measurements in NMIP1 and NMP2, and with the core shroud measurements in NMP 1 [
Reference:
Letter No.
NMP1L 1373, dated 10/22/98]. In all cases, agreement with measured results within the uncertainty was obtained. The uncertainties were shown to be less than +/-20%,
which meets the criterion set forth in Regulatory Guide 1.190 for acceptability of the calculations.
Based on the acceptable results of the verifications and benchmarking comparisons of the RPV neutron fluence values and calculational methodology, NMPNS has concluded that the neutron fluence values calculated for the proposed P-T limit curves and tables are consistent with the requirements of Regulatory Guide 1.190. Accordingly, the ART value calculated using these neutron fluence values satisfy 10 CFR 50, Appendix G, and Regulatory Guide 1.99, Revision 2, and thereby satisfy Generic Letter 88-11.
4.3 Conclusion NRC regulations require that P-T limit curves provide an adequate margin of safety to the conditions at which brittle fracture may occur. These requirements are set forth in GDC 31 and 10 CFR 50, Appendices G and H. Generic Letter 88-11 and Regulatory Guides 1.99 and 1.190 provide guidance for compliance with the requirements of GDC 31 and Appendices G and H. The Appendices reference the requirements and guidance of Section XI, Appendix G, of the ASME B&PV Code for the development of P-T limit curves. The methodologies described in Regulatory Guides 1.99 and 1.190 and the Page 7 of 11
ASME Code will provide P-T limit curves with the requisite margin against brittle fracture. The proposed P-T limit curves and associated P-T limit tables are consistent with these methodologies, as modified by application of ASME Code Case N-640. The proposed change to Specification 3.2.2.c is a conformance change which serves only to update the requirements to reflect the proposed changes to the P-T limit curves and tables.
ASME Code Case N-640 proposes an alternative to a requirement contained in Section XI, Appendix G, of the ASME B&PV Code. The alternate fracture toughness for RPV materials permitted by the Code Case is based on the additional knowledge gained since the inception of 10 CFR 50, Appendix G. The more appropriate assumptions and provisions allowed by the Code Case maintain a margin of safety that is consistent with the intent of 10 CFR 50, Appendices G and H.
The NRC has granted similar exemptions and approved the associated TS changes for a number of other Boiling Water Reactor (BWR) plants, including: Pilgrim (ADAMS Accession Numbers ML010720448 and ML010790519), Brunswick Units 1 and 2 (ADAMS Accession Numbers ML012760157 and ML012780286), and Susquehanna Units 1 and 2 (ADAMS Accession Numbers M1L013520568 and ML013520605).
The comparisons of the RPV neutron fluence values and calculational methodology for NMP1 with the Regulatory Guide 1.190 Position 1.4 methodology and uncertainty estimates are anticipated to require supplemental review by the NRC staff. This review is considered to be independent of the proposed application of the ASME B&PV N-640 Code Case and associated changes to the TS P-T limit curves and tables. However, in order to assist in the NRC staffs review, NMPNS will submit to the NRC the report documenting the results of the benchmark measurements and calculations applicable to the methods used for NMIP1 by January 15, 2003. Pending approval of the submittal, NMP1 is prepared to accept, if required, an application period restriction for the proposed P-T limit curves and tables similar to that recently imposed on other facilities requesting application of the ASME N-640 Code Case. It is requested that the application period for the proposed P-T limit curves and tables allow plant operation through the remainder of the current operating cycle (Cycle 15) and also through the next operating cycle (Cycle 16). The estimated EFPY at the end of Cycle 16 is 23.38, which assures that the P-T limits will remain conservatively defined since the proposed P-T limit curves and tables are based on neutron fluence values that are currently accepted for 28 EFPY.
5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Analysis The proposed changes to the Technical Specifications (TSs) would replace the current Reactor Coolant System (RCS) Pressure-Temperature (P-T) limit curves and associated tables with revised curves and tables that are based, in part, on the alternate methodology of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Page 8 of 11
(B&PV) Code Case N-640. The TS Bases have been revised to reflect the proposed changes to the TSs.
Nine Mile Point Nuclear Station, LLC, (NMPNS) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1.
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No.
The proposed changes do not involve physical changes to the plant or alter the RCS pressure boundary (i.e., there are no changes in operating pressure, materials, or seismic loading). The proposed P-T limit curves and tables and supporting changes provide continued assurance that the fracture toughness of the Reactor Pressure Vessel (RPV) is consistent with analysis assumptions and NRC regulations. The proposed P-T curves and tables were developed in accordance with the fracture toughness requirements of 10 CFR 50, Appendix G, and ASME B&PV Code,Section XI, Appendix G, as modified by the alternate criteria and methods of ASME B&PV Code Case N-640. The more appropriate assumptions and provisions allowed by the Code Case maintain sufficient margins of safety to assure that, when stressed, the RPV boundary will behave in a non-brittle manner.
Use of this methodology provides assurance that the probability of a rapidly propagating fracture will be minimized. The proposed P-T limit curves and tables and supporting changes will prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2.
Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No.
The proposed P-T limit curves and tables and supporting changes do not affect the design or assumed accident performance of any structure, system, or component, or introduce any new modes of system operation or failure modes. Compliance with the proposed P-T curves and tables and supporting requirements will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Page 9 of 11
- 3.
Does the proposed change involve a significant reduction in a margin of safety?
Response
No.
NRC regulations require that P-T limit curves provide an adequate margin of safety to the conditions at which brittle fracture may occur. These requirements are set forth in 10 CFR 50, Appendix A, General Design Criterion (GDC) 31 and 10 CFR 50, Appendices G and H. Generic Letter 88-11 and Regulatory Guides 1.99 and 1.190 provide guidance for compliance with the requirements of GDC 31 and Appendices G and H. The Appendices reference the requirements and guidance of Section XI, Appendix G, of the ASME B&PV Code for the development of P-T limit curves. The methodologies described in Regulatory Guides 1.99 and 1.190 and the ASME Code will provide P-T limit curves with the requisite margin against brittle fracture. The proposed P-T limit curves and associated P-T limit tables are consistent with these methodologies, as modified by the application of ASME Code Case N-640.
ASME Code Case N-640 proposes an alternative to a requirement contained in Section XI, Appendix G, of the ASME B&PV Code. The alternate fracture toughness for RPV materials permitted by the Code Case is based on the additional knowledge gained since the inception of 10 CFR 50, Appendix G. The more appropriate assumptions and provisions allowed by the Code Case maintain a margin of safety that is consistent with the intent of 10 CFR 50, Appendices G and H. The proposed P-T limit curves and tables and supporting requirements provide assurance that the established P-T limits are not exceeded. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based on the above, NMPNS concludes that the proposed amendment presents no significant hazards considerations under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
5.2 Applicable Regulatory Requirements/Criteria The proposed P-T limit curves and associated tables are consistent with the alternate assessment criteria and methods of ASME B&PV Code Case N-640, and satisfy the requirements of GDC 31; 10 CFR 50.60; 10 CFR 50, Appendix G; and the 1989 Edition of ASME B&PV Code,Section XI, Appendix G, as modified by the Code Case. The proposed P-T limit curves and tables also satisfy Generic Letter 88-11 by using methods consistent with Regulatory Guide 1.99, Revision 2, and Regulatory Guide 1.109.
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.
Page 10 of 11
5.3 Commitments The following table identifies those actions committed to by NMPNS in this document.
Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
REGULATORY COMMITMENTS Due Date/Event NMPNS will submit to the NRC the report documenting the results To be submitted by of the RPV neutron fluence benchmark measurements and January 15, 2003 calculations applicable to the methods used for NMP1.
6.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) 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.
Page 11 of 11
ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
The current versions of Technical Specification pages 84 through 94d have been marked up by hand to reflect the proposed changes.
LIMITING CONDITION FOR OPERATION aW n cn i Arjtr-,D lhiWncm j
- c.
During leakage and hydrostatic testing, the reactor vessel temperature and pressure shall satisfy the requirements of Figure@ 3.2.2.e, (3,2.2A, ot 3.2.2. asop Ootia if the c6re is not critical. During reactor vessel heatup and cooldown for the purpose of leakage and hydrostatic testing, the reactor vessel temperature and pressure shall satisfy the requirements of Figures 3.2.2.a and 3.2.2.b for non-critical heatup and cooldown, respectively.
- d.
The reactor vessel head bolting studs shall not be under tension unless the temperature of the
"ýessel head flange and the head are equal to or greater than 100 0 F.
In order to generate additional plant-specific data, a capsule containing irradiated and unirradiated material will be re-inserted at the B capsule location. Re-insertion capsules have already been installed at the A and C locations.
A prime (') is used to indicate a re-insertion capsule. The withdrawal schedule for the re Insertion capsules is as follows:
Fourth capsule (A') - 24 EFPY Fifth capsule (C') - 32 EFPY Sixth capsule (B') - 40 EFPY AMENDMENT NO. 1h /K8 84
4 co
INSERT FIGURE 3.2.2.a HEATUP - CORE NOT CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
*1
-5I S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S Minimum Temperature for Boltup:
1000F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE ('F)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been Included in this figure)
FIGURE 3.2.2.a MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (HEATING RATE _ 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION
L ORN CRTI F/RATIO 304 2369 FOR~A UP TO NWIGEFECIVE OFW C
OERATV 257RE 90 h
-a 2O 197
,100 1979 120 (intr9e 7
130tit/hvebe/
.2icu~
nthstbe
- 19.
AT 14 0 1_.
2O UPT1ET5EGE0ETIpL OE 13~r 160 22 OPE7T AMENiDMET Wt.' 1.4z /%
886
INSERT TABLE 3.2.2.a LIMIT FOR NON-CRITICAL OPERATION HEATUP AT UP TO 1000F/HR REACTOR PRESSURE (psig)
IN TOP DOME 0
298 298 298 298 298 298 298 298 300 304 311 319 329 340 354 369 387 406 406 429 454 483 515 547 582 622 665 713 767 840 840 895 969 1050 1140 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 102 107 112 117 122 127 132 137 142 147 152 157 162 167 172 177 182 182 187 192 197 202 207 212 217 222 227 232 238 238 242 247 252 257 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP (CORE NOT CRITICAL) (HEATING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION
INSERT FIGURE 3.2.2.b COOLDOWN 1500 REACTOR PRESSURE (psig) 1000 500 0
- CORE NOT CRITICAL Minimum Temperature for Boltup:
100OF 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (fF)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.b MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (COOLING RATE < 100°FIHR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION S
a a
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S a
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a a
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a a
a S
a a
a a
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S a
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a a
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S a
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a S
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a
AMENDMENT NO. 1AZ,8, COOLDO*
AT 'UP TO 1
° F/HtR
/
S160 100 171 110 184 120 291 00 6130 391 200 49 220 957 280 1062 290 (ratory tcibeltline downco r water temperath esue t
rulto loop cin (isrm tuncertainties h bee Included in thisble)
/
[IIMMTE ORT RERPRESS
- TON D G71 OLI[W (ONT CL(COO GAE<
' O*F/E[R)
FOLDPOW(CNT EIGI EFFE
- FULL POW YEAR S*RE OPERI ON r
3,2.2.
88
INSERT TABLE 3.2.2.b LIMIT FOR NON-CRITICAL OPERATION COOLDOWN AT UP TO 100 0F/HR REACTOR PRESSURE (psig)
IN TOP DOME 0
205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 455 471 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 100 100 100 100 100 100 100 100 100 102 107 112 117 122 127 132 137 140 160 160 163 163 167 172 177 182 187 192 197 202 207 212 217 222 227 232 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.b MINIMUM TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE NOT CRITICAL) (COOLING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION
INSERT FIGURE 3.2.2.c HEATUP 1500 1000 REACTOR PRESSURE (psig) 500 0
- CORE CRITICAL Minimum Temperature for Boltup:
100°F W ater Level Must Be in Range For Power Operation If Core Is Critical Below 2220F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (fF)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.c MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (HEATING RATE < 100°F/HR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION
VT....
9OF CORE OPiPON AMENDMEN FO. ZJ
,'90
INSERT TABLE 3.2.2.c LIMIT FOR POWER OPERATION (CORE CRITICAL)
HEATUP AT UP TO looTF/HR REACTOR PRESSURE (psig)
INTOPDOME REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (-F) 0 298 298 300 304 311 319 329 340 354 360 360 360 406 429 454 483 515 547 582 622 665 713 767 840 840 895 969 1050 1140 100 100 172 177 182 187 192 197 202 207 212 217 222 a 222 a 227 232 237 242 247 252 257 262 267 272 278 278 282 287 292 297 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(awater level must be in range for power operation if core is critical below 222°F)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEATUP (CORE CRITICAL) (HEATING RATE < 1000FIHR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION
I 2
0 to Oka
INSERT FIGURE 3.2.2.d COOLDOWN - CORE CRITICAL 1500 1000 REACTOR PRESSURE (psig)
Soo I 0
Minimum Temperature for Boltup:
100°F Water Level Must Be in Range For Power Operation If Core Is Critical Below 203°F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (°F)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.d MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (COOLING RATE<1000 F/HR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION
LIMI FOR,POWE OPERATI N (CORE ICAL)
COOLD WN ATup 1000 F/HIR
-w 130 130 S136
-110 143 120 1510 14 71 150 1841 M10 216 L80 235 190 258 200
-7"?,SLE.3.2.2.,I 284 210 315 220 35 230/
3 233 O 0 240 060 25 360 9
493 60" 556 270 630 280 708 290 786 300 8
310 9 7 320 062 330
(&a er eemust be in range fo,- oer operation, If co-critical below 2 F)
(istum uncertainties have bee included ithsta e)
OR U._*0 CORE OPE__
ON_./
AKHDWMKT No. IAZ fo9 92
INSERT TABLE 3.2.2.d LIMIT FOR POWER OPERATION (CORE CRITICAL)
COOLING AT UP TO 100WF/HR REACTOR PRESSURE (psig)
IN TOP DOME REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 0 205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 360 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 1199 100 100 102 107 112 117 122 127 132 137 140 142 147 152 157 162 167 172 177 180 200 203 a 203 a 207 212 217 222 227 232 237 242 247 252 257 262 267 272 277 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
('water level must be in range for power operation if core is critical below 203'F)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.d MINIMUM TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE CRITICAL) (COOLING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION
cz
INSERT FIGURE 3.2.2.e LEAK/HYDRO TEST - CORE NOT CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
100°F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE ('F)
(reactor vessel beltline downcomer water temperature Is measured at recirculation loop suction)
(instrument uncertainties have been Included In this figure)
FIGURE 3.2.2.e MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (CORE NOT CRITICAL) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION a
a a
a a
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a a.-----.1-----a.
-a.
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a a
a
AMENDHENT NO. 19Z /AX' 94
INSERT TABLE 3.2.2.e LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL)
REACTOR PRESSURE (psig)
IN TOP DOME 0
360 360 688 704 722 742 764 788 815 844 877 913 953 997 1046 1100 1160 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 130 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.e MINIMUM TEMPERATURE FOR PRESSURIZATION DURING LEAK/HYDROSTATIC TESTING (CORE NOT CRITICAL)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION
z 0
685 724 769 821 881 9
94b
LEM FOR IN-SE VICE TEST (CORE N CRITIC
, FUEL IN /E EL) 36 00 3 0 110 60 120 360 130 582 140 604 u-/
PA R
DE L.7 150 631 160 661 170 697 180 1
79 9
10 92 220 974 230 1033 237 1058 240 1154 250 (rei or w ent e~euncertaintiesblln o
menve been aeincludedte
,is table)
- mainio n
TALE 3.g
/*LEAK/HYDR*
STTICTE IG (CR NO 'CRITICAL)
ORUPTO ~
FORE FTVE POWER OR~ ~
O CPT
- T OU EOPERATION
/
END*.s'r o. 16 4 94d
//
/
ATTACHMENT 3 CHANGES TO TECHNICAL SPECIFICATION BASES PAGES (FOR INFORMATION ONLY)
The current version of Technical Specification Bases page 95 has been marked-up by hand to reflect the proposed changes. These Bases pages are provided for information only and do not require NRC issuance.
BASES FOR 3.2.2 AND 4.2.2 MINIMUM REACTOR VESSEL TEMPERATURE FOR PRESSURIZATION Figures 3.2.2.a,'3.2.2.b; 3.2.2.c,aend 3.2.2.d are plots of pressure versus temperature for heatup and cooldown rates of up to 100O maximum (Specification 3.2.1). Figurea 3.2.2.e.
,e l..
@6p1ot6gtof pressure versus temperature for leakage and hydrostatic testing. When the minimum temperature for leakage and hydrostatic testing Is reached, a thermal soak shall be performed to ensure that t*he thermal gradient across the vessel wall is negligible. These curves are based on calculations of stress Intensity factors according to Appendix G of Sectionf of the ASME Boiler and Pressure Vessel Cod %,=Edition (wh Wlnter/1992 Addend In addition, temperature shifts due to fast neutron fluence at twenty-eight effective full power years of operation were Incorporated into {he figures. These shifts were calculated using the procedure presented In Regulatory Guide 1.99, Revision 2. Reactor vessel flange/reactor head flange boltup is governed by other criteria as stated in Specification 3.2.2.d. The pressure readings on the figures have been adjusted to account for instrument uncertainties and to reflect the calculated elevation head difference between the pressure sensing Instrument locations and the pressure sensitive area of the core beltline region. The temperature readings on the figures have been adjusted to account for instrument uncertainties.
-, t. 0 The reactor vessel head flange and vessel flange in combination with the double "0" ring type seal are designed to provide a leak-tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flanges. Both the head and vessel flanges have an NDT temperature of 40 0 F and they.are not subject to any appreciable neutron radiation exposure. Therefore, the minimum vessel flange and head flange temperature for bolting Is established at 400 F + 60 0 F or 100 0 F.
Figures 3.2.2.a, 3.2.2.b,.3.2.2.c, 3.2.2.d',39.2.2*e,3...2.
nd
.2..
have Incorporated a temperature shift due to the calculated fast neutron fluence. The neutron flux at the vessel wall Is calculated from core physics data and has been determined using flux monitors installed inside the vessel. The curves; a eeq,( foY3.
o
.2.f
.nd 7.2.7.g-,are applicable for up to twenty-eight effective full power years of operation.
urvs 32.2. ano 3.2.g a
i bl for pt tw ty n tw nty oureff tiv ful owýr yersrdsptivqy.
Vessel material surveillance samples are located within the core region to permit periodic monitoring of exposure and changes in material properties. The material sample program conforms with ASTM El85-66 except for the material withdrawal schedule which is specified in Specification 4.2.2.b.
AMENDMENT NO. W X%,'
95
ATTACHMENT 4 EXEMPTION REQUEST The Reactor Coolant System (RCS) Pressure-Temperature (P-T) limits proposed by Nine Mile Point Nuclear Station, LLC, (NMPNS) for Nine Mile Point Unit 1 (NMP1) are calculated using an alternative method to that described in 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and Appendix H, "Reactor Vessel Material Surveillance Program Requirements." The alternative method is based, in part, on the use of an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Case. Specifically, ASME Code Case N-640, "Alternate Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1," is used in calculating the RCS P-T limits proposed for the NMP1 Technical Specifications (TSs). Since this Code Case has not yet received formal approval from the NRC for generic application, the use of the alternative method requires an exemption from the current requirements of 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," which implements 10 CFR 50, Appendices G and H.
Pursuant to 10 CFR 50.12, "Specific exemptions," the NRC may grant an exemption from requirements contained in 10 CFR 50 (10 CFR 50.60 for this exemption) provided the following four conditions are satisfied:
- 1. The requested exemption is authorized by law,
- 2. The requested exemption does not present an undue risk to the public health and
- safety,
- 3. The requested exemption will not endanger the common defense and security, and
- 4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.
Previous exemptions permitting use of the ASME N-640 Code Case have been granted by the NRC to a number of nuclear facilities, including: Pilgrim (ADAMS Accession Numbers ML010720448 and ML010790519), Brunswick Units 1 and 2 (ADAMS Accession Numbers ML012760157 and ML012780286), and Susquehanna Units 1 and 2 (ADAMS Accession Numbers ML013520568 and ML013520605). In addition, the NRC is currently in the process of providing generic approval of ASME Code Case N-640 by including it in Revision 13 of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1 [
Reference:
Draft Regulatory Guide DG 1091 (66 FR 67335, 12/28/01)].
Page 1 of 5
ASME B&PV Code Case N-640 10 CFR 50.12(a) Requirements The requested exemption to permit use of ASME B&PV Code Case N-640 in conjunction with ASME B&PV Code,Section XI, Appendix G, to determine the RCS P T limit curves and tables meets the criteria of 10 CFR 50.12 as further discussed below.
- 1.
The requested exemption is authorized by law:
The provisions of 10 CFR 50.60(b) permit the use of alternatives to 10 CFR 50.60, Appendices G and H, when an exemption is granted by the NRC (Commission) under 10 CFR 50.12.
- 2.
The requested exemption does not present an undue risk to the public health and safety:
The RCS P-T limit curves and tables proposed for the NMP1 TSs rely, in part, on the requested exemption. In accordance with ASME Code Case N-640, the proposed P-T limit curves and associated tables have been developed using the fracture toughness K1, values as defined in ASME B&PV Code,Section XI, Appendix A, Figure A-4200-1, in lieu of the KIa values defined in ASME B&PV Code,Section XI, Appendix G, Figure G-2210-1. Curve KIc is used as the lower bound for fracture toughness. Except for the changes associated with the use of ASME Code Case N-640, the other margins involved with the ASME B&PV Code,Section XI, Appendix G, process of determining P-T limit curves remain unchanged.
Use of the K1c curve in determining the lower bound fracture toughness in the development of P-T limit curves is more technically correct than the KIa curve.
The K1c curve models the slow heatup and cooldown processes that a Reactor Pressure Vessel (RPV) normally undergoes. These slow heatup and cooldown limits are enforced by NMP1 TS Sections 3.2.1 and 3/4.2.2. Specifications 3.2.1, 3.2.2.a, b, and c, and 4.2.2.a provide assurance that the heatup and cooldown rate limit of < 1000 F/HR, as specified in Updated Final Safety Analysis Report (UFSAR) Section V-CA and Table V-i, is met.
Use of this approach is justified by the initial conservatism of the Kia curve when it was incorporated into the ASME B&PV Code in 1974. This initial conservatism was necessary due to the limited knowledge of RPV material fracture toughness at the time. Since that time, considerable knowledge has been gained regarding fracture toughness of RPV materials and their fracture response to applied loads. This additional knowledge has served to demonstrate that the fracture toughness provided by the Kia curve is well beyond the margin of safety Page 2 of 5
required to protect against potential RPV failure, and the Kic fracture toughness curve provides an adequate margin of safety for such a failure.
Use of the K1c fracture toughness limits as a basis for the proposed P-T limit curves and tables will enhance overall plant safety by widening the P-T operating window, especially in the region of low temperature operations. Safety benefits that would be realized during pressure tests include a reduction in the challenges to operators in maintaining a high temperature in a limited operating band, personnel safety while conducting inspections in primary containment at elevated temperatures, and increased availability of plant systems, including shutdown cooling, due to a reduction of the heatup and test time.
Based on the above justification, NMPNS believes that this requested exemption does not present an undue risk to the public health and safety.
- 3.
The requested exemption will not endanger the common defense and security:
This exemption request is limited to the revision of P-T operating and test limits for the NMPNS NMP1 commercial power reactor in accordance with industry proposed guidance. As such, this exemption request has no impact on common defense and security. Therefore, the common defense and security are not endangered by approval of this exemption request.
- 4.
Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60:
In accordance with 10 CFR 50.12(a)(2), the NRC will consider granting an exemption to the regulations if "special circumstances" are present. The regulation provides six criteria which licensees can us to provide the basis for the "special circumstance" provision of the regulation. The following three criteria are applicable to this exemption request:
"(a)(2)(ii) - Application of the regulation in the particular cicumstances would not serve the underlying purpose of the rule or it is not necessary to achieve the underlying purpose of the rule; or (a)(2)(iii) - Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated; or (a)(2)(v) - The exemption would provide only temporary relief from the applicable regulation and the licensee or applicant has made good faith efforts to comply with the regulations."
Each of the above three requirements is addressed below.
Page 3 of 5
Appendix G of 10 CFR 50 endorses ASME B&PV Code,Section XI, Appendix G, as providing a conservative method for developing reactor vessel P-T limits.
Application of this methodology in the development of P-T operating and pressure test limit curves satisfies the underlying requirement that the RCS pressure boundary be operated in a regime having sufficient margin to assure that, when stressed, the RPV boundary will behave in a non-brittle manner. Use of this methodology provides assurance that the probability of a rapidly propagating fracture will be minimized. Therefore, RCS P-T limit curves developed using this methodology provide assurance that adequate margin exists considering the uncertainties in determining the effects of irradiation on material properties.
The ASME B&PV Code,Section XI, Appendix G, methodology was conservatively developed based on the limited knowledge of RPV material fracture toughness that existed in 1974. Since that time, considerable knowledge has been gained regarding fracture toughness of RPV materials and their fracture response to applied loads. This increased knowledge serves to permit relaxation of the ASME B&PV Code,Section XI, Appendix G, requirements by application of ASME B&PV Code Case N-640. Relaxation of the Appendix G requirements will have no impact on the underlying purpose of the ASME B&PV Code or the regulations of 10 CFR 50.60. Therefore, the associated safety margins are maintained.
The RCS P-T operating window is defined by the RCS P-T operating and test limit curves and associated P-T limit tables that are contained in the NMP1 TSs.
As previously discussed, the P-T limit curves and tables have been developed in accordance with the ASME B&PV Code,Section XI, Appendix G, methods.
Continued operation of NMP1 with the current P-T limit curves and tables, without the relief provided by ASME B&PV Code Case N-640, would unnecessarily restrict the P-T operating band. This restriction challenges the operations staff during pressure tests to maintain a high temperature within a limited operating band. It also subjects inspection personnel to increased safety hazards while conducting inspections of systems with the potential for steam leaks at elevated temperatures.
This constitutes an unnecessary burden that can be alleviated by the application of ASME B&PV Code Case N-640 in the development of the proposed RCS P-T limit curves and tables. Furthermore, implementation of the proposed P-T limit curves and tables, developed using the N-640 Code Case, will not significantly reduce the margin of safety below that established by the original ASME B&PV Code,Section XI, Appendix G, requirements.
Page 4 of 5
The requested exemption provides only temporary relief since it is anticipated that the provisions of ASME B&PV Code Case N-640 will be incorporated into (or reconciled with) the requirements of 10 CFR 50, Appendix G, in response to ongoing to industry efforts to do so. NRC approval of the N-640 Code Case is pending; however, additional action may be required to permit use of the Code Case without requiring an exemption to 10 CFR 50, Appendix G. The estimated time for the completion of the, as yet, unspecified additional action(s) is not known. Therefore, the effective period for the requested exemption is indefinite.
Summary of Bases for Acceptability of ASME B&PV Code Case N-640 Compliance with 10 CFR 50, Appendix G, as required by 10 CFR 50.60(a), would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety. ASME B&PV Code Case N-640 permits a reduction in the lower bound fracture toughness used in ASME B&PV Code,Section XI, Appendix G, in the determination of RCS P-T limit curves. Use of the alternate methodology of the N-640 Code Case is acceptable based on the margin of safety being maintained commensurate with that which existed in 1974 when the original requirements of ASME B&PV Code,Section XI, Appendix G, were approved. Therefore, application of ASME B&PV Code Case N-640 for the development of the RCS P-T limit curves and tables for NMP1 will maintain an acceptable margin of safety and does not present an undue risk to the public health and safety.
Page 5 of 5
ATTACHMENT 5 REPORT NO. SIR-02-129
Report No.: SIR-02-129 Revision No.: 0 Project No.: NMP-05Q File No.: NMP-05Q-401 November 2002 Revised Pressure-Temperature Curves for Nine Mile Point Unit 1 Prepared for:
Constellation Generation Group Prepared by:
Structural Integrity Associates Greenwood Village, CO Prepared by:
Reviewed by:
Approved by:
G.* Stevens, P. E.
K.K'-ikawa, P. E.
"L Stevens, P. E.
Date:
I'I s/!o Date:
1__/__/_
Date:
t115102.
0 Structural Integrity Associates
REVISION CONTROL SHEET Document Number:
SIR-02-129
Title:
Revised Pressure-Temperature Curves for Nine Mile Point Unit 1 Client: Constellation Generation Group SI Project Number:
NMP-05Q Section Pages Revision Date Comments 1.0 - 5.0 1-18 A
10/04/02 Initial draft for client review.
1.0 1-1 0
11/5/02 Initial issue.
2.0 2 2-2 3.0 3 3-26 4.0 4-1 5.0 5-1 App. A Al -All
Table of Contents Section Page 1.0 IN TR O D U CTIO N..........................................................................................................
1-1 2.0 A R T ESTIM ATE...........................................................................................................
2-1 3.0 P-T C UR V E M ETH O D O L O G Y..................................................................................
3-1 3.1 B enchm ark A nalysis.....................................................................................................
3-1 3.2 R evised P-T Curves....................................................................................................
3-15 4.0 C O N CLU SIO N S.............................................................................................................
4-1 5.0 RE FE REN CES...............................................................................................................
5-1 APPENDIX A: P-T CURVE PLOTS AND TABULATIONS IN TECHNICAL SPE CIFICATIO N FO R M A T..............................................................................................
A l SIR-02-129, Rev. 0 ioi
List of Tables Table Page Table 2-1. A RT C alculations.......................................................................................................
2-2 Table 3-1. Tabulated Values for Benchmark Pressure Test Curve.............................................
3-7 Table 3-2. Tabulation for Benchmark Core Not Critical (100°F/hr Cooldown)
P-T Curve for 28 EFPY............................................................................................
3-8 Table 3-3. Tabulation for Benchmark Core Not Critical (O0 F/hr) P-T Curve for 28 EFPY....... 3-9 Table 3-4. Tabulation for Benchmark Core Not Critical (100°F/hr Heatup)
P-T Curve for 28 EFPY..........................................................................................
3-10 Table 3-5. Tabulated Values for Benchmark Core Critical Curve............................................
3-11 Table 3-6. Tabulated Values for Revised Pressure Test P-T Curve for 28 EFPY.................... 3-17 Table 3-7. Tabulated Values for Revised Core Not Critical (100l F/hr Cooldown)
P-T Curve for 28 EFPY..........................................................................................
3-18 Table 3-8. Tabulated Values for Revised Core Not Critical (0°F/hr) P-T Curve for 28 E FPY............................................................................................................
3-19 Table 3-9. Tabulated Values for Revised Core Not Critical (100°F/hr Heatup)
P-T Curve for 28 EFPY..........................................................................................
3-20 Table 3-10. Tabulated Values for Revised Core Critical P-T Curves for 28 EFPY................. 3-21 List of Figures Figu Page Figure 3-1. Benchmark Pressure Test P-T Curve......................................................................
3-12 Figure 3-2. Benchmark Core Not Critical Curve......................................................................
3-13 Figure 3-3. Benchmark Core Critical Curve.............................................................................
3-14 Figure 3-4. Revised Pressure Test P-T Curve for 28 EFPY......................................................
3-22 Figure 3-5. Revised Cooldown Core Not Critical P-T Curve for 28 EFPY.............................. 3-23 Figure 3-6. Revised Heatup Core Not Critical P-T Curve for 28 EFPY...................................
3-24 Figure 3-7. Revised Cooldown Core Critical P-T Curve for 28 EFPY.....................................
3-25 Figure 3-8. Revised Heatup Core Critical P-T Curve for 28 EFPY..........................................
3-26 SIR-02-129, Rev. 0 iv
1.0 INTRODUCTION
This report documents the revised set of pressure-temperature (P-T) curves developed for the Nine Mile Point Unit 1 (NMP-1). This work includes a full set of updated P-T curves (i.e.,
pressure test, core not critical, and core critical conditions) for NMP-1 for 28 effective full power years (EFPYs). The curves were developed using the methodology specified in ASME Code Case N-640 [1], as well as 10CFR50 Appendix G [2], Welding Research Council (WRC)
Bulletin No. 175 [3], and the 1989 Edition of ASME Code,Section XI, Appendix G [4]. The revised P-T curves show an increase in the operating window of as much as 50'F or more, which is obtained from using the reference fracture toughness, K1,, in accordance with Code Case N-640.
SIR-02-129, Rev. 0 1-1
2.0 ART ESTIMATE Reference [5] provides adjusted reference temperature (ART) estimates for the NMP-1 reactor pressure vessel (RPV) materials in accordance with Regulatory Guide 1.99, Revision 2 (RG 1.99) [6] for 28 EFPY. The limiting value for an inside surface (1/4t) postulated flaw is:
ART'= 167.7°F for Plate G-307-4/5
[5, Table 4-4]
This value is reproduced in Table 2-1 in accordance with RG 1.99. In addition, the value for a 3/4t flaw is also determined in Table 2-1, with a value of 136.8°F.
Note that per Reference [7], fluence analysis subsequent to the Reference [5] report lowered the best estimate fluence and reduced the limiting plate ART. The conservative 28 EFPY limiting ART value of 167.7°F applied in Reference [5] is again used in this calculation.
SIR-02-129, Rev. 0 2-1
Table 2-1. ART Calculations Chemistry Chemistry Adjustments For 114t Initial RTmT Factor A RTIhDT Margin Terms ARTHDT Location rF)
Cu(wt%)
Ni (wt%)
(F)
CF) 0 YA CF) a, (*F) EFPY (7F) 40 U21 1I.5u I vl13 U
I lU UU I*
1ou t a,
403 027 I 053 17385 937 170 00 200 1loI 40 07 053 178 628 17 0O 00 280 I1368 Plate G-307A/5 (1/4t)
Fluence Information"
/
Wall Thickness (inches)
Fluence at !
Attenuaton, 114t Fluence @ 114t Fluence Factor, FF Location Ful 114t 314t EFPY (n/cm2) e"24x (nrcrn2)
(02.- Log 1)
Plate G-307-415 (1/4t) 7125 1781 280 270E+18 0652 1 759E+18 0539 Plate G-307-4/5 (3/4t) 7125 5344 280 270E+18 0277 7479E+17 0361 SIR-02-129, Rev. 0 2-2
3.0 P-T CURVE METHODOLOGY 3.1 Benchmark Analysis As a first step in computing the revised P-T curves, a benchmark evaluation was performed using ASME Code,Section XI, Appendix G methodology without application of ASME Code Case N 640 for comparison against NMP-1's previously developed P-T curves.
The intent was to reproduce the existing NMP-1 P-T curves for the pressure test and heatup/cooldown conditions so that consistent methodology could be applied to the revised P-T curves using Code Case N-640.
The P-T curve methodology is based on the requirements of References [2] through [4]. The supporting calculations for the curves are contained in References [8] and [9]. From the previous work performed for NMP-1 [5], the beltline region bounds all other regions with respect to brittle fracture.
The approach used for reproducing the previously developed NMP-1 P-T curves is summarized below:
- a.
Assume a fluid temperature, T.
- b.
For the temperature, T, assumed in step (a), compute the temperature at the assumed flaw tip, Tl/4t (i.e., for an ID 1/4t flaw or an OD 3/4t flaw). This is accomplished by adding a temperature drop term, AT1,4t, to T. AT11 4t values were obtained from the heat transfer analysis performed for the Reference [5] report for the appropriate heatup/cooldown conditions, as follows:
ATl/4t for pressure test curve: 0°F (no thermal for pressure test curve)
AT11 4t for core not critical curve: 23.151'F (for 100lF/hr cooldown curve)
AT1 /4t for core not critical curve: 00F (for 0°F/hr cooldown curve)
AT11 4t for core not critical curve: variable (for 100°F/hr heatup curve)
SIR-02-129, Rev. 0 3-1
- c.
Calculate the allowable stress intensity factor, Kia [4] based on Tl/4t using the following relationship:
Ka - 26.78 = 1.223 exp [0.0145 (T1 /4t - ART + 160)]
where:
T1t4t ART Kla
= flaw tip temperature (OF)
= limiting ART value, as defined above (°F)
=
allowable stress intensity factor (ksi4inch)
- d.
Calculate the thermal stress intensity factor, Krr, using the appropriate relationship from Figure G-2214-2 of Reference [4]:
Krr = Mt ATw where:
ATw Mt
=
through-wall temperature drop (OF)
= factor from Figure G-2214-2 of Reference [4].
= 0.3144 The values for AT, were obtained from the heat transfer analysis performed for the Reference [5] report for the appropriate heatup/cooldown conditions, as follows:
AT, for pressure test curve: 0°F (no thermal for pressure test curve)
ATw for core not critical curve: 47.169°F (for 100°F/hr cooldown curve)
AT, for core not critical curve:
0°F (for 0 °FIhr cooldown curve)
AT, for core not critical curve: variable (for 100OF/hr heatup curve)
- e.
Calculate the allowable pressure stress intensity factor, Kip, using the appropriate relationship for the P-T curve under consideration from ¶G-2215 and ¶G-2400 of Reference [4]:
SIR-02-129, Rev. 0 3-2
1.5K1 p = Kra for Curve A (i.e., pressure-test curve) 2.OKwp + Krr = Kja for Curves B and C (i.e., core not critical and core critical curves) where:
Krr
= thermal stress intensity factor (ksiqinch)
K1p
= allowable pressure stress intensity factor (ksiIinch)
- f.
Compute the pressure, P. The relationship for the pressure, P, to the allowable pressure stress intensity factor, Kip, is as follows from ¶G-2214 of Reference [4]:
Kw = Mm Gm + Mb Cab where:
Mm a'm P
R t
Mb ab
= membrane stress correction factor from Figure G-2214-1 of Reference [4].
= membrane stress due to pressure (ksi)
= PR/t for the beltline region.
= pressure (ksi)
= vessel inside radius (inches)
=
106.5" [5, Table 4-5]
= vessel wall thickness (inches)
= 7.125" [5, Table 4-5]
= bending stress correction factor
=
(2/3)Mm from Figure G-2214-1 of Reference [4].
= bending stress due to pressure (ksi)
= 0 for a thin-walled vessel Thus, P = Kipt/(RMm) for the beltline region.
The values for Mm were selected so that the previous P-T curve results from Reference [5] were matched, as follows:
SIR-02-129, Rev. 0 3-3
Mm for pressure test curve: 2.60 Mm for core not critical curve: 2.60 (for 100°F/hr cooldown curve)
Mm for core not critical curve: 2.60 (for 0°F/hr cooldown curve)
Mm for core not critical curve: variable (for 100°F/hr heatup curve)
(varied between 2.80 at lower temperatures to 2.60 at upper temperatures and linear interpolation)
- g.
Repeat steps (a) through (f) for other temperatures to generate a series of P-T points for each region.
- h.
Adjust for any applicable instrument errors for temperature and pressure from T and P, respectively. Instrument errors were documented to be 4.0°F for temperature and 10.0 psig for pressure for the leak/hydro curve [5, Table 4-5], and 12.2°F for temperature and 52.2 psig for pressure for the heatup/cooldown curves
[5, Table 4-5]*. An additional static head pressure adjustment of 15.4 psig was used to account for the weight of water in a full vessel.
The temperature and instrument uncertainties were not applied to the 10CFR50 Appendix G limits described below.
The following additional requirements were used to define the lower portion of the P-T curves.
These limits are established by the discontinuity regions of the vessel (i.e., flanges), and are specified in Reference [2]:
For Pressure Test Conditions:
"/ Thermal stresses were assumed to be negligible during the pressure test condition and were therefore not considered.
V If P is greater than 20% of the pre-service hydro test pressure, the upper vessel temperature must be greater than RTNDT of the limiting flange material + 90 0F. The pre-service hydro test pressure was 1,800 psig [5, Table 4-5].
/ If P is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is conservatively assumed to be greater than or equal to the RTNDrT of the limiting flange material + 60'F. This additional 60'F margin is not recommended in SIR-02-129, Rev. 0 3-4
Reference [2], but has been a standard recommendation by GE for the BWR industry and was conservatively used in the Reference [5] work. For the NMP-1 flange material, this minimum temperature is 100°F (based on a limiting RTNDT of 40'F for non-beltline materials per Table 4-5 of Reference [5]). Since the 60'F margin is only a recommendation, application of this extra limit is not required by References [2] or [4],
but was used in the benchmark evaluation to demonstrate reproduction of the Reference
[5] results.
For Core Not Critical Conditions:
" If P is greater than 20% of the pre-service hydro test pressure, the upper vessel temperature must be greater than RTNDT of the limiting flange material + 120'F.
/ If P is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is conservatively assumed to be greater than or equal to the RTNDT of the limiting flange material + 60'F. As identified above for the pressure test, this limit is only a recommendation, application of this extra limit is not required by References [2]
or [4], but was used in the benchmark evaluation to demonstrate reproduction of the Reference [5] results.
For Core Critical Conditions:
" Per the requirements of Table 1 of Reference [2], the core critical P-T limits must be 40'F above any Pressure Test or Core Not Critical curve limits. Core Not Critical conditions are more limiting than Pressure Test conditions, so Core Critical conditions are equal to Core Not Critical conditions plus 40'F.
" Another requirement of Table 1 of Reference [2] (or actually an allowance for the BWR), concerns minimum temperature for initial criticality in a startup. Given that water level is normal, BWRs are allowed initial criticality at the closure flange region temperature (RTNDT + 60'F) if the pressure is below 20% of the pre-service hydro test pressure. This corresponds to 100°F for NMP-1, as identified above.
,/ Also per Table 1 of Reference [2], at pressures above 20% of the pre-service hydro test pressure, the Core Critical curve upper vessel temperature must be at least that required for the pressure test. The temperature was determined in Reference [5] to be 260'F for SIR-02-129, Rev. 0 3-5
the cooldown condition and 278'F for the heatup condition. As a result of this requirement, the Core Critical curve must have at least a temperature of 260'F (for cooldown 1/4t flaws) or 278'F (for heatup 3/4t flaws) for pressures greater than 20% of the pre-service hydro pressure.
The resulting pressure and temperature series constitutes the P-T curve. The P-T curve relates the minimum required fluid temperature to the reactor pressure.
Tabulated values for the resulting benchmark P-T curves for 28 EFPYs are shown in Tables 3 1 through 3-5. The resulting P-T curves are plotted in Figures 3-1 through 3-3.
Based on the results shown in Figures 3-1 through 3-3, the following conclusions can be made with respect to the benchmark analysis:
/ For the Pressure Test curve, the results shown in Figure 3-1 demonstrate that the previous methodology was successfully reproduced. The independently derived curve is essentially identical to the pressure test curve developed in Reference [5]. Any differences are due to round-off and are insignificant.
" For the Core Not Critical curve, the results shown in Figure 3-2 demonstrate that the previous methodology was successfully reproduced. The independently derived curves are essentially identical to the core not critical curves developed in Reference [5]. Any differences are due to round-off and are insignificant.
v/ For the Core Critical curve, the results shown in Figure 3-3 demonstrate that the previous methodology was successfully reproduced. The independently derived curves are essentially identical to the core critical curves developed in Reference [5]. Any differences are due to round-off and are insignificant.
It is therefore concluded that the P-T curve methodology described above is consistent with and the same as that previously used to develop the P-T curves in the Reference [5] report.
SIR-02-129, Rev. 0 3-6
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~~Pressure-Temperature Curve Calculation-------
(Core Not Cnbcal - Bounding Cwve)
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336 13796 3360 397 197 SIR-02-129, Rev. 0 3-11
Figure 3-1. Benchmark Pressure Test P-T Curve 1,200 1,100 1,000 900 800.
700 Lu 600 500 400 300.
200-K 100 0
0 40 80 120 160 200 TEMPERATURE (°F)
NMP-1 Pressure and Temperature Limits Hydrostatic and Leak Tests < 28 EFPY 240 280 SIR-02-129, Rev. 0 3-12
Figure 3-2. Benchmark Core Not Critical Curve 1,200 1,100 1,000 900 800 700 600 500 400 300 200 100 0
40 80 120 160 200 240 TEMPERATURE ('F)
NMP-1 Pressure and Temperature Limits Care Not Critical Conditions : 28 EFPY SIR-02-129, Rev. 0
- 0.
uJ w
u) u)
ILl 0.
280 320 3-13
Figure 3-3. Benchmark Core Critical Curve 1,200 1,100 1,000 700-1 uJ cc 600 00 uLJ D.
500 400.
300 200 100 0
-I 0
40 80 120 160 200 240 280 320 360 400 TEMPERATURE ('F)
NMP-1 Pressure and Temperature Limits Core Critical Conditions <28 EFPY SIR-02-129, Rev. 0 3-14
3.2 Revised P-T Curves Revised P-T curves for 28 EFPYs were developed for NMP-1 using ASME Code Case N-640
[1]. The same methodology described above was used with three exceptions. First, Step (c) was modified to use Klc in place of Kia, as follows:
KIc = 20.734 etO 02(Tr1I4-ARTNDT)] + 33.2
[10, A-4200]
Second, for the Pressure Test case only, the static head described in Step (h) was determined to be 20.8 psig, for a 576" water column (bottom of beltline to inside surface of top head) at a density of 62.4lb/ft3 at 70°F. All other curves used a static head based on the operating temperature.
Third, the minimum temperature requirements for the core critical curve were changed as follows:
For Core Critical Conditions:
v" Per Table 1 of Reference [6], at pressures above 20% of the pre-service hydro test pressure, the Core Critical curve upper vessel temperature must be at least that required for the pressure test. The temperature requirements is as follows:
Minimum Temperature for Critical Core Operation A=
P=
Sm =
ays =
Mm =
K, =
ART(1/4t) =
ART(3/4t) =
113.625 106.5 1,030 17,520 69.4 2.52 66.299 167.7 136.8 inches inches psig psi ksi ksi(inch)ub using KIc OF OF SIR-02-129, Rev. 0 Tcra =
Tcrt =
AT1/4T + Terror 191 OF 12.2 160 OF 62.0 T.
203 222 I 3-15
As a result of this requirement, the Core Critical curve must have at least a temperature of 203'F (for cooldown 1/4t flaws) or 222°F (for heatup 3/4t flaws) for pressures greater than 20% of the pre-service hydro pressure.
Tabulated values for the resulting P-T curves are shown in Tables 3-6 through 3-10. The resulting P-T curves are shown in Figures 3-4 through 3-8.
P-T curve plots and tabulations formatted consistent with the NMP-1 plant Technical Specifications are provided in Appendix A.
SIR-02-129, Rev. 0 3-16
Table 3-6. Tabulated Values for Revised Pressure Test P-T Curve for 28 EFPY Pressure-Temperature Curve Calculation (Pressure Test)
Plant =
Component =
Vessel thickness, t =
M, Vessel Radius, R =
AT.=
AT1/4t Safety Factor =
Mm =
Temperature Instrument Error =
Pressure Instrument Error =
Hydro Test Pressure =
Flange RTNDT Fluid Temperature T
(1F) 100 100 126 126 131 136 141 146 151 156 161 166 171 176 181 186 191 196 201 206 211 216 221 226 231 236 241 1/4t Temperature
(*F) 1000 1000 126.0 126.0 131.0 1360 141.0 1460 151.0 1560 161.0 166.0 171.0 176.0 181.0 1860 191.0 196.0 201.0 206.0 211.0 216.0 221.0 226.0 231.0 236 0 241.0 Beitline 7.125" inches 0.3144 (per Figure G-2214-2 of Section Xl, Appendix G) 106.5 inches 167.7 OF===>
ýE
'Fpy 0.0
°F (no thermal for pressure test) 0.0 ksi*nch'u (no thermal for pressure test) 0.0 OF (no thermal for pressure test) 1.50 (for pressure test) 2.60 (per Figure G-2214-1 of App. G, assumed value to match Reference [1] results) 4.0 OF 36.2 l psig (instrument uncertainty + 20 8 psig static head) 1,800
~psig 40 -OF Ki.
(ksl*Inch 11 2) 3855 3855 42.20 42.20 43.15 44.20 45.36 4663 4805 4961 51.33 5324 55.35 57.68 60.25 63.10 66.24 69.72 73.56 77.80 82.49 87.68 93.41 99.74 10674 114.47 12302 K~p (ksi*inchl")
25.70 25.70 28.14 2814 28.77 2947 30.24 31.09 32.03 3307 3422 3549 36.90 38.45 40.17 42.07 44.16 4648 4904 51.87 5500 5845 62.27 6649 71.16 76.31 82.01 Calculated Pressure P
(psig) 0 360 360 724 740 758 778 800 824 851 881 913 949 989 1034 1082 1136 1196 1262 1335 1415 1504 1602 1711 1831 1964 2110 Adjusted Temperature for P-T Curve
(°F) 100 100 130 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 Adjusted Pressure for P-T Curve (psig) 0 360 360 688 704 722 742 764 788 815 844 877 913 953 997 1046 1100 1160 1226 1298 1379 1468 1566 1675 1795 1927 2074 SIR-02-129, Rev. 0 3-17
Table 3-7. Tabulated Values for Revised Core Not Critical (100°F/hr Cooldown) P-T Curve for 28 EFPY Pressure-Temperature Curve Calculation (Core Not Cntcal -- 100*F/hr cooldown)
Plant =
Component =
Vessel thickness, t =
M, Vessel Radius, R =
ART=
AT.T =
KIT =
ATVl4i
=
Safety Factor =
MT =
Temperature Instrument Error=
Pressure Instrument Error =
Hydro Test Pressure =
Flange RT140r =
Fluid Temperature 114t T
Temperature
('F)
F) 40 63.2 45 68.2 50 732 55 782 60 832 65 882 70 932 75 982 80 1032 85 1082 88 1110 90 1132 95 1182 100 1232 105 1282 110 1332 115 1382 120 1432 125 1482 128 151 1 138 1612 140 1632 145 1682 148 1712 148 171.2 1508 1740 1508 1740 155 1782 160 1832 165 1882 170 1932 175 1982 180 2032 185 2082 190 2132 195 2182 200 2232 205 2282 210 2332 215 2382 220 2432 225 2482 230 2532 235 2582 240 2632 245 2682 7.125
- jInches 0.3144 (per Figure G-2214-2 of Section XI, Appendix G) 1065.5 inches 167.7
- F MI-!eEFPY_=
47.169
' F (use value at 70'F from past work) 14.8 ý ksi'inch" 23.151
'F (use value at 70*F from past work) 2.006<.
(for heatup/cooldown) 2.60 (per Figure G-2214-1 of App G, assumed to be the same as Curve A) 12.2
¶ 67.6 *
, psig (instrument uncertainty + 15 4 psig stabc head) 1,600
'psg Calculated Adjusted Adjusted Adjusted Pressure Temperature Pressure for Temperature P
for P-T Curve P-T Curve for Curve C Kk KP (ksl*lnch'na I
ksl*inch'a I 3576 3603 3633 3666 3702 3742 3787 3836 3890 3950 3986 4016 4090 41 71 4260 4359 4468 4589 4722 4806 5139 52.13 54.12 5542 5542 5670 5670 5875 61.44 6441 6769 7132 7533 7976 8466 9007 9605 10266 10997 11804 12697 13683 147.73 15977 17308 18779 1047 1060 1075 1091 1110 1130 1152 11 76 1204 1234 1252 1267 1303 1344 1389 1438 1493 1553 1620 1662 1828 1865 1965 2029 20.29 2093 2093 2196 2331 2479 2643 2825 3025 3247 3491 37.62 4061 4392 4757 51 61 5607 61 00 6645 7247 7913 8648 (psi9) 0 273 277 281 286 291 296 303 310 317 322 326 335 346 357 370 384 400 417 428 470 480 506 522 522 539 539 565 600 638 680 727 778 835 898 968 1045 1130 1224 1328 1443 1570 1710 1865 2036 2225 (1F) 100 100 100 100 100 100 100 100 100 100 100 102 107 112 117 122 127 132 137 140 150 152 157 160 160 163 163 167 172 177 182 187 192 197 202 207 212 217 222 227 232 237 242 247 252 257 (psfq) 0 205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 360 360 360 455 471 471 498 532 570 613 659 711 768 831 900 977 1062 1156 1260 1375 1502 1642 1797 1968 2158
(*F) 52 57 62 67 72 77 82 87 92 97 100 102 107 112 117 122 127 132 137 140 150 152 157 160 160 163 163 167 172 177 182 187 192 197 202 207 212 217 222 227 232 237 242 247 252 257 SIR-02-129, Rev. 0 3-18
Table 3-8. Tabulated Values for Revised Core Not Critical (O0F/hr) P-T Curve for 28 EFPY Pressure-Temperature Curve Calculation (Core Not Cnrtcal - O0F4hr cooldown)
Plant Component Vessel thickness, I
?A, Vessel Radius, R ART AT.
Krr ATv41=
Safety Factor Mm.
Temperature Instrument Error Pressure Instrument Error Hydro Test Pressure =
Fluid Temperature 114t T
Temperature C'F) 40 45 50 55 60 65 70 75 80 85 88 90 95 100 105 110 115 120 125 130 135 140 145 148 149 151 152 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 400 450 500 550 600 650 700 75 0 80 0 85 0 87 8 900 95 0 1000 105 0 1100 1150 1200 1250 1300 1350 1400 1450 1480 1490 1508 1520 1550 1600 1650 1700 1750 1800 1850 1900 1950 2000 205 0 2100 2150 220 0 225 0 2300 235 0 2400 245 0 250 0 255 0 2600 BeltilneJ 7.125 rinches 0.3144 J(per Figure G-2214-2 of Section Xl, Appendix G) 165 Inches
- 16.
F
-=>
2-8"EFy72 0.0
- F (use value at 100°F from past work) o0o ksiinrch"'
7 00
- F (use value at 1 00F from past work) 2 00
~(for heatup/cooldown) 2.60 f(per Figure G-2214-1 of App G, assumed to be the same as Curve A) 12.2 TF 67-2 Ipsig (instrument uncertainty + 15 4 psig static head) 1,500 psig F
K, I~p (ksl~lneh""i flsl~lneh""!
3481 3498 3517 3538 3561 3586 3614 3645 3679 3717 3739 3758 3804 3855 3912 3974 4043 4119 4203 4295 4398 4511 4637 4718 4746 4799 4835 4928 5097 5284 5491 5719 5972 6251 6559 6899 7276 7692 81 52 8660 9221 9842 10528 11286 121.24 13050 14073 152 04 16454 1741 1749 1758 1769 1780 1793 1807 1822 1839 1858 1870 1879 1902 1928 1956 1987 2021 2059 2101 2148 2199 22 56 2318 2359 2373 2400 2417 24 64 2549 2642 2746 2860 2986 3125 3279 3450 3638 3846 4076 4330 4611 4921 5264 5643 6062 6525 7037 7602 8227 Calculated Adjusted Adjusted Adjusted Pressure Temperature Pressure for Temperature P
forP-TCurve P-TCurve for Curve C (psig) 0 450 452 455 458 461 465 469 473 478 481 484 489 496 503 511 520 530 541 553 566 580 597 607 611 617 622 634 656 680 706 736 768 804 844 888 936 990 1049 1114 1186 1266 1354 1452 1560 1679 1811 1956 2117
('F)
(pslg) 100 0
100 383 100 385 100 388 100 391 100 394 100 398 100 402 100 406 100 411 100 414 102 416 107 422 112 429 117 436 122 444 127 453 132 463 137 474 142 485 147 499 152 513 157 529 160 540 161 543 163 550 164 555 167 567 172 589 177 613 182 639 187 669 192 701 197 737 202 777 207 820 212 869 217 922 222 982 227 1047 232 1119 237 1199 242 1287 247 1385 252 1493 257 1612 262 1743 267 1889 272 2050 C'F) 52 57 62 67 72 77 82 87 92 97 100 102 107 112 117 122 127 132 137 142 147 152 157 160 161 163 164 167 172 177 182 187 192 197 202 207 212 217 222 227 232 237 242 247 252 257 262 267 272 SIR-02-129, Rev. 0 3-19 r
Table 3-9. Tabulated Values for Revised Core Not Critical (100°F/hr Heatup) P-T Curve for 28 EFPY Pressure-Temperature Curve Calculation (Core Not Critical - 100°Flhrheatup)
Plant Component Vessel thickness, t M4 Vessel Radius, R ART Safety Factor Temperature instrument Error Pressure Instrument Error.
Hydro Test Pressure Flange RT-rD 7.125 inches 0.1144 (per Figure G-2214-2 of Section X), Appendix G) 2.00 (for heatup/cooldown) 12.2 Pg 67.6 psig (instrument uncertainty.+ 15 4 psig static head) 1,8&00 psig 140' I-F 114t AT,,
Temperature
('F)
C'F) 0000 400 0000 450 0000 500 0000 550 0000 600 0000 650
- 0078 699
-4801 702
-9523 705
-12123 729
-14861 729
-17324 727
-20336 747
-23348 767
-25662 793
-27975 820
-29754 852
-31532 685
-32901 921
-34270 957
-35328 997
-36386 1036
-37207 1078
-38027 1120
-38667 1163
-39306 1207
-39810 1252
-40313 1295
.40.313 1295
-40714 1341
-41115 1387
.41441 1434 767 1480
-42037 1528
.42307 1575
-42532 1623
-42757 1670
-42939 1719
-43120 1767
.43265 1825
-43265 1825
-43409 1866
-43526 1915
-43643 1964
-43 742 201 3
-43840 2062
-43923 211.1
-44006 2160
-44078 2209
-44150 2259
-44214 2308
-44278 2357
-44337 2407
-44395 2456
-44452 2505
-44508 2555 K.
(ksl*lnchw) 3619 3651 36 85 3724 3766 3813 3864 3867 3870 3897 3898 3895 3918 3943 3977 4013 4059 4109 4168 4232 4307 4388 4481 4582 4697 4822 4964 5111 5112 5284 54 73 5684 5916 6173 6456 6771 7116 7500 7923 8495 8496 8933 9508 10143 10846 11621 12479 13425 14472 15627 16904 18314 19871 20000 20000 20000 AT.
Kg i p
(*F)
(kli'tnch") (ksl*lnch' 5 )
0000 000 0000 000 0000 000 0000 000 0000 000 0000 000 0078 002 4897 154 9715 305 12480 392 15398 484 18009 566 21.236 668 24463 769 26944 847 29424 925 31331 985 33238 1045 34707 1091 36175 11.37 37310 1173 38444 1209 39324 1236 40204 1264 40890 1286 41576 1307 42115 1324 42654 1341 42654 1341 43083 1355 43512 1368 43860 1379 44207 1390 44472 1398 44720 1406 44955 1413 45.267 1423 45463 1429 45659 1436 45814 1440 45969 1445 46095 1449 46220 1453 46325 1456 46430 1460 46520 1463 46610 1465 46688 1468 46765 1470 46834 1472 46902 1475 46965 1477 47027 1479 47087 1480 47147 1482 47206 1484 1810 1825 1843 1862 1883 1907 1931 1857 1782 1753 1707 1664 1625 1587 1565 1544 1537 1532 1538 1547 1567 1589 1622 1659 1706 1758 18.20 18 85 18 85 19 65 2053 21 53 2263 2388 2525 2679 2847 3036 3244 3527 3526 3742 4028 4343 4693 5079 5507 5979 6501 7077 7715 8419 9196 9260 9259 9258 280 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 2 80 280 280 280 280 278 277 275 274 272 2 71 269 268 266 2 66 2 65 263 262 260 260 2 60 2 60 260 2 60 260 260 2 60 2 60 2 60 2 60 2 60 2 60 2 60 2 60 2 60 2 60 260 260 2 60 260 260 260 Calculated Adjusted Adjusted Adjusted Pressure Temperature Pressure for Temperature P
for P-T Curve P-T Curve for Curve C (pslg)
(°F)
(pslg)
(*F) 0 100 0
52 436 100 298 57 440 100 298 62 445 100 298 67 450 100 298 72 456 100 298 77 461 100 298 82 444 100 298 87 426 100 298 92 419 100 298 97 408 100 298 100 398 102 298 102 388 107 298 107 379 112 298 112 374 117 298 117 369 122 298 122 367 127 298 127 366 132 298 132 368 137 300 137 372 142 304 142 379 147 311 147 386 152 319 152 396 157 329 157 408 162 340 162 421 167 354 167 437 172 369 172 455 177 387 177 474 182 406 182 474 182 406 182 497 187 429 187 522 192 454 192 551 197 483 197 582 202 515 202 614 207 547 207 650 212 582 212 689 217 622 217 732 222 665 222 781 227 713 227 83M 232 767 232 908 238 840 238 907 238 840 238 963 242 895 242 1036 247 969 247 1118 252 1050 252 1208 257 1140 257 1307 282 1239 262 1417 267 1349 267 1538 272 1471 272 1673 277 1605 277 1821 282 1754 282 1985 287 1918 287 2166 292 2099 292 2366 297 2299 297 2383 302 2315 302 2382 307 2315 307 2382 312 2315 312 SIR-02-129, Rev. 0 Fluid Temperature T
40 45 50 55 60 65 70 75 80 85 88 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 169 8 169 8 175 180 185 190 195 200 205 210 215 220 225 8 225 8 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 3-20 I
I T
Table 3-10. Tabulated Values for Revised Core Critical P-T Curves for 28 EFPY Pressure-Temperature Curve Calculatio (Core Critical)
CrIdnwn rurye IpulsPlant
-=
-n NMlIiOPa1ý EFRY 28 Cntjcal Temperature -
203
°F Hydro Test Pressure 1,800 psig Flange RTýDT =
4 F
Adjusted Curve B Temperature (IF) 52 57 62 67 72 77 82 87 92 97 100 102 107 112 117 122 127 132 137 140 147 152 157 160 160 163 163 167 172 177 182 187 192 197 202 207 212 217 222 227 232 237 242 247 252 257 262 267 272 Adjusted Curve B Pressure for (psl2) 0 205
,209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 360 360 360 455 471 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 1199 1287 1385 1493 1612 1743 1889 2050 Curve C Temperature
('F) 100 100 102 107 112 117 122 127 132 137 140 142 147 152 157 162 167 172 177 180 187 192 197 200 200 203 203 207 212 217 222 227 232 237 242 247 252 257 262 267 272 277 282 287 292 297 302 307 312 Curve C Pressure (psfg) 0 205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 360 360 360 360 360 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 1199 1287 1385 1493 1612 1743 1889 2050 Adjusted Curve B Temperature (IF) 52 57 62 67 72 77 82 87 92 97 100 102 107 112 117 122 127 132 137 142 147 152 157 162 167 172 177 182 182 187 192 197 202 207 212 217 222 227 232 238 238 242 247 252 257 262 267 272 277 282 287 292 297 302 307 312 317 322 327 HatupJ Curve Plant -
IU1P-i~37~IiV EFPY '
28 Cntlcal Temperature =
222 F
Hydro Test Pressure =
- 1,BX00 ps1g Flange RTNoT =
F Adjusted Curve 5 Curve C Pressure far Temperature (pslg)
(IF) 0 100 298 100 298 102 298 107 298 112 298 117 298 122 298 127 298 132 298 137 298 140 298 142 298 147 298 152 298 157 298 162 298 167 298 172 300 177 304 182 311 187 319 192 329 197 340 202 354 207 369 212 387 217 406 222 406 222 429 227 454 232 483 237 515 242 547 247 582 252 622 257 665 262 713 267 767 272 840 278 840 278 895 282 969 287 1050 292 1140 297 1239 302 1349 307 1471 312 1605 317 1754 322 1918 327 2099 332 2299 337 2315 342 2315 347 2315 352 2314 357 2314 362 2314 367 Curve C Pressure (pslQ) 0 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 298 300 304 311 319 329 340 354 360 360 360 406 429 454 483 515 547 582 622 665 713 767 840 840 895 969 1050 1140 1239 1349 1471 1605 1754 1918 2099 2299 2315 2315 2315 2314 2314 2314 3-21 SIR-02-129, Rev. 0
Figure 3-4. Revised Pressure Test P-T Curve for 28 EFPY 1,200 1,100 1,000 900 800.
""o 700
- 0.
1400 2 600 U,
00 LU
- a.
500 400 300 200 100-0-
0 40 80 120 160 200 TEMPERATURE (°F)
NMP-1 Pressure and Temperature Limits Hydrostatic and Leak Tests < 28 EFPY 240 280 SIR-02-129, Rev. 0 3-22
Figure 3-5. Revised Cooldown Core Not Critical P-T Curve for 28 EFPY 1,200 1,100-,
1,000 100 F/hr 900 800 700 CC 600 Co 400 300 200 100 0
i 0
40 80 120 160 200 240 2
TEMPERATURE (°F)
NMP-1 Pressure and Temperature Limits Core Not Critical Conditions (COOLDOWN) < 28 EFPY SIR-02-129, Rev. 0
.80 320 3-23
Figure 3-6. Revised Heatup Core Not Critical P-T Curve for 28 EFPY 1,200 1,100 1,000 900 800 700 600 50 400 300 200 100 0
0 40 80 120 160 200 240 TEMPERATURE ('F)
NMP-1 Pressure and Temperature Limits Core Not Critical Conditions (HEATUP) _ 28 EFPY SIR-02-129, Rev. 0
- 0.
U, IU w:
0.
280 320 3 -24
Figure 3-7. Revised Cooldown Core Critical P-T Curve for 28 EFPY (n
LIu U,
aw9 Q.l 1,200 1,100 1,000 900 80o 700 600 50 400 300 200 100 0
0 40 80 120 160 200 240 280 320 360 400 TEMPERATURE (°F)
NMP-1 Pressure and Temperature Limits Core Critical Conditions (COOLDOWN) _ 28 EFPY SIR-02-129, Rev. 0 3-25
Figure 3-8. Revised Heatup Core Critical P-T Curve for 28 EFPY 1,200 1,100 1,000 900 800 700 600 500 400 300 200 100 0
0 40 80 120 160 200 240 280 TEMPERATURE ('F)
NMP-1 Pressure and Temperature Limits Core Critical Conditions (HEATUP) < 28 EFPY 320 360 400 SIR-02-129, Rev. 0 IUJ 0.
3-26
4.0 CONCLUSION
S The revised P-T curves for NMP-1 are shown in Figures 3-4 through 3-8 for 28 EFPYs for incorporation into the NMP-1 plant Technical Specifications. The curves utilize the same methodology as was used for the previously approved P-T curves with the exception that K1c was applied in place of KI as allowed by ASME Code Case N-640 [1].
SIR-02-129, Rev. 0 4-1
5.0 REFERENCES
- 1.
ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,"Section XI, Division 1, Approved February 26, 1999.
- 2.
U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," 1-1-98 Edition.
- 3.
Welding Research Council Bulletin No. 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," PVRC Ad Hoc Group on Toughness Requirements, Welding Research Council, August 1972.
- 4.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1989 Edition.
- 5.
MPM Report No. MPM-59838, "Pressure-Temperature Operating Curves for Nine Mile Point Unit 1," May 1998, SI File No. NMP-05Q-207.
- 6.
USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 305-4), May 1988.
- 7.
Niagara Mohawk Power Corporation Document No. FA98-195, "Pressure-Temperature Operating Curves for Nine Mile Point Unit 1, Rev. 1," 12/16/98, SI File No. NMP-05Q 207.
- 8.
Structural Integrity Associates Calculation No. NMP-05Q-301, Revision 0, "Benchmark Analysis," 11/5/02.
- 9.
Structural Integrity Associates Calculation No. NMP-05Q-302, Revision 0, "P-T Curves Generated Using Code Case N-640," 11/5/02.
- 10.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix A, "Analysis of Flaws,"
1995 Edition.
SIR-02-129, Rev. 0 5-1
APPENDIX A P-T CURVE PLOTS AND TABULATIONS IN TECHNICAL SPECIFICATION FORMAT SIR-02-129, Rev. 0 Al
LIMIT FOR NON-CRITICAL OPERATION HEATUP AT UP TO 1000FIHR REACTOR PRESSURE (psig)
IN TOP DOME 0
298 298 298 298 298 298 298 298 300 304 311 319 329 340 354 369 387 406 406 429 454 483 515 547 582 622 665 713 767 840 840 895 969 1050 1140 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE ('F) 100 100 102 107 112 117 122 127 132 137 142 147 152 157 162 167 172 177 182 182 187 192 197 202 207 212 217 222 227 232 238 238 242 247 252 257 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.a MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEAT-UP (CORE NOT CRITICAL) (HEATING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION SIR-02-129, Rev. 0 A2
LIMIT FOR NON-CRITICAL OPERATION COOLDOWN AT UP TO 1000F/HR REACTOR PRESSURE (psig)
IN TOP DOME 0
205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 455 471 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 100 100 100 100 100 100 100 100 100 102 107 112 117 122 127 132 137 140 160 160 163 163 167 172 177 182 187 192 197 202 207 212 217 222 227 232 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.b MINIMUM TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE NOT CRITICAL) (COOLING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION SIR-02-129, Rev. 0 A3
LIMIT FOR POWER OPERATION (CORE CRITICAL)
HEATUP AT UP TO 1000FIHR REACTOR PRESSURE (psig)
IN TOP DOME 0
298 298 300 304 311 319 329 340 354 360 360 360 406 429 454 483 515 547 582 622 665 713 767 840 840 895 969 1050 1140 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 172 177 182 187 192 197 202 207 212 217 222 a 222 a 227 232 237 242 247 252 257 262 267 272 278 278 282 287 292 297 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
("water level must be in range for power operation if core is critical below 222'F)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.c MINIMUM TEMPERATURE FOR PRESSURIZATION DURING HEATUP (CORE CRITICAL) (HEATING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION SIR-02-129, Rev. 0 A4
LIMIT FOR POWER OPERATION (CORE CRITICAL)
COOLING AT UP TO 100°F/HR REACTOR PRESSURE (psig)
IN TOP DOME 0
205 209 213 218 223 229 235 242 250 254 258 268 278 290 302 316 332 349 360 360 360 471 498 532 570 613 659 701 737 777 820 869 922 982 1047 1119 1199 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 100 100 102 107 112 117 122 127 132 137 140 142 147 152 157 162 167 172 177 180 200 203 a 203 a 207 212 217 222 227 232 237 242 247 252 257 262 267 272 277 (reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(awater level must be in range for power operation if core is critical below 203°F)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.d MINIMUM TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN (CORE CRITICAL) (COOLING RATE < 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION SIR-02-129, Rev. 0 A5
LIMIT FOR IN-SERVICE TEST (CORE NOT CRITICAL, FUEL IN VESSEL)
REACTOR PRESSURE (psig)
IN TOP DOME REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF) 0 360 360 688 704 722 742 764 788 815 844 877 913 953 997 1046 1100 1160 100 100 130 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 (reactor vessel beltline downcomer water temperature Is measured at recirculation loop suction)
(instrument uncertainties have been included in this table)
TABLE 3.2.2.e MINIMUM TEMPERATURE FOR PRESSURIZATION DURING LEAKIHYDROSTATIC TESTING (CORE NOT CRITICAL)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF CORE OPERATION SIR-02-129, Rev. 0 A6
HEATUP - CORE NOT CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
100°F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (*F)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.a MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HEATUP AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (HEATING RATE _ 100°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION SIR-02-129, Rev. 0 A7
COOLDOWN - CORE NOT CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
100°F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE ('F)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.b MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING COOLDOWN AND LOW-POWER PHYSICS TESTS (CORE NOT CRITICAL) (COOLING RATE *1 00°F/HR)
FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION SIR-02-129, Rev. 0
- a.
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a a
a
.mI A8
HEATUP - CORE CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
I OOOF If Cor Is Critical SJ-.-.------------J I
- 1.
, //
PBeowe 222rtio 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.c MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (HEATING RATE < 100°FIHR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION SIR-02-129, Rev. 0 A9
COOLDOWN - CORE CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
100OF Water Level Must Be in Range For Power Operation If Core Is Critical Below 2030F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been included in this figure)
FIGURE 3.2.2.d MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING CORE OPERATION (CORE CRITICAL) (COOLING RATE t100°F/HR) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION SIR-02-129, Rev. 0 A10
LEAK/HYDRO TEST - CORE NOT CRITICAL 1500 1000 REACTOR PRESSURE (psig) 500 0
Minimum Temperature for Boltup:
100°F 0
50 100 150 200 250 300 350 400 REACTOR VESSEL BELTLINE DOWNCOMER WATER TEMPERATURE (OF)
(reactor vessel beltline downcomer water temperature is measured at recirculation loop suction)
(instrument uncertainties have been Included In this figure)
FIGURE 3.2.2.e MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING IN-SERVICE HYDROSTATIC TESTING AND LEAK TESTING (CORE NOT CRITICAL) FOR UP TO 28 EFFECTIVE FULL POWER YEARS OF OPERATION SIR-02-129, Rev. 0 a
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