NLS2012103, Calculation Nedc 07-048, Ccn 5C3, Revised Pressure Temperature Curves.

Calculation Nedc 07-048, Ccn 5C3, Revised Pressure Temperature Curves.

ML12278A139
Person / Time
Site:	Cooper
Issue date:	08/27/2012
From:	Thomas K Nebraska Public Power District (NPPD)
To:	Office of Nuclear Reactor Regulation
References
NLS2012103 NEDC07-048 CCN5C3, Rev 2
Download: ML12278A139 (9)

Text

NLS2012103 Enclosure Enclosure Calculation NEDC 07-048 CCN 5C3, Revised Pressure Temperature Curves (49 pages)

Page 1 of 8

Title:

Revised Pressure Temperature Curves Calculation Number: NEDC07-048 CCN5C3 CEDIONumber 10-028 rev. 2 System/Structure: NB Setpoint Change/Part Eval Number: N/A Component: RPV Discipline: Civil Classification: X] Essential; [ Non-Essential SQAP Requirements Met? [ ] YES; [X] N/A Proprietary Information Included? [ YES; [X] NO Stand-Alone? [X] YES; [ ]NO

==

Description:==

This calculation develops the pressure temperature (P-T) curves for the beltline for 32 effective full power years (EFPY). The calculation includes the 2% power uprale, which was implemented after RE24. The calculation also provides P-T curves for 54 EFPY.

Revision 1 changed the minimum lemperature in Table 3 and Figure 1 to 70°F since the moderator temperature for the shutdown margin calculation is 68°F. Also revised reference 2 to the latest version.

Revision 2 revises curves A and B to add the lower head and upper vessel (all EFPY) regions to show thal the beltline curves are limiting.

Revision 3 corrects the graph of Curve A for the FW Nozzle and Curve C. Ref. CR-2010-2176.

Revision 4 changed the implementing EE number and noted the revisions to Design Inputs 3 and 4. Changed Design Input 2 to NEDC 07-032. These changes did not affect the calculation. None of the previous revisions were docketed with the NRC.

Revision 5 corrects Curve B for the lower head region (Ref CR-2010-4548). Also corrected a typo in a TS reference listed in the first paragraph under methodology.

Revision 5C1 accepts the curves developed by Structural Integrity for 24 month fuel cycles Revision 5C2 corrects typographical errors on page 5 of revision SCI. There are no changes to the SIA calculations.

Revision 5C3 accepts the revised curves developed by Structural Integrity for 24 month fuel cycles, which removed the discussion on the instrument nozzles.

Conclusions and Recommendations: The calculation to revise the P-T curves to 32 EFPY and 54 EFPY is consistent with the design basis of the plant. The P-T curves were developed in accordance with the methods in the 2001 Edition, 2003 Addenda of ASME Section XI, Appendix G. The curves generated from these methods ensure that the P-T limits will not be exceeded during any phase of reactor operation. These curves are suitable for inclusion in the Technical Specifications, USAR and other affected documents following NRC approval.

5C3 3 Structural Integrity K ./ ,././omo /,A Associates 8/24/2011 K.

5C2 3 Structural AssociatesIntegrity K.B. Thomas 2/1512012 N/A Bri Wolken 8&05/2011 3/30/2012 501 3 Structural Integrity K. B. Thomas 8/07/2011 N/A Brian Wolken Associates 8/05/2011 8118/2011 William 5W27/2010Green 5 3 K.B. Thomas Kirby Woods Kirby Woods William 5W27/2010Green 4 3 K.B. Thomas Kirby Woods Kirby Woods Rev.

Number Status Prepared By/Date Reviewed By/Date IDVed By/Date Approved By/Date Status Codes

1. Active 4. Superseded or Deleted 7. PRA/PSA

2. Information Only 5. OD/OE Support Only

3. Pending 6. Maintenance Activity Support Only

Page: 2 of 8 NEDC: 07-048 Rev. Number: 5C3 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REVISION PENDING CHANGES NUMBER DESIGN INPUTS NUMBER TO DESIGN INPUTS ASME Code,Section XI, Rules for 1 Inservice Inspection of Nuclear N/A 2001 Edition, 2003 Addenda Power Plant Components, Appendix G 2 NEDC 07-032, Vessel Fluence 1 iC1 3 NEDC 07-045, ART 1 C01 4 NEDC 07-046, USE 1 None 5 BWRVIP-74-A 0 None Memo DED 2003-005, Alan Able to Ken Thomas, dated August 14, 6 2003, "Instrument Uncertainty 0 None Associated With Technical Specification 3.4.9 7 CE Dwg No. E-232-230 3 None 8 GE Dwg No. 729E479-B, Sheet 1 2 None of 3 MISC-PENG-ER-012, Cooper 9 Reactor Pressure Vessel, Plates, 0 None Forgings, Welds and Cladding 10 OP 2.2.47, HVAC Reactor Building 45 None NEDC99-020, Review Of Structural 11 Integrity Report Sir-99-069 And 1 1CI Calculations No. NPPD-13Q-301,

_NPPD-13Q-302, NPPD-13Q-303 12 CE Stress Report, CENC-1150 NIA April 1971 13 CE Dwg No. E-232-243 7 None GE-NE-523-159-1292, Vessel 14 Surveillance MaterialsAnalysis Testing and NIA February 1993 Fracture Toughness

_Report

Page: 3 of 8 NEDC: 07-048 Rev. Number: 5C3 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM REVISION NUMBER AFFECTED DOCUMENTS NUMBER 1 TS 3.4.9 and Bases 07/16/08 2 6.MISC.502, ASME Class 1 System Leakage Test 40 3 6.MISC.504, ASME Class 1 Hydrostatic Test 13 6.RCS.601, Technical Specification Monitoring of RCS Heatup/Cooldown 18 Rate 5 PMIS 07, PMIS HYDRO display N/A 6 OP 2.1.1, Startup Procedure 164 7 USAR, Section IV2.6.3 8/07(08 8 USAR, Section IV2.6.2 8/07/08

Page: 4 of 8 NEDC: 07-048 Rev. Number: 503 The purpose of this form is to assist the Preparer in screening2 new and revised design calculations to determine potential impacts to procedures and plant operations.-

SCREENING QUESTIONS YES NO UNCERTAIN

1. Does it involve the addition, deletion, or manipulation of a component or I[ [X] [1 components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?

2. Could it impact system operating parameters (e.g., temperatures, flowrates, [X] [I II pressures, voltage, or fluid chemistry)?

3. Does it impact equipment operation or response such as valve closure [ [XI I[I time?

4. Does it involve assumptions or necessitate changes to the sequencing of II [XJ II operational steps?

5. Are there new assumptions/requirements that 7 need to be translated to new [X1 [] []

or existing site procedures/documents?

6. Does it transfer an electrical load to a different circuit, or impact when [] (x] []

electrical loads are added to or removed from the system during an event?

7. Does it influence fuse, breaker, or relay coordination? [1 I] [X1 []

8. Does it have the potential to affect the analyzed conditions of the environment for any part of the Reactor Building, Containment, or Control Room?

9. Does it affect TS/TS Bases, USAR, or other Licensing Basis documents? [Xl [I []

10. Does it affect a Dry Fuel Storage (CoC) or associated Technical [I I X] I Specification, Dry Fuel Storage UFSAR, or CNS 10CFR72.212 Report?

11. Does it constitute any change or addition to, or removal from, the [I I X] []

Independent Spent Fuel Storage Installation (ISFSI) facility or spent fuel storage cask design, or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished?

12. Does it affect DCDs? II [X] [I

13. Does it impact the Maintenance Rule Program? Refer to Procedure EDP-06 IDR selection checklist. II jX] II

14. Does it have the potential to affect procedures in any way not already H] (X] []

mentioned (refer to review checklists in Procedure EDP-06)? If so, identify:

If all answers are NO, then additional review or assistance is not required.

Ifany answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2.

Page: 5 of 8 NEDC: 07-048 Rev. Number: 5C3 Record of Revisions Revision 1 changed the minimum temperature in Table 3 and Figure 1 to 70°F since the moderator temperature for the shutdown margin calculation is 68'F. Also revised reference 2 to the latest version.

Revision 2 revises curves A and B to add the lower head and upper vessel (all EFPY) regions to show that the beltline curves are limiting.

Revision 3 corrects the graph of Curve A for the FW Nozzle and Curve C. Ref. CR-2010-2176.

Revision 4 changed the implementing EE number and noted the revisions to Design Inputs 3 and 4. Changed Design Input 2 to NEDC 07-032. These changes did not affect the calculation. None of the previous revisions were docketed with the NRC.

Revision 5 corrects Curve B for the lower head region (Ref CR-2010-4548). Also corrected a typo in a TS reference listed in the first paragraph under methodology.

Revision 5C1 accepts the curves developed by Structural Integrity for 24 month fuel cycles.

Revision 5C2 corrects typographical errors on page 5 of revision 5C1. There are no changes to the SIA calculations.

Revision 5C3 accepts the revised curves developed by Structural Integrity for 24 month fuel cycles, which removed the discussion on the instrument nozzles.

Page: 6 of 8 NEDC: 07-048 Rev. Number. 5C3 Nebraska Public Power District DESIGN CALCULATIONS SHEET PURPOSE:

This calculation reviews and accepts Structural Integrity Associates (SIA) Calculation 1100445.303, revision 1, which determines the pressure temperature (P-T) curves for the beltline for 32 effective full power years (EFPY). The calculation includes the 2% power uprate, which was implemented after RE24. The calculation also addresses 24 month fuel cycles that are scheduled to be implemented after the 2012 refueling outage. The calculation also provides P-T curves for 54 EFPY.

ASSUMPTIONS:

ART and RTNDT Determination The calculation includes the 2% power uprate that was implemented after RE24. The adjusted reference temperature (ART) values were determined in accordance with Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (RGI.99). The limiting beltrine material has an ART value of 105.8°F and 131.2°F for 32 and 54 EFPY, respectively. Non-beltline regions are not subjected to fluence in excess of 1 0E+1 7 n/cm 2 ; therefore, the RTNOT values are valid substitutions for corresponding ART values. These values are considered to be the most appropriate for computation of updated P-T curves.

Vessel Dimensions and Fluid Properties The inner radius of the cylindrical portion of the reactor pressure vessel (RPV) is 110.375 inches. The vessel shell base metal varies in thickness from 5.375 to 6.375 inches, with 5.375 the predominant base metal thickness in the beltline region. Therefore, 5.375 inches is the smallest thickness (i.e.,

bounding input for P-T calculations).

The normal water level in the RPV is 551.75 inches. The maximum potential water level in the RPV during the pressure test is 825.1875 inches.

The wall thickness in the lower head varies from 3.1875 inches, where it connects to the lower shell, to 6.8125 inches. The thicker section provides the reinforcement for the CRD penetrations.

The Feedwater nozzle penetration in the upper vessel is considered the limiting location for stress in this region. The nozzle inner diameter is 11.969 inches with an outside diameter of 23.0625 inches. The nozzle comer radius, both inside and outside, is 5Y in. and the thickness at the comer is also 5/ in.

The Heatup and Cooldown temperature rate-of-change assumed for the vessel metal for normal operation is 100°F/hr.

Pressure and Instrument Uncertainty The pre-service system hydrostatic test pressure was 1,563 psig, which corresponds to 1.25 times the design pressure of 1,250 psig. This is consistent with previous calculations. The following are the results of the instrument uncertainty for CNS pressure and temperature measurements:

Reactor Vessel Metal Temperature is bounded by +/- 5 0F Reactor Vessel Pressure is bounded by +1- 2%

Page: 7 of 8 NEDC: 07-048 Rev. Number: 5C3 The pressure uncertainty is conservatively evaluated for 1,250 psig, bounding all normal operating pressures and corresponds to 25 psig.

METHODOLOGY:

The limits in the SIA calculation were derived from the NRC-approved methods listed in TS 3.4.9, using the specific revisions listed below:

1) The neutron fluence was calculated using the RAMA computer code (See Design Input 2).

2) Instrument uncertainties have been considered in the referenced calculations.

3) The pressure and temperature limits were calculated consistent with ASME Section XI, Appendix G.

a ART (Adjusted Reference Temperature) is defined in 10CFR50 Appendix G as the temperature shift in the Charpy curve for the irradiated material relative to that for the unirradiated material measured at the 30-foot-pound energy level.

• The feedwater nozzles at CNS are not subjected to fluences above 1.OOE+17 n/cm 2. The feedwater nozzles are well above the highest elevation (370.13 inches) that is projected to exceed 1.00E+17 n/cm 2 fluence for 54 EFPY.

2

" Similarly, since the lower head region is not subjected to fluences above 1.OOE+17 n/cm '

stresses in the lower head penetration ligaments do not have to be evaluated. However, it is included to show that the beltline is bounding. The lower head is below the lowest elevation (197.07 inches) that is projected to exceed 1.OOE+17 n/cm 2 fluence for 54 EFPY. The limiting RTNDT of the lower head plate materials is 28 0 F, which is significantly less than the limiting ART of the beltline materials (Tables 1 and 2).

" USE is the standard industry parameter used to indicate the maximum toughness of a material at high temperature. 10CFR50 Appendix G requires the predicted end-of-life Charpy Impact test USE for RPV materials to be at least 50 ft-lb (absorbed energy), unless an approved analysis supports a lower value. The predicted USE drop is determined in accordance with NRC Regulatory Guide 1.99, Revision 2 (RGI.99). For Boiling Water Reactors (BWRs) that cannot meet the 50 ft-lb criterion, the BWR Vessel and Internals Project (BWRVIP) has provided a bounding Equivalent Margins USE Analysis (EMA) for plants in BWRVIP-74-A, which is valid for up to 54 EFPY of operation.

The calculation methodology is described in Section 4.0 of the SIA calculation. The calculation also addressed the Standby Liquid Control Nozzle in the lower head. The Standby Liquid Control Nozzle is bounded by the lower head CRD penetrations and does not affect the curves.

The following additional minimum temperature requirements apply to the Upper Vessel Region, per Table I of 10CFR50, Appendix G, to accommodate flange region temperature limits:

" If the pressure is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than the RTNDT of the limiting flange material + 90 0F.

" If the pressure is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature must be greater than or equal to the RTNoT of the limiting flange material, 20°F.

Page: 8 of 8 NEDC: 07-048 Rev. Number. 5C3 CONCLUSION:

These calculations of the P-T curves to 32 EFPY and 54 EFPY are consistent with the design basis of the plant. The P-T curves were developed in accordance with the methods in the 2001 Edition, 2003 Addenda of ASME Section XI. The ART and USE were calculated in accordance with RG 1.99. The curves generated from these methods ensure that the P-T limits will not be exceeded during any phase of reactor operation. These curves are suitable for inclusion in the Technical Specifications, USAR and other affected documents following NRC approval.

REFERENCES:

1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

2. BWRVIP-135 Revision 2: BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations"

3. Technical Specification Section 1.1, Definition of Shutdown Margin ATTACHMENTS:

Structural Integrity Associates Calculation 1100445.303, revision 1.