HNP-07-104, Responses to Requests for Additional Information - License Renewal Application Section 4.2 and Subsection B.2.17

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Responses to Requests for Additional Information - License Renewal Application Section 4.2 and Subsection B.2.17
ML072350080
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 08/16/2007
From: Natale T
Progress Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-06-136, HNP-07-104
Download: ML072350080 (13)
v • d • e


Text

AUG 1 6 2007 SERIAL: HNP-07-104 10 CFR 54 U. S. Nuclear Regulatory Commission ATTENTION:

Document Control Desk Washington, DC 20555

Subject:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400 / LICENSE NO. NPF-63 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION

-LICENSE RENEWAL APPLICATION SECTION 4.2 AND SUBSECTION B.2.17

References:

1. Letter from Cornelius J. Gannon to the U. S. Nuclear Regulatory Commission (Serial: HNP-06-136), "Application for Renewal of Operating License," dated November 14, 2006 2. Letter from Maurice Heath (NRC) to Robert J. Duncan 11, "Requests for Additional Information for the Review of the Shearon Harris Nuclear Power Plant, Unit 1, License Renewal Application," dated July 20, 2007 Ladies and Gentlemen:

On November 14, 2006, Carolina Power & Light Company, now doing business as Progress Energy Carolinas, requested the renewal of the operating license for the Shearon Harris Nuclear Power Plant, Unit No. 1, also known as the Harris Nuclear Plant (HNP), to extend the tenn of its operating license an additional 20 years beyond the current expiration date.By letter dated July 20, 2007, the Nuclear Regulatory Commission (NRC) provided requests for additional information (RAls) concerning the HNP License Renewal Application (LRA).The enclosure to this letter provides responses to the RAls. The response to each of the RAls indicates that a change to the LRA is required, and the response to RAI-B.2. 17 involves a modification to existing License Renewal Commitment

  1. 13 described in Enclosure 1 of Reference
1. A transmittal to document these changes will be provided at a later date. This document contains no new Regulatory Commitments.

Any actions discussed in this letter should be considered intended or planned actions that are included for information.

Progress Energy Carolinas, Inc.Harris Nuclear Plant P. 0. Box 165 New Hill, NC 27562 4Ic c2 Document Control Desk J-NP-07-104

/ Page 2 Please refer any questions regarding this submittal to Mr. Roger Stewart, Supervisor

-License Renewal, at (843) 857-5375.1 declare, under penalty of perjury, that the foregoing is true and correct (Executed on AUG 1 6 2007 )Sincerely, TJ. tale Manager -Support Services Harris Nuclear Plant TJN/mhf

Enclosure:

Responses to Requests for Additional Infonination dated July 20, 2007 cc: Mr. P. B. O'Bryan (NRC Senior Resident Inspector, I-NP)Ms. B. 0. Hall (Section Chief, N.C. DENR)Mr. M. L. Heath (NRC License Renewal Project Manager, HNP)Dr. W. D. Travers (NRC Regional Administrator, Region 11)Ms. M. G. Vaaler (NRC Project Manager, I-NP)

H-NP-07-104 Enclosure Page 1 of I11 Responses to Requests for Additional Information dated July 20, 2007 Background On November 14, 2006, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., requested the renewal of the operating license for the Shearon Harris Nuclear Power Plant, Unit No. 1, also known as the Harris Nuclear Plant (HNP), to extend the term of its operating license an additional 20 years beyond the current expiration date.By letter dated July 20, 2007, the Nuclear Regulatory Commission (NRC) provided a request for additional information (RAI) concerning the HNP License Renewal Application.

This enclosure provides the responses to the NRC RAI. Note that NRC RAI numbers 4.2.1 and 4.2.2 are not used.Table of Contents Page NRC RAI-B3.2.17

........................................................................................

1..NRC RAL-4.2.3 (Editorial Correction)

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4 NRC RAI-4.2.4

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5 NRC RAI-4.2.5

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6 NRC RAI-4.2.6

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6 NRC RAI-B.2.17 (A) The applicant states that the surveillance capsule that is to be withdrawn during the 1 6 t1h refueling outage would have been exposed to a neutron fluence value that is equivalent to the peak reactor pressure vessel (R-PV) fluence at 55 effective full power year (EFPY). Please confirm this statement.

The staff requests that the applicant provide the following inform-ation related to this test: (1) Lead factor of the Surveillance capsule (2) Identification number of the capsule, and (3) Heat number of the surveillance material in the capsule (B) Program element 6, item 2 of aging management program (AMP) B.2. 17 states that the applicant intends to test one surveillance capsule after the 1 6 1h refueling outage. The staff requests that the applicant submit the following information that pertains to the test: (1) The projected refueling outage of withdrawal (2) Projected capsule neutron fluence value at the time of withdrawal (3) Corresponding EFPY for the peak RPV fluence to equal the capsule fluence (4) The identification number of the capsule, and (5) Heat number of the surveillance material in the capsule HNP 104 Enclosure Page 2 of I11 (C) The staff requests that the applicant confirm that the withdrawal schedule of the final two capsules for the extended period of operation is consistent with the requirements, specifically the limitations on lead factor, specified in paragraph 7.6.2 of the American Society of Testing Materials E 185 (ASTM E 185), 'Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." (D) Section 5.3.1.6 of the final safety analysis report states that the applicant intends to use two standby capsules with identifications Y and Z for future tests. However, the "operating experience" Section of AMP B.2.17 indicates that three capsules will remain in the RPV for future tests to manage neutron embrittlement during the extended period of operation.

The staff requests that the applicant provide an explanation for this inconsistency.(E) The staff requests that the applicant include the following statements in the commitment table of the license renewal application (LRA): (1) The applicant will notify the staff if there is any change in the withdrawal schedules of the surveillance capsules.(2) If a standby capsule is removed from the RPV without the intent to test it, the capsule will be stored in manner which maintains it in a condition which would permit its future use, including during the period of extended operation, if necessary.

RAI-B.2.17 Response (A) Confirmation of statement in (A): The withdrawal of the next capsule will occur during Refueling Outage (RFO)-16, at which time the capsule fluence is projected to be equivalent to the 60-year (i.e., 55 EFPY) maximum vessel fluence of 6.8x 10'9 n/cm 2 in accordance with ASTM E 185-82.A(I) The lead factor of the surveillance capsule to be withdrawn during REO- 16 is 2.68.A(2) The identification of this capsule is Capsule W.A(3) The heat number of the surveillance base metal in Capsule W is B4197-2. The surveillance weld wire in Capsule W is 5P677 1, Linde 124 flux, lot number 03 42.(B) Program Element 6, Item 2 of AMP B.2.17 states:... I-NP will evaluate neutron exposure for the remaining capsules, based on the analysis of the capsule withdrawn during RFO- 16. The neutron exposure and withdrawal schedule for the capsules remaining after RFO- 16 will be optimized to provide meaningful metallurgical

  • data.In addition to the response to RAI-B3.2.17 questions B(l) through B(5), plans for the capsules remaining inside the reactor vessel are discussed in the response to RAI-B.2. 17(C).B(1) The projected refueling outage for the withdrawal after RFO- 16 (i.e., after the withdrawal of capsule W) will be determined after the analysis of Capsule W.

I-NP 104 Enclosure Page 3 of 11I B(2) The projected neutron fluence for the next capsule to be withdrawn after RFO- 16 will not exceed twice the 60-year maximum vessel fluence of 6.8x 10 1 9 n/cm 2 in accordance with ASTM E 185-82.B(3) The corresponding EFPY for the peak reactor vessel fluence equal to the capsule fluence for the capsule to be withdrawn after RFO- 16 will be determined after the analysis of Capsule W.B(4) The identification number of the capsule to be withdrawn after RFO-16 is either Capsule Y or Z.B(5) The heat number of the surveillance base metal in Capsules Y and Z is B4197-2. The surveillance weld wire in Capsules Y and Z is 5P6771, Linde 124 flux, lot number 0342.(C) HINP is currently in Cycle 14. At the current time, Capsules W, Y, and Z remain inside the reactor vessel. Capsules W, Y, and Z are projected to exceed the 60-year (i.e., 55 EFPY)maximum vessel fluence prior to the end of 40 years. All remaining capsules currently have a lead factor of 2.68, which is used to determine the withdrawal schedule of Capsules W, Y, and Z in accordance with ASTM E 185-82. Capsule W is scheduled to be withdrawn during RFO-16, at which time the capsule fluence is projected to be equivalent to the 55 EFPY maximum vessel fluence of 6.8x 10 19 n/cm 2.Capsules Y and Z will remain in the reactor vessel after Capsule W is withdrawn.

Based on the above, the following changes to Section B.2.17 of the HNP LRA are required.These changes require a modification to HNP LR Commitment

  1. 13) from Enclosure I of the letter from Cornelius J. Gannon to the U. S. Nuclear Regulatory Commission (Serial: I-NP-06-136), "Application for Renewal of Operating License," dated November 14, 2006.Enhancement 1 to Element 6 in the 14NP LIZA will be modified as follows: Withdrawal of the next capsule (i.e., Capsule W) will occur during RFO- 16, at which time the capsule fluence is projected to be equivalent to the 60-year maximum vessel fluence of 6.8xI 10 9 n/cm 2 in accordance with ASTM E 185-82.Enhancement 2 to Element 6 in the I-NP LRA will be modified as follows: The analysis of Capsule W, to be withdrawn during RFO- 16, will be used to evaluate neutron exposure for remaining Capsules Y and Z, as required by 10 CFR 50 Appendix H. The withdrawal schedule for one of the remaining capsules (i.e., Capsule Y or Z) will be adjusted, based on the analysis of Capsule W, so that the capsule fluence will not exceed twice the 60-year maximum vessel fluence in accordance with ASTM E 185-82. The neutron exposure and withdrawal schedule for the last capsule will be optimized to provide meaningful metallurgical data. If the last capsule is projected to significantly exceed a meaningful fluence value, it will either be relocated to a lower flux position or withdrawn for possible testing or re-insertion.

The remaining Capsules Y and Z (and archived test specimens available for reconstitution) will be available for the monitoring of neutron exposure if additional license renewals are sought (i.e., 80 years of operation).

1HNP 104 Enclosure Page 4 of I11 (D) HNP FSAR Section 5.3.1.6 states that Capsules U, V, and X have been withdrawn from the reactor vessel and tested, and that Capsules W, Y, and Z remain inside the vessel. FSAR Section 5.3.1.6 also states that Capsule W is scheduled for removal from the vessel, and that Capsules Y and Z are standby capsules.

Therefore, FSAR Section 5.3.1.6 is stating that three capsules (i.e., Capsules W, Y, and Z) are currently in the reactor vessel. Section B.2.17 of the 1-INP LRA states that three capsules remain inside the vessel, exposed to additional neutron flux, providing a source for future data that will be used to manage neutron embrittlement aging effects for the period of extended operation.

Therefore the statements in FSAR Section 5.3.1.6 and HNP LRA Section B.2.17 are consistent.(E) Request for further commitments: (E)( 1) The HNP procedure entitled "Technical Specification Equipment List Program and Core Operating Limits Report," Attachment 3, states: Changes to the reactor materials surveillance schedule must receive NRC approval prior to implementation. (

Reference:

Section III.B.3 of 10 CFR 50, Appendix H).Therefore, an additional commitment in the 1-NP LRA is not needed.(E)(2) Both Commitment

  1. 1 3 , Item 1, from Enclosure I of the letter from Cornelius J.Gannon to the U. S. Nuclear Regulatory Commission (Serial: I-NP-06-136),"Application for Renewal of Operating License," dated November 14, 2006, and HNP LRA Section B.2. 17, Enhancements, indicate that the Reactor Vessel Surveillance Program will be enhanced such that tested and untested specimens from all capsules pulled from the reactor vessel must be kept in storage to permit future reconstitution use and that the identity, traceability, and recovery of the capsule specimens shall be maintained throughout testing and storage. Therefore, an additional commitment in the H-NP LRA is not needed.An amendment to the LRA is required in response to RAI B.2.17 (C).NRC RAI-4.2.3 (Editorial Correction)

In Tables 4:2-2 and 4.2-3 of the LRA, the chemical composition values of elements Copper and Nickel for the surveillance capsule test sample representing the intermediate shell plate (heat number-B34197-2) and the RPV's intermediate shell plate (heat number-B34197-2) are identical.

However, the chemistry factors are different.

The staff requests that the applicant add a footnote stating that the chemistry factor for the surveillance capsule test sample representing the intermediate shell plate is derived from the surveillance test data.

HNP 104 Enclosure Page 5 of 11I RAI-4.2.3 Response A footnote will be added to Tables 4.2-2 and 4.2-3 in the LRA, stating that the chemistry factors for the surveillance test capsule representing the Intermediate Shell Plate (heat number B4197-2)and the Intermediate Shell-to-Lower Shell Circumferential Weld (100%) (heat number 5P677 1)are derived from the surveillance data.An amendment to the LRA is required in response to RAL-4.2.3.

NRC RAI-4.2.4 The staff requests that the applicant include the following items in Section 4.2.4 of the LRA: (A) The current pressure-temperature (P-T) limits are valid through 36 EFPY. The P-T limits for the extended period of operation will be managed by using approved fluence calculations when there are changes in power of core design in conjunction with surveillance capsule results.(B) Any change in P-T curves will be implemented by the license amendment process (i.e., modifications of technical specifications) and will meet the requirements of Title 10 of the Code of Federal Regulations Section 50.60 (10 CFR 50.60) and 10 CFR Part 50, Appendix G.RAI-4.2.4 Response (A) The following will be added to the HNP LRA, Section 4.2.4, at the end of the"Analysis" Subsection:

The current P-T limits are valid through 36 EFPY. The P-T limits for the extended period of operation will be managed by using approved fluence calculations when there are changes in power or core design in conjunction with surveillance capsule results.(B) The following will be added to the HNP LRA, Section 4.2.4 at the end of the"Analysis" Subsection:

P-T limits have been imposed on operational parameters at I-NP, thereby assuring that the reactor vessel is operated within required safety margins in accordance with the requirements of 10 CFR 5 0.60 and 10 CFR 5 0, Appendix G. HNP has implemented changes in the P-T curves throughout the current period of operation using the license amendment process, and expects to continue to use the license amendment process to implement future changes in P-T curves for the remainder of the current period of operation and for the extended period of operation.

HNP 104 Enclosure Page 6 of 11I An amendment to the LRA is required in response to RAI-4.2.4.

NRC RAI-4.2.5 Since the P-T limits for the extended period of operation are not yet developed, the applicant should make a statement in the LRA that they will submit the appropriate analysis for the low temperature overpressure (LTOP) setpoints that will be valid for the license renewal period. Any change in the LTOP setpoints will be implemented by the license amendment process (i.e., modifications of technical specifications) and will meet the requirements of 10 CFR 50.60 and 10 CFR Part 50, Appendix G.RAI-4.2.5 Response The following will be added to Section 4.2.5 of the I-NP LRA: HNP will submit the appropriate analysis for LTOP setpoints that will be valid for the period of extended operation.

LTOP setpoints have been imposed on operational parameters at I-NP, thereby assuring that the reactor vessel is operated within required safety margins in accordance with the requirements of 10 CFR 50.60 and 10 CFR 50, Appendix G. 1HNP has implemented changes in the LTOP setpoints throughout the current period of operation using the license amendment process, and expects to continue to use the license amendment process to implement future changes in LTOP setpoints for the remainder of the current period of operation and for the extended period of operation.

An amendment to the LRA is required in response to RAI-4.2.5.

NRC RAI-4.2.6 During the audit at the Harris Nuclear Plant, the staff was informed by the applicant that one reactor vessel nozzle was projected to achieve a neutron fluence greater than IlX 1017 n/cm 2 (E > 1 MeV) at the end of the extended period of operation.

This nozzle material was not listed in Tables 4.2-1, 4.2-2 and 4.2-3 of the LRA. According to Table IV A-2 of NUREG- 180 1, Revision 1, ferritic materials are subject to neutron embrittlement when they are exposed to neutron fluence greater than I X 1017 n/cm 2 (E > 1 MeV) at the end of the extended period of operation.

Therefore, the staff requests that the applicant provide the following for this nozzle material and its associated welds: (1) The RTPTS value of the nozzle material and its associated welds per the requirements of Title 10 of the Code of Federal Regulations (CFR) Section 50.6 1.(2) The adjusted reference temperature value of the nozzle material and its associated welds that will be used for developing pressure-temperature limits per the requirements of 10 CFR Part 50, Appendix G.

FINP 104 Enclosure Page 7 of 11 (3) The upper shelf energy value of the nozzle material and its associated welds per the requirements of 10 CFR Part 50, Appendix G.RAI-4.2.6 Response The reactor vessel materials outside the traditional beltline region that are exposed to a 55 BEPY fluence greater than 1017 n/cm 2 (E > 1.0 MeV) were evaluated to determine if these materials should be considered beltline materials for the period of extended operation.

The beitline is defined in 10 CFR 50.61 (a)(3) as the region of the reactor vessel that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection for the most limiting material.The evaluation found five reactor vessel materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 that were not previously analyzed for irradiation damage. The materials were: I1) Upper to Intermediate Circumferential Weld AC (Heat 4P4784, Linde 124), 2) Upper Shell (conservatively C0123-1), 3) Inlet Nozzle Weld 15-A, 15-B3, 15-C (Heat 3P4966, Linde 124), 4) Inlet Nozzle (conservatively 438B3-5).

and 5) Upper Shell Longitudinal Welds BE/BF (Heat 4P4784, Linde 124).The reactor vessel materials below the traditional beltline region did not include any additional materials that required analysis for irradiation damage, in accordance with 10 CFR 50.6 1.Table 4.2-4 summarizes the decrease in Charpy upper shelf energy (CvUSE) for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 10" 7 n/cm 2 (EB> 1.0 MeV). These values were calculated per the requirements of 10 CFR 50, Appendix G. Based on the CvUSE analysis, the material locations above the traditional beltline region are not limiting since they are projected to maintain Charpy upper shelf energies greater than that of the intermediate shell plate B4197-2, which is located inside the traditional beltline region. CVUSE for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 52.8 fl-lbs at Y/4-thickness.

Table 4.2-5 surmmarizes the reactor vessel pressurized thermal shock (PTS) reference temperature (RTpTs) for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 (E > 1.0 MeV). These values were calculated per the requirements of 10 CFR 50.61. Based on the RTPT5 analysis, none of the material locations above the traditional beltline region are limiting, since they are projected to maintain RTPTS values less than that of the intermediate shell plate, heat number B4 197-2, which is located inside the traditional beitline region. RTPi-S for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 199.9'F.

I-INP-07-104 Enclosure Page 8 of 11I Table 4.2-6 summarizes the reactor vessel adjusted reference temperatures (ART) at the '/4-thickness and %/-thickness wall locations for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 (EB> 1.0 MeV). These values were calculated per the requirements of 10 CFR 50, Appendix G. Based on the ART analysis, none of the material locations above the traditional beltline region are limiting, since they are projected to maintain ART values less than that of the intermediate shell plate, heat number B4 197-2, which is located inside the traditional beltline region. ART for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 195.3'F at 1/4/-thickness and 183.6'F at 3/4-thickness.

An amendment to the LRA is required in response to RAI-4.2.6.

J-INP 104 Enclosure Page 9 of 11I TABLE 4.2-4 UPPER SHELF ENERGY (Cv~USE) EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 101 7 n/cm 2 (E > 1.0 MeV)Material Description CuT (lunc ) Initial PrPredicted CVUSE_________________________________

__________

Per_______

R.G. 1.99, Revision 2 Reactor Vessel Heat Twt% (x 101 C\/USE CVUSE%Beitline Region Location Number j Type n/cm2 ft-lbs ft-lbs I Decrease Upper to Intermediate Circumferential 4P4784 ASA/Linde 124 0.06 0.2073 95 (4) 8191 13.8 Weld (AC)Upper Shell C0123- 11 2) SA-533 Gr. B 0.12 0.2073 84 71.8 14.5 Inlet Nozzle Weld (1 5-A, 15-B3, 15-C) 3P4966 ASA/Linde 124 0.02 0.0113 63(4 59.5(5)5.Inlet Nozzle 43813-5 (3) SA-508 Cl. 2 0.35 0.0113 128 108.5 15.2 Upper Shell Longitudinal WeldS (BE! BF) 4P4784 ASA/Linde 124 0.06 0.2073 95 (4) 811 13.8 I. Calculated based on guidelines in RG 1.99, Revision 2. The 55 EFPY inside surface fluence is the calculated value at the 'wetted' surface of the reactor vessel. The 'AT location fluence value is determined by calculating the IAT depth into the vessel and adding the minimum cladding thickness.

2. Upper Shell Plate CO0123-1 ex-hibited a higher value for initial RTNDT than the other Upper Shell Plate C0224- 1. Therefore, Upper Shell Plate CO 123-1 was chosen as the more conservative plate for the purpose of the embrittlement evaluation.
3. Inlet Nozzle 438B3-5 exhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 438B3-5 was chosen as the most conservative inlet nozzle for the purpose of the embrittlement evaluation.

4. As defined by ASTM E 185-82, these Charpy data are from the transition region and not the upper shelf region of the Charpy curve, since the specimen fracture surfaces exhibit less than 95% shear. Therefore, these Charpy data do not represent upper shelf energy levels and are considered conservative.
5. Predicted value is conservative, since initial Charpy data are from the transition region and not the upper shelf region of the Charpy curve.

I-INP 104 Enclosure Page 10 oflII TABLE 4.2-5 PTS REFERENCE TEMPERATURE EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 1iO'7 n/cm 2 (E > 1.0 MeV)Reactor Vessel Beltline Region Location RTPTS Calculation Per 10 CFR 50.61 Using Tables Upper to Intermediate A4P74 ASA/Linde0.6 09 2 8.0.3 1 073 5.6 55 131 30 Circumferential Weld A4P74 124 0.6 09 -2 820.41 0.3 576 55 131 30 Upper Shell C0I23-I1(2 1 C0123- 1(2) A-3 0.12 0.60 42 83.0 0.3401 0.703 58.4 34.0 134.4 270 Gr. B__Inlet Nozzle Weld 1A,5- 3P4966 A /Lne0. 02 0.92 -56 27.0 0.0203 0.173 4.7 34.3 -17.0 270 B, 15-C 124 __Inlet Nozzle 43813-5")

43813-5(3) A50 0.35 0.89 0 255.0 0.0203 0.173 44.1 34.0 78.1 270 UprSelBE/BF1 4P4784 0SA06i0.91

-20 82.0 0.3401 0.703 57.6 65.5 103.1 270 Longitudinal Welds 124 I .RTPTS is normally calculated using the fluence at the clad/base metal interface in accordance with 10 CFR 50.6 1. However, HNP calculated RTPTS using the 55 EFPY inside wetted surface fluence, which is higher than the 55 EFPY fluence at the clad/base metal interface.

2. Upper Shell Plate C0123-1 exhibited a higher value for initial RTNDT than the other Upper Shell Plate C0224-1. Therefore, Upper Shell Plate C0123-1 was chosen as the more conservative plate for the purpose of the embrittlemnent evaluation.
3. Inlet Nozzle 438B3-5 e~xhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 438B3-5 was chosen as the most conservative inlet nozzle for the purpose of the embrittlernent evaluation.

HINP-07-104 Enclosure Page I11 of I11 TABLE 4.2-6 ADJUSTED REFERENCE TEMPERATURE EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 107 nlecM (E > 1.0 MeV)Matril DsciptonChemical 55 EFPY Fluence ARTNDT, OF MagnART, OF Reactor Vessel 1Cornposition Initial Chemit. 1019 n/cm at 55 EFPY Iat 55 EFPY Beitline Re-ion Material Heat Type Cu Ni RTNDT Factor Inside A/TtI) 3/AT t) 1/4 T 1/4T 1/4T %/4T 1/4T 1/4T C I D Number wt% wt% surface Location Location Location Location Location Location Location Location Location____________

RG 1.99, Revision 2, Position I. 1 * .4 %, Upper to AA Inemdae AC 4P4784 Linde 0.06 0.91 -20 82.0 0.3401 0.2073 0.0818 47.4 31.0 58.3 46.0 85.7 57.0 Circumferential 124 Weld Upper Shell C0123- 1(2 1C0]23-l1 t 2) SA-533 0.12 0.60 42 83.0 0.3401 0.2073 0.0818 48.0 31.4 34.0 31.4 124.0 104.8 Gr. B Inlet Nozzle 15-A 15- ASA/Wed1,1ýC 3P4966 Linde 0.02 0.92 -56 27.0 0.0203 0.0113 0.00372 3.2 1.5 34.2 34.0 -18.6 -20.5 Weld B 15-C124 Inlet Nozzle 43813-5 (3) 43813-5 (3) SA-508 0.35 0.89 0 255.0 0.0203 0.0113 0.00372 30.3 13.8 30.3 13.8 60.6 27.6 Cl. 2 Upper Shell ASA/Longitudinal BE/B3F 4P4784 Linde 0.06 0.91 -20 82.0 0.3401 0.2073 0.0818 47.4 31.0 58.3 46.0 85.7 57.0 Welds ______ __ 124 I .Calculated based on guidelines in RG 1.99, Revision 2. The 55 EFPY inside surface fluence is the calculated value at the "wetted' surface of the reactor vessel. The 1/4/T and %/T location fluence values are determined by calculating the '/4T and 1/T depth into the vessel and adding the minimum cladding thickness.

2. Upper Shell Plate C0123-1 exhibited a higher value for- initial RTNDT than the other Upper Shell Plate C0224-1. Therefore, Upper Shell Plate CO0123-I was chosen as the more conservative plate for the purpose of the embrittlement evaluation.
3. Inlet Nozzle 438B3-5 exhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 43813-5 wvas chosen as the most conservative inlet nozzle for the purpose of the ernbrittlemnent evaluation.

Text

AUG 1 6 2007 SERIAL: HNP-07-104 10 CFR 54 U. S. Nuclear Regulatory Commission ATTENTION:

Document Control Desk Washington, DC 20555

Subject:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400 / LICENSE NO. NPF-63 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION

-LICENSE RENEWAL APPLICATION SECTION 4.2 AND SUBSECTION B.2.17

References:

1. Letter from Cornelius J. Gannon to the U. S. Nuclear Regulatory Commission (Serial: HNP-06-136), "Application for Renewal of Operating License," dated November 14, 2006 2. Letter from Maurice Heath (NRC) to Robert J. Duncan 11, "Requests for Additional Information for the Review of the Shearon Harris Nuclear Power Plant, Unit 1, License Renewal Application," dated July 20, 2007 Ladies and Gentlemen:

On November 14, 2006, Carolina Power & Light Company, now doing business as Progress Energy Carolinas, requested the renewal of the operating license for the Shearon Harris Nuclear Power Plant, Unit No. 1, also known as the Harris Nuclear Plant (HNP), to extend the tenn of its operating license an additional 20 years beyond the current expiration date.By letter dated July 20, 2007, the Nuclear Regulatory Commission (NRC) provided requests for additional information (RAls) concerning the HNP License Renewal Application (LRA).The enclosure to this letter provides responses to the RAls. The response to each of the RAls indicates that a change to the LRA is required, and the response to RAI-B.2. 17 involves a modification to existing License Renewal Commitment

  1. 13 described in Enclosure 1 of Reference
1. A transmittal to document these changes will be provided at a later date. This document contains no new Regulatory Commitments.

Any actions discussed in this letter should be considered intended or planned actions that are included for information.

Progress Energy Carolinas, Inc.Harris Nuclear Plant P. 0. Box 165 New Hill, NC 27562 4Ic c2 Document Control Desk J-NP-07-104

/ Page 2 Please refer any questions regarding this submittal to Mr. Roger Stewart, Supervisor

-License Renewal, at (843) 857-5375.1 declare, under penalty of perjury, that the foregoing is true and correct (Executed on AUG 1 6 2007 )Sincerely, TJ. tale Manager -Support Services Harris Nuclear Plant TJN/mhf

Enclosure:

Responses to Requests for Additional Infonination dated July 20, 2007 cc: Mr. P. B. O'Bryan (NRC Senior Resident Inspector, I-NP)Ms. B. 0. Hall (Section Chief, N.C. DENR)Mr. M. L. Heath (NRC License Renewal Project Manager, HNP)Dr. W. D. Travers (NRC Regional Administrator, Region 11)Ms. M. G. Vaaler (NRC Project Manager, I-NP)

H-NP-07-104 Enclosure Page 1 of I11 Responses to Requests for Additional Information dated July 20, 2007 Background On November 14, 2006, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., requested the renewal of the operating license for the Shearon Harris Nuclear Power Plant, Unit No. 1, also known as the Harris Nuclear Plant (HNP), to extend the term of its operating license an additional 20 years beyond the current expiration date.By letter dated July 20, 2007, the Nuclear Regulatory Commission (NRC) provided a request for additional information (RAI) concerning the HNP License Renewal Application.

This enclosure provides the responses to the NRC RAI. Note that NRC RAI numbers 4.2.1 and 4.2.2 are not used.Table of Contents Page NRC RAI-B3.2.17

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1..NRC RAL-4.2.3 (Editorial Correction)

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4 NRC RAI-4.2.4

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5 NRC RAI-4.2.5

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6 NRC RAI-4.2.6

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6 NRC RAI-B.2.17 (A) The applicant states that the surveillance capsule that is to be withdrawn during the 1 6 t1h refueling outage would have been exposed to a neutron fluence value that is equivalent to the peak reactor pressure vessel (R-PV) fluence at 55 effective full power year (EFPY). Please confirm this statement.

The staff requests that the applicant provide the following inform-ation related to this test: (1) Lead factor of the Surveillance capsule (2) Identification number of the capsule, and (3) Heat number of the surveillance material in the capsule (B) Program element 6, item 2 of aging management program (AMP) B.2. 17 states that the applicant intends to test one surveillance capsule after the 1 6 1h refueling outage. The staff requests that the applicant submit the following information that pertains to the test: (1) The projected refueling outage of withdrawal (2) Projected capsule neutron fluence value at the time of withdrawal (3) Corresponding EFPY for the peak RPV fluence to equal the capsule fluence (4) The identification number of the capsule, and (5) Heat number of the surveillance material in the capsule HNP 104 Enclosure Page 2 of I11 (C) The staff requests that the applicant confirm that the withdrawal schedule of the final two capsules for the extended period of operation is consistent with the requirements, specifically the limitations on lead factor, specified in paragraph 7.6.2 of the American Society of Testing Materials E 185 (ASTM E 185), 'Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels." (D) Section 5.3.1.6 of the final safety analysis report states that the applicant intends to use two standby capsules with identifications Y and Z for future tests. However, the "operating experience" Section of AMP B.2.17 indicates that three capsules will remain in the RPV for future tests to manage neutron embrittlement during the extended period of operation.

The staff requests that the applicant provide an explanation for this inconsistency.(E) The staff requests that the applicant include the following statements in the commitment table of the license renewal application (LRA): (1) The applicant will notify the staff if there is any change in the withdrawal schedules of the surveillance capsules.(2) If a standby capsule is removed from the RPV without the intent to test it, the capsule will be stored in manner which maintains it in a condition which would permit its future use, including during the period of extended operation, if necessary.

RAI-B.2.17 Response (A) Confirmation of statement in (A): The withdrawal of the next capsule will occur during Refueling Outage (RFO)-16, at which time the capsule fluence is projected to be equivalent to the 60-year (i.e., 55 EFPY) maximum vessel fluence of 6.8x 10'9 n/cm 2 in accordance with ASTM E 185-82.A(I) The lead factor of the surveillance capsule to be withdrawn during REO- 16 is 2.68.A(2) The identification of this capsule is Capsule W.A(3) The heat number of the surveillance base metal in Capsule W is B4197-2. The surveillance weld wire in Capsule W is 5P677 1, Linde 124 flux, lot number 03 42.(B) Program Element 6, Item 2 of AMP B.2.17 states:... I-NP will evaluate neutron exposure for the remaining capsules, based on the analysis of the capsule withdrawn during RFO- 16. The neutron exposure and withdrawal schedule for the capsules remaining after RFO- 16 will be optimized to provide meaningful metallurgical

  • data.In addition to the response to RAI-B3.2.17 questions B(l) through B(5), plans for the capsules remaining inside the reactor vessel are discussed in the response to RAI-B.2. 17(C).B(1) The projected refueling outage for the withdrawal after RFO- 16 (i.e., after the withdrawal of capsule W) will be determined after the analysis of Capsule W.

I-NP 104 Enclosure Page 3 of 11I B(2) The projected neutron fluence for the next capsule to be withdrawn after RFO- 16 will not exceed twice the 60-year maximum vessel fluence of 6.8x 10 1 9 n/cm 2 in accordance with ASTM E 185-82.B(3) The corresponding EFPY for the peak reactor vessel fluence equal to the capsule fluence for the capsule to be withdrawn after RFO- 16 will be determined after the analysis of Capsule W.B(4) The identification number of the capsule to be withdrawn after RFO-16 is either Capsule Y or Z.B(5) The heat number of the surveillance base metal in Capsules Y and Z is B4197-2. The surveillance weld wire in Capsules Y and Z is 5P6771, Linde 124 flux, lot number 0342.(C) HINP is currently in Cycle 14. At the current time, Capsules W, Y, and Z remain inside the reactor vessel. Capsules W, Y, and Z are projected to exceed the 60-year (i.e., 55 EFPY)maximum vessel fluence prior to the end of 40 years. All remaining capsules currently have a lead factor of 2.68, which is used to determine the withdrawal schedule of Capsules W, Y, and Z in accordance with ASTM E 185-82. Capsule W is scheduled to be withdrawn during RFO-16, at which time the capsule fluence is projected to be equivalent to the 55 EFPY maximum vessel fluence of 6.8x 10 19 n/cm 2.Capsules Y and Z will remain in the reactor vessel after Capsule W is withdrawn.

Based on the above, the following changes to Section B.2.17 of the HNP LRA are required.These changes require a modification to HNP LR Commitment

  1. 13) from Enclosure I of the letter from Cornelius J. Gannon to the U. S. Nuclear Regulatory Commission (Serial: I-NP-06-136), "Application for Renewal of Operating License," dated November 14, 2006.Enhancement 1 to Element 6 in the 14NP LIZA will be modified as follows: Withdrawal of the next capsule (i.e., Capsule W) will occur during RFO- 16, at which time the capsule fluence is projected to be equivalent to the 60-year maximum vessel fluence of 6.8xI 10 9 n/cm 2 in accordance with ASTM E 185-82.Enhancement 2 to Element 6 in the I-NP LRA will be modified as follows: The analysis of Capsule W, to be withdrawn during RFO- 16, will be used to evaluate neutron exposure for remaining Capsules Y and Z, as required by 10 CFR 50 Appendix H. The withdrawal schedule for one of the remaining capsules (i.e., Capsule Y or Z) will be adjusted, based on the analysis of Capsule W, so that the capsule fluence will not exceed twice the 60-year maximum vessel fluence in accordance with ASTM E 185-82. The neutron exposure and withdrawal schedule for the last capsule will be optimized to provide meaningful metallurgical data. If the last capsule is projected to significantly exceed a meaningful fluence value, it will either be relocated to a lower flux position or withdrawn for possible testing or re-insertion.

The remaining Capsules Y and Z (and archived test specimens available for reconstitution) will be available for the monitoring of neutron exposure if additional license renewals are sought (i.e., 80 years of operation).

1HNP 104 Enclosure Page 4 of I11 (D) HNP FSAR Section 5.3.1.6 states that Capsules U, V, and X have been withdrawn from the reactor vessel and tested, and that Capsules W, Y, and Z remain inside the vessel. FSAR Section 5.3.1.6 also states that Capsule W is scheduled for removal from the vessel, and that Capsules Y and Z are standby capsules.

Therefore, FSAR Section 5.3.1.6 is stating that three capsules (i.e., Capsules W, Y, and Z) are currently in the reactor vessel. Section B.2.17 of the 1-INP LRA states that three capsules remain inside the vessel, exposed to additional neutron flux, providing a source for future data that will be used to manage neutron embrittlement aging effects for the period of extended operation.

Therefore the statements in FSAR Section 5.3.1.6 and HNP LRA Section B.2.17 are consistent.(E) Request for further commitments: (E)( 1) The HNP procedure entitled "Technical Specification Equipment List Program and Core Operating Limits Report," Attachment 3, states: Changes to the reactor materials surveillance schedule must receive NRC approval prior to implementation. (

Reference:

Section III.B.3 of 10 CFR 50, Appendix H).Therefore, an additional commitment in the 1-NP LRA is not needed.(E)(2) Both Commitment

  1. 1 3 , Item 1, from Enclosure I of the letter from Cornelius J.Gannon to the U. S. Nuclear Regulatory Commission (Serial: I-NP-06-136),"Application for Renewal of Operating License," dated November 14, 2006, and HNP LRA Section B.2. 17, Enhancements, indicate that the Reactor Vessel Surveillance Program will be enhanced such that tested and untested specimens from all capsules pulled from the reactor vessel must be kept in storage to permit future reconstitution use and that the identity, traceability, and recovery of the capsule specimens shall be maintained throughout testing and storage. Therefore, an additional commitment in the H-NP LRA is not needed.An amendment to the LRA is required in response to RAI B.2.17 (C).NRC RAI-4.2.3 (Editorial Correction)

In Tables 4:2-2 and 4.2-3 of the LRA, the chemical composition values of elements Copper and Nickel for the surveillance capsule test sample representing the intermediate shell plate (heat number-B34197-2) and the RPV's intermediate shell plate (heat number-B34197-2) are identical.

However, the chemistry factors are different.

The staff requests that the applicant add a footnote stating that the chemistry factor for the surveillance capsule test sample representing the intermediate shell plate is derived from the surveillance test data.

HNP 104 Enclosure Page 5 of 11I RAI-4.2.3 Response A footnote will be added to Tables 4.2-2 and 4.2-3 in the LRA, stating that the chemistry factors for the surveillance test capsule representing the Intermediate Shell Plate (heat number B4197-2)and the Intermediate Shell-to-Lower Shell Circumferential Weld (100%) (heat number 5P677 1)are derived from the surveillance data.An amendment to the LRA is required in response to RAL-4.2.3.

NRC RAI-4.2.4 The staff requests that the applicant include the following items in Section 4.2.4 of the LRA: (A) The current pressure-temperature (P-T) limits are valid through 36 EFPY. The P-T limits for the extended period of operation will be managed by using approved fluence calculations when there are changes in power of core design in conjunction with surveillance capsule results.(B) Any change in P-T curves will be implemented by the license amendment process (i.e., modifications of technical specifications) and will meet the requirements of Title 10 of the Code of Federal Regulations Section 50.60 (10 CFR 50.60) and 10 CFR Part 50, Appendix G.RAI-4.2.4 Response (A) The following will be added to the HNP LRA, Section 4.2.4, at the end of the"Analysis" Subsection:

The current P-T limits are valid through 36 EFPY. The P-T limits for the extended period of operation will be managed by using approved fluence calculations when there are changes in power or core design in conjunction with surveillance capsule results.(B) The following will be added to the HNP LRA, Section 4.2.4 at the end of the"Analysis" Subsection:

P-T limits have been imposed on operational parameters at I-NP, thereby assuring that the reactor vessel is operated within required safety margins in accordance with the requirements of 10 CFR 5 0.60 and 10 CFR 5 0, Appendix G. HNP has implemented changes in the P-T curves throughout the current period of operation using the license amendment process, and expects to continue to use the license amendment process to implement future changes in P-T curves for the remainder of the current period of operation and for the extended period of operation.

HNP 104 Enclosure Page 6 of 11I An amendment to the LRA is required in response to RAI-4.2.4.

NRC RAI-4.2.5 Since the P-T limits for the extended period of operation are not yet developed, the applicant should make a statement in the LRA that they will submit the appropriate analysis for the low temperature overpressure (LTOP) setpoints that will be valid for the license renewal period. Any change in the LTOP setpoints will be implemented by the license amendment process (i.e., modifications of technical specifications) and will meet the requirements of 10 CFR 50.60 and 10 CFR Part 50, Appendix G.RAI-4.2.5 Response The following will be added to Section 4.2.5 of the I-NP LRA: HNP will submit the appropriate analysis for LTOP setpoints that will be valid for the period of extended operation.

LTOP setpoints have been imposed on operational parameters at I-NP, thereby assuring that the reactor vessel is operated within required safety margins in accordance with the requirements of 10 CFR 50.60 and 10 CFR 50, Appendix G. 1HNP has implemented changes in the LTOP setpoints throughout the current period of operation using the license amendment process, and expects to continue to use the license amendment process to implement future changes in LTOP setpoints for the remainder of the current period of operation and for the extended period of operation.

An amendment to the LRA is required in response to RAI-4.2.5.

NRC RAI-4.2.6 During the audit at the Harris Nuclear Plant, the staff was informed by the applicant that one reactor vessel nozzle was projected to achieve a neutron fluence greater than IlX 1017 n/cm 2 (E > 1 MeV) at the end of the extended period of operation.

This nozzle material was not listed in Tables 4.2-1, 4.2-2 and 4.2-3 of the LRA. According to Table IV A-2 of NUREG- 180 1, Revision 1, ferritic materials are subject to neutron embrittlement when they are exposed to neutron fluence greater than I X 1017 n/cm 2 (E > 1 MeV) at the end of the extended period of operation.

Therefore, the staff requests that the applicant provide the following for this nozzle material and its associated welds: (1) The RTPTS value of the nozzle material and its associated welds per the requirements of Title 10 of the Code of Federal Regulations (CFR) Section 50.6 1.(2) The adjusted reference temperature value of the nozzle material and its associated welds that will be used for developing pressure-temperature limits per the requirements of 10 CFR Part 50, Appendix G.

FINP 104 Enclosure Page 7 of 11 (3) The upper shelf energy value of the nozzle material and its associated welds per the requirements of 10 CFR Part 50, Appendix G.RAI-4.2.6 Response The reactor vessel materials outside the traditional beltline region that are exposed to a 55 BEPY fluence greater than 1017 n/cm 2 (E > 1.0 MeV) were evaluated to determine if these materials should be considered beltline materials for the period of extended operation.

The beitline is defined in 10 CFR 50.61 (a)(3) as the region of the reactor vessel that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection for the most limiting material.The evaluation found five reactor vessel materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 that were not previously analyzed for irradiation damage. The materials were: I1) Upper to Intermediate Circumferential Weld AC (Heat 4P4784, Linde 124), 2) Upper Shell (conservatively C0123-1), 3) Inlet Nozzle Weld 15-A, 15-B3, 15-C (Heat 3P4966, Linde 124), 4) Inlet Nozzle (conservatively 438B3-5).

and 5) Upper Shell Longitudinal Welds BE/BF (Heat 4P4784, Linde 124).The reactor vessel materials below the traditional beltline region did not include any additional materials that required analysis for irradiation damage, in accordance with 10 CFR 50.6 1.Table 4.2-4 summarizes the decrease in Charpy upper shelf energy (CvUSE) for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 10" 7 n/cm 2 (EB> 1.0 MeV). These values were calculated per the requirements of 10 CFR 50, Appendix G. Based on the CvUSE analysis, the material locations above the traditional beltline region are not limiting since they are projected to maintain Charpy upper shelf energies greater than that of the intermediate shell plate B4197-2, which is located inside the traditional beltline region. CVUSE for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 52.8 fl-lbs at Y/4-thickness.

Table 4.2-5 surmmarizes the reactor vessel pressurized thermal shock (PTS) reference temperature (RTpTs) for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 (E > 1.0 MeV). These values were calculated per the requirements of 10 CFR 50.61. Based on the RTPT5 analysis, none of the material locations above the traditional beltline region are limiting, since they are projected to maintain RTPTS values less than that of the intermediate shell plate, heat number B4 197-2, which is located inside the traditional beitline region. RTPi-S for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 199.9'F.

I-INP-07-104 Enclosure Page 8 of 11I Table 4.2-6 summarizes the reactor vessel adjusted reference temperatures (ART) at the '/4-thickness and %/-thickness wall locations for the five materials above the traditional beltline region with 55 EFPY fluence values greater than 1017 n/cm 2 (EB> 1.0 MeV). These values were calculated per the requirements of 10 CFR 50, Appendix G. Based on the ART analysis, none of the material locations above the traditional beltline region are limiting, since they are projected to maintain ART values less than that of the intermediate shell plate, heat number B4 197-2, which is located inside the traditional beltline region. ART for the intermediate shell plate B4197-2, the limiting reactor vessel material, is 195.3'F at 1/4/-thickness and 183.6'F at 3/4-thickness.

An amendment to the LRA is required in response to RAI-4.2.6.

J-INP 104 Enclosure Page 9 of 11I TABLE 4.2-4 UPPER SHELF ENERGY (Cv~USE) EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 101 7 n/cm 2 (E > 1.0 MeV)Material Description CuT (lunc ) Initial PrPredicted CVUSE_________________________________

__________

Per_______

R.G. 1.99, Revision 2 Reactor Vessel Heat Twt% (x 101 C\/USE CVUSE%Beitline Region Location Number j Type n/cm2 ft-lbs ft-lbs I Decrease Upper to Intermediate Circumferential 4P4784 ASA/Linde 124 0.06 0.2073 95 (4) 8191 13.8 Weld (AC)Upper Shell C0123- 11 2) SA-533 Gr. B 0.12 0.2073 84 71.8 14.5 Inlet Nozzle Weld (1 5-A, 15-B3, 15-C) 3P4966 ASA/Linde 124 0.02 0.0113 63(4 59.5(5)5.Inlet Nozzle 43813-5 (3) SA-508 Cl. 2 0.35 0.0113 128 108.5 15.2 Upper Shell Longitudinal WeldS (BE! BF) 4P4784 ASA/Linde 124 0.06 0.2073 95 (4) 811 13.8 I. Calculated based on guidelines in RG 1.99, Revision 2. The 55 EFPY inside surface fluence is the calculated value at the 'wetted' surface of the reactor vessel. The 'AT location fluence value is determined by calculating the IAT depth into the vessel and adding the minimum cladding thickness.

2. Upper Shell Plate CO0123-1 ex-hibited a higher value for initial RTNDT than the other Upper Shell Plate C0224- 1. Therefore, Upper Shell Plate CO 123-1 was chosen as the more conservative plate for the purpose of the embrittlement evaluation.
3. Inlet Nozzle 438B3-5 exhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 438B3-5 was chosen as the most conservative inlet nozzle for the purpose of the embrittlement evaluation.

4. As defined by ASTM E 185-82, these Charpy data are from the transition region and not the upper shelf region of the Charpy curve, since the specimen fracture surfaces exhibit less than 95% shear. Therefore, these Charpy data do not represent upper shelf energy levels and are considered conservative.
5. Predicted value is conservative, since initial Charpy data are from the transition region and not the upper shelf region of the Charpy curve.

I-INP 104 Enclosure Page 10 oflII TABLE 4.2-5 PTS REFERENCE TEMPERATURE EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 1iO'7 n/cm 2 (E > 1.0 MeV)Reactor Vessel Beltline Region Location RTPTS Calculation Per 10 CFR 50.61 Using Tables Upper to Intermediate A4P74 ASA/Linde0.6 09 2 8.0.3 1 073 5.6 55 131 30 Circumferential Weld A4P74 124 0.6 09 -2 820.41 0.3 576 55 131 30 Upper Shell C0I23-I1(2 1 C0123- 1(2) A-3 0.12 0.60 42 83.0 0.3401 0.703 58.4 34.0 134.4 270 Gr. B__Inlet Nozzle Weld 1A,5- 3P4966 A /Lne0. 02 0.92 -56 27.0 0.0203 0.173 4.7 34.3 -17.0 270 B, 15-C 124 __Inlet Nozzle 43813-5")

43813-5(3) A50 0.35 0.89 0 255.0 0.0203 0.173 44.1 34.0 78.1 270 UprSelBE/BF1 4P4784 0SA06i0.91

-20 82.0 0.3401 0.703 57.6 65.5 103.1 270 Longitudinal Welds 124 I .RTPTS is normally calculated using the fluence at the clad/base metal interface in accordance with 10 CFR 50.6 1. However, HNP calculated RTPTS using the 55 EFPY inside wetted surface fluence, which is higher than the 55 EFPY fluence at the clad/base metal interface.

2. Upper Shell Plate C0123-1 exhibited a higher value for initial RTNDT than the other Upper Shell Plate C0224-1. Therefore, Upper Shell Plate C0123-1 was chosen as the more conservative plate for the purpose of the embrittlemnent evaluation.
3. Inlet Nozzle 438B3-5 e~xhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 438B3-5 was chosen as the most conservative inlet nozzle for the purpose of the embrittlernent evaluation.

HINP-07-104 Enclosure Page I11 of I11 TABLE 4.2-6 ADJUSTED REFERENCE TEMPERATURE EVALUATION THROUGH YEAR 60 (55 EFPY) FOR MATERIALS ABOVE THE TRADITIONAL BELTLINE REGION WITH 55 EFPY FLUENCE VALUES GREATER THAN 107 nlecM (E > 1.0 MeV)Matril DsciptonChemical 55 EFPY Fluence ARTNDT, OF MagnART, OF Reactor Vessel 1Cornposition Initial Chemit. 1019 n/cm at 55 EFPY Iat 55 EFPY Beitline Re-ion Material Heat Type Cu Ni RTNDT Factor Inside A/TtI) 3/AT t) 1/4 T 1/4T 1/4T %/4T 1/4T 1/4T C I D Number wt% wt% surface Location Location Location Location Location Location Location Location Location____________

RG 1.99, Revision 2, Position I. 1 * .4 %, Upper to AA Inemdae AC 4P4784 Linde 0.06 0.91 -20 82.0 0.3401 0.2073 0.0818 47.4 31.0 58.3 46.0 85.7 57.0 Circumferential 124 Weld Upper Shell C0123- 1(2 1C0]23-l1 t 2) SA-533 0.12 0.60 42 83.0 0.3401 0.2073 0.0818 48.0 31.4 34.0 31.4 124.0 104.8 Gr. B Inlet Nozzle 15-A 15- ASA/Wed1,1ýC 3P4966 Linde 0.02 0.92 -56 27.0 0.0203 0.0113 0.00372 3.2 1.5 34.2 34.0 -18.6 -20.5 Weld B 15-C124 Inlet Nozzle 43813-5 (3) 43813-5 (3) SA-508 0.35 0.89 0 255.0 0.0203 0.0113 0.00372 30.3 13.8 30.3 13.8 60.6 27.6 Cl. 2 Upper Shell ASA/Longitudinal BE/B3F 4P4784 Linde 0.06 0.91 -20 82.0 0.3401 0.2073 0.0818 47.4 31.0 58.3 46.0 85.7 57.0 Welds ______ __ 124 I .Calculated based on guidelines in RG 1.99, Revision 2. The 55 EFPY inside surface fluence is the calculated value at the "wetted' surface of the reactor vessel. The 1/4/T and %/T location fluence values are determined by calculating the '/4T and 1/T depth into the vessel and adding the minimum cladding thickness.

2. Upper Shell Plate C0123-1 exhibited a higher value for- initial RTNDT than the other Upper Shell Plate C0224-1. Therefore, Upper Shell Plate CO0123-I was chosen as the more conservative plate for the purpose of the embrittlement evaluation.
3. Inlet Nozzle 438B3-5 exhibited a higher value for initial RTNDT than the other Inlet Nozzles (438B3-4 and 438B3-6).

Therefore, Inlet Nozzle 43813-5 wvas chosen as the most conservative inlet nozzle for the purpose of the ernbrittlemnent evaluation.

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