ML15209A791

Issuance of Amendment License Amendment Request to Implement 10 CFR 50.61a, Alternative Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

ML15209A791
Person / Time
Site:	Palisades
Issue date:	11/23/2015
From:	Jennivine Rankin Plant Licensing Branch III
To:	Entergy Nuclear Operations
Rankin J
References
TAC MF4528
Download: ML15209A791 (31)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 23, 2015 Vice President, Operations Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043-9530

SUBJECT:

PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO IMPLEMENT 10 CFR 50.61a, "ALTERNATE FRACTURE TOUGHNESS REQUIREMENTS FOR PROTECTION AGAINST PRESSURIZED THERMAL SHOCK EVENTS" (CAC NO. MF4528)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 257 to Renewed Facility Operating License No. DPR-20 for the Palisades Nuclear Plant. The amendment consists of changes to the operating license in response to your application dated July 29, 2014, as supplemented by letters dated February 13, April 1, and August 14, 2015.

The amendment approves the request to implement Title 10 of the Code of Federal Regulations (10 CFR) 50.61 a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," in lieu of 10 CFR 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."

A copy of the NRC's safety evaluation is provided in Enclosure 2. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, J~~nager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255

Enclosures:

1. Amendment No. 257 to DPR-20

2. Safety Evaluation cc: ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-255 PALISADES NUCLEAR PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 257 License No. DPR-20

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee),

dated July 29, 2014, as supplemented by letters dated February 13, April 1, and August 14, 2015, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public; and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, Renewed Facility Operating License No. DPR-20 is hereby amended to authorize revision to license condition 2.C.(8) as set forth in Entergy Nuclear Operations, lnc.'s letter dated July 29, 2014.

3. This license amendment is effective as of the date of issuance and shall be implemented within 120 days.

F R TH~ CL,EAR REGULATORY COMMISSION Da *d . elton, Ch f Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License Date of Issuance: November 23, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 257 RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Replace the following page of Renewed Facility Operating License No. DPR-20 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT 5b 5b

-5b-(8) Amendment 257authorizes the implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61.

Renewed License No. DPR-20 Amendment 2a+, 257

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELATED TO AMENDMENT NO. 257 FOR RENEWED FACILITY OPERATING LICENSE NO. DPR-20 ENTERGY NUCLEAR OPERATIONS. INC.

PALISADES NUCLEAR PLANT DOCKET NO. 50-255

1.0 INTRODUCTION

By application dated July 29, 2014 (Reference 1), as supplemented by letters dated February 13 (Reference 2), April 1 (Reference 3), and August 14, 2015 (Reference 4), Entergy Nuclear Operations, Inc. (ENO, the licensee), submitted a license amendment request (LAR) for proposed changes to the Renewed Facility Operating License No. DPR-20 for Palisades Nuclear Plant (PNP). The proposed amendment would approve the licensee's request to implement Section 50.61 a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," of Title 10 of the Code of Federal Regulations ( 10 CFR) in lieu of 10 CFR 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events."

The supplements dated February 13, April 1, and August 14, 2015, supplied additional information that clarified the application, but did not expand the scope of the application as originally noticed and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staff's original proposed no significant hazards consideration (NSHC) determination as published in the Federal Register on September 30, 2014 (79 FR 58814).

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements The requirements in 10 CFR 50.61 a (Alternate PTS rule) provide PTS screening criteria that may be implemented as an alternative to the requirements of 10 CFR 50.61. The regulation specifies that each licensee shall submit a license amendment request for review and approval by the Director of the Office of Nuclear Reactor Regulation (NRR) before implementation.

10 CFR 50.61 a includes PTS screening criteria in the form of an embrittlement reference temperature, RTMAx-x, which characterizes the subject reactor pressure vessel (RPV) material's resistance to fracture initiating from flaws, based on more comprehensive and Enclosure 2

up-to-date analysis methods and data than were used to support 10 CFR 50.61. The RTMAx-x value is the sum of RT NDT(uJ, the unirradiated reference temperature and b.T 30, the shift in the Charpy V-notch transition temperature at the 30 foot-pound (ft-lb) energy level produced by irradiation. The b.T 30 value is estimated in 10 CFR 50.61 a as the sum of two terms: matrix damage (MO) and copper rich precipitates (CRP), where MD represents the temperature shift attributed to matrix damage and CRP represents the temperature shift attributed to the copper rich precipitates. Part of the technical basis for the RTMAx-x screening criteria in 10 CFR 50.61 a is the estimation of the RPV through-wall cracking frequency (TWCF), as described in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Limit in the PTS Rule (10 CFR 50.61)," (Reference 5) and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (Reference 6). The RT MAx-x screening criteria values correspond to a TWCF value of 1 x 1o-6 per reactor operating year. This TWCF screening criteria is more than five-times as restrictive than that corresponding to the RT Prs screening criteria of 10 CFR 50.61.

The 10 CFR 50.61 a screening criteria are based on vessel wall thickness and the region of the vessel where the material is located. The region specific acceptance criteria defined in 10 CFR 50.61 a, Table 1, "PTS Screening Criteria," is then compared with the calculated values of RTMAx-x for the subject pressure vessel.

10 CFR 50.61 a also requires the licensee to perform additional evaluations, as follows:

1. To demonstrate that the date of construction and design requirements meet the 10 CFR 50.61 a requirements for applicability.

2. To evaluate plant-specific surveillance data, which includes statistical assessments of the b.T30 values for each material in the beltline. 1

3. To demonstrate that the flaw distribution in the beltline region of the subject RPV is adequately represented by the flaw distribution in Table 1 of 10 CFR 50.61 a, which represents the flaw distribution used in the probabilistic fracture mechanics (PFM) calculations that established the technical basis for the RTMAx-x limits.

2.2 Applicable Regulatory Guidance Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (Reference 7) describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence.

NUREG-0800, Revision 2, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Branch Technical Position 5-3 (Reference 8) provides guidance for applying the NRC requirements for fracture toughness, pressure-temperature limits, material surveillance, and PTS to older plants.

If the RPV beltline material is not represented by surveillance material, its L1T30 should be determined using the equations in 10 CFR 50.61a without modification.

2.3 Proposed Changes 10 CFR 50.61 a may be implemented as an alternative to the requirements of 10 CFR 50.61. By letter dated July 29, 2014, as supplemented, the licensee proposes to revise PNP's Renewed Facility Operating License Section 2.C.(8). The existing license condition states the following:

Upon implementation of Amendment 237, within one year of completing each of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Category B-A and B-D reactor vessel weld inspections, submit information and analyses requested in Section (e) of the final 10 CFR 50.61 a (or proposed 10 CFR 50.61 a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61 a) to the NRC.

The above license condition was a result of Amendment No. 237, which supported a change to the in-service inspection program that is based on topical report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." In the safety evaluation of the topical report, the NRC required licensees to amend their licenses to submit the above information (Reference 9).

The license condition would be replaced with the following:

Amendment 257 authorizes the implementation of 10 CFR 50.61a in lieu of 10 CFR 50.61.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the application and organized the evaluation into the following four topics:

1. Applicability of 10 CFR 50.61 a to the PNP RPV

2. Calculation of RTMAX-x

3. Plant-Specific Surveillance Data Assessment

4. Plant-Specific Flaw Assessment By letters dated January 20, and March 19, 2015 (References 10 and 11, respectively), the NRC issued requests for additional information (RAls) to support the review of the licensee's application. The licensee responded to the RAls by letters dated February 13, and April 1, 2015. The RAls were focused in the following three areas:

• RAI 1: related to assessment of flaws from the last ASME Section XI in-service inspection (ISi). The NRC staff's review and evaluation of RAI 1 is discussed in Section 3.4 of the safety evaluation (SE).

• RAI 2: related to inputs for the calculation of projected values of RTMAx-x. The NRC staff's review and evaluation of RAI 2 is discussed in Section 3.2 of the SE.

• RAI 3: related to clarification of the reference used in evaluating neutron fluence. The NRC staff's review and evaluation of RAI 2 is discussed in Section 3.2 of the SE.

3.1 Applicability of 10 CFR 50.61a Section (b) of 10 CFR 50.61 a describes the applicability of the rule to a given facility and states that it applies to holders of an operating license for a pressurized water reactor (PWR) whose construction permit was issued before February 3, 2010, and whose reactor vessel was designed and fabricated to the ASME Code, 1998 Edition or earlier. The purpose of this applicability restriction is that the structural and thermal hydraulic analyses that established the basis for the Alternate PTS Rule embrittlement limits only represented plants constructed before this date. The PNP construction permit was issued in 1967 and it was designed to the 1965 Edition of the ASME Code, Winter 1965 Addendum. Based on this information, the NRC staff confirms that 10 CFR 50.61 a is applicable to the PTS assessment of the PNP RPV.

Under Section (c) of 10 CFR 50.61a, each licensee is required to submit a license amendment application together with the appropriate documentation for review and approval by the Director of NRR. The application must be submitted for review and approval by the Director at least three years before the limiting RT PTs value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61 for plants licensed under Part 50, "Domestic Licensing of Production and Utilization Facilities." Based on the information contained in Sections 1.0 and 2.0 of this safety evaluation, the plant is scheduled to exceed the 10 CFR 50.61 limits in August 2017 and the licensee submitted this LAR to apply 10 CFR 50.61 a on July 29, 2014. Therefore, the NRC staff confirms that the licensee has followed the required timeline for consideration.

3.2 Calculation of RTMAx-x Under Section (c)(1) of 10 CFR 50.61 a, the licensee is required to calculate the projected values of RTMAx-x for each reactor vessel beltline material, using the calculation procedures given in 10 CFR 50.61 a, and to specify the bases for the projected RTMAX-x value of the subject material. The licensee has addressed the calculation of the projected values of RT MAX-x for each RPV beltline material in Section 8 of WCAP-17628-NP, Revision 1, "Alternate Pressurized Thermal Shock (PTS) Rule Evaluation for Palisades," which was included as an enclosure to the LAR. The required inputs to the calculations are found in Sections 4 and 5 of the WCAP document.

3.2.1 Material Property Inputs The licensee supplied material property inputs for the calculation of RTMAX-x in Sections 4 and 5 of WCAP-17628-NP, Revision 1. The staff noted that all of the inputs were previously approved by letter dated December 7, 2011 (Reference 12).

One aspect of the inputs was not clearly documented in the LAR. This related to the values of RT NDT(uJ for the plate materials. In Section 4 of the enclosure to the LAR, the licensee stated that the values of RT NDT(uJ for the plate materials were determined in accordance with the fracture toughness requirements in NUREG-0800, Revision 2, Branch Technical Position 5-3 (BTP 5-3), formerly MTEB 5-2, and the requirements of Subparagraph NB-2331 of Section Ill of the ASME Code. However, in Table 4-1, "Details of RT MAx-x Calculation Inputs for Palisades" of WCAP-17628-NP, Revision 1, all of the plates are described as using plant-specific methods to

obtain the RT NDT(u) value, without a direct reference to whether it was determined using the procedures of BTP 5-3 or of the ASME Code. Therefore, by letter dated January 20, 2015, the NRC staff requested the licensee (RAI 2a) to clarify how the RT NDT(u) values for the plates were determined by documenting which method was used for each RT NDT(u) value in Table 4-1, and by providing data demonstrating how the value of RT NDT{u) was calculated when a provision of BTP 5-3 was used.

By letter dated February 13, 2015, the licensee responded to RAI 2a with a table that summarized the specific method used for determining the RT NDT(u) value for each RPV plate.

Seven of the nine plates used BTP 5-3, paragraph 81 .1 (3)(b), which allows the licensee to estimate the temperature associated with 50 ft-lbs from Charpy impact test results (Tso) in the transverse orientation from test results obtained in the longitudinal orientation. To demonstrate how plant-specific data were used to determine the RT NDT(u) value for the seven plates, the licensee supplied copies of the certified material test record for each plate.

The NRC staff has reviewed the response to RAI 2a and notes that the licensee has followed the NRC guidance for determining the unirradiated properties. However, a letter from AREVA Inc. dated January 30, 2014 (Reference 13), to the NRC has called into question the conservative nature of BTP 5-3 in some cases. In response to the letter from AREVA, Inc., the staff has assessed the use of BTP 5-3 with an analysis of 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," surveillance data. 2 The staff found that the use of BTP 5-3, paragraph 81 .1 (3)(b), methodology to estimate the Tso was potentially not conservative in all cases. The staff proposed an adjustment factor, which is based on the current BTP 5-3 value, to be added to the current value in order to establish a conservative estimate (defined as mean+2cr, where cr is the standard deviation of the available data). The proposed adjustment factor for the A302 modified plates at PNP would raise the RT NDT(u) value for the seven plates that use BTP 5-3, paragraph 81 .1 (3)(b).

The Tso adjustment factor discussed in Reference 14 was considered here as part of the staff's independent review. This adjustment could affect the evaluation of the LAR for PNP because the Fracture Analysis of Vessels, Oak Ridge PFM model does not consider the uncertainty associated with a plant-specific value for RT NTD(u) (from Reference [6], page A-9) like the seven plates at PNP that use BTP 5-3. Given the potential for a non-conservative estimate of RT NDT(u) from BTP 5-3 for the seven PNP plates, the NRC staff assessed the effect of adding the proposed adjustment factor to the plant-specific value of RT NDT(u) for the seven plates at PNP on the implementation of 10 CFR 50.61 a at PNP.

First, the RTMAx-Aw, or the maximum reference temperature for axial welds, in Table 8-1, "RTMAX-Aw Calculation Results for Palisades at 42.1 EFPY [Effective Full Power Year]" of the enclosure to the LAR for the upper shell longitudinal welds is controlled by the plate properties, which were determined with the use of BTP 5-3. Therefore, the value reported in Table 8-1 for the upper shell longitudinal welds could be non-conservative. With the Tso adjustment factor, the maximum RTMAX-Aw value for the upper shell longitudinal welds would increase to 157 degrees Fahrenheit (F) from 135 degrees F.

2 A presentation on the NRC staff assessment of BTP 5-3 protocols was given in a public meeting on February 19, 2015 (Reference 14).

Furthermore, the staff notes that the RTMAX-Aw in Table 8-1 of the LAR enclosure for the intermediate and lower shell longitudinal welds is controlled by the weld properties, where the RTNDT(uJ values were determined with the generic method and did not use BTP 5-3. Adding the adjustment factor to the reported value for the plates that used BTP 5-3 does not change the limiting RTMAX-Aw for the intermediate and lower shell longitudinal welds. Therefore, the RTMAX-Aw values reported in Table 8-1 for the intermediate and lower shell longitudinal welds are not affected by the potential non-conservatism in the BTP 5-3 estimates. In a similar manner, the staff examined the calculations summarized in Tables 8-2, "RTMAX-PL Calculation Results for Palisades at 42.1 EFPY" and 8-3, "RTMAx-cw Calculation Results for Palisades at 42.1 EFPY" of the LAR enclosure for the plates and for the circumferential welds. Where the plates used BTP 5-3, the staff reassessed the RT NDT(uJ value with the Tso adjustment factor added. With the Tso adjustment factor, the maximum RTMAX-PL value for the upper shell plates would increase to 172 degrees F from 150 degrees F; the maximum RTMAX-PL value for the intermediate and lower shell plates is unchanged. For the circumferential welds, the maximum RTMAx-cw value for the intermediate-to lower weld does not change while the value for the upper-to-intermediate weld increases to 172 degrees F from 150 degrees F.

To summarize, consideration of the Tso adjustment factor would increase the limiting values of RTMAx-x for the upper shell materials. Given the revised RTMAx-x values for upper shell region from the staff's independent analysis, the conclusion based on the results in Table 8-4, "RTMAx-x values for Palisades at 42.1 EFPY" in the submittal would not change. Specifically, the potential uncertainty related to the use of BTP 5-3 for the unirradiated properties of the seven plates could change some details in Tables 8-1, 8-2, and 8-3, but would not change the conclusion from Table 8-4. Therefore, based on the response to RAI 2a and the staff's independent analysis of Tables 8-1, 8-2, and 8-3 in the LAR enclosure, the staff's concern expressed in RAI 2a about the material property inputs used in the calculations is resolved. Even accounting for the potential non-conservatism in BTP 5-3, the NRC staff concludes that the licensee has demonstrated that the limiting values of RTMAx-x are below the screening criteria of 10 CFR 50.61a.

3.2.2 Neutron Fluence Inputs As part of its review, the NRC staff noted two specific issues related to the fluence values summarized in Table 5-1, "Maximum Neutron Fluence on the RV Clad-to-Base Metal Interface for Palisades at 42.1 EFPY" of the LAR enclosure. First, in Table 5-1, the values listed for the intermediate and lower longitudinal (axial) welds in the vessel are the same: 2.161 x 1019 neutron/cm 2 (E > 1.0 MeV), which is 63% of the peak values for the adjacent intermediate and lower shell plates. The methodology for determining the maximum fluence at the axial welds relative to the maximum fluence in the adjacent plates is not discussed in the LAR. Therefore, in RAI 2b of the staff's letter dated January 20, 2015, the staff asked the licensee to identify how the maximum fluence for each region and component in Table 5-1 was determined from the detailed fluence information contained in References 8 and 13 listed in Section 10 of the LAR enclosure. For clarity, the staff requested the licensee to illustrate the axial position of the active fuel and the azimuthal position of the peak fluence values for the adjacent intermediate and lower shell plates on Figure 4-1, "Details of RTMAx-x Calculation Inputs for Palisades" of the LAR enclosure.

By letter dated February 13, 2015, the licensee responded to RAI 2b with the following summary:

The spatial distribution of the neutron fluence for the Palisades reactor vessel was determined on a cycle-by-cycle basis using the synthesis methodology described in Reference 8 of the LAR enclosure. This analysis provided a fine spatial grid of fluence values over the inner surface of the reactor vessel. A scan of these fine grid fluence distributions coupled with the knowledge of the locations (axial and azimuthal extent) of individual reactor vessel materials allowed the determination of the maximum for each weld and plate in the beltline region.

The licensee also provided the axial locations of active fuel and the azimuthal locations of peak fluence in a revision to Figure 4-1, "Identification and Location of Beltline Region Materials for the Palisades Reactor Vessel" of the LAR enclosure.

The second issue is related to the neutron flux used to calculate the reported values for fluence.

The flux is one of the inputs used in 10 CFR 50.61a to estimate LlT30. The LAR lacks neutron flux values in Tables 8-1, 8-2, and 8-3, for the calculation of RT MAx-x for the materials listed in Table 4-1. Therefore, in RAI 2c of letter dated January 20, 2015, the NRC staff asked the licensee to supply neutron flux values on a per cycle basis for the limiting material/region, which is weld heat W5214, the intermediate shell longitudinal weld.

By letter dated February 13, 2015, the licensee responded to RAI 2c with a table of neutron flux and end-of-cycle neutron fluence for the limiting axial weld that covers the past operating history and projected operating conditions.

The NRC staff reviewed the responses to RAI 2b and 2c and noted that the flux values in the response to RAI 2c match those for the 60° azimuth in Table 2.2-3 of WCAP-15353-Supplement 2-NP, Revision 0 (Reference 15). However, in the revised Figure 4-1 included in the response to RAI 2b, there is no intermediate shell longitudinal weld at the 60° azimuthal location. Therefore, based on the responses to RAI 2b and 2c, the staff asked for further clarification by letter dated March 19, 2015, related to the frame of reference for the azimuthal locations of peak fluence and the axial welds.

By letter dated April 1, 2015, the licensee noted that the azimuthal coordinate system shown in Figure 4-1 from the response to RAI 2b pertains to a reactor design azimuthal coordinate system, which differs from the azimuthal coordinate system used in the fluence evaluation. The fluence evaluation used a system that is offset from the reactor design coordinate system by 90 degrees. For example, longitudinal welds 1-112A and 3-112A shown at 90 degrees in Figure 4-1, which depicts the reactor design azimuthal coordinate system, are oriented at O degrees in the azimuthal coordinate system used for the fluence evaluation, and longitudinal weld 2-112C, which is shown at 150 degrees in Figure 4-1, is located at 60 degrees in the azimuthal coordinate system used for the fluence evaluation. A revision of Figure 4-1 was given in the April 1, 2015, letter to clarify the location of reactor vessel materials within the azimuthal coordinate system used in the reactor quadrant model used for the fluence evaluation.

In addition, the licensee noted in the RAI response dated April 1, 2015, that Table 8-1 in the LAR, which provides calculated RT MAX-Aw values for the axial welds, applies the maximum

fluence calculated for any of the axial welds in a given shell course to all of the axial welds in that course. Therefore, all of the calculated RTMAX-Aw values for the intermediate and lower shell axial welds in Table 8-1 use the same fluence values even though they are not all located at the same relative location in the azimuthal coordinate system used for the fluence evaluation.

The NRC staff has reviewed the responses for RAI 2b and 2c and the RAI response dated April 1, 2015, and notes that the licensee has explained how the azimuthal fluence variation was determined. Furthermore, the revision to Figure 4-1 in the April 1, 2015, letter offers added clarity to the analysis. Finally, the table of flux values in the response to RAI 2c allows the staff to verify the calculation of RTMAX-Aw values for the intermediate and lower shell axial welds. The NRC staff's verification of the RTMAX-Aw values is consistent with the licensee's values listed in Table 8-1 of the LAR enclosure. For the stated reasons, the RAI responses satisfy the staff's concerns and the issues expressed in RAls 2b and 2c are resolved.

3.2.3 Fluence Calculations For input to its PTS analysis, Entergy performed fluence calculations in accordance with Westinghouse Topical Report (TR) WCAP-15353, Revision 0, "Palisades Reactor Vessel Neutron Fluence Evaluation," (Reference 16). The TR was further supplemented in July 2011 with WCAP-15353- Supplement 2- NP, Revision 0, and in June 2013 with WCAP-15353-Supplement 3 - NP, Revision 0, "Palisades Reactor Pressure Vessel Fluence Evaluation" (Reference 17).

Section 3.0 of the LAR states the following:

In 2013, ENO re-calculated PNP reactor vessel fluence, based on actual reactor operation through fuel Cycle 22, and expected fluence based on projected operations through Cycle 26. This evaluation was documented in "Palisades Reactor Pressure Vessel Fluence Evaluation, "WCAP-15353- Supplement 3-NP, Revision 0, which was submitted [by Reference 17). The evaluation concluded that the PNP RV limiting welds would not reach the 10 CFR 50.61 PTS screening criteria limit until August 2017.

The NRC staff reviewed the application and determined that additional information was required. In RAI 3 of the staff's letter dated January 20, 2015, the NRC staff asked the licensee to clarify which supplement of WCAP-15353-NP was used in evaluating PNP's alternate fracture toughness requirements for protection against PTS. By letter dated February 13, 2015, the licensee responded that WCAP-15353 - Supplement 2 - NP, Revision 0 was used for the RPV neutron fluence values in the alternate PTS rule elevation. WCAP-15353- Supplement 3 - NP, Revision O was used to determine when the 10 CFR 50.61 screening criterion date would be reached for the limiting RPV material, based on recent plant operations. The licensee provided the NRC staff clarification for which WCAP-15353-NP supplement was used for the neutron fluence evaluation. Based on the considerations discussed above, the NRC staff has determined the licensee has adequately clarified the WCAP-15353-NP supplement for alternate fracture toughness requirements for protection against PTS.

3.2.4 Regulatory Guide 1.190 Guidance The NRC staff compared the methods used to calculate neutron fluence in WCAP-15353, Supplement 2-NP, Revision 0 with the guidance provided in RG 1.190. The guidance provided in RG 1.190 indicates that the following attributes compose an acceptable fluence calculation:

• a fluence calculation performed using an acceptable methodology

• analytic uncertainty analysis identifying possible sources of uncertainty

• benchmark comparison to approved results of a test facility

• plant-specific qualification by comparison to measured fluence values The licensee stated that the neutron calculations provided in WCAP-15353, Supplement 2-NP, Revision O are performed in a manner consistent with the guidance set forth in RG 1.190. A solution to the Boltzmann transport equation is approximated using the two-dimensional discrete ordinates transport (DORT) code. The licensee uses a cross-section library based on the ENDF/B-VI nuclear data. Numeric approximations include a Ps Legrende expansion for anisotropic scattering and the modeling uses S16 order of angular quadrature. These cross-section data and approximations are in accordance with the modeling guidance contained in RG 1.190. Since the licensee used NRC-approved RG 1.190 adherent methods to determine the RPV neutron fluence, the NRC staff determined that the fluence calculations are acceptable.

Space and energy dependent core power (neutron source) distributions and associated core parameters are treated on a fuel-cycle-specific basis. Three-dimensional flux solutions are constructed using a synthesis of azimuthal, axial, and radial flux. Source distributions include cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions, which are used to develop spatial and energy dependent core source distributions that are averaged over each fuel cycle. This method accounts for source energy spectral effects by using an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history for each fuel assembly. The neutron source and transport calculations, as described above, are performed in accordance with the guidance set forth in RG 1.190. Based on the consistency with the guidance contained in RG 1.190, the NRC staff determined that the source and transport calculations are acceptable.

The fluence methods are supported by an analytic uncertainty analysis and the estimated uncertainty is less than 20 percent which is in accordance with RG 1.190 and hence acceptable.

Details of the analytic uncertainty analysis are provided in WCAP-15353- Supplement 2 - NP, Revision 0.

WCAP-15353, Revision 0 describes the methods qualification using the standard benchmark problems found in RG 1.190. WCAP-15353, Revision 0 compared the calculations with the benchmark measurements from the Poolside Critical Assembly simulator at the Oak Ridge National Laboratory and the surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment. The comparison of PNP demonstrated that the plant-specific measurement data is less than 20 percent different from the analytical prediction. The NRC staff determined that these constitute acceptable test facilities, as they are specifically referenced in RG 1.190 and are within the acceptable analytical prediction.

WCAP-15353 - Supplement 2 - NP, Revision 0, contains acceptable plant-specific benchmarking for PNP as it contains a database of PWR dosimetry benchmarking. The PNP unit-specific geometry, a PWR RPV, is well represented within the database. PWR-specific benchmarking documented in WCAP-15353- Supplement 2- NP, Revision 0, indicates that surveillance capsule fluence can be calculated within 20 percent of measured values, which is in accordance with RG 1.190. The NRC staff has determined, therefore, that these uncertainties are acceptable.

The NRC staff has reviewed the neutron fluence methods used in the LAR and the RAI responses supplied by the licensee to implement 10 CFR 50.61a. The licensee's RAI response adequately clarified which supplement of WCAP-15353-NP was used in evaluating PNP's alternate fracture toughness requirements for protection against PTS. The NRC staff concludes that the neutron fluence methods used to support the implementation of 10 CFR 50.61 a are acceptable.

3.2.5 Conclusion To summarize, the NRC staff has reviewed the inputs required for the calculation of RT MAx-x and the neutron fluence methods used in the LAR, as supplemented by the RAI responses. The response to RAI 2a clearly identified those plates that used BTP 5-3 to obtain estimated RT NDT(u)- The NRC staff performed independent calculations to evaluate the potential non-conservatism associated with the use of BTP 5-3 and concluded the adjustment factors to account for this non-conservatism would not change the conclusions of the PNP request for implementation of 10 CFR 50.61 a. The responses to RAI 2b and RAI 2c clarified the azimuthal and axial fluence/flux distributions. The response to RAI 3 clarified that WCAP-15353-NP, Supplement 2 - NP, Revision 0 was used for the neutron fluence evaluation. The NRC staff reviewed the fluence evaluation and determined the calculations were performed in a manner consistent with RG 1.190. Therefore, the NRC staff concludes that the material and neutron fluence inputs along with the neutron fluence methods used to support the implementation of 10 CFR 50.61 a are acceptable.

3.3 Evaluation of Plant-Specific Surveillance Data Under Section (f)(6) of 10 CFR 50.61a, the licensee is required to verify that its projected embrittlement as quantified by the value of RTMAx-x for each material is appropriate. The appropriateness of these values is determined by comparing plant-specific surveillance data to the projections of 10 CFR 50.61 a. For each heat of material, where there are three or more surveillance data points measured at different neutron fluences, the licensee is required to determine if the surveillance data show a significantly different trend than the embrittlement trend curve (ETC) from 10 CFR 50.61 a; significance is determined by following the procedures specified in paragraphs (f)(6)(ii), (f)(6)(iii), and (f)(6)(iv) of 50.61 a. When fewer than three surveillance data points exist for a specific heat of material, then the ETC projections are used without this consistency check.

Section 8.2, "Surveillance Capsule Data Statistical Checks" of the LAR enclosure describes the three plant-specific heats of material: one plate material (Heat C-1279) and two welds (Heat W5214 for the axial welds and Heat 27204 for the circumferential welds) that were

monitored in surveillance programs covered by 10 CFR Part 50, Appendix H. The heat-specific Lff30 values for each fluence level, determined as specified by the requirements of 10 CFR Part 50, Appendix H, were summarized in Section 6, "Surveillance Capsule Data" and assessed for consistency with the ETC in Section 8.2 of the LAR enclosure. The licensee's analysis demonstrates that the surveillance data considered are consistent with the overall ETC. As a result, 10 CFR 50.61 a requires that the projected values of Lff30 be used without modification.

The staff notes that the surveillance data for the three materials includes three or more different fluence levels and comes from either the original PNP surveillance program, the supplemental PNP program, or other light water reactor (LWR) Appendix H surveillance programs. The surveillance data for the limiting material, weld Heat W5214, comes from 11 different fluence levels where the values range from 12 to 208 percent of the maximum axial weld fluence in Table 8-:7, "Surveillance Data Evaluation for Palisades Weld Wire Heat W5214" of the LAR enclosure. Furthermore, the staff has reviewed CE NPSD-119, Revision 1 (Reference 18) and noted that no additional surveillance data for Weld Heat W5214 exists, verifying the completeness of the surveillance checks for the welds included in Table 4-1. The review confirmed that only those plants used in Section 8.2 of the submittal were heat-specific matches for the PNP weld heats in Table 4-1.

For comparison, the NRC staff has reviewed previous PTS submittals made by the licensee under 10 CFR 50.61. The staff finds that the surveillance data for Heat W5214 summarized in the LAR is the same as that used in the submittal dated December 7, 2011, (Reference 12) that considered surveillance data for PNP under 10 CFR 50.61. In that 2011 submittal, following the rules for 10 CFR 50.61, the licensee used the surveillance data to calculate a plant-specific chemistry factor that reduced the projected embrittlement of the limiting axial weld fabricated from Heat W5214 and extended the period of time under which PNP could operate in compliance with 10 CFR 50.61. Under 10 CFR 50.61 a, the same surveillance data is used as a check on the projected embrittlement of the limiting axial weld fabricated from Heat W5214.

This check confirms the embrittlement projections of the 10 CFR 50.61 a ETC and, thus, does not change the values of RTMAx-x.

To summarize, the NRC staff has reviewed industry documents and past PNP submittals and finds the comparison of surveillance data in Table 8-5, "Surveillance Capsule Materials for Palisades," includes all of the available results from 10 CFR Part 50, Appendix H surveillance programs. The staff finds the assessments for plate Heat C-1279 and weld Heats W5214 and 27204 are consistent with the overall ETC. Therefore no modification of the projected RTMAx-x values is required.

3.4 Plant-Specific Flaw Assessment Under Section (c)(2) of 10 CFR 50.61a, the licensee is required to perform an examination and an assessment of flaws in the RPV beltline as required by paragraph (e) of 10 CFR 50.61 a.

The licensee must verify that the requirements of paragraphs (e), (e)(1), (e)(2), and (e)(3) of this section have been met. The licensee has summarized the inspection results in Tables 7-1, "Reactor Vessel ISi History for Palisades Beltline and Extended Beltline Materials," and 7-2, "ISi Information for Reactor Vessel Beltline and Extended Beltline Flaws for Palisades," of the LAR enclosure and illustrated the approximate location for each flaw in Figure 8-1. The licensee's

assessment of flaws is included in Tables 8-11, "Alternate PTS Rule Allowable Number of Flaws in Plates and Forgings Scaled for Palisades," and 8-12, "Alternate PTS Rule Allowable Number of Flaws in Welds Scaled for Palisades." Both assessments show that the plant-specific data are bounded by the data in Tables 2 and 3 of 10 CFR 50.61 a.

The NRC staff reviewed the inspection results in Section 7, "lnservice Inspection Data," and the assessment in relevant portions of Section 8, "Determination of RTMAx-x Values for all Beltline and Extended Beltline Region Materials," of the LAR enclosure. The staff notes that in February 2014, the licensee conducted inservice inspection (ISi) ultrasonic testing (UT) along with eddy current testing (ET). The ET examinations documented that the flaws discovered with UT near the inner diameter (ID) of the beltline and extended beltline regions were not connected to the ID surface. Furthermore, the staff noted two issues that needed clarification. First, there are a large number of flaws (17) grouped together in weld 2-112B (Figure 8-1 ). Given the fact that so many flaws are relatively close together, the NRC staff expressed concern that the separation between flaws could be relatively small, and that the licensee had not offered any discussion of the proximity of the indications to each other. Therefore, in RAI 1a of letter dated January 20, 2015, the staff asked the licensee to address the proximity of the indications in weld 2-112B, given the requirements in ASME Section XI, paragraph IWA-3300(b).

By letter dated February 13, 2015, the licensee responded to RAI 1a by noting that, according to Figure IWA-3300-1, if the separation between flaws is smaller than the through-wall dimension of the largest flaw, the two flaws shall be considered as one larger flaw. For the flaws in weld 2-112B, the largest through-wall dimension in the group of 17 flaws was 0.06 inches while the smallest separation between flaws in the group was 0.1725 inches; therefore, flaws within the group of 17 were treated individually.

The second issue relates to Table 8-11 in the submittal. All the flaws detected in the February 2014 ISi at PNP have been dispositioned in accordance with 10 CFR 50.61a, Table 3, "Allowable Number of Flaws in Plates and Forgings," and placed into different bins and compared, on a unit area basis, to the scaled maximum number used in the technical basis for 10 CFR 50.61 a as recommended in NRC memorandum (Reference 19), and as noted in the 10 CFR 50.61a Final Rule announced in the Federal Register on January 4, 2010 (75 FR 13).

To use 10 CFR 50.61 a without further evaluation, the plant-specific results must be bounded by the flaw distributions in Tables 2 and 3 of 10 CFR 50.61 a. The licensee has assumed that all of the flaws detected in the 2014 ISi occur in the plate rather than in the weld. It was not clear to the staff if the surface area associated with the weld was included in the surface area inspected for the plate. Therefore, in RAI 1b of the letter dated January 20, 2015, the NRC staff asked the licensee to clarify how the total plate area was calculated and whether it includes the surface area of the weld.

By letter dated February 13, 2015, the licensee responded to RAI 1b by saying that the total inspected plate area was determined with the dimensions given in Table 8-10, "Inspection Length and Area for Palisades," in the LAR enclosure along with the requirements for inspection volume in Figure IWB-2500-1 of the ASME Code,Section XI, subtracting the surface area associated with the weld itself. As noted in Table 8-1 O of the submittal, the total plate area was rounded down for comparison to Table 3 in 10 CFR 50.61 a.

The NRC staff has reviewed the RAI 1a and 1b response and performed independent calculations to confirm the flaw density. The staff finds that the responses to RAI 1a and 1b have clearly demonstrated that ( 1) each detected flaw has been considered with respect to its proximity to adjacent flaws, and (2) the surface area of the plate inspection was calculated using plant-specific dimensions and following Code inspection requirements, subtracting the weld area. The staff notes that the ISi results are not definitive with respect to the location of the flaw; the assumption by the licensee that all of the flaws occur in plate and subtracting the weld area from the inspection area is a conservative approach because 10 CFR 50.61a allows more numerous and larger flaws in welds than plates. For example, in the weld regions at PNP, 10 CFR 50.61 a would allow three flaws greater than 0.375 inches, but less than 0.475 inches, in through-wall extent (TWE) in the inspection, where no flaws greater than 0.375 inches in TWE are allowed in the plate regions. For these reasons, the staff considers that the flaws detected in the February 2014 ISi are bounded by the assumptions used in the development of 10 CFR 50.61 a. The issues related to RAI 1a and 1bare therefore resolved.

To summarize, the NRC staff finds that the submittal, as supplemented by the responses to RAI 1a and 1b, has demonstrated that the flaws detected in the February 2014 ISi are bounded by the assumptions used in the development of 10 CFR 50.61a.

3.5 Overall Conclusion Based on the NRC staff's review of (1) the applicability of 10 CFR 50.61 a to PNP, (2) the calculations of RT MAx-x, (3) the surveillance checks, and (4) the flaw density and size distributions supplied in the licensee's application dated July 29, 2014, as supplemented by licensee letters dated February 13, April 1, and August 14, 2015, the staff concludes that the PNP RPV beltline and extended beltline materials have end-of-license extension RTMAX-x values that are below the 10 CFR 50.61a screening criteria. Hence, the NRC staff grants ENO's license amendment request to use 10 CFR 50.61 a in lieu of 10 CFR 50.61 through the end of the current license, which is March 24, 2031.

4.0 PUBLIC COMMENTS As discussed in SE Section 1.0, the NRC staff published a public notice concerning the proposed amendment in the Federal Register on September 30, 2014. The notice included the NRC staff's proposed NSHC determination. The notice also gave an opportunity for public comments on the staff's proposed NSHC determination.

Public comments were received on the proposed amendment. Specifically, Beyond Nuclear (BN) supplied comments dated October 30, 2014 (Reference 20).

Consistent with the requirements in 10 CFR 50.91 (a)(2)(ii), and as noted above, the NRC staff's public notice solicited comments specifically on the proposed NSHC determination.

BN supplied comments on its objections to NRC's issuance of the proposed amendment and to the regulations in 10 CFR 50.61a. BN also gave comments pertaining to PNP that were not related to the proposed amendment. Not all of BN's comments pertained specifically to the NRC staff's proposed NSHC determination. The staff's responses below are in response to comments that pertain to the proposed amendment.

4.1 Public Comment Nos. 1 and 9 out of a Total of 23 Comments (summarized)

BN Comment:

BN asserts in numerous portions of its submittal that the NRC has weakened the PTS regulations to allow embrittled reactors to continue to operate NRC Response:

The NRC requires licensees to meet either 10 CFR 50.61 or 10 CFR 50.61 a to demonstrate fracture toughness against PTS events. Section 50.61 was published in the Federal Register on July 23, 1985 (50 FR 29937) and Section 50.61a was published in the Federal Register on January 4, 2010 (75 FR 13). 10 CFR 50.61a was developed based on many more years of operational experience than was available when 10 CFR 50.61 was published. Also, during the 25 years between 1985 and 2010, the state of knowledge on embrittlement, fracture mechanics, and the populations of flaws found in reactor pressure vessels all increased substantially. When coupled with increased computer modeling capabilities, these advances in the state of knowledge permitted the 10 CFR 50.61a calculations to capture much more accurately the details of a PTS event than was possible when 10 CFR 50.61 was adopted. As a result, the technical basis document for 10 CFR 50.61 a concluded that the risks of through-wall cracking from a PTS event were much lower than previously estimated. The agency increased the reference temperature screening criteria in 10 CFR 50.61a (relative to those in 10 CFR 50.61) while at the same time decreasing the risk of vessel failure associated with these limits. As described in Chapter 10 of NUREG-1806, the risk associated with 10 CFR 50.61a embrittlement limits ( 1x10- 5 events/reactor operating year) is five times less than the risk associated with 10 CFR 50.61 embrittlement limits (5x10-5 events/reactor operating year). Additional information concerning the relationship between vessel failure risk and reference temperature limits can be found in NUREG-1874 for 10 CFR 50.61a and in SECY-82-465 (Reference 21) for 10 CFR 50.61. As a result of the work described in NUREG-1806, NUREG-1874, and related documents referenced in these two reports, the Commission found that the reference temperature limits found in 10 CFR 50.61a provide a more accurate and realistic linkage between embrittlement and reactor vessel fracture risk than the limits established in 10 CFR 50.61.

ENO has elected to use 10 CFR 50.61 a to establish the fracture toughness of its reactor vessel and as stated in SE Section 2.0, this election is permitted by NRC regulations. As described in this SE, the NRC staff has concluded that the licensee's evaluation satisfies the requirements of 10 CFR 50.61a.

4.2 Public Comment Nos. 3. 16. and 19 out of a Total of 23 Comments (summarized)

BN Comment:

BN questions in numerous portions of its submittal the adequacy of the schedule for ENO's withdrawal of surveillance capsules and testing of metal coupons from these capsules.

Specifically, BN questions if ENO and the NRC are relying on extrapolations of data to assess the current embrittlement state of the Palisades reactor pressure vessel, and asks what assurance can be given that the steel in the Palisades reactor pressure vessel is not embrittling more rapidly than expected.

NRC response:

Appendix H of 10 CFR 50 establishes the requirements for the reactor vessel material surveillance program. These requirements, and the licensee's compliance with these requirements, are administered separately from the licensee's compliance with the PTS rules (10 CFR 50.61 and 10 CFR 50.61 a). As such, the adequacy of the surveillance capsule program is not directly related to this license amendment request. Nevertheless, information relevant to this comment is provided in the following paragraphs.

Appendix H to 10 CFR Part 50, "Reactor Vessel Material Surveillance Program Requirements,"

requires licensees to have a material surveillance program to monitor changes in the mechanical properties of the reactor pressure vessel materials. Surveillance programs include a number of capsules that contain Charpy and tensile specimens (sometimes called "coupons").

These capsules are placed inside the RPV and thus lie closer to the core than the vessel itself.

Because of their location, the amount of irradiation experienced by the capsules, and by the samples inside the capsules, exceeds that experienced by the RPV wall itself at any moment in time. For example, the amount of irradiation accumulated by the specimens in a capsule after 10 years may be the same as that experienced by the RPV wall after 30 or 50 years of operation. The exact difference in irradiation exposure received by the capsules and the wall depends on capsule placement. This acceleration is intentional, and is called a "lead factor."

Having the capsules "lead" the amount of neutron exposure and, thereby, embrittlement experienced by the RPV facilitates safe life management of nuclear RPVs. This practice ensures that regulatory decisions and operational limits, both of which need information on strength and toughness, are informed years in advance, and rely on data that is interpolated between measured points, not extrapolated beyond the last measured point. With specific regard to Palisades, the surveillance data for the limiting W5214 weld heat have been collected up to an irradiation exposure corresponding to 87.6 EFPY of exposure for the reactor pressure vessel itself. Since the renewed license for PNP is only for 42.1 EFPY it is clear that this license amendment request and safety evaluation are not based on data that has been extrapolated.

It has been noted in the public's comments that the schedule for surveillance capsule removal and testing from the Palisades vessel has changed over time. This is common and is needed as part of license renewal. For PNP, the most recent adjustment was submitted to the NRC by letter dated September 19, 2006 (Reference 22). Nuclear Management Co., LLC (the licensee at the time) requested to revise the Appendix H withdrawal schedule. By letter dated August 14, 2007 (Reference 23), the NRC approved the plant's modified withdrawal schedule because it satisfied the requirements of ASTM E-185-82 and 10 CFR Part 50, Appendix H.

There is a great amount of collective experience concerning the effects of neutron embrittlement and its effect on reactor pressure vessel steels (e.g., from all of the 10 CFR 50 Appendix H surveillance programs in the USA, from other surveillance programs in operating reactors worldwide, and from the national and international research communities). Nevertheless, it remains important to compare the data collected from the 10 CFR Part 50, Appendix H surveillance programs for a particular plant to the predictions of the embrittlement trend equations that reflect this much larger body of data. This comparison provides a verification that the PNP reactor pressure vessel is not em brittling more rapidly than expected. 10 CFR 50.61 a requires such a comparison in the form of several statistical tests that must be performed on RPV surveillance data to determine whether the surveillance data are sufficiently close to the predictions of the ETC. These comparisons determine whether plant-specific surveillance data deviate significantly from the predictions in a manner that suggests the predictions to be non-

conservative. SE section 3.3 concludes that the plant-specific plate and welds are consistent with the overall ETC and no modification of the projected RTMAx-x values is required.

4.3 Public Comment Nos. 2. 3. and 13 out of a Total of 23 Comments (summarized)

BN comment:

BN questions if the NRC staff is aware of the following international activities:

1. Articles regarding reactor pressure vessel embrittlement and PTS risks at Japanese PWRs. Specifically, articles report that embrittlement predictions are non-conservative.

2. Efforts to anneal RPVs in other countries and the results. In addition, BN questions if annealing would be successful at PNP.

NRC response:

International activities are not directly related to ENO's license amendment request to adopt 10 CFR 50.61a; however, information on BN's questions and comments on international activities is described below.

1. The NRC staff is aware of the evaluation in Japan of the embrittlement at the Genkai nuclear reactor. As a result of the embrittlement measurement that was significantly underpredicted, the Japanese industry and the Japanese regulatory community undertook an extensive investigation. A report summarizing this investigation was published by the Japanese agency NISA (Nuclear and Industrial Safety Agency), in Japanese. Presentations on this report (made in English) at meetings attended by the NRC staff show no microstructural evidence of the cause for the measured embrittlement. The Japanese have re-calibrated their embrittlement prediction models using the Genkai data as part of the dataset. Additionally, the American Society for Testing and Materials (ASTM) has recently updated its embrittlement prediction model (ASTM E900-15). This effort considered embrittlement data from Japan. For the specific conditions of the PNP RPV, the predictions of ASTM E-900-15 are quite close to those of 10 CFR 50.61a.

2. The NRC staff is aware of annealed RPVs in the former Soviet Union. Annealing is one of the many embrittlement management tools available to licensees in the United States.

The NRC has regulations under 10 CFR 50.66 and associated regulatory guidance in RG 1.162 (Reference 24) concerning annealing. While annealing is an option, no utility in the United States has, to date, pursued annealing as part of its aging management plan.

4.4 Public Comment No. 4 out of a Total of 23 Comments BN comment:

WHY does Palisades have the worst embrittled reactor pressure vessel in the U.S., as NRC's Nuclear Regulatory Research staffer Jennifer Uhle admitted was the case at a public meeting at the Beach Haven Event Center in South Haven on Feb. 29, 2012? Is it due to the impurities in the RPV steel from its initial fabrication?

NRC response:

Within the population of reactors currently operating in the United States, PNP is among the most embrittled. Nevertheless, ranking of the embrittlement in one reactor relative to another is not important from a licensing or a safety perspective because NRC regulations establish screening limits on embrittlement (which is quantified by the "reference temperature") that must be met by every operating reactor. Compliance with these regulations depends on how each reactor's reference temperature compares with these screening limits; comparisons between reactors are not relevant to licensing decisions, or to operating safety.

The magnitude of embrittlement that occurs in RPVs depends on many factors. As correctly mentioned in Public Comment No. 4, it depends on the presence of impurity elements in the RPV steel during its initial fabrication, most notably copper and phosphorus. Also, embrittlement depends on elements such as nickel and manganese; these are intentional alloying additions to the reactor vessel steel without which adequate strength and toughness properties could not be achieved. Finally the magnitude of embrittlement depends on the total amount of irradiation exposure of the vessel, and on the temperature at which the vessel operates. The embrittlement prediction formula in 10 CFR 50.61 a accounts for all of these factors.

The embrittlement in the PNP reactor is among the highest of those currently operating in the United States because of a relatively high copper content coupled with a relatively high total exposure to neutron irradiation. However, as documented in SE Section 3.5, because the magnitude of embrittlement at PNP is below the regulatory screening criteria, the licensee has demonstrated that PNP meets NRC regulations in 10 CFR 50.61a.

4.5 Public Comment Nos. 5 and 6 out of a Total of 23 Comments (summarized)

BN comment:

BN questions how a PTS accident progresses and the consequences of a PTS event (e.g. would it result in a catastrophic event that would result in a breach of containment, resulting in a large number of casualties and/or a large loss of property) such as that noted in NUREG/CR-2239, "Technical Guidance for Siting Criteria Development" (ADAMS Accession No. ML072320420).

NRC response:

This question pertains to the consequence of a PTS event and not to the license amendment, which allows a licensee to employ an alternative methodology to assess how well its reactor vessel materials respond to a PTS event. As such, no specific response is given herein; however, Chapter 10, Section 10.3, of NUREG-1806 has information on the NRC's evaluation of

the potential post-accident sequence progression following a PTS event.

4.6 Public Comment No. 17 out of a Total of 23 Comments BN comment:

Since risk is probability times consequences, isn't it accurate to say that embrittlement/pressurized thermal shock risks are significantly worse than they were in 1971, pre-operations?

NRC response:

The NRC staff agrees that reactor pressure vessel at PNP is more embrittled than it was in its pre-service condition. As with every nuclear reactor, the RPV becomes more embrittled throughout the operating life. Although the likelihood that a PTS event could fracture the vessel is higher today than in 1971, the regulations in 10 CFR 50.61and10 CFR 50.61a contain requirements for reactor vessel fracture toughness that provide adequate protection against PTS events and ensure that the risk of vessel fracture is maintained at a low level (see SE Section 4.1). As seen in SE Section 3.5, the NRC staff has concluded the licensee has met the requirements of 10 CFR 50.61a.

4.7 Public Comment No. 18 out of a Total of 23 Comments BN comment:

What about the synergistic effects of Palisades' many problems? Not only is the RPV the worst embrittled in the US, but Palisades needs the 2nd steam generator replacement in its history, its badly corroded reactor lid is now 6 years overdue for replacement, and a diversity of leaks, breakdowns, and failures have occurred in the past few years. Might not all these problems add up to a catastrophic failure at Palisades? Why doesn't NRC address the totality of all these risks as a whole, as that is the reality of the situation, instead of just one system, structure, or component at a time, in isolation? To the equipment problems, there are the safety culture violations at Palisades.

NRC response:

The NRC staff is not aware of another steam generator replacement or a reactor vessel head replacement at Palisades. At the request of Commissioner Magwood by letter dated April 25, 2013 (Reference 25), the licensee explained the status of capital improvements at PNP in a letter dated May 16, 2013 (Reference 26). The licensee explained that there were not any near-term plans to replace the steam generators or the reactor vessel head. The NRC staff notes the NRC's mission is to protect the health and safety of the public and the environment.

The NRC performs independent inspections of the reactor vessel head and steam generators.

The results of those inspections indicates the licensee is operating PNP safely.

The Reactor Oversight Process (ROP) action matrix reflects overall performance issues at the plant. More information about the ROP Action Matrix can be found at http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/actionmatrix_summary.html. The ROP integrates the NRC's inspection, assessment, and enforcement programs. The fundamental building blocks that form the framework for the regulatory oversight process are seven cornerstones of safety: initiating events, mitigating systems, barrier integrity, emergency

preparedness, occupational radiation safety, public radiation safety, and security. These cornerstones have been grouped into three strategic performance areas: reactor safety, radiation safety, and safeguards. This framework is based on the principle that the agency's mission of assuring public health and safety is met when the agency has reasonable assurance that licensees are meeting the objectives of the seven cornerstones of safety. In addition to the seven cornerstones, the ROP features three cross-cutting areas (human performance, safety conscious work environment, and problem identification and resolution) because they can affect each of the cornerstones. The NRC examines these cross-cutting areas in aggregate to find potential performance deficiencies. The reactor inspection program is an integral part, along with performance indicators (Pis), assessment, and enforcement, of the ROP. Acceptable performance in the cornerstones, as measured by the Pis and the risk-informed baseline inspection program, is indicative of overall licensee performance that provides for adequate protection of public health and safety. As noted in the PNP mid-cycle assessment letter dated September 1, 2015 (Reference 27), the NRC determined the performance at PNP during the most recent quarter was within the Regulatory Response Column of the NRC's ROP Action Matrix which indicates that the cornerstone objectives have been met with minimal degradation in safety performance. Additional information on the recent PNP performance summary can be found at http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/PALl/pali_chart.html.

4.8 Public Comment No. 11 out of a Total of 23 Comments BN comment:

On May 25, 2012, 25-30 of us met with NRC Chairman Jaczko and many other NRC staff, including from Region 3 (including Regional Administrator Chuck Casto). Embrittlement of the RPV and PTS risks were a primary subject matter discussed during the meeting. Dr. Barbara J. Pellegrini was one of the concerned local residents who attended. On May 30, she wrote NRC Chairman Jaczko her ideas for how Palisades' RPV embrittlement could be measured, and PTS risks defended against, including consultation with many experts in the field of materials science. NRC never responded to her letter. Why not?

NRC Response:

The NRC responded to Dr. Pellegrini's letter dated May 30, 2012 by letter dated April 24, 2013.

However, the NRC notes that the response to Dr. Pellegrini was not properly made publicly available. The NRC staff has taken the appropriate steps to properly docket the letter from Dr.

Brian Sheron dated April 24, 2013, which responded to Dr. Pellegrini's concerns. This response can now be found at ADAMS Accession Nos. ML13105A006 and ML13099A293 (Reference 28).

4.9 Public Comment No. 21 out of a Total of 23 Comments BN Comment:

Isn't yet another INCREASE in reactor disaster risk at Palisades the sheer 42 year old age?

NRC Response:

BN is correct to note that the age of the RPV at PNP influences how em brittled the vessel has become and, thereby, how close the vessel is to the NRC's embrittlement screening criteria,

such as that expressed in 10 CFR 50.61 a. The NRC notes that the age of the RPV is fully accounted for in the 10 CFR 50.61 a requirement to assess fluence which is used to calculate the RTMAx-x values. As discussed in SE section 3.2, the licensee has satisfied the NRC's requirements for determining fluence and, in doing so, has properly accounted for the age of the PNP vessel in its amendment request.

In addition, PNP was issued, consistent with NRC regulation, a renewed license on January 17, 2007 (Reference 29). During license renewal, the staff evaluated the licensee's aging management programs for passive and long-lived structures and components. As documented in NUREG-1871, (Reference 30), the NRC concluded that the licensee has demonstrated that the aging effects will be adequately managed so that the intended functions of a structure or component will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21 (a)(3). The NRC requires licensees to test, monitor, and inspect the condition of safety equipment and to maintain that equipment in reliable operating condition over the life of the plant. The NRC also requires licensees to continuously correct deficiencies that could affect plant safety (e.g., degraded or failed components caused by aging or operation events). Over the years, the licensee has replaced and/or overhauled plant equipment when needed. The testing, monitoring, inspection, maintenance, and replacement of plant equipment provides reasonable assurance that this equipment will perform its intended safety functions during the life of the plant, regardless of its age. This conclusion applies regardless of whether the licensee is approved to implement the 10 CFR 50.61 a screening criteria.

4.10 Miscellaneous Issues (Public Comment Nos. 1. 10. 12. 13. 14. 20. 22 out of a Total of 23 Comments)

BN Comments:

Comments were received pertaining to the following licensee's actions/operation:

• Licensee's violation of PTS safety standards in 1981.

• Licensee response to a design basis accident.

• Licensee's financial qualifications and interests.

• Licensee's staffing.

• Licensee's choice to not anneal the RPV.

• Licensee's safety culture.

NRC response:

The issues raised pertain to PNP general operation, and are not within the scope of the NRC staff's NSHC determination nor are they related to the proposed amendment. As such, no specific response is given herein.

The staff notes that BN's comment Nos. 1 and 1O are addressed in a public meeting webinar summary dated April 18, 2013 (Reference 31 ). With regard to the comments that do not pertain to the proposed amendment, the NRC staff notes that to the extent that BN believes that PNP is operating unsafely, it can request that the Commission institute a proceeding pursuant to 10 CFR 2.206(a) to modify, suspend, or revoke a license, or for any other action as may be proper.

4.11 Miscellaneous Issues (Public Comment Nos. 7. 8. 9. 15. 23 out of a Total of 23 Comments)

BN Comments:

Comments were received pertaining to the following:

• Hypothetical accident conditions at PNP.

• NRC's response to the March 29, 1982, op-ed in New York Times by Demetrios Basdekas.

• Issues related to the PNP license renewal proceedings.

• The previously approved PNP power uprate.

• Pressure adding to PTS risks.

NRC response:

This license amendment would allow the licensee to employ an alternative methodology to assess how well its reactor vessel materials respond to a PTS event. The issues raised above are not within the scope of the NRC staff's NSHC determination nor are they related to the proposed amendment. As such, no specific response is given herein.

The staff notes that BN's comment Nos. 7 and 15 are addressed in a public meeting webinar summary dated April 18, 2013. With regard to the comments that do not pertain to the proposed amendment, the NRC staff notes that to the extent that BN believes that PNP is operating unsafely, it can request that the Commission institute a proceeding pursuant to 10 CFR 2.206(a) to modify, suspend, or revoke a license, or for any other action as may be proper. In addition, to the extent the BN believes the existing PTS rules are deficient, it can petition the Commission to issue, amend or rescind any regulation under to 10 CFR 2.802, "Petition for Rulemaking."

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, "Part 20-Standards for Protection against Radiation." The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration (79 FR 58814, September 30, 2014). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, on the basis of the considerations discussed above, that (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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

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ML15209A791 *via memo **via email OFFICE DORL/LPL3-1 /PM DORL/LPL3-1 /LA DE/EVIB/BC DSS/SRXB/BC* Tech Editor**

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