L-13-223, License Amendment Request to Implement 10 CFR 50.61a. Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.

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License Amendment Request to Implement 10 CFR 50.61a. Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
ML13212A027
Person / Time
Site: Beaver Valley
Issue date: 07/30/2013
From: Emily Larson
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-13-223
Download: ML13212A027 (48)


Text

Beaver Valley Power Station P.O. Box 4 FirstEnergy Nuclear Operating Company Shippingporl, PA 15077 Eric A. Larson 724-682-5234 Site Vice President Fax: 724-643-8069 July 30, 2013 L-13-223 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 1 Docket Number 50-334, License Number DPR-66 License Amendment Request to Implement 10 CFR 50.61 a. "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" Pursuant to 10 CFR 50.61 a(c) and 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) hereby submits an amendment application for the Beaver Valley Power Station, Unit No.1 operating license. The proposed amendment would authorize the implementation of 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."

An evaluation of the proposed change is enclosed. To allow for normal NRC processing, FENOC requests approval of the proposed license amendment by July 31, 2014. Also, an implementation period of 120 days following the effective date of the amendment is requested.

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at (330) 315-6810.

Beaver Valley Power Station, Unit No. 1 L-13-223 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on July__&, 2013.

Sincerely, r: at~~-

Eric A. Larson

Enclosure:

FENOC Evaluation of the Proposed Amendment cc: Director, Office of Nuclear Reactor Regulation (NRR)

NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

FENOC Evaluation of the Proposed Amendment

Subject:

License Amendment Request for Approval to Implement 10 CFR 50.61 a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" Table of Contents 1.0

SUMMARY

DESCRIPTION ................................................................................. 2 2.0 DETAILED DESCRIPTION .................................................................................. 2

3.0 TECHNICAL EVALUATION

................................................................................. 5 3.1 Alternate Pressurized Thermal Shock Rule .................................................. 5 3.2 Method Discussion ........................................................................................ 7 3.3 Plant Specific Material Properties ............................................................... 18 3.4 Neutron Fluence Values ........................................... ~ .................................. 21 3.5 Surveillance Capsule Data .......................................................................... 23

3. 6 Inservice Inspection Data ............................................................................ 26 3.7 Determination of RTMAX-X Values for All Beltline Region Materials .......... 27 3.8 Evaluation of Extended Beltline Materials ................................................... 34 3.9 Conclusion .................................................................................................. 37

4.0 REGULATORY EVALUATION

........................................................................... 37 4.1 Significant Hazards Consideration .............................................................. 37 4.2 Applicable Regulatory Requirements I Criteria ............................................ 39 4.3 Conclusions ................................................................................................. 40

5.0 ENVIRONMENTAL CONSIDERATION

.............................................................. 40

6.0 REFERENCES

................................................................................................... 41 Attachments:

1. Proposed Facility Operating License Change (Mark-Up)
2. Proposed Facility Operating License Change (Re-Typed)

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 2 of 42 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend the Beaver Valley Power Station, Unit No. 1 (BVPS-1) Operating License No. DPR-66. The proposed amendment would authorize implementation of 10 CFR 50.61a, "Alternate fracture toughness requirements for protection against pressurized thermal shock [PTS] events," (alternate PTS rule).

Proposed text for a new license condition to be added to the existing Facility Operating License (FOL) is included in this license amendment request.

BVPS-1 currently complies with 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," which establishes screening criteria below which the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The 10 CFR 50.61 screening criteria define a limiting level of embrittlement beyond which plant operation cannot continue without further evaluation. 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 14) the screening criteria in the PTS rule is overly conservative and the risk of through wall cracking due to a PTS event is much lower than previously estimated.

As such, the specified screening limits and associated compensatory actions may

  • impose an unnecessary burden on licensees whose pressurized water reactor (PWR) vessel is projected to exceed the PTS rule screening criteria.

The alternate PTS rule, which was included in the Federal Register with an effective date of February 3, 2010, provides fracture toughness requirements for protection against PTS events for PWR pressure vessels that are less burdensome than the requirements of the PTS rule.

BVPS-1 is expected to exceed the screening criteria of the PTS rule prior to its extended license expiration (2036). Compliance with the alternate PTS rule as an alternative to the requirements of the PTS rule would support the BVPS-1 license extension and reduce regulatory burden while maintaining adequate protection to public health and safety.

2.0 DETAILED DESCRIPTION FENOC proposes to implement the requirements of the alternate PTS rule in lieu of the current requirements of the PTS rule and to add a new license condition. Proposed text for a new license condition to be added to the existing FOL, Appendix C, "Additional Conditions Operating License No. DPR-66," that authorizes the implementation of the alternate PTS rule is included in the attachments.

During plant operation, the walls of reactor pressure vessels (RPVs) are exposed to neutron radiation, resulting in localized embrittlement of the vessel steel and weld materials in the core area. If an embrittled RPV had a flaw of critical size and certain severe system transients were to occur, the flaw could propagate through the vessel,

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 3 of 42 resulting in a through-wall crack. The severe transients of concern are known as pressurized thermal shock events. PTS events in PWRs are caused by severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel.

As summarized in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)," in the early 1980s, the nuclear industry and the Nuclear Regulatory Commission (NRC) staff performed a number of investigations to assess the risk of vessel failure posed by PTS and to establish the operational limits needed to ensure that the likelihood of RPV failures caused by PTS transients is maintained at an acceptably low level. These efforts led to the development of the PTS rule. The nil ductility (fracture toughness) transition reference temperature (RTNor) of the reactor vessel material increases as a result of irradiation throughout the operational life of the vessel. The PTS rule establishes screening criteria (or maximum values of RTNDT permitted during the operating life of the plant) of 270 degrees Fahrenheit (°F) for axial welds, plates, and forgings, and 300 oF for circumferential welds. The reference temperature value RTNDT evaluated for the end-of-life (EOL) fluence for each of the vessel beltline materials, using the procedures in paragraph (c) of the PTS rule, is referred to as (RTPrs).

The PTS rule requires licensees to take compensatory actions when the value of RT Prs for any material in the beltline is projected to exceed the PTS screening criterion using the plant's projected EOL fluence. It requires the licensee to implement flux reduction programs that are reasonably practical to avoid exceeding the PTS screening criteria. If a licensee has no reasonably practical flux reduction program that will prevent RT Prs from exceeding the PTS screening criteria using the EOL fluence, the licensee is required to submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of the postulated PTS events if continued operation beyond the screening criteria is allowed. Reactor vessel annealing may also be implemented by a licensee to prevent exceeding the screening criteria.

During the Beaver Valley Power Station license renewal process, FENOC confirmed by letter dated April 2, 2008 that BVPS-1 was expected to exceed the 270°F screening limit of the PTS rule prior to the EOL extension. As a result, FENOC committed that prior to exceeding the PTS screening criteria for BVPS-1, a flux reduction measure to manage PTS in accordance with the requirements of the PTS rule would be selected and submitted to the NRC for review and approval. This commitment is contained in the BVPS-1 Updated Final Safety Analysis Report Table 16-1, "Unit 1 License Renewal Commitments," as well as in Appendix A of the NRC, "Safety Evaluation Report Related to the License Renewal of Beaver Valley Power Station, Units 1 and 2 (NUREG-1929),"

dated October 2009.

Advancements in understanding and knowledge of materials behavior, ability to realistically model plant systems and operational characteristics, and to better evaluate

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 4 of 42 PTS transients to estimate loads on vessel walls have shown that earlier analyses, performed some 20 years ago as part of the development of the PTS rule and screening criteria, were overly conservative.

In 1999, the NRC undertook a project to develop a technical basis to support a risk-informed revision of the existing PTS rule. Realistic input values and models and an explicit treatment of uncertainties were used to develop the alternate PTS rule, which was approved by the NRC and included in the Federal Register with an effective date of February 3, 2010. In order to implement the alternate PTS rule, a licensee must submit a request for approval in the form of an application for a license amendment request in accordance with 10 CFR 50.90 and include documentation required by alternate PTS rule paragraphs (c)(1), (2), and (3).

The BVPS-1 PTS EOL extension analysis is documented in Westinghouse Report WCAP-15571 Supplement 1, Revision 2, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," dated September 2011 (Reference 5). The projected RTprs values for EOL extension at BVPS-1 meet the PTS rule screening criteria for beltline and extended beltline materials with the exception of lower shell plate B6903-1 (heat C6317-1). Plate B6903-1 has a 50 effective full power year (EFPY) RTPTS value of 277* °F, which exceeds the PTS rule screening criteria of 270 °F. Based on the fluence information provided in WCAP-15571 Supplement 1, Revision 2, the PTS rule screening limit is expected to be reached at 39.6 EFPY.

Paragraph (a) of the alternate PTS rule defines the reference temperature RTMAX-X* The reference temperature RTMAX-X means any or all of the reactor vessel material properties that characterize the resistance to fracture initiation from flaws found along axial weld fusion lines (RTMAX-Aw), in plates (in regions not associated with welds) (RTMAX-PL), in forgings (in regions not associated with welds) (RTMAX-FO), along circumferential weld fusion lines (RTMAX-cw), or the sum of RTMAX-AW and RTMAX-PL.

This license amendment request documents the basis for implementing the requirements of the alternate PTS rule up to the end of the 60-year operating license.

The evaluation concludes the following:

1) The BVPS-1 reactor vessel beltline materials have end-of-license (EOL, 50 EFPY) RTMAX-X values below the alternate PTS rule screening criteria.
2) The surveillance data for the vessel base metal did not pass the surveillance data statistical tests. However, as permitted by the alternate PTS rule, an adjustment was made to the calculation of the RTMAX-X values for this material and incorporated into the analysis. The adjustment did not change the limiting RTMAX-X values. The adjustment is discussed further in Section 3. 7.2.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 5 of 42

3) The criteria for reactor vessel beltline weld flaw density and size distribution did not require evaluation based on the latest BVPS-1 inspection results. This is acceptable per the alternate PTS rule criteria. Analysis of inspection results are discussed further in Section 3.7.3.

FENOC's implementation of the alternate PTS rule in lieu of the PTS rule would provide new screening criteria for PTS, resulting in a burden reduction while continuing to provide adequate protection to public health and safety.

3.0 TECHNICAL EVALUATION

This section documents evaluations performed for the BVPS-1 reactor vessel to meet the requirements of the alternate PTS rule. Section 3.1 discusses the alternate PTS rule and its requirements. Section 3.2 provides the methodology for calculating RTMAX-X and performing the examination and flaw assessment required by the alternate PTS rule. Sections 3.3 through 3.6 provide inputs necessary to conduct the alternate PTS rule evaluations described in Section 3.2. Specifically, these sections provide the material properties, neutron fluence values, surveillance capsule analysis results, and inservice inspection data of the reactor vessel beltline materials. The results of the RTMAX-X calculations and flaw assessment are presented in Section 3.7. An evaluation of extended beltline materials is provided in Section 3.8. The conclusion and references for the PTS evaluation follow in Sections 3.9 and 6.0, respectively.

3.1 Alternate Pressurized Thermal Shock Rule The alternate PTS rule primary requirements consist of the following:

  • Each licensee shall have projected values of RTMAX-X for each reactor vessel beltline material for the EOL fluence of the material. The assessment of RTMAX-X values must use the calculation procedures described in Section 3.2.1 of this evaluation that are equivalent expressions of those calculations prescribed by the alternate PTS rule. The assessment must specify the bases for the projected value of RTMAX-x for each reactor vessel beltline material, including the assumptions regarding future plant operation (for example, core loading patterns, projected capacity factors); the copper (Cu), phosphorus (P), manganese (Mn), and nickel (Ni) contents; the reactor cold leg temperature (Tc); and the neutron flux and fluence values used in the calculation for each beltline material.
  • Each licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the criteria described in 10 CFR 50.61 a (f)(6)(i)(A) and (f)(6)(i)(B).
  • Each licensee shall perform an examination and an assessment of flaws in the reactor vessel beltline as described in Section 3.2.3 of this evaluation. The licensee shall verify that the requirements described in Section 3.2.3 have been met.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 6 of 42

  • Each licensee shall compare the projected RTMAX-X values for plates, forgings, axial welds, and circumferential welds to the PTS screening criteria in Table 3.2-2 of this evaluation, for the purpose of evaluating a reactor vessel's susceptibility to fracture due to a PTS event.
  • If any of the projected RTMAX-x values are greater than the PTS screening criteria in Table 3.2-2, then the licensee may propose the compensatory actions or plant-specific analyses as required in the alternate PTS rule paragraphs (d)(3) through (d)(?), as applicable, to justify operation beyond the PTS screening criteria in Table 3.2-2. The licensee shall implement those flux reduction programs that are reasonably practicable to avoid exceeding the PTS screening criteria. If this analysis indicates that no reasonably practicable flux reduction program will prevent the RTMAX-X value for one or more of the reactor vessel beltline materials from exceeding the PTS screening criteria, then the licensee shall perform a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent the potential for an unacceptably high probability of failure of the reactor vessel as a result of postulated PTS events. In the analysis, the licensee may determine the properties of the reactor vessel materials based on available information, research results and plant surveillance data, and may use probabilistic' fracture mechanics techniques.

Two alternate PTS rule subsequent requirements consist of the following:

  • Whenever there is a significant change in projected values of RTMAX-x, so that the previous value, the current value, or both values, exceed the screening criteria before the expiration of the plant operating license; or upon the licensee's request for a change in the expiration date for operation of the facility; a re-assessment of RTMAX-x values must be conducted. If the surveillance data used to perform there-assessment of RTMAX-X values meet the requirements discussed in alternate PTS rule paragraphs (f)(6)(v) or (f)(6)(vi), the data must be analyzed in accordance with the alternate PTS rule and the RTMAX-X values must be recalculated and resubmitted for approval.
  • The licensee shall verify that the requirements of alternate PTS rule paragraphs (e),

(e)(1), (e)(2), and (e)(3) have been met. The licensee must submit, within 120 days after completing a volumetric examination of reactor vessel beltline materials as required by American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, the adjustments made to the volumetric test data to account for NDE-related uncertainties as described in paragraph (e)(1) and all information required by paragraph (e)(1)(iii) for review and approval. If a licensee is required to implement paragraphs (e)(4), (e)(5), and (e)(6) of the alternate PTS rule, the information required in these paragraphs must be submitted within one year after completing a volumetric examination of reactor vessel materials as required by ASME Code,Section XI.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 7 of 42 3.2 Method Discussion This section describes the methodology for calculating RTMAX-X and is derived from the alternate PTS rule requirements.

3.2.1 Calculation of RTMAX-X Values In accordance with paragraph (f) of the alternate PTS rule, each licensee shall calculate RTMAX-X values for each reactor vessel beltline material using the fast neutron fluence

(<pt). The values of RTMAX-Aw, RTMAX-PL. RTMAX-Fo, and RTMAX-cw must be determined using equations 1 through 4 (Reference 1). Reference 2 provides additional information on these equations, which is included below. RTMAX-X values are calculated in °F as follows:

RT _ = MAX nAWFL [

MAX .

{(RTadj-aw(i)

NDT(u)

+ /1Tadj-aw(i)(,l,t 30

)\}]

'I' FL }

(1)

MAX AW i=l AWFL(t) (RTadj-p/(i) + /1Tadj-p/(i)(,l,t ))

  • NDT(u) 30 * 'I' FL Where:

nAwFL is the number of axial weld fusion lines in the beltline region of the vessel, is a counter that ranges from 1 to nAwFL,

¢t FL is the maximum fluence occurring on the vessel ID along a particular axial weld fusion line, RT:J%;c~;<i) is the unirradiated RT NOT of the weld adjacent to the i1h axial weld fusion line, RT:J%;c~~<i) is the unirradiated RT NOT of the plate adjacent to the i1h axial weld fusion line, 11~~dJ*-aw(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢tFL of the weld adjacent to the i1h axial weld fusion line, and 11~~dj-pt(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation due to ¢tFL of the plate adjacent to the i1h axial weld fusion line.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 8 of 42 nPL RTMAX-PL = MAX[RTPL(i) + 11TPL(i)(J.tPL(i))ll (2) i~l NDT(u) 30 'f MAX ~

Where:

nPL is the number of plates in the beltline region of the vessel, is a counter that ranges from 1 to npL,

¢t:riJJ is the maximum fluence occurring over the vesseiiD occupied by a particular plate, RT:~n) is the unirradiated RT NoT of a particular plate, and 11T3~L(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢t:f})) of a particular plate.

_ FO(i) FO(i) FO(i) )~

RTMAX-FO =MAX nFo [

RTNDT(u) + 11T30

(

¢tMAX (3) i~l Where:

nFO is the number of forgings in the beltline region of the vessel, is a counter that ranges from 1 to nFO,

¢t;~~) is the maximum fluence occurring over the vessel ID occupied by a particular forging, RT/:f}ig,) is the unirradiated RT NOT of a particular forging, and

/1T3~o(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢t;~~) of a particular forging.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 9 of 42 RT,acfi-cw(i) + /1Tadj-cw(i)(A.t )\)

'!' FL p .

l(

NDT(u) 30 ncwFL

_ acfj-pl(i) acfj-p/(i)

RTMAX-CW = Mb-X MAXCWFL(i) (RTNDT(u) + /11;0 (¢tFJ1 (4)

(RTacfj-fo(i)

NDT(u)

+ /1Tacfj-30 fo(i) ("'t ))

'!' FL Where:

ncwFL is the number of circumferential weld fusion lines in the beltline region of the vessel, is a counter that ranges from 1 to ncwFL,

¢tFL is the maximum fluence occurring on the vessel ID along a particular circumferential weld fusion line, RT;%;<~~Ul is the unirradiated RT NOT of the weld adjacent to the i1h circumferential weld fusion line, RT;%;tz:?l is the unirradiated RT NoT of the plate adjacent to the i1h circumferential weld fusion line (if there is no adjacent plate this term is ignored),

RT;%;(,~)Ul is the unirradiated RTNOT of the forging adjacent to the i1h circumferential weld fusion line (if there is no adjacent forging this term is ignored),

!1T3~cfi-cw(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢tFL of the weld adjacent to the i1h circumferential weld fusion line,

!1T3~dJ-pt(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢tFL of the plate adjacent to the i1h axial weld fusion line (if there is no adjacent plate this term is ignored), and

!1T3~dJ-fo(i) is the shift in the Charpy V-Notch 30-foot-pound energy produced by irradiation to ¢tFL of the forging adjacent to the i1h axial weld fusion line (if there is no adjacent forging this term is ignored).

The values of L1T3o must be determined using Equations 5, 6 and 7, for each axial weld, plate, forging, and circumferential weld. The L1T3o value for each axial weld calculated as specified by Equation 1 must be calculated for the maximum fluence (<ptFL) occurring along a particular axial weld at the clad-to-base metal interface. The L1T3o value for each adjacent plate calculated as specified by Equation 1 must also be calculated using

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 10 of 42 the same value of <ptFL used for the axial weld. The .6.T3o value for each plate or forging calculated as specified by Equations 2 and 3 must be calculated for the maximum fluence (<ptMAX) occurring at the clad-to-base metal interface over the entire area of each plate or forging. In Equation 4, the fluence (<ptFL) value used for calculating the circumferential weld .6. T 30 value is the maximum fluence occurring along the circumferential weld at the clad-to-base metal interface. The .6.T30 values in Equation 4 shall also be calculated for the adjoining plates or forgings using the same maximum circumferential weld fluence. If the conditions specified in alternate PTS rule paragraph (f)(6)(v) are not met, licensees must propose .6.T30 and RTMAX-X values in accordance with paragraph (f)(6)(vi) of the alternate PTS rule.

The equation used to calculate the 11 T3o shift is displayed below:

!1T3o =MD+ CRP (5)

Where:

MD= A(1-0.001718T c)(1 +6.13PMn 2.4 71 ) <pte0*5 (6)

CRP= 8(1 +3.77Ni 1*191 ) f(Cue,P)g(Cue,Ni, <pte) (7)

A= 1.140x1.0-7 for forgings 1.561x1 o-7 for plates 1.417x1 o-7 for welds B= 102.3 for forgings 102.5 for plates in non-CE manufactured vessels 135.2 for plates in CE manufactured vessels 155.0 for welds 2

<pte= <pt for <p ;;:: 4.39x1 0 10 n/cm /sec 10

<pt(4.39x1 0 / <p)

  • 0 2595 for <p < 4.39 x 10 10 n/cm 2/sec f(Cue,P)= 0 for Cu :::; 0.072

[Cue- 0.072] 0*668 for Cu > 0.072 and P :5 0.008

[Cue - 0.072+1.359(P-0.008)] 0*668 for Cu > 0.072 and p > 0.008 Cue= 0 for Cu:::; 0.072 MIN (Cu, Maximum Cue) for Cu > 0.072 Max. Cue= 0.243 for Linde 80 welds 0.301 for all other materials g(Cue,Ni,

<pte)=

1 1

-+-ta nh[ log 10 (qyte) + 1.1390Cue- 0.448Ni -18.120 J 2 2 0.629 Tc = Cold leg temperature under normal full power operating conditions (°F) as a time weighted average 2

<p= Average neutron flux (n/cm /sec)

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 11 of 42 t Time that the reactor has been in full power operation (sec) cpt Neutron fluence (n/cm 2)

P= Phosphorous content (weight percent [wt%])

Ni= Nickel content (wt%)

Cu= Copper content (wt%)

Mn= Manganese content (wt%)

The values of Cu, Mn, P, and Ni in Equations 6 and 7 must represent the best estimate values for the material. For a plate or forging, the best estimate value is normally the mean of the measured values for that plate or forging. For a weld, the best estimate value is normally the mean of the measured values for a weld deposit made using the same weld wire heat number as the critical vessel weld. If these values are not available, either the upper limiting values given in the material specifications to which the vessel material was fabricated, or conservative estimates (that is, mean plus one standard deviation) based on generic data as shown in Table 3.2-1 for P and Mn, must be used. *

  • Table 3.2-1 Conservative Estimates for Chemical Element Weight Percentages Materials p Mn Plates 0.014 1.45 Forgings 0.016 1.11 Welds 0.019 1.63 The values of RT NDT(U) must be evaluated according to the procedures in the ASME Code, Section Ill, paragraph NB-2331. If any other method is used for this evaluation, the licensee shall submit the proposed method for review and approval by the Director of the Office of Nuclear Reactor Regulation along with the calculation of RTMAX-X values.
  • If a measured value of RTNDT(U) is not available, a generic mean value of RTNDT(U) for the class of material must be used if there are sufficient test results to establish a mean.
  • The following generic mean values of RT NDT(U) must be used unless justification for different values is provided: 0 oF for welds made with Linde 80 weld flux; and -56 oF for welds made with Linde 0091, 1092, and 124 and ARCOS B-5 weld fluxes.

The value ofTc in Equation 6 of this section must represent the time-weighted average of the reactor cold leg temperature under normal operating full power conditions from

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 12 of 42 the beginning of full power operation through the end of licensed operation. For the surveillance capsule statistical tests, T c is a time-weighted average from the beginning of full power operation up to the time of capsule withdrawal.

If any of the calculated RTMAX-X values for BVPS-1 are greater than the PTS screening criteria, defined in Table 3.2-2, further evaluation or action, consistent with paragraphs (d)(3) through (d)(?) of the alternate PTS rule, is required.

Table 3.2-2 PTS Screening Criteria RTMAX-x limits [°F] for different vessel wall 1

Product form and RTMAX-x thicknesses (TwALL values 9.5 in. < TWALL 10.5 in. < TwALL TwALL::;; 9.5 in.

10.5 in.
;; 11.5 in.

Axial Weld-RTMAX-Aw 269 230 222 Plate-RTMAX-PL 356 305 293 2

Forging without underclad cracks-RTMAX-Fo 356 305 293 Axial Weld and Plate-RTMAX-Aw + RTMAX-PL 538 476 445 3

Circumferential Weld-RTMAx-cw 312 277 269 4

Forging with underclad cracks-RT MAX Fo 246 241 239 Note:

1. Wall thickness is the beltline wall thickness including the clad thickness in inches (in.).
2. Forgings without underclad cracks apply to forgings for which no underclad cracks have been detected and that were fabricated in accordance with Regulatory Guide 1.43.
3. RT PTS limits contribute 1 x 10-a per reactor year to the reactor vessel TWCF.
4. Forgings with underclad cracks apply to forgings that have detected underclad cracking or were not fabricated in accordance with Regulatory Guide 1.43.

3.2.2 Surveillance Capsule Data Statistical Checks As a condition of the alternate PTS rule, the licensee must consider plant-specific information that could affect the use of this equation for the determination of a material's L1T 30 value. In order to make this determination, the alternate PTS rule provides requirements for evaluation of surveillance capsule data in paragraphs (f)(6)(i), (f)(6)(ii),

(f)(6)(iii), and (f)(6)(iv). The requirements consist of a mean deviation test, a slope deviation test, and an outlier deviation test.

Specifically, the alternate PTS rule states that the licensee shall verify that an appropriate RTMAX-X value has been calculated for each reactor vessel beltline material by considering plant-specific information that could affect the use of the model (that is, Equations 5, 6 and 7) for the determination of a material's L1 T3o value.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 13 of 42 The licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the following criteria

  • The surveillance material must be a heat-specific match for one or more of the materials for which RTMAX-X is being calculated. The 30-foot-pound transition temperature must be determined as specified by the requirements of 10 CFR 50, Appendix H.
  • If three or more surveillance data points measured at three or more different neutron fluences exist for a specific material, the licensee shall determine if the surveillance data show a significantly different trend than the embrittlement model predicts. If fewer than three surveillance data points exist for a specific material, then the embrittlement model must be used without performing the consistency check.

The licensee shall estimate the mean deviation from the embrittlement model for the specific data set (that is, a group of surveillance data points representative of a given material). The mean deviation from the embrittlement model for a given data set must be calculated using Equations 8 and 9. The mean deviation for the data set must be compared to the maximum heat-average residual given in Table 3.2-3 or derived using Equation 10. The maximum heat-average residual is based on the material group into which the surveillance material falls and the number of surveillance data points. For surveillance data sets with greater than 8 data points, the maximum credible heat-average residual must be calculated using Equation 10. The value of a used in Equation 10 must be obtained from Table 3.2-3.

Residual (r) =Measured Ll T3o- Predicted Ll T3o (8)

II Mean deviation for a data set of n data points = (1 In) x I ri (9) i=l Maximum credible heat average residual= 2.33a/n°* 5 (1 0)

Where:

n= number of surveillance data points (sample size) in the specific data set a = standard deviation of the residuals about the model for a relevant material group given in Table 3.2-3.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 14 of 42 Table 3.2-3 Maximum Heat-Average Residual [°F] for Relevant Material Groups by Number of Available Data Points (Significance Level= 1%)

Number of available data points Material group a [°F]

3 4 5 6 7 8 Welds, for Cu > 0.072 26.4 35.5 30.8 27.5 25.1 23.2 21.7 Plates, for Cu > 0.072 21.2 28.5 24.7 22.1 20.2 18.7 17.5 Forgings, for Cu > 0.072 19.6 26.4 22.8 20.4 18.6 17.3 16.1 Weld, Plate or Forging, for Cu :s: 0.072 18.6 25.0 21.7 19.4 17.7 16.4 15.3 The licensee shall estimate the slope of the embrittlement model residuals (estimated using Equation 8) plotted as a function of the base 10 logarithm of neutron fluence for the specific data set. The licensee shall estimate the T-statistic for this slope (TsuRv) using Equation 11 and compare this value to the maximum permissible T-statistic (TMAX) in Table 3.2-4. For surveillance data sets with greater than 15 data points, the T MAX value must be calculated using Student's T distribution with a significance level (a) of 1 percent for a one-tailed test.

~ = m SURV (se(m)) (11)

Where:

m= the slope of a plot of all of the rvalues (estimated using Equation 8) versus the base 10 logarithm of the neutron fluence for each r value. The slope shall be estimated using the method of least squares.

se(m) = the least-squares estimate of the standard-error associated with the estimated slope value m.

Table 3.2-4 TMAX Values for the Slope Deviation Test (Significance Level= 1%)

Number of available data points (n) TMAX 3 31.82 4 6.96 5 4.54 6 3.75 7 3.36 8 3.14 9 3.00 10 2.90 11 2.82

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 15 of 42 Table 3.2-4 TMAX Values for the Slope Deviation Test (Significance Level = 1%)

Number of available data points (n) TMAX 12 2.76 13 2.72 14 2.68 15 2.65 The licensee shall estimate the two largest positive deviations (outliers) from the embrittlement model for the specific data set using Equations 8 and 12. The licensee shall compare the largest normalized residual (r *) to the appropriate allowable value from the third column in Table 3.2-5 and the second largest normalized residual to the appropriate allowable value from the second column in Table 3.2-5.

r r*=- (12)

(J" Where r is defined using Equation 8 and a is given in Table 3.2-3.

Table 3.2-5 Threshold Values for the Outlier Deviation Test (Significance Level = 1%)

Number of available data Second largest allowable Largest allowable normalized points (n) normalized residual value (r*) residual value (r*)

3 1.55 2.71 4 1.73 2.81 5 1.84 2.88 6 1.93 2.93 7 2.00 2.98 8 2.05 3.02 9 2.11 3.06 10 2.16 3.09 11 2.19 3.12 12 2.23 3.14 13 2.26 3.17 14 2.29 3.19 15 2.32 3.21 The L1 T 30 value must be determined using Equations 5, 6, and 7 if all three of the following criteria are satisfied:

  • The mean deviation from the embrittlement model for the data set is equal to or less than the value in Table 3.2-3 or the value derived using Equation 10 of this section;

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 16 of 42

  • The T-statistic for the slope (TsuRv) estimated using Equation 11 is equal to or less than the maximum permissible T-statistic (TMAX) in Table 3.2-4; and
  • The largest normalized residual value is equal to or less than the appropriate allowable value from the third column in Table 3.2-5 and the second largest normalized residual value is equal to or less than the appropriate allowable value from the second column in Table 3.2-5.

If any of these criteria are not satisfied, the licensee shall review the data base for that heat in detail, including all parameters used in Equations 5, 6, and 7 of this section and the data used to determine the baseline Charpy V-notch curve for the material in an unirradiated condition. The licensee shall propose ~ T 30 and RTMAX-x values, considering their plant-specific surveillance data, to be used for evaluation relative to the acceptance criteria of this rule.

3.2.3 Reactor Vessel Beltline lnservice Inspection Data Evaluation The licensee must have performed an examination of the reactor vessel beltline welds using procedures, equipment, and personnel that have been qualified under the ASME Code Section XI, Appendix VIII, Supplement 4 and Supplement 6, as specified in 10 CFR 50.55a(b)(2)(xv). The licensee shall verify that the flaw density and size distributions within the volume described in ASME Code,Section XI, Figures IWB-2500-1 and IWB-2500-2 and limited to a depth from the clad-to-base metal interface of 1-inch or 10 percent of the vessel thickness, whichever is greater, do not exceed the limits in Tables 3.2-6 and 3.2-7 based on the test results from the volumetric examination. For BVPS-1, the clad-to-base metal interface of 1-inch is greater than 10 percent of the vessel thickness. The verification of the flaw density and size distributions shall be performed line-by-line for Tables 3.2-6 and 3.2-7.

Table 3.2-6 Allowable Number of Flaws in Welds Through-Wall Extent (TWE) of Flaw (inches) Maximum number of flaws per 1, 000 inches of weld length in the inspection volume that are greater than or equal to TWEMIN and less than TWEMAX TWEMIN TWEMAX 0 0.075 No Limit 0.075 0.475 166.70 0.125 0.475 90.80 0.175 0.475 22.82 0.225 0.475 8.66 0.275 0.475 4.01 0.325 0.475 3.01 0.375 0.475 1.49 0.425 0.475 1.00 0.475 Infinite 0.00

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 17 of 42 The licensee shall determine the allowable number of weld flaws in the reactor vessel beltline by multiplying the values in Table 3.2-6 by the total length of the reactor vessel beltline welds that were volumetrically inspected and dividing by 1,000 inches of weld length.

Table 3.2-7 Allowable Number of Flaws in Plates or Forgings Through-Wall Extent (TWE) of Flaw (inches) Maximum number of flaws per 1,000 square inches of inside surface area in the inspection volume that are greater than or equal to TWEMIN and less than TWEMAX TWEMIN TWEMAX 0 0.075 No Limit 0.075 0.375 8.05 0.125 0.375 3.15 0.175 0.375 0.85 0.225 0.375 0.29 0.275 0.375 0.08 0.325 0.375 0.01 0.375 Infinite 0.00 The licensee shall determine the allowable number of plate or forging flaws in their reactor vessel beltline by multiplying the values in Table 3.2-7 by the total surface area of the reactor vessel beltline plates or forgings that were volumetrically inspected and dividing by 1000 square inches.

For each flaw detected within the inner 1-inch of the inspection volume, measured from the clad-to-base metal interface, with a through-wall extent equal to or greater than 0.075 inches, the licensee shall document the dimensions of the flaw, including through-wall extent and length, whether the flaw is axial or circumferential in orientation and its location within the reactor vessel, including its azimuthal and axial positions and its depth embedded from the clad-to-base metal interface.

The licensee shall verify that axially oriented flaws located at the clad-to-base metal interface do not open to the vessel inside surface using surface or visual examination techniques capable of detecting and characterizing service induced cracking of the reactor vessel cladding. The licensee shall verify that all flaws between the clad-to-base metal interface and three-eighths of the reactor vessel thickness from the interior surface are within the allowable values in ASME Code,Section XI, Table IWB-351 0-1.

The licensee shall perform analyses to demonstrate that the reactor vessel will have a through-wall cracking frequency (TWCF) of' less than 1 x 1o-6 per reactor year if the ASME Code,Section XI volumetric examination indicates any of the following:

  • The flaw density and size in the inspection volume exceed the limits in Tables 3.2-6 and 3.2-7;

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 18 of 42

  • There are axial flaws that penetrate through the clad into the low alloy steel reactor vessel shell, at a depth equal to or greater than 0.075 inches in through-wall extent from the clad-to-base metal interface; or
  • Any flaws between the clad-to-base metal interface and three-eighths of the vessel thickness exceed the size allowable in ASME Code,Section XI, Table IWB-351 0-1.

These analyses must address the effects on TWCF of the known sizes and locations of all flaws detected by the ASME Code,Section XI, Appendix VIII, Supplement 4 and Supplement 6 ultrasonic examination out to three-eighths of the vessel thickness from the inner surface, and may also take into account other reactor vessel-specific information, including fracture toughness information. The licensee shall also prepare and submit a neutron fluence map, projected to the date of license expiration, for the reactor vessel beltline clad-to-base metal interface and indexed in a manner that allows the determination of the neutron fluence at the location of the detected flaws.

3.3 Plant Specific Material Properties Before performing the alternate pressurized thermal shock evaluation, a review of the latest plant-specific beltline region material properties for the BVPS-1 reactor vessel was performed. The beltline region of a reactor vessel, per the PTS rule (Reference 4),

is defined as:

... the region of the reactor vessel (shell material including welds, heat-affected zones and plates or forgings) 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 of the most limiting material with regard to radiation damage.

A summary of the best estimate copper, manganese, phosphorus, and nickel contents and RT NDT(U) values of the beltline materials for the BVPS-1 reactor vessel are summarized in Table 3.3-1. RTNDT(U) values for plate materials were determined using methodology contained in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NRC Branch Technical Position 5-2, "Overpressurization Protection of Pressurized-Water Reactors While Operating at Low Temperatures," and weld material values were generic for Linde 0091 and 1092 weld fluxes. Figure 3.3-1 shows the location of the beltline materials.

The fabrication of the BVPS-1 reactor vessel was initiated by Babcock and Wilcox (B&W) (Reference 6) and completed by Combustion Engineering (CE). In the alternate PTS rule b.T30 correlation, there are two options for plate material coefficient "B", a term used to calculate b.T30 ; one forCE manufactured vessels and one for non-CE manufactured vessels. According to the NRC reactor vessel integrity database (RVID,

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 19 of 42 Reference 7), BVPS-1 is considered to be aCE manufactured vessel. However, since B&W purchased, specified the requirements, and did the testing for the plate materials, the non-CE manufactured value for coefficient "B" will be used to calculate IJ.. T30 for the intermediate and lower shell plates.

Table 3.3-1 Details of RTMAx-x Calculation Inputs for BVPS-1 Region and Material Cu( 1l Ni( 1l p(2) Mn(3l RT NDT(u)

Material Material No. Component Heat Identification Type [wt%] [wt%] [wt%] [wt%]

Description No. [oF](1) Method(2l Intermediate 1 86607-1 A5338 C4381-1 0.14 0.62 0.015 1.40 43 MTE8 5-2 Shell Plate Intermediate 2 86607-2 A5338 C4381-2 0.14 0.62 0.015 1.40 73 MTE8 5-2 Shell Plate Lower Shell 3 86903-1 A5338 C6317-1 0.21 0.54 0.010 1.31 27 MTE8 5-2 Plate Lower Shell 4 87203-2 A5338 C6293-2 0.14 0.57 0.015 1.30 20 MTE8 5-2 Plate Intermediate Shell Linde 5 19-714A 305424 0.28 0.63 0.013 1.63 -56 Generic Longitudinal 1092 Weld Intermediate Shell Linde 6 19-7148 305424 0.28 0.63 0.013 1.63 -56 Generic Longitudinal 1092 Weld Lower Shell Linde 7 Longitudinal 20-714A 305414 0.34 0.61 0.012 1.63 -56 Generic 1092 Weld Lower Shell Linde 8 Longitudinal 20-7148 305414 0.34 0.61 0.012 1.63 -56 Generic 1092 Weld' Intermediate to Linde 9 Lower Shell Circ.11-714 90136 0.27 0.07 0.013 1.63 -56 Generic 0091 Weld Note:

1. Material chemistry and initial RTNorobtained from WCAP-15571-NP, Supplement 1, Revision 2 (Reference 5)
2. Phosphorus content and initial RTNOT determination method are obtained from RVI D (Reference 7)
3. Plate material manganese content is from plant-specific certified material test reports (Reference 6). Weld material manganese content are conservative estimates provided in Table 3.2-1

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 20 of 42 Cl~tuMFER£N1lAl i . - - ..

SE~.~. VERT JCAl SE AI"S 270° ~6607 ...7 '

(2}

tOR£ B6607~1 ___/

(1)

-H 11-7\4 (9)

B72QJ... 2 20-7147*.

(8)

(4)

-l i

..J 49.1" 8690~

(3)

Figure 3.3-1: Identification and Location of Beltline Region Materials for the BVPS-1 Reactor Vessel Note: Numbers in parentheses correspond to "No." column in Table 3.3-1

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 21 of 42 3.4 Neutron Fluence Values The projected maximum neutron fluence (ionization energy [E] greater than[>] 1.0 mega-electron volts [MeV]) values at the clad-to-base metal interface of the BVPS-1 reactor vessel for 48, 50, and 54 EFPY are shown in Table 3.4-1 for the beltline materials. BVPS-1 is projected to have a total operating time of less than 50 EFPY at end-of-life (EOL). Neutron fluence values at 50 EFPY were interpolated from the 48 and 54 EFPY data available in WCAP-15571-NP, Supplement 1, Revision 2 (Reference 5).

In addition to neutron fluence data, the BVPS-1 reactor vessel cold leg temperature under normal operating full-power conditions from the beginning of full-power operation through the current operating cycle is presented in Table 3.4-2. The temperatures and cycle times for Fort Calhoun Unit 1 and St. Lucie Unit 1 are also presented, as they are required in order to complete the surveillance capsule data evaluation for the available sister plant material data. These temperatures will be used to determine the time-weighted average of the reactor cold leg temperature, T c, used in Equation 6.

Table 3.4-1 Maximum Neutron Fluence on the Reactor Vessel Clad-to-Base Metal Interface for BVPS-1 at 48, 50 EOL, and 54 EFPY 19 Maximum Flue nee [1 0 Neutron/cm 2 , E > 1.0 MeV]

No. Region and Component Description 48 EFPY 50 EFPY 54 EFPY 1 Intermediate Shell Plate 5.35 5.,57 6.02 2 Intermediate Shell Plate 5.35 5.57 6.02 3 Lower Shell Plate 5.35 5.57 6.02 4 Lower Shell Plate 5.35 5.57 6.02 5 Intermediate Shell Longitudinal Weld 1.04 1.08 1.17 6 Intermediate Shell Longitudinal Weld 1.04 1.08 1.17 7 Lower Shell Longitudinal Weld 1.05 1.09 1.17 8 Lower Shell Longitudinal Weld 1.05 1.09 1.17 9 Intermediate to Lower Shell Circ. Weld 5.33 5.55 6.00

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 22 of 42 Table 3.4-2 Reactor Vessel Cold Leg Temperature per Operating Cycle BVPS-1 Fort Calhoun Unit 1 St. Lucie Unit 1 Cycle Cycle Cumulative Cycle Cumulative Cycle Cumulative Tc Tc Tc Time Cycle Time Time Cycle Time (oF) Time Cycle Time (oF) (oF)

(EFPY) (EFPY)_ (EFPY) (EFPY) (EFPY) (EFPY) 1 542.5 1.16 1.16 529 0.67 0.67 539 2 542.5 0.72 1.88 529 0.81 1.48 539 3.510 3.510 3 542.5 0.79 2.67 532 0.58 2.06 539 4 542.5 0.92 3.59 536 0.62 2.68 539 5 542.5 1.19 4.78 536 0.78 3.46 549 1.119 4.630 6 542.5 1.11 5.89 545 0.81 4.27 549 7 542.5 1.25 7.14 545 0.72 4.99 549 4.890 9.520 8 542.5 1.10 8.24 545 0.72 5.71 549 9 542.5 1.38 9.62 544 0.96 6.67 549 10 542.5 1.19 10.81 539 0.87 7.54 549 1.300 10.820 11 542.5 0.97 11.78 545 0.99 8.53 549 1.210 12.030 12 542.5 1.14 12.92 543 0.80 9.33 549 1.270 13.300 13 542.5 1.37 14.29 543 1.11 10.44 549 1.140 14.440 14 542.5 1.32 15.61 543 1.00 11.44 549 1.180 15.620 15 541.6 1.33 16.94 549 1.620 17.240 16 541.6 1.44 18.38 17 541.6 1.23 19.61 18 544.0 1.38 20.99 19 544.0 1.47 22.46 20 544.1 1.35 23.81 21 544.0 1.34 25.15

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 23 of 42 3.5 Surveillance Capsule Data The BVPS-1 surveillance materials are a heat-specific match for lower shell plate 86903-1 (Heat C6317-1) and intermediate shell longitudinal weld wire heat 305424.

The 30-foot-pound transition temperatures were determined using measured Charpy V-notch data plotted using CVGRAPH, Version 4.1 software which uses the requirements of 10 CFR 50, Appendix H.

There have been four surveillance capsule analyses conducted for BVPS-1. As a result, there are eight surveillance data points for material heat C6317-1 and four surveillance data points for weld wire heat 305424 measured at four different neutron fluences.

Tables 3.5-1 through 3.5-4 contain surveillance data of the BVPS-1 beltline materials required to perform the surveillance data evaluation. Tables 3.5-3 and 3.5-4 contain sister plant material data from Fort Calhoun Unit 1 and St. Lucie Unit 1, respectively.

The BVPS-1 and St. Lucie Unit 1 surveillance material data were obtained from the latest surveillance capsule analyses (References 8, 9 and 10). The Fort Calhoun Unit 1 surveillance weld data were obtained from the surveillance material baseline evaluation (Reference 11) and RVID (Reference 7). Time-averaged coolant temperatures at the time of each surveillance capsule removal were determined using the data from Table 3.4-2.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 24 of42 Table 3.5-1 Surveillance Data for BVPS-1 Base Metal C6317-1 (Lower Shell Plate 86903-1)

Time- Measured .6T30 Transition Chemical Composition Fluence Averaged Temperature (°F)< 1l 19 Withdraw Capsule p (x10 n/cm2, EFPY Coolant Cu Ni Mn Cycle E > 1.0MeV) Temperature Longitudinal Transverse

[wt%] [wt%] [wt%] [wt%] (oF) v 0.200 0.540 0.010 1.310 0.299 1.16 1 542.5 128.49 137.81 u 0.200 0.540 0.010 1.310 0.604 3.59 4 542.5 118.93 131.84 w 0.200 0.540 0.010 1.310 0.930 5.89 6 542.5 148.52 179.99 y 0.200 0.540 0.010 1.310 2.05 14.3 13 542.5 142.18 166.93 Note:

1. Values are reported to the precision level calculated in Reference 8 Table 3.5-2 Surveillance Data for BVPS-1 Weld Wire Heat 305424 (Intermediate Shell Longitudinal Weld)

Chemical Composition Fluence Time-Averaged Measured .ilTso 19 Withdraw Capsule Cu Ni p Mn (x10 n/cm2, EFPY Coolant Transition Cycle

[wt%] [wt%] [wt%] [wt%] E > 1.0MeV) Temperature (°F) Temperature (°F)< 1l v 0.260 0.620 0.018 1.370 0.299 1.16 1 542.5 159.72 u 0.260 0.620 0.018 1.370 0.604 3.59 4 542.5 166.32 w 0.260 0.620 0.018 1.370 0.930 5.89 6 542.5 187.73 y 0.260 0.620 0.018 1.370 2.05 14.3 13 542.5 179.69 Note:

1. Values are reported to the precision level calculated in Reference 8 --

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 25 of42 Table 3.5-3 Surveillance Data for Fort Calhoun Unit 1 Weld Wire Heat 305414 (BVPS-1 lower Shell longitudinal Welds)

Chemical Composition Fluence Time-Averaged Measured L1T3o 19 Withdraw Capsule Cu Ni p Mn (x1 0 n/cm2, EFPY Coolant Transition Cycle

[wt%] [wt%] [wt%] [wt%] E > 1.0MeV) Temperature (°F) Temperature (°F)< 1*2>

W-225 0.35 0.60 0.013 1.57 0.488 2.06 3 529.8 210 .

W-265 0.35 0.60 0.013 1.57 0.847 4.99 7 536.2 225 W-275 0.35 0.60 0.013 1.57 1.54 11.44 14 540.1 219 Notes:

1. TANH (hyperbolic tangent) curve fit data are obtained from RVI D (Reference 7)
2. Values are reported to the precision level used in Reference 5 Table 3.5-4 Surveillance Data for St. lucie Unit 1 Weld Wire Heat 90136 (BVPS-1 Intermediate Shell to longitudinal Shell Circumferential Weld)

Chemical Composition Fluence Time-Averaged Measured L1T3o Withdraw Capsule Cu Ni p Mn (x10 19 n/cm2, EFPY Coolant Transition Cycle 1

[wt%] [wt%] [wt%] [wt%] E > 1.0MeV) Temperature (°F) Temperature (°F)< >

97° 0.2291 0.0699 0.013 1.02 0.5174 4.63 5 541.4 72.34 104° 0.2291 0.0699 0.013 1.02 0.7885 9.52 9 545.3 67.4 284° 0.2291 0.0699 0.013 1.02 1.243 17.24 15 547.0 68.0 Note:

1. Values are reported to the precision level calculated in Reference 10

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 26 of 42 3.6 lnservice Inspection Data Three 10-year inservice inspections have been performed for the BVPS-1 reactor vessel welds. The most recent inservice inspection was performed to ASME Code,Section XI, Appendix VIII, 1989 Edition with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv and xvi) (Reference 3). Table 3.6-1 and 3.6-2 contain data on the welds and the characteristics of any indications within the beltline region of the reactor vessel obtained from the latest inservice inspection report (Reference 13).

Table 3.6-1 Reactor Vessel lnservice Inspection History for BVPS-1 Beltline Materials Weld Region and Percent Number of Number of lnservice Material Date Last Component Coverage Recordable Reportable Inspection Identification Inspected Flaws( 1)

Description Obtained Indications No.

Intermediate Shell RC-R-1-L-3 19-714A Longitudinal Weld at 2007 100% 1 None 45° Intermediate Shell No RC-R-1-L-4 19-7148 Longitudinal Weld at 2007 100% None Indications 225° Intermediate to Lower RC-R-1-C-5 11-714 2007 100% 6 None Shell Circ. Weld Lower Shell RC-R-1-L-6 20-714A Longitudinal Weld at 2007 100% 1 None 135° Lower Shell RC-R-1-L-7 20-7148 Longitudinal Weld at 2007 100% 1 None 315° Note:

1. Flaws that are reportable are those that exceed the ASME Code,Section XI, Table IWB-3510-1 acceptance standards Table 3.6-2 lnservice Inspection Information for Reactor Vessel Beltline Flaws for BVPS-1 Weld Weld Table In service Type Weld Indication UT Beam t L s 2a a (in.) IWB-3510-1 lnspeciton Width (in.) No. Direction( 1) (in.) (in.) (in.) (in.)

(A or C)( 1) Disposition No.

RC-R-1-L-3 A 1.78 1 ccw 8.0 1.25 3.97 0.125 0.060 Allowable 1 UP 8.0 1.25 3.34 0.2 0.1 Allowable 2 UP 8.0 0.75 3.0 0.125 0.06 Allowable 3 UP 8.0 0.75 3.55 0.2 0.1 Allowable RC-R-1-C-5 c 1.25 4 UP 8.0 0.75 2.34 0.125 0.06 Allowable 5 UP 8.0 1.1 1.22 0.125 0.06 Allowable 6 UP 8.0 0.75 2.73 0.18 0.09 Allowable RC-R-1-L-6 A 1.78 1 ccw 8.0 0.6 2.06 0.125 0.06 Allowable RC-R-1-L-7 A 1.78 1 ccw 8.0 1.1 2.22 0.14 0.07 Allowable Note:

1. A =Axial, C =Circumferential, CCW =Counter-clockwise, UT =Ultrasonic

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 27 of 42 3.7 Determination of RTMAX-X Values for Beltline Region Materials 3.7.1 Calculation of RTMAX-X Values Using the alternate PTS rule methodology described in Section 3.2.1, RTMAX-X values were generated for the beltline region materials of the BVPS-1 reactor vessel using fluence values at the EOL (50 EFPY). These values were calculated using reactor vessel beltline material copper, nickel, phosphorus, and manganese content, unirradiated RTNDT, projected EOL neutron fluence values, and time-weighted average reactor vessel cold-leg temperature as described in Sections 3.3 and 3.4. Tables 3.7-1 through 3.7-3 summarize the results of Equations 1 through 7 when used to calculate RTMAX-X for the BVPS-1 axial welds, plates, and circumferential welds.

The calculated RTMAX-X values and applicable PTS screening criteria are provided in Table 3.7-4:

Table 3.7-1 RT MAX-Aw Calculation Results for BVPS-1 at 50 EFPY Fluence b.T3o Total 8eltline Region Material 19 b.T3o RTNDT(u) RT MAX-AW(i)

Weld Group (x10 n/cm2, Adj. RTNW Location Heat No. (oF) (oF)(1) (oF) (oF) 2) (oF)

E > 1.0MeV)

Intermediate Shell 305424 1.08 222.7 0.0 -56 Intermediate Longitudinal Weld 166.7 Shell Longitudinal Intermediate Shell C4381-1 1.08 109.5 0.0 43 182.5 Weld Plate 152.5 19-714A Intermediate Shell C4381-2 1.08 109.5 0.0 73 Plate 182.5 Intermediate Shell 305424 1.08 222.7 0.0 -56 Intermediate Longitudinal Weld 166.7 Shell Longitudinal Intermediate Shell C4381-1 1.08 109.5 0.0 43 182.5 Weld Plate 152.5 19-7148 Intermediate Shell C4381-2 1.08 109.5 0.0 73 Plate 182.5 Lower Shell Lower Shell 305414 1.09 228.0 0.0 -56 Longitudinal Weld 172.0 Longitudinal Weld 178.6 Lower Shell Plate C6317-1 1.09 125.2 26.36 27 178.6 20-714A Lower Shell Plate C6293-2 1.09 104.5 0.0 20 124.5 Lower Shell Lower Shell 305414 1.09 248.0 0.0 -56 Longitudinal Weld 172.0 Longitudinal Weld 178.6 Lower Shell Plate C6317-1 1.09 125.2 26.36 27 178.6 20-7148 Lower Shell Plate C6293-2 1.09 104.5 0.0 20 124.5 Note:

1. Adjustment to b. T 30 due to surveillance data statistical checks. Details regarding this adjustment can be seen in Section 3. 7 .2.
2. The sum of b. T3o, the adjustment to b. T3o, and RTNDTiul*

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 28 of 42 Table 3.7-2 RTMAX-PL Calculation Results for BVPS-1 at 50 EFPY Fluence Total Material 19 LlT3o LlT3oAdj. RTNDT(u) 8eltline Region Location (x10 n/cm2, (oF) (oF)(1) (oF) RTNDT RTMAX-PL (°F)

Heat No. (oF)(2)

E > 1.0MeV)

Intermediate Shell Plate C4381-1 5.57 156.0 0.0 43 199.0 86607-1 Intermediate Shell Plate C4381-2 5.57 156.0 0.0 73 229.0 229.0 86607-2 Lower Shell Plate 86903-1 C6317-1 5.57 168.1 26.36 27 221.4 Lower Shell Plate 87203-2 C6293-2 5.57 149.4 0.0 20 169.4 Note:

1. Adjustment toLl T 30 due to surveillance data statistical checks. Details regarding this adjustment can be seen in Section 3.7.2.
2. The sum of Ll T3o, the adjustment toLl T3o, and RTNDTiul*

Table 3.7-3 RTMAx-cw Calculation Results for BVPS-1 at 50 EFPY Fluence Material 19 LlT3o Total 8eltline Region (x1 0 AT3o RTNDT(u) RTMAX-CW Weld Group Heat (oF) Adj. RTN8T Location n/cm2, E > (oF)(1) (oF) (oF) 2) (oF)

No.

  • 1.0MeV)

Intermediate to Lower 90136 5.55 155.2 0.0 -56 99.2 Shell Circ. Weld Intermediate Intermediate Shell to Lower C4381-1 5.55 155.8 0.0 43 198.8 Plate Shell Circ. 228.8 Intermediate Shell Weld C4381-2 5.55 155.8 0.0 73 228.8 Plate 11-714 Lower Shell Plate C6317-1 5.55 168.0 26.36 27 221.3 Lower Shell Plate C6293-2 5.55 149.3 0.0 20 169.3 Note:

1. Adjustment to AT30 due to surveillance data statistical checks. Details regarding this adjustment can be seen in Section 3. 7 .2.
2. The sum of AT3o, the adjustment to AT3o, and RTNortul*

Table 3.7-4 RTMAx-x values for BVPS-1 at 50 EFPY 10 CFR 50.61a 8VPS-1 Screening Criteria Axial Weld-RTMAX-AW (°F) 182.5 269 Plate-RTMAX-PL (°F) 229.0 356 Axial Weld and Plate-RTMAX-AW +

411.5 538 RTMAX PL Circumferential Weld-RTMAx-cw (°F) 228.8 312 The RTMAX-X values calculated for BVPS-1 are less than PTS screening criteria and therefore meet this requirement of the alternate PTS rule. Upon NRC issuance of the license amendment, FENOC intends to document Table 3.7-4 values in the BVPS-1 Pressure and Temperature Limits Report.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 29 of 42 3.7.2 Surveillance Capsule Data Statistical Checks As discussed in Section 3.2.2, the alternate PTS rule (Reference 1) requires that surveillance data that could affect the calculation of L\T3o be evaluated. This requirement is only applicable for materials for which three or more points of surveillance data exist at three or more unique fluence values.

For BVPS-1, there have been four surveillance capsules withdrawn to date. The BVPS-1 and sister plant surveillance materials, along with their calculated values of L\T30, can be seen below in Table 3.7-5. This table includes a list of the tested and analyzed capsules for each material.

Table 3.7-5 Surveillance Capsule Materials for BVPS-1 Region and Material Fluence 19 Calculated No. Component Identification Capsule Direction (x10 n/cm2, I:J.T3o (°F)

Description (Heat No.) E > 1.0MeV) v Longitudinal 0.299 75.85 u Longitudinal 0.604 94.79 w Longitudinal 0.930 105.34 Lower Shell Plate 86903-1 y Longitudinal 2.05 126.06 3

Base Metal (C6317-1) v Transverse 0.299 75.85 u Transverse 0.604 94.79 w Transverse 0.930 105.34 y Transverse 2.05 126.06 v N/A 0.299 148.40 5 Surveillance Program 19-717A&B u N/A 0.604 178.76 6 Weld Metal (305424) w N/A 0.930 193.11 y N/A 2.05 217.24 W-225 N/A 0.488 194.42 7 Ft. Calhoun 20-717A & B Surveillance Program W-265 N/A 0.847 210.28 8 (305414)

Weld Metal W-275 N/A 1.54 225.58 97° N/A 0.5174 75.34 St. Lucie Surveillance 11-714 9 (90136) 104° N/A 0.7885 81.00 Program ~eld Metal 284° N/A 1.243 88.28 All of the materials listed have at least three data points at three or more different neutron fluences, and therefore this data can be used to determine if the surveillance data shows a significantly different trend than the embrittlement model predicts. Using the methodology described in Section 3.2.2, a mean deviation test, a slope deviation test, and an outlier deviation test were conducted for each surveillance material. The inputs for the surveillance data evaluations, including the measured values of L\ T30 , are

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 30 of 42 provided in Tables 3.5-1 through 3.5-4 for the four surveillance materials. The results of the evaluations are shown in Tables 3.7-6a, 3.7-7, 3.7-8, and 3.7-9.

Table 3.7-6a shows that both the mean and outlier deviation tests are not satisfied by the base metal surveillance capsule results for lower shell plate 86903-1. As described in Section 3.2.2, the alternate PTS rule requires proposing L1 T3o and RTMAX-X values, considering their plant-specific surveillance data, to be used for evaluation relative to the acceptance criteria. Since the eight base-metal surveillance capsule results in Table 3.5-1 were included in the database used for development of Equations (5) to (7) in the alternate PTS rule, the proposed adjustment to the bT3o and RTMAX-X values should not be excessive. That is, the variability in the prediction of mean values (residual values) for only eight measurements is expected to be higher than the variability (standard deviation) for all 309 measurements (Reference 12) used to develop the equation constants for plate materials with the higher copper content.

Therefore, a minimum adjustment of 26.36°F to just pass the mean deviation test in Table 3.7-6a was selected. This adjustment was added to the calculated values of bT3o in Table 3. 7-5 and thus subtracted from each of the eight residual values in Table 3.7-6a. As shown in Table 3.7-6b, these adjusted residual values all pass the mean deviation test, slope deviation test, and outlier deviation test. The surveillance capsule data adjustment of 26.36°F was applied to the calculated values of b T3o for lower shell plate 86903-1 (Heat C6317-1) in Tables 3.7-1 to 3.7-3, but it did not affect the calculated values of RTMAX-X in Table 3.7-4 because there were other materials without any adjustment to the calculated values of b T 30 that were more limiting.

Table 3.7-Ga Surveillance Data Evaluation for BVPS-1 Base Metal Heat C6317-1 (Lower Shell Plate 86903-1) without any Adjustment Log of Fluence 2 Capsule Direction llx" Residual "r" (x- Xavg) r* (r/sigma) v Longitudinal 18.48 52.64 0.167 2.48 u Longitudinal 18.78 24.14 0.011 1.14 w Longitudinal 18.97 43.18 0.007 2.04 y Longitudinal 19.31 16.12 0.183 0.76 v Transverse 18.48 61.96 0.167 2.92 u Transverse 18.78 37.05 0.011 1.75 w Transverse 18.97 74.65 0.007 3.52 y Transverse 19.31 40.87 0.183 1.93 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard Deviation 21.2 Slope (m) -25.64 Largest r* 3.52 (sigma)

Mean Deviation 43.8 Standard Error of Fit 18.64 Largest allowable r* 3.02 Maximum Mean Standard Error of 17.5 21.74 Pass/Fail? Fail Residual Slope Pass/Fail? Fail T -Statistic -1.18 Second largest r* 2.92 Second largest Critical T-Statistic 3.14 2.05 allowable r*

Pass/Fail? Pass Pass/Fail? Fail

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 31 of 42 Table 3.7-6b Surveillance Data Evaluation for BVPS-1 Base Metal Heat C6317-1 (Lower Shell Plate 86903-1) with an Adjustment of 26.36°F Log of Fluence Adjusted 2 Adjusted r*

Capsule Direction (X- Xav9)

"x" Residual "r" (r/sigma) v Longitudinal 18.48 26.28 0.167 1.24 u Longitudinal 18.78 -2.22 0.011 -0.10 w Longitudinal 18.97 16.82 0.007 0.79 y Longitudinal 19.31 -10.24 0.183 -0.48 v Transverse 18.48 35.60 0.167 1.68 u Transverse 18.78 10.69 0.011 0.50 w Transverse 18.97 48.29 0.007 2.28 y Transverse 19.31 14.51 0.183 0.68 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard Deviation 21.2 Slope (m) -25.64 Largest r* 2.28 (sigma)

Mean Deviation 17.5 Standard Error of Fit 18.64 Largest allowable r* 3.02 Maximum Mean Standard Error of 17.5 21.74 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T -Statistic -1.18 Second largest r* 1.68 Second largest Critical T-Statistic 3.14 2.05 allowable r*

Pass/Fail? Pass Pass/Fail? Pass Table 3.7-7 Surveillance Data Evaluation for BVPS-1 Weld Wire Heat 305424 (Intermediate Shell Longitudinal Weld)

Log of Fluence 2 Capsule Direction ux" Residual "r" (X- Xav9) r* (r/sigma) v N/A 18.48 11.32 0.167 0.43 u N/A 18.78 -12.44 0.011 -0.47 w N/A 18.97 -5.38 0.007 -0.20 y N/A 19.31 -37.55 0.183 -1.42 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard 26.4 Slope (m) -54.02 Largest r* 0.43 Deviation (sigma)

Standard Error of Mean Deviation -11.0 9.08 Largest allowable r* 2.81 Fit Maximum Mean Standard Error of 30.8 14.97 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T -Statistic -3.61 Second largest r* -0.20 Second largest Critical T-Statistic 6.96 1.73 allowable r*

Pass/Fail? Pass Pass/Fail? Pass

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 32 of 42 Table 3.7-8 Surveillance Data Evaluation for Fort Calhoun Unit 1 Weld Wire Heat 305414 (BVPS-1 Lower Shell Longitudinal Welds)

Log of Fluence 2 Capsule Direction Residual "r" (x- Xav9) r* (r/sigma)

"x" W-225 N/A 18.69 15.58 0.061 0.59 W-265 N/A 18.93 14.72 0.000 0.56 W-275 N/A 19.19 -6.58 0.064 -0.25 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard 26.4 Slope (m) -44.93 Largest r* 0.59 Deviation (sigma)

Standard Error of Mean Deviation 7.9 7.98 Largest allowable r* 2.71 Fit Maximum Mean Standard Error of 35.5 22.59 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T-Statistic -1.99 Second largest r* 0.56 Second largest Critical T-Statistic 6.96 1.55 allowable r*

Pass/Fail? Pass Pass/Fail? Pass Table 3.7-9 Surveillance Data Evaluation for St. Lucie Unit 1 Weld Wire Heat 90136 (BVPS-1 Intermediate Shell to Longitudinal Shell Circumferential Weldt Log of Fluence 2 Capsule Direction IIXIJ Residual "r" (X- Xav 9) r* (r/sigma) 97° N/A 18.71 -3.00 0.035 -0.11 104° N/A 18.90 -13.60 0.000 -0.52 284° N/A 19.09 -20.28 0.037 -0.77 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Standard 26.4 Slope (m) -45.25 Largest r* -0.12 Deviation (sigma)

Standard Error of Mean Deviation -12.3 1.87 Largest allowable r* 2.71 Fit Maximum Mean Standard Error of 35.5 6.95 Pass/Fail? Pass Residual Slope Pass/Fail? Pass T-Statistic -6.51 Second largest r* -0.52 Second largest Critical T-Statistic 31.82 1.55 allowable r*

Pass/Fail? Pass Pass/Fail? Pass As shown in Tables 3.7-6b through 3.7-9, the surveillance results for the plate and weld surveillance materials now satisfy the criteria in the alternate PTS rule for all three tests.

Therefore, the use of Equations (5) to (7) in the alternate PTS rule (Reference 1) for calculation of~ T30 with the surveillance data adjustment of Table 3. 7 -6b is acceptable for BVPS-1.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 33 of 42 3.7.3 Reactor Vessel Beltline lnservice Inspection Data Evaluation In accordance with the requirements discussed in Section 3.2.3, the results of the latest inservice inspection of the reactor vessel of BVPS-1 were analyzed in detail to ensure that the recorded indications met the acceptance criteria. The reactor vessel inservice inspection data specified in Section 3.6 indicates that the Category B-A examinations of the BVPS-1 reactor vessel beltline region welds have been performed to ASME Code,Section XI, Appendix VIII requirements. At least one inspection has been performed on each weld that is inspected per Code requirements. Inspection coverage of the welds within the beltline region has been 100 percent. No inside surface flaws were found in the beltline welds because all indications were reported to be subsurface.

After reviewing the data from the last inservice inspection, conducted in 2007 (Reference 13), nine indications with the potential to be located in the beltline region of the BVPS-1 reactor vessel were recorded. Seven of these indications were adjacent to the core and therefore within the reactor vessel beltline region. Four of those indications fall within the inner 3/St of the reactor vessel thickness and are allowable per ASME Code,Section XI, 1989 edition (Reference 3), Table IWB-3510-1. None of the indications fall within the inner 1-inch of the reactor vessel. Therefore, no further

  • evaluation of the alternate PTS rule Table 2 and 3*flaw limits is required and BVPS-1 inherently meets the examination and flaw assessment requirements detailed in Section 3.2.3. This evaluation of these inservice inspection indications is summarized in Table 3.7-10.

Table 3.7-10 Evaluation of lnservice Inspection Information for Reactor Vessel Beltline Flaws for BVPS-1 Weld Inner Inner Flaw Limit lnservice Indication TWE Location Adjacent (3/8) Flaw 1

(in.)< ) (Plate/Weld) to Core? thickness (1/10)t Evaluation Inspection No. Orientation or 1"? Required?

No. _(_t) ?

RC-R-1-L-3 1 0.125 Plate Yes No No Axial No 1 0.2 Weld Yes No No Circ. No 2 0.125 Plate Yes Yes No Circ. No 3 0.2 Plate Yes No No Circ. No RC-R-1-C-5 4 0.125 Plate Yes Yes No Circ. No 5 0.125 Plate Yes Yes No Circ. No 6 0.18 Weld Yes Yes No Circ. No RC-R-1-L-6 1 0.125 Plate No Yes No Axial No RC-R-1-L-7* 1 0.14 Weld No Yes No Axial No Note:

1. Through-wall extent (TWE) is the same as the dimension 2a from Table 3.6-2

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 34 of 42 3.8 Evaluation of Extended Beltline Materials The alternate PTS rule requires that RT MAX-x values be projected for all materials in the reactor vessel beltline. The "beltline" is defined in Section 3.3.

Historically, before the consideration of extended operating licenses, and consistent with the definition above, the beltline was considered to only be those materials that were immediately adjacent to the core. Consideration of these materials was also sufficient for consideration of the adjacent regions as these beltline materials typically extend for some vertical distance above and below the immediate core region. With the advent of the extended operating license, a concern arose that fluence accumulation, and subsequently embrittlement, could become significant in regions outside of those that had been previously considered as part of the "beltline." A guideline was adopted in NUREG-1801, "Generic Aging Lessons Learned (GALL) Report" (GALL report), that irradiation effects be considered for all components with fluence accumulations greater than 1 x 10 17 n/cm 2 , E > 1.0 MeV. Materials with fluence levels meeting this criterion, other than those materials immediately adjacent to the core, are now typically referred to as extended beltline materials.

In light of the GALL report, it was determined that the BVPS-1 upper shell forging and the upper to intermediate shell girth weld would accumulate fluence levels equal to or greater than 1 x 10 17 n/cm 2 , E > 1.0 MeV. Appropriate material properties were determined for these materials and their 11 T3o and RTPTS values are shown in Table 3.8-1. Materials below the core that have not been previously considered as part of the traditional beltline will not accumulate fluence levels equal to or greater than 1 x 10 17 n/cm 2 , E > 1.0 MeV.

The upper shell forging has a RT PTs value approximately one half of the value for the alternate PTS rule limiting beltline plate (Intermediate Shell Plate 86607-2).

Furthermore, the RTMAX screening criteria in Table 1 of the alternate PTS rule is the same for plates and forgings not susceptible to underclad cracking. Surveillance data is not available for the upper shell forging. Therefore, Intermediate Shell Plate 86607-2 will remain the limiting beltline material for the alternate PTS rule.

The upper to intermediate shell girth weld has an RT PTS value that is slightly higher than the intermediate to lower shell girth weld. However, this weld is the same heat (305414) as the lower shell longitudinal welds. Since the upper to intermediate shell girth weld accumulates lower fluence than the lower shell longitudinal welds, and the welds have the same initial RT NOT, the RT NoT + 11T 30 for this girth weld will be less than that for lower shell longitudinal welds. Since the calculation of RTMAX-cw considers adjacent materials, including the lower shell longitudinal welds, but is controlled by the higher RTNOT + 11 T30 of the Intermediate Shell Plate (86607 -2), inclusion of the upper to intermediate shell girth weld will have no impact on the calculation and resulting value for RT MAX-cw.

Furthermore, the evaluation of surveillance capsule data for the lower shell longitudinal welds is applicable and acceptable for the upper to intermediate shell girth weld.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 35 of 42 The inservice inspection results have been reviewed for the intermediate to upper shell girth weld, and it has been determined that two indications were recorded during the last inservice inspection. Both indications are allowable per Table IWB-351 0-1 of ASME Section XI. However, only one of these indications is within the inner 3/St of the reactor vessel. Neither of these indications is within the inner 1-inch or 1/1 Ot of the reactor vessel and therefore, the limits of Tables 2 and 3 of the alternate PTS rule do not apply.

The RTMAX-X values calculated for the traditional beltline materials, as identified in Section 3.7.1, remain the limiting values for BVPS-1 implementation of the alternate PTS rule.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 36 of 42 Table 3.8-1 RTPTS Values for BVPS-1 Materials with Fluence > 1 x 1017 n/cm2 (E > 1.0 MeV) at 50 EFPY Surface Chemistry Fluence Initial (c)

Margin!dl (e)

Material Heat Number Neutron Factor (b) ARTPTS O"u 0",1 RTPTS Material Description Factor, RTNDT ID (lot Number) Fluence (CF) (oF) (oF) (oF) (oF) (oF) 19 2 FF<a> (oF)

(x10 n/cm ) (oF)

Upper Shell Forqinq 86604 123V339VA1 0.625 0.8685 84.2 40 73.1 0 17 34 147.1 305414 Upper to Intermediate 10-714 (3951 & 0.625 0.8685 209.11 -56 181.6 17 28 65.5 191.1 Shell Girth Weld 3958) __________ .,.I


-------------- --------------- ----------- ------------- -28m- -------------

~ Using non-credible surveillance data rJ 1 0.625 0.8685 216.9 -56 188.4 17 65.5 197.9 AOFJ 0.625 0.8685 41.0 10 35.6 17 17.8 49.2 94.8 Upper to Intermediate FOIJ 0.625 0.8685 41.0 10 35.6 17 17.8 49.2 94.8 10-714 Shell Girth Weld EODJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 HOCJ 0.625 0.8685 27.0 10 23.4 17 11.7 41.3 74.8 Notes:

a. FF = fluence factor= f 10*28 - 0 "10 109 <fll_
b. Initial RTNDT value for the upper shell forging is a measured value. All other values are generic.
c. ARTPTS = CF X FF.

2 2 112

d. Margin (M) = 2( cru [standard deviation for RTNDT(UJ] + cr,1 [standard deviation for RTNoT]) .
e. RTPTs =Initial RTNoT + ARTPTs +Margin.
f. The Fort Calhoun surveillance weld metal is the same weld heat as the BVPS-1 upper to intermediate shell girth weld (heat 305414). The Fort Calhoun surveillance weld data is non-credible; therefore, the higher cr11 term of 28°F was utilized for BVPS-1 weld heat 305414.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 37 of 42

3.9 CONCLUSION

Based on this evaluation, the BVPS-1 reactor vessel is acceptable per the alternate PTS rule acceptance criteria. As shown in Section 3.7.1, all of the beltline region materials in the BVPS-1 reactor vessel have EOL (50 EFPY) RTMAX-X values below the screening criteria values. After conducting surveillance data statistical tests, it was determined that the surveillance data did not satisfy the alternate PTS rule requirements. Adjustments were made to the calculations of the RT MAX-X values and incorporated into the analysis. These adjustments did not change the limiting RTMAX-X values. A review of the latest reactor vessel inservice inspection report for BVPS-1 showed that the flaw density and size distribution is acceptable per the alternate PTS rule requirements.

4.0 REGULATORY EVALUATION

FirstEnergy Nuclear Operating Company (FENOC) proposes to amend the Beaver Valley Power Station, Unit No. 1 Operating License DPR-66. The requested amendment involves a change to the operating license that would authorize implementation of 10 CFR 50.61 a (alternate pressurized thermal shock [PTS] rule),

"Alternate fracture toughness requirements for protection against pressurized thermal shock events," in lieu of the requirements of 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events."

Implementation of the alternate PTS rule would result in a burden reduction while continuing to provide adequate protection to public health and safety.

4.1 Significant Hazards Consideration FirstEnergy Nuclear Operating Company has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No This amendment request would allow implementation of the alternate PTS rule in lieu of 10 CFR 50.61 and would not involve a significant increase in the probability or consequences of an accident. Application of the alternate PTS rule in lieu of 10 CFR 50.61 would not result in physical alteration of a plant structure, system or component, or installation of new or different types of equipment. Further, application of the alternate PTS rule would not significantly affect the probability of accidents previously evaluated in the Updated Final Safety Analysis Report (UFSAR) or cause a change to any of the dose analyses associated with the UFSAR accidents because accident mitigation functions would remain unchanged. Use of the alternate PTS rule would change how fracture toughness of the reactor

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 38 of 42 vessel is determined and does not affect reactor vessel neutron radiation fluence. As such, implementation of the alternate PTS rule in lieu of 10 CFR 50.61 would not increase the likelihood of a malfunction.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The amendment request would allow implementation of the alternate PTS rule in lieu of 10 CFR 50.61. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. No physical plant alterations are made as a result of the proposed change. The proposed change does not challenge the performance or integrity of any safety-related system.

Therefore, the proposed change does not create the possibility of a new or

  • different kind of accident from any previously evaiuated.
3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The amendment request would authorize implementation of the alternate PTS rule in lieu of 10 CFR 50.61. The alternate PTS rule would maintain the same functional requirements for the facility as 10 CFR 50.61. The alternate PTS rule establishes screening criteria that limit levels of embrittlement beyond which operation cannot continue without further plant-specific evaluation or modifications. Sufficient safety margins are maintained to ensure that any potential increases in core damage frequency and large early release frequency resulting from implementation of the alternate PTS rule are negligible. As such, there would be no significant reduction in the margin of safety as a result of use of the alternate PTS rule. The margin of safety associated with the acceptance criteria of accidents previously evaluated in the UFSAR is unchanged.

The proposed change would have no affect on the availability, operability, or performance of the safety-related systems and components.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 39 of 42 Based on the above, FENOC concludes that the proposed amendment does not involve a significant hazards consideration under the criteria set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements I Criteria An assessment of the proposed changes concluded that there are no exceptions to any of the following regulations. Therefore, FENOC would remain in compliance with the following regulations and guidance:

10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 1, "Quality Standards and Records," requires the structures, systems, and components important to safety to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function .

. GDC 31, "Fracture prevention of the reactor coolant pressure boundary," requires that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

GDC 32, "Inspection of the reactor coolant pressure boundary," requires components that are part of the reactor coolant pressure boundary be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight intergrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," requires that alllightwater reactors meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR 50, Appendix G and Appendix H.

10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," ensures that changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 40 of 42 environment are monitored. Under the program, fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel.

Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001, describes methods for determining reactor pressure vessel fluence.

Beaver Valley Power Station (BVPS), Unit No. 1 Updated Final Safety Analysis Report (UFSAR) Table 16-1, "Unit 1 License Renewal Commitments," includes a commitment, Item Number 24, associated with flux reduction plans to manage PTS.

The same commitment is contained in Appendix A of the NRC, "Safety Evaluation Report Related to the License Renewal of Beaver Valley Power Station, Units 1 and 2 (NUREG-1929)," dated October 2009 and states:

Prior to exceeding the PTS screening criteria for BVPS Unit 1, FENOC will select a flux reduction measure to manage PTS in accordance with the requirements of 10 CFR 50.61. A flux reduction plan will be submitted to the NRC for review and approval.

This commitment will no longer be necessary following NRC approval of this license amendment request. FirstEnergy Nuclear Operating Company plans to withdraw this commitment via the appropriate process following approval of the license amendment.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion setforth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 41 of 42

6.0 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. 1, dated January 4, 2010, and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.
2. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010.
3. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with no Addenda, American Society of Mechanical Engineers, New York.
4. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events,"

Federal Register, Volume 60, No. 243, dated December 19, 1995 and last updated on January 4, 2010.

5. WCAP-15571-NP, Supple.ment 1, Revision 2, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program,"

September 2011.

6. CMTR-RV-DLW, "Reactor Vessel Certified Material Test Reports for DLW."
7. Nuclear Regulatory Commission Reactor Vessel Integrity Database (RVID),

Version 2.0.1, July 6, 2000.

8. WCAP-15571-NP, Revision 1, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," April 2008, (Accession No. ML082740207).
9. WCAP-12751, Revision 0, "Analysis of the Capsule at 104° from the Florida Power and Light Company St. Lucie Unit No. 1 Reactor Vessel Radiation Surveillance Program," November 1990.
10. WCAP-15446, Revision 1, "Analysis of Capsule 284° from the Florida Power and Light Company St. Lucie Unit No. 1 Reactor Vessel Radiation Surveillance Program," January 2002, (Accession No. ML021280606).
11. TR-0-MCD-001, "Omaha Public Power District Fort Calhoun Station Unit No.

1, Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program," March 1977.

12. ORNL/TM-2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," November 2007, (Accession No. ML081000630).

FENOC Evaluation of the Proposed Amendment Beaver Valley Power Station, Unit No. 1 Page 42 of 42

13. Wesdyne lSI Report, "1 0 Year Reactor Vessel In-Service Inspection for Beaver Valley Unit #1 Power Station," performed in October 2007.
14. NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (1 0 CFR 50.61 ),"August 2007.

Attachment 1 Proposed Facility Operating License Change (Mark-Up)

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APPENDIXC ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-66 FirstEnergy Nuclear Operating Company and FirstEnergy Nuclear Generation, LLC shall comply with the following conditions on the schedules noted below:

Amendment Additional Condition Implementation Number Date 281 Initial Performance of New Surveillance and Assessment Requirements Upon implementation of Amendment No. 281 adopting TSTF-448, The Revision 3, the determination of control room envelope (CRE) amendment unfiltered air inleakage as required by Surveillance Requirement shall be (SR) 3. 7.1 0.4, in accordance with Specification 5.5.14.c(i), the implemented assessment of CRE habitability as required by Specification within 120 days 5.5.14.c(ii), and the measurement of CRE pressure as required by from date of Specification 5.5.14.d, shall be considered met. Following issuance implementation:

(a) The first performance of SR 3.7.10.4, in accordance with Specification 5.5.14.c(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from the date of the most recent successful tracer gas test, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE

  • habitability, Specification 5.5.14.c(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from the date of the most recent successful tracer gas test, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test.

Alternate Fracture Toughness Requirements for Protection The Against Pressurized Thermal Shock Events amendment shall be License Amendment No. TBD authorizes the implementation of implemented 10 CFR 50.61 a in lieu of 10 CFR 50.61. within 120 days from date of issuance Beaver Valley Unit 1 C-5 Amendment No. ,TI3P_ _ l-- {Deleted: 290

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Attachment 2 Proposed Facility Operating License Change (Re-Typed)

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APPENDIXC ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-66 FirstEnergy Nuclear Operating Company and FirstEnergy Nuclear Generation, LLC shall comply with the following conditions on the schedules noted below:

Amendment Additional Condition Implementation Number Date 281 Initial Performance of New Surveillance and Assessment Requirements Upon implementation of Amendment No. 281 adopting TSTF-448, The Revision 3, the determination of control room envelope (CRE) amendment unfiltered air inleakage as required by Surveillance Requirement shall be (SR) 3. 7.1 0.4, in accordance with Specification 5.5.14.c(i), the implemented assessment of CRE habitability as required by Specification within 120 days 5.5.14.c(ii), and the measurement of CRE pressure as required by from date of Specification 5.5.14.d, shall be considered met. Following issuance implementation:

(a) The first performance of SR 3. 7.1 0.4, in accordance with Specification 5.5.14.c(i), shall be. within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from the date of the most recent successful tracer gas test, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from the date of the most recent successful tracer gas test, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 18 months, plus the 138 days allowed by SR 3.0.2, as measured from the date of the most recent successful pressure measurement test.

TBD Alternate Fracture Toughness Requirements for Protection The Against Pressurized Thermal Shock Events amendment shall be License Amendment No. TBD authorizes the implementation of implemented 10 CFR 50.61 a in lieu of 10 CFR 50.61. within 120 days from date of issuance Beaver Valley Unit 1 C-5 Amendment No. TBD