NL-17-0955, Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06

Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06

ML17156A831
Person / Time
Site:	Hatch
Issue date:	06/05/2017
From:	Wheat J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-0955
Download: ML17156A831 (41)

Text

~ Southern Nuclear Justin T. Wheat Nuclear Licensing Manager JUN 0 5 2011 40 lnvcmrss Cenler PJr!..wdy Posl Oflice llox 1295 Birmingham. Al. 35242 205 992 5998 lei 205 992 760 I fax jlwhcal@>~oulh~rnco.com Docket Nos.: 50-321 50-366 NL-17-0955 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Alternatives HNP-ISI-Alf-05-05 and HNP-181-AL T-05-06 Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(z)(1) Southern Nuclear Operating Company (SNC) hereby submits Alternative HNP-ISI-ALT-05-05 to allow a reduced percentage requirement for nozzle-to-vessel weld and inner radius examinations based on the applicability criteria of Code Case N-702.

In addition, in accordance with 10 CFR 50.55a(z)(2) SNC hereby submits Alternative HNP-ISI-ALT-05-06. This Alternative proposes alternate test pressures for certain Class 1 pressure tests using non-nuclear heat. This Alternative would be used following repair/replacement activities (excluding those on the reactor vessel) which occur subsequent to the periodic Class 1 pressure test required by Table IWB-2500-1, Category 8-P and prior to the next refueling outage on those components that cannot be isolated.

Components which can be isolated will be pressure tested at a pressure in accordance with IWB-5221 (a).

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

Respectfully submitted, JTW/RMJ

Enclosures:

1. Alternative HNP-ISI-ALT-05-05

2. Alternative HNP-181-AL T-05-06

U. S. Nuclear Regulatory Commission NL-17-0955 Page 2 cc:

Regional Administrator, Region II NRR Project Manager-Hatch Senior Resident Inspector-Hatch RTYPE: CHA02.004

Edwin I. Hatch Nuclear Plant Alternatives HNP-181-ALT-05-05 and HNP-181-ALT-05-06 Alternative HNP-181-AL T 05 to NL-17-0955 Alternative HNP-ISI-AL T-05-05

1. ASME Code Component(s) Affected Code Class:

ASME Section XI Code Class 1 Component Numbers:

Code

References:

Various (see Table 1 for detailed list of components)

ASME Section XI, 2007 Edition with 2008 Addenda Code Case N-702 Examination Category:

B-D Item Number(s):

83.90 and 83.100

2. Requested Approval Date Approval is requested by May 31, 2018.

3. Applicable! ASME Code Requirements ASME Section XI, 2007 Edition through the 2008 Addenda, Table IWB-2500-1, Examination Category 8-D, "Full Penetration Welded Nozzles in Vessels" requires a volumetric examination of all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles each 1 0-year interval. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented. SNC has also implemented Code Case N-648-1 to permit a visual VT-1 examination of the three RPV Closure Head nozzle inside radius sections.

The RPV nozzle-to-vessel welds and nozzle inside radius sections subject to this request are listed below in Table 1 (applies to both Hatch Nuclear Plant (HNP) units):

TABLE 1 Identification Description Total Minimum Number to Number Number be examined N1 Recirculation Outlet 2

1 N2 Recirculation Inlet 10 3

N3 Main Steam Outlet 4

1 N5 Core SQra_y_

2 1

N6 Head Spray 2

1 N7 Head Vent 1

1 N8 Jet Pump 2

1 Instrumentation

4. Reason for Request

NRC Regulatory Guide 1.147, Rev. 17 conditionally accepts the use of Code Case N-702.

The NRC condition required a Licensee to develop an evaluation to justify that certain criteria be satisfied and the evaluation must be reviewed and accepted by the NRC prior to the implementation of the Code Case.

This code case provides an alternative to performing examination of 1 00% of the nozzle-to-vessel welds and inside radius sections for Examination Category B-D nozzles with the exception of the Feedwater and Control Rod Drive Return Line (CRDRL) Nozzles. The alternative is to perform examinations on a minimum of 25% of the nozzle inside radius E1-1 to NL-17-0955 Alternative HNP-ISI-ALT-05-05 sections and nozzle-to-vessel welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles.

5. Proposed Alternative and Basis for Use

Proposed Alternative Pursuant to 10 CFR 50.55a(z)(1 ), SNC requests approval to implement the alternative of Code Case N-702 in lieu of the code required 100% examination of all nozzles identified in Table 1. As an alternative, for the nozzle-to-vessel welds and inside radius sections identified in Table 1, SNC proposes to examine a minimum of 25% of the nozzle-to-vessel welds and the inside radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702.

Basis for Use Balling Water Reactor (BWR) Vessel Internals Project (B~RVIP) has issued two topical reports:

BWRVIP-108 Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1003557, October 2002 (ML023330203) and BWRVI P-241 "Probabilistic Fracture Mechanics Evaluation for the Boling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," EPRI Technical Report 1021005, October 2010 (ML11119A041)

The BWRVIP-1 08NP report contains the technical basis supporting American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-702 "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds" for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 1 00 percent to 25 percent of the nozzles for each nozzle type during each 1 0-year interval.

BWRVIP-241 provides supplemental analyses for BWR RPV recirculation inlet and outlet nozzle-to-shell welds and nozzle blend radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation (SE) for the BWRVIP-108NP report.

Based on the two evaluations (BWRVIP-241 and BWRVIP-1 08NP), the failure probabilities due to a low temperature over pressure (L TOP) event at the nozzle inside radius section and the nozzle-to-vessel shell weld for HNP Units 1 and 2 recirculation inlet and outlet nozzles are very low and meet the NRC safety goal.

Regulatory Guide 1.147, Revision 17 conditionally accepts the use of Code Case N-702 with the following condition: The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18,2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April19, 2013 (ML13071A240) are met."

Section 5.0 of the NRC Safety Evaluation for BWRVIP-241 states:

E1-2 to NL-17-0955 Alternative HNP-ISI-AL T-05-05 "Licensees who plan to request relief from the ASME Code Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for use of ASME Code Case N-702 as an alternative.

However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:"

(1)

The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour HNP Unit 1 and Unit 2, Reactor Coolant System heatup and cooldown rates provided in the respective units' Pressure and Temperature Limits Reports (which is referenced in Technical Specification 3.4.9.1) are limited to a maximum of 1 ooaF when averaged over any one hour period and thus meets the requirement of condition 1.

Note: Inputs to address the SE conditions 2 through 5 are described below using bold text.

UNIT1 The HNP details are divided by Recirculation Inlet and Outlet Nozzles for Unit-1

{Responses (2) through (5)} and for Unit-2 {Responses (6) through (9)}.

Recirculation Inlet Nozzles (N2)

(2) p=

r=

t=

CRPV =

(pr/t)/CRPV ~ 1.15 RPV normal operating pressure (psig) (SNC conservatively used 1058 psig per HNP-1 Technical Specifications 3.4.10 for Reactor Steam Dome Pressure)

RPV inner radius (inch) (11 0.375)

RPV wall thickness (inch) (6.875) and 19332 (based on the BWRVI P-1 08 recirculation inlet nozzle/RPV finite element method (FEM) model);

HNP-1 specific calculations for Condition 2 above:

[(1 058 X 11 0.375)/6.875]/19332 = 0.88 ~ 1.15 The HNP-1 result is 0.88 and thus meets the requirement of condition 2 to be~ 1.15.

p =

RPV normal operating pressure (psig) (1 058) ro =

nozzle outer radius (inch) (10.219) ri =

nozzle inner radius (inch) (6.187) and CNozzLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model);

HNP-1 specific calculations for Condition 3 above:

[1058(10.2192 + 6.1872)/(10.2192 - 6.1872)]/1637 = 1.39 ~ 1.47 The HNP-1 result is 1.39 and thus meets the requirement of condition 3 to be ~_1.47.

E1-3 to NL-17-0955 Alternative HNP-ISI-AL T-05-05 Recirculation Outlet Nozzles (N1)

(4)

P=

r=

t=

CRPv=

(pr/t)/CRPV $. 1.15 RPV normal operating pressure (psig) (1058)

RPV inner radius (inch) (110.375)

RPV wall thickness (inch) (6.875) and 16171 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model);

HNP-1 specific calculations for Condition 4 above:

[(1 058 X 11 0.375)/6.875]/16171 = 1.05 ~ 1.15 The HNP-1 result is 1.05 and thus meets the requirements of condition 4 to be~ 1.15.

p =

RPV normal operating pressure (psig) (1 058) ro =

nozzle outer radius (inch) (19.75) n =

nozzle inner radius (inch) (12.75) and CNozzLE = 1977 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model).

UNIT2 HNP-1 specific calculation for Condition 5 above:

[1058(19.752 + 12.752)/(19.752 -12.752)]/1977 = 1.30 ~ 1.59 The HNP-1 result is 1.30 and thus meets the requirements of condition 5 to be~ 1.59.

Recirculation Inlet Nozzles (2N2)

(2)

P=

r=

t=

CRPv-(pr/t)/CRPV $. 1.15 RPV normal operating pressure (psig) (SNC conservatively used 1058 psig per HNP-2 Technical Specifications 3.4.1 0 for Reactor Steam Dome Pressure)

RPV inner radius (inch) (110.375)

RPV wall thickness (inch) (6.875) and 19332 (based on the BWRVI P-1 08 recirculation inlet nozzle/RPV finite element method (FEM) model);

HNP-2 specific calculations for Condition 2 above:

[(1 058 X 11 0.375)/6.875]/19332 = 0.88 ~ 1.15 The HNP-2 result is 0.88 and thus meets the requirement of condition 2 to be~ 1.15.

E1-4 to NL-17-0955 Alternative HNP-ISI-ALT-05-05 (3)

[p(ro2 + r?) I (ro2 - r?)]/CNoZZLE ~ 1.47 p =

RPV normal operating pressure (psig) (1 058) ro =

nozzle outer radius (inch) (10.219) n =

nozzle inner radius (inch) (6.187) and CNoZZLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model);

HNP-2 specific calculations for Condition 3 above:

[1058(10.2192 + 6.1872)/(10.2192 - 6.1872)]/1637 = 1.39 ~ 1.47 The HNP-2 result is 1.39 and thus meets the requirement of condition 3 to be ~1.47.

Recirculation Outlet Nozzles (2N1)

(4)

P=

r=

t=

CRPV=

(pr/t)/CRPV.$. 1.15 RPV normal operating pressure (psig) (1 058)

RPV inner radius (inch) (11 0.375)

RPV wall thickness (inch) (6.875) and 16171 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model);

HNP-2 specific calculations for Condition 4 above:

[(1 058 X 11 0.375)/6.875]/16171 = 1.05 ~ 1.15 The HNP-2 result is 1.05 and thus meets the requirements of condition 4 to be~ 1.15.

p =

RPV normal operating pressure (psig) (1 058) ro =

nozzle outer radius (inch) (19.75) ri =

nozzle inner radius (inch) (12.75) and CNozzLE = 1977 (based on the BWRVIP-108 recirculation outlet nozzle/RPV FEM model).

HNP-2 specific calculation for Condition 5 above:

[1 058(19. 752 + 12. 752)/(19. 752 - 12. 752)]/1977 = 1.30 ~ 1.59 The HNP-2 result is 1.30 and thus meets the requirements of condition 5 to be~ 1.59.

Conclusions The analyses for the HNP Recirculation Nozzles in BWRVIP-108 and BWRVIP-241 assume that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address the extended operating period. Based on analysis performed in support of the use of Code Case N-702, Structural Integrity (SIA) evaluated the beltline for 60 years based on the axial flux profile and the active fuel and nozzle elevations.

E1-5 to NL-17-0955 Alternative HNP-ISI-ALT-05-05 The plates and welds in the beltline remain the limiting materials for the period of extended operation. Therefore, the fluence assumptions used in BWRVIP-108 and BWRVIP-241 remain valid and are applicable to HNP.

The analyses in BWRVIP-108 and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth. HNP is projected to exceed the total number of pressure/thermal cycles used in the BWRVIP analysis during the extended operating period.

However, the usage factor for the Reactor Recirculation nozzles remains below 1.0.

Previous BWRVIP documents have demonstrated that stress corrosion crack growth represents the majority of the crack growth and that crack growth due to additional mechanical/thermal fatigue cycles introduced by the extended operation time is insignificant compared to hypothetical stress corrosion crack growth. Thus, the amount of thermal cycle driven fatigue crack growth due to the extended operation to 60 years is not a controlling factor in the probability of failure of the BWR reactor vessel nozzles.

SIA determined the maximum probability of failure (PoF) as 1.31x10"7 per year for the nozzle inside radius section and 1.5 x1 o-10 per year for the nozzle-to-vessel weld; both of these values are less than the 5x1 o-s per year criteria established in BW RVI P-1 08N P.

In addition to the justification provided above, a non-proprietary version of the SIA Calculation Number 1500231.303, "Probability Fracture Mechanics Evaluation for Hatch Units 1 and 2 Recirculation Inlet (N2) and Outlet (N 1) Nozzles" is included as Attachment 1.

Also, the most recent ultrasonic examinations of the nozzle to vessel welds and nozzle inside radius sections were performed with Appendix VIII techniques.

6. Duration of Proposed Alternative

This proposed alternative will be used for the Fifth Ten-Year Interval of the lnservice Inspection Program for HNP Units 1 and 2.

7. Precedents

Similar requests have been approved for:

Peach Bottom Atomic Power Station, Units 1 and 2 (Reference Accession Number ML112770217), dated January 24, 2012.

Columbia Generating Station (Reference Accession Number ML15036A220), dated February 13, 2015.

E1-6

Edwin I. Hatch Nuclear Plant Alternatives HNP-181-AL T-05-05 and HNP-181-AL T-05-06 to Enclosure 1 81A Calculation Probability Fracture Mechanics Evaluation for Hatch Units 1 and 2 Recirculation Inlet (N2) and Outlet (N1) Nozzles

~Structural Integrity Associates, Inc.'"

CALCULATION PACKAGE PROJECT NAME:

Hatch Units 1/2 N702 Relief Request CONTRACT NO.:

PO No. SNG19354-0021, PO Revision I, Contract Revision 5 CLIENT:

PLANT:

File No.: 1500231.303 Project No.: 1500231 Quality Program: [8] Nuclear D Commercial Southern Nuclear Operating Company Hatch Nuclear Power Plant Units I and 2 CALCULATION TITLE:

Probability Fracture Mechanics Evaluation for Hatch Units 1 and 2 Recirculation Inlet (N2) and Outlet (N1)

Nozzles I

I Document Revision 0

Affected Pages 1 -23 A A-2 Revision Description Initial Issue Project Manager Approval Si2nature & Date

~

Terry J. Herrmann PE TJH 12/28/16 Preparer(s) &

Checker(s)

Sian'ttures & Date Responsible Engineer:

~41r-s.s. Tang SST 12/28/16 Responsible Verifier:

VJ~~~

Wilson Wong ww 12/28/16 Page 1 of23 F0306-0IR2

lJ Structural Integrity Associates, Inc.<<

Table of Contents

1.0 INTRODUCTION

......................................................................................................... 4 2.0 OBJECTIVE.................................................................................................................. 4 3.0 METHODOLOGY........................................................................................................ 4 3.1 Fatigue Cycles................................................................................................... 4 3.2 Probabilistic Fracture Mechanics Evaluation.................................................... 5 4.0 DESIGN INPUT............................................................................................................ 6 4.1 Deterministic Parameters................................................................................... 6 4.1.1

/S/....................................................................................................................... 6 4.1.2 Stresses.............................................................................................................. 6 4.1.3 Fatigue Cycles...........................................................................1........................ 6 4.2 Random Variables............................................................................................. 6 4.2.1 Material Chemistry............................................................................................ 7 4.2.2 Fluence.............................................................................................................. 7 4.2.3 SCC Initiation.................................................................................................... 7 4.2.4 SCC Growth....................................................................................................... 7 4.2.5 Fatigue Crack Growth....................................................................................... 8 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS....................................... 9 6.0 ASSUMPTIONS............................................................................................................ 9 7.0 RESULTS OF ANALYSES........................................................................................ 10

8.0 CONCLUSION

S......................................................................................................... 1 0

9.0 REFERENCES

............................................................................................................ 11 Appendix A LIST OF SUPPORTING FILES...................................................................... A-1 File No.: 1500231.303 Revision: 0 Page 2 of23 F0306-0IR2

s; Structural Integrity Associates, Inc!

List of Tables Table 1 Deterministic Parameter Summary............................................................................ 13 Table 2: Probability of Detection Distribution [10]................................................................ 13 Table 3 Stress Coefficients..................................................................................................... 14 Table 4: Thermal Transients.................................................................................................... l5 Table 5: Lumping ofMonitored Cycles................................................................................. 16 Table 6: Random Variables Parameter Summary................................................................... 16 Table 7: Units I and 2 Nozzle and Vessel to Shell Material Chemistry................................. 17 Table 8: Fluence for Units 1 and 2......................................................................................... 17 Table 9: Gnits 1 and 2 PoF for Period of Extended Operation..........l.................................... 18 List of Figures Figure I: Stress Extraction Path Orientations in theN 1 (top) and N2 (bottom) Nozzle........ 19 Figure 2: Pressure Stress Distributions................................................................................... 20 Figure 3: Through-wall Stress Distributions, FWP Transient................................................ 21 Figure 4: Through-wall Stress Distribution, SSPC Transient................................................ 22 Figure 5: Weld Residual Stress Distributions for Paths 2 and 4............................................. 23 File No.: 1500231.303 Revision: 0 Page 3 of23 F0306-0IR2

e Structural Integrity Associates, Inc.~

1.0 INTRODUCTION

Southern Company intends to extend the applicability of Code Case N-702 [I] for Hatch Nuclear Plant Units I and 2 through the end ofthe period of extended operation (PEO). The Code Case allows reduction of in-service inspection from I 00% to 25% of all nozzle blend radii and nozzle-to-shell welds every I 0 years, including one nozzle from each system and pipe size, except for feedwater and control rod drive return nozzles.

Technical documents BWRVIP-108 [2, 3] and BWRVIP-241 [4] provide the basis for the code case, but only consider 40 year plant operation. In order to extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PFM) evaluation, consistent with the methods ofBWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure (PoF) remains acceptable.

The evaluation consists of two parts: Finite Element Model (FEM~ Stress Analysis and Probabilistic FracturJ Mechanics (PFM) Analysis. The FEM stress analysis is performed in separate calculations [5]

while this calculation package documents the PFM analysis.

2.0 OBJECTIVE The objective of the evaluations documented in this calculation package is to perform a plant specific analysis ofthe Hatch Units I and 2, recirculation outlet (N1) and inlet (N2) nozzles to extend applicability of the existing relief request to 60 years of operation.

3.0 METHODOLOGY This evaluation considers the nozzle-to-shell weld and nozzle blend radius on the Hatch Units 1 and 2 N I and N2 nozzles per Reference [3] and [ 4] and confirms that the nozzle still meets the acceptable failure probability considering the bounding fluence at the end of the PEO. The probability of failure expressed in this calculation is equivalent to the through-wall cracking frequency.

The acceptance criterion limits the difference in probability of failure per year due to the low temperature over pressure (LTOP) event to be no more than 5xl o-6 when changing from full (100%) in-service inspection to 25% inspection for the PEO. In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. If the resulting probability of failure per year due to an L TOP event (including I x 1 o-3 probability of L TOP event occurrence per year

[3, pg 5-13]) is less than 5xto*6, then the inspection reduction based on Code Case 702 can be extended to 60 years.

3.1 Fatigue Cycles In the FEM calculation [5], two bounding transients were defined to conservatively include fatigue crack growth contributions from all the normal and upset (Service Levels A and B) thermal transients defined for theN 1 and N2 nozzles. The N2 specific transient "Sudden Start of Pump in Cold Recirculation Loop (SSPC)" and Nl specific transient "Improper Start of Pump in Cold Recirculation Loop (ISPC)" [6]

have an instantaneous temperature down shock to 130°F from the normal operating temperature, before instantaneously returning to the normal operating temperature after only 34 and 26 seconds, respectively, bounding all Service Level AlB transients. Due to the severity of the transients, the vessel "Loss of File No.: 1500231.303 Revision: 0 Page 4 of23 F0306-0IR2

~Structural Integrity Associates, Inc.£ Feedwater Pump, Isolation Valve Closed (FWP)" transient [6] is selected to bound all other vessel transients for both the N I and N2 nozzles. The number of cycles used in the N-702 evaluation is defined in Section 5.0.

3.2 Probabilistic Fracture Mechanics Evaluation The probabilistic evaluation is performed for the case of25% inspection for the extended operating period (with zero inspection coverage conservatively assumed for the initial 40 years of operation).

For the nozzle blend radius region, a nozzle blend radius crack model from Reference [13] is used in the probabilistic fracture mechanics evaluation. For this location and crack model, the applicable stress is the stress perpendicular to a path defined 90 degrees from the tangent drawn at the blend radius.

For the nozzle-to-shell weld, either a circumferential or an axial crack, depending on weld orientation, I can initiate due to either component fabrication (i.e. cons~dering only welding process) or stress corrosion cracking. The probability of failure for a circumferential crack is less than an axial crack, due to the difference in the stress (hoop versus axial) and the influence on the crack model. However, this probabilistic fracture mechanics evaluation for the nozzle and vessel shell weld considers both circumferential and axial cracks (depending on weld orientation).

An axial elliptical crack model with a crack aspect ratio of all = 0.5 is used in the evaluation for the nozzle-to-shell weld. The inspection probability of detection (PoD) curve from BWRVIP-05 [8] (Table

2) is utilized with a ten-year inspection interval. The calculation of stress intensity factor is at the deepest point of the crack.

The approach used for this evaluation is consistent with the methodology presented in BWRVIP-05 [8].

A Monte Carlo simulation is performed using a variant of the VIPER program [9]. The Monte Carlo method can be used to solve probabilistic problems using deterministic computation. A mean value, standard deviation, and distribution curve as defined in the random variables summary (Table 6) defines a set of possible inputs and their probabilities of occurring. Using this domain of possible inputs, a set of inputs are generated for use in determining whether the nozzle will fail using conventional deterministic fracture mechanics methodology. This is repeated 2 million times. The number of simulations in which the nozzle is determined to fail divided by the number of simulations run gives the probability of failure.

The VIPER program was developed as part of the BWRVIP-05 effort for Boiling Water Reactor (BWR) reactor pressure vessel (RPV) shell weld inspection recommendations. The software was modified into a separate version, identified as VIPERNOZ, for use in this evaluation. The detailed description of the methodology incorporated in the VIPERIVIPERNOZ program is documented in References [8] and [3].

The modified software for this project is identified as VIPERNOZ to distinguish from the original VIPER software, and is verified on a project specific basis [7] to ensure the modifications made to the VIPER software are fully quality assured.

File No.: 1500231.303 Revision: 0 Page 5 of23 F0306-0IR2

~Structural Integrity Associates, Inc.*

4.0 DESIGN INPUT The plant specific input is described below. Section 4.1 presents all inputs modeled deterministically as constants while Section 4.2 describes the probabilistic treatment of inputs considered to be random variables (RV) in the VIPERNOZ code.

4.1 Deterministic Parameters Table I summarizes the dimensional and operational inputs used in the N-702 evaluation [5, 6].

Subsections 4.1.1 through 4.1.3 describe the more detailed input parameters used for in service inspection (lSI) interval, stress distributions and fatigue cycles, respectively.

4.1.1 lSI In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. The probability of detection (POD) distribution function associated with inspection is shown in Table 2 [I 0].

I 4.1.2 Stresses Stresses due to vessel pressure and bounding thermal transients are determined in previous calculations

[5]. In that calculation package, through wall stress distributions are presented at four locations in the region of the N2 nozzle for use in the N-702 evaluation. Figure 1 shows the locations and orientations of these four through-wall stress paths.

For vessel pressure, an internal pressure of 1,000 psig is applied to the inside surfaces ofthe RPV and N2 nozzle FE model. A bounding transient is also analyzed and the maximum cyclic stress ranges, based on a linearized through wall stress distribution, are identified. Figure 2 through Figure 4 show the distributions of the stress component acting normal to the crack plane (e.g. hoop or axial depending on the Path location) for the unit pressure, full power thermal expansion (steady state first load step of transient analysis) and the bounding transient load case step, respectively. Details of the analysis can be found in [ 5].

Table 3 presents the polynomial coefficients for the maximum and minimum through-wall stress distributions for all the N 1 and N2 nozzles for the 2 bounding thermal transients.

4.1.3 Jfati~e C7JVcles The thermal transients were obtained from the thermal cycle diagrams in Reference [6]. The number of cycles considered for each bounding transient is explained in Section 5.0.

4.2 Random Variables Random variables (RV) used in the N-702 evaluation are summarized in Table 6. Subsections 4.2.1 through 4.2.5 describe the more detailed input parameters used for SCC Initiation, SCC Growth and fatigue crack growth respectively. Table 6 identifies the specific references for each RV used in this N-702 evaluation.

File No.: 1500231.303 Revision: 0 Page 6 of23 F0306-0IR2

~Structural Integrity Associates, Inc.~

4.2.1 Material Chemistry The weld chemistries for the Nl and N2 nozzles and nozzle to vessel welds are presented in Table 7 along with the standard deviation and distribution used in the PFM evaluation.

4.2.2 Fluence.

The fluence for extended operation are obtained from Reference [6]. No best-estimated fast neutron fluence (> I.OMeV) was available at the recirculation inlet and outlet nozzle for the period of extended operation. Thus, the peak fluence at the inside vessel surface from the closest vessel circumference weld (C4) and the lower Shell were used. The fluence values are summarized in Table 8. The tluence values are provided for 49.3 EFPY, Unit I and 50.1 EFPY, Unit 2. These tluence values are taken to be applicable to the end of the extended period of operation of 60 years, consistent with [ 11].

4.2.3 SCC Initiation The cladding stress corrosion crack (SCC) initiation model in the VIPERNOZ program is a power law relationship. Since there is no cladding specific SCC initiation data, the cast stainless steel SCC data in a BWR environment is used as specified in Reference 8, Section 8.2.2.2, and used in References 3 and 4.

This model has the form; T = 84.2 *I 018 a -w.s (I) where:

T = time, hours cr =applied stress, ksi The residual plot shows that a lognormal distribution produces the best fit for the data. The lognormal residual plot with the linear fit of the data is shown below:

where:

4.2.4 SCC Growth

<1> = 0.9248x - 0.003

= (x - cr) I Jl cr = data mean Jl = data standard deviation X= In (TactuaJ/Tpredicted)

(2)

The SCC growth model in VIPERNOZ program is also a power law relationship [11]. The relationship used is; File No.: 1500231.303 Revision: 0 da = 6.82

• 1 o-12 K 4 dt (3)

Page 7 of23 F0306-0IR2

~Structural Integrity Associates, Inc.&

where:

da/dt = stress corrosion crack growth rate, inlhr K

= sustained crack tip stress intensity factor, ksi.,Jin The residual plot shows that a Weibull distribution produces the best fit for the data. The Weibull residual plot with the linear fit of the data is shown below:

where:

y F

X Y = 0.9085x - 0.3389

=In (In (II (1-F)))

1 cumulative distribution from 0 to I

=In ((da/dt) actual! (da/dt) predicted) 4.2.5 Fatigue Crack Growth (4)

The fatigue crack growth data for SA-533 Grade B Class I and SA-508 Class 2 (carbon moly steels) in a reactor water environments are reported in Reference [12] for weld metal testing at an R-ratio (algebraic ratio ofKminiKmax, "R") of0.2 and 0.7. To produce a fatigue crack growth law and distribution for the VIPERNOZ software, the data for R= 0.7 was fitted into the form of Paris Law. The R= 0.7 fatigue crack growth law was chosen for conservatism. The curve fit results of the mean fatigue crack growth law is presented with the Paris law shown as follows:

where da = 3.817*10-9{A.K)2.927 dn a = crack depth, in n = cycles

~K = Kmax - Kmin. ksi-in°*5 (5)

A comparison to the ASME Section XI fatigue crack growth law in a reactor water environment is documented in Reference [I 0] and it shows a reasonable comparison where the Section XI law is more conservative on growth rate at high ~K.

Using the rank ordered residual plot, it is shown that a Weibull distribution is representative for the data. The Weibull residual plot with the linear curve fit of the data is shown below:

y = -0.3712 + 4.15x (6)

File No.: 1500231.303 Revision: 0 Page 8 of23 F0306-0IR2

~Structural Integrity Associates, lnc.a where y=ln(ln(I/(1-F))

x = ln((da/dn)actua)l(da/dn)mean)

F =cumulative probability distribution 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS The stress analyses for the nozzle-to-shell weld and the nozzle blend radius for theN I and N2 nozzle are presented in Reference [5]. The stress analyses are performed for the load cases of unit pressure, and the bounding normal and upset (Service Levels A and B) thermal transient. The azimuthal locations evaluated are 0° and 90°, which also represents the symmetric un-modeled 180° and 270° locations of the nozzle. Two thr~ugh-wall sections are selected at each azimuthal location.

1 One is at the location of the weld between the RPV and nozzle and the other is at the blend radius locatibn of the nozzle.

The load cases analyzed for the N 1 and N2 nozzles include:

I. Unit pressure (1000 psi)

2. ISPC or SSPC Transient [6]

3. FWP Transient [6]

For the thermal transients, only the maximum or minimum through-waH linearized membrane plus bending stress profiles that produce the largest stress ranges for thermal fatigue crack growth are used in the evaluation. These through wall stress profiles are shown in Figures 3 and 4.

The nozzle ISPC and SSPC transients are the bounding Service Level AlB transients and used for all Level AlB thermal transients except those for the FWP and StartUp/ShutDown. The bounding number of cycles from both units are used in the PFM evaluation as shown in Table 5.

Weld residual stresses (WRS) are assumed present in the nozzle-to-shell welds. The WRS distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation. Figure 5 shows the assumed cosine distribution and the 3rd order polynomial fit used in the evaluation for Paths 2 and 4. No WRS is present in the nozzle blend radius region.

6.0 ASSUMPTIONS The following assumptions used in the evaluation are based on previous BWRVIP development projects.

Details of each assumption are provided.

1. Flaws are assumed to be aligned parallel with the weld direction as justified in BWRVIP-05 [8].

2. One stress corrosion crack initiation and 0.1 fabrication flaws is assumed per nozzle blend radius as justified in BWRVIP-108NP [3] and BWRVIP-108 SER [2].

File No.: 1500231.303 Revision: 0 Page 9 of23 F0306-0IR2

~Structural Integrity Associates, lnc.a

3. One stress corrosion crack initiation and 1.0 fabrication flaw is assumed per nozzle/shell weld as justified in BWRVIP-1 08NP [3].

4. The NRC Pressure Vessel Research Users' Facility (PVRUF) flaw size distribution is assumed to apply as justified in the W-EPRI-180-302 [10] report.

5. The weld residual stress distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation as justified in BWRVIP-108NP [3].

6. Upper shelf fracture toughness is set to 200 ksi.Vin with a standard deviation ofO ksi.Vin for on-irradiated material [2].

7. Standard deviation of the mean Ktc is set to 15 percent of the mean value of the Ktc as justified in BWRVIP-1 08 SER [2].

8. Zero inspection coverage conservatively assumed for the initial 40 years of operation.

7.0 RESULTS OF kNAL YSES The reliability evaluation is presented using plant specific inspection coverage. The probabilities of failure (Po F) per year due to the limiting L TOP event with 25% inspection for the extended operating term (with zero inspection coverage for the initia140 years of operation) are summarized in Table 9.

The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to L TOP events are both less than the 5 x 1 0-6 per year NRC safety goal from Reference [ 14].

Hatch Units 1 and 2 both performed a power uprate that increased operating pressures and temperatures (up to 1048 psig and 552°F) [19, 20]. These uprated numbers are not reflected in the thermal cycle diagrams provided in Reference [6]. However, since the pressure increase is less than 5%, the small effects of the power uprate is well within the margin of the acceptance criteria.

8.0 CONCLUSION

S The probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii in the Hatch Units 1 and 2 N I and N2 nozzles are below the acceptance criterion of 5x 1 o*6 per year. This analysis shows that theN 1 and N2 nozzles meet the acceptable failure probability even when considering elevated fluence level for reduced inspection using ASME Code Case N-702 to the end of the period of extended operation.

File No.: 1500231.303 Revision: 0 Page 10 of23 F0306*01R2

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9.0 REFERENCES

I. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (8 WR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20,2004.

2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction oflnspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19,2007, SI File No. BWRVIP.I08P.

3. BWRVIP-108NP: BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii.

EPRI, Palo Alto, CA: 2007. 1016123.

4. BWRV/P-241: BWR Vessel Internal Project, Probabilistic Fracture Mechanics Evaluation/or the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA.

1021005. fPRI PROPRIETARY INFORMATION.

I

5. SI Finite Element Analysis Calculations

a. 1500231.30 I, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N2 Nozzle," Revision 0.

b. 1500231,302, "Finite Element Model Development and Thermal/Mechanical Stress Analysis for the N1 Nozzle," Revision 0.

6. Southern Nuclear Response to SIA Design Input Request, Sl File No. 1500231.214.

7. SI Calculation 1500231.304, "Verification of Software VIPERNOZ Version 2.0," Revision 0, December 2016.

8. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.

9. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates.

10. SI Calculation W -EPRI-180-302, "Evaluation of effect of inspection on the probability of failure for BWR Nozzle-to-Shell-Welds and Nozzle Blend Radii Region," Revision 0.

11. NUREG/CR-6923, Appendix B.8, "Expert Panel Report on Proactive Materials Degradation Assessment," Published February 2007.

12. Bamford, W. H., "Application of corrosion fatigue crack growth rate data to integrity analyses of nuclear reactor vessels," Journal of Engineering Materials and Technology, Vol. 101, 1979, SI File No. 1300341.213.

13. Delvin, S. A., Riccardella, P. C., 'Fracture mechanics analysis of JAERI model pressure vessel test,'

ASME PVP Conference, Paper 78-PVP-91, 1978.

14. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), NUREG-1806, Vol. 1, August 2007.

File No.: 1500231.303 Revision: 0 Page 11 of23 F0306-0IR2

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15. USNRC Report, "Final Safety Evaluation ofthe BWR Vessel Internals Project BWRVIP-05 Report," T AC No. M93925, Division of Engineering Office ofNuclear Reactor Regulation, Nuclear Regulatory Commission, July 28, 1998.

16. EPRI Letter 2012-138, '*BWRVIP Support of ASME Code Case N-702 lnservice Inspection Relief,"

August 31,2012, Sl File No. 1300341.213.

17. SI RPV Adjusted Reference Temperature Calculations:

a. Calculation No. 1001527.301, 'Hatch Unit 1 RPV Material Summary and ART Calculation; Rev. 1, November, 2011

b. Calculation No. 1001527.302, 'Hatch Unit 2 RPV Material Summary and ART Calculation,'

Rev. 1, December, 2011.

18. S~ FatiguePro Calculations:

1 I a. Calc No. FP-HTCH-303, "Fatigue Update for Edwin I. Hatch Nuclear Power Plant, Unit 1,"

Revision 2, 2015.

b. Calc No. FP-HTCH-304, "Fatigue Update for Edwin I, Hatch Nuclear Power Plant, Unit 2,"

Revision 2, 2015.

19. GE Document 25A5541, Revision 3, "Reactor Vessel-Power Uprate," SJ File No. GPC0-42Q-237.

20. GE Document 25A5565, Revision 2, "Reactor Vessel-Power Uprate," Sl File No. GPC0-42Q-237.

File No.: 1500231.303 Revision: 0 Page 12 of23 F0306-0IR2

~Structural Integrity Associates, lnc.8 Table 1 Deterministic Parameter Summary VIPERNOZ Variable Value RPV Thickness 6.875 inches (excluding clad)

RPV Radius 110.375 inches (to vessel surface)

Clad Thickness 0.21875 inches Operating Temperature 522 °F{Region B)

L TOP Event Temperature J00°f Operating Pressure I

1005 psig L TOP Event Pressure I

1200 psig Table 2: Probability of Detection Distribution [10]

File No.: 1500231.303 Revision: 0 Flaw Size, in.

0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 0.60 Cumulative POD 0.00 0.10 0.46 0.80 0.92 0.95 0.98 0.99 0.99 1.00 1.00 1.00 1.00 Reference

[6]

[6]

[6]

[6]

[15]

[6]

[15]

Page 13 of23 F0306-0JR2

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Table 3 Stress Coefficients Nozzle Nl N2 File No.: 1500231.303 Revision: 0 Load Case PI.FWP. Max P2, FWP,Max P3,FWP,Max P4,FWP.Max PI, Pres P2, Pres P3, Pres P4. Pres PI, ISPC, Max P2, ISPC. Max P3, ISPC, Max P4, ISPC, Max PI,FWP, Min P2,FWP,Min P3,FWP,Min P4,FWP,Min PI,SS P2,SS P3,SS P4,SS P1, ISPC, Min P2, ISPC, Min P3, ISPC, Min P4, ISPC, Min PI, FWP, Max P2,FWP, Max P3,FWP, Max P4, FWP, Max PI, Pres P2, Pres P3, Pres P4, Pres PI, SSPC, Max P2, SSPC, Max P3, SSPC, Max P4, SSPC, Max PI, FWP, Min P2,FWP, Min P3, FWP, Min P4,FWP, Min P1,SS P2,SS P3. SS P4,SS PI, SSPC, Min P2, SSPC, Min P3, SSPC, Min P4, SSPC, Min Co Cl 19.04

-0.76 23.84

-3.79 19.92

-0.93 23.18

-3.37 43.12

-6.69 7.95

-0.29 5.68 0.96 1.55 2.64 51.55

-39.79

-16.48 21.09 59.31

-43.05

-18.35 23.92

-33.77 20.52

-3$.31 28.60

-34.54 22.49

-31.15 29.25

-22.82 16.93

-26.55 23.95

-23.93 19.15

-26.12 24.49

-22.92 16.98

-26.68 24.02

-23.93 19.15

-26.19 24.52 19.04

-0.18 21.75

-0.75 17.83 0.87 19.89

-0.68 45.11

-10.51 9.27 0.34 9.01 0.44 27.30

-2.44 59.41

-44.57

-23.44 27.91 60.97

-47.54

-20.56 24.87

-36.58 24.50

-40.95 33.54

-36.97 26.68

-37.73 29.91

-25.74 20.99

-29.06 28.76

-26.19 23.08

-26.82 25.64

-25.85 21.04

-29.21 28.83

-26.19 23.08

-26.91 25.68 C2 CJ

-138 0.12

-1.61 0.20

-1.61 0.15

-1.71 0.21 0.57

-0.03 0.14

-0.01

-0.11 0.01

-0.35 0.04 7.22

-0.39

-5.70 0.43 8.34

-0.48

-6.61 0.51

-3.41 0.18

-6.01 0.40

-4.06 0.23

-6.44 0.45

-3.23 0.18

-5.12 0.40

-3.96 0.24

-6.08 0.45

-3.23 0.18

-5.12 0.40

-3.96 0.24

-6.07 0.45

-1.91 0.18

-2.86 0.32

-2.65 0.24

-2.47 0.27 1.41

-0.10

-0.03 0.00

-0.16 0.01 0.27

-0.04 8.94

-0.53

-7.71 0.61 10.12

-0.64

-6.69 0.51

-4.58 0.27

-7.88 0.59

-5.42 0.35

-6.77 0.49

-4.48 0.28

-7.62 0.60

-5.32 0.36

-6.55 0.50

-4.49 0.28

-7.63 0.60

-5.32 0.36

-6.55 0.50 Page 14 of23 F0306-0IR2

s; Structural Integrity Associates, Inc.£ Table 4: Thermal Transients TransientiD Bolt Up Design Hydro Test Start Up Loss ofFeedwater Heater, Turbine Trip at25% Power (Unit 2)

Turbine Trip with 100% Stearn By Pass Loss ofFeedwater Heater, Feedwater Heater by pass Scram: Loss ofFeedwater Pumps, Isolation Valves Close Scram: Turbine Generator Trip Feedwater on Isolation Valves Stay Open Scram: Overpressure with Delayed Scram, Feedwater stays on, Isolation valves stay open Scram: Single Relief or Safety Valve Blowdown All other scrams Improper Start of Cold Recir Lo~

Sudden Start of Pump in Recirc Loop Reductio to 0% Power/Hot Standby/Shutdown-Vessel Flooding Pre-Op Blowdown Scram: Loss ofFeedwater Pumps Isolation Valves Close Scram: Reactor Overpressure with delayed Scram, feedwater stays on, isolation valves stay open Scram: Single relief or Safety Valve Blowdown Scram: Automatic Blowdown Improper start of cold recirc loop Sudden start of pump in cold recirc loop Improper startup with recirculation pumps off & drain shut off Grand total number of cycles (not included transients with zero internal cycles)

File No.: 1500231.303 Revision: 0 Unitt Service

1. .of Level Transients 125 130 120 10 I 70 10 40 I

2 147 5

5 118 783

1. of Total Service Internal Cycles Level Cycles I

125 Nonnall Upset I

130 Nonnall Upset I

120 Nonnall Upset 2

20 NonnaV Upset I

70 Nonnall Upset 4

40 I

40 I

I I

I I

147 I

5 I

5 1

118 NonnaV Upset Nonnall Upset Emergency Emergency Emergency Emergency Emergency Emergency Emergency 822 Unit2 No. of

1. of Total Transients Internal Cycles Cycles 125 I

125 130 I

130 117 I

117 10 I

10 70

)

70 40 2

80 140 I

140 111 I

111 10 I

10 5

4 20 1

2 2

8 I

8 I

I I

1 I

1 I

I I

1 I

I 771 827 Page 15 of23 F0306-0IR2

s; Structural Integrity Associates, lnc.a Table 5: Lumping of Monitored Cycles Unit 1 Unit2 Used in PFM Transient 37 per 33 per per

/10 years year years year year years StartUp/Shutdown 559 15.1 407 12.3 15.1 151 Loss ofFeedwater Pump, Isolation Valve Closed (FWP) 1601 43.2 1929 58.5 58.5 585 Sudden Start of Pump in Cold Recirculation Loop (SSPC) 5 0.14 1

.03.

0.14 1

Total 2165 58.5 2337 70.8 73.7 737 Table 6: Random Variables Parameter Summary Random Parameter I Mean Std Dev Distrfibution Ref.

Flaw density, nozzle/shell weld I perweld

...JMean Poisson

[3]

(fabrication)

Flaw density, nozzle/shell weld I perweld

...JMean Poisson

[3]

(SCC initiation)

Flaw density, nozzle blend radius 0.1 perweld

...JMean Poisson

[2]

(fabrication)

Flaw size (fabrication) n/a n/a PVRUF

[3]

Flaw size (stress corrosion)

Clad thickness n/a Constant

[31 Weld residual stress, through-wall 8

inside surface 5

Normal

[31 (ksi) cosine distribution Clad residual stress (ksi)*

32 5

Normal

[31 K1c upper shelf (ksi...Jin) 200 0

Normal

[2, I6,31 sec initiation time (hr) t = 84.2xi0 18(crY Residual Lognormal

[2]

10.5 y=0.9248x-0.0003 K deQendent da/dt = 6.82x I o-Residual Wei bull

[II 1

~

y=0.9085x-0.3389 SCCG (in/hr)

K>50 ksi...Jin K inde12endent da/dt = 2.8xio*6, na na

[II 1 K <50 ksi...Jin sec threshold (ksi...Jin)

IO 2

Normal

[21 Fatigue crack growth (FCG) da/dn=3.82 Residual Wei bull

[3]

(in/cycle) xI o-9( dK)2.927 y=4.155x-0.37I2 FCG threshold (ksi...Jin) 0 0

Normal

[3,3,3]

• Note: The mean clad stress used already includes the effects of post-weld heat treatment.

File No.: 1500231.303 Revision: 0 Page I6 of23 F0306-0IR2

~Structural Integrity Associates, Inc.&

Table 7: Units 1 and 2 Nozzle and Vessel to SheD Material Chemistry Random Parameter Mean Std Dev N1 Nozzleto

%Cu 0.35 0.045

%Ni 1.0 0.0165 shell weld Initial RT ndt ef)

-20 13 N1 Nozzle

%Cu 0.18 0.04407

%Ni 0.81 0.068 forging Initial RT ndt (

0 f)

+30 26.48

%Cu 0.10 0.045 N2 Nozzle to

%Ni 1.10 0.0165 shell weld lpitial RT ndt (

0 f)

-20 13 N2Nozzle

%Cu 0.18 0.04407

%Ni 0.81 0.068 forging Initial RTndt (

0 f)

+30 26.48 Table 8: Fluence for Units 1 and 2 Unit EFPY Weld Beltline ID Azimuth 49.3 2

50.1 File No.: 1500231.303 Revision: 0 C4 Distribution Normal Std Distribution Dev 20%

Normal Ref.

[6,17,2,3]

Ref

[6,3]

Page 17 of23 F0306-01R2

e Structural Integrity Associates, lnc.c Table 9: Units 1 and 2 PoF for Period of Extended Operation PoF due to L TOP Event Nozzle Location (with zero inspection for initia140 _years and 25% for PEO)

Conditional PoF PoF per year*

Allowable PoF per for60 years year [14]

Path I Nozzle Blend Radius 7.85 x w-3 t.3I x w-7 Path 2 Vessel to Shell Weld 9.0 X )0-6 1.5 x w-10 Nl Path 3 Nozzle Blend Radius 3.0 X )0-6 5.o x w-11 Path4 Vessel to Shell Weld 5.0 X 10-6 8.33 x w-11 5.0x 10-6 rath I Nozzle Blend Radius 5.25 x to-3 1 8.70 x w-8 Path2 Vessel to Shell Weld

< 5.0 x w-7 1 < 8.33 x w-12 N2 Nozzle Blend Radius 2.11 x w-10 Path3 1.3 x w-s Path4 Vessel to Shell Weld 6.0 X 10-6 1.00 x 1 o-10

• Note: Values include I x I o*3 probability of L TOP event occurrence per year [3, pg 5-13].

File No.: 1500231.303 Revision: 0 Page 18 of23 F0306-0IR2

s; Structural Integrity Associates, Inc!

Figure 1: Stress Extraction Path Orientations in the N1 (top) and N2 (bottom) Nozzle File No.: 1500231.303 Revision: 0 Page 19 of23 F0306-0IR2

~Structural Integrity Associates, Inc.<<

50 45 40 35 30

~

z; 25

~

~

20 15 10 5

0 50 45 40 35 30

~

~25

~

~

20 15 10 5

0 File No.: 1500231.303 Revision: 0

~

0 0

2 4

6 8

Distance from Inside Surface (m)

(a) Nl Nozzle 2

3 4

5 6

Distance from Inside Surface (in)

(b) N2 Nozzle Figure 2: Pressure Stress Distributions 10 Pl..Pft:t.

7 8

9 Page 20 of23 F0306-0IR2

S} Structural Integrity Associates, Inc.*

30 20 10 0

-20

-40

-so 20 10 0

-20

-30

• SO File No.: 1500231.303 Revision: 0 0

2 0

.... -.: $,*:~:::":"":.":.:".::.".:.~:;_:;;.~~;_~::_:"-.=~~.~.~.:.::~~:.*

Pl. FWP, Max P2, FWP, Max P3, FWP, Max P4, FWP, Max

• • • • • • • • • Pl, FWP, Min

- P2, FWP, Min

---

*P3, FWP, Min

-- - P4, FWP, Min 4

6 Distance from Inside Surface (in)

P2, FWP, Max P3, FWP, Max P4, FWP, Max Pl, FWP, Min

• • • • • • • • • P2, FWP, Min

- P3, FWP, Min

- * - P4. FWP, Min (a) Nl Nozzle

---

"::'.~

4 Distance from inside surface (in1 (b) N2 Nozzle 8

10 9

Figure 3: Through-wall Stress Distributions, FWP Transient 10 Page 21 of23 F0306-0IR2

e Structural Integrity Associates, Inc.<<

70 Pl, SSPC, Max 60 P2, SSPC, Max P3, SSPC, Max 50 P4, SSPC, Max 40 Pl, SSPC, Min P2, SSPC, Min 30 P3, SSPC, Min P4, SSPC, Min

'iii 20

~

~

~

~ 10

~

0

-10

-20

-30

-40 0

2 4

6 8

10 Distance from Inside Surface (in)

(a) Nl Nozzle 70 P1, SSPC, Max 60 P2, SSPC, Max so P3, SSPC, Max

-P4, SSPC, Max 40 Pl, SSPC, Min P2, SSPC, Min 30 P3, SSPC, Min

~ 20 P4, SSPC, Min

~

~ 10 In 0

-10

-20

-30

-40 0

1 2

3 4

s 6

7 8

9 Distance from Inside Surface (in)

(b) N2 Nozzle Figure 4: Through-wall Stress Distribution, SSPC Transient File No.: 1500231.303 Revision: 0 10 Page 22 of23 F0306-0IR2

s; Structural Integrity Associates, Inc.*

15 10 5

Ill Ill Ill QJ...

+"

Vl 0

-5

-10 0

File No.: 1500231.303 Revision: 0 1

2 y = -o.0012x3 + 1.0704x2 - 8.3874x + 10.788 R% = 0.9048 3

4 5

Distance from Inside surface (in) 6 Figure 5: Weld Residual Stress Distributions for Paths 2 and 4 7

8 Page 23 of23 F0306*01R2

s; Structural Integrity Associates, Inc.*

Appendix A LIST OF SUPPORTING FILES File No.: 1500231.303 Revision: 0 Page A-I of A-2 F0306-0IR2

lJ Structural Integrity Associates, Inc...

FileName N 1-Path I.INP NI-Path2.1NP NI-Path3.INP NI-Path4.1NP N 1-Path I.OUT N I-Path2.0UT NI-Path3.0UT

~I-Path4.0UT N2-Path I.INP N2-Path2.1NP N2-Path3.1NP N2-Path4.1NP N2-Path I.OUT N2-Path3.0UT N2-Path2.0UT N2-Path4.0UT VIPERNOZ_v2.EXE ISPCTPOD.EXE FL WDSTRB.EXE PostStress.xlsx File No.: 1500231.303 Revision: 0 Description VIPERN~Z input file for N I Path I at nozzle blend radii.

VIPERNOZ input file for Nl Path 2 at nozzle-to-shell-weld.

VlPERNOZ input file for N I Path 3 at nozzle blend radii.

VIPERNOZ input file for N I Path 4 at nozzle-to-shell-weld.

VIPERNOZ output file for Nl Path I at nozzle blend radii.

VIPERNOZ output file for Nl Path 2 at nozzle blend radii.

VIPERNOZ output file for N I Path 3 at nozzle-to-shell-weld.

VIPERNOZ output file for N ~ Path 4 at nozzle-to-shell-weld.

VIPERNOZ input file for N2 Path 1 at nozzle blend radii.

VIPERNOZ input file for N2 Path 2 at nozzle-to-shell-weld.

VIPERNOZ input file for N2 Path 3 at nozzle blend radii.

VIPERNOZ input file for N2 Path 4 at nozzle-to-shell-weld.

VIPERNOZ output file for N2 Path 1 at nozzle blend radii.

VIPERNOZ output file for N2 Path2 at nozzle blend radii.

VIPERNOZ output file for N2 Path 3 at nozzle-to-shell-weld.

VIPERNOZ output file for N2 Path 4 at nozzle-to-shell-weld.

VIPERNOZ executable program VIPERNOZ probability of detection curve input file VIPERNOZ flaw size distribution curve input file Stress postprocessing spreadsheet Page A-2 of A-2 F0306-0IR2

Edwin I. Hatch Nuclear Plant Alternatives HNP-ISI-ALT-05-05 and HNP-ISI-ALT-05-06 Alternative HNP-ISI-AL T-05-06 to NL-17-0955 Alternative HNP-ISI-ALT-05-06

1. ASME Code Component(s) Affected Code Class:

Component Numbers:

Code

References:

Examination Category:

Item Number(s):

2. Requested Approval Date ASME Section XI Code Class 1 Various ASME Section XI, 2007 Edition with 2008 Addenda, IWB-5221 (a) 8-P 815.10 Approval is requested by May 31, 2018

3. Applicable ASME Code Requirements 10 CFR 50.55a(b)(2)(xxvi) requires the use of the 1998 Edition, IWA-4540(c) for pressure testing of Class 1, 2, & 3 mechanical joints The 1998 Edition of ASME Section XI, IWA-4540(c) states: "Mechanical joints made in installation of pressure retaining items shall be pressure tested in accordance with IWA-5211 (a). Mechanical joints for component connections, piping, tubing (except heat exchanger tubing), valves, and fittings, NPS-1 and smaller, are exempt from the pressure test." Southern Nuclear Operating Company (SNC) understands that this means a pressure test is required for a mechanical joint when a new valve or flange greater than NPS-1 is installed as part of the repair/replacement activity, and does not include those items covered by IWA-4132 "Items Rotated From Stock."

Note that the 1998 Edition, IWA-5211 (a) states "a system leakage test conducted during operation at nominal operating pressure, or when pressurized to nominal operating pressure and temperature." SNC has defined this to be a minimum of 1 045 psig for components within the Reactor Coolant Pressure Boundary (RCPB).

The applicability for Code Case N-795 begins with the 1998 Edition with the 1999 Addenda and includes applicability to the 2013 Edition; although the 1998 Edition specified in 10 CFR 50.55a(b)(2)(xxvi) is not included in the published applicability, SNC believes that the comparison of IWB-5211 (a) from the 1998 Edition and IWB-5221 (a) of the 2007 Edition with the 2008 Addenda is compatible when the pressure has been defined specifically as described above. Therefore, SNC concludes that Code Case N-795 may be used for the 1998 Edition specified by the NRC condition found in 10 CFR 50.55a.

ASME Section XI, 2007 Edition with the 2008 Addenda IWA-4540(a) states: "Unless exempted by IWA-4540(b), repair/replacement activities performed by welding or brazing on a pressure-retaining boundary shall include a hydrostatic or system leakage test in accordance with IWA-5000, prior to, or as part of, returning to service. Only brazed joints and welds made in the course of a repair/replacement activity require pressurization and VT-2 visual examination during the test."

E2-1 to NL-17-0955 Alternative HNP-ISI-AL T-05-06 ASME Section XI, 2007 Edition with the 2008 Addenda IW8-5221 (a) states: The system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power."

4. Reason for Request

At the Hatch Nuclear Plant, Units 1 and 2 (hereinafter referred to as HNP), Class 1 pressure tests for repair/replacement activities (in accordance with IWA-4540) when performed after Table IW8-2500-1, Category 8-P testing has been completed, requires abnormal plant conditions/alignments. Testing at these abnormal plant conditions/alignments results in additional risks and delays while providing little added benefit beyond tests which could be performed at slightly reduced pressures under normal plant conditions.

Relief is requested from the test pressure requirement of IW8-5221 (a) (i.e., 1045 psig) on the basis of hardship as cited below.

Normal operating pressure (i.e. 1 045 psig) will not be reached for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 920 psig during the startup sequence:

o Control Rod Drive withdrawal limitations and the associated gradual increases in reactor power, pressure and temperature.

o Technical Specification-required Pressure versus Temperature Limitations.

o Main Steam line piping, turbine control and stop valve warming requirements.

o Main turbine warming requirements.

VT -2 leakage examination inside the drywell (primary containment) represents a hardship at the nominal operating pressure of 1045 psig during start-up because of high ambient and component temperatures.

o Data was retrieved for two startups (9/2013 & 3/2015) using instrumentation approximately 8ft. higher than the Main Steam Safety Relief Valves (SRVs).

Ambient temperature was approximately 144 degrees Fahrenheit once reaching 920 psig.

The data shows ambient temperature increases to approximately 150 degrees Fahrenheit over a 6-hour period while holding pressure steady at 920 psig.

o The local drywall temperature seen once nominal reactor pressure of 1045 psig is attained is approximately 170 degrees Fahrenheit or higher.

Reactor coolant system nominal operating pressure results in drywell ambient temperatures that require special safety precautions such as ice vests and cool air supply lines for personnel performing the VT-2 examinations.

Code Case N-795 is intended to provide alternative test pressure for certain Class 1 pressure tests using non-nuclear heat. The code case would be used following repair/replacement activities (excluding those on the reactor vessel) which occur subsequent to the periodic Class 1 pressure test required by Table IW8-2500-1, Category 8-P and prior to the next refueling outage on those components that cannot be isolated.

Components which can be isolated will be pressure tested at a pressure in accordance with IW8-5221 (a).

Performance of the Category 8-P pressure test each refueling outage places HNP in a position of significantly reduced margin, approaching the fracture toughness limits defined in the Technical Specification Pressure Temperature (P-T) Curves. To violate these curves E2-2 to NL-17-0955 Alternative HNP-ISI-AL T-05-06 would place the vessel in a Low Temperature Over Pressure (L TOP) condition. With strict operational control procedures, specific component alignment and operations staff training regarding LTOP, it may be considered acceptable to be at this reduced margin condition for the purpose of verifying the leakage status/integrity of the primary system in order to meet the ASME Section XI, Category B-P requirements prior to startup from a refueling outage.

However, to perform this evolution more frequently would increase the overall risks to the plant.

5. Proposed Alternative and Basis for Use

Proposed Alternative Pursuant to 10 CFR 50.55a(z)(2), an alternative is requested on the basis that compliance to the specified requirements would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.

SNC proposes to perform the system leakage test and associated VT-2 examination following repair/replacement activities on those components that cannot be isolated in accordance with Code Case N-795, while using longer hold times than specified in the code case. The system leakage test will be performed during the normal operational start-up sequence at a minimum of 920 psig [88% of the pressure required by IWB-5221 (a)] following a one hour hold time (for uninsulated components) and a six hour hold time (for insulated components) in lieu of the nominal operating pressure associated with 1 00% reactor power of approximately 1 045 psig.

Note that this code case is not applicable to Class 1 pressure tests performed to satisfy the periodic requirement of Table IWB-2500-1, Category B-P and is not applicable to pressure tests required following repair/replacement activities on the reactor vessel. SNC will continue to conduct the periodic system leakage tests required by IWB-2500-1, Category B-P at the end of each refueling outage at a pressure corresponding to 100% rated reactor power.

Basis for Use By the end of a normal refueling outage the core decay heat has had time to decrease and some spent fuel has been removed and some new fuel has been added. The result is a much lower decay heat load and much lower heatup rates. At the end of a normal refueling outage, the rate of temperature increase is able to be tolerated during the system leakage test. During normal performance of this system leakage test, the pressurization phase of the test is taken at a slow and very controlled pace. The pressurization phase normally takes several hours to reach test conditions.

However, following a maintenance or forced outage, there is a much larger decay heat load from the reactor core. That heat load is difficult to control once Shutdown Cooling (SOC) has been removed from service. Once SOC is removed from service, heatup starts immediately. During a short term mid-cycle shutdown, the projected heatup rate could be in the order of 0.5°F per minute. Under those conditions, the time available to pressurize up to test conditions, perform the VT-2 exam and return to SOC will be greatly reduced. The hurried time frames may create a more error-likely environment.

During short mid-cycle outages, the core does have a large decay heat load. Considering only the actions of isolating SOC from the vessel under high decay heat loads, there is some inherent risk. There would be some probability that once isolated, mechanical, control or operational problems could occur which could delay return to SOC.

E2-3 to NL-17-0955 Alternative HNP-ISI-AL T-05-06 The required VT-2 examinations performed following repair/replacement activities are limited to the areas affected by the work thereby allowing for a focused exam. The VT-2 exams, therefore, have a much smaller examination boundary than the periodic test.

Indication of leakage identified through the VT-2 examinations during a test at either the 1 00% rated reactor power level or at 88% of that value will not be significantly different.

Higher pressure under the otherwise same conditions will produce a higher flow rate but the difference is not significant. Code Case N-795 proposes increased hold times, as compared to a test performed at normal operating pressure, to allow for more leakage from the pressure boundary if a through-wall or mechanical joint leakage condition exists. Further, SNC proposes to implement longer hold times than specified by the Code Case. SNC believes these longer hold times are justified to allow for additional leakage to accumulate at the area of interest so as to be more evident during the VT-2 examination, should a through-wall or mechanical joint leakage condition exist. This alternate test pressure, when combined with longer hold times, will provide adequate evidence of leakage, should a leak exist.

While SNC does not expect that leakage will occur, any leakage will be related to the differential pressure at the point of leakage, or across the connection. A 12% reduction in the test pressure is not expected to result in the arrest of a leak that would occur at nominal operating pressure. In the unlikely event that leakage would occur subsequent to the VT-2 examination, at higher pressures associated with 1 00% rated reactor power, leakage would be detected by the drywell monitoring systems, which include drywell pressure monitoring, the containment atmosphere monitoring system, and the drywell floor drain sumps.

Leakage monitoring is required by Technical Specifications.

Code Case N-795 and the SNC proposed hold times allow for an adequate pressure test to be performed; ensuring the safety margin is not reduced due to VT-2 examination being performed at the slightly reduced pressure. There is no physical benefit withheld by testing at the slightly reduced pressure. The affected pressure boundary will be tested and will be otherwise fully capable of performing its intended safety function as part of the Reactor Coolant Pressure Boundary.

The use of Code Case N-795 will only be applied if the System Leakage Test required by IWB-2500-1, Category B-P has been completed for the cycle on components that cannot be isolated and will not be implemented for any repair/replacement activity performed on the reactor pressure vessel.

In summary, the proposed alternative is to perform the system leakage test and VT-2 examination in accordance with Code Case N-795 at 920 psig with a minimum hold time of one hour for uninsulated components and a six hour hold time for insulated components during maintenance, forced outages, or following the performance of the periodic pressure test required by Table IWB-2500-1, Category B-P during refueling outages. The provisions of this alternative are not applicable to the Examination Category B-P pressure test performed during refueling outages or to pressure tests of repair/replacement activities of the reactor pressure vessel or components that can be isolated. Considering the discussion above, SNC believes that this alternative will provide an acceptable verification of the leak integrity of the locations having repair/replacement activities performed without putting the plant in a non-conservative operational condition and without unnecessary radiation exposure and safety challenges to personnel.

E2-4 to NL-17-0955 Alternative HNP-ISI-ALT-05-06

6. Duration of Proposed Alternative

This proposed alternative will be used from approval date through December 31, 2025.

7. Precedents

1. 1 0 CFR 50.55a(a)(3)(ii) request was approved for Susquehanna Steam Electric Station, Units 1 and 2 Relief Request for the Fourth 10-Year lnservice Inspection Interval (TAC NOS. MF2705 through MF-2714) dated June 9, 2014 (ADAMS Accession No. ML14141A073).

2. 1 0 CFR 50.55a(a)(3)(ii) request was approved for Columbia Generating Station relief request 31SI-12 proposed alternative using Code Case N-795 (TAC NO. MF0319) dated August 9, 2013 (ADAMS Accession No. ML13191A054).

3. 10 CFR 50.55a(a)(3)(ii) request was approved for Monticello Nuclear Generating Plant relief from the requirements of the American Society of Mechanical Engineers code for the Fifth 10-Year lnservice Inspection Program Interval (TAC NOS. ME8068, ME8070, and ME8701) dated February 26,2013 (ADAMS Accession No. ML13035A158).

4. 10 CFR 50.55a(a)(3)(ii) request was approved for the MNGP during their Fourth 10-Year lnservice Inspection Interval as a one-time relief by NRC letter "Monticello Nuclear Generating Plant-One Time lnservice Inspection Program Plan Relief Request No. 8 for Leak Testing the "8" and "G" Main Steam Safety Relief Valves (TAC No. MB9538)",

dated June 13,2003 (ADAMS Accession No. ML031640464).

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