RA-18-072, Pressure and Temperature Limit Report Revision 1

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Pressure and Temperature Limit Report Revision 1
ML18187A199
Person / Time
Site: Oyster Creek
Issue date: 06/29/2018
From: Moore T
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-18-072
Download: ML18187A199 (26)
v • d • e


Text

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~~* Exelon Generation Oyster Creek 741 Route 9 South Forked River, NJ 08731 10 CFR Part 50 Appendix G RA-18-072 June 29, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Oyster Creek Nuclear Generating Station Pressure and Temperature Limit Report Revision 1 The purpose of this letter is to transmit the Pressure and Temperature Limits Report (PTLR)

Revision 1 for the Oyster Creek Nuclear Generating Station (OCNGS) in accordance with Technical Specification (TS) 6.23.c, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)."

The OCNGS PTLR was revised to update the Eflfective Full-Power Years (EFPY) to operate the station to the end of plant life.

Please direct any questions you may have regarding this matter to Mr. Gary Flesher, Regulatory Assurance Manager, at (609) 971-4232.

Respectfully,

~

Timothy A. Moore Site Vice President Oyster Creek Nuclear Generating Station

Attachment:

Oyster Creek Generating Station Pressure and Temperature Limit Report (PTLR)

Revision 1 cc Regional Administrator - NRC Region I NRC Senior Resident Inspector - Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station Manager, Bureau of Nuclear Engineering - New Jers~y Department of Environmental Protection Mayor of Lacey Township, Forked River, NJ

Oyster Creek PTLR Revision 1 Page 1 of 25 Exelon Nuclear Corporation Oyster Creek Generating Station Pressure and: Temperature Limits Report (PTLR) for 40 Effective Full'."Power Yea.rs (EFPY)

Prepared by: ~

~

I Program Manager Approved by: ~~

Engineering rograms Manager Concurred by: ~~~~~.;;..;.;a.'---

Corpora Date:

/S> / 1i /i S Date: It.I,zo/J t L

Date: ffhfe

section 1.0 2.0 3,0 4.0 5.0 6.0 FigUre 1 Figure2 Figure 3 Table 1 Table 2 Table 3 Table4 Appendix A Oyster Creek PTLR Revision 1 Page2 of25 Purpose Applicability Methodology Operating ~imits.

Discussion References Table of Contents Oyster Creek Pressure Test (Curve A) P-TCurve (40.EFPY)

Oyster Creek Core :No,t Critical (Curve B) p.;;r Curve ( 40 EFPY)

Oyster Creek Core Critical (Curve C)_ P-T Curve (40 EFPY)

Oyster Creek Pressure Test (Curve A) P-T Gurve (40 EFPY)

Oyster Creek Core Not Critical (Cur:ve B) P-T Curve (40 EFPY)

Oy~ter Cree~ Ccire Critical (Curve C) p.:T Curve (40 EFPY)

Oyster Creek ART C~lculations for 40 EFPY Oyster Creek Reactor Vessel Material Surveillance Programs 3

3 3

4 6

10 12 13 14

15.

18 21 24 25

Oyster Creek PTLR Revision 1 Page 3 of 25 PRESSURE AND TEMPERATURE LIMITS REPORT CPTLR)

FOR 40 EFFECTIVE FULL-POWER YEARS 1.0 PURPOSE The purpose of the Oyster Creek Generating Station (OCGS) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing; RCS Heatup and Cooldown rates; Reactor Pressure Vessel (RPV) head flange bolt-up temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Report SlR-05-044-A, Revision 1-A, contained within BWROG-TP::11-022-A, Revision 1 (CM-1) (Reference 6.1).

2.0 APPLICABILITY This report is applicable to the OCGS RPV.for 40 Effective Full-Power Years (EFPY). The following OCGS Technical Specification (TS) is affected by the information contained in this report:

TS Limiting Conditions for Operation 3.3 ("Reactor Coolant")

TS Surveillance Requirement 4.3 ("Reactor Coolant")

The Oyster Creek Reactor Vessel Pressure and Temperature Limits for 32 to 50 EFPY have been developed per Reference 6.2. Only the 40 EFPY limits are incorporated in this revision of the PTLR.

Future revisions of the PTLR must be revised per the 1 OCFR50.59 Review process as applicable.

3.0 METHODOLOGY The limits in this report were derived as follows:

1) The methodology used is in accordance with Reference 6.1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," Revision 1-A, incorporating the NRC Safety Evaluation in Reference 6.3.
2) The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190), Reference 6.4 using the RAMA computer code, as documented in Reference 6.5.

4.0 Oyster Creek PTLR Revision 1 Page 4 of 25

3) The adjusted ~eference temperature (ART) values for the limiUng beltline materials are calculated in accordance with NRG Regulatory C3uide 1.99, Revision 2 (RG 1.99), Reference 6.6, as documentect*in References 6.7 and 6.8;
4) The pressure and temperature (P-T) limits were calculated in accordance with the 8\\/VROG P-Tlimits topical report, Revision 0:-A (Refererice.6.9), asdocumented in Referen~es 6'.2 am;:16.8.However, the P-T limits in References Ef2 and 6.8 also meet.the requirements of

I the most recent NRC-:approved revision of the BWROG topical report, Revision 1-A (Reference 6.1 ),

  • 5) This revision of the prt9ssure and temperature *limits is to incorporate the following changes:

Revision O: Initial issue of PTLR.

Revision 1: Incorporated P-T limits for 40 EFPY and removed curves for 32 and 36 EFPY.

Changes to the curves, limits, or parameters ~ittliin this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions.in the Updated Firial Safety Analysis Report I

(UFSAR), can be m~de pursuant to 1o CFR50.59, provided the above methodologies are utilized.

The *revised PTLR shall be submitted to the NRri" upon issuan_c~.

Changes to the curves, Hmits, or parameters within this PTLR, based upon revised RPV fluence calculation methodology, cannot be maqe without prior NRG approval. Such analysis and revisions.

shail be submitted to the NRG for review prior to inc~rporation into the PTLR.

OPERA TING LIMITS The P-T curves included in this report represent steam dome pressure versus minimum vessel coolant temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C, in accordance with 10 CFR 50 Appendix G (Reference 6.10).

Complete P-T curves were.developed for 40 EFPY for OCGS, as documented in Reference 6.2.

Composite curves bounding all RPV component curves for Curves A, B, and C for OCGS for 40

Oyster Creek.PTLR Revision 1 Page 5 of 25 EFPY ar~ provided in Figures 1 through 3 of this report. Tabulation of the curves is included in Tables 1 through 3. The ART tables for the OCGS vessel beltline materials are shown in Table 4 for

. 40 EFPY (Reference 6.8).* The P-T Curves A, B, and C presented in this report each repr!3serit.a bounding curve fat the RPV beltline, upper vessel, and bottom. head re~ions. The resulting P-T curves8re based on the geometry, design and materials information for the OCGS vessel with the following conditions:

l-leatup and Copldown rate limit du~ing Hydrostatic and Class 1 Leak Testing (Fig~re 1:

Curve A): ~ 2~°F/h,our1.

Normal Operating H~atup and Cooldown rate limit (Figure 2: Curve B - nori~nuclear heating, and Figure 3: Curve C - nuclear heating): ~ 100°F/hour2.

  • RPV h~ad inst~ilation temperature limit (Figi;ire 1: Curve A-Hydrostatic and Class.1
  • Leak Testing; Figure 2: Curve B - non-nuclear heating): ~ 60°F.

Minimum temperature limits are set in accordance with 1 Cl CFR 50, Appendix G (Table 1 in Reference 6.10). Regarding the RPV head installation temperature limit, the minimum bolt-up temperature is selected ~o address the NRG condition in Section 4.0 of Reference 6.3 regarding lowest service temperature (LST) for all ferritic components of the reactor coolant pressure boundary (RCPB), including piping and other non-RPV components. The minimum temperature is set to 60°F for Cu~es A and B, which bounds the maximum RT~or for the closure flange m~terial, 36~F for the upper shell plate (Reference 6.2). The minimum criticality temperature is 96"F for Curve C, Which Is eq~ai:to RT Nor.max + 60°F.. However, from Reference 6.2, the non-bel.tline

/

(feedwater nozzle) P-T limits are more limiting than the closure flange limits for Curves Band C, and the minimum temperature is 76°F for Curve 1B and 116"F for Curve C, based on the non-beltline P-T limits. These temperatures a're consistent with the minimum temperature limits arid minimum bolt-up temperatures in the current docketed P-T curv~s (Reference 6.11, approved by the NRG in Reference 6.12). These temperatures also bou111d the non-RPV ferritic components of the RCPB, such as piping. Non-ductile fracture was considered in the design of ferritic RCPB piping, in accordance with the requirements of the piping code of construction and specifications identified in the OCGS FSAR Table 5.2-1 (Reference 6.13). Consequently, the P-T limits have considered all 1 Interpreted as the temperature change in any 1-hour period is less than or equal to 25°F, based on Reference 6.1.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to l00°F, the bounding heatup/cooldown rate on which the P-T curves are evaluated, discussed in Sectioau 5.0.

Oyster Creek PTLR Revision 1 Page 6 of 25 ferritic RCPB components, consistent with the requirements of 10CFR50 Appendix G, as described in Reference 6: 13..

5.0 DISCUSSION The ART of the limiting beltline. material is used to adjust the beltline P-T curves to account for irrad.iation effecti. HG 1.99 provides.the n,ethods for determining the ART (Reference 6.6). The RG..1.99 methods for dete;mining the limiting material and. adj~stirig the P-T curves using ART are' discussed in this.section. Appendix A in this report provides additional detail regarding RPV material surveill~nce, programs at Oyster Creek.

,,. ~

The OCGSRPv beitline copper (Cu) and nickel (Ni) values were obtained from the evaluatiol'!.of the QCGS veis~I plate and *weld.materials (Reference 6.7).. The Cu and Ni v~lue~ were used 1With, Tables 1 and 2:ofRG 1.99to'aeterrnirie a ch$misfryfactor (CF) per Regulatory Position*1.1 of RG.**

1_'99' for welds and plates, resp~ctively.

  • The peak RPV ID fluence used in the P-T curve evaluation for 40 EFPY is 5.69x1018 ri/cm2 for OCGS (Table 4): Fluence values were linearly int~rpol~ted for 40 EFPY b~sed on the fluence *.

values for 32 *and 50 EFPY provided in Reference 6:.5, which were calculated using' methods that com~ly with the guidelines of RG 1.190 (Referen~e 6.4).. This fluence value applies to the limiUng belti1nelower-:intermediate shell plate 564-03C for OCGS.. The ID fluence value was adjusted based upon anattenuation factor ~f 0.630 for a postulated 1/4Ula~: As a result, the 1/4t 40 EFPY fluence for the limiting lower-intermediate plate is 3.58x1018 n/cr'n2 for OCGS. The limiUng 1/4t ART for the OCGS beltline *for 40 EFPY is 172.5°F.

The RPV fluence values in Reference 6.5 were used as input to the currently docketed P-T curves.

An updated fluence evaluation for the OCGS RPV was subsequently performed in Reference 6.14.

Based on review of the updated fluence values for 40 EFPY, the previously calculated limiting 1/4t ART value bounds the limiting 1/4t ART calculated using the updated fluence. Accordingly, the 40 EFPY ART values previously documented in Reference 6.8 are provided in Table 4 and used as input to the P-T curves.

The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. The latest NRG-approved revision of the BWROG P-T limits topical report, Revision 1-A (Reference 6.1 ), was developed to incorporate additional requirements related to the effects of nozzles within the beltline region. Revision O of this

. Revision 1 Page 7 of 25 PTLR (Reference 6.11) and the supporting P-T curves calculation (Reference*6.2) predate the approval of the Revision 1-A BWROG topical report and were prepared to an earlier revision of the topical report, Revision O:A (Reference 6.9). However; the OCGS P-T limits developed in Reference 6.2 already appropriately addressed the effects of RPV nozzles. The limiting.RPV nozzle

\\

configuration, the feedwater (FVV) nozzle, is considered in the evaluation of the non-beltline (upper vessel) region P-T limits. The OCGS small bore instrument nozzles are outside the be!tline region and are bou.nded by the non~beltline/FW:nozzle P-T limits, as agreed by the NRC in Reference.

6.12: OCGShas.no.notiles withiffthe RPVbeittiin~. Ther~fore'. the 40 EFPY P-T limitsare alS'O

  • consistent withth~ latest NRG-approved BWROG topical report methodology (Reference 6.1 ).

Accordingly, both topical reports are referenced in this PTLR..

The.P-T curves for the core not critical and core critic~I operating conditions at a given EFPY apply.

for both the 1/4t and 3/4t locations: When combi1riin'g pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4t location (in~ide surface flaw) and the 3/4t location. *

(outside surface flaw). This is because the thermal gradient tensile stress of intereistis in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification; the thermal gradient stress at the 1/4t location is assumed to be. tensile for both*

heatup and cooldown. This results in the approach of applying the maximum tensile stress at the*.

1/4t location. This approach is conservative because irradiation effects cause the allowable toughness ~t 1/4t to be less than. that at 3/4t for a given metal temperature. This approa~h causes no operationaLdifficulties, since the BWR is at steam saturation conditions during normal operation, which is well below the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves are developed based on a coolant heatup and cooldown temperature rate of~ 100°F/hr for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/B RPV thermal transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of::;; 25°F/hr must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RT Nor, the chemistry (weight-percent Cu and Ni) and adjusted reference temperature at the 1/4 thickness location for all RPV beltline materials significantly affected by fluence (i.e., fluence

  • Oyster Creek PTLR Revision1 Page 8 of 25

> 1017 n/cm2.for E > 1 MeV) are shown in Table 4 for 40 EFPY. The initial RT NDT values shown in Table4 (obtained from-Reference 6.15) Were developed usitig the procedures of Branch Technical P~sition MTEB 5'.'2 in Standard Revievv Plan 5.3.2 in. NUREG-0800, and they have been previously.

approved for use by the. NRG (Reference 6.16)..

  • Per the Bornng VVater Reactor Vessel and lntemals Project (BWRVIP) lntegratedSurveilla~ce. :..
  • Program (ISP), representative weld and plate surveillance materials data for OCGS'were revielAled from BWRViP-135, Revision 3 (Reference 6.17), and iri: accordance \\/Vith Appendix A of Reference.

. 6:1. Use of the BWRVIPISP forOCGS was approved by th'e NRG in Reference 6:18., The.

  • representative heats of weld and plate material in _the ISP are:riot.the. same as thetarget weld and plate heats in the vessel, and no surveillance heats are present in the ()CGS b~ltlirie. Therefore, the chemistry factors (CFs) from the tables in Regulatory Guide -1.99, Revision2 (Ref~rence 6.6),

I were used in the determination *of the ART values' for all materi~ls for the OCGS vessel>

The only computer cod~ us~d in the de~ermination of the OCGS P-T curveswas the*A~sys***..

(Release 8.1 with Service Pack, 1) finite element computer program for the feedwater nozzle (non-beltline) stresses. This analysis was performed to deter~ineJhrough-wall thermal and pressure

  • stress distributions for the OCGS feedwater nozzl_es due to a step-change thermai transient.*

(Refe~ence 6.19). The ANSYS program was controlled under the vendor's 1 o CFR 50 Appendix B

  • . Quality Assura~ce Program for nuclear quality-related work. Benchmarking consisterit_with NRG GL 83-11, Supplement,1*.*(Refere~ce 6.20) was~performed*as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations* for several probiems.
  • The following inputs were used,as input to the finite element analysis:

With respect to operating conditions, stress distributions were developed for a thermal shock of 450°F, which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions. The stress results for a 450°F shock are appropriate for use in developing.the non~beltline P-T curves based on the limiting feedwater.nozzle, as a shock of 450°F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 100°F/hr startup transient. Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 1 Q0°F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region. The boundary integral equation/influence function (BIE/IF) methodology as presented in Reference 6.1 was used in Reference 6.2 to calculate the thermal stress intensity factor K11 by fitting a third order polynomial equation to the path stress distribution for the thermal load case.

Oyster Creek PTLR Revision 1 Page 9 of 25 Heat transfer coefficients were calculated from the governing design basis stress report for theOCGS feedwater nozzle ~nd from a model' of the heat transfer coefficient as a function of flow rate. The heat transfer coefficients were evaluated.at flow rates that bound the8ctual operating conditions in the feedwater nozzles at OCGS. *.

  • . With respe9tJo µ*res.sure stress, a unit pressure of 1,000 psig was applied to the internal surfaces of the fi~ite element* model. The pressure stress distribution* was taken along.

the sam~ path a~ the thermal stress distribution. The BIE/IF methodology presented in Reference 6~1 was used in Reference 6.2 to calc'ulate the pressure stress intensity factor

. K1p by fitting a t~ird order polynomial equ~tion tothe path stress distribution for the pressure load case.The resultirig.K1l!Jcan be linearly scaled to determine.the K1p f~r various RpV 'internal ~ressures.

A two~dimensional, axisymmetric finite elementmodel of the feedwater nozzle was constructed using the same modeling techniques that were employed to evaluate the feedwater nozzle in the governing d~ign basis stress report. Te> model the feedwater nozzle using a two~dimensional model, the analysis was performed as a penetration iii a

. sphere and not in a cylincie~. To account for three;.cJimens.ional effects on the pressure stresses at the nozzle blend radiu~. a con\\/ersion factor of 3.2 ti~es the' cylinder radius was used to model the sphere (Reference 6.19). Material properties were evaluated at 325°F to conservatively bound the 100°F condition where the maximum stresses*

occurred.

6.0 6.1 6.2 6.3 6.4 6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 6.13 6.14 6.15 6.16 Oyster Creek PTLR Revision 1 Page 10 of 25 REFERENCES Licensing Topical Report (LTR) BWROG-TP-11-022-A, Revision 1 (SIR-05-044, Revision 1-A),

"Pressure-Temperature Limits RepOrt Methodology fdr Boiling Water Reactor~tAugust 2013, ADAMS Accession No.ML13277A557/

Structural lrite~rity Associates; Inc. Calculation No.. QC-,05Q-313,.Revision 3, "Revised P-T Curves Based on New Fluence,!' Novemb~r 28, 2007.

U.S. NRC Letter.to BWROG dated May:.1a, 2013, "Final Safety Evaluation for Bomng Water Reactor Owners;, Grqup Topical Report. BWIROG-TP-: 11-022,. Revision. 1, November 2011,

  • 'Pressure.,TemperatufeUmitsRepor:t Methodology for Boiling Water Reactors"' (TAC NO.

l\\/1E7649, ADAMS Accession No. ML13107A062).

U. S: Nuclear Regulatory Commission Regulatory Guide 1*.190, IICalculational and Dosirnetry Methods for Determining Pressure Vesse.1 Neutron Fluence," March 2001.

TransWare Enterpris~s Inc. Report No. EXL~IFLU-001-R:.002, RevisionO, "Flu~nce Evaluation for Oyster Creek Reactor Pressure Vessel,l!SI File No.,OC-'050~257:.

U. S. Nucl~at Regulatory Commission Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May.1988.

Structural *Integrity Associates, Inc. Calculati~n No. OC-05Q-301, Revi~ion* 1, "Adju§lted.

Reference Temperature Evaluation," May 11, 2006.

Letter T JG~08:.001 from T.J: Griesb~ch (SI) to Greg Harttraft (Exelon), Revised Calculation of*

P-T Limit cu*rves for the Oyster Creek*Generating Station, dated February 26, 2008.

SI Report NQ. SIR.-05~044~A. Revision 0, "Pressure-Temperature Limits Report Methodology for

10Q-401.

Pa~SOofTitle 10 ~fthe Code ofFedei'al R~guiatiohs, Appendix G, "Fracture Toughness Requirements'."

. to AmerGe~ Letter No: RA-08-004, "Oyster Creek Generating Station, Pressure and Temperature Limits Report (PTLR) for 32 and 36 Effective Full-Power Years (EFPY),"

Exelon Nuclear Corporation, Revision Oc, March 4, 2008 (ADAMS Accession No. ML080740287).

Oyster Creek Nuclear Generating Station License Amendment No. 269, "Relocation of Pressure and Temperature Curves to the Pressure and Temperature Limits Report," September 30, 2008 (ADAMS Accession No. ML082390685).

Oyster Creek Nuclear Generating Station Updated Final Safety Analysis Report, Section 5.2, Revision 17, October 2011. SI File No. 1601200.201.

TransWare Enterprises Inc. Report No. OYC-FLU-001-R-005, Revision 0, "Non-Proprietary Version of Oyster Creek Generating Station Reactor Pressure Vessel Fluence Evaluation," SI File No. 1401181.211.

General Electric Report GENE-B13-01769, "Pressure-Temperature Curves per Regulatory Guide 1.99, Revision 2 for the Oyster Creek Nuclear Generating Station," July 1995, SI File No.

OC-05Q-210.

Letter from Pao-Tsin Kuo (U.S. NRG) to Mr. Timothy Rausch (AmerGen Energy Company, LLC), "Safety Evaluation Report Related to the License Renewal of Oyster Creek Nuclear Generating Station," Docket No. 50-219, dated March 30, 2007.

Oyster Creek PTLR

  • Revision 1.

Page 11 of 25 6.17 BWRVIP-135, Revision 3: BWR Vessel and lntemals Project, Integrated Surveillance Program (ISP) Data Source BookandPlant Evaluations. EPRI, Palo Alto, CA: 2014. 3002003144. EPRI PROPRIETARY INFORMATION.

6.18 Letterfrom P. S, Tam (NRC) to C. M. Crane (AmerGen Energy Company; LLC), "Oyster Creek

.Nuclear Generating $tation (OCNGS)- Issuance of Amendment RE: Use of Integrated Surveillance* Program for Reactor Vessel Specirnen Surveillance (TAC NO, MB700q)", dated

, April 27, 2004:,

6.19 Str~cturallntegrity,Associates Calculation No, OC-050;307, Revision 0, "Feedwater Nozzle 6.20 6.21 6.22 6.23 6.24 Green's Functions, July 20, 2005.

U: S. Nuclear Regui~tory C~m~ission, Gerieric L~tter 83-11, Supplement 1, "Licensee

.* Qualification for Performing Safety Analyses/ June 24, 1999.

Part 50 of Title 10 of the Code of Federal Reguiations, Appendix H, "Reactor Ves'sel Material Surveillance Program Requirements,"

GPU NuclearTechnical Data Report TDR-725, Revision 4, "Testing and Evaluation of Irradiated Reactor Vessel Materials Surveillance Program.Specimens," January 3, 19,96, SI File No. OC-050-219.

Manahan, M. P.,' et aL, Examination, Testing, and Evaluation of Specimens fro~ the 210° Irradiated Pressur~Vessel Surveillance Capsule for the Oyster Creek Nuclear Generating Station, Battelle Columbus Laboratories Report BCL-382-85-1, Rev. 1, October 1985, SI File No. GPUN-270-215.

BWRVIP-86, Revision 1-A:BWR Ve;sel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Altq, CA: 2012. 1025144. EPRI PROPRIETARY INFORMATION.

Oyster Creek PTLR Revision 1 Page 12 of 25 Figure 1: Oyster Creek Pressure Test (Curve A) P-T Curve (40 EFPY) 1,400.

1,300 1,200 1,100 ci 1,000

'iii

a. --

c

~ 900

c C.

0 1-800

..J w

en en

~ 700

~

0 I-

~ 600

~

~

I-

~

i w

~

en en w

~

C.

500 400 300 200 100 0

~

0 l,

. j. l I I I

/

Bolt-up Temp 60°F 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE {°F)

Oyster Creek PTLR

. Revision 1

. Page 13 of 25 Figure 2: Oyster Creek Core Not Critical (Curve B) P-T Curve (40 EFPY) 1,400 1,300

. 1,200.

1,100

_ 1,000 C) l -

C 90Q

~

I:

D..

g 800

.;.I w

en.

Cl)

~

  • 700 0:: g

~ 600

£111::

~

~

i LIJ 0::
J en en w

0::

D.

500 400 300 200 100 0

0 50 r

'-L I,

J I

I l

r*

I Bolt-up Temp = 60°F I

Minimum Non-Beltline Temp = 76°F J

100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Cl

'iii

a. -

C t5

i::
0.

0 I-

..J w

U)

U) w a::

0 I-0 t5 a::.

6:

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J w a::

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Oyster Creek PTLR Revision 1 Page 14 of 25 Figure 3: Oyster Creek Core Critical (Curve C) P-T Curve (40 EFPY) 1,400 I

1,300 1,200 1,100 1,000 900 I I I

~

800 700 600 500 J

I

/

I I

400 300 200 I

I Minimum I

100 Criticality J

Temp 116°F I

0 0

50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Oyster Creek PTLR Revision 1 Page 15 of 25 Table 1: Oyster Creek Pressure Test(Curve A) P-T Curve (40 EFPY)

Temperature Pressure for for P-T Curve P-T.Curve

{OF)

{~sig}

60 0.0 60 375.0 62 375.0 64 375.0 66 375.0 68 375.0 70 375.0 72 375.0 74 375.0 76 375.0 78 375.0 80 375.0 82 375.0 84 375.0 86 375.0 88 375.0 90 375.0 92 375.0 94 375.0

96.

375.0 98 375.0.

100 375.0 102 375.0 104 375.0 106.

375.0 108 375.0 110 375.0 112 375.0 114 375.0 116 375.0 118 375.0 120 375.0 122 375.0 124 375.0 126 375.0 126 412.4 126 419.6 126 427.0 126 434.7 126 442.7 126 451.1 126 459.8 126 468.8 126 478.3 126 488.1 126 498.3

Temperature for P-T Curve (OF}

126 126.

1.26 126 126 126 126 126 126 126 126 126.

  • 126 126 126 125*

126 126 126 126 126 128 130 132 134 136 138 140 142 144 146 148 150 152 154 156 158 160 162 164 166 168 170 172 174 176 178 180 Pressure for P-T Curve

(~sig}

508.9

  • 520.0 531.5 543.5

.555.9.

568.9 582.4 596.5 611.1 626:3 642.2 658.7 675.9 693.7 712.3 725.2

  • 725.2 725.2 725.2 725;2 725.2 731.2 737.5.

744:o*

750.8 757.8

. 765.2 772.8 780.8 789.0 797.7 806.6 816.0 825.7 835.8 846.3 857.3 868.7 880.5 892.9 905.8 919.1 933.1 947.6 962.7 978.4 994.7 1011.7 Oyster Creek PTLR

  • . Revision 1 Page 16 of 25

Temperature for P-T Curve

{OF}

182.

184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220

.222 224 Pressure for P-T Curve

{~sig}

1029.4 1047.8 1067.0 1087.0 1107.8 1129.4 1151.9 1175.3 1199.7 1225.1 1251.5 1279.0 1307.6 1337.4 1368.4 1400.7 1434.2 1469.2 1505:6 1543.4 1582.8 1623.8 Oyster Creek PTLR Revision 1 Page 17 of 25

Oyster Creek PTLR

  • Revision 1
  • Page 18 of 25 Table 2: Oyster Creek Core Not Critical (Curve B) P-T Curve (40 EFPY)
  • Temperature Pressure for for P-T Curve*. P-TCurve (OF)
  • <psig) 60 0.0 60 O'.O 62 0.0 64

. 0.0 66 0.0 68 0.0 70 0.0 72 0.0 74 0.0 76 0.0 78 3.2.'

80 10.2 82 17.6 84 25.2

86.

33.2

.88 41.5 90_

50.1 92 59.1 94

  • 68.5 96

'78.2 98 88.3 100

-98.9 102 109.9 104 12f3 106

.133.2 108 145.5 110 158.4 112 171.8 114 185.8 116 200.3 118 215.4 120 231.2 122 247.5 124 264.6 126 282.3 128 300.8 130 320.0 132 340.0 134 360.8 134 375.0 136 375.0 138 375.0 140 375.0 142 375.0 144 375.0

Temperature for P-T Curve (OF}

. 1.46 148

.

  • 150 152 154.

156 156 156 156 156 156

'156 156.

156.

156 158 *.

160 162

  • 164 166 168 170 172 174 176 178 180 182 184 186 188 190 192 194 196 198 200 202 204 206 208 210 212 214 216 218 220 222 Pressure for P-T Curve

(~sig}

  • 375.0 375.0 375;0.

375.0 375.0 375.0 428.5 452.9 455.3 455.3

. 455.3 455.3 455.3 455.3 455.. 3

.463.5 472.0 480.9 490.2 49,9.9 509'.9 520.3 531.2 542.5 554.3 566.. 6 579.3

  • 592.6 606.4 620~8 635.8 651.4 667.6 684.5 702.0 720.3 739.4 759.2 779.8 801.3 823.6 846.8 871.0 896.2 922.4 949:7 978.1 1007.7 Oyster Creek PTLR Revision 1 Page 19 of 25

Temperature for P-T Curve

{OF) 224

  • 226 228 230 232 234 236 238 240 242 244 246 248
  • 250 252 Pressure for P-TCurve

{~sig}

1038.4 1070.4 1103.8 1138.4 1174.5 1212.1 1251.2 1291.9

  • 1334.3 1378.4 1424.3 1472.0 1521.7 1573.5 1627.3 Oyster Creek PTLR Revision 1 Page 20 of 25

Oyster Creek PTLR Revision 1 Page 21 of 25 Table 3: Oyster Creek Core Critical (Curve C) P-T Curve (40 EFPY)

Temperature Pressure for for P-T Curve P-T Curve (OF)

(psig) 100 0

100 0

102 0

104 0

106 0

108 0

110 0

112 0

114 0

116 0

118 3

120 10 122 18 124 25 126 33 128 42 130 50 132 59 134 68 136 78 138 88 140

99.

142 110 1,44 121 146 133 148 146 150 158 152 172 154 186 156 200 158 215 160 231 162 248 164 265 166 282 168 301 170 320 172 340 174 375 176 375 178 375 180 375 182 375

Oyster Creek PTLR Revision 1 Page 22 of 25 Temperature Pressure for for P-T Curve P-T Curve

{OF}

{~sig}

184 375 186 375 188 375 190 375 192 375 194 375 196 375 198 463 200 472 202 481 204 490 206 500 208 510 210 520 212 531 214 543

  • 216 554 218 567 220 579 222 593 224 606 226 621 228 636 230 651 232 668

. 234 684 236 702 238 720.

240 739 242 759 244 780 246 801 248 824 250 847 252 871 254 896 256 922 258 950 260 978 262 1008 264 1038 266 1070 268 1104 270 1138 272 1175 274 1212 276 1251 278 1292

Temperature for P-T Curve (OF) 280 282 284 Pressure for P-T Curve (psig) 1334 1378 1424 Oyster Creek PTLR Revision 1 Page 23 of 25

Oyster Creek PTLR Revision 1 Page 24 of 25 Table 4: Oyster Creek ART Calculations for 40 EFPY (Reference 6.8)

-~~-~~~/.: ~ ':.,r:~~~:'.~f;> :. ~-.:*~;./,~~~-~ -.?' ;,:::~:_?:,. :*~:>;, ;(.;~ic;;L*2.:.. ~l!.~~ :~:*~:. -~~" ~-;.. ~;;*'~~'.'~ / '!* -::;":~i;[(.r ~- ;}:P~'1:;ES~~-'.-;,::;r:.:x... ~~~~/~\\:~"~J-; :*\\*:~~~;;~tiJ~'.!)::;i~%;~~~;.':}{~-~~~t~ 5:;_:;~~~::~:7~'. ~~:~tf; -~:J~;::**-~~;. * ~~J1;x~:* ;)?.. :,L:,.~

HEAT NO. I ADJUSTMENTS FOR MARGIN AND PIECE CODE FORGING INITIAL CHEMISTRY BEL TLINE IRRADIA,ioN SHIFT NO.

NO.

SIN RTmn Cu NI CF ID Fluencliil Attenuation 1/4~T Fluanca FF ART110r a0 o1 Mirgln ART PART NAME (Note 3)

'F o/,,

°F 1019 ri/cm2 1018 n/cm2

'F

  • tp op

'F.

'F 564-03A G-8-7 P-2161-1 17 O.at 0.48 139.4. 6.69E+l8 0.630 3.50E+l8 0.717 99.9.

17.0 10.7 40.2 167.1 LoWer lnterm~1ate Shell Plates.

HN*-w.,::039******** ****-*-a.*a:a*.. -*.............. p:21as:2*-*-**""'......... a""....... 0.12 0.46 '120.1 6.89E-t-f8........... 0.830 *-*-"...... 3.68E1'18........ 0. 717..** 86.5.... *11.o

  • 120..... 42.e'"' -* 137. t _..

':--:664'-03C:*::*e1:Vc,.B,fi;*.:,,**: ;es!-*/.P-2161);.kcc" _'.\\'.:.'31,:,;.,. ;;o:~<i :o:61.. :13.9,z: i6.68E~18e' '\\" 0.830.'0'(,- ;"'}3.68Efl9X:.,0;1,r-.,_::,99,f~.17,0,-12.7: "1:42.ft" 'cil.7M~

664-030 G-307-1 T-1937-2 30

.0.17 0.11 79.45 3.07E+l9

-0.030 l.93E+l9 0.661 44.6 17.0 12.6 42.3 116.9 lower SheO Plates 584"-03E G-308-1 T-1837-1.. -.. ***....... 2f*-.. 0.17 0.1 {

  • 79.45... 3.07E+18........... 0.830 *--*...... 1.93E-t-1 e*-*-*. 0.681,.... 44.6...... 1,.0 °14.2..... 44_3'"' -* 109.9...

664-03F G-307-5 P-2076-2 3

0.27 0.63 173.9 3.07E+l9° 00

........ 0.630 *-*-** ****** 1.93E+l0 -*-*. 0.661..... 97.6.*.. *11.0 °13.8 ****4;.9**** -*144_5""

PARTNAME l11'N8r Intermediate Shell Axial Welds Weld No.*

2-564A, 2-5648, and 2-564C Lower Shell Axial Welds Z-5S~~;_;~E, PART NAME**

Lower Shell to Lower Intermediate Shell Circumferential Weld W1!1ld No.

3-66-1 WeldMatal ADJUSTMENTS FOR MARGIN AND Typo/ Hoat Flux Lot Number/

INITIAL I-C-:H:-BI-...IS-:TR::,Y+=-rc=a--..-.,..,--.,,--,-,,==---.-::::-!-,-:::8:-EL-Tl"T"IN_E...,l,..R_RAD....,..IA"',,_o...,N_s.,.H_IFT=*,-I Number FluxTyp9 RTNOT C11 NI CF IDFlu*nca Attllr,uatlon 114-TPlu1nc1 FF.

&RTiiUT u4 o1 Margin ART*

op op n/cmZ n/cmZ op op op -

'F op

  • :::: -::::- -::::* ---::::::::-***** *-*:::::********.. -:::~::::****** -::::: ****:::***** ::::1-:::.. :::*** <:1 9

0" Weld Metal.

Type/He!t Flux lot Numb,r I INITIAL* CHEMISTRY Number Flux Type RT"or Cu NI "F

'I.

II 1248 ARCOSB-5

  • 60 0.21)6 0.o7 CF ID Flu*nc*

Att*nuatlan

'F 101*n1cm2 90.2 3.07E+18 0.630 114-TFlu,nca 101*n1cm2 1.93E+l0 ADJUSTMENTS FOR MARGIN AND BEL TLINE IRRADIA,iON SHIFT FF MTmrr era a1 Margin ART ap*

'F' qp

'P ap 0.661 64.0 27.0 o.o 64.0 60.0

Oyster Creek PTLR

  • Revision 1 Page*25 of 25.

APPENDIX A Oyster Creek Reactor Vessel Material Surveillance Programs Oyster Creek:

lri acbordance with 1 O CFR 50, Appendix H, Reactor Vessel Materic1I Surveillance Program Requirements (Reference 6J~1), one surveiHance capsule has been removed from the OCGS RPV. The first surveillance capsule was removed from the OCGS RPV on February 12, 1983 after 8.38 EFPY (Reference 6.22). The surv~illance capsule* contained fiux wire~ for neutron.

  • fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials fron, th~ vessel materials within the core beltline region.

The flux wires and test specimens removed from the capsule were tested according to ASTM E185-82. The methods and results of testing are prese~ted in. References 6.22 and 623, as required by 10 CFR 50, Appendices G and H (Referen~es 6.10 and 6.21 ). There are two remaining OCGS surveillance ccipsules which will remain in place to serve as backup surveillance material for the BWRVIP program; Qr as otherwise needed.

Currently, OCGS has made a licensing commitment to replace the existing material surveillance program with the BWRVIP ISP (Reference 6.24) in the lice.nse amendment issued by the NRC regarding _implementaUon of the BVVRVIP ISP; dated April 27, 2004 (Reference 6.18). The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Pr~grams, and has been approved by NRC. U111der the ISP, there are no furth'er capsules from OCGS to be tested. Representative surveillance capsule materials for the OCGS limiting beltline plate and weld are in the Cooper and Hatch Unit 2 surveillance capsule programs, respectively. The next Cooper surveillance capsule is scheduled to be withdrawn and tested under the ISP in approximately 2029 at 40 EFPY. Th1e next Hatch Unit 2 surveillance capsule is I

scheduled for withdrawal and testing under the ISP in approximately 2027 at 37 EFPY.

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