Text

License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits
	ML19183A038
Person / Time
Site:	Catawba
Issue date:	07/01/2019
From:	Simril T; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA-18-0036
	Download: ML19183A038 (148)
	v • d • e

Tom Simril Vice President Catawba Nuclear Station Duke Energy CN01VP / 4800 Concord Road York, SC 29745 o: 803.701.3340 f: 803.701.3221 Tom.Simril@duke-energy.com Serial: RA-18-0036 10 CFR 50.90 July 2, 2019 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 CATAWBA NUCLEAR STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-413 AND 50-414 RENEWED LICENSE NOS. NPF-35 AND NPF-52

SUBJECT:

LICENSE AMENDMENT REQUEST PROPOSING TO REVISE TECHNICAL SPECIFICATION 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy) hereby requests an amendment to the Catawba Nuclear Station (CNS) Unit 1 Renewed Facility Operating License.

The proposed change will revise CNS Technical Specification (TS) 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect an update to the P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY). The proposed change will also reflect that the revised CNS Unit 1 P/T limit curves in TS 3.4.3 are applicable until 42.7 effective full power years (EFPY).

The proposed change is necessary because the existing CNS Unit 1 P/T limit curves in TS 3.4.3 are only applicable up to 30.7 EFPY. CNS Unit 1 is expected to reach 30.7 EFPY during cycle 26, which is projected for early 2021.

Note that although the proposed change only impacts CNS Unit 1, the amendment request is being docketed for both CNS Units 1 and 2 since the TS are common to both units.

The Enclosure provides a description and assessment of the proposed change. Attachment 1 provides the existing CNS TS pages marked up to show the proposed change. Attachment 2 provides a non-proprietary version of Westinghouse report WCAP-15448, Revision 1, Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY. Attachment 3 provides a non-proprietary version of Westinghouse report WCAP-17669, Revision 0, Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate:

Reactor Vessel Integrity and Neutron Fluence Evaluations.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed change involves no

U.S. Nuclear Regulatory Commission RA-18-0036 Page 2 significant hazards consideration. The bases for these determinations are included in the Enclosure.

Duke Energy requests approval of the proposed amendment within one year of the date this submittal is accepted by the NRC staff for review. Once approved, Duke Energy will implement the license amendment within 120 days.

There are no new regulatory commitments contained in this submittal.

In accordance with 10 CFR 50.91, Duke Energy is notifying the State of South Carolina of this license amendment request by transmitting a copy of this letter and enclosure to the designated State Official.

If there are any questions or if additional information is needed, please contact Mr. Art Zaremba, Director-Nuclear Fleet Licensing at 980-373-2062 or Arthur.Zaremba@duke-energy.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 2, 2019.

Sincerely, Tom Simril Vice President, Catawba Nuclear Station

Enclosure:

Description and Assessment of the Proposed Change Attachments:

1. Technical Specifications Markup

2. Westinghouse Report WCAP-15448, Revision 1, "Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY"

3. Westinghouse Report WCAP-17669, Revision 0, "Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations"

U.S. Nuclear Regulatory Commission RA-18-0036 Page 3 cc (with Enclosure/Attachments):

L. Dudes, USNRC Region II - Regional Administrator J.D. Austin, USNRC Senior Resident Inspector - CNS M. Mahoney, NRR Project Manager - CNS L. Garner, Manager, Radioactive and Infectious Waste Management (SC)

A. Nair-Gimmi, Director, Nuclear Response (SC)

U.S. Nuclear Regulatory Commission RA-18-0036 Page 1 ENCLOSURE Description and Assessment of the Proposed Change

Subject:

License Amendment Request Proposing to Revise Technical Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination Analysis 4.4 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS:

1. Technical Specifications Markup

2. Westinghouse Report WCAP-15448, Revision 1, Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY

3. Westinghouse Report WCAP-17669, Revision 0, Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations

U.S. Nuclear Regulatory Commission RA-18-0036 Page 2

1.

SUMMARY

DESCRIPTION Duke Energy Carolinas, LLC (Duke Energy) hereby requests an amendment to the Catawba Nuclear Station (CNS) Unit 1 Renewed Facility Operating License. The proposed change will revise CNS Technical Specification (TS) 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY). The proposed change will also reflect that the revised CNS Unit 1 P/T limit curves in TS 3.4.3 are applicable until 42.7 effective full power years (EFPY).

2.

DETAILED DESCRIPTION 2.1 System Design and Operation All components of the CNS Reactor Coolant System (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients and reactor trips. CNS is required to limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation.

The CNS TS contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The typical use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

Operating limits are established that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).

CNS Updated Final Safety Analysis Report Section 5.3.2 provides additional details regarding the methodology used to develop the P/T limit curves that are contained in the CNS TS.

2.2 Current Technical Specifications Requirements CNS Limiting Condition for Operation (LCO) 3.4.3 requires that RCS pressure, RCS temperature and RCS heatup and cooldown rates be maintained within the limits specified in Figures 3.4.3-1 and 3.4.3-2 (P/T curves) that are associated with this proposed change. The two elements of LCO 3.4.3 are:

The limit curves for heatup, cooldown and ISLH testing; and Limits on the rate of change of temperature.

LCO 3.4.3 limits apply to all components of the RCS, except the pressurizer and define allowable operating regions and permit a large number of operating cycles while providing a wide margin to non-ductile failure. Violating LCO 3.4.3 limits would result in placing the reactor vessel outside of the bounds of the stress analyses and could increase stresses in other reactor coolant pressure boundary (RCPB) components.

U.S. Nuclear Regulatory Commission RA-18-0036 Page 3 2.3 Reason for the Proposed Change The proposed change is necessary because the existing CNS Unit 1 P/T limit curves in TS 3.4.3 are only applicable up to 30.7 EFPY. CNS Unit 1 is expected to reach 30.7 EFPY during cycle 26, which is projected for early 2021. A new set of P/T limit curves with a longer term of applicability is required.

2.4 Description of the Proposed Change TS 3.4.3, Figure 3.4.3-1, (UNIT 1 ONLY) RCS Heatup Limitations is revised as follows:

The existing RCS heatup limitations curve is superseded entirely by a new curve applicable up to 42.7 EFPY.

The words Upper Shell Forging 06, Intermediate Shell Forging 05, and Bottom Head Ring 03 for the Limiting Material are revised to state Lower Shell Forging 04, Intermediate Shell Forging 05.

The words Limiting ART at 30.7 EFPY are revised to state Limiting ART at 42.7 EFPY.

The 1/4-T value of 42F is revised to state 47F.

The 3/4-T value of 31F is revised to state 34F.

TS 3.4.3, Figure 3.4.3-2, (UNIT 1 ONLY) RCS Cooldown Limitations is revised as follows:

The existing RCS cooldown limitations curve is superseded entirely by a new curve applicable up to 42.7 EFPY.

The words Upper Shell Forging 06, Intermediate Shell Forging 05, and Bottom Head Ring 03 for the Limiting Material are revised to state Lower Shell Forging 04, Intermediate Shell Forging 05.

The words Limiting ART at 30.7 EFPY are revised to state Limiting ART at 42.7 EFPY.

The 1/4-T value of 42F is revised to state 47F.

The 3/4-T value of 31F is revised to state 34F.

The TS markup provided in Attachment 1 reflects the proposed change described above.

3.

TECHNICAL EVALUATION Pressure/Temperature Limit Curves Development and Applicability Term Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility transition temperature) corresponding to the limiting material in the beltline region of the reactor pressure vessel (RPV). The most limiting RTNDT of the material in

U.S. Nuclear Regulatory Commission RA-18-0036 Page 4 the core (beltline) region of the RPV was determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift ( RTNDT).

RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactors life, RTNDT due to the radiation exposure associated with that time period was added to the original unirradiated RTNDT. Using the Adjusted Reference Temperature (ART) values, P/T limit curves were determined in accordance with the requirements of 10 CFR 50, Appendix G.

The existing P/T limit curves in CNS TS 3.4.3 for Unit 1 were originally developed in Westinghouse report WCAP-15203, Revision 1 for 34 EFPY (Reference 1). As a result of the NRC staffs issuance of amendments regarding Measurement Uncertainty Recapture Power Uprate (MUR) in April 2016 (Reference 2), the applicability of the TS 3.4.3 CNS Unit 1 P/T limit curves was decreased from 34 EFPY to 30.7 EFPY. The limiting material referenced in TS Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY) was also revised due to the issuance of amendments regarding MUR. The reduction to 30.7 EFPY is supported by Westinghouse report WCAP-17699-NP (Reference 3).

The proposed TS 3.4.3 P/T limit curves for CNS Unit 1 provided in Attachment 1 (i.e., the proposed change) were drawn using data points from Tables 11 and 14 of Westinghouse report WCAP-15448, Revision 1 for 51 EFPY (Reference 4). The 51 EFPY P/T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. The projected ART values in WCAP-15448 are based on the original thermal power rating of CNS Unit 1 prior to the MUR uprate in thermal power.

As documented in Westinghouse report WCAP-17699-NP, to determine the new proposed applicability term of 42.7 EFPY for the TS 3.4.3 CNS Unit 1 P/T limit curves that were originally developed in Reference 4 for 51 EFPY, the limiting reactor vessel material ART values with consideration of the MUR power uprate was compared to the limiting beltline material ART values used in development of the 51 EFPY P/T limit curves. The Regulatory Guide 1.99, Revision 2 (Reference 5) methodology was used along with the surface fluence in Section 2 of WCAP-17699-NP to calculate ART values for the CNS Unit 1 reactor vessel materials at 54 EFPY. It is important to note that the 54 EFPY ART values calculated as part of the MUR uprate evaluation in WCAP-17699-NP were used when assessing the applicability of the 51 EFPY P/T limit curves. The 54 EFPY ART values (as opposed to 51 EFPY values) were calculated to correspond with the other reactor vessel integrity evaluations calculated as part of the MUR power uprate analysis at end-of-life extension (EOLE). With re-evaluation of surveillance data credibility, a recalculation of chemistry factors and the consideration of all reactor vessel materials projected to achieve surface fluence levels of 1x1017 n/cm2 or higher, the applicability date of the 51 EFPY P/T limit curves for CNS Unit 1 decreased to 42.7 EFPY with the MUR power uprate. A more detailed discussion is contained in Section 8.2 of Reference 3, which is being provided as Attachment 3 to this submittal. Reference 4, which contains the data points for the proposed TS 3.4.3 CNS Unit 1 P/T limit curves, is being provided as Attachment 2 to this submittal.

10 CFR 50, Appendix G requires that P/T limits be developed to bound all ferritic materials in the RPV. Regulatory Issue Summary 2014-11 (Reference 6) clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may define P/T curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure. As part of the response to a NRC request for additional information (RAI) regarding

U.S. Nuclear Regulatory Commission RA-18-0036 Page 5 the CNS MUR license amendment request (Reference 7), Westinghouse letter MCOE-LTR 103, Rev. 0 was included as an attachment. The letter included an evaluation of nozzle P/T limits that demonstrated the RV nozzle P/T limits to be bounded by the normal beltline P/T limits. In Reference 2, the NRC confirmed that the P/T curves for the nozzles would not be more limiting than the normal beltline P/T curves. According to Reference 6, if a material does not exceed 1x1017 n/cm2 at the end of the design life, then the material does not need to be adjusted for neutron embrittlement in the P/T limit curve calculation. According to Reference 3 (Attachment 3 of this submittal), the fluence at the lowest extent of the nozzles is below the 1x1017 n/cm2 threshold at the end of design life (54 EFPY) and thus consideration of RV nozzle embrittlement is not required in the P/T limit curve development for the proposed change.

Therefore, the evaluation of RV nozzle P/T limit curves in MCOE-LTR-13-103, which conservatively considered embrittlement based on fluence for 34 EFPY is still bounded by the proposed P/T limit curves provided in Attachment 1 of this submittal.

Low Temperature Overpressure (LTOP) Setpoint Considerations The Low Temperature Overpressure Protection (LTOP) analysis calculated a maximum pressure of 609.5 psig for CNS Unit 1 (including location adjustment and instrument uncertainty). The maximum pressure calculation included mass injection from a high head injection pump and intermediate head injection pump. This calculation assumed the current LTOP PORV setpoint of 400 psig and residual heat removal (RHR) suction relief valve setpoint of 463 psig. The new limiting pressure determined for the heatup and cooldown curves at 42.7 EFPY (with measurement uncertainty recapture) is 1089 psig for the reactor vessel beltline region and remained 621 psig for the Closure Head/Vessel Flange region. Therefore, the current PORV low-pressure setpoint (400 psig) and RHR suction relief valve setpoint (463 psig) remains acceptable for plant operation to 42.7 EFPY for CNS Unit 1.

The minimum 42.7 EFPY LTOP enable temperature is 87°F for CNS Unit 1. This value remains lower than the required minimum enable temperature of at least 200°F plus uncertainty (10°F).

Therefore, the current enable temperature of 210°F remains acceptable for plant operation to 42.7 EFPY for CNS Unit 1.

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following regulatory requirements and guidance documents are applicable to the proposed change.

10 CFR 50.36 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, establishes the requirements related to the content of the TSs. Pursuant to 10 CFR 50.36(c)

TSs will include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) LCOs, (3) Surveillance Requirements (SRs), (4) design features; and (5) administrative controls.

CNS LCO 3.4.3 limits the pressure and temperature changes during RCS heatup and cooldown (i.e., to the right and below the P/T curves in Figures 3.4.3-1 and 3.4.3-2), to prevent non-ductile RPV failure. The proposed change revises the CNS Unit 1 P/T limit curves in TS 3.4.3 and reflects that the curves are applicable until 42.7 EFPY. Based on the determination that the

U.S. Nuclear Regulatory Commission RA-18-0036 Page 6 proposed TS Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY) are acceptable up to 42.7 EFPY (see Section 3 above), Duke Energy concludes that CNS LCO 3.4.3 will continue to meet 10 CFR 50.36(c)(2)(i) with the proposed change.

10 CFR 50.60 Section 50.60 of 10 CFR, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, imposes fracture toughness and material surveillance program requirements, which are set forth in 10 CFR 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements. With the proposed change, CNS meets the requirements set forth in 10 CFR 50, Appendices G and H. Therefore, CNS also satisfies the requirements of 10 CFR 50.60 for the proposed change.

10 CFR 50, Appendix G Appendix G to 10 CFR 50 requires that the P/T limits for the facilitys reactor pressure vessel (RPV) be at least as conservative as those obtained by following the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Using the ART values, P/T limits curves were determined in accordance with the requirements of 10 CFR 50, Appendix G.

Therefore, Duke Energy concludes for the proposed change that the CNS RPV will continue to meet RPV integrity regulatory requirements through 42.7 EFPY.

10 CFR 50, Appendix H Appendix H to 10 CFR 50 establishes requirements for each facility related to its RPV material surveillance. These regulatory requirements will continue to be met for the proposed change with the surveillance capsule removal schedule prescribed in Section 9 (Table 9-1) of to this submittal.

Regulatory Guide (RG) 1.99, Revision 2 RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains guidance on methodologies the NRC considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. This RG was used along with the surface fluence in Section 2 of WCAP-17669-NP to calculate ART values for the CNS Unit 1 reactor vessel materials at 54 EFPY. As previously noted in Section 3 of this LAR above, the 54 EFPY ART values (as opposed to 51 EFPY values) were calculated to correspond with the other reactor vessel integrity evaluations calculated as part of the MUR power uprate analysis at end-of-life extension (EOLE). Therefore, the proposed change has no effect on how Duke Energy applies RG 1.99, Revision 2 for CNS Unit 1.

RG 1.190 RG. 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001, describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence. As noted in WCAP-17669-NP, the neutron transport evaluation methodologies utilized for the CNS Unit 1 neutron fluence evaluation follow the guidance of RG 1.190.

U.S. Nuclear Regulatory Commission RA-18-0036 Page 7 Regulatory Issue Summary (RIS) 2014-11 RIS 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Duke Energy appropriately considered RIS 2014-11 for the proposed change in the Technical Evaluation section of the LAR above (Section 3).

The proposed change does not affect plant compliance with any of the above regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedent The NRC has previously approved changes similar to the proposed change in this License Amendment Request for other nuclear power plants including:

H.B. Robinson Steam Electric Plant, Unit No. 2: Application dated November 2, 2015 (ADAMS Accession No. ML15307A069); NRC Safety Evaluation dated November 22, 2016 (ADAMS Accession No. ML16285A404).

Browns Ferry Nuclear Plant, Unit 2: Application dated June 19, 2014 (ADAMS Accession No. ML14175A307); NRC Safety Evaluation dated June 2, 2015 (ADAMS Accession No. ML15065A049).

4.3 No Significant Hazards Consideration Determination Analysis Duke Energy requests an amendment to the Catawba Nuclear Station (CNS) Unit 1 Renewed Facility Operating License. The proposed change will revise CNS Technical Specification (TS) 3.4.3, RCS Pressure and Temperature (P/T) Limits, to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY). The proposed change will also reflect that the revised CNS Unit 1 P/T limit curves in TS 3.4.3 are applicable until 42.7 effective full power years (EFPY).

Duke Energy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

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

Response: No.

The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY) that are applicable until 42.7 EFPY.

The proposed change does not involve physical changes to the plant or alter the reactor coolant system (RCS) pressure boundary (i.e., there are no changes in operating pressure, materials or seismic loading). The proposed P/T limit curves and Adjusted

U.S. Nuclear Regulatory Commission RA-18-0036 Page 8 Reference Temperature (ART) values for TS 3.4.3 with an applicability term of 42.7 EFPY provide continued assurance that the fracture toughness of the reactor pressure vessel (RPV) is consistent with analysis assumptions and NRC regulations. The methodology used to develop the proposed P/T limit curves provides assurance that the probability of a rapidly propagating failure will be minimized. The proposed P/T limit curves, with the applicability term of 42.7 EFPY, will continue to prohibit operation in regions where it is possible for brittle fracture of reactor vessel materials to occur, thereby assuring that the integrity of the RCS pressure boundary is maintained.

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

2.

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

Response: No.

The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY) that are applicable until 42.7 EFPY.

The proposed change does not affect the design or assumed accident performance of any structure, system or component or introduce any new modes of system operation or failure modes. Compliance with the proposed P/T limit curves will provide sufficient protection against brittle fracture of reactor vessel materials to assure that the RCS pressure boundary performs as previously evaluated.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises TS 3.4.3 to reflect updated P/T limit curves in Figures 3.4.3-1 (UNIT 1 ONLY) and 3.4.3-2 (UNIT 1 ONLY) that are applicable until 42.7 EFPY.

CNS complies with applicable regulations (i.e., 10 CFR 50, Appendices G and H) and adheres to Nuclear Regulatory Commission (NRC)-approved methodologies (i.e.,

Regulatory Guides 1.99 and 1.190) with respect to the proposed P/T limit curves in TS 3.4.3 in order to provide an adequate margin of safety to the conditions at which brittle fracture may occur. The proposed P/T limit curves for CNS Unit 1, with an applicability term of 42.7 EFPY, will continue to provide assurance that the P/T limits are not exceeded.

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

Based on the above, Duke Energy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

U.S. Nuclear Regulatory Commission RA-18-0036 Page 9 4.4 Conclusions In conclusion, based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commissions regulations; and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.

ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.

REFERENCES

1. Westinghouse Report WCAP-15203, Revision 1, Catawba Unit 1 Heatup and Cooldown Curves for Normal Operation Using Code Case N-640, April 2001 (ADAMS Accession No. ML030900046).

2. NRC letter, Catawba Nuclear Station, Units 1 and 2 - Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (CAC NOS. MF4526 and MF4527), April 29, 2016 (ADAMS Accession No. ML16081A333).

3. Westinghouse Report WCAP-17669-NP, Revision 0, Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations, June 2013 (ADAMS Accession No. ML14353A029).

4. Westinghouse Report WCAP-15448, Revision 1, Catawba Unit 1 Heatup and Cooldown Curves for Normal Operation Using Code Case N-640 for 51 EFPY, April 2001.

5. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

6. Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, October 14, 2014.

7. Duke Energy letter, License Amendment Request (LAR) for Measurement Uncertainty Recapture (MUR) Power Uprate Response to NRC Requests for Additional Information (RAIs) (TAC Nos. MF4526 and MF4527), January 22, 2015 (ADAMS Accession No. ML15029A417).

to RA-18-0036 Technical Specifications Markup

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-f--

en f---

Q)

I f--f-- -f--

I,,_

I, Cl.

-f--

rs..

I "O

1000 -

~ Cooldown Rat~t~/hr)

I Q)

'\\ 0, 20, 40, 60, 100

+-'

ro 0

I/

I "O

I'-

C

~

I(

500 I/

f--

Closur~iad ~

Vessel Fl ge Limit,

f---

-f---

~-

I/

I

\\

0 V

0 50 100 0 200 250 300 50 400 450 500 In icated Temperature (De. F)

Figure 3.4.3-2 (UNIT 1 ONLY)

RCS Cooldown Limitations Catawba Units 1 and 2 3.4.3-5 Amendment Nos.-28-tfz-rr

INSERT 2 0

500 1000 1500 2000 2500 0

50 100 150 200 250 300 350 400 450 500 PRESSURE (PSIG)

TEMPERATURE (DEG F)

Acceptable Operation Cooldown Rates (0F/hr) 0, 20, 40,60 & 100 Unacceptable Operation MATERIALS PROPERTY BASIS Limiting Material: Lower Shell Forging 04, Intermediate Shell Forging 05 Limiting ART at 42.7 EFPY:

1/4-T 470F 3/4-T 340F Figure 3.4.3-2 (UNIT 1 ONLY)

RCS Cooldown Limitations (Without Margins for Instrument Errors)

Closure Head and Vessel Flange Limit to RA-18-0036 Westinghouse Report WCAP-15448, Revision 1, Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY

Westinghouse Non-Proprietary Class 3 Catawba Unit 1 Heatup and Cooldown Curves for Normal Operation Using Code Case N-640 for 51 EFPY Westinghouse Electric Company LLC

0s,:;-J>3J/

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15448, Revision 1 Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY J. H. Ledger April 2001 Prepared by the Westinghouse Electric Company LLC for the Duke Power Company Approved:;:sJ; H-*z___::t FoR..,

C. H. Boyd, Manager Engineering and Materials Technology Westinghouse Electric Company LLC P.O. Box355 Pittsburgh, PA 15230-0355

PREFACE This report has been technically reviewed and verified by:

Revision 1:

An error was detected in the "OPERLIM" Computer Program that Westinghouse uses to generate pressure-temperature (PT) limit curves. This error potentially effects the heatup curves when the 1996 Appendix G Methodology is used in generating the PT curves. It has been determined that WCAP-15448 Rev. 0 was impacted by this error. Thus, this revision provides corrected curves from WCAP-15448 Rev. 0.

Note that only the heatup curves and associated data point tables have changed. The cooldown curves and data points remain valid and were not changed.

ii TABLE OF CONTENTS LIST OF TABLES................................................................................................................................. m LIST OF FIGURES................................................................................................................................ iv EXECUTIVE

SUMMARY

.......................................... V INTRODUCTION...................................................................................................................... I 2

PURPOSE.................................................................................................................................. 2 3

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS................... 3 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE.......................................... 7 5

HEATUPAND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES....................... 16 6

REFERENCES......................................................................................................................... 26

Table l Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Table 8 Table 9 Table \\0 Table 11 Table 12 Table 13 Table 14 Ill LIST OF TABLES Summary of the Vessel Surface. I/4T and 3/4T Fluence Values used for the Generation of the 5 I EFPY Heatup/Cooldown Curves..................................................... 8 Integrated Neutron Exposure of the Catawba Unit L McGuire Unit 2, and Watts Bar Unit l Surveillance Capsules Tested To Date.................................................. 8 Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data........ l 0 Reactor Vessel Beltline Maten al Unirradiated Toughness Properties............................... 11 Calculation of Chemistry Factors using Catawba Unit l Surveillance Capsule Data....... 12 Summary of the Catawba Unit l Reactor Vcssd Beltlinc Material Chemistry Factors..... 13 Summary of the Calculated Fluence Factors used for the Generation of the 51 EFPY Hcatup and Cooldown Curves...................................................................................... 13 Calculation of the ART Values for the I/4T Location @51 EFPY................................. 14 Calculation of the ART Values for the 3/4T Location (ct 51 EFPY.................................. 14 Summary of the Limiting ART Values Used in the Generation of the Catawba Unit l Heatup/Cooldown Curves.............................................................................................. 15 51 EFPY 60°F/hr. Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)............................................................. 22 51 EFPY 80°F/hr. Heatup Curve Data Points Using 1996 App. G

(\\vithout Uncertainties for Instrumentation Errors)............................................................. 23 51 EFPY 100°F/hr. Hcatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)............................................................. 24 51 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors).............................................................. 25

Figure 1 Figure 2 Figure 3 Figure 4 LIST OF FIGURES Catawba Unit l Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for the First 5 J EFPY iv (Without Margins for Instrumentation Errors)................................................................ 18 Catawba Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of 80°F /hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 19 Catawba Unit I Reactor Coolant System Heatup Limitations (Heatup Rate of l00°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 20 Catawba Unit l Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)................................................................ 21

V EXECUTIVE

SUMMARY

The purpose of this report is to generate pressure-temperature limit curves for Catawba Unit 1 for normal operation at 51 EFPY using the methodology from the 1996 ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, ASME Code Case N-640P01, which allows the use of the K1c methodology and the elimination of the reactor vessel flange temperature requirement (Ref, WCAP-15315£111). Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values at the l/4T and 3/4T location. The lower shell forging is limiting for the 3/4T location, while the intermediate shell forging is limiting for the 1/4T ~ocation. The pressure-temperature limit curves were generated without margins for instrumentation errors for heatup rates of 60, 80 and l00°F/hr and cooldown rates of 0, 20, 40, 60 and 100°F/hr. These curves can be found in Figures I through 4.

1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RT NOT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel, The adjusted RT NDT of the limiting material in the core region of the reactor vessel is detennined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced

.1RT Nor, and adding a margin. The unirradiated RT NDT is designated as the higher of either the drop weight nil-ductility transition temperature (NOTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (nonnal to the major working direction) minus 60°F.

RT NOT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT NDT at any time period in the reactor's life,.1RT NDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NOT (IRT NDr)- The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide t.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."141 Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRT NOT+.1RT NDT + margins for uncertainties) at the t/4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface. The most limiting ART values are used in the generation ofheatup and cooldown pressure-temperature limit curves.

2 2

PURPOSE Duke Power Company contracted Westinghouse to generate new heatup and cooldown curves for 51 EFPY using the latest Code Methodologies and the elimination of the flange requirement (as modified in Section 3.3 of this report). The heatup and cooldown curves were generated without margins for instrumentation errors. The curves include a hydrostatic leak test limit curve from 2485 to 2000 psig.

The purpose of this report is to present the calculations and the development of the Duke Power Company Catawba Unit l heatup and cooldown curves for 51 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2[11, for all the beltline materials and the deyelopment of the heatup and cooldown pressure-temperature limit curves for nonnal operation. Note that the Pressure-Temperature curves herein were generated at 51 EFPY to be consistent with the Catawba Unit 2 curves, which at the time had a capacity factor of 85%. Later in time Pressure-Temperature curves were generated for the McGuire Units at 54 EFPY; which equates to an increased capacity factor of 90%.

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K1c is obtained from the reference fracture toughness curve, defined in Code Case N-640, "Alternative Reference Fracture Toughness for Development of PT Limit Curves for Section XI"13 & 6J of the ASME.Appendix G to Section XI. The K1c curve is given_by the following equation:

where, K1c== 33.2 + 20.734

e[0.02(T-RTNor)I (1) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT This K 1c curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

where, Kim K11 K1c C

C

=

stress intensity factor caused by membrane (pressure) stress stress intensity factor caused by the thermal gradients function of temperature relative to the RT NDT of the material 2.0 for Level A and Level B service limits (2) 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical 3

4 For membrane tension, the corresponding K1 for the postulated defect is:

K Im = Mm X (pR; / t)

(3) where, Mm for an inside surface flaw is given by:

Mm

=

1.85 for t < 2, Mm

o.926 ! for 2~ Ji~ 3.464, Mm

=

3.21 for.fi > 3.464 Similarly, Mm for an outside surface flaw is given by:

Mm 1.77 for.fi < 2, Mm

=

0.893 Ji for 2 ~ Ji :<::; 3.464, Mm

=

3.09 for Ji > 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding K1 for the postulated defect is:

K1b = Mb

Maximum Stress, where Mb is two-thirds of Mm The maximum K1 produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K 11 = 0.953xto*3 x CR x t2*5, where CR is the cooldown rate in °F/hr., or for a postulated outside surface defect, K11 = 0.753xl0"3 x HU x t2*5, where HU is the heatup rate in °F/hr.

The through-wall temperature difference associated with the maximum thermal K1 can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal K1 *

(a)

The maximum thermal K1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).

(b)

Alternatively, the K1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness inside surface defect using the relationship:

K1t = {l.0359Co+ 0.63220 + 0.4753C2 + 0.3855C3) *&

(4)

or similarly, K1T during heatup for a 1/4-thickness outside surface defect using the relationship:

K1t = (l.043Co+ 0.630C1 + 0.481C2 + 0.401C3) *,.[;i; (5) where the coefficients C0, C 1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

o-(x) = Co+ C1(x /a)+ C2(x I a)2 + CJ(x I a) 3 (6) and xis a variable that_represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

5 Note, that equations 3, 4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-14040, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldwon Limit Curves"181 Section 2.6

( equations 2.6.2-4 and 2.6.3-l) with the exceptions just described above.

At any time during the heatup or cooldown transient, K1e is determined by the metal temperature at the tip of a postulated flaw at the l/4T and 3/4T location, the appropriate value for RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K1i, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cool down procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the l/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the ~T (temperature) developed during cooldown results in a higher value ofK1c at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Kie exceeds K11, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the I/4T location and, therefore, allowable pressures may unlrnowingly be violated if the rate of cooling is decreased at various intervals along a cool down ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

6 Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1c for the 1/4T crack during heatup is lower than the K1c for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1c values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a l/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate ofheatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3.3 Closure HeadNessel Flange Requirements 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RT NOT by at least 120°F for normal operation when the pressure exceeds 20 percent of the pre service hydrostatic test pressure (3106 psi), which is 621 psig for Catawba Unit 1. However, per WCAP-15315, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation For Operating PWR and BWR Plants"1131, this requirement is no longer necessary when using the methodology of Code Case N-640131* Hence, the Catawba Unit 1 heatup and cooldown limit curves will be generated without flange requirements included.

4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RT NDT + tiRT NDT + Margin (7) 7 Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Bo~ler and Pressure Vessel Code171. If measured values of initial RT NDT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

t.\\RT NDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

~T NDT== CF * (0 0.10 log f)

To calculate L'!RT NOT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

f.

_ f.

(-0.24>.)

(depth x) -

sunace e

(8)

(9) where x inches (vessel beltline thiclmess is 8.465 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant tluence is then placed in Equation 8 to calculate the L'!RT Nor at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections as a part ofWCAP-15117 and are also presented in a condensed version in Table 1 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"[81* Table l contains the calculated vessel surface fluences values along with the Regulatory Guide 1.99, Revision 2, l/4T and 3/4T calculated fluences used to ca1culate the ART values for all beltline materials in the Catawba Unit 1 reactor vessel.

Additionally, the surveillance capsule fluence values are presented in Table 2.

Notes:

TABLE 1 Summary of the Vessel Surface, l/4T and 3/4T Fluence Values used for the Generation of the 51 EFPY Heatup/Cooldown Curves Material Surface 1/4T Intermediate Shell Forging 05 2.95 X 1019 1.78 X 1019 Lower Shell Forging 04 2.95 X 1019 1.78 X 1019 Intermediate to Lower Shell 2.95 X 1019 1.78 X 1019 Circumferential Weld Seam 3/4T 6.43 X 1018 6.43 X 1018 6.43 X 1018 (a) These fluence values were obtained from the calculated fluence values given in Table 6-13 of WCAP-15117.

(b) l/4T and 3/4T = Fcsurfs<:,,J *e<-0*24*xi, where xis the depth into the vessel wall (i.e. 8.465*0.25 or 0.75)

TABLE2*

Integrated Neutron Exposure of the Catawba Unit l, McGuire Unit 2, and Watts Bar Unit 1 Surveillance Capsules Tested To Date Plant Capsule Fluence Catawba Unit l z

2.993 x 1018 n/cm2, (E > 1.0 MeV)

Catawba Unit 1 y

1.318 x 1019 n/cm2, (E > l.0 MeV)

Catawba Unit l V

2.334 x 1019 n/cm2, (E > 1.0 MeV)

Catawba Unit l u

2.439 x 1019 n/cm2, (E > 1.0 MeV)

Catawba Unit l X

2.439 x 1019 n/cm2, (E > 1.0 MeV)

McGuire Unit 2 V

3.23 x 1018 n/cm2, (E > 1.0 MeV)

McGuire Unit 2 X

1.47 x 1019 n/cm2, (E > 1.0 MeV)

McGuire Unit 2 u

2.04 x 1019 n/cm2, (E > 1.0 MeV)

McGuire Unit 2 w

3.07 x 1019 n/cm2, (E > 1.0 MeV)

Watts Bar Unit l u

5.05 x 1018 n/cm2, (E > 1.0 MeV)

This data was taken from Table 4-2 in WCAP-15118 except for McGuire Unit 2 which were updated in Table 7.1-1 ofWCAP-15334.

8

9 Margin is calculated as, M = 2.JO': + O' ~. The standard deviation for the initial RT NDT margin tenn, is cr; 0°F when the initial RT NDT is a measured value, and l 7°F when a generic value is available. The standard deviation for the 6RT NDT margin term, cr6, is 17°F for plates or forgings, and 8.5°F for plates or forgings when surveillance data is used. For welds, cr" is equal to 28°F when surveillance capsule data is not used, and is 14°F (half the value) when credible surveillance capsule data is used. cr6 need not exceed 0.5 times the mean value of ~TNDT*

Contained in Table 3 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials111* These measured shift values were obtained using CVGRAPH, Version 4.1 191, which is a hyperbolic tangent curve-fitting program.

10 TABLE3 Measured 30 ft-lb Transition Temperature Shifts of all Available Surveillance Data Material Capsule Measured 30 ft-lb Transition Temperature Shift Intermediate Shell Forging 05 z

-14.9°F(a)

(Tangential Orientation) y 19.09°F (Reference WCAP-15117)

V 25.61°F lntennediate Shell Forging 05 z

15.74°F (Axial Orientation) y 48.63°F (Reference WCAP-15117)

V 50.58°F Catawba Unit I Surveillance z

1.91°F Weld Metal Data y

l 7.79°F (Reference WCAP-15117)

V 26.5°F McGuire Unit 2 Surveillance V

38.51°F Weld Metal Data X

35.93°F (Reference WCAP-14 799[ 101) u 23.81°F w

43.76°F Watts Bar Unit I Surveillance Weld Metal Data u

-6.oop(a)

(Reference WCA P-150461 111)

Notes:

(a)

This value will be assumed to be 0°F in this evaluation for conservatism (i.e. higher CF value).

Table 4 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RT NOT of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4 were used to generate the calculated chemistry factor (CF) values based on Tables l and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 6. Table 5 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 6.

11 TABLE 4Cc)

Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Cu(%)

Ni(%)

Initial RT NDT(b)

Closure Head Flange 0.05 0.83

-4op Vessel Flange 0.86

-31°F Intermediate Shell Forging 05 0.09 0.86

_gop Lower Shell Forging 04 0.04 0.83

_13op Intennediate to Lower Shell Girth Weld SeamCb>

0.04 0.72

-51°F Catawba Unit l Surveillance Weld MetalCb) 0.05 0.73 McGuire Unit 2 Surveillance Weld MetaJCb) 0.04 0.74 Watts Bar Unit l Surveillance Weld MetatCb>

0.03 0.15 Watts Bar Unit 2 Surveillance Weld MetaJCb) 0.02 0.69 Notes:

(a) Based on measured data. The RVID Database contains four IRT NDT Values (-51 °F, -68°F, -43°F and -50°F) for weld wire heat# 895075. The average of these IRTNDr values is -53°F. However, -51°F is the measured value for the Catawba Unit 1 weld metal. -51 °F is used in this evaluation since it is more conservative than the average of -53°F.

(b) The surveillance weld was made with the same weld wire and flux as the intennediate to lower shell girth weld (weld wire heat# 895075, type Grau L.O. # LW320 flux, Lot #P46).

(c) This data was taken from Table 4-7 in WCAP-15118.

TABLE 5(g)

Calculation of Chemistry Factors using Catawba Unit 1 Surveillance Capsule Data Material Capsule Capsule f<*>

FFCb>

t.RTNDT(c)

FF*ARTNDT Intermediate Shell z

0.2993 0.670 0 (-14.9)°F(c) 0°F Forging 05 y

1.318 1.077 l9.09°F 20.56°F (Tangential)

V 2.334 1.229 25.61°F 31.47°F Intennediate Shell z

0.2993 0.670 l5.74°F I0.55°F Forging 05 y

1.318 1.077 48.63°F S2.37°F (Axial)

V 2.334 1.229 50.58°F 62.l6°F SUM 177.11 °F CFFo,gingos = :E(FF

RT Nor) + °2:(FF2) = (177.11) + (6.24) = 28.4°F Z(DCP) 0.2993 0.670 1.51 °F (l.91°F) 1.01 Of Beltline Region Y(DCP) 1.318 1.077 14.0S°F (17.79°F) 15.13°F Weld Metal(d. o V(DCP) 2.334 1.229 20.94°F (26.5°F) 25.74°F V(DBP) 0.323 0.689 39.51 Of (38.51 °F) 27.22°F X (DBP) 1.47 l.11 36.93"F(35.93"F) 40.99°F U(DBP) 2.04 1.19 24.81 °F (23.81°F) 29.52°F W(DBP) 3.07 1.30 44.76°F (43.76°F) 58.19°F U(WAT) 0.505 0.809 l.3 I 7°F (-6.0°F) 1.07°F SUM l98.87°F CFS/Pwtld = L(FF

RT Nor)+ ~(FF2) = (198.87) + (8.585) = 23.2°F Notes:

(a) f= Integrated neutron fluence from References I, 10 and 11, (x 1019 n/cm2, E > 1.0 MeV).

(b) FF = fluence factor = f (o.28

o.1 °108 0 (c) Af{TNDT values are measured[11.

12 FF2 0.45 l.16 I.SI 0.45 1.16 1.51 6.24 0.4S 1.16 1.51 0.475 l.23 1.42 1.69 0.65 8.585 (d) McGuire Unit 2 operates with an inlet temperature of approximately 554°F, Catawba Unit l operates with an inlet temperature of approximately 553°F, and Watts Bar Unit I operates with an inlet temperature of approximately 560"F.

The measured Af{T NOT values from the McGuire Unit 2 surveillance program were adjusted by adding I °F to each measured Af{T NOT and the Watts Bar Unit 1 surveillance program were adjusted by adding 7°F to each measured Af{T NOT value before applying the ratio procedure. The surveillance weld metal Af{ T Nor values have been adjusted by a ratio factor of:

0.79 (CFvw + CFsw = 54.0 + 68.0"' 0.79) for the Catawba Unit l data.

1.0 (CFvw + CFsw"' 54.0 + 54.0"' 1.0) for the McGuire Unit 2 data.

1.317 (CFvw + CFsw = 54.0 + 41.0 = 1.317) for the Watts Bar Unit I data.

Pre-adjusted values are in parenthesis.

( e) Assumed to be 0°F for conservatism (i.e. results in a higher CF).

(f) DCP = Catawba Unit 1 (Data is from WCAP-15117);

DBP = McGuire Unit 2 (Data is from WCAP-14799).

WAT= Watts Bar Unit I (Data is from WCAP-15046).

(g) This data was taken from Table 4-8 ofWCAP-15118.

TABLE6*

Summary of the Catawba Unit I Reactor Vessel Beltline Material Chemistry Factors Material Reg. Guide 1.99, Rev. 2 Reg. Guide 1.99, Rev. 2 Position 1.1 CF's Position 2.1 CF's Intermediate Shell Forging 05 58.0°F 28.4°F Lower Shell Forging 04 26.0°F Beltline Region Weld Metal 54.0°F 23.2°F This data was taken from Table 4-9 in WCAP-15118 and Table 5 herein.

- No surveillance material for forging 04, thus Position 2.1 does not apply.

13 Contained in Table 7 is a summary of the 1/4T and 3/4T fluences along with the associated fluence factors (FF) used in the calculation of adjusted reference temperatures for the Catawba Unit I reactor vessel beltline materials. Note that since this is a forging vessel (i.e. no longitudinal welds) the peak fluence at 51 EFPY was used for the upper/lower Shell forging and the circwnferential weld.

TABLE 7 Summary of the Calculated Fluences and Fluence factors Used for the Generation of the 51 EFPY Heatup and Cooldown Curves EFPY 1/4Tf l/4T FF 3/4Tf 3/4T FF

?I 1.78 x 1019 n/cm2 1.16 6.43 x 1018 n/cm2 0.876 Based on the surveillance program credibility evaluation presented in Appendix D to WCAP-15117, the Catawba Unit 1 surveillance program data is credible. In addition, the surveillance program weld metal is representative of all of the beltline region girth weld seam. Hence, the adjusted reference temperature (ART) must be calculated for 51 EFPY for each beltline material at the l/4T and 3/4T locations. In addition, ART values must be calculated per Regulatory Guide l. 99, Revision 2, Position 1.1 and 2.1 Contained in Table 8 and 9 are the calculations of the 51 EFPY ART values used for generation of the heatup and cooldown curves.

TABLE 8 Calculation of the ART Values for the l/4T Location@ 51 EFPY Material RG 1.99, R2Method Intermediate Shell Position 1.1 Forging OS Position 2.1 Lower Shell Forging 04 Position 1.1 Intermediate to Lower Shell Position I.I Circumferential Weld Seam Position 2.1 Notes:

(a)

Initial RTNDr values measured values.

CF

{°F) 58.0 28.4 26.0 54.0 23.2 (b)

(c)

ART = Initial RT NDT + 6RT NDr + Margin (°F)

ARTNDr = CF* FF (d)

M = 2 *(cr? + cra2)1'2 (See Page 9).

FF 1.16 l.16 1.16 1.16 1.16 TABLE9 ARTNDr<*>

Margin<dl (OF)-

{°F) 67 34.0 33 17.0 30 30 63 56.0 27 27

Calculation of the ART Values for the 3/4T Location@ 51 EFPY Material RG 1.99, R2 Method Intermediate Shell Position 1.1 Forging05 Position 2.1 Lower Shell Forging 04 Position 1.1 Intermediate to Lower Shell Position 1.1 Circumferential Weld Seam Position 2.1 Notes:

(a)

Initial RTNDr values measured values.

CF (OF) 58.0 28.4 26.0 54.0 23.2 (b)

(C)

ART = Initial RT NDT + ART NDT + Margin (°F)

'1RT NDT "' CF

FF (d)

M = 2 *(o-;2 + o,.2)112 (See Page 9).

FF ARTNDr<*l Margin(d)

(Of)

(OF) 0.876 51 34.0 0.876 25 17.0 0.876 23 23 0.876 47 47 0,876 20 20 14 IRTNDrC*J ART°'>

(Of)

(OF)

-8 93

-8 42

-13 47

-51 68

-51 3

IRTND/*l ART°'l (Of)

(Of)

-8 77

-8 34

-13 33

-51 43

-51

-11

15 The intermediate shell forging 05 is the limiting beltline material for the 3/4T case and the lower shell forging 04 is the limiting beltline material for the 1/4T case (See Tables 8 and 9). Contained in Table I 0 is a summary of the limiting ARTs to be used in the generation of the Catawba Unit I reactor vessel heatup and cooldown curves.

TABLE IO Summary of the Limiting ART Values Used in the Generation of the Catawba Unit 1 Heatup/Cooldown Curves EFPY 1/4T Limiting ART 3/4T Limiting ART 51 47°F 34°f

5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for nonnal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the rnethods18* 121 discussed in Sections 3.0 and 4.0 of this report. The pressure difference between the wide-range pressure transmitter and the limiting beltline region has not been accounted for in the pressure-temperature limit curves generated for nonnal operation.

16 Figures l through 3 present the heatup curves without margins for possible instrumentation errors using heatup rates of 60, 80 and lOO?F/hr applicable for the first 51 EFPY. Figure 4 presents the cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40, 60 and 100°F/hr applicable for 51 EFPY. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures l through 4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures l through 3. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-640[3!

(approved in February 1999) as follows:

where, K1m is the stress intensity factor covered by membrane (pressure) stress, Kie= 33.2 + 20.734 e[002(T-RTNur)l, T is the minimum permissible metal temperature, and RT NDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 5. The pressure-temperature limits for core operation ( except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the in service hydrostatic leak tests for the Catawba Unit 1 reactor vessel at 51 EFPY is 108°F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Catawba Unit l reactor vessel.

The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 11 through 14.

17

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL FORGING 05 & 04 LIMITING ART VALUES AT 51 EFPY:

1/4T, 47°F 3/4T, 34°F 2500 2250 2000 1750 8'

ff 1500 en en ! 1250

0.

'0 I

i 1000 u

~

750 500 250 0

I I

I IOpertim Verslon:5.1 Run:12694 I i.---

i---I Leak Test Limit I

~

J Acceptable ~

I Unacceptable Operation Operation J '

reatup Rate I I 60 Deg. F/Hr

'--:.1

)

Critical Limit I I

60 Deg. F/Hr I I '

Criticality Limit based on

. - inservice hydrostatic test temperature (108 F) for the service period up to 51 EFPY I Boltup I I Temp I 0

50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 1 Catawba Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors) 18

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL Forging 05 & 04 LIMITING ART VALUES AT 51 EFPY:

1/4T, 47°F 3/4T, 34°F 2500 2250 2000 1750 C>

~ 1500

-e

I en

{ft e 1250 D.

i 11000

~

750 500 250 0

0 I

I I

IOperlim Verslon:5.1 Run:12694 J j Leak Test Llmltj J

Acceptable I Operation I

Unacceptable Operation j

I I

Heatup Rate I I 80 Deg. F/Hr

.i..

~

J Critical Limit I (

1 so Deg. F/Hr I

I Criticality Limit based on

~

inservlce hydrostatic test temperature (108 F) for the service period up to 51 EFPY Boltup 1

~

I Temp 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 2 Catawba Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 80°F/hr)

Applicable to the First 51 EFPY (Without Margins for Instrumentation Errors) 19

MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL Forging 05 & 04 LIMITING ART VALUES AT 51 EFPY:

1/4T, 47°F 3/4T,34°F 2500 2250 2000 1750 8 t 1500 I!!

0 0

~ 1250 D.

"C I

i 1000 CJ a 750 500 250 0

I I

Leak Test Limitl IOperlim Version:5.1 Run:12694 I

~

j Acceptable ~

)

Operation I

j Unacceptable Operation j I

Heatup Rate 100 Deg. F/Hr f I

'N Critical Limit.I 100 Deg. F/Hr I

I Criticality Limit based on inservice hydrostatic test temperature (108 F) for the service period up to 51 EFPV I

I Boltup I

~

I Temp I

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 3 Catawba Unit l Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

Applicable to the First 51 EFPY (Without Margins for Instrumentation Errors) 20

21 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE & LOWER SHELL FORGING 05 & 04 LIMITING ART VALUES AT 51 EFPY:

1/41, 47°F 3/4T, 34°F 2500 Q.() 2250 lfJ P-. 2000 CL.)

1750 s-

, 1500 lfJ lfJ Cl) 1250 s-0...

1000

'"O Cl)

.._;)

750 ce -::,

500 C_) -

C'O 250 u

0 Figure 4 I

I I

I I I I 1608780SD6 I

I I

I

~

I I

UNACCEPTABLE I

I OPERATION I

I I

I I I

I I

I I I I

I I I

J I

I I I I

I I

I I I

I I I I

I I

I I I I

I I I I I I

COOLDOWN I I I

1 1

I 1

1 I

I 1 RATES I I I

I I I

I I

F/Hr.

I I

I I I

I I

I i

I I I I I I

I I

I I I I

I I

I I I

I I

-a 0

I I i

I I

I I I I

I I 20

}

I I I

I i i I 1

i I I

40 I

I I

6 0 I

I 1 0 0 I

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-=r:

I I

ACCEPTABLE I

I I I I

I i I I

OPERATION I I I

I I I I l I

l I I

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I I I I I

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I I i

I I

l I I I

I I

I I w I I I

I I I I

I I I l

I l : I I

I I I I I I I

I I

i I I I I

I I

I 77 I I I

I I 1 l

I I I I

I I

I I I I

I I

I

! I I

I I

I I I I

I I

I 0

50 100 150 M o d e r a t o r I

I I

I l

I I

I I I I I

I I

I I I I

I I

I l I I

I I I I

I i I I I I I I I

I I I I I I I

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I I I I i I I I I I I I l I I I

I l I I

I I I I I I I I I I

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I l I I I I

I I I I I I

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I I I I 200 250 300 350 Temperature I

I I

I I I I

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l l I I

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I i I

--Ll+

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I I I I

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! I '

I 400 450 500

( D e g. F )

Catawba Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 51 EFPY (Without Margins for Instrumentation Errors)

22 TABLE 11 51 EFPY 60 °F/hr. Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Heatup Curves Conforuration #: 12694 60 Heatup Critical Limit Leak Test Limit T

p T

p 60 0

108 0

91 2000 60 1089 108 1141 108 2485 65 1141 110 1197 70 1197 115 1235 75 1235 120 1301 80 1301 125 1336 85 1366 130 1437 90 1437 135 1475 95 1475 140 1525 100 1525 145 1587 105 1587 150 1661 110 1661 155 1749 115 1749 160 1849 120 1849 165 1963 125 1963 170 2093 130 2093 175 2238 135 2238 180 2400 140 2400

Heatup Curves 80 Heatup T

60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 TABLE12 51 EFPY 80 °Flhr. Heatup Curve Data Points Using 1996 App. G

( without Uncertainties for Instrumentation Errors)

Confilroration #: 12694 Critical Limit Leak Test Limit p

T p

T 0

108 0

91 1089 108 1141 108 1141 110 1197 1197 115 1231 1231 120 1293 1293 125 1357 1357 130 1413 1413 135 1435 1435 140 1468 1468 145 1512 1512 150 1567 1567 155 1632 1632 160 1709 1709 165 1798 1798 170 1899 1899 175 2014 2014 180 2143 2143 185 2288 2288 190 2449 2449 23 p

2000 2485

Heatup Curves 100 Heatup T

60 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 TABLE13 51 EFPY 100 °Flhr. Heatup Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Confiruration #: 12694 Critical Limit Leak Test Limit p

T p

T 0

108 0

91 1089 108 1141 108 1141 110 1197 1197 115 1227 1227 120 1286 1286 125 1349 1349 130 1399 1399 135 1411 1411 140 1432 1432 145 1463 1463 150 1503 1503 155 1553 1553 160 1613 1613 165 1682 1682 170 1763 1763 175 1854 1854 180 1958 1958 185 2075 2075 190 2207 2207 195 2353 2353 24 p

2000 2485

Cooldown Cutves Steady State T

p 60 0

60 1089 65 1141 70 ll97 75 1260 80 1329 85 1406 90 1490 95 1583 100 1687 105 1801 110 1927 115 2066 120 2220 125 2391 TABLE 14 51 EFPY Cooldown Curve Data Points Using 1996 App. G (without Uncertainties for Instrumentation Errors)

Configuration#: 1508760398 20F 40F 60F T

p T

p 60 0

60 0

60 1086 60 1086 60 1086 65 1141 65 1141 65 1141 70 1197 70 1197 70 1197 75 1260 75 1260 75 1260 80 1329 80 1329 80 1329 85 1406 85 1406 85 1406 90 1490 90 1490 90 1490 95 1583 95 1583 95 1583 100 1687 100 1687 100 1687 105 1801 105 1801 105 1801 110 1927 110 1927 110 1927 ll5 2066 115 2066 115 2066 120 2220 120 2220 120 2220 125 2391 125 2391 125 2391 25 lOOF T

p 60 0

60 1086 65 1141 70 1197 75 1260 80 1329 85 1406 90 1490 95 1583 100 1687 105 1801 110 1927 115 2066 120 2220 125 2391

6 REFERENCES

1.

WCAP-15117, "Analysis of Capsule V and the Dosimeters from Capsules U and X from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program," Ed Terek, et al., Dated October 1998.

26

2.

WCAP-15118, "Catawba Unit 1 Heatup and Cooldown Limit Curves For Normal Operation", Ed Terek, Dated October, 1998.

3.

ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Secti_on XI, Division l ", February 26, 1999.

4.

Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

5.

Code of Federal Regulations, IO CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

6.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, "Fracture Toughness Criteria for Protection Against Failure.", Dated 1989 & December 1995.

7.

1989 Section m, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels."

8.

WCAP-14040-NP-A, Revision 2, "Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

9.

CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1999.

10.

WCAP-14799, "Analysis of Capsule W from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et. al., March 1998.

11.

WCAP-15046, "Analysis of Capsule U from the Tennessee Valley Authority Watts Bar Unit I Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., June 1998.

12.

WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves," W. S. Hazelton, et al., April 1975.

13.

WCAP-15315, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation For Operating PWR and BWR Plants", W. Bamford, et.al., October 1999.

to RA-18-0036 Westinghouse Report WCAP-17669, Revision 0, Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations

WCAP-17669-NP Revision 0 Westinghouse Non-Proprietary Class 3 June 2013 Catawba Unit 1 Measurement Uncertainty Recapture {MUR)

Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations 8 Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066

© 2013 Westinghouse Electric Company LLC All Rights Reserved WCAP-17669-NP Revision 0 Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations Amy E. Freed*

Materials Center of Excellence - I Jianwei Chen*

Radiation Engineering and Analysis June 2013 Reviewer:

Elliot J. Long*

Materials Center of Excellence - I Reviewer:

Greg A. Fischer*

Radiation Engineering and Analysis Approved: Frank C. Gift*, Manager Materials Center of Excellence - I Approved: Laurent P. Houssay*, Manager Radiation Engineering and Analysis

Electronically approved records are authenticated in the electronic document management system.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii WCAP-17669-NP June 2013 Revision 0 TABLE OF CONTENTS LIST OF TABLES....................................................................................................................................... iv LIST OF FIGURES..................................................................................................................................... vi EXECUTIVE

SUMMARY

......................................................................................................................... vii 1

METHOD DISCUSSION............................................................................................................. 1-1 2

CALCULATED NEUTRON FLUENCE..................................................................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS........................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES........................................................................ 2-3 3

MATERIAL PROPERTY INPUT................................................................................................. 3-1 4

SURVEILLANCE DATA............................................................................................................. 4-1 5

CHEMISTRY FACTORS............................................................................................................. 5-1 6

PRESSURIZED THERMAL SHOCK CALCULATIONS.......................................................... 6-1 7

UPPER-SHELF ENERGY CALCULATIONS............................................................................ 7-1 8

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES APPLICABILITY......................................................................................................................... 8-1 8.1 MUR POWER UPRATE ART CALCULATIONS.......................................................... 8-2 8.2 P-T LIMIT CURVES APPLICABILITY EVALUATION............................................... 8-6 9

SURVEILLANCE CAPSULE WITHDRAWAL

SUMMARY

..................................................... 9-1 10 REFERENCES........................................................................................................................... 10-1 APPENDIX A SURVEILLANCE DATA CREDIBILITY EVALUATION............................................ A-1 APPENDIX B EMERGENCY RESPONSE GUIDELINE LIMITS...................................................... B-1 APPENDIX C VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS............................................................................................. C-1 C.1 NEUTRON DOSIMETRY............................................................................................. C-1 C.2 REFERENCES............................................................................................................. C-36

iv WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 LIST OF TABLES Table 1-1 Minimum Recommended Number of Surveillance Capsules and Their Withdrawal Schedule (Schedule in Terms of EFPY of the RV).......................................................... 1-5 Table 2-1 Pressure Vessel Material Locations for Catawba Unit 1.................................................. 2-4 Table 2-2 Catawba Unit 1 Calculated Neutron Fluence Projections at the RV Clad/Base Metal Interface at 26, 34, 48, and 54 EFPY............................................................................... 2-5 Table 2-3 Catawba Unit 1 Calculated Neutron Fluence at the RV Clad/Base Metal Interface for Cycles 1 through 22 and Future Projections.................................................................... 2-6 Table 2-4 Calculational Uncertainties.............................................................................................. 2-7 Table 3-1 Material Properties for the Catawba Unit 1 RV(a)............................................................ 3-2 Table 4-1 Catawba Unit 1 Surveillance Capsule Data..................................................................... 4-2 Table 4-2 McGuire Unit 2 and Watts Bar Unit 1 Surveillance Capsule Data for Weld Heat # 895075

......................................................................................................................................... 4-3 Table 5-1 Calculation of Catawba Unit 1 Position 2.1 Chemistry Factor Values Using Surveillance Capsule Test Results........................................................................................................ 5-2 Table 5-2 Summary of Catawba Unit 1 Positions 1.1 and 2.1 Chemistry Factors........................... 5-3 Table 6-1 RTPTS Calculations for the Catawba Unit 1 RV Materials at 54 EFPY(a)......................... 6-2 Table 7-1 Catawba Unit 1 Predicted Positions 1.2 and 2.2 USE Values at 54 EFPY...................... 7-2 Table 8.1-1 Calculation of the Catawba Unit 1 ART Values at the 1/4T Location for 34 EFPY........ 8-2 Table 8.1-2 Calculation of the Catawba Unit 1 ART Values at the 3/4T Location for 34 EFPY........ 8-3 Table 8.1-3 Calculation of the Catawba Unit 1 ART Values at the 1/4T Location for 54 EFPY........ 8-4 Table 8.1-4 Calculation of the Catawba Unit 1 ART Values at the 3/4T Location for 54 EFPY........ 8-5 Table 8.2-1 Summary of the Catawba Unit 1 Limiting ART Values used in the Applicability Evaluation of the Existing 34 EFPY RV Heatup and Cooldown Curves......................... 8-6 Table 8.2-2 Summary of the Catawba Unit 1 Limiting ART Values used in the Applicability Evaluation of the Existing 51 EFPY RV Heatup and Cooldown Curves......................... 8-6 Table 9-1 Catawba Unit 1 Surveillance Capsule Withdrawal Summary.......................................... 9-1 Table A-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Catawba Unit 1 Surveillance Capsule Data Only.......................................................................... A-4 Table A-2 Best-Fit Evaluation for Catawba Unit 1 Surveillance Materials Only............................ A-5 Table A-3 Mean Chemical Composition and Operating Temperature for Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1........................................................................................... A-6 Table A-4 Operating Temperature Adjustments for the Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Surveillance Capsule Data............................................................................ A-7

WESTINGHOUSE NON-PROPRIETARY CLASS 3 v

WCAP-17669-NP June 2013 Revision 0 Table A-5 Calculation of Weld Heat # 895075 Interim Chemistry Factor for the Credibility Evaluation Using Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Surveillance Capsule Data................................................................................................................... A-8 Table A-6 Best-Fit Evaluation for Surveillance Weld Metal Heat # 895075 Using Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Data................................................................... A-9 Table B-1 Evaluation of Catawba Unit ERG Limit Category......................................................... B-1 Table C-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors.................................. C-11 Table C-2 Calculated Fast Neutron Flux (E > 1.0 MeV) at Catawba Unit 1 Surveillance Capsule Center Core Midplane Elevation.................................................................................. C-12 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule Z.................................................................................................................................... C-14 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule A......

...................................................................................................................................... C-19 Table C-5 Comparison of M/C and BE Reaction Rates at the Surveillance Capsule Center........ C-25 Table C-6 Comparison of Calculated and BE Exposure Rates at the EVND Capsule Center from Catawba Unit 1............................................................................................................. C-27 Table C-7 Comparison of Calculated and BE Exposure Rates at the Surveillance Capsule Center from Catawba Unit 1..................................................................................................... C-29 Table C-8 Comparison of M/C Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions from Catawba Unit 1.................................................................................... C-30 Table C-9 Comparison of BE/C Exposure Rate Ratios for Surveillance Capsules from Catawba Unit 1.................................................................................................................................... C-30 Table C-10 Comparison of Calculated and BE Exposure Rates at the EVND Capsules at the Core Midplane of Catawba Unit 1......................................................................................... C-31 Table C-11 Comparison of Calculated and BE Exposure Rates at the EVND Capsules at Off-Midplane Positions of Catawba Unit 1......................................................................... C-32 Table C-12 Comparison of M/C Sensor Threshold Reaction Rates from Ex-Vessel Midplane Capsules at Catawba Unit 1......................................................................................................... C-33 Table C-13 Comparison of M/C Sensor Threshold Reaction Rates from Ex-Vessel Off-Midplane Capsules at Catawba Unit 1.......................................................................................... C-34 Table C-14 Comparison of BE/C Exposure Rate Ratios from Ex-Vessel Midplane Capsules at Catawba Unit 1............................................................................................................. C-35 Table C-15 Comparison of BE/C Exposure Rate Ratios from Ex-Vessel Off-Midplane Capsules at Catawba Unit 1............................................................................................................. C-35

vi WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 LIST OF FIGURES Figure 2-1 Catawba Unit 1 Reactor Geometry in r-at the Core Mid-plane - 12.5° Neutron Pad Configuration................................................................................................................... 2-8 Figure 2-2 Catawba Unit 1 Reactor Geometry in r-at the Core Mid-plane - 20.0° Neutron Pad Configuration................................................................................................................... 2-9 Figure 2-3 Catawba Unit 1 Reactor Geometry in r-at the Core Mid-plane - 22.5° Neutron Pad Configuration................................................................................................................. 2-10 Figure 2-4 Catawba Unit 1 Reactor Geometry in r-z Plane at 40 Azimuthal Angle...................... 2-11 Figure 2-5 Catawba Unit 1 Reactor Geometry in r-z Plane at 29 Azimuthal Angle...................... 2-12 Figure 7-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in USE as a Function of Copper and Fluence for Catawba Unit 1...................................................................................... 7-3

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii WCAP-17669-NP June 2013 Revision 0 EXECUTIVE

SUMMARY

This report presents the reactor vessel (RV) integrity and neutron fluence evaluations for the Catawba Unit 1 measurement uncertainty recapture (MUR) power uprate to 3469 MWt. The RV integrity evaluations must be shown to meet the applicable U.S. Nuclear Regulatory Commission (NRC) requirements through the end of the licensed operating period. Catawba Unit 1 is licensed for 60 years of operation, which pertains to 54 effective full power years (EFPY) and is deemed end-of-life extension (EOLE).

Appendix A contains the credibility evaluation for the Catawba Unit 1 surveillance materials.

Conclusions for the surveillance data credibility evaluation are contained in Appendix A of this report.

Appendix B contains the Emergency Response Guideline (ERG) limits classification for Catawba Unit 1.

The ERG limits were developed in order to establish guidance for operator action in the event of an emergency situation, such as a pressurized thermal shock (PTS) event. Conclusions for the ERG limits evaluation are contained in Appendix B of this report.

The conclusions to the RV integrity evaluations are as follows:

EOLE Pressurized Thermal Shock All of the Catawba Unit 1 RV materials are projected to remain below the 10 CFR 50.61 screening criteria values of 270F for forgings, and 300F for circumferentially oriented welds, through EOLE (54 EFPY).

See Section 6 for more details.

EOLE Upper-Shelf Energy All of the Catawba Unit 1 RV materials are projected to remain above the upper-shelf energy (USE) screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through EOLE (54 EFPY). See Section 7 for more details.

Applicability of Pressure-Temperature Limit Curves The current Catawba Unit 1 pressure-temperature (P-T) limit curves are contained in Technical Specifications Figures 3.4.3-1 and 3.4.3-2. With a re-evaluation of surveillance data credibility, a recalculation of chemistry factors, the consideration of MUR power uprate fluence projections, along with consideration of all RV materials that are projected to achieve surface fluence levels of 1 x 1017 n/cm2 or higher at 34 EFPY, the applicability of the current P-T limit curves decreased from 34 EFPY to 30.7 EFPY. See Section 8 for more details.

Surveillance Capsule Withdrawal Schedules All in-vessel surveillance capsules have been removed from the Catawba Unit 1 RV. The guidelines of ASTM E185-82 are met, as required by 10 CFR 50, Appendix H, with consideration of the MUR power uprate. Ex-Vessel Neutron Dosimetry is installed in Catawba Unit 1 such that neutron fluence may be monitored during the period of extended operation in accordance with the Generic Aging Lessons Learned (GALL) Report. See Section 9 for more details.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 WCAP-17669-NP June 2013 Revision 0 1

METHOD DISCUSSION Adjusted Reference Temperature Per Regulatory Guide 1.99, Revision 2 (Reference 1), the following equations and variables are to be used for calculating Adjusted Reference Temperature (ART) values at the clad/base metal interface and at the RV 1/4-thickness (1/4T) and 3/4-thickness (3/4T) locations.

ART (°F) = Initial RTNDT + RTNDT + Margin

[Eqn. 1]

Where, Initial RTNDT (°F) = Reference temperature of the unirradiated material Margin (°F) 2 2

2

I

[Eqn. 2]

Where, I is the standard deviation for the Initial RTNDT (note that I is referred to as U in 10 CFR 50.61). If the initial RTNDT (reference nil-ductility transition temperature) is a measured value, I is estimated from the precision of the test method; per WCAP-14040-A, Revision 4 (Reference 2), I = 0°F when the initial RTNDT is a measured value. Per 10 CFR 50.61 (Reference 3), when the initial RTNDT is not a measured value and a generic mean initial RTNDT value is used for welds with the welding flux types identified in 10 CFR 50.61, then I = 17°F.

is the standard deviation for RTNDT.

For plates and forgings:

= 17°F when surveillance capsule data are not credible or not used*

= 8.5°F when credible surveillance capsule data are used*

For welds:

= 28°F when surveillance capsule data are not credible or not used*

= 14°F when credible surveillance capsule data are used*

not to exceed 0.5*RTNDT per Regulatory Guide 1.99, Revision 2 Shift in Reference Temperature (RTNDT) Calculations:

RTNDT (°F) = CF

FF

[Eqn. 3]

1-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0

Where, CF (°F) = chemistry factor based on the copper (Cu) and nickel (Ni) weight % of the material or based on the results of surveillance capsule test data. If the weight percent of Cu and Ni is used to determine the CF, then the CF is obtained from Table 1 or Table 2 of Regulatory Guide 1.99, Revision 2. If surveillance capsule data are used to determine the CF, then the CF is determined as follows:

[Eqn. 4]

Where:

n = The number of surveillance data points Ai = The measured value of RTNDT (°F)**

fi = fluence for each surveillance data point (x1019 n/cm2 (E > 1.0 MeV))

- If the surveillance weld copper and nickel content differs from that of the vessel weld, then the measured values of RTNDT (Ai in the preceding equation for CF) shall be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld (CFVW) to that for the surveillance weld (CFSW) based on the Cu and Ni content of the materials.

RTNDT (°F) = (measured RTNDT) * (CFVW / CFSW)

[Eqn. 5]

FF = fluence factor = f(0.28 - 0.10*log(f))

[Eqn. 6]

Where, f

= Vessel inner wall surface fluence, 1/4T fluence, or 3/4T fluence [x1019 n/cm2 (E > 1.0 MeV)]. The neutron fluence at any depth in the vessel wall is calculated as follows:

f (x1019 n/cm2 (E > 1.0 MeV)) = fsurf

e-0.24*(x)

[Eqn. 7]

Where, fsurf =

Vessel inner wall surface fluence, x1019 n/cm2 (E > 1.0 MeV) x

=

The depth into the vessel wall from the inner surface, inches Upper-Shelf Energy (USE)

The predicted decrease in USE is determined as a function of fluence and copper content using either of the following:

1. Figure 2 of Regulatory Guide 1.99, Revision 2, Position 1.2, or

2. Surveillance program test results and Figure 2 of Regulatory Guide 1.99, Revision 2, Position 2.2.

Both methods require the use of the 1/4 thickness (1/4T) vessel fluence.

])

[

])

[

1 log 20

.0 56

.0

(

1 log 10

.0 28

.0

(

n i

f i

n i

f i

i i

i f

f A

CF

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-3 WCAP-17669-NP June 2013 Revision 0 Reactor Vessel/Core Inlet Temperature (Tcold)

Regulatory Guide 1.99, Revision 2, Position 1.3 identifies limitations of applicability for the calculations of reference temperature and upper-shelf energy. Nominal irradiation temperature is one of the limitations, wherein Regulatory Guide 1.99 indicates that the nominal irradiation temperature for which the procedures are valid is 550°F. Irradiation below 525°F should be considered to produce greater embrittlement, and irradiation over 590°F may be considered to produce less embrittlement.

It is concluded that Catawba Unit 1 operates within the 525°F and 590°F range. Thus, the Regulatory Guide 1.99, Revision 2 correlations are applicable.

Pressurized Thermal Shock The PTS Rule, 10 CFR 50.61 (Reference 3), requires that for each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected PTS reference temperature (RTPTS) values accepted by the U.S. NRC for each RV beltline material at the end-of-life (EOL) fluence of the plant. This assessment must specify the basis for the projected value of RTPTS for each vessel beltline material, including the assumptions regarding core-loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation. This assessment must be updated whenever there is a significant change in projected values of RTPTS, or upon request for a change in the expiration date for operation of the facility. Changes to RTPTS values are considered significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewed term, if applicable, for the plant.

Per 10 CFR 50.61 (Reference 3), the following equations and variables are to be used for calculating RTPTS values at the clad/base metal interface of the vessel. RTPTS is also referred to as the EOL RTNDT.

RTPTS (°F) = IRTNDT + M + RTNDT

[Eqn. 8]

Where, IRTNDT (°F) = RTNDT(u) = Initial Unirradiated RTNDT value M = Margin (°F) 2 2

2

U

[Eqn. 9]

Where, U = 0°F when Initial RTNDT is a measured value U = 17°F when Initial RTNDT is a generic value

1-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 For plates and forgings:

= 17°F when surveillance capsule data are not credible or not used***

= 8.5°F when credible surveillance capsule data are used***

For welds:

= 28°F when surveillance capsule data are not credible or not used***

= 14°F when credible surveillance capsule data are used***

- - not to exceed 0.5*RTNDT per 10 CFR 50.61 (Reference 3)

RTNDT (°F) = CF

FF

[Eqn. 10]

Where, CF = chemistry factor (°F) calculated generically for copper (Cu) and nickel (Ni) content based on Tables 1 and 2 in Reference 3 for welds and base metal, respectively (also referred to as Position 1.1). It can also be calculated using credible surveillance capsule data per Equation 5 of Reference 3 (also referred to as Position 2.1).

FF = fluence factor = f(0.28 - 0.10*log(f)), where the normalized neutron fluence at the clad/base metal interface on the inside surface of the vessel is f = / (1.0 x 1019). The units for are n/cm2, E > 1.0 MeV.

The RTPTS screening criteria values are 270F for plates, forgings and axial weld materials and 300F for circumferential weld materials. All available surveillance data must be considered in the evaluation.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-5 WCAP-17669-NP June 2013 Revision 0 Surveillance Capsule Withdrawal Schedule Per ASTM E185-82 (Reference 4), Section 4.15, the RTNDT or adjustment in reference temperature is the difference in the 41 J (30 ft-lbf) index temperatures from the average Charpy curves measured before and after irradiation.

Per ASTM E185-82, Section 4.18, the USE level is the average energy value for all Charpy specimens (normally three) whose test temperature is above the upper end of the transition region. For specimens tested in sets of three at each test temperature, the set having the highest average may be regarded as defining the upper-shelf energy.

The surveillance capsule withdrawal schedule is generated based upon the guidelines specified in ASTM E185-82, Section 7.6. The minimum recommended number of surveillance capsules and their withdrawal times are identified in Table 1-1.

Table 1-1 Minimum Recommended Number of Surveillance Capsules and Their Withdrawal Schedule (Schedule in Terms of EFPY of the RV)

Predicted Transition Temperature Shift at Vessel Inside Surface

< 100°F

> 100°F & < 200°F

> 200°F Minimum Number of Capsules 3

4 5

Withdrawal Sequence EFPY First 6(a) 3(a) 1.5(a)

Second 15(b) 6(c) 3(d)

Third EOL(e) 15(b) 6(c)

Fourth EOL(e) 15(b)

Fifth EOL(e)

Notes:

(a) Or at the time when the accumulated neutron fluence of the capsule exceeds 5 x 1018 n/cm2, or at the time when the highest predicted RTNDT of all encapsulated materials is approximately 50°F, whichever comes first (b) Or at the time when the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the RV inner wall location, whichever comes first (c) Or at the time when the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the RV 1/4T location, whichever comes first (d) Or at the time when the accumulated neutron fluence of the capsule corresponds to a value midway between that of the first and third capsules (e) Not less than once or greater than twice the peak EOL vessel fluence. This may be modified on the basis of previous tests. This capsule may be held without testing following withdrawal

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 WCAP-17669-NP June 2013 Revision 0 2

CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates SN transport analysis was performed for the Catawba Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant-and fuel-cycle-specific basis. The neutron dosimetry sensor sets removed from the five previously withdrawn surveillance capsules [Z, Y, V, U, and X] and Ex-Vessel Neutron Dosimetry (EVND) capsules [A, B, C, D, E, and F] were re-analyzed using the current dosimetry evaluation methodology and updated neutron transport calculations. These dosimetry evaluations were used to validate the plant-specific neutron transport calculations applicable to Catawba Unit 1 and are described in Appendix C. These validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessels for operating periods extending to 60 EFPY.

All of the calculations described in this section were based on nuclear cross-section data derived from ENDF/B-VI.3 and made use of the latest available calculational tools. Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 (Reference 5). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-16083-NP-A, Revision 0 (Reference 6) using the state-of-the-art three-dimensional parallel discrete ordinates radiation transport code, RAPTOR-M3G (Reference 7).

RAPTOR-M3G is a three-dimensional (3-D) parallel discrete ordinates (SN) radiation transport code. The methodology employed by RAPTOR-M3G is identical to the methodology employed by the TORT code, with a number of evolutionary solution enhancements resulting from the last two decades of research.

RAPTOR-M3G has been designed from the ground-up as a parallel processing code, and adheres to modern best practices of software development. It has been rigorously tested against the TORT code (Reference 8) and benchmarked on an extensive set of real-world problems. The detailed benchmark for RAPTOR-M3G is described in Reference 7.

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Catawba Unit 1 RV, a series of fuel cycle-specific forward transport calculations were carried out using RAPTOR-M3G for each operating cycle at Catawba Unit 1.

For the Catawba Unit 1 transport calculations, the models depicted in Figures 2-1 through 2-3 were utilized since, with the exception of the neutron pads, the reactor is octant symmetric. In each of these figures, a single octant is depicted showing the arrangement of neutron pads and surveillance capsules as applicable. With regard to these three geometries, it should be noted that the maximum exposure of the pressure vessel occurs in octants with the 12.5° neutron pad span where no surveillance capsules are present. Further, the surveillance capsules are located in octants with either the 20.0° or 22.5° neutron pad span.

In addition to the core, reactor internals, pressure vessel, and primary biological shield, the RAPTOR-M3G models developed for these octant geometries included explicit representations of the surveillance capsules, the pressure vessel cladding, and the insulation located external to the pressure vessel.

2-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 From a neutronic standpoint, the inclusion of the surveillance capsules and associated support structure in the analytical model is significant. Since the presence of the capsules and structure has a marked impact on the magnitude of the neutron flux as well as on the relative neutron and gamma ray spectra at dosimetry locations within the capsules, a meaningful evaluation of the radiation environment internal to the capsules can be made only when these perturbation effects are properly accounted for in the analysis.

In developing the RAPTOR-M3G analytical models of the reactor geometry shown in Figures 2-1 through 2-3, nominal design dimensions were employed for the various structural components. Likewise, water temperatures and coolant density in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. These coolant temperatures were varied on a cycle-specific basis and are described in more detail later in this section. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera.

The RAPTOR-M3G geometric mesh description of the reactor models shown in Figures 2-1 through 2-3 consisted of 209 radial by 195 azimuthal by 179 vertical intervals for each neutron pad configuration.

Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the RAPTOR-M3G calculations was set at a value of 0.001.

A section view of the RAPTOR-M3G model of the Catawba Unit 1 reactor in the r,z plane is shown in Figures 2-4 and 2-5. The model extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation approximately six feet below the active fuel to approximately five feet above the active fuel. As can be seen in the figures, the stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The volume fractions utilized for the fuel region, the bypass region, the downcomer region, the reactor pressure vessel insulation region, and the volume fractions for the regions above and below the active core were treated as a homogeneous mixture of composing materials. In regions containing reactor coolant, the coolant temperatures were varied on a cycle-specific basis and are described in more detail later in this section.

The data utilized for the core power distributions in the plant-specific transport analyses included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. This information was used to develop spatial-and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron flux, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies.

From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the RAPTOR-M3G parallel discrete ordinates code, version 2.0 (Reference 7), and the BUGLE-96 cross-section library (Reference 9). The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 WCAP-17669-NP June 2013 Revision 0 scattering was treated with a P3 Legendre expansion or higher and angular discretization was modeled with an S8 order of angular quadrature or higher. Energy-and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

In Table 2-1, locations of the lower shell B (bottom head ring) to lower vessel head circumferential weld, lower shell A to lower shell B circumferential weld, intermediate shell to lower shell A circumferential weld, and upper shell to intermediate shell circumferential weld, and outlet/inlet nozzle to upper shell welds are given for Catawba Unit 1. These locations are given relative to the origin of the radiation transport model. Note that the Catawba Unit 1 RV does not have any longitudinal welds.

Selected results from the neutron transport analyses are provided in Tables 2-2 and 2-3. In Table 2-2, calculated fast neutron (E > 1.0 MeV) fluence for selected RV materials at the pressure vessel clad/base metal interface is provided at future projections to 26, 34, 48 and 54 EFPY. Cycle-specific calculations were performed for Cycles 1 to 21, where a core thermal power of 3411 MWt was used. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 22 will be representative of future plant operation with an uprated core power at 3469 MWt. In Table 2-3, calculated fast neutron (E > 1.0 MeV) fluence at the pressure vessel clad/base metal interface is provided for Cycles 1 through 21 and future projections for Catawba Unit 1, at various azimuthal locations. Please note that the fluence values reported in Tables 2-2 and 2-3 did not consider 0.11 EFPY of neutron exposure resulting from pre-commercial operation of Catawba Unit 1, between January 1985 and June 1985. The impact of this omission on the fluence evaluation results has been assessed to be negligible (less than 0.3% of the cumulative fast neutron fluence at 60 EFPY). Future fluence evaluations will consider the pre-commercial operation phase of Catawba Unit 1.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Catawba Unit 1 reactor pressure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments.

4. Comparisons of the plant-specific calculations with all available dosimetry results from the Catawba Unit 1 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily method-related and would tend to apply generically to all fast neutron exposure evaluations.

2-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Catawba Unit 1 analyses was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Catawba Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures.

Table 2-4 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 6. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix C of this report support these uncertainty assessments for Catawba Unit 1.

Table 2-1 Pressure Vessel Material Locations for Catawba Unit 1 Material Axial Location(b)

(cm)

Azimuthal Location (°)

r-Neutron Pad Configuration used in Exposure Calculations Lower shell B (bottom head ring) to lower vessel head circumferential weld

-312.577 0 to 360 12.5° neutron pad Lower shell A to lower shell B (bottom head ring) circumferential weld

-208.877 0 to 360 12.5° neutron pad Intermediate shell to lower shell A circumferential weld 12.023 0 to 360 12.5° neutron pad Upper shell to intermediate shell circumferential weld 224.723 0 to 360 12.5° neutron pad Outlet Nozzle to Upper Shell Weld 1(a) 275.713 22 12.5° neutron pad Outlet Nozzle to Upper Shell Weld 2(a) 158 12.5° neutron pad Outlet Nozzle to Upper Shell Weld 3(a) 202 12.5° neutron pad Outlet Nozzle to Upper Shell Weld 4(a) 338 12.5° neutron pad Inlet Nozzle to Upper Shell Weld 1(a) 264.713 67 12.5° neutron pad Inlet Nozzle to Upper Shell Weld 2(a) 113 12.5° neutron pad Inlet Nozzle to Upper Shell Weld 3(a) 247 12.5° neutron pad Inlet Nozzle to Upper Shell Weld 4(a) 293 12.5° neutron pad Notes:

(a) Lowest extent (b) Relative to the origin of the radiation transport model

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-5 WCAP-17669-NP June 2013 Revision 0 Table 2-2 Catawba Unit 1 Calculated Neutron Fluence Projections at the RV Clad/Base Metal Interface at 26, 34, 48, and 54 EFPY RV Material Fluence(a)

(n/cm2, E > 1.0 MeV) 26 EFPY 34 EFPY 48 EFPY 54 EFPY Outlet Nozzle to Upper Shell Welds (Lowest Extent) 1, 2, 3, and 4 1.49E+16 1.85E+16 2.47E+16 2.74E+16 Inlet Nozzle to Upper Shell Welds (Lowest Extent) 1, 2, 3, and 4 3.08E+16 3.82E+16 5.12E+16 5.67E+16 Upper Shell Forging 6.57E+17 8.01E+17 1.05E+18 1.16E+18 Intermediate Shell Forging 1.37E+19 1.72E+19 2.34E+19 2.60E+19 Lower Shell Forging A 1.37E+19 1.72E+19 2.34E+19 2.60E+19 Lower Shell Forging B (Bottom Head Ring) 1.10E+18 1.34E+18 1.77E+18 1.95E+18 Upper Shell to Intermediate Shell Circumferential Weld 6.57E+17 8.01E+17 1.05E+18 1.16E+18 Intermediate Shell to Lower Shell A Circumferential Weld 1.37E+19 1.72E+19 2.34E+19 2.60E+19 Lower Shell A to Lower Shell B Circumferential Weld 1.10E+18 1.34E+18 1.77E+18 1.95E+18 Lower Shell B to Lower Vessel Head Circumferential Weld 1.65E+14 2.03E+14 2.71E+14 3.00E+14 Lower Shell B to Lower Vessel Head Circumferential Weld at Outside of Reactor Pressure Vessel(b) 2.99E+15 3.73E+15 5.02E+15 5.58E+15 Notes:

(a) Extended beltline materials are currently interpreted to be the RV materials that will be exposed to a neutron fluence greater than or equal to 1 x 1017 n/cm2 (E > 1.0 MeV) at the end of design life of the vessel (54 EFPY). Only the materials that are projected to experience a fluence value of at least 1 x 1017 n/cm2 (E > 1.0 MeV) will be included in the subsequent evaluations contained within this report.

(b) For locations far away from the reactor active core midplane, the neutron streaming along the reactor cavity causes the maximum fast neutron fluence exposure occur at outside of the reactor pressure vessel instead of clad/base metal interface.

2-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 2-3 Catawba Unit 1 Calculated Neutron Fluence at the RV Clad/Base Metal Interface for Cycles 1 through 22 and Future Projections Cycle ID Cumulative Cycle Time (EFPY)

Fluence (n/cm2, E > 1.0 MeV)

Maximum Fluence (n/cm2, E >

1.0 MeV) 0° 15° 21° 22° 30° 45° 1

0.79 3.67E+17 5.64E+17 6.71E+17 6.74E+17 6.66E+17 7.59E+17 7.59E+17 2

1.54 7.19E+17 1.02E+18 1.14E+18 1.14E+18 1.11E+18 1.18E+18 1.18E+18 3

2.31 1.04E+18 1.51E+18 1.70E+18 1.70E+18 1.61E+18 1.67E+18 1.70E+18 4

3.17 1.42E+18 2.03E+18 2.28E+18 2.28E+18 2.13E+18 2.15E+18 2.28E+18 5

3.96 1.74E+18 2.49E+18 2.79E+18 2.78E+18 2.60E+18 2.62E+18 2.79E+18 6

4.98 2.17E+18 3.13E+18 3.52E+18 3.51E+18 3.21E+18 3.15E+18 3.52E+18 7

5.93 2.53E+18 3.68E+18 4.15E+18 4.13E+18 3.79E+18 3.73E+18 4.15E+18 8

7.00 2.90E+18 4.30E+18 4.87E+18 4.85E+18 4.41E+18 4.24E+18 4.87E+18 9

8.17 3.28E+18 4.93E+18 5.59E+18 5.56E+18 5.03E+18 4.80E+18 5.59E+18 10 9.29 3.63E+18 5.47E+18 6.21E+18 6.19E+18 5.66E+18 5.45E+18 6.21E+18 11 10.48 3.94E+18 5.94E+18 6.76E+18 6.74E+18 6.19E+18 5.95E+18 6.76E+18 12 11.85 4.32E+18 6.52E+18 7.45E+18 7.43E+18 6.91E+18 6.74E+18 7.45E+18 13 13.25 4.70E+18 7.10E+18 8.13E+18 8.11E+18 7.59E+18 7.50E+18 8.13E+18 14 14.69 5.06E+18 7.64E+18 8.77E+18 8.76E+18 8.26E+18 8.22E+18 8.77E+18 15 15.99 5.40E+18 8.18E+18 9.41E+18 9.40E+18 8.86E+18 8.74E+18 9.41E+18 16 17.35 5.71E+18 8.70E+18 1.00E+19 1.00E+19 9.50E+18 9.36E+18 1.00E+19 17 18.68 6.02E+18 9.18E+18 1.06E+19 1.06E+19 1.01E+19 9.90E+18 1.06E+19 18 20.05 6.36E+18 9.69E+18 1.12E+19 1.12E+19 1.06E+19 1.04E+19 1.12E+19 19 21.38 6.68E+18 1.02E+19 1.18E+19 1.18E+19 1.12E+19 1.09E+19 1.18E+19 20 22.82 7.01E+18 1.07E+19 1.24E+19 1.24E+19 1.17E+19 1.14E+19 1.24E+19 21 24.18 7.33E+18 1.11E+19 1.29E+19 1.29E+19 1.23E+19 1.21E+19 1.29E+19 22 25.62 7.67E+18 1.17E+19 1.35E+19 1.35E+19 1.29E+19 1.26E+19 1.35E+19 26.00 7.76E+18 1.18E+19 1.37E+19 1.37E+19 1.30E+19 1.27E+19 1.37E+19 30.00 8.73E+18 1.33E+19 1.54E+19 1.55E+19 1.47E+19 1.42E+19 1.55E+19 34.00 9.69E+18 1.48E+19 1.72E+19 1.72E+19 1.63E+19 1.56E+19 1.72E+19 40.00 1.11E+19 1.70E+19 1.98E+19 1.99E+19 1.88E+19 1.78E+19 1.99E+19 44.00 1.21E+19 1.85E+19 2.16E+19 2.16E+19 2.05E+19 1.93E+19 2.16E+19 48.00 1.31E+19 2.00E+19 2.34E+19 2.34E+19 2.21E+19 2.07E+19 2.34E+19 54.00 1.45E+19 2.22E+19 2.60E+19 2.60E+19 2.46E+19 2.29E+19 2.60E+19 60.00 1.60E+19 2.45E+19 2.86E+19 2.87E+19 2.71E+19 2.51E+19 2.87E+19

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-7 WCAP-17669-NP June 2013 Revision 0 Table 2-4 Calculational Uncertainties Description Uncertainty Capsule Vessel IR PCA Comparisons 3%

3%

H. B. Robinson Comparisons 3%

3%

Analytical Sensitivity Studies 10%

11%

Additional Uncertainty for Factors not Explicitly Evaluated 5%

5%

Net Calculational Uncertainty 12%

13%

2-8 q-

~-

N-aL "I-

- c..

WESTINGHOUSE NON-PROPRIETARY CLASS 3 XR/YT TORT Cross-Section t.fesnes: 209R, 195, 1792 Sectr on at Z =

38.00 cm

- -Sh,I --

'-llolll -- - -llodoi tMlool The stainless steel regions include the core baffle, core barrel, thermal shield, and vessel clad.

33.1

&7.5 1D1.2

'IM.D 16&.a 202.s 2!6.2 -2~10~.o~m~...+-,~~~.......... ~R

[cm]

Figure 2-1 Catawba Unit 1 Reactor Geometry in r-0 at the Core Mid-plane -12.5° Neutron Pad Configuration WCAP-17669-NP June 2013 Revision 0

l.:

N--

J.:

l-

"'1 -

111-... -

WESTINGHOUSE NON-PROPRIETARY CLASS 3 XR/YT TORT Cross-Section Weal\\tit i.09R,1&59.179Z Sectfon Cit Z =

38.00 cm fllllll,1111 -- -

The stainless steel regions include the core baffle, core barrel, neutron pad, surveillance capsule holder, and vessel clad The carbon steel regions include the surveillance capsule specimens and pressure vessel.

2-9 d~o

!!.a

&7.!1 111U m.o aa 202.s 2H.2-.~21~0~.o~~~,+~........ ~.,.,...,......,R Figure 2-2 ltml Catawba Unit 1 Reactor Geometry in r-0 at the Core Mid-plane - 20.0° Neutron Pad Configuration WCAP-17669-NP June 2013 Revision 0

2-10 Figure 2-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 XR/YT TORT Cross-Section Mfflles: 209R,19~,179Z Sec:ffon of Z =

38.00 cm Nlllltllll -- -

The stainless steel regions include the core baffle, core barrel, neutron pad, surveillance capsule holder, and vessel clad. The carbon steel regions include the surveillance capsule specimens and pressure vessel.

Catawba Unit 1 Reactor Geometry in r-0 at the Core Mid-plane - 22.5° Neutron Pad Configuration WCAP-17669-NP June 2013 Revision 0

Figure 2-4 WCAP-17669-NP 0

E

(.)

N 0...

LLI 00 co 0,-..:

WESTINGHOUSE NON-PROPRIETARY CLASS 3 R/Z TORT Cross-Section Meshes: 209R,t958,179Z Sectlon at angle 8 :: 40° bi.L...j..__ _____

J ll.llOE+OO

-'.11E-t02 2-11 R

Catawba Unit 1 Reactor Geometry in r-z Plane at 40° Azimuthal Angle June 2013 Revision 0

2-12 Figure 2-5 WCAP-17669-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 R/Z TORT Cross-Section M~hes: 209R. 1958. 179Z Section at an!Jle 8 = 29° Catawba Unit 1 Reactor Geometry in r-z Plane at 29° Azimuthal Angle June 2013 Revision 0

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 WCAP-17669-NP June 2013 Revision 0 3

MATERIAL PROPERTY INPUT The fracture toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan (Reference 10). The beltline region of a RV, per 10 CFR 50.61 (Reference 3), is defined as:

the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

The beltline materials, as described in the paragraph above, must be considered in the RV integrity evaluations. Additionally, as described in Item IV.A2.R-84 of NUREG-1801, Revision 2 (Reference 11),

any materials with an EOLE fluence value exceeding 1.0 x 1017 n/cm2 (E > 1.0 MeV) must be considered in the RV integrity evaluations. The materials that exceed this threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met through EOLE.

Summaries of the best-estimate (BE) copper and nickel contents, RTNDT(U) values, and initial USE values for the RV materials are provided in Table 3-1 for Catawba Unit 1.

3-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 3-1 Material Properties for the Catawba Unit 1 RV(a)

Material Description Chemical Composition Fracture Toughness Properties RV Material and Identification Number Cu Wt. %

Ni Wt. %

Initial RTNDT (b)

(°F)

Initial USE (ft-lb)

Upper Shell (US) Forging 06 0.16(c) 0.85

-26 101 Intermediate Shell (IS) Forging 05 0.09 0.86

-8 134 Lower Shell (LS) Forging 04 0.04(d) 0.83

-13 134 Bottom Head Ring 03 0.06 0.77 14 68 US to IS Circumferential (Circ.) Weld W06 (Heat # 899680) 0.03(e) 0.75(e) 10(e) 92(f)

IS to LS Circ. Weld W05 (Heat # 895075) 0.04 0.72

-51 130 LS to Bottom Head Ring Weld W04 (Heat # 899680) 0.03(e) 0.75(e) 10(e) 92(f)

Catawba Unit 1 Surveillance Weld (Heat # 895075) 0.05(g) 0.73(g)

McGuire Unit 2 Surveillance Weld (Heat # 895075) 0.04(g) 0.74(g)

Watts Bar Unit 1 Surveillance Weld (Heat # 895075) 0.03(g) 0.75(g)

Notes:

(a) Values obtained from Tables 3.1.2-1 and 3.1.2-2 of WCAP-17175-P (Reference 12), unless otherwise noted.

(b) All initial RTNDT values are based on measured data.

(c) No weight percent copper value was reported in the Certified Material Test Report (CMTR). Therefore, the maximum copper weight percent value for A508 Class 2 forging materials is conservatively applied based on the generic data provided in Appendix G of ORNL document ORNL/TM-2006/530 (Reference 14).

(d) According to WCAP-17175-P (Reference 12), the weight percent copper value is 0.05. However, according to the CMTR, the weight percent copper value is 0.04, which is consistent with the value used in the Catawba License Renewal Application (LRA, [Reference 13]). Therefore, 0.04 wt. % copper is utilized in this evaluation.

(e) Values are based on measured data from the weld certification records for weld Heat # 899680. The initial RTNDT was determined using the measured data and the method described in Section B.1.1(4) of NUREG-0800 Branch Technical Position 5-3 (Reference 10). Note that the RTNDT for this value is higher than that reported in Table 3.1.2-2 of Reference 12; however, based on the CMTR data, the Charpy tests were performed at 10°F; therefore, the initial RTNDT should be 10°F, rather than 0°F.

(f) Value obtained from Table 4.2-3 of the McGuire and Catawba LRA.

(g) Information for the surveillance welds is taken from Table 4 of WCAP-15203, Revision 1 (Reference 15).

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 WCAP-17669-NP June 2013 Revision 0 4

SURVEILLANCE DATA Per 10 CFR 50.61, calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. Furthermore, Regulatory Guide 1.99, Revision 2 allows the use of data from the plant-specific surveillance program in determining the decrease in USE.

In addition to the plant-specific surveillance data, 10 CFR 50.61 also requires the use of data from surveillance programs at other plants which include the same limiting beltline material when calculating Position 2.1 chemistry factors. Data from a surveillance program at another plant are often called sister plant data.

Tables 4-1 and 4-2 summarize the Catawba Unit 1 surveillance data as well as surveillance data from McGuire Unit 2 and Watts Bar Unit 1. The McGuire Unit 2 and Watts Bar Unit 1 surveillance programs include weld Heat # 895075, which is the same weld heat as the Catawba Unit 1 Intermediate Shell to Lower Shell circumferential weld. Thus, the McGuire Unit 2 and Watts Bar Unit 1 data will be used in calculation of the Position 2.1 chemistry factor value for Catawba Unit 1 weld Heat # 895075.

4-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 4-1 Catawba Unit 1 Surveillance Capsule Data Material Capsule Withdrawal EFPY Lead Factor(a)

Capsule Fluence(a)

(x1019 n/cm2, E > 1.0 MeV)

Measured 30 ft-lb Transition Temperature Shift(b) (°F)

Measured USE Decrease(b)

(%)

IS Forging 05 (Axial)

Z 0.79 3.85 0.292 15.74 0(d)

Y 4.98 3.73 1.31 48.63 4

V 9.29 3.72 2.31 50.58 1

X(c) 9.29 3.88 2.41 U(c) 9.29 3.88 2.41 W(c) 14.69 4.00 3.51 IS Forging 05 (Tangential)

Z 0.79 3.85 0.292 0.0(e) 0 Y

4.98 3.73 1.31 19.09 9

V 9.29 3.72 2.31 25.61 10 X(c) 9.29 3.88 2.41 U(c) 9.29 3.88 2.41 W(c) 14.69 4.00 3.51 Surveillance Weld Material (Heat # 895075)

Z 0.79 3.85 0.292 1.91 5

Y 4.98 3.73 1.31 17.79 2

V 9.29 3.72 2.31 26.5 3

X(c) 9.29 3.88 2.41 U(c) 9.29 3.88 2.41 W(c) 14.69 4.00 3.51 Notes:

(a) The calculated fluence values and lead factors were updated as part of the MUR power uprate analysis.

(b) Values obtained from Table 5-10 of WCAP-15117 (Reference 16).

(c) Capsules X and U were removed and the dosimeters were analyzed, but the specimens were not. Capsule W was placed in the spent fuel pool following removal.

(d) Original value was measured as a +1 % increase, but an increase should not occur; therefore, a conservative value of zero will be used.

(e) Original value was -14.9°F, but physically a reduction should not occur; therefore, a value of zero will be used.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 WCAP-17669-NP June 2013 Revision 0 Table 4-2 McGuire Unit 2 and Watts Bar Unit 1 Surveillance Capsule Data for Weld Heat # 895075 Material Capsule Capsule Fluence (x1019 n/cm2, E > 1.0 MeV)

Measured 30 ft-lb Transition Temperature Shift

(°F)

Inlet Temperature

(°F)

Temperature Adjustment

(°F)

McGuire Unit 2 Surveillance Weld(a) (Heat # 895075)

V 0.302 38.51 557(b)

+4(d)

X 1.38 35.93 U

1.90 23.81 W

2.82 43.76 Watts Bar Unit 1 Surveillance Weld(a) (Heat # 895075)

U 0.447 0.0(c) 560

+7(d)

W 1.08 30.5 X

1.71 25.8 Z

2.40 13.9 Notes:

(a) Data pertaining to the McGuire Unit 2 and Watts Bar Unit 1 surveillance welds were taken from Tables 4.2-1 and 4.2-2 of WCAP-17455-NP (Reference 17),

respectively, unless otherwise noted.

(b) McGuire Unit 2 capsules were irradiated between Cycles 1 through 10. An average inlet temperature of 557°F was determined over the period of capsule irradiation.

(c) Original value was -6.4°F, but physically a reduction should not occur; therefore, a value of zero will be used.

(d) Temperature adjustment = 1.0*(Tcapsule - Tplant), where Tplant = 553°F for Catawba Unit 1 (applied to the weld RTNDT data for each of the McGuire Unit 2 and Watts Bar Unit 1 capsules in the Position 2.1 Chemistry Factor calculation).

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 WCAP-17669-NP June 2013 Revision 0 5

CHEMISTRY FACTORS As described in Section 1 of this report, Position 1.1 chemistry factors for each RV material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of 10 CFR 50.61. The best-estimate copper and nickel weight percents for the Catawba Unit 1 RV materials were provided in Table 3-1 of this report.

The Position 2.1 chemistry factors are calculated for the materials that have available surveillance program results. The calculation is performed using the method described in 10 CFR 50.61, which is also summarized in Section 1 of this report. The Catawba Unit 1 surveillance data as well as any applicable sister plant data were summarized in Section 4 of this report, and will be utilized in the Position 2.1 chemistry factor calculations in this Section.

The Position 2.1 chemistry factor calculations are presented in Table 5-1 for the Catawba Unit 1 RV materials contained in the radiation surveillance program. These values were calculated using the surveillance data summarized in Section 4 of this report. Additionally, surveillance data from McGuire Unit 2 and Watts Bar Unit 1 are utilized in Table 5-1 as it is applicable to the Catawba Unit 1 Intermediate Shell to Lower Shell circumferential weld (Heat # 895075).

All of the surveillance weld data are adjusted for irradiation temperature and chemical composition differences in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 (Reference 18).

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-2 for Catawba Unit 1.

Appendix A contains the credibility evaluation for each of the surveillance materials for which a chemistry factor is calculated in this Section. Margin will be applied to the calculations of ART and RTPTS according to the conclusions of the credibility evaluation for each of the surveillance materials.

5-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 5-1 Calculation of Catawba Unit 1 Position 2.1 Chemistry Factor Values Using Surveillance Capsule Test Results Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV)

FF(b)

RTNDT (c)

(°F)

FF*RTNDT

(°F)

FF2 IS Forging 05 (Axial)

Z 0.292 0.663 15.74 10.44 0.440 Y

1.31 1.075 48.63 52.28 1.156 V

2.31 1.226 50.58 62.03 1.504 IS Forging 05 (Tangential)

Z 0.292 0.663 0.0(d) 0.00 0.440 Y

1.31 1.075 19.09 20.52 1.156 V

2.31 1.226 25.61 31.41 1.504 SUM:

176.68 6.199 CFIS Forging = (FF

RTNDT) ÷ (FF2) = (176.68) ÷ (6.199) = 28.5F Catawba Unit 1 Surveillance Weld (Heat # 895075)

Z 0.292 0.663 1.51 (1.91) 1.00 0.440 Y

1.31 1.075 14.05 (17.79) 15.11 1.156 V

2.31 1.226 20.94 (26.5) 25.67 1.504 McGuire Unit 2 Surveillance Weld (Heat # 895075)

V 0.302 0.672 42.51 (38.51) 28.57 0.452 X

1.38 1.089 39.93 (35.93) 43.50 1.187 U

1.90 1.176 27.81 (23.81) 32.70 1.382 W

2.82 1.276 47.76 (43.76) 60.93 1.628 Watts Bar Unit 1 Surveillance Weld (Heat # 895075)

U 0.447 0.776 9.24 (0.0(d))

7.17 0.602 W

1.08 1.022 49.50 (30.5) 50.57 1.044 X

1.71 1.148 43.30 (25.8) 49.69 1.317 Z

2.40 1.236 27.59 (13.9) 34.10 1.528 SUM:

349.00 12.238 CFWeld Metal = (FF

RTNDT) ÷ (FF2) = (349.00) ÷ (12.238) = 28.5F Notes:

(a) f = fluence.

(b)

FF = fluence factor = f(0.28 - 0.10*log f).

(c)

RTNDT values are the measured 30 ft-lb shift values. All values are taken from Tables 4-1 and 4-2 of this report. The McGuire Unit 2 and Watts Bar Unit 1 surveillance weld RTNDT values have been adjusted according to the temperature adjustments summarized in Table 4-2 of this report then by using the ratio procedure to account for differences in the surveillance weld chemistry and the beltline weld chemistry (pre-adjusted values are listed in parentheses). For McGuire Unit 2, Ratio = CFVessel Weld / CFSurv. Weld = 54°F / 54°F = 1.00. For Watts Bar Unit 1, Ratio = 54°F / 41°F = 1.32. The Catawba Unit 1 surveillance weld RTNDT values are not adjusted for temperature differences, but are adjusted by a ratio for chemistry differences (pre-adjusted values are listed in parentheses). For Catawba Unit 1, Ratio = 54°F / 68°F = 0.79.

(d)

This RTNDT value was determined to be negative, but physically a reduction should not occur; therefore, a value of zero is used.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 WCAP-17669-NP June 2013 Revision 0 Table 5-2 Summary of Catawba Unit 1 Positions 1.1 and 2.1 Chemistry Factors RV Material and Identification Number Chemistry Factor (F)

Position 1.1(a)

Position 2.1(b)

US Forging 06 123.5 IS Forging 05 58 28.5 LS Forging 04 26 Bottom Head Ring 03 37 US to IS Circ. Weld W06 (Heat # 899680) 41 IS to LS Circ. Weld W05 (Heat # 895075) 54 28.5 LS to Bottom Head Ring Weld W04 (Heat # 899680) 41 Catawba Unit 1 Surveillance Weld (Heat # 895075) 68 McGuire Unit 2 Surveillance Weld (Heat # 895075) 54 Watts Bar Unit 1 Surveillance Weld (Heat # 895075) 41 Notes:

(a)

Position 1.1 Chemistry Factors were calculated using the copper and nickel weight percents presented in Table 3-1 of this report and Tables 1 and 2 of 10 CFR 50.61.

(b)

Position 2.1 Chemistry Factors taken from Table 5-1 of this report. Per Appendix A, the Catawba Unit 1 surveillance forging and weld data are credible.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 WCAP-17669-NP June 2013 Revision 0 6

PRESSURIZED THERMAL SHOCK CALCULATIONS A limiting condition on RV integrity known as PTS may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the RV under the following conditions:

Severe overcooling of the inside surface of the vessel wall followed by high repressurization, Significant degradation of vessel material toughness caused by radiation embrittlement, The presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling (10 CFR 50.61) on PTS (Reference 3) that established screening criteria on RV embrittlement, as measured by the maximum RTNDT in the limiting beltline component at the end-of-license, termed reference temperature for PTS (RTPTS). RTPTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic pressurized water reactor (PWR) vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end-of-license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions made the procedure for calculating the RTPTS values consistent with the methods given in Regulatory Guide 1.99, Revision 2 (Reference 1).

These accepted methods were used with the surface fluence of Section 2 to calculate the following RTPTS values for the Catawba Unit 1 RV materials at 54 EFPY (EOLE). The EOLE RTPTS calculations are summarized below in Table 6-1 for Catawba Unit 1.

PTS Conclusion For Catawba Unit 1, the limiting RTPTS value at 54 EFPY is 63°F (see Table 6-1); this value corresponds to US Forging 06. Therefore, all of the beltline and extended beltline materials in the Catawba Unit 1 reactor vessel are below the RTPTS screening criteria values of 270F for forgings, and 300F for circumferentially oriented welds through EOLE (54 EFPY).

The Alternate PTS Rule (10 CFR 50.61a [Reference 19]) was published in the Federal Register by the NRC in 2010. This alternate rule is less restrictive than the Mandatory PTS Rule (10 CFR 50.61) and is intended to be used for situations where the 10 CFR 50.61 criteria cannot be met. Catawba Unit 1 currently meets the criteria for the Mandatory PTS Rule through EOLE and therefore does not need to utilize the Alternate PTS Rule at this time.

6-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 6-1 RTPTS Calculations for the Catawba Unit 1 RV Materials at 54 EFPY(a)

RV Material and Identification Number R.G. 1.99, Rev. 2 Position CF(b)

(F)

Fluence(c)

(x1019 n/cm2, E > 1.0 MeV)

FF IRTNDT (d)

(F)

RTNDT (F)

U (d)

(°F)

(e)

(°F)

Margin (F)

RTPTS (F)

US Forging 06 1.1 123.5 0.116 0.4472

-26 55.2 0

17.0 34.0 63 IS Forging 05 1.1 58 2.60 1.2559

-8 72.8 0

17.0 34.0 99 Using credible surveillance data 2.1 28.5 2.60 1.2559

-8 35.8 0

8.5 17.0 45 LS Forging 04 1.1 26 2.60 1.2559

-13 32.7 0

16.3 32.7 52 Bottom Head Ring 03 1.1 37 0.195 0.5634 14 20.8 0

10.4 20.8 56 US to IS Circ. Weld W06 (Heat # 899680) 1.1 41 0.116 0.4472 10 18.3 0

9.2 18.3 47 IS to LS Circ. Weld W05 (Heat # 895075) 1.1 54 2.60 1.2559

-51 67.8 0

28.0 56.0 73 Using credible surveillance data 2.1 28.5 2.60 1.2559

-51 35.8 0

14.0 28.0 13 LS to Bottom Head Ring Weld W04 (Heat # 899680) 1.1 41 0.195 0.5634 10 23.1 0

11.5 23.1 56 Notes:

(a)

The 10 CFR 50.61 methodology was utilized in the calculation of the RTPTS values. See Section 1 of this report for details.

(b)

Taken from Table 5-2 of this report.

(c)

Taken from Table 2-2 of this report.

(d)

Initial RTNDT values are measured and are taken from Table 3-1 of this report.

(e)

Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible. Per the guidance of 10 CFR 50.61, the base metal = 17°F for Position 1.1 and, with credible surveillance data, = 8.5°F for Position 2.1; the weld metal = 28°F for Position 1.1 and, with credible surveillance data, = 14°F for Position 2.1.

However, need not exceed 0.5*RTNDT.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 WCAP-17669-NP June 2013 Revision 0 7

UPPER-SHELF ENERGY CALCULATIONS The requirements for USE are contained in 10 CFR 50, Appendix G (Reference 20). 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the USE of any RV material is predicted to drop below 50 ft-lb.

Regulatory Guide 1.99, Revision 2 defines two methods that can be used to predict the decrease in USE due to irradiation. The method to be used depends on the availability of credible surveillance capsule data. For RV beltline materials that are not in the surveillance program or are not credible, the Charpy USE (Position 1.2) is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2 (Reference 1).

When two or more credible surveillance data sets become available from the RV, they may be used to determine the Charpy USE of the surveillance materials. The surveillance data are then used in conjunction with Figure 2 of the Regulatory Guide to predict the decrease in USE (Position 2.2) of the RV materials due to irradiation. If the EOLE USE values calculated using Position 2.2 are most limiting, then they must be used regardless of the credibility of the surveillance data.

The 54 EFPY (EOLE) Position 1.2 USE values of the RV materials can be predicted using the corresponding 1/4T fluence projection, the copper content, and Figure 2 in Regulatory Guide 1.99, Revision 2.

The predicted Position 2.2 USE values are determined for the RV materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection. The reduced plant surveillance data were obtained from Table 5-10 of WCAP-15117 (Reference 16) for Catawba Unit 1. The surveillance data were plotted on Regulatory Guide 1.99, Revision 2, Figure 2 (see Figure 7-1 of this report) using the updated surveillance capsule fluence values documented in Table 4-1 of this report for Catawba Unit 1. These data were fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines were used instead of the existing lines to determine the Position 2.2 EOLE USE values.

The projected USE values were calculated to determine if the Catawba Unit 1 RV materials remain above the 50 ft-lb limit at EOLE. These calculations are summarized in Table 7-1 for Catawba Unit 1.

USE Conclusion For Catawba Unit 1, the limiting USE value at 54 EFPY is 60 ft-lb (see Table 7-1); this value corresponds to Bottom Head Ring 03. Therefore, all of the beltline and extended beltline materials in the Catawba Unit 1 RV are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through EOLE (54 EFPY).

7-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 7-1 Catawba Unit 1 Predicted Positions 1.2 and 2.2 USE Values at 54 EFPY RV Material and Identification Number Wt. %

Cu(a) 1/4T EOLE Fluence(b)

(x1019 n/cm2, E > 1.0 MeV)

Unirradiated USE(a)

(ft-lb)

Projected USE Decrease

(%)

Projected EOLE USE (ft-lb)

US Forging 06 0.16 0.070 101 14 87 IS Forging 05 0.09 1.565 134 21(c) 106 Using surveillance data 0.09 1.565 134 10(d) 121 LS Forging 04 0.04 1.565 134 21(c) 106 Bottom Head Ring 03 0.06 0.117 68 12(c) 60 US to IS Circ. Weld W06 (Heat # 899680) 0.03 0.070 92 10(c) 83 IS to LS Circ. Weld W05 (Heat # 895075) 0.04 1.565 130 21(c) 103 Using surveillance data 0.04 1.565 130 8(d) 120 LS to Bottom Head Ring Weld W04 (Heat # 899680) 0.03 0.117 92 12(c) 81 Notes:

(a)

From Table 3-1 of this report.

(b) 1/4T fluence was calculated using Equation (3) of Regulatory Guide 1.99, Revision 2, and the Catawba Unit 1 RV beltline wall thickness of 8.465 inches.

(c)

Percentage USE decrease is conservatively based on lowest Cu Wt. % chemistry line (0.05% for weld and 0.10% for base metal) delineated in Figure 2 of Regulatory Guide 1.99, Revision 2.

(d)

Percentage USE decrease is based on Position 2.2 of Regulatory Guide 1.99, Revision 2 using data from Table 4-1. Credibility Criterion 3 in the Discussion section of Regulatory Guide 1.99, Revision 2, indicates that even if the surveillance data are not considered credible for determination of RTNDT, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82. Regulatory Guide 1.99, Revision 2, Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of the Guide) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 WCAP-17669-NP June 2013 Revision 0 Figure 7-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in USE as a Function of Copper and Fluence for Catawba Unit 1 1

10 100 1.00E+17 1.00E+18 1.00E+19 1.00E+20 Percentage Drop in USE Neutron Fluence, n/cm2 (E > 1 MeV)

Surveillance Material: IS Forging 05 Surveillance Material: Weld Heat # 895075 Upper Limit

% Copper Base Metal Weld 0.35 0.30 0.30 0.25 0.25 0.20 0.20 0.15 0.15 0.10 0.10 0.05 weld line ISandLSForgingsand IStoLSCirc.Weld 54 EFPY1/4TFluence=

1.565x1019 n/cm2 BottomHeadRingandLSto Bottom HeadRingWeld54EFPY 1/4TFluence=0.117x1019 n/cm2 forging line USForgingandUStoISCirc.

Weld54EFPY1/4T Fluence=

0.070x1019 n/cm2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 WCAP-17669-NP June 2013 Revision 0 8

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES APPLICABILITY Heatup and cooldown limit curves are calculated using the most limiting values of RTNDT corresponding to the limiting RV material. The most limiting RV material RTNDT values are determined by using the unirradiated RV material fracture toughness properties and estimating the irradiation-induced shift (RTNDT).

RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactor's life, RTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. Using the ART values, pressure-temperature (P-T) limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G (Reference 20), as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 21).

The P-T limit curves for normal heatup and cooldown of the primary reactor coolant system for Catawba Unit 1 were previously developed in WCAP-15203, Revision 1 (Reference 15) for 34 EFPY and WCAP-15448, Revision 1 (Reference 22) for 51 EFPY. The existing 34 and 51 EFPY P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material.

To determine the applicability of the Catawba Unit 1 P-T limit curves developed in WCAP-15203, Revision 1 (Reference 15) for 34 EFPY and in WCAP-15448, Revision 1 (Reference 22) for 51 EFPY, the limiting reactor vessel material ART values with consideration of the MUR power uprate are compared to the limiting beltline material ART values used in development of the existing 34 EFPY and 51 EFPY P-T limit curves contained in References 15 and 22. The Regulatory Guide 1.99, Revision 2 (Reference 1) methodology was used along with the surface fluence of Section 2 to calculate ART values for the Catawba Unit 1 reactor vessel materials at 34 EFPY and 54 EFPY. Note that the 54 EFPY ART values calculated as part of this MUR uprate evaluation will be used when assessing the applicability of the existing 51 EFPY P-T limit curves. 54 EFPY ART values (as opposed to51 EFPY values) were calculated to correspond with the other RVI evaluations calculated as part of this MUR power uprate analysis at EOLE. The ART calculations with consideration of the MUR power uprate are summarized in Tables 8.1-1 through 8.1-4 for Catawba Unit 1.

Existing P-T Limit Curves Applicability Conclusions Comparisons of the limiting MUR power uprate ART values to those used in calculation of the existing P-T limit curves are contained in Tables 8.2-1 and 8.2-2 for Catawba Unit 1. With a re-evaluation of surveillance data credibility, a recalculation of chemistry factors, and the consideration of all RV materials projected to achieve surface fluence levels of 1 x 1017 n/cm2 or higher, the applicability date of the existing Catawba Unit 1 P-T limit curves decreased from 34 EFPY to 30.7 EFPY with the MUR power uprate. Similarly, the applicability date of the 51 EFPY P-T limit curves for Catawba Unit 1 decreased from 51 EFPY to 42.7 EFPY with the MUR power uprate. For more detailed conclusions, refer to Section 8.2 for Catawba Unit 1.

8-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 8.1 MUR POWER UPRATE ART CALCULATIONS Table 8.1-1 Calculation of the Catawba Unit 1 ART Values at the 1/4T Location for 34 EFPY RV Material and Identification Number R.G. 1.99, Rev. 2 Position CF(a)

(F) 1/4T Fluence(b)

(x1019 n/cm2, E > 1.0 MeV) 1/4T FF(b)

IRTNDT (c)

(F)

RTNDT (F)

I (c)

(°F)

(d)

(°F)

Margin (F)

ART (F)

US Forging 06 1.1 123.5 0.048 0.2870

-26 35.4 0

17.0 34.0 43 IS Forging 05 1.1 58 1.035 1.0096

-8 58.6 0

17.0 34.0 85 Using credible surveillance data 2.1 28.5 1.035 1.0096

-8 28.8 0

8.5 17.0 38 LS Forging 04 1.1 26 1.035 1.0096

-13 26.3 0

13.1 26.3 40 Bottom Head Ring 03 1.1 37 0.081 0.3752 14 13.9 0

6.9 13.9 42 US to IS Circ. Weld W06 (Heat # 899680) 1.1 41 0.048 0.2870 10 11.8 0

5.9 11.8 34 IS to LS Circ. Weld W05 (Heat # 895075) 1.1 54 1.035 1.0096

-51 54.5 0

27.3 54.5 58 Using credible surveillance data 2.1 28.5 1.035 1.0096

-51 28.8 0

14.0 28.0 6

LS to Bottom Head Ring Weld W04 (Heat # 899680) 1.1 41 0.081 0.3752 10 15.4 0

7.7 15.4 41 Notes:

(a)

Taken from Table 5-2 of this report.

(b) 1/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2, and the Catawba Unit 1 RV beltline wall thickness of 8.465 inches.

(c)

Initial RTNDT values are measured and are taken from Table 3-1 of this report.

(d)

Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and, with credible surveillance data, = 8.5°F for Position 2.1; the weld metal = 28°F for Position 1.1 and, with credible surveillance data, = 14°F for Position 2.1. However, need not exceed 0.5*RTNDT.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-3 WCAP-17669-NP June 2013 Revision 0 Table 8.1-2 Calculation of the Catawba Unit 1 ART Values at the 3/4T Location for 34 EFPY RV Material and Identification Number R.G. 1.99, Rev. 2 Position CF(a)

(F) 3/4T Fluence(b)

(x1019 n/cm2, E > 1.0 MeV) 3/4T FF(b)

IRTNDT (c)

(F)

RTNDT (F)

I (c)

(°F)

(d)

(°F)

Margin (F)

ART (F)

US Forging 06 1.1 123.5 0.017 0.1580

-26 19.5 0

9.8 19.5 13 IS Forging 05 1.1 58 0.375 0.7286

-8 42.3 0

17.0 34.0 68 Using credible surveillance data 2.1 28.5 0.375 0.7286

-8 20.8 0

8.5 17.0 30 LS Forging 04 1.1 26 0.375 0.7286

-13 18.9 0

9.5 18.9 25 Bottom Head Ring 03 1.1 37 0.029 0.2162 14 8.0 0

4.0 8.0 30 US to IS Circ. Weld W06 (Heat # 899680) 1.1 41 0.017 0.1580 10 6.5 0

3.2 6.5 23 IS to LS Circ. Weld W05 (Heat # 895075) 1.1 54 0.375 0.7286

-51 39.3 0

19.7 39.3 28 Using credible surveillance data 2.1 28.5 0.375 0.7286

-51 20.8 0

10.4 20.8

-9 LS to Bottom Head Ring Weld W04 (Heat # 899680) 1.1 41 0.029 0.2162 10 8.9 0

4.4 8.9 28 Notes:

(a)

Taken from Table 5-2 of this report.

(b) 3/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2, and the Catawba Unit 1 RV beltline wall thickness of 8.465 inches.

(c)

Initial RTNDT values are measured and are taken from Table 3-1 of this report.

(d)

Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and, with credible surveillance data, = 8.5°F for Position 2.1; the weld metal = 28°F for Position 1.1 and, with credible surveillance data, = 14°F for Position 2.1. However, need not exceed 0.5*RTNDT.

8-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table 8.1-3 Calculation of the Catawba Unit 1 ART Values at the 1/4T Location for 54 EFPY RV Material and Identification Number R.G. 1.99, Rev. 2 Position CF(a)

(F) 1/4T Fluence(b)

(x1019 n/cm2, E > 1.0 MeV) 1/4T FF(b)

IRTNDT (c)

(F)

RTNDT (F)

I (c)

(°F)

(d)

(°F)

Margin (F)

ART (F)

US Forging 06 1.1 123.5 0.070 0.3488

-26 43.1 0

17.0 34.0 51 IS Forging 05 1.1 58 1.565 1.1237

-8 65.2 0

17.0 34.0 91 Using credible surveillance data 2.1 28.5 1.565 1.1237

-8 32.0 0

8.5 17.0 41 LS Forging 04 1.1 26 1.565 1.1237

-13 29.2 0

14.6 29.2 45 Bottom Head Ring 03 1.1 37 0.117 0.4496 14 16.6 0

8.3 16.6 47 US to IS Circ. Weld W06 (Heat # 899680) 1.1 41 0.070 0.3488 10 14.3 0

7.2 14.3 39 IS to LS Circ. Weld W05 (Heat # 895075) 1.1 54 1.565 1.1237

-51 60.7 0

28.0 56.0 66 Using credible surveillance data 2.1 28.5 1.565 1.1237

-51 32.0 0

14.0 28.0 9

LS to Bottom Head Ring Weld W04 (Heat # 899680) 1.1 41 0.117 0.4496 10 18.4 0

9.2 18.4 47 Notes:

(a)

Taken from Table 5-2 of this report.

(b) 1/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2, and the Catawba Unit 1 RV beltline wall thickness of 8.465 inches.

(c)

Initial RTNDT values are measured and are taken from Table 3-1 of this report.

(d)

Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and, with credible surveillance data, = 8.5°F for Position 2.1; the weld metal = 28°F for Position 1.1 and, with credible surveillance data, = 14°F for Position 2.1. However, need not exceed 0.5*RTNDT.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-5 WCAP-17669-NP June 2013 Revision 0 Table 8.1-4 Calculation of the Catawba Unit 1 ART Values at the 3/4T Location for 54 EFPY RV Material and Identification Number R.G. 1.99, Rev. 2 Position CF(a)

(F) 3/4T Fluence(b)

(x1019 n/cm2, E > 1.0 MeV) 3/4T FF(b)

IRTNDT (c)

(F)

RTNDT (F)

I (c)

(°F)

(d)

(°F)

Margin (F)

ART (F)

US Forging 06 1.1 123.5 0.025 0.1984

-26 24.5 0

12.3 24.5 23 IS Forging 05 1.1 58 0.567 0.8410

-8 48.8 0

17.0 34.0 75 Using credible surveillance data 2.1 28.5 0.567 0.8410

-8 24.0 0

8.5 17.0 33 LS Forging 04 1.1 26 0.567 0.8410

-13 21.9 0

10.9 21.9 31 Bottom Head Ring 03 1.1 37 0.042 0.2678 14 9.9 0

5.0 9.9 34 US to IS Circ. Weld W06 (Heat # 899680) 1.1 41 0.025 0.1984 10 8.1 0

4.1 8.1 26 IS to LS Circ. Weld W05 (Heat # 895075) 1.1 54 0.567 0.8410

-51 45.4 0

22.7 45.4 40 Using credible surveillance data 2.1 28.5 0.567 0.8410

-51 24.0 0

12.0 24.0

-3 LS to Bottom Head Ring Weld W04 (Heat # 899680) 1.1 41 0.042 0.2678 10 11.0 0

5.5 11.0 32 Notes:

(a)

Taken from Table 5-2 of this report.

(b) 3/4T fluence and FF were calculated using Regulatory Guide 1.99, Revision 2, and the Catawba Unit 1 RV beltline wall thickness of 8.465 inches.

(c)

Initial RTNDT values are measured and are taken from Table 3-1 of this report.

(d)

Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal = 17°F for Position 1.1 and, with credible surveillance data, = 8.5°F for Position 2.1; the weld metal = 28°F for Position 1.1 and, with credible surveillance data, = 14°F for Position 2.1. However, need not exceed 0.5*RTNDT.

8-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 8.2 P-T LIMIT CURVES APPLICABILITY EVALUATION Tables 8.1-1 through 8.1-4 summarize the 1/4T and 3/4T ART calculations for Catawba Unit 1 at 34 and 54 EFPY.

The applicability of the existing Catawba Unit 1 34 EFPY P-T limit curves, contained in WCAP-15203, Revision 1 (Reference 15) and 51 EFPY P-T limit curves, contained in WCAP-15448, Revision 1 (Reference 22) is evaluated by comparing the updated ART values contained in Section 8.1.1 with those used in the References 15 and 22 calculations. The existing P-T limit curves for Catawba Unit 1 are based on the limiting beltline material ART values, which are influenced by both fluence and initial material properties of that material. Using the MUR power uprate fluence projections, the 1/4T and 3/4T ART values were recalculated in Tables 8.1-1 through 8.1-4 as part of this applicability evaluation for Catawba Unit 1. Since the capsule fluence values were also revised as part of the MUR power uprate, the Position 2.1 chemistry factor values were updated in Section 5 of this report. The comparison of limiting ART values is contained in Table 8.2-1 for the existing Catawba Unit 1 34 EFPY P-T limit curves, and Table 8.2-2 for the existing 51 EFPY P-T limit curves.

Table 8.2-1 Summary of the Catawba Unit 1 Limiting ART Values used in the Applicability Evaluation of the Existing 34 EFPY RV Heatup and Cooldown Curves 1/4T Location 3/4T Location Existing 34 EFPY Curves documented in WCAP-15203, Revision 1 MUR Uprate Evaluation at 34 EFPY (Table 8.1-1)

Existing 34 EFPY Curves documented in WCAP-15203, Revision 1 MUR Uprate Evaluation at 34 EFPY (Table 8.1-2)

Limiting ART (°F) 42 43 31 30 Limiting Material LS Forging 04 US Forging 06 IS Forging 05 (using credible surveillance data)

IS Forging 05 (using credible surveillance data) & Bottom Head Ring 03 Table 8.2-2 Summary of the Catawba Unit 1 Limiting ART Values used in the Applicability Evaluation of the Existing 51 EFPY RV Heatup and Cooldown Curves 1/4T Location 3/4T Location Existing 51 EFPY Curves documented in WCAP-15448, Revision 1 MUR Uprate Evaluation at 54 EFPY (Table 8.1-3)

Existing 51 EFPY Curves documented in WCAP-15448, Revision 1 MUR Uprate Evaluation at 54 EFPY (Table 8.1-4)

Limiting ART (°F) 47 51 34 34 Limiting Material LS Forging 04 US Forging 06 IS Forging 05 (using credible surveillance data)

Bottom Head Ring 03

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-7 WCAP-17669-NP June 2013 Revision 0 34 EFPY P-T Limit Curves Table 8.2-1 compares the MUR power uprate limiting ART values at 34 EFPY to the limiting ART values used in development of the existing 34 EFPY P-T limit curves that are documented in WCAP-15203, Revision 1 (Reference 15). The limiting ART values used to develop the existing P-T limit curves are documented in Table 10 of Reference 15.

Based on the comparison of the ART values in Table 8.2-1, the limiting 1/4T ART value (42°F) used in the development of the current P-T limit curves at 34 EFPY is slightly less than the MUR power uprate limiting 1/4T ART value (43°F) at 34 EFPY for Catawba Unit 1. However, the limiting 3/4T ART value (30°F) calculated for the MUR evaluation is bounded by the limiting 3/4T value (31°F) used in the development of the current P-T limit curves, as shown in Table 8.2-1. Note that due to the consideration of all materials projected to achieve fluence levels of 1 x 1017 n/cm2 or higher at 34 EFPY, the limiting materials have changed since the existing 34 EFPY P-T limit curves were developed.

Using the MUR fluence values, the fluence in which the limiting material at the 1/4T location (Upper Shell Forging 06) would have an associated ART value of 42°F was calculated for Catawba Unit 1. This fluence was used to determine a revised EFPY for the current P-T limit curves by linearly interpolating the MUR uprate fluence projections in Table 2-2 of this report. With consideration of all RV materials that are projected to achieve surface fluence levels of 1 x 1017 n/cm2 or higher at 34 EFPY, the applicability date for which the current heatup and cooldown P-T limit curves were developed decreased from 34 EFPY to 30.7 EFPY with the MUR power uprate for Catawba Unit 1.

51 EFPY P-T Limit Curves Table 8.2-2 compares the MUR power uprate limiting ART values at 54 EFPY to the limiting ART values used in development of the existing 51 EFPY P-T limit curves that are documented in WCAP-15448, Revision 1 (Reference 22). The limiting ART values used to develop the existing P-T limit curves are documented in Table 10 of Reference 22.

Based on the comparison of the ART values in Table 8.2-2, the limiting 1/4T ART value (47°F) used in the development of the current P-T limit curves at 51 EFPY is less than the MUR power uprate limiting 1/4T ART value (51°F) at 54 EFPY for Catawba Unit 1. However, the limiting 3/4T ART value (34°F) calculated for the MUR evaluation is equal to the limiting 3/4T value (34°F) used in the development of the current P-T limit curves, as shown in Table 8.2-2. Note that due to the consideration of all materials projected to achieve fluence levels of 1 x 1017 n/cm2 or higher at 51 EFPY, the limiting materials have changed since the existing 51 EFPY P-T limit curves were developed.

Using the MUR fluence values, the fluence in which the limiting material at the 1/4T location (Upper Shell Forging 06) would have an associated ART value of 47°F was calculated for Catawba Unit 1. This fluence was used to determine a revised EFPY for the current P-T limit curves by linearly interpolating the MUR uprate fluence projections in Table 2-2 of this report. With consideration of all RV materials that are projected to achieve surface fluence levels of 1 x 1017 n/cm2 or higher at 51 EFPY, the applicability date for which the current heatup and cooldown P-T limit curves were developed decreased from 51 EFPY to 42.7 EFPY with the MUR power uprate for Catawba Unit 1.

8-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 P-T Limits Applicability Conclusion With consideration of all RV materials that are projected to achieve fluence levels of 1 x 1017 n/cm2 or higher, it is concluded that the MUR uprate evaluation does require a reduction of the existing Catawba Unit 1 P-T limit curves applicability dates. The revised applicability date of the 34 EFPY P-T limit curves decreases to 30.7 EFPY. Similarly, the revised applicability date of the 51 EFPY P-T limit curves decreases to 42.7 EFPY.

The new MUR uprate applicability dates are as follows:

Catawba Unit 1 - WCAP-15203, Revision 1 (Reference 15)

Current applicability: 34 EFPY Revised applicability using MUR uprate fluence: 30.7 EFPY Catawba Unit 1 - WCAP-15448, Revision 1 (Reference 22)

Current applicability: 51 EFPY Revised applicability using MUR uprate fluence: 42.7 EFPY

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 WCAP-17669-NP June 2013 Revision 0 9

SURVEILLANCE CAPSULE WITHDRAWAL

SUMMARY

Table 9-1 summarizes the removal of the six surveillance capsules from the Catawba Unit 1 RV, meeting the requirements of ASTM E185-82 (Reference 4), as required by 10 CFR 50, Appendix H (Reference 23).

Table 9-1 Catawba Unit 1 Surveillance Capsule Withdrawal Summary Capsule Capsule Location Lead Factor(a)

Withdrawal EFPY(b)

Fluence(a)

(x1019 n/cm2, E > 1.0 MeV)

Z 301.5° 3.85 0.79 0.292 Y

241° 3.73 4.98 1.31 V

61° 3.72 9.29 2.31 X(c) 238.5° 3.88 9.29 2.41 U(c) 58.5° 3.88 9.29 2.41 W(d) 121.5° 4.00 14.69 3.51 Notes:

(a)

Updated as part of the MUR power uprate fluence evaluation.

(b)

EFPY from plant startup. The fluence evaluation supporting this effort did not consider 0.11 EFPY of neutron exposure resulting from pre-commercial operation of Catawba Unit 1, between January 1985 and June 1985. The impact of this omission on the fluence evaluation results has been assessed to be negligible, and the results remain valid within the 20%

uncertainty criterion for fluence calculations.

(c)

Capsules X and U were removed from the RV at 9.29 EFPY and the dosimeters were tested. The material specimens were not tested and are being stored for potential future testing or further irradiation.

(d)

Capsule W was removed from the RV at 14.69 EFPY. This capsule was placed in the spent fuel pool following removal.

The removed and untested specimens may either be tested or re-inserted into the RV to be further irradiated.

Based on the limiting RTNDT value (55.2°F) for Catawba Unit 1 documented in Table 6-1 of this report, Catawba Unit 1 is required to withdraw three surveillance capsules, with the third capsule able to be held without testing following withdrawal. To date, three capsules (Z, Y, and V) were withdrawn and tested per ASTM E185-82. Two capsules (X and U) were withdrawn and the dosimeters were tested, but the material specimens were not tested and are being stored for potential future testing or further irradiation.

The last capsule (W) was withdrawn and placed in the spent fuel pool following removal.

The withdrawal schedule for these surveillance capsules meets the current recommendations of ASTM E185-82 as required by 10 CFR Part 50, Appendix H for a license extension through 60 years of operation, with consideration of the MUR power uprate. Furthermore, although Catawba Unit 1 does not have any capsules remaining in the RV, an EVND program is in place, meeting the requirements of Section XI.M31 of the GALL Report (Reference 11). If necessary in the future, the removed and untested specimens may either be tested or re-inserted into the RV to be further irradiated. Decisions regarding the stored specimens should be made when Duke Energy seeks to obtain a second 20-year extension to 80 years of plant operation.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 10-1 WCAP-17669-NP June 2013 Revision 0 10 REFERENCES

1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.

2. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

3. Code of Federal Regulations, 10 CFR Part 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

4. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials, 1982.

5. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

6. WCAP-16083-NP-A, Revision 0, Benchmark Testing of the FERRET Code for the Least Squares Evaluation of Light Water Reactor Dosimetry, May 2006.

7. WCAP-16083-NP, Revision 1, Benchmark Testing of the FERRET Code for the Least Squares Evaluation of Light Water Reactor Dosimetry, April 2013.

8. RSICC Computer Code Collection CCC-650, DOORS 3.2: One-, Two-, and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System, April 1998.

9. RSICC Data Library Collection DLC-185, BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, March 1996.

10. Branch Technical Position 5-3, Revision 2, Fracture Toughness Requirements, Chapter 5 of Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, March 2007.

11. NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, U.S. Nuclear Regulatory Commission, December 2010.

12. WCAP-17175-P, Revision 0, Catawba Units 1 and 2 Reactor Vessel Integrity Program Plans, January 2011.

13. Application to Renew the Operating Licenses of McGuire Nuclear Station, Units 1 & 2 and Catawba Nuclear Station, Units 1 & 2, June 2001. [NRC ADAMS Accession # ML011660145]

10-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0

14. Oak Ridge National Laboratory document ORNL/TM-2006/530, A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels, November 2007.

15. WCAP-15203, Revision 1, Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640, April 2001.

16. WCAP-15117, Revision 0, Analysis of Capsule V and the Dosimeters from Capsules U and X from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, October 1998.

17. WCAP-17455-NP, Revision 0, McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations, February 2012.

18. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

19. Code of Federal Regulations, 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010.

20. Code of Federal Regulations, 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

21. Appendix G to the 1995 through the 1996 Addendum Edition of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

22. WCAP-15448, Revision 1, Catawba Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Using Code Case N-640 for 51 EFPY, April 2001.

23. Code of Federal Regulations, 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 WCAP-17669-NP June 2013 Revision 0 APPENDIX A SURVEILLANCE DATA CREDIBILITY EVALUATION Introduction Regulatory Guide 1.99, Revision 2 (Reference A-1) describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the ART and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date there have been three surveillance capsules removed and tested from the Catawba Unit 1 reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.

The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Catawba Unit 1 RV surveillance data and determine if that surveillance data is credible.

Evaluation Criterion 1:

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, Fracture Toughness Requirements, (Reference A-2) as follows:

the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

The Catawba Unit 1 reactor vessel beltline region consists of the following materials:

1.

Intermediate Shell Forging 05

2.

Lower Shell Forging 04

3.

Intermediate Shell Forging to Lower Shell Forging Circumferential Weld Seam W05 (Weld Wire Heat # 895075, Flux Type Grau L.O. # LW320, Flux Lot # P46)

A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 The Catawba Unit 1 surveillance program utilizes axial and tangential test specimens from Intermediate Shell Forging 05. The surveillance weld metal was fabricated with weld wire Heat # 895075, Flux Type Grau L.O., Flux Lot # P46.

Intermediate Shell Forging 05 had the highest initial RTNDT and highest weight percent copper out of the two beltline forgings in the Catawba Unit 1 reactor vessel. Thus, it was selected as the surveillance base metal.

The weld material in the Catawba Unit 1 surveillance program was made of the same material as the reactor vessel beltline circumferential weld. In accordance with the definition of the reactor vessel beltline at that time, this was the only weld in the beltline region.

Therefore, the materials selected for use in the Catawba Unit 1 surveillance program were those judged to be most likely limiting with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed.

Based on the discussion, Criterion 1 is met for the Catawba Unit 1 surveillance program.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP-15117 (Reference A-3).

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Catawba Unit 1 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Catawba Unit 1 surveillance program.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 WCAP-17669-NP June 2013 Revision 0 Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 (Reference A-4).

The functional form of the least-squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these RTNDT values about this line is less than 28°F for the weld and less than 17°F for the forging.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A-5). At this meeting the NRC presented five cases. Of the five cases, Case 1 (Surveillance data available from plant but no other source) most closely represents the situation for the Catawba Unit 1 surveillance forging material. However, Catawba Unit 1 has a weld that will be evaluated for credibility using the guidance for the appropriate case as explained in Reference A-5. Weld Heat #

895075 pertains to IS to LS circumferential weld W05 in the Catawba Unit 1 reactor vessel. This weld heat is contained in the Catawba Unit 1 surveillance program as well as the McGuire Unit 2 and Watts Bar Unit 1 surveillance programs. NRC Case 4 per Reference A-5 is entitled Surveillance Data from Plant and Other Sources and most closely represents the situation for Catawba Unit 1 weld Heat #

895075.

A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Case 1: IS Forging 05 and Case 4: Weld Heat # 895075 (Catawba Unit 1 data only)

Following the NRC Case 4 guidelines, the Catawba Unit 1 surveillance weld metal (Heat # 895075) will be evaluated first using Catawba Unit 1 data only. The Catawba Unit 1 surveillance forging data will also be evaluated here since only Catawba Unit 1 data are considered. Table A-1 contains these evaluations.

Table A-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Catawba Unit 1 Surveillance Capsule Data Only Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV)

FF(b)

RTNDT (c)

(°F)

FF*RTNDT

(°F)

FF2 IS Forging 05 (Axial)

Z 0.292 0.663 15.74 10.44 0.440 Y

1.31 1.075 48.63 52.28 1.156 V

2.31 1.226 50.58 62.03 1.504 IS Forging 05 (Tangential)

Z 0.292 0.663 0.0(d) 0.00 0.440 Y

1.31 1.075 19.09 20.52 1.156 V

2.31 1.226 25.61 31.41 1.504 SUM:

176.68 6.199 CFIS Forging = (FF

RTNDT) ÷ (FF2) = (176.68) ÷ (6.199) = 28.5F Surveillance Weld Material (Heat # 895075)

Z 0.292 0.663 1.91 1.27 0.440 Y

1.31 1.075 17.79 19.13 1.156 V

2.31 1.226 26.5 32.50 1.504 SUM:

52.89 3.100 CFSurv. Weld = (FF

RTNDT) ÷ (FF2) = (52.89) ÷ (3.100) = 17.1F Notes:

(a) f = capsule fluence taken from Table 4-1 of this report.

(b)

FF = fluence factor = f(0.28 - 0.10*log f).

(c)

RTNDT values are the measured 30 ft-lb shift values taken from Table 4-1 of this report. These measured RTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Reg. Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Catawba Unit 1 data are being considered; therefore, no temperature adjustment is required.

(d)

Original value was negative, but physically a reduction should not occur; therefore, a value of zero will be used.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-5 WCAP-17669-NP June 2013 Revision 0 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A-2.

Table A-2 Best-Fit Evaluation for Catawba Unit 1 Surveillance Materials Only Material Capsule CF (Slopebest-fit)

(°F)

Capsule f (x1019 n/cm2, E > 1.0 MeV)

FF Measured RTNDT

(°F)

Predicted RTNDT (a)

(°F)

Scatter RTNDT (b)

(°F)

<17°F (Base Metal)

<28°F (Weld)

IS Forging 05 (Axial)

Z 28.5 0.292 0.663 15.74 18.9 3.2 Yes Y

28.5 1.31 1.075 48.63 30.6 18.0 No V

28.5 2.31 1.226 50.58 34.9 15.6 Yes IS Forging 05 (Tangential)

Z 28.5 0.292 0.663 0.0 18.9 18.9 No Y

28.5 1.31 1.075 19.09 30.6 11.6 Yes V

28.5 2.31 1.226 25.61 34.9 9.3 Yes Surveillance Weld Material (Heat # 895075)

Z 17.1 0.292 0.663 1.91 11.3 9.4 Yes Y

17.1 1.31 1.075 17.79 18.3 0.6 Yes V

17.1 2.31 1.226 26.5 20.9 5.6 Yes Notes:

(a)

Predicted RTNDT = CFbest-fit

FF.

(b)

Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 17°F for base metal. Table A-2 indicates that two of the six surveillance data points fall outside the +/- 1 of 17F scatter band for surveillance forging materials.

From a statistical point of view, +/- 1 would be expected to encompass 68% of the data. Since 66.7%

(two-thirds) of the forging data fall within the +/- 1 scatter band, it is concluded that this is approximately 68% and meets the intent of the requirement. Also, note that the two surveillance forging data points (Capsule Y axial and Capsule Z tangential) that fall outside the scatter band are only slightly outside the criteria by approximately 1°F or 2°F, respectively. The net effect of these deviations relative to the 1 bounds is not considered to be statistically significant at a typical level of confidence.

Therefore, based on engineering judgment, the IS Forging 05 surveillance data are deemed credible per the third criterion. Note that the data shown above in Table A-2 along with the credibility conclusions contained herein are consistent with the previous credibility evaluation contained Appendix D of Reference A-3.

The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A-2 indicates that all three surveillance data points fall within the +/- 1 of 28°F scatter band for surveillance weld materials; therefore, the weld metal (Heat # 895075) is deemed credible per the third criterion when only the Catawba Unit 1 data are considered.

A-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Case 4: Weld Heat # 895075 (All data)

In accordance with the NRC Case 4 guidelines, the data from Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 will now be analyzed together. Since the data are from multiple sources, the data are adjusted to the mean chemical composition and operating temperature of the surveillance capsules. This is performed in Table A-3 below.

Table A-3 Mean Chemical Composition and Operating Temperature for Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Material Capsule Cu Wt. %(a)

Ni Wt. %(a)

Plant Inlet Temperature(b)

(°F)

Weld Metal Heat # 895075 (Catawba Unit 1 data)

Z 0.05 0.73 553 Y

0.05 0.73 553 V

0.05 0.73 553 Weld Metal Heat # 895075 (McGuire Unit 2 data)

V 0.04 0.74 557 X

0.04 0.74 557 U

0.04 0.74 557 W

0.04 0.74 557 Weld Metal Heat # 895075 (Watts Bar Unit 1 data)

U 0.03 0.75 560 W

0.03 0.75 560 X

0.03 0.75 560 Z

0.03 0.75 560 MEAN 0.04 0.74 557.0 Notes:

(a) All copper and nickel weight percent values are documented in Table 3-1 of this report, except for the calculated mean values.

(b) All inlet temperature values are documented in Table 4-2 of this report, except for the calculated mean value.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 WCAP-17669-NP June 2013 Revision 0 Therefore, the Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 surveillance capsule data will be adjusted to the mean chemical composition and operating temperature calculated in Table A-3.

Catawba Unit 1 data CFMean

=

54°F (calculated per Table 1 of Regulatory Guide 1.99, Revision 2 using Cu Wt. % = 0.04 and Ni Wt. % = 0.74 per Table A-3)

CFSurv. Weld (Catawba Unit 1)

=

68°F (per Table 5-2 of this report)

Ratio = 54 68 = 0.79 (applied to Catawba Unit 1 surveillance data for weld Heat # 895075 in the credibility evaluation)

McGuire Unit 2 data CFMean

=

54°F CFSurv. Weld (McGuire Unit 2)

=

54°F (per Table 5-2 of this report)

Ratio = 54 54 = 1.00 (no ratio is applied to McGuire Unit 2 surveillance data for weld Heat # 895075 in the credibility evaluation since ratio = 1.00)

Watts Bar Unit 1 data CFMean

=

54°F CFSurv. Weld (Watts Bar Unit 1) =

41°F (per Table 5-2 of this report)

Ratio = 54 41 = 1.32 (applied to Watts Bar Unit 1 surveillance data for weld Heat # 895075 in the credibility evaluation)

The capsule-specific temperature adjustments are as shown in Table A-4 below:

Table A-4 Operating Temperature Adjustments for the Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Surveillance Capsule Data Material Plant Inlet Temperature

(°F)

Mean Operating Temperature (°F)

Temperature Adjustment (°F)

Weld Metal Heat # 895075 (Catawba Unit 1 data) 553 557.0

-4 Weld Metal Heat # 895075 (McGuire Unit 2 data) 557 557.0 0

Weld Metal Heat # 895075 (Watts Bar Unit 1 data) 560 557.0

+3

A-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Using the chemical composition and operating temperature adjustments described and calculated previously, an interim chemistry factor is calculated for weld Heat # 895075 using the Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 data. This calculation is shown in Table A-5.

Table A-5 Calculation of Weld Heat # 895075 Interim Chemistry Factor for the Credibility Evaluation Using Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Surveillance Capsule Data Material Capsule Capsule f(a)

(x1019 n/cm2, E > 1.0 MeV)

FF(b)

RTNDT (c)

(°F)

FF*RTNDT

(°F)

FF2 Catawba Unit 1 Surveillance Weld (Heat # 895075)

Z 0.292 0.663

-1.65 (1.91)

-1.10 0.440 Y

1.31 1.075 10.89 (17.79) 11.71 1.156 V

2.31 1.226 17.78 (26.5) 21.80 1.504 McGuire Unit 2 Surveillance Weld (Heat # 895075)

V 0.302 0.672 38.51 (38.51) 25.88 0.452 X

1.38 1.089 35.93 (35.93) 39.14 1.187 U

1.90 1.176 23.81 (23.81) 27.99 1.382 W

2.82 1.276 43.76 (43.76) 55.83 1.628 Watts Bar Unit 1 Surveillance Weld (Heat # 895075)

U 0.447 0.776 3.96 (0.0(d))

3.07 0.602 W

1.08 1.022 44.22 (30.5) 45.17 1.044 X

1.71 1.148 38.02 (25.8) 43.63 1.317 Z

2.40 1.236 22.31 (13.9) 27.57 1.528 SUM:

300.71 12.238 CFHeat # 895075 = (FF

RTNDT) ÷ (FF2) = (300.71) ÷ (12.238) = 24.6F Notes:

(a) f = capsule fluence taken from Tables 4-1 and 4-2 of this report.

(b)

FF = fluence factor = f(0.28 - 0.10*log f).

(c)

RTNDT values are the measured 30 ft-lb shift values. Pre-adjusted values are taken from Tables 4-1 and 4-2 of this report.

RTNDT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences between each surveillance welds chemistry and the mean surveillance weld chemistry for Heat # 895075 (pre-adjusted values are listed in parentheses). The temperature adjustments are shown in Table A-4 of this report. The ratios applied are 0.79 for Catawba Unit 1 and 1.32 for Watts Bar Unit 1. No ratio is applied to the McGuire Unit 2 data since the ratio is equal to 1.

(d)

This RTNDT value was determined to be negative, but physically a reduction should not occur; therefore, a value of zero is used.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9 WCAP-17669-NP June 2013 Revision 0 The scatter of RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A-6.

Table A-6 Best-Fit Evaluation for Surveillance Weld Metal Heat # 895075 Using Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 Data Material Capsule CF (Slopebest-fit)

(°F)

Capsule f (x1019 n/cm2, E > 1.0 MeV)

FF Measured RTNDT

(°F)

Predicted RTNDT (a)

(°F)

Scatter RTNDT (b)

(°F)

<28°F (Weld)

Catawba Unit 1 Surveillance Weld (Heat # 895075)

Z 24.6 0.292 0.663

-1.65 16.3 18.0 Yes Y

24.6 1.31 1.075 10.89 26.4 15.5 Yes V

24.6 2.31 1.226 17.78 30.2 12.4 Yes McGuire Unit 2 Surveillance Weld (Heat # 895075)

V 24.6 0.302 0.672 38.51 16.5 22.0 Yes X

24.6 1.38 1.089 35.93 26.8 9.2 Yes U

24.6 1.90 1.176 23.81 28.9 5.1 Yes W

24.6 2.82 1.276 43.76 31.4 12.4 Yes Watts Bar Unit 1 Surveillance Weld (Heat # 895075)

U 24.6 0.447 0.776 3.96 19.1 15.1 Yes W

24.6 1.08 1.022 44.22 25.1 19.1 Yes X

24.6 1.71 1.148 38.02 28.2 9.8 Yes Z

24.6 2.40 1.236 22.31 30.4 8.1 Yes Notes:

(a)

Predicted RTNDT = CFbest-fit

FF.

(b)

Scatter RTNDT = Absolute Value [Predicted RTNDT - Measured RTNDT].

The scatter of RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A-6 indicates that all eleven surveillance data points fall within the +/- 1 of 28°F scatter band for surveillance weld materials; therefore, the weld material (Heat # 895075) is deemed credible per the third criterion when all available data are considered.

In conclusion, the combined surveillance data from Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1 for weld Heat # 895075 may be applied to the Catawba Unit 1 reactor vessel weld. The chemistry factor calculation as applicable to the McGuire Unit 2 reactor vessel weld is contained in Table 5-1 of this report.

A-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.

The Catawba Unit 1 capsule specimens are located in the reactor between the neutron pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pads. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25F.

Hence, Criterion 4 is met for the Catawba Unit 1 surveillance program.

Criterion 5:

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The Catawba Unit 1 surveillance program does not contain correlation monitor material; therefore, this criterion is not applicable to the Catawba Unit 1 surveillance program.

Conclusion Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Catawba Unit 1 surveillance forging data are deemed credible and the surveillance weld data are deemed credible when considering Catawba Unit 1 data only and also when considering all available data (Catawba Unit 1, McGuire Unit 2, and Watts Bar Unit 1).

Appendix A References A-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

A-2 10 CFR 50, Appendix G, Fracture Toughness Requirements, Federal Register, Volume 60, No.

243, December 19, 1995.

A-3 WCAP-15117, Revision 0, Analysis of Capsule V and the Dosimeters from Capsules U and X from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, October 1998.

A-4 ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society of Testing and Materials, 1982.

A-5 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-1 WCAP-17669-NP June 2013 Revision 0 APPENDIX B EMERGENCY RESPONSE GUIDELINE LIMITS The Emergency Response Guideline (ERG) limits (Reference B-1) were developed in order to establish guidance for operator action in the event of an emergency situation, such as a PTS event. Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.

The highest RTNDT for which the generic category ERG limits were developed is 250F for a longitudinal flaw and 300F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250F for a longitudinal flaw or 300F for a circumferential flaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTPTS values are calculated in Section 6 of this report. The material with the highest RTNDT defines the limiting material, which for Catawba Unit 1 is US Forging 06. Table B-1 identifies ERG category limits and the limiting material RTNDT values at 54 EFPY for Catawba Unit 1.

Table B-1 Evaluation of Catawba Unit ERG Limit Category ERG Pressure-Temperature Limits (Reference B-1)

Applicable RTNDT Value(a)

ERG P-T Limit Category RTNDT < 200F Category I 200F < RTNDT < 250F Category II 250F < RTNDT < 300F Category III b Limiting RTNDT Values Reactor Vessel Material RTNDT Value @ 54 EFPY US Forging 06 63F(b)

Notes:

(a)

Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.

(b)

Value taken from Table 6-1 of this report.

B-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Emergency Response Guideline Limits Conclusion Per Table B-1, the limiting material for Catawba Unit 1 (Upper Shell Forging 06) has an RTNDT less than 200°F through 54 EFPY. Therefore, Catawba Unit 1 remains in ERG Category I through EOLE (54 EFPY).

Appendix B Reference B-1 HF04BG, Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 2, Westinghouse Owners Group, April 30, 2005.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1 WCAP-17669-NP June 2013 Revision 0 APPENDIX C VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS C.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to date at Catawba Unit 1 are described herein. Similarly, comparisons of measured EVND capsule results to both the calculated and least-squares adjusted values for EVND Capsules withdrawn from Catawba Unit 1 are also described in this section. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Reference C-1). One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within 20% for in-vessel neutron dosimetry sensors and within +/-30% for cavity (ex-vessel) neutron dosimetry sensors as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 2.0 of this report.

C.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the five neutron sensor sets analyzed to date as part of the Catawba Unit 1 Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:

Capsule ID Azimuthal Location Withdrawal Time Irradiation Time [EFPY]

Z 31.5° Single End of Cycle 1 0.79 Y

29.0° Dual End of Cycle 6 4.98 V

29.0° Dual End of Cycle 10 9.29 U

31.5° Dual End of Cycle 10 9.29 X

31.5° Dual End of Cycle 10 9.29 The passive neutron sensors included in the evaluations of Surveillance Capsules Z, Y, U, V, and X for Catawba Unit 1 are summarized as follows:

C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Sensor Material Reaction Of Interest Capsule Z

Capsule Y

Capsule V

Capsule U

Capsule X

Copper 63Cu(n,)60Co X

X X

Iron 54Fe(n,p)54Mn X

X X

Nickel 58Ni(n,p)58Co X

X X

Uranium-238 238U(n,f)137Cs X

X X

Neptunium-237 237Np(n,f)137Cs X

X X

Cobalt-Aluminum*

59Co(n,)60Co X

X X

Since all of the dosimetry monitors located at the radial center of the material test specimen array, radial gradient corrections were not required for these reaction rates. Pertinent physical and nuclear characteristics of the passive neutron sensors are listed in Table C-1.

The use of passive monitors such as those listed above does not yield a direct measure of the energy-dependent neutron flux at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:

1. the measured specific activity of each monitor,

2. the physical characteristics of each monitor,

3. the operating history of the reactor,

4. the energy response of each monitor, and

5. the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules Z, Y, V, U, and X are documented in References C-2 through C-4, respectively for Catawba Unit 1. Results from the radiometric counting of the neutron sensors from EVND Capsules A, B, C, D, E, and F are documented in Reference C-5. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules Z, Y, V, U, and X withdrawn from Catawba Unit 1 was based on the monthly power generation of Catawba Unit 1 from initial reactor commercial operation through the end of the dosimetry evaluation period. Please note the fluence evaluation supporting this effort did not consider 0.11 EFPY of neutron exposure resulting from pre-commercial operation of Catawba Unit 1, between January 1985 and June 1985. The impact of this

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-3 WCAP-17669-NP June 2013 Revision 0 omission on the fluence evaluation results has been assessed to be negligible, and the results remain valid within the +/- 20% uncertainty criterion for fluence calculations. Future fluence evaluations will consider the pre-commercial operation phase of Catawba Unit 1. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules Z, Y, V, U, and X withdrawn from Catawba Unit 1 is given in Table 6-7 of Reference C-4.

Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

]

e[ ]

e

[1 C

P P

Y F

N A

=

R j

t

j ref j

0 j

d, j

where:

R

=

Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus).

A

=

Measured specific activity (dps/g).

N0

=

Number of target element atoms per gram of sensor.

F

=

Atom fraction of the target isotope in the target element.

Y

=

Number of product atoms produced per reaction.

Pj

=

Average core power level during irradiation period j (MW).

Pref

=

Maximum or reference power level of the reactor (MW).

Cj

=

Calculated ratio of (E > 1.0 MeV) during irradiation period j to the time weighted average (E > 1.0 MeV) over the entire irradiation period.

=

Decay constant of the product isotope (1/sec).

tj

=

Length of irradiation period j (sec).

td,j

=

Decay time following irradiation period j (sec).

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

C-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 In the equation describing the reaction rate calculation, the ratio [Pj]/[Pref] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 2.0, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low-leakage to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.

The fuel cycle-specific neutron flux values along with the computed values for Cj are listed in Table C-2 for Catawba Unit 1. These flux values represent the cycle-dependent results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238U measurements to account for the presence of 235U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238U and 237Np sensor reaction rates to account for gamma ray induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Catawba Unit 1 fission sensor reaction rates are summarized as follows:

Correction Capsule Z Capsule Y Capsule V Capsule U Capsule X 235U Impurity/Pu Build-in 0.8728 0.8328 0.7991 0.7957 0.7957 238U(,f) 0.9673 0.9680 0.9681 0.9669 0.9669 Net 238U Correction 0.8443 0.8062 0.7736 0.7694 0.7694 237Np(,f) 0.9903 0.9903 0.9903 0.9900 0.9900 These factors were applied in a multiplicative fashion to the decay-corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules Z, Y, V, U, and X from Catawba Unit 1 are given in Table C-3. In Table C-3, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

Results of the sensor reaction rate determinations for EVND Capsules A, B, C, D, E and F from Catawba Unit 1 are given in Table C-4. In Table C-4, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238U impurities, plutonium build-in, and gamma ray induced fission effects.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-5 WCAP-17669-NP June 2013 Revision 0 C.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a BE neutron energy spectrum with associated uncertainties. BEs for key exposure parameters such as (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties.

For example,

)

)(

(

R g

ig i

g

g ig R

i

relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross section, ig, each with an uncertainty. The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the Catawba Unit 1 surveillance capsule dosimetry and EVND, the FERRET code (Reference C-6) was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine BE values of exposure parameters ((E > 1.0 MeV) and dpa) along with associated uncertainties for the capsules analyzed to date.

The application of the least-squares methodology requires the following input:

1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Catawba Unit 1 application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section C.1.1.

The dosimetry reaction cross sections and uncertainties were obtained from the SNLRML dosimetry cross-section library (Reference C-7). The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E1018, Application of ASTM Evaluated Cross-Section Data File, Matrix E 706 (IIB).

C-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 The uncertainties associated with the measured reaction rates, dosimetry cross sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E 944, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance.

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Catawba Unit 1 surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation:

Reaction Uncertainty 46Ti(n,p)46Sc 5%

63Cu(n,)60Co 5%

54Fe(n,p)54Mn 5%

58Ni(n,p)58Co 5%

238U(n,f)137Cs 10%

237Np(n,f)137Cs 10%

59Co(n,)60Co 5%

These uncertainties are given at the 1 level.

Dosimetry Cross-Section Uncertainties The reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-7 WCAP-17669-NP June 2013 Revision 0 tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the Catawba Unit 1 surveillance programs, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.

Reaction Uncertainty 46Ti(n,p)46Sc 4.51-4.87%

63Cu(n,)60Co 4.08-4.16%

54Fe(n,p)54Mn 3.05-3.11%

58Ni(n,p)58Co 4.49-4.56%

238U(n,f)137Cs 0.54-0.64%

237Np(n,f)137Cs 10.32-10.97%

59Co(n,)60Co 0.79-3.59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutron spectra input to the least-squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

gg' g'

g 2

n gg' P

R R

R M

where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

e

+

]

[1

=

P

-H g

g g

g

where

C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 2

2 2

)

g' (g

H

The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term). The value of is 1.0 when g = g, and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Catawba Unit 1 calculated spectra was as follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 C.1.3 Comparisons of Measurements and Calculations Results of the least-squares evaluations of the dosimetry from the Catawba Unit 1 surveillance capsules and EVND capsules withdrawn to date are provided in Tables C-5 through C-7, C-10, and C-11. In Tables C-5 and C-6, measured, calculated, and BE values for sensor reaction rates are given for each surveillance and EVND capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least-squares adjusted reaction rates. These ratios of measured/calculated (M/C) and measured/best estimate (M/BE) illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Tables C-7, C-10 and C-11, comparison of the calculated and BE values of neutron flux (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the best-estimate/calculated (BE/C) ratios observed for each of the capsules.

The data comparisons provided in Tables C-5 through C-7, C-10, and C-11 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-9 WCAP-17669-NP June 2013 Revision 0 spectra, measured sensor reaction rates, and dosimetry reaction cross sections. Further, these results indicate that the use of the least-squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 2.0 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displacements at the surveillance capsule locations is specified as 12% at the 1 level. From Tables C-7, C-10, and C-11, it is noted that the corresponding uncertainties associated with the least-squares adjusted exposure parameters have been reduced to 6% for neutron flux (E > 1.0 MeV) and 8-9% for iron atom displacement rate. Again, the uncertainties from the least-squares evaluation are at the 1 level.

Further comparisons of the measurement results (from Tables C-5 through C-7, C-10, and C-11) with calculations are given in Tables C-8 and C-9 for the surveillance capsules and Tables C-12 through C-15 for EVND capsules. These comparisons are given on two levels. In Tables C-8, C-12 and C-13, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Tables C-9, C-14 and C-15, calculations of fast neutron exposure rates in terms of (E > 1.0 MeV) and dpa/s are compared with the BE results obtained from the least-squares evaluation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to-calculation comparisons falling well within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the M/C comparisons for fast neutron reactions range from 0.83 to 1.25 for the 25 samples included in the surveillance capsules data set. The overall average M/C ratio for the entire set of Catawba Unit 1 surveillance capsules data is 1.00 with an associated standard deviation of 12.8%. Similarly for Catawba Unit 1 EVND capsules at core midplane, the M/C comparisons for fast neutron reactions range from 0.81 to 1.10 for the 24 samples included in the data set. The overall average M/C ratio for the entire set of Catawba Unit 1 EVND capsules at core midplane data is 0.93 with an associated standard deviation of 8.1%. For Catawba Unit 1 EVND capsules off core midplane, the M/C comparisons for fast neutron reactions range from 0.69 to 1.29 for the 12 samples included in the data set. The overall average M/C ratio for the entire set of Catawba Unit 1 EVND capsules off core midplane data is 0.88 with an associated standard deviation of 19.0%.

In the comparisons of BE and calculated fast neutron exposure parameters for Catawba Unit 1 surveillance capsules, the corresponding BE/C comparisons for the capsule data sets range from 0.90 to 1.04 for neutron flux (E > 1.0 MeV) and from 0.92 to 1.06 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.96 with a standard deviation of 6.1% and 0.98 with a standard deviation of 6.4%, respectively. Similarly, for Catawba Unit 1 EVND capsules at core midplane, the corresponding BE/C comparisons for the capsule data sets range from 0.90 to 0.98 for neutron flux (E > 1.0 MeV) and from 0.92 to 0.98 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.93 with a standard deviation of 3.9% and 0.95 with a standard deviation of 3.2%,

respectively. For Catawba Unit 1 EVND capsules off core midplane, the corresponding BE/C comparisons for the capsule data sets range from 0.79 to 1.02 for neutron flux (E > 1.0 MeV) and from

C-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 0.81 to 1.03 for iron atom displacement rate. The overall average BE/C ratios for neutron flux (E > 1.0 MeV) and iron atom displacement rate are 0.91 with a standard deviation of 18.0% and 0.92 with a standard deviation of 16.9%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 2.0 of this report are validated for use in the assessment of the condition of the materials comprising the beltline region and extended beltline region of the Catawba Unit 1 reactor pressure vessel.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-11 WCAP-17669-NP June 2013 Revision 0 Table C-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors Monitor Material Reaction of Interest Target Atom Fraction 90% Response Range (MeV)

Product Half-life Fission Yield

(%)

Titanium 46Ti (n,p) 0.0825 3.70 - 9.43 83.79 d Copper 63Cu (n,)

0.6917 4.53-11.0 5.271 y Iron 54Fe (n,p) 0.0585 2.27 - 7.54 312.1 d Nickel 58Ni (n,p) 0.6808 1.98 - 7.51 70.82 d Uranium-238 238U (n,f) 1.0000 1.44 - 6.69 30.07 y 6.02 Neptunium-237 237Np (n,f) 1.0000 0.68 - 5.61 30.07 y 6.17 Cobalt-Aluminum 59Co (n,)

0.0015 non-threshold 5.271 y Note:

The 90% response range is defined such that, in the neutron spectrum characteristic of the Catawba Unit 1 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.

C-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-2 Calculated Fast Neutron Flux (E > 1.0 MeV) at Catawba Unit 1 Surveillance Capsule Center Core Midplane Elevation Fuel Cycle Cycle Length

[EFPS]

(E > 1.0 MeV) [n/cm2-s]

Capsule Z Capsule Y Capsule V Capsule U Capsule X 1

2.49E+07 1.17E+11 1.09E+11 1.09E+11 1.18E+11 1.18E+11 2

2.37E+07 7.68E+10 7.68E+10 8.27E+10 8.27E+10 3

2.43E+07 8.61E+10 8.61E+10 8.97E+10 8.97E+10 4

2.71E+07 7.75E+10 7.75E+10 7.99E+10 7.99E+10 5

2.49E+07 7.61E+10 7.61E+10 8.10E+10 8.10E+10 6

3.22E+07 7.83E+10 7.83E+10 7.81E+10 7.81E+10 7

3.00E+07 7.83E+10 8.25E+10 8.25E+10 8

3.38E+07 7.45E+10 7.42E+10 7.42E+10 9

3.69E+07 6.87E+10 7.03E+10 7.03E+10 10 3.53E+07 7.16E+10 7.74E+10 7.74E+10 Average 1.17E+11 8.37E+10 7.88E+10 8.23E+10 8.23E+10

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-13 WCAP-17669-NP June 2013 Revision 0 Table C-2 Calculated Cj Factors at Catawba Unit 1 Surveillance Capsule Center Core Midplane Elevation (cont.)

Fuel Cycle Cycle Length

[EFPS]

Cj Capsule Z Capsule Y Capsule V Capsule U Capsule X 1

2.49E+07 1.00 1.30 1.38 1.43 1.43 2

2.37E+07 0.92 0.98 1.01 1.01 3

2.43E+07 1.03 1.09 1.09 1.09 4

2.71E+07 0.93 0.98 0.97 0.97 5

2.49E+07 0.91 0.97 0.98 0.98 6

3.22E+07 0.94 0.99 0.95 0.95 7

3.00E+07 0.99 1.00 1.00 8

3.38E+07 0.95 0.90 0.90 9

3.69E+07 0.87 0.85 0.85 10 3.53E+07 0.91 0.94 0.94 Average 1.00 1.00 1.00 1.00 1.00

C-14 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule Z Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 63Cu (n,) 60Co Top 4.39E+04 4.84E+05 7.38E-17 Middle 4.46E+04 4.92E+05 7.50E-17 Bottom 4.81E+04 5.30E+05 8.09E-17 Average 7.66E-17 54Fe (n,p) 54Mn Top 1.21E+06 4.34E+06 6.89E-15 Bottom 1.32E+06 4.74E+06 7.51E-15 Average 7.20E-15 58Ni (n,p) 58Co Top 6.17E+06 5.68E+07 8.13E-15 Middle 7.34E+06 6.75E+07 9.67E-15 Bottom 7.02E+06 6.46E+07 9.25E-15 Average 9.01E-15 238U (n,f) 137Cs (Cd)

Middle 1.44E+05 8.07E+06 5.30E-14 Including 235U, 239Pu, and fission corrections:

4.47E-14 237Np (n,f) 137Cs (Cd)

Middle 1.20E+06 6.72E+07 4.29E-13 Including fission corrections:

4.25E-13 59Co (n,) 60Co Top 7.33E+06 8.08E+07 5.27E-12 Middle 8.51E+06 9.38E+07 6.12E-12 Bottom 8.99E+06 9.91E+07 6.47E-12 N/A*

8.67E+06 9.56E+07 6.24E-12 Average 6.03E-12 59Co (n,) 60Co (Cd)

Top 4.35E+06 4.80E+07 3.13E-12 Middle 3.08E+06 3.40E+07 2.22E-12 Bottom 4.29E+06 4.73E+07 3.09E-12 Average 2.81E-12 Notes:

1. Measured specific activities are indexed to a counting date of February 2, 1987.

2. The average 238U (n,f) reaction rate of 4.47E-14 includes a correction factor of 0.8728 to account for plutonium build-in and an additional factor of 0.9673 to account for photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rate of 4.25E-13 includes a correction factor of 0.9903 to account for photo-fission effects in the sensor.

The location of the dosimeter sensor cannot be determined from reference document.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-15 WCAP-17669-NP June 2013 Revision 0 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule Y (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 63Cu (n,) 60Co Top 1.41E+05 3.47E+05 5.29E-17 Middle 1.29E+05 3.17E+05 4.84E-17 Bottom 1.28E+05 3.15E+05 4.80E-17 Average 4.98E-17 54Fe (n,p) 54Mn Top 1.67E+06 3.00E+06 4.75E-15 Bottom 1.55E+06 2.78E+06 4.41E-15 Average 4.58E-15 58Ni (n,p) 58Co Top 1.12E+07 4.98E+07 7.13E-15 Middle 1.04E+07 4.62E+07 6.62E-15 Bottom 1.02E+07 4.53E+07 6.49E-15 Average 6.74E-15 238U (n,f) 137Cs (Cd)

Middle 5.34E+05 5.09E+06 3.34E-14 Including 235U, 239Pu, and fission corrections:

2.69E-14 237Np (n,f) 137Cs (Cd)

Middle 4.27E+06 4.07E+07 2.60E-13 Including fission corrections:

2.57E-13 59Co (n,) 60Co Top 2.07E+07 5.09E+07 3.32E-12 Top 2.42E+07 5.95E+07 3.88E-12 Middle 2.06E+07 5.07E+07 3.31E-12 Middle 2.39E+07 5.88E+07 3.84E-12 Bottom 2.33E+07 5.73E+07 3.74E-12 Bottom 2.03E+07 4.99E+07 3.26E-12 Average 3.56E-12 59Co (n,) 60Co (Cd)

Top 1.28E+07 3.15E+07 2.05E-12 Middle 1.29E+07 3.17E+07 2.07E-12 Bottom 1.29E+07 3.17E+07 2.07E-12 Average 2.07E-12 Notes:

1. Measured specific activities are indexed to a counting date of November 28, 1992.

2. The average 238U (n,f) reaction rate of 2.69E-14 includes a correction factor of 0.8328 to account for plutonium build-in and an additional factor of 0.9680 to account for photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rate of 2.57E-13 includes a correction factor of 0.9903 to account for photo-fission effects in the sensor.

C-16 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule V (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 63Cu (n,) 60Co Top 1.67E+05 2.98E+05 4.55E-17 Middle 1.52E+05 2.72E+05 4.14E-17 Bottom 1.49E+05 2.66E+05 4.06E-17 Average 4.25E-17 54Fe (n,p) 54Mn Top 1.36E+06 2.78E+06 4.41E-15 Bottom 1.21E+06 2.48E+06 3.92E-15 Average 4.17E-15 58Ni (n,p) 58Co Top 4.56E+06 4.19E+07 6.00E-15 Middle 4.17E+06 3.83E+07 5.49E-15 Bottom 4.10E+06 3.77E+07 5.39E-15 Average 5.63E-15 238U (n,f) 137Cs (Cd)

Middle 9.84E+05 5.35E+06 3.52E-14 Including 235U, 239Pu, and fission corrections:

2.72E-14 237Np (n,f) 137Cs (Cd)

Middle 6.43E+06 3.50E+07 2.23E-13 Including fission corrections:

2.21E-13 59Co (n,) 60Co Top 2.89E+07 5.16E+07 3.37E-12 Top 2.75E+07 4.91E+07 3.21E-12 Middle 2.42E+07 4.32E+07 2.82E-12 Middle 2.85E+07 5.09E+07 3.32E-12 Bottom 2.46E+07 4.39E+07 2.87E-12 Bottom 2.82E+07 5.04E+07 3.29E-12 Average 3.14E-12 59Co (n,) 60Co (Cd)

Top 1.52E+07 2.72E+07 1.77E-12 Middle 1.40E+07 2.50E+07 1.63E-12 Bottom 1.44E+07 2.57E+07 1.68E-12 Average 1.69E-12 Notes:

1. Measured specific activities are indexed to a counting date of July 1, 1998.

2. The average 238U (n,f) reaction rate of 2.72E-14 includes a correction factor of 0.7991 to account for plutonium build-in and an additional factor of 0.9681 to account for photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rate of 2.21E-13 includes a correction factor of 0.9903 to account for photo-fission effects in the sensor.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-17 WCAP-17669-NP June 2013 Revision 0 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule U (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 63Cu (n,) 60Co Top 1.78E+05 3.19E+05 4.87E-17 Middle 1.60E+05 2.87E+05 4.37E-17 Bottom 1.56E+05 2.80E+05 4.26E-17 Average 4.50E-17 54Fe (n,p) 54Mn Top 1.44E+06 2.90E+06 4.60E-15 Middle 1.33E+06 2.68E+06 4.25E-15 Bottom 1.30E+06 2.62E+06 4.15E-15 Average 4.33E-15 58Ni (n,p) 58Co Top 4.95E+06 4.40E+07 6.30E-15 Middle 4.60E+06 4.09E+07 5.85E-15 Bottom 4.43E+06 3.94E+07 5.63E-15 Average 5.93E-15 238U (n,f) 137Cs (Cd)

Middle 1.05E+06 5.72E+06 3.76E-14 Including 235U, 239Pu, and fission corrections:

2.89E-14 237Np (n,f) 137Cs (Cd)

Middle 6.51E+06 3.55E+07 2.26E-13 Including fission corrections:

2.24E-13 59Co (n,) 60Co Top 3.05E+07 5.47E+07 3.57E-12 Middle 2.91E+07 5.21E+07 3.40E-12 Middle 2.54E+07 4.55E+07 2.97E-12 Bottom 2.66E+07 4.77E+07 3.11E-12 Bottom 2.31E+07 4.14E+07 2.70E-12 Average 3.15E-12 59Co (n,) 60Co (Cd)

Top 1.58E+07 2.83E+07 1.85E-12 Middle 1.56E+07 2.80E+07 1.82E-12 Bottom 1.43E+07 2.56E+07 1.67E-12 Average 1.78E-12 Notes:

1. Measured specific activities are indexed to a counting date of July 1, 1998.

2. The average 238U (n,f) reaction rate of 2.89E-14 includes a correction factor of 0.7957 to account for plutonium build-in and an additional factor of 0.9669 to account for photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rate of 2.24E-13 includes a correction factor of 0.9900 to account for photo-fission effects in the sensor.

C-18 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-3 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 Surveillance Capsule X (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 63Cu (n,) 60Co Top 1.74E+05 3.12E+05 4.76E-17 Middle 1.65E+05 2.96E+05 4.51E-17 Bottom 1.62E+05 2.90E+05 4.43E-17 Average 4.57E-17 54Fe (n,p) 54Mn Top 1.43E+06 2.88E+06 4.57E-15 Middle 1.33E+06 2.68E+06 4.25E-15 Bottom 1.33E+06 2.68E+06 4.25E-15 Average 4.35E-15 58Ni (n,p) 58Co Top 5.04E+06 4.48E+07 6.41E-15 Middle 4.68E+06 4.16E+07 5.95E-15 Bottom 4.61E+06 4.10E+07 5.86E-15 Average 6.08E-15 238U (n,f) 137Cs (Cd)

Middle 1.16E+06 6.32E+06 4.15E-14 Including 235U, 239Pu, and fission corrections:

3.19E-14 237Np (n,f) 137Cs (Cd)

Middle 8.56E+06 4.66E+07 2.97E-13 Including fission corrections:

2.94E-13 59Co (n,) 60Co Top 3.12E+07 5.59E+07 3.65E-12 Top 2.94E+07 5.27E+07 3.44E-12 Middle 2.63E+07 4.71E+07 3.07E-12 Middle 3.15E+07 5.64E+07 3.68E-12 Bottom 3.08E+07 5.52E+07 3.60E-12 Bottom 2.76E+07 4.95E+07 3.23E-12 Average 3.44E-12 59Co (n,) 60Co (Cd)

Top 1.58E+07 2.83E+07 1.85E-12 Middle 1.57E+07 2.81E+07 1.84E-12 Bottom 1.58E+07 2.83E+07 1.85E-12 Average 1.84E-12 Notes:

1. Measured specific activities are indexed to a counting date of July 1, 1998.

2. The average 238U (n,f) reaction rate of 3.19E-14 includes a correction factor of 0.7957 to account for plutonium build-in and an additional factor of 0.9669 to account for photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rate of 2.94E-13 includes a correction factor of 0.9900 to account for photo-fission effects in the sensor.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-19 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule A Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Middle 5.48E+02 5.07E+03 4.88E-18 Average 4.88E-18 54Fe (n,p) 54Mn Middle 7.82E+03 1.69E+04 2.69E-17 Middle 7.94E+03 1.72E+04 2.73E-17 Average 2.71E-17 58Ni (n,p) 58Co Middle 1.87E+04 2.53E+05 3.63E-17 Average 3.63E-17 63Cu (n,) 60Co Middle 6.32E+02 2.39E+03 3.64E-19 Average 3.64E-19 238U (n,f) 95Zr (Cd)

Middle 8.41E+02 1.55E+04 1.36E-16 Including fission corrections:

1.30E-16 238U (n,f) 103Ru (Cd)

Middle 1.60E+02 1.70E+04 1.22E-16 Including fission corrections:

1.17E-16 238U (n,f) 137Cs (Cd)

Middle 9.74E+02 1.76E+04 1.31E-16 Including fission corrections:

1.26E-16 237Np (n,f) 95Zr (Cd)

Middle 1.34E+04 2.47E+05 1.96E-15 Including fission corrections:

1.94E-15 237Np (n,f) 103Ru (Cd)

Middle 2.35E+03 2.49E+05 2.02E-15 Including fission corrections:

2.00E-15 237Np (n,f) 137Cs (Cd)

Middle 1.51E+04 2.72E+05 1.99E-15 Including fission corrections:

1.97E-15 59Co (n,) 60Co Middle 5.22E+05 1.97E+06 3.99E-14 Average 3.99E-14 59Co (n,) 60Co (Cd)

Middle 2.45E+05 9.25E+05 1.87E-14 Average 1.87E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

C-20 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule B (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Middle 7.89E+02 7.19E+03 6.93E-18 Average 6.93E-18 54Fe (n,p) 54Mn Middle 1.16E+04 2.49E+04 3.95E-17 Middle 1.15E+04 2.47E+04 3.92E-17 Average 3.93E-17 58Ni (n,p) 58Co Middle 2.78E+04 3.71E+05 5.31E-17 Average 5.31E-17 63Cu (n,) 60Co Middle 8.30E+02 3.13E+03 4.77E-19 Average 4.77E-19 238U (n,f) 95Zr (Cd)

Middle 1.26E+03 2.29E+04 2.00E-16 Including fission corrections:

1.93E-16 238U (n,f) 103Ru (Cd)

Middle 2.70E+02 2.82E+04 2.02E-16 Including fission corrections:

1.95E-16 238U (n,f) 137Cs (Cd)

Middle 1.50E+03 2.71E+04 2.02E-16 Including fission corrections:

1.95E-16 237Np (n,f) 95Zr (Cd)

Middle 2.11E+04 3.84E+05 3.04E-15 Including fission corrections:

3.02E-15 237Np (n,f) 103Ru (Cd)

Middle 4.07E+03 4.25E+05 3.44E-15 Including fission corrections:

3.42E-15 237Np (n,f) 137Cs (Cd)

Middle 2.35E+04 4.24E+05 3.10E-15 Including fission corrections:

3.08E-15 59Co (n,) 60Co Middle 7.60E+05 2.86E+06 5.80E-14 Average 5.80E-14 59Co (n,) 60Co (Cd)

Middle 3.92E+05 1.48E+06 2.99E-14 Average 2.99E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-21 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule C (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Middle 8.37E+02 7.27E+03 7.00E-18 Average 7.00E-18 54Fe (n,p) 54Mn Middle 1.29E+04 2.70E+04 4.28E-17 Middle 1.25E+04 2.62E+04 4.15E-17 Average 4.21E-17 58Ni (n,p) 58Co Middle 2.90E+04 3.69E+05 5.28E-17 Average 5.28E-17 63Cu (n,) 60Co Middle 8.71E+02 3.27E+03 4.98E-19 Average 4.98E-19 238U (n,f) 95Zr (Cd)

Middle 1.49E+03 2.58E+04 2.25E-16 Including fission corrections:

2.18E-16 238U (n,f) 103Ru (Cd)

Middle 3.01E+02 2.99E+04 2.15E-16 Including fission corrections:

2.08E-16 238U (n,f) 137Cs (Cd)

Middle 1.69E+03 3.05E+04 2.27E-16 Including fission corrections:

2.20E-16 237Np (n,f) 95Zr (Cd)

Middle 2.59E+04 4.48E+05 3.55E-15 Including fission corrections:

3.53E-15 237Np (n,f) 103Ru (Cd)

Middle 4.83E+03 4.80E+05 3.89E-15 Including fission corrections:

3.86E-15 237Np (n,f) 137Cs (Cd)

Middle 2.72E+04 4.90E+05 3.58E-15 Including fission corrections:

3.56E-15 59Co (n,) 60Co Middle 9.12E+05 3.42E+06 6.93E-14 Average 6.93E-14 59Co (n,) 60Co (Cd)

Middle 4.70E+05 1.76E+06 3.57E-14 Average 3.57E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

C-22 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule D (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Top 2.50E+02 2.13E+03 2.05E-18 Average 2.05E-18 54Fe (n,p) 54Mn Top 3.54E+03 7.33E+03 1.16E-17 Top 3.74E+03 7.74E+03 1.23E-17 Average 1.20E-17 58Ni (n,p) 58Co Top 9.58E+03 1.19E+05 1.71E-17 Average 1.71E-17 63Cu (n,) 60Co Top 2.24E+02 8.39E+02 1.28E-19 Average 1.28E-19 238U (n,f) 95Zr (Cd)

Top 6.01E+02 1.02E+04 8.90E-17 Including fission corrections:

8.58E-17 238U (n,f) 103Ru (Cd)

Top 1.18E+02 1.15E+04 8.25E-17 Including fission corrections:

7.95E-17 238U (n,f) 137Cs (Cd)

Top 5.71E+02 1.03E+04 7.68E-17 Including fission corrections:

7.40E-17 237Np (n,f) 95Zr (Cd)

Top 1.13E+04 1.92E+05 1.52E-15 Including fission corrections:

1.51E-15 237Np (n,f) 103Ru (Cd)

Top 2.15E+03 2.09E+05 1.70E-15 Including fission corrections:

1.69E-15 237Np (n,f) 137Cs (Cd)

Top 1.08E+04 1.95E+05 1.42E-15 Including fission corrections:

1.41E-15 59Co (n,) 60Co Top 2.43E+05 9.10E+05 1.84E-14 Average 1.84E-14 59Co (n,) 60Co (Cd)

Top 1.57E+05 5.88E+05 1.19E-14 Average 1.19E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-23 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule E (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Middle 5.99E+02 4.98E+03 4.80E-18 Average 4.80E-18 54Fe (n,p) 54Mn Middle 9.23E+03 1.89E+04 2.99E-17 Middle 9.29E+03 1.90E+04 3.01E-17 Average 3.00E-17 58Ni (n,p) 58Co Middle 2.21E+04 2.69E+05 3.85E-17 Average 3.85E-17 63Cu (n,) 60Co Middle 5.84E+02 2.18E+03 3.33E-19 Average 3.33E-19 238U (n,f) 95Zr (Cd)

Middle 1.27E+03 2.10E+04 1.84E-16 Including fission corrections:

1.78E-16 238U (n,f) 103Ru (Cd)

Middle 2.59E+02 2.46E+04 1.77E-16 Including fission corrections:

1.71E-16 238U (n,f) 137Cs (Cd)

Middle 1.32E+03 2.38E+04 1.78E-16 Including fission corrections:

1.72E-16 237Np (n,f) 95Zr (Cd)

Middle 2.48E+04 4.11E+05 3.25E-15 Including fission corrections:

3.23E-15 237Np (n,f) 103Ru (Cd)

Middle 4.18E+03 3.97E+05 3.22E-15 Including fission corrections:

3.20E-15 237Np (n,f) 137Cs (Cd)

Middle 2.45E+04 4.41E+05 3.22E-15 Including fission corrections:

3.21E-15 59Co (n,) 60Co Middle 6.17E+05 2.30E+06 4.67E-14 Average 4.67E-14 59Co (n,) 60Co (Cd)

Middle 3.62E+05 1.35E+06 2.74E-14 Average 2.74E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

C-24 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-4 Measured Sensor Activities and Reaction Rates for Catawba Unit 1 EVND Capsule F (cont.)

Reaction Location Measured Activity (dps/g)

Saturated Activity (dps/g)

Adjusted Reaction Rate (rps/atom) 46Ti (n,p) 46Sc Bottom 2.01E+02 1.69E+03 1.63E-18 Average 1.63E-18 54Fe (n,p) 54Mn Bottom 3.06E+03 6.29E+03 9.97E-18 Bottom 2.88E+03 5.92E+03 9.38E-18 Average 9.68E-18 58Ni (n,p) 58Co Bottom 8.07E+03 9.92E+04 1.42E-17 Average 1.42E-17 63Cu (n,) 60Co Bottom 1.83E+02 6.84E+02 1.04E-19 Average 1.04E-19 238U (n,f) 95Zr (Cd)

Bottom 4.05E+02 6.78E+03 5.92E-17 Including fission corrections:

5.71E-17 238U (n,f) 103Ru (Cd)

Bottom 8.56E+01 8.22E+03 5.90E-17 Including fission corrections:

5.70E-17 238U (n,f) 137Cs (Cd)

Bottom 3.86E+02 6.95E+03 5.19E-17 Including fission corrections:

5.01E-17 237Np (n,f) 95Zr (Cd)

Bottom 8.14E+03 1.36E+05 1.08E-15 Including fission corrections:

1.07E-15 237Np (n,f) 103Ru (Cd)

Bottom 1.36E+03 1.31E+05 1.06E-15 Including fission corrections:

1.05E-15 237Np (n,f) 137Cs (Cd)

Bottom 7.11E+03 1.28E+05 9.36E-16 Including fission corrections:

9.30E-16 59Co (n,) 60Co Bottom 3.14E+05 1.17E+06 2.38E-14 Average 2.38E-14 59Co (n,) 60Co (Cd)

Bottom 1.65E+05 6.17E+05 1.25E-14 Average 1.25E-14 Notes:

1. Measured specific activities are indexed to a counting date of July 24, 2007.

2. The average 238U (n,f) reaction rates include correction factors to account for plutonium build-in and photo-fission effects in the sensor.

3. The average 237Np (n,f) reaction rates include correction factors to account for photo-fission effects in the sensor.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-25 WCAP-17669-NP June 2013 Revision 0 Table C-5 Comparison of M/C and BE Reaction Rates at the Surveillance Capsule Center Catawba Unit 1 Capsule Z Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 63Cu(n,)60Co 7.66E-17 6.11E-17 7.16E-17 1.25 1.08 54Fe(n,p)54Mn 7.20E-15 6.93E-15 7.33E-15 1.04 0.98 58Ni(n,p)58Co 9.01E-15 9.71E-15 9.93E-15 0.93 0.91 59Co(n,)60Co 6.02E-12 5.16E-12 5.94E-12 1.17 1.01 59Co(n,)60Co (Cd) 2.81E-12 3.53E-12 2.86E-12 0.80 0.98 238U(n,f)137Cs (Cd) 4.47E-14 3.72E-14 3.90E-14 1.20 1.15 237Np(n,f)137Cs (Cd) 4.25E-13 3.58E-13 3.99E-13 1.19 1.06 Catawba Unit 1 Capsule Y Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 63Cu(n,)60Co 4.98E-17 4.67E-17 4.77E-17 1.07 1.04 54Fe(n,p)54Mn 4.58E-15 5.12E-15 4.83E-15 0.89 0.95 58Ni(n,p)58Co 6.74E-15 7.15E-15 6.81E-15 0.94 0.99 59Co(n,)60Co 3.56E-12 3.71E-12 3.53E-12 0.96 1.01 59Co(n,)60Co (Cd) 2.06E-12 2.57E-12 2.09E-12 0.80 0.99 238U(n,f)137Cs (Cd) 2.69E-14 2.70E-14 2.56E-14 1.00 1.05 237Np(n,f)137Cs (Cd) 2.57E-13 2.55E-13 2.49E-13 1.01 1.03 Catawba Unit 1 Capsule V Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 63Cu(n,)60Co 4.25E-17 4.43E-17 4.11E-17 0.96 1.03 54Fe(n,p)54Mn 4.17E-15 4.84E-15 4.29E-15 0.86 0.97 58Ni(n,p)58Co 5.62E-15 6.76E-15 5.93E-15 0.83 0.95 59Co(n,)60Co 3.14E-12 3.47E-12 3.11E-12 0.91 1.01 59Co(n,)60Co (Cd) 1.69E-12 2.41E-12 1.72E-12 0.7 0.98 238U(n,f)137Cs (Cd) 2.72E-14 2.55E-14 2.30E-14 1.07 1.18 237Np(n,f)137Cs (Cd) 2.21E-13 2.40E-13 2.20E-13 0.92 1.00 Note:

See Section C.1.2 for details describing the BE reaction rates.

C-26 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-5 Comparison of M/C and BE Reaction Rates at the Surveillance Capsule Center (cont.)

Catawba Unit 1 Capsule U Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 63Cu(n,)60Co 4.50E-17 4.57E-17 4.33E-17 0.99 1.04 54Fe(n,p)54Mn 4.33E-15 5.02E-15 4.49E-15 0.86 0.96 58Ni(n,p)58Co 5.93E-15 7.01E-15 6.23E-15 0.85 0.95 59Co(n,)60Co 3.15E-12 3.67E-12 3.13E-12 0.86 1.01 59Co(n,)60Co (Cd) 1.78E-12 2.53E-12 1.81E-12 0.70 0.98 238U(n,f)137Cs (Cd) 2.89E-14 2.65E-14 2.41E-14 1.09 1.20 237Np(n,f)137Cs (Cd) 2.24E-13 2.51E-13 2.27E-13 0.89 0.99 Catawba Unit 1 Capsule X Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 63Cu(n,)60Co 4.56E-17 4.57E-17 4.37E-17 1.00 1.04 54Fe(n,p)54Mn 4.35E-15 5.02E-15 4.61E-15 0.87 0.94 58Ni(n,p)58Co 6.07E-15 7.01E-15 6.45E-15 0.87 0.94 59Co(n,)60Co 3.44E-12 3.67E-12 3.41E-12 0.94 1.01 59Co(n,)60Co (Cd) 1.84E-12 2.53E-12 1.88E-12 0.73 0.98 238U(n,f)137Cs (Cd) 3.19E-14 2.65E-14 2.58E-14 1.20 1.23 237Np(n,f)137Cs (Cd) 2.94E-13 2.51E-13 2.75E-13 1.17 1.08 Note:

See Section C.1.2 for details describing the BE reaction rates.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-27 WCAP-17669-NP June 2013 Revision 0 Table C-6 Comparison of Calculated and BE Exposure Rates at the EVND Capsule Center from Catawba Unit 1 Catawba Unit 1 EVND Capsule A Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 4.88E-18 5.57E-18 4.92E-18 0.88 0.99 54Fe(n,p)54Mn 2.70E-17 2.96E-17 2.66E-17 0.91 1.02 58Ni(n,p)58Co 3.63E-17 4.11E-17 3.67E-17 0.88 0.99 63Cu(n,)60Co 3.64E-19 4.12E-19 3.64E-19 0.88 1.00 59Co(n,)60Co 3.99E-14 7.54E-14 4.06E-14 0.53 0.98 59Co(n,)60Co (Cd) 1.87E-14 2.25E-14 1.87E-14 0.83 1.00 238U(n,f)137Cs (Cd) 1.24E-16 1.44E-16 1.30E-16 0.86 0.95 237Np(n,f)137Cs (Cd) 1.97E-15 1.92E-15 1.87E-15 1.03 1.05 Catawba Unit 1 EVND Capsule B Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 6.92E-18 7.38E-18 6.86E-18 0.94 1.01 54Fe(n,p)54Mn 3.93E-17 4.09E-17 3.87E-17 0.96 1.02 58Ni(n,p)58Co 5.31E-17 5.71E-17 5.38E-17 0.93 0.99 63Cu(n,)60Co 4.77E-19 5.33E-19 4.85E-19 0.90 0.98 59Co(n,)60Co 5.80E-14 1.03E-13 5.90E-14 0.56 0.98 59Co(n,)60Co (Cd) 2.99E-14 3.61E-14 2.98E-14 0.83 1.00 238U(n,f)137Cs (Cd) 1.94E-16 2.06E-16 1.99E-16 0.94 0.98 237Np(n,f)137Cs (Cd) 3.17E-15 2.87E-15 3.00E-15 1.10 1.05 Catawba Unit 1 EVND Capsule C Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 7.00E-18 7.75E-18 7.00E-18 0.90 1.00 54Fe(n,p)54Mn 4.21E-17 4.39E-17 4.05E-17 0.96 1.04 58Ni(n,p)58Co 5.28E-17 6.18E-17 5.56E-17 0.85 0.95 63Cu(n,)60Co 4.98E-19 5.55E-19 4.99E-19 0.90 1.00 59Co(n,)60Co 6.93E-14 1.22E-13 7.05E-14 0.57 0.98 59Co(n,)60Co (Cd) 3.57E-14 4.31E-14 3.56E-14 0.83 1.00 238U(n,f)137Cs (Cd) 2.15E-16 2.30E-16 2.15E-16 0.93 1.00 237Np(n,f)137Cs (Cd) 3.65E-15 3.37E-15 3.43E-15 1.08 1.06 Note:

See Section C.1.2 for details describing the BE reaction rates.

C-28 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-6 Comparison of Calculated and BE Exposure Rates at the EVND Capsule Center from Catawba Unit 1 (cont.)

Catawba Unit 1 EVND Capsule D Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 2.05E-18 2.19E-18 1.98E-18 0.94 1.04 54Fe(n,p)54Mn 1.19E-17 1.31E-17 1.21E-17 0.91 0.99 58Ni(n,p)58Co 1.71E-17 1.88E-17 1.75E-17 0.91 0.98 63Cu(n,)60Co 1.28E-19 1.55E-19 1.32E-19 0.83 0.97 59Co(n,)60Co 1.84E-14 5.20E-14 1.93E-14 0.35 0.95 59Co(n,)60Co (Cd) 1.19E-14 1.54E-14 1.17E-14 0.78 1.02 238U(n,f)137Cs (Cd) 7.98E-17 7.55E-17 7.48E-17 1.06 1.06 237Np(n,f)137Cs (Cd) 1.54E-15 1.19E-15 1.39E-15 1.29 1.11 Catawba Unit 1 EVND Capsule E Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 4.79E-18 5.54E-18 4.76E-18 0.87 1.01 54Fe(n,p)54Mn 3.00E-17 3.32E-17 2.90E-17 0.90 1.03 58Ni(n,p)58Co 3.85E-17 4.76E-17 4.07E-17 0.81 0.94 63Cu(n,)60Co 3.33E-19 3.92E-19 3.34E-19 0.85 1.00 59Co(n,)60Co 4.67E-14 1.11E-13 4.81E-14 0.42 0.97 59Co(n,)60Co (Cd) 2.74E-14 3.47E-14 2.71E-14 0.79 1.01 238U(n,f)137Cs (Cd) 1.74E-16 1.92E-16 1.71E-16 0.91 1.01 237Np(n,f)137Cs (Cd) 3.21E-15 3.00E-15 2.99E-15 1.07 1.08 Catawba Unit 1 EVND Capsule F Reaction Reaction Rate [rps/atom]

M/C M/BE Measured Calculated BE 46Ti(n,p)46Sc 1.63E-18 2.13E-18 1.59E-18 0.76 1.02 54Fe(n,p)54Mn 9.68E-18 1.28E-17 9.67E-18 0.76 1.00 58Ni(n,p)58Co 1.42E-17 1.84E-17 1.41E-17 0.77 1.01 63Cu(n,)60Co 1.04E-19 1.50E-19 1.08E-19 0.69 0.96 59Co(n,)60Co 2.38E-14 5.21E-14 2.43E-14 0.46 0.98 59Co(n,)60Co (Cd) 1.25E-14 1.54E-14 1.24E-14 0.81 1.01 238U(n,f)137Cs (Cd) 5.47E-17 7.44E-17 5.79E-17 0.74 0.94 237Np(n,f)137Cs (Cd) 1.02E-15 1.19E-15 9.89E-16 0.86 1.03 Note:

See Section C.1.2 for details describing the BE reaction rates.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-29 WCAP-17669-NP June 2013 Revision 0 Table C-7 Comparison of Calculated and BE Exposure Rates at the Surveillance Capsule Center from Catawba Unit 1 Capsule ID (E > 1.0 MeV) [n/cm2-s]

Calculated BE Uncertainty (1)

BE/C Z

1.17E+11 1.23E+11 6%

1.04 Y

8.40E+10 7.94E+10

a.

6%

0.94 V

7.91E+10 7.19E+10 6%

0.90 U

8.26E+10 7.52E+10 6%

0.91 X

8.26E+10 8.25E+10 6%

0.99 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

Capsule ID Iron Atom Displacement Rate [dpa/s]

Calculated BE Uncertainty (1)

BE/C Z

2.23E-10 2.38E-10 8%

1.06 Y

1.59E-10 1.53E-10 8%

0.96 V

1.50E-10 1.39E-10 8%

0.92 U

1.57E-10 1.45E-10 8%

0.92 X

1.57E-10 1.61E-10 8%

1.02 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

C-30 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-8 Comparison of M/C Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions from Catawba Unit 1 Reaction M/C Ratio Average Standard Deviation Capsule Z

Capsule Y

Capsule V

Capsule U

Capsule X

63Cu(n,a)60Co 1.25 1.07 0.96 0.99 1.00 1.05 11.1 54Fe(n,p)54Mn 1.04 0.89 0.86 0.86 0.87 0.90 8.5 58Ni(n,p)58Co 0.93 0.94 0.83 0.85 0.87 0.88 5.5 238U(n,f)137Cs (Cd) 1.20 1.00 1.07 1.09 1.20 1.11 7.8 237Np(n,f)137Cs (Cd) 1.19 1.01 0.92 0.89 1.17 1.04 13.4 Average 1.12 0.98 0.93 0.94 1.02 1.00

% Standard Deviation 11.8 7.0 10.1 10.9 15.5 12.8 Note:

The average and % standard deviation values in bold face type represent the average and standard deviation of the entire twenty-five sample threshold foil data set.

Table C-9 Comparison of BE/C Exposure Rate Ratios for Surveillance Capsules from Catawba Unit 1 Capsule ID BE/C Ratio (E > 1.0 MeV) dpa/s Z

1.04 1.06 Y

0.94 0.96 V

0.90 0.92 U

0.91 0.92 X

0.99 1.02 Average 0.96 0.98

% Standard Deviation 6.1%

6.4%

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-31 WCAP-17669-NP June 2013 Revision 0 Table C-10 Comparison of Calculated and BE Exposure Rates at the EVND Capsules at the Core Midplane of Catawba Unit 1 Capsule ID (E > 1.0 MeV) [n/cm2-s]

Calculated BE Uncertainty (1)

BE/C A

4.62E+08 4.24E+08 6%

0.91 B

6.77E+08 6.64E+08 6%

0.98 C

7.73E+08 7.33E+08 6%

0.94 E

6.70E+08 6.09E+08 6%

0.90 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

Capsule ID Iron Atom Displacement Rate [dpa/s]

Calculated BE Uncertainty (1)

BE/C A

1.43E-12 1.33E-12 9%

0.92 B

2.17E-12 2.14E-12 9%

0.98 C

2.56E-12 2.46E-12 9%

0.96 E

2.24E-12 2.08E-12 9%

0.92 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the core midplane following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

C-32 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-11 Comparison of Calculated and BE Exposure Rates at the EVND Capsules at Off-Midplane Positions of Catawba Unit 1 Capsule ID (E > 1.0 MeV) [n/cm2-s]

Calculated BE Uncertainty (1)

BE/C D

2.63E+08 2.70E+08 6%

1.02 F

2.60E+08 2.06E+08 6%

0.79 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the EVND capsule center following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

Capsule ID Iron Atom Displacement Rate [dpa/s]

Calculated BE Uncertainty (1)

BE/C D

8.99E-13 9.34E-13 9%

1.03 F

8.97E-13 7.28E-13 9%

0.81 Note:

Calculated results are based on the RAPTOR-M3G transport calculations taken at the EVND capsule center following the completion of each respective capsules irradiation period and are the average neutron exposure over the irradiation period for each capsule. See Section C.1.2 for details describing the BE exposure rates.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-33 WCAP-17669-NP June 2013 Revision 0 Table C-12 Comparison of M/C Sensor Threshold Reaction Rates from Ex-Vessel Midplane Capsules at Catawba Unit 1 Reaction M/C Ratio Average

% Standard Deviation Capsule A Capsule B Capsule C Capsule E 46Ti(n,p)46Sc 0.88 0.94 0.90 0.87 0.90 3.4 54Fe(n,p)54Mn 0.91 0.96 0.96 0.90 0.93 3.4 58Ni(n,p)58Co 0.88 0.93 0.85 0.81 0.87 5.8 63Cu(n,a)60Co 0.88 0.90 0.90 0.85 0.88 2.7 238U(n,f)137Cs (Cd) 0.86 0.94 0.93 0.91 0.91 3.9 237Np(n,f)137Cs (Cd) 1.03 1.10 1.08 1.07 1.07 2.8 Average 0.91 0.96 0.94 0.90 0.93

% Standard Deviation 6.9 7.3 8.5 10.0 8.1 Note:

The average and % standard deviation values in bold face type represent the average and standard deviation of the entire twenty-four sample threshold foil data set.

C-34 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 Table C-13 Comparison of M/C Sensor Threshold Reaction Rates from Ex-Vessel Off-Midplane Capsules at Catawba Unit 1 Reaction M/C Ratio Average

% Standard Deviation Capsule D Capsule F 46Ti(n,p)46Sc 0.94 0.76 0.85 15.0 54Fe(n,p)54Mn 0.91 0.76 0.84 12.7 58Ni(n,p)58Co 0.91 0.77 0.84 11.8 63Cu(n,)60Co 0.83 0.69 0.76 13.0 238U(n,f)137Cs (Cd) 1.06 0.74 0.90 25.1 237Np(n,f)137Cs (Cd) 1.29 0.86 1.08 28.3 Average 0.99 0.76 0.88

% Standard Deviation 16.6 7.3 19.0 Note: The average and % standard deviation values in bold face type represent the average and standard deviation of the entire twelve sample threshold foil data set.

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-35 WCAP-17669-NP June 2013 Revision 0 Table C-14 Comparison of BE/C Exposure Rate Ratios from Ex-Vessel Midplane Capsules at Catawba Unit 1 Capsule ID BE/C Ratio (E > 1.0 MeV) dpa/s A

0.91 0.92 B

0.98 0.98 C

0.94 0.96 E

0.90 0.92 Average 0.93 0.95

% Standard Deviation 3.9%

3.2%

Table C-15 Comparison of BE/C Exposure Rate Ratios from Ex-Vessel Off-Midplane Capsules at Catawba Unit 1 Capsule ID BE/C Ratio (E > 1.0 MeV) dpa/s D

1.02 1.03 F

0.79 0.81 Average 0.91 0.92

% Standard Deviation 18.0%

16.9%

C-36 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17669-NP June 2013 Revision 0 C.2 REFERENCES C-1 Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.

C-2 WCAP-11527, Rev. 0, Analysis of Capsule Z from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, June 1987.

C-3 WCAP-13720, Analysis of Surveillance Capsule Y from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, June 1993.

C-4 WCAP-15117, Rev. 0, Analysis of Capsule V and the Dosimeters from Capsules U and X from the Duke Power Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program, October 1998.

C-5 WCAP-16869-NP, Rev. 1, Ex-Vessel Neutron Dosimetry Program for Catawba Unit 1 Cycles 15 and 16, May 2009.

C-6 A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

C-7 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

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