ML20238F618

Safety Evaluation Supporting Amends 170 & 162 to Licenses NPF-35 & NPF-52,respectively

ML20238F618
Person / Time
Site:	Catawba
Issue date:	08/28/1998
From:	NRC (Affiliation Not Assigned)
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Download: ML20238F618 (9)

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1 UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20061H1001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.170 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO.162 TO FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CORPORATION. ET AL.

l CATAWBA NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By letter dated September 15,1997, as supplemented by letters dated March 5, April 27, June 15, July 22, and August 10,1998, Duke Energy Corporation, et al. (the licensee, DEC) submitted a request for amendments to the Catawba Nuclear Station, Units 1 and 2, Technical Specifications (TS), revising the units' pressure-temperature (P-T) limits curves in the TS, and extending the period of their applicability from 10 to 15 effective full power years (EFPYs) of operation. On February 6 and March 26,1998, the staff issued requests for additional information. On March 5, April 27, June 15, July 22, and August 10,1998, the licensee supplemented the amendment request. The June 15,1998, supplement also expanded the scope of the licensee's amendment request, proposing changes and additional requirements to address low-temperature overpressure protection of the primary coolant system.

The March 5, April 27, July 22, and August 10,1998, letters provided additional information that did not change the scope of the September 15,1997, application and the initial proposed no significant hazards consideration determination.

2.0 DISCUSSION AND EVALUATION The staff reviewed the licensee's submittals in three specific areas. These areas are addressed as follows in Sections 2.1,2.2, and 2.3.

2.1 Neutron Fluence The licensee reported that the fluence evaluation, which is the basis for the current as well as

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the proposed revised P-T curves, was performed when the first four surveillance capsules were removed and evaluated at the end of about 1 and 4 EFPYs. The results are documented in Westinghouse Topical Reports WCAP-13720 (" Analysis of Capsule Y from the Duke Power i

Company Catawba Unit 1 Reactor Vessel Radiation Surveillance Program" by J. M. Chicots, et al., Westinghouse Electric Corporation, June 1993), and WCAP-13875 (" Analysis of Capsule X from the Duke Power Company Catawba Unit 2 Reactor Vessel Radiation Surveillance Program," by E. Terek, et al., Westinghouse Electric Corporation, February 9909040139 990828

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, 1994). Both these topical reports were previously submitted to the NRC docket files and are l'

available to the public.

l At the time these' evaluations were performed, ENDF/B-IV based nuclear cross sections were

' used. Since that time, the staff has identified a number of nonconservative values in the iron inelastic. cross sections.'. Currently the staff-recommended cross sections are based on the

' ENDF/B-VI cross section file. For plants like Catawba, which are equipped with neutron pads

- (partial thermal shields), ENDF/B-VI based cross sections yield calculated fluence values about

'15 percent higher than those based on ENDF/B-IV. This observatbn is based on the staffs experience with a large number of computations of fluence values.

In the calculation of the current P-T curves, the licensee ' assumed fluence values based on the measured values, which were higher than the calculated fluence. The new proposed fluence values are based on the measured values. However, had the fluences been recalculated using cross sections based on ENDF/B-VI, the resulting values would have been up to 3 percent

. lower than the measured values. Therefore, the licensee's proposed values are up to 3 percent more conservative compared to the calculated values.

' At the Catawba reactors, dosimeters were located in the surveillance capsules, which were fastened outside the neutron pads (partial thermal shields). Based on the staff's experience,

. fluence values calculated using the ENDF/8-VI based cross sections and accounting for the presence of the thermal shield, would be about 15 percent higher. WCAP-13720 and WCAP-13875, which describe the analysis of surveillance capsule Y and summary of capsule Z for Catawba Unit 1, and analysis of surveillance capsule X and summary of capsule Z for

' Unit 2, established that the measured values were indeed higher than the corresponding calculated values by 15 to 18 percent. As previously stated, the fluence for the current P-T -

curves and the proposed P-T curves is based on the measured values. Therefore, the proposed fluence values would be up to 3 percent higher than the values calculated using the ENDF/B-VI cross sections. Th'e staff, therefore, concludes that the proposed fluence values are conservative and are, thus, acceptable.

2.2 Proposed New P-T Limits Curves

The staff evaluated the P-T limits based on the following NRC regulations and guidance:

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix G, " Fracture Toughtness Requirements"; Generic Letter (GL) 88-11, "NRC Position on Radiation

. Embrittlement of Reactor Vessel Materials and its impact on Plant Operations"; GL 92-01,

- Revision 1, Supplement 1, " Reactor Vessel Structural Integrity"; Regulatory Guide (RG) 1.99,

Revision 2 (Rev. 2),' " Radiation Embrittlement of Reactor Vessel Materials;" and Standard Review Plan (SRP) Section 5.3.2, " Pressure-Temperature Limits.." GL 88-11 advised licensees that the staff would use RG 1.99, Rev. 2, to review P-T limit curves. RG 1.99, Rev. 2, contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy (USE) resulting from neutron radiation. GL 92-01, Revision 1, requested that licensees submit their reactor pressure vessel (RPV) data for their plants to the staff.

GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. These data are used by the

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staff as the basis for the staffs review of P-T limit curves, and as the basis for the staffs review of pressurized thermal shock (PTS) assessments (10 CFR 50.61 assessments). Appendix G to 10 CFR Part 50 requires that P-T limit curves for the RPV be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

SRP 5.3.2 provides an acceptable method of calculating the P-T limits for ferritic materials in

. the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of i

Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor K,, which is a function of the stress state and flaw configuration. The methods of Appendix G postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress. This flaw is postulated to have a depth that is equal to one-fourth of the RPV beltline thickness and a length equal to 1.5 times the RPV beltline thickness. The critical locations in the RPV beltline region for calculating heatup and cooldown P-T limit curves are the one-fourth thickness (1/4T) and three-fourth thickness (3/4T) locations, which correspond to the depth of the maximum postulated flaw, if initiated and grown i

from the inside and outside surfaces of the RPV, respectively.

The Appendix G, ASME Code methodology requires that licensees determine the adjusted I

reference temperature (ART or RTuo7). The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTuor), the mean value of the adjustment in reference temperature caused by irradiation (ARTuor), and a margin (M) term.

The ARTuor is a product of a chemistry factor and a fluence factor. The chemistry factor is

'4 dependent upon the amount of copper and r.ickelin the material and may be determined from tables in RG 1.99, Rev. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTuoy is a plant-specific or a generic value and whether the chemistry factor was determined using the tables in RG 1.99, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of initial RTuor, copper and nickel contents, fluence and calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term.

Unit 1 For the Unit 1 reactor vessel, the licensee determined that the most limiting material at the 1/4T and 3/4T locations is the lower shell forging plate 04. This plate was fabricated using plate heat 527708. The licensee calculated an ART of 43 *F at the 1/4T location and 26 'F at the 3/4T location at 15 EFPYs. The neutron fluence used in the ART calculation was 7.1 X 10 n/cm at 2

the 1/4T location and 2.57 X 10 n/cm at the 3/4T location. The initial RTuor for the limiting 2

plate was -13 'F. The margin term used in calculating the ART for the limiting plate was 28 *F at the 1/4T location and 19.6 'F at the 3/4T location, as permitted by Position 1.1 of RG 1.99, Rev.2.

The ART is determined using the chemistry values for each beltline material of Unit 1. The Reactor Vessel Integrity Database (RVlD) contains chemistry values for each beltline material for all light water reactors in the U.S. The licensee provided updated chemistry data for the

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. beltline materials of Unit 1 by letters dated August 12,1993 (D. L. Rehn to NRC, transmitting WCAP-13720) and July 3,1997 (M. S. Tuckman to NRC). It should be noted that the staff and the licensee used the updated chemistry values to determine the ART for the beltline materials in Unit 1.

The beltline welds in the Unit 1 RPV were all fabricated using weld wire hest 895075. The f

staffs review of data in the RVID indicates that the initial RTuor for weld wire heat 895075, as J

reported by the licensee for Unit 1, may be nonconservative. The staff does not agree with the l

initial RTuor of-51 *F for weld wire heat 895075, because this initial RTwor'value is nonconservative when compared to data from other plants for this weld wire heat. However, the ART for the subject weld using the most conservative value of the initial RTuor, as indicated

- in the RVID, for heat number 895075, is -40 *F, which still results in an ART that is less than the' ART for the limiting material in the RPV. Therefore, using that value will have no impact on the P-T limits evaluation.

j The staff performed an independent calculation of the ART values for the limiting material using the methodology in RG 1.29, Rev. 2. Based on these calculations, the staff verified that the licensee's limiting material for the Unit i reactor vessel is the lower shell forging plate 04 that was fabricated using plate heat 527708. The staff's calculated ART value for the limiting

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material agreed with the licensee's calculated ART value. Substituting the ART values for the Unit 1 limiting plate into the equations in SRP 5.3.2, the staff verified that the proposed P-T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G of 10 CFR Part 50.

i in addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange 1

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material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of I

the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 *F for normal operation and by 90 *F for hydrostatic pressure tests and leak tests. Based on the flange RTuor of 10 *F for Unit 1 provided by the licensee, the staff has determined that the proposed P-T limits, as depicted in the proposed Figures 3.4-2 and 3.4-3, satisfy the requirement for the closure flange region during normal operation and hydrostatic pressure test and leak test.

Unit 2 For the Unit 2 reactor vessel, the licensee determined that the most limiting material at ths 1/4T and 3/4T locations is the intermediate shell plate B8605-2. The licensee calculated an ART of 113 *F at the 1/4T location and 96 *F at the 3/4T location at 15 EFPYs. The neutron fluence used in the ART calculation was 7.0 X 10 n/cm at the 1/4T location and 2.5 X 10 n/cm at 2

2 the 3/4T location. The initial RTuor for the limiting plate was 33 *F. The margin term used in calculating the ART for the limiting plate was 34 *F at the 1/4T location and 32.2 *F at the 3/4T location, as permitted by Position 1.1 of RG 1.99, Rev. 2.

The ART is determined using the chemistry values for each beltline material of Unit 2. The RVID contains chemistry values for each beltline material for all light water reactors in the U.S.

The licensee provided updated chemistry data for the beltline materials of Unit 2 by letters

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, dated March 2,1994 (D. L. Rehn to NRC, transmitting WCAP-13875) and July 3,1997

. (M. S. Tuckman to NRC). It should be noted that the staff and the licensee used the updated chemistry values to determine the ART for the beltline a:aterials in Unit 2. In addition, the staff compared the licensee's best estimate chemistry data for weld wire heat 83648 against the best estimate chemistry values in the Combustion Engineering Owners Group (CEOG) Report CE NPSD-1039, Revision 2. The staff verified that the licensee's best estimate Cu and Ni values were the same as the values in the CEOG Report.

The beltline welds in the Unit 2 RPV were all fabricated using weld wire heat 83648. The staff reviewed the initial RTer values in the RVID for welds made of weld wire heat 83648 for all

. plants. The staff found that the initial RT, value of -80*F for the Unit 2 axial and girth welds was acceptable, since there were no other plants with the same weld wire heat.

The staff performed an independent calculation of the ART values for the limiting material using the methodology in RG 1.99, Rev. 2. Based on these calculations, the staff verified that the licensee's limiting material for the Unit 2 reactor vessel is the intermediate shell plate B8605-2.

The staff's calculated ART value for the limiting material agreed with the licensee's calculated ART value. Substituting the ART values for the Unit 2 limiting plate into the equations in SRP 5.3.2, the staff verified that the proposed P-T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G of 10 CFR Part 50, in addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes a minimum temperature at the closure head flange based on the reference temperature for the flange material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 'F for normal operation and by 90 'F for hydrostatic pressure tests and leak tests. Based on the flange RT 7 of 10 'F for Unit 2 provided by the licensee, the staff has determined that the proposed P-T limits satisfy the requirement for the closure flange region during normal operation and hydrostatic pressure test and leak test.

Prooosed TS Chanoes Th'e staff concludes that the proposed P-T limits for the Catawba Nuclear Station, Units 1 and 2, reactor coolant systems for heatup, cooldown, leak test, and criticality satisfy the requirements in Appendix G to Section XI of the ASME Code and Appendix G of 10 CFR Part 50 for 15 EFPYs. The proposed P-T limits also satisfy GL 88-11 because the method in RG 1.99, Rev. 2, was used to calculate the ART. Therefore, the proposed P-T limits, depicted in revised i

TS Figures 3.4-2 and 3.4-3, may be incorporated into the TS.

The licensee also proposed to revise Table 4.4-5, " Reactor Vessel Material Surveillance Schedule," modifying the actual capsule number, lead factors and withdrawal schedule. For Unit 1, the proposed withdrawal schedule is in accordance with the ASTM E185-73 Standard.

L Appendix H of 10 CFR Part 50 states that "...the design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E185 that is current on l

the issue date of the ASME Code to which the reactor vessel was purchased. Later editions of l

ASTM E 185 may be used, but including only those editions through 1982.".The staff

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! concludes that the revised Catawba Unit i surveillance capsule withdrawal schedule satisfies the requirements of Appendix H to 10 CFR Part.50, and is acceptable.

For Unit 2, the proposed withdrawal schedule in the revised Table 4.4-5 meets the intent of the ASTM E185-82 Standard. Since this edition of the standard is a later edition of the standard that was current on the issue date of the ASME Code to which the reactor vessel was purchased and is included through the 1982 Edition, the Catawba revised surveil'ance capsule withdrawal schedule satisfies the requirements of Apnendix H to10 CFR Part 50 and is acceptable.

2.3 Low-Temperature Overpressure (LTOP) Considerations The LTOP system mitigates overpressure transients at low temperatures so that the integrity of the reactor coolant pressure boundary is not compromised by violating 10 CFR Part 50, Appendix G. Catawba's LTOP system uses the pressurizer power-operated relief valves-(PORVs) to accomplish this function. The system is manually enabled by operators and uses a l

single setpoint as the lift pressure for the PORVs. The design basis of Catawba's LTOP system considers both mass-addition and heat-addition transients during water-solid reactor coolant system (RCS) conditions. Two mass-addition cases are analyzed. One analysis accounts for injection from a single safety injection pump while the other accounts for injection from a single centrifugal charging pump. The heat-addition analysis accounts for heat input from the secondary sides of all steam generators (SGs) into the RCS, upon starting a single reactor -

coolant pump (RCP). The heat-addition transient analysis assumes the SGs' secondary side temperatures are 50 'F higher than the RCS temyrature.

LTOP Enable Temperature The LTOP enable temperature is the temperature below which the LTOP system is required to be operable. The licensee's proposed enable temperature accounts for instrumentation -

uncertainties associated with the instrumentation used to enable the LTOP system and the temperature difference between the reactor coolant and the metal at a distance one-fourth of the vessel wall thickness from the inside surface in the beltline region. Therefore, the minimum allowed enable temperature is calculated as RTuor + 90 'F + E, + delta-Ty; where, E,, refers to instrument error and delta-Tu refers to the ter,5perature difference between the reactor coolant and the metal at a distance one-fourth of the vessel wall thickness from the inside surface in the beltline region. This method is consistent with Branch Technical Position (BTP)

RSB 5-2, which is attached to SRP Section 5.2.2, " Overpressure Protection," Revision 2.

The licensee proposed an LTOP enable temperature of a 285 'F for both units. Based on vessel material data (RTuor values of 43 *F for Unit 1 and 112.6 'F for Unit 2), instrumentation uncertainties (the licensee conservatively assumed 10 *F), and delta-Tu (the licensee used 19.1 *F based on a Westinghouse evaluation), the minimum allowed enable temperatures for Cata'wba Units 1 and 2 are 162.1 'F and 231.7 'F, respectively. Limiting RTuor determinations

- were reviewed by the staff (see above). However, for the values presented above, the proposed LTOP enable temperature of a 285 'F is conservative with respect to the calculated minimum LTOP enable temperatures based on BTP RSB 5-2. Therefore, the staff finds the licensee's proposed enable temperature acceptable.

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, LTOP Actuation Setooint For the mass-addition cases for Catawba' Units 1 and 2 and for the heat-addition case for Catawba Unit 2, the proposed LTOP actuation setpoint of s 400 psig was verified to protect the P-T limits using the methods provided in Westinghouse reports, " Pressure Mitigating System Transient Analysis Results," July 1977, and " Supplement to the July 1977 Report, Pressure Mitigating System Transient Analysis Results," September 1977. These reports were utilized in determining the initial LTOP actuation setpoint. For the heat-addition case for Catawba Unit 1, the licensee performed plant-specific analyses using the RETRAN-02 MOD 5.1 Code to verify that the proposed setpoint for Unit 1 protects the P-T limits. This additional analysis was performed because Unit 1 has SGs with heat transfer areas that placed them outside of the p

limitations placed on the heat addition analyses in the previously mentioned Westinghouse l

reports.

For the mass-addition case, the licensee performed two analyses. One with injection from a i

safety injection pump and one with injection from a centrifugal charging pump. For the heat-addition case, the licensee assumed the start of a single RCP with the temperature of the secondary side of the SGs 50 *F higher than the RCS. In its analyses, the licensee assumed that only one PORV was available for pressure relief. Additionally, the licensee did not credit the RHR suction relief valves for mitigating the pressure transient in the analyses. The licensee evaluated the results of the heat-addition and mass-addition cases with a water solid RCS and determined that the mass addition cases remained limiting. The licensee considered the impact of SG tube plugging and accounted for static and dynamic head effects as well as instrumentation uncertainties in the final determination of the LTOP setpoints.

The licensee's analyses did not account for injection from the cold leg accumulators.

Additionally, the dynamic head effect was divided into several temperature regions with different RCP operational restrictions. For Unit 1, the licensee used the following assumptions with regard to RCP operation:

RCS Maximum Number of RCPs Cold Leg Temperature Allowed in Operation 267*F 1

268*F 2

273*F 4

For Unit 2, the licensee used the following assumption:

RCS Maximum Number of RCPs Cold Leg Temperature Allowed in Operation 267*F 1

l 273'F 2

295*F 4

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, e-The licensee's analyses demonstrated that the proposed actuation setpoint of s 400 psig is adequate to protect the P-T limits. The staff has reviewed the licensee's analyses and finds the licensee's proposed LTOP actuation setpoint acceptable. Additionally, the licensee proposed additional TS restrictions on the number of pumps capable of injection, accumulator discharge valves, and RCP operation. The staff has reviewed the proposed additional TS changes. The staff finds that the proposed changes are necessary to bring the TS into agreement with the assumptions in the analyses and, therefore, are acceptable.

With regard to the use of the RETRAN-02 MODS.1 Code to perform the heat-addition transient for Catawba Unit 1, the staff has reviewed and accepted the use of this Code for non-LOCA transient analyses in letters dated April 12,1994, November 1,1991 October 19,1988, and September 4,1984. The RETRAN RCS model used for this analysis has also been approved i

by the NRC and is described in Duke Power Company Topical Report DPC-NE-3000-PA, l

Revision 1, dated December 1997. The licensee has reviewed the restrictions and limitations placed in the staff's safety evaluation reports associated with RETRAN-02 MODS.1 computer Code and the RETRAN RCS model described in DPC-NE-3000-PA. The licensee concluded that the use of these methods for the heat-addition analysis was appropriate. The staff reviewed the associated reports and concurs with the licensee's assessment.

Minimum Temperature The licensee evaluated the minimum temperature at which the LTOP system can protect the l

P-T limits. In this evaluation the licensee considered the mass-addition transient since it was more limiting and accounted for static and dynamic head effects as well as instrumentation uncertainties. Additionally, since the licensee's P-T curves did not extend below 85 *F, the licensee used extrapolated data for the lower temperatures. The extrapolations were conservatively calculated based on a constant slope of a straight line using the last two data points from the steady state curve. The licensee's evaluation determined that the minimum p

temperatures at which the LTOP system can protect the P-T curves are 63.2 'F for Unit 1 and 62.1 'F for Unit 2. Accordingly, the licensee proposed a note to TS 3.4.9.3 that limits the temperature to a minimum of 65 'F when the PORVs are relied upon for LTOP. Extrapolation of the P-T curves should be reviewed by the staff. However, for the values provided in the licensee's submittals, the staff finds the minimum temperature of 65 *F acceptable.

Summarv of LTOP Review The staff has reviewed the licensee's proposed LTOP system enable temperature and actuation setpoint. The staff has also reviewed the licensee's proposed TS related to the number of pumps capable of injection into the RCS, accumulator discharge valves, RCP I

l operation, and the minimum temperature protected by the LTOP system. Based on the evaluation provided in Section 2 of this SE, the staff finds the licensee's proposed changes acceptable. The staff further finds acceptable-tiis licensee's use the RETRAN-02 MOD 5.1 Code for analyzing the heat-addition scenario for Catawba Unit 1.

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3.0 STATE CONSULTATION

I in accordance with the Commission's regulations, South Carolina State official Virgil Autrey was notified of the proposed issuance of the amendments. The State official had no comments.

4.0. ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no i

significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued two proposed findings that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 52580 dated October 8,1997; 63 FR 40553 dated

. July 29,1998). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

5.0 CONCLUSION

The Commission has concluded, based on the consideradons discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by

- operation in the proposed manner, (2) such activities will be conducted in compliance with the l

Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.-

Principal Contributors: Meena K. Khanna Lambros Lois Mohammed A. Shuaibi Peter S. Tam -

Date:

August 28, 1998-N----.---.-____

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