ML20217K824
| ML20217K824 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/21/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217K822 | List: |
| References | |
| NUDOCS 9910260251 | |
| Download: ML20217K824 (9) | |
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.8 UNITED STATES yo 1
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C 20bo5-0001 Q*...+]'
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.' 143 TO FACILITY OPERATING LICENSE NO. NPF-12 1
SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PU^BLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION. UNIT NO 1 DOCKET NO. 50-395
1.0 INTRODUCTION
j By application dated August 19,1999, as supplemented by letter dated October 8,1I99, South f
Carolina Electric & Gas Company (SCE&G, the licensee) requested changes to the Technical j
Specifications (TS) for the Virgil C. Summer Nuclear Station (VCSNS). The proposed changes
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would revise TS 3/4.4.9, " Reactor Coolant System Pressure /Temperaturo Limus" to incorporate J
the new Pressure-Temperature (P-T) Limits Curves consistent wim the analysis results of reactor vessel surveillance capsule W. This change is required since the c'urrent curves are only applicable to 13 Effective Full Power Years (EFPY) and will expire r3n or about October 23,
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1999. In conjunction with the requested arnendment, the licensee has requested an exemption from the requirements of Title 10 of the Code of Federal Reoulationi(10 CFR) Part 50, Appendix 0. The licensee requests, as an altemate requirement..to utilize the methodology q
presented in the 1996 ASME Section XI, Appendix G and the ASME Code Case N-640, "Altemative Reference Fracture Toughness for Development ct P-T Curves for Section XI, Division I," dated Feb uary 1999. The staff's review of the licensee's exemption request is the subject of a separate safety evaluation dated October 20,4999. The Bases section for the Pressure-Temperature Limits is be;ng revised to accuratsly reflect current industry standards and regulations. The October 8,1999, submittal conta'ned clarifying information only, and did not change the initial no significant hazards consideration determination.
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2.0 BACKGROUND
AND EVALUATION 2.1 Reactor Pressure Vessel Fluence l
2.1.1. Background j
l The determination of the fluence was based on the results of the dcsimetry analysis for in-vessel surveillance capsule W, which was removed at the end of the previous cycle (the tenth cycle). The analysis report (WOAP 15101, Ravision 0) also included re-analysis of the results of tne three previous capsu'es U, V, and X removed at the end of the first, third and fifth cycles respectively. The analysis was performed using the DORT two-dimensional discrete ordinates program. The analytical methodclogy (the DORT code) and the cross sections used (based on the ENDF/B VI data set)in the evaluation of the capsule dosimetry are those recommended by the staff.
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The associated approx!mations were: P for the expansion of the scattering cross section and 3
S. for the modeling of the angular quadrature. These approximations are recommended by the staff. With DORT the licensee used BUGLE 96, a 47 energy group cross sections library which is based on the ENDF/B-VI data set. Forward as well as adjoint calculations were performed for the determination of the fluence. The plant-specific sources associated with the adjoint calculations were derived from fud cycle design reports and the plant operating history.
The capsule dosimetry for a set of six dosimeters was reported for measured and calculated values.
2.1.2 Evaluation The above analysis complies with staff recommendations and the provisions of Draft Guide (DG) 1053 regarding pressure vessel dosimetry. The staff found the comparison of the measured to the calculated values acceptable. The capsule report went on to derive a "best estimate" value employing the FERRET code. However, FERRET, a least squares averaging code, incorporates built-in covariance factors and has not been reviewed nor approved by the staff. Therefore, we do not account for the FERRET results in this review.
2.1.3 Conclusion The staff reviewed the VCSNS surveillance capsule W analysis report regarding the estimation of projected values of the vessel fluence to 32 effective full-power years (EFPYs). The analytical mahodology meets staff recommendations and the comparison of measured and calculated results are acceptable; therefore, the staff finds the projected fluence values acceptable:
2.2 Low Temperature Overpressure Protection System
<2.2.1 Background The low-teroperature overpressure protection (LTOP) system mitigates overpressure transients at low temperatures sc that the integrity of the reactor coolant pressure boundary is not compromised by exceeding the 10 CFR Part 50, Appendix G, P-T limits under steady-state operating conditions. The VCSNS LTOP system uses the residual haat removal (RHR) system relief valves or a reactor coolant system (RCS) vent with the reactor depressurized to accomplish this funet!cni Tne system is manually enabled by operators. The design basis of VCSNS considers both mass-addition and heat-addition transients for the LTOP system. The limiting mass addition analysis accounts for the injection from one centrifugal charging pump to the water solid RCS with letdown isolated. The heat-addition analysis accounts for heat input j
from the secondary side of the steam generators bto the RCS upon an inadvertent startup of f
one inactive reactor coolant pump (RCP) with a 50*F mismatch between the primary and secondary side temperature and a pressurizer water volume of less than 1,288 cubic feet.
The current Limiting Condition for Operation (LCO) in TS 3.4.9.3," Overpressure Protection Systems," requires that the LTOP system shall be operable with two operable RHR relief valves with a lif ting setpoint of less than or equal to 450 psig, or that the RCS be depressurized with an RCS vent of 2_2.7 square inches. This LCO ;s applicable when any RCS cold leg ten,7erature is s 300*F when the head is on the reactor vessel. Also, the current TS 3.5.3 provided a restriction that allows a maximum of one operable centrifugal charging pumo and no operable
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, safety injection pumps when any RCS cold leg temperature is less than or equal to 300*F. TS
' 33.1.3 provides restrictions to preclude starting an RCP when any RCS cold leg temperature is
~ less than or equal to 300*F unless the pressurizer water volume is less than 1,288 cubic feet and the secondary water temperature of each steam generator is less than 50*F above the RCS cold leg temperatures. These TS restrictions would assure that VCSNS will be operated within the coniiguration assumed in the analysis for LTOP system design. In addition, the plant
' operating procedures will maximize the use of a pressurizer cushion (steam / nitrogen bubble) ;
during perbJs of low temperature operation. These administrative controls would minimize the severity of potential overpressure transients and provide assurance that most translents can be
- terminated by operator actions'before the RHR relief valves are challenged.'
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~ 2.2.2 Evaluation l
2.2.2.1 Enable Temperature I
The LTOP enable temperature is the temperature below which the LTOP system is required to j
' be operable; The licensee establishes an LTOP enable temperature in accordance with the i
requirements in the NRC branch technical position (BTP) RSB 5 2, which uses an enable RCS.
. water temperature corresponding to a metal temperature of at least RTsor + 90*F at the belt line location (1/4t or 3/4t). Therefore, the licensee considers the enable temperature as RTwor +
'90*F + temperature dife.r nce between RCS and metal + Instrument Uncertainties. Using the above equation, the calcuWed LTOP enable temperature is 235.2'F. Since the current enable temperature within the Technical Specifications (300*F) is greater than the required enable temperature of 235.2'F, it remains conservative and need not be changed to support use of the new PT curves applicable to 32 EFPY, Thus, the licensee has concluded that the current LTOP enable temperature of 300'F remains applicable and supports the use of the new reactor coolant system P/T limits up to 32 EFPYs. The staff finds that the current LTOP enable temperature of 300'F is conservative with respect to the enable temperature of 235.2*F allowed by NRC (BTP) RSB 5-2 and is, thelefore, acceptable.
2.2.2.2 LTOP Actuation Setpoint The LTOP is designed to mitigate overpressure transients at low temperatures to prevent exceeding RCS P-T limits at VCSNS. Additionally, since overpressure events most likely occur during isothermal conditions in the RCS the NRC has accepted the use of the RCS P-T limits during steady-state conditions for the design of LTOP. The LTOP actuation setpoint !c the f pressure at which the pressure relief valves will lift, when the LTOP is enabled, to limit the peak
' RCS pressure during a design basis pressurization transient.
VCSNS uses RHR relief valves to provide pressure relief capacity for LTOP. The current RHR relief valve lifting setpoint is 450 psig. The licensee, in a letter dated October 8,1999, provided the results of its analysis in a tabulation which lists RCS temperatures, mLximum reactor vess_el pressures for mass addition and heat adWion transient, and corresponding P T limit j
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- under various temperature cona:tions netween 60*F and 350*F. The data presented in this list
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covered the entire range of temperatures in which the LTOP is required. The licensee's
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- analysis assumes an RHR relief valve lifting setpoint of 468 psig which includes 4 percent margin to cover potential uncertainties associated with the relief valve lifting. The data presented in the licensee's tabulation confirms that the currer.t RHR relief valve setpoints will
, provide adequate protection relative to the new P T limits. Therefore, the licensee propos 3s E
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4-that the current RHR relief valve lifting setpoint of 450 psig remains unchanged and supports the new P-T limits applicable up to 32 EFPYs. The staff finds that the licensee's proposal regarding the LTOP setpoint is acceptable because the setpoints fall within a range considered
. to be. acceptable.
2.2.2.3 RCS Vent Size.
In the event one or both of the RHR relief valves become inoperable for an extended time period as specified in TS 3.4.9.3, the RCS will be depressurized and an RCS vent of greater than or equal to 2.7 square inches will be provided for overpressure protection With the RCS j
depressurized, the licensee's analysis states that:
In the event one of the RHR relief valves becomes inoperable, the RCS will be depressurized with an RCS vent of greater than or equal to 2.7 in to provide a
overpressure protection. Potential vents include open!ng two or more of the pressurizer PORVs, removal of one or more of the pressurizer safety valves, or removal of the pressurizer manway. With the RCS depressurized, the RCPs will be stopped with power removed thereby making the mass addition transient the only credible transient. Assuming the minimum required vent at the top nf the pressurizer, peak RV pressures for the design basis mass addition event are
<250 psig; this is well below the new PT limits. Thus, the current minimum vent size remains adequate for low temperature overpressure protection.
The licensee proposes that the current required vent size remains unchanged to support the new P/T limits applicable up to 32 EFPYs. A vent size of 2.7 square inches is capable of
- mitigating the most limiting low temperature overpressure transient with a peak reactor vessel pressure below 250 psig. Therefore, the staff finds this acceptable, i
2.2.3 Conclusions The staff has re'.iewed the licensee's proposed TS 3.4.9.3 for the LTOP enable temperature, actuation setpoint and vent size. The staff also reviewed the licensee's analyses alated to the propcsed enable temperature of 300*F, actuation setpoint of 450 psig and the vent size of 2.7 square inches as discussed above. The licensee has considered instrument uncertainties in its setpuint calculation. The staff finds that the licensee's analyses were performed in a manner consistent with the approved methodology and that the results of tne analyses cor;servatively demonstrated that the new P-T limits applicable up to 32 EFPYs at VCSNS will l
be adequately protected with these setpoints. Therefore, the staff finds that the proposed TS l
3.4.9.3 with its associated bases regarding LTOP design acceptable.
2.3 Pressure Temperature Limits The NRC has sstablished requirements in 10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The staff evaluates the P-T limit curv9 I
based on the following NRC' regulations and guidance: 10 Ci:R 50, Appendix G; Generic Letter (GL) 8811; GL 92-01, Revision 1; GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) l 1.99, Revision 2 (Rev. 2); and Standard Review Plan (3RP) Section 5.3.2. GL 88-11 advised
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licensees that the staff would use RG 1.99, nev. 2, to review P T limit curves. RG 1.99, Rev. 2, contains methodologies for determining the increase in transition temperature and the decrease
j l in upper-shelf energy (USE) resulting from neutron radiation. GL 92-01, Rev.1, requested that licensees submit their reactor pressure vessel (RPV) data for their plants to the staff for review.
GL 92-01, Rev.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 staff as the basis for the staff's review of P-T 'imit curves and as the basis for the staff's 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 (ASME) Boiler and Pressure Vessel Code (Code).
SRP 5.3.2 provides an acceptable method of determining the P-T limit curves for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor, K,, wh;ch is a function of the stress state and flaw configuration. Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions; for hydrostatic testing curves, Appendix G requires a safety factor of 1.5.
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 1/4 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 curves are the 1 A thickness (1/4T) and 3/4 thickness (3/4T) locations, which corresponc the d6pth of the maximum postulated flaw, if initiated and grown from the inside and outside surfaces of the RPV. respectively.
The Appendix G ASME Code methodology requires that licensees determine the adjusted reference temperature (ART or adjusted RTno7). The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTsor), the mean value of the adjustment in reference temperature caused by irradiation (ARTno7), and a margin (M) term.
The ARTor is a product of a chemistry factor and a fluence factor. The chemistry factor is depenrient upon the amount of copper and nickelin 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 RTsor is a plant-specific or a generic value and whether the chamistry f actor
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(CF) was determined using the tablos in RG 1.99, Rev. 2,'or surveillance data. The margin 1
l term is used to account for uncertainties in the values of the initial RTsor, the copper and nickel contents, the fluence and the calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term.
As detailed in the staff evaluation below, the licensee's P-T lim;t curves satisfy the requirements
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of 10 CFR 50.60(a), with the additional provisions allowed by ASME Code Case N-640. This Code case allows the P i wa cuMs tu ce developed using the K,c fracture toughness curve of ASME Section XI, Appendix A,instead of the Km curve of Appendix G, as authorized ar.d explained in the V. C. Summer exemption from 10 CFR Part 50, Appendix G requirements to be used for generation of pressure-temperature limits curves which was granted October 20,1999.
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2.3.1 Licensee Evaluation As stated above, the licensee submitted ART calculations and P-T limit curves valid for up to 32 EFPYs. For the V. C. Summer reactor vessel, the licensee determined that the most limiting material at the 1/4T and 3/4T locations is in the lower shell plate that was f abricated using plate heat number C9923-1,2. The licensee calculated an ART of 107'F at the 1/4T location and 94*F at the 3/4T location at 32 EFPYs. The neutron fluence used in the ART calculation is 2.41 X 10" n/cm at the 1/4T location and 9.52 X 10* n/cm at the 3/4T location. The ARTnor 2
2 values at the 1/4T and 3/4T locations are 63.2*F and 50.3'F, respectively. The initial RTnor for the limiting plate is 10*F. The margin term used in calculating the ART for the limitirig plate is 34*F at the 1/4T and 3/4T locations, as permitted by Position 1.1 of RG 1.99, Revision 2. The licensee's limiting ART for the vessel flange, head flange, and upper shell plate and weld
. material is 10*F.
2.3.2 Staff Evaluation The staff independently calculated the ARTS using the staff reviewed and approved data and calculations of the NRC Reactor VesselIntegrity Database (RVID). In addition, the staff independently generated P-T limit curves for normal operations and inservice hydrostatic testing condition effective to 32 EFPYs for V. C. Summer. Although the staff's calculations using the NRC approved data and methodology differed slightly from the licensee's data and j
methodology, the licensee's P-T limit curves were found to be conservative with respect to the staff's determinations, and therefore, acceptable. The details of this evaluatity are provided below.
The ART was det: mined using the chemistry values of percent copper and percent nickel for each beltline material of V. C. Summer. The RVID contains chemistry values for each beltline inaterla! for all light water reactors in the U.S. The licensee's and the vendor's data were verified by the staff before incorporation into the RVID. It should be notcm ; hat the staff used the most recent updated chemistry data for the beltline materials in the e"aluation of the V. C.
Summer P-T limit curves. The staff independently verified the data for the chemical compositiona, initial RTnor, fluence, and margin values for V. C. Summer. In addition, the staff found that the calculated values, as proposed in SCE&G's submittal, were at least as
. conservative as those values derived by the staff.
l The. staff also performed an independent evaluation of the surveillance data for V. C. Summer.
In calculating the P-T limits, the margin term was 34'F. This value was calculated from the sigma delta value, which in this case was 17'F; the 2-sigma delta value would be 34'F. The
. staff verified that the measured ARTnorvalues from all of the surveillance data were within 2-standard deviations of the predicted ARTuorvalues. Therefore, the surveillance data indicates that the predicted AF.Tso1 values and margin term f rom RG 1.99, Rev. 2, are appropriate.
The staff evaluated each of the licensee's P-T limit curves for acceptability by performing independent calculations, using the methodology referenced in the Code (as indicated by SRP 5.3.2), and verified that the licensee's proposed P-T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G,10 CFR Part 50. The staff independently generated P T limit curves for normal operations and hydrostatic test pressures effective to 32 EFPYs for V. C.
Summer. In comparing the staff's generated P-T limit curves to the licensee's generated P-T
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. limit curves, the staff determined that the licensee's proposed P-T limit curves for V. C. Summer meet the requirements of Appendix G of Section XI of the A?ME Code, as modified by Code
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Case N-640. Therefore, the staff determined that the licenseG proposed P T limit curves are j
acceptable, since they meet the requirements of 10 CFR 50.60 and Appendix G, 10 CFR Part 50, as modified by Codo Case N-640.
In addition to beltline materials, Appendix G of 10 CFR 50 also imposes a minimum temperature at the RPV 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 region, highly stressed by the bolt preload must exceed the adjusted 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 RT, of 10*F tor the limiting flange and upper i
shell materials, as stated in the RVID, and also confirmed by the licensee, the staff has determined that the proposed P-T limits satisfy the requirement for the closure flange region of Appendix G,10 CFR Part 50, during normal operation and hydrostatic pressure test and leak test for V. C. Summer.
2.3.3 Conclusions The staff concludes that the proposed P-T limits for the RCS 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 50 for 32 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. Hence, the proposed P-T limit curves may be incorporated into the V. C. Summer technical specifications.
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3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the State of Scuth Carolina official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a f acility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no 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 a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (64 FR 48865). Accordingly, the amerMment meets 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 staternent or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
l The Commission has concluded, based on the considerations discussed above, that (1) there is
. reasonable assurt.1ce that the health and safety of the public will not be endangered by L
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, operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissioris 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.
Principal Contributors: Lambros Lois Chu Liang Meena Khanna Date: October 21,1999 l
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' AMENDMENT NO.143 TO FACILITY OPE'RATIN'G LICENSE NO. NPF SUMMER, UNIT NO.1 :
DISTRIBUTION'
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Docket File; A
PUBLIC '
PD 11-1 R/F.
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j R. Emch j
4 E. Dunnington K. Cotton OGC
- . 3 G. Hill (2)i 4
W. Beckner vt
- L. Lois C. Liang M.Khanna ACRS R. Haag, Ril
' R. Scholl (e-mail SE only)'
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