ML041700466
| ML041700466 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 06/16/2004 |
| From: | Kalyanam N NRC/NRR/DLPM/LPD4 |
| To: | Venable J Entergy Operations |
| Kalyanam N,NRR/DLPM,415-1480 | |
| References | |
| TAC MC1156 | |
| Download: ML041700466 (14) | |
Text
June 16, 2004 Mr. Joseph E. Venable Vice President Operations Entergy Operations, Inc.
17265 River Road Killona, LA 70066-0751
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: PRESSURE TEMPERATURE LIMIT CURVES TO 32 EFFECTIVE FULL POWER YEARS WITH POWER UPRATE (TAC NO. MC1156)
Dear Mr. Venable:
The Commission has issued the enclosed Amendment No. 196 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 22, 2003.
The amendment requested extension of the pressure temperature curves from 16 effective full power years (EFPYs) to 32 EFPYs. In addition, the request proposed to change: (1) the maximum heatup and cooldown rates to 60 Fahrenheit/hour (F/hr) and 100 F/hr, respectively, (2) the low temperature overpressure protection enable temperature, and (3) the evaluation of RTPTS.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.
Sincerely,
/RA/
N. Kalyanam, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-382
Enclosures:
- 1. Amendment No. 196 to NPF-38
- 2. Safety Evaluation cc w/encls: See next page
June 16, 2004 Mr. Joseph E. Venable Vice President Operations Entergy Operations, Inc.
17265 River Road Killona, LA 70066-0751
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: PRESSURE TEMPERATURE LIMIT CURVES TO 32 EFFECTIVE FULL POWER YEARS WITH POWER UPRATE (TAC NO. MC1156)
Dear Mr. Venable:
The Commission has issued the enclosed Amendment No. 196 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 22, 2003.
The amendment requested extension of the pressure temperature curves from 16 effective full power years (EFPYs) to 32 EFPYs. In addition, the request proposed to change: (1) the maximum heatup and cooldown rates to 60 Fahrenheit/hour (F/hr) and 100 F/hr, respectively, (2) the low temperature overpressure protection enable temperature, and (3) the evaluation of RTPTS.
A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.
Sincerely,
/RA/
N. Kalyanam, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-382
Enclosures:
- 1. Amendment No. 196 to NPF-38
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC PDIV-1 Reading RidsNrrDlpmLdiv (HBerkow)
RidsNrrDlpmLpdiv1 (RGramm) RidsNrrPMNKalyanam RidsNrrLADJohnson RidsOgcRp RidsAcrsAcnwMailCenter GHill (2)
TBoyce LLois/JUhle TMcLellan/MMitchell RidsRgn4MailCenter (AHowell) RidsNrrDlpmDpr (SWall)
RidsNrrLADBaxley RidsNrrPMAMcMyrtray Accession No.:ML041700466 *Staff provided SE Wed. with minor Editorial changes OFFICE PDIV-1/PM PDIV-1/LA DSSA/SRXB*
DE/EMCB DIPM/IROB OGC PDIV-1/SC NAME NKalyanam DBaxley LLois/JUhle TMcLellan/
MMitchell TBoyce for KKavanagh TSmith(NLO)
RGramm DATE 6/1/04 6/10/04 12/15/03 6/2/04 6/7/04 6/8/04 6/10/04 OFFICIAL RECORD COPY
ENTERGY OPERATIONS, INC.
DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 196 License No. NPF-38 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (EOI) dated October 22, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2. of Facility Operating License No. NPF-38 is hereby amended to read as follows:
2.
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 16, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 196 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert XIX XIX 3/4 4-3 3/4 4-3 3/4 4-5 3/4 4-5 3/4 4-5a 3/4 4-28 3/4 4-28 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-34 3/4 4-34
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 196 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.
WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382
1.0 INTRODUCTION
Technical specifications (TSs) include limiting conditions for operation (LCOs) that establish pressure/temperature (P/T) and low temperature overpressure (LTOP) system limits for the reactor coolant system (RCS). The limits are defined by figures and values that provide an acceptable range of operating temperatures and pressures for heatup, cooldown, LTOP, criticality, and inservice leak and hydrostatic testing conditions. These parameters are valid for a specified number of effective full-power years (EFPYs) or for a specified fluence period (the period for which the curves are calculated to be applicable). License amendments are required at the end of the effective period for P/T limit curves or when surveillance specimens are withdrawn and tested. Also, each time the P/T curves are revised, the LTOP system is reevaluated to ensure that its functional requirements can still be met.
By application dated October 22, 2003 (ADAMS Accession No. ML041620063), Entergy Operations, Inc. (the licensee), requested changes to the TSs for Waterford Steam Electric Station, Unit 3 (Waterford 3). The licensee, requested changes to TSs 3.4.1.3, Reactor Coolant System, Hot Shutdown; 3.4.1.4, Reactor Coolant System, Cold Shutdown - Loops Filled; 3.4.8.1, Pressure/Temperature Limits; and 3.4.8.3, Overpressue Protection Systems.
Specifically, the existing P/T limits are changed from 16 to 32 EFPYs. In addition, the maximum heatup rate is changed to 60 ofahrenheit per hour (oF/hr) and the maximum cooldown rate is changed to 100 oF/hr for all RCS temperatures.
1.1 Index Page, XIX The description for figures 3.4-2 and 3.4-3 in the index page are revised to reflect the new figures.
1.2 TS 3.4.1.3, Mode 4, Reactor Coolant and/or Shutdown Cooling Loops and TS 3.4.1.4, Mode 5 with reactor coolant loops filled, Reactor Coolant and/or Shutdown Cooling Loop The second footnote ** states in part "A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 272 F unless..." The reference to 272 F is changed to 200 F to be consistent with the new analysis associated with the LTOP system enable temperature.
The associated TS Bases are revised to reflect the lower temperature.
1.3 TS 3.4.8.1, Pressure/Temperature Limits The maximum heatup and cooldown rates that are included in the LCO items "a" and "f are changed. The proposed changes include deleting existing LCO items b, c, e, and "f. Item a," which will specify a maximum heatup rate of 60 F/hr and item "d" is now renumbered as b that will specify a maximum cooldown rate of 100 F/hr.
The TS Bases for TS 3.4.8.1 is modified to reflect the new P/T limit curves and analyses. Additionally, Table B 3/4 4-1, Reactor Vessel Fracture Toughness, is removed.
1.4 TS Figures 3.4-2 and 3.4-3 TS Figures 3.4-2 and 3.4-3 are revised to reflect the new P/T limit curves valid to 32 EFPY.
1.5 TS 3.4.8.3, Overpressure Protection Systems The APPLICABILITY is changed. Currently the TS is applicable in MODE 4 when the temperature of any RCS cold leg is less than or equal to 272 F, in MODE 5, and in MODE 6 when the head is on the reactor vessel and the RCS is not vented through a 5.6 square inch or larger vent. Based on the analyses, the LTOP system enable temperature is changed to 200 F.
The footnote # in the APPLICABILITY section which states... 260 F during inservice leak and hydrostatic testing with Reactor Coolant System temperature changes restricted in accordance with Specification 3.4.1g., is deleted.
2.0 REGULATORY EVALUATION
2.1 Fluence Calculations The staff issued Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, in March 2001. The RG provides guidance regarding acceptable methods for the benchmarking of vessel fluence methodologies based on the requirements of General Design Criterion (GDC) 31 and in part GDC 14 and GDC 30.
Therefore, the staff is basing the review of the peak vessel fluence evaluation for Waterford 3 on the adherence of the calculational method to the guidance in RG 1.190.
The irradiated material properties corresponding to the fluence estimated using RG 1.190 are used to calculate the P/T curves which also define the LTOP enable temperature which must satisfy the requirements of Appendix G to Title 10 of the Code of Federal Regulations (10 CFR)
Part 50. The value for RTPTS corresponding to 32 EFPYs must satisfy the requirements of 10 CFR 50.61.
2.2 Revision to P/T and LTOP limits for 32 EFPYs The Nuclear Regulatory Commission (NRC) has established requirements in 10 CFR Part 50 to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The staff evaluates P/T limit curves based on the following NRC regulations and guidance:
Appendix G to 10 CFR Part 50. This requires that P/T limit curves for the reactor pressure vessel (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).
Generic Letter (GL) 88-11. GL 88-11 advised licensees that the staff would use RG 1.99, Revision (Rev.) 2 to review P/T limit curves.
RG 1.99, Rev. 2. This contains methodologies for determining the increase ductile-to-brittle in transition temperature and the decrease in upper-shelf energy resulting from neutron irradiation.
GL 92-01, Rev. 1 and GL 92-01, Rev. 1, Supplement 1. GL 92-01, Rev. 1 requested that licensees submit their 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 review of P/T limit curves.
Standard Review Plan (SRP) Section 5.3.2, which 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 methodology of Appendix G to Section XI of the ASME Code.
The basic parameter of the methodology of Appendix G to Section XI of the ASME Code is the stress intensity factor (KI), which is a function of the stress state and flaw configuration.
Appendix G to Section XI of the ASME Code requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on stress intensities resulting from hydrostatic testing. Appendix G to Section XI of the ASME Code also requires a safety factor of 1.0 on stress intensities resulting from thermal loads for normal and transient operating conditions as well as for hydrostatic testing.
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 six times its depth. The critical locations in the RPV beltline region for calculating heatup and cooldown P/T limit curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
The methodology found in Appendix G to Section XI of the ASME Code requires that licensees determine the adjusted reference temperature (ART or adjusted RTNDT) at the 1/4T and 3/4T locations. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the shift in reference temperature caused by irradiation
( RTNDT), and a margin term. Guidance on the determination of RTNDT and the margin term is given in RG 1.99, Rev. 2. RTNDT is a product of a chemistry factor (CF) and a fluence factor (FF). The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Rev. 2, or from surveillance data. The FF is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF 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 the initial RTNDT, the copper and nickel contents, the fluence, and the calculation procedures.
The licensee stated that it was invoking Code Case N-641, which apply to P/T limit curves development, and the provisions of the Code Case have been incorporated into the 1998 Edition of the ASME Code endorsed in 10 CFR 50.55a. Therefore, no request for use of Code Case N-641 is needed from the licensee.
3.0 TECHNICAL EVALUATION
3.1.1 Fluence Calculations The results of the testing and analysis of the second surveillance capsule are documented in WCAP-16002NP (Reference 1). The Appendix G related analyses and a summary of the capsule fluence results are documented in WCAP-16088NP, Rev. 1 (Reference 2).
Capsule 263 was removed at the end of Cycle 11 at 13.83 EFPYs. All of the neutron transport calculations and dosimetry evaluations were derived from the ENDF/B-VI cross section data file. The transport calculations were performed using the DORT discrete ordinates code version 3.1 (Reference 3) and the BUGLE-96 (Reference 4) cross sections. In these analyses, the anisotropic scattering was treated with a P3 Legendre approximation and the angular discretization was modeled with an S16 angular quadrature. Neutron source distribution and system operating temperatures were treated on a cycle specific basis. All of the above conditions satisfy the guidance in RG 1.190 and are acceptable.
In addition, the measurement uncertainty recapture power uprate of 1.5 percent which was implemented at the beginning of Cycle 12 (3390 megawatts thermal (MWt) to 3441 MWt), and an eight percent extended power uprate (EPU) (3441 MWt to 3716 MWt) scheduled for Cycle 14, were accounted for in the recalculation of the neutron sources. The method of calculation was conservative and followed the guidance in RG 1.190, and is acceptable.
3.1.2 Neutron Dosimetry Cu-63, Fe-54, and Ni-58 dosimeters were analyzed. Direct comparison of measured to calculated specific activities demonstrated that the measured to calculated ratios are well within the 20 percent guidance in the RG. This confirms the result of the calculation for the peak vessel fluence at the midplane for 32 EFPYs of 2.48x1019 neutrons per square centimeter (n/cm2) which is used for the P/T limit curves.
3.2 Revision to P/T and LTOP limits for 32 EFPYs 3.2.1 Licensees Evaluation The licensee submitted ART values and P/T limit curves valid for up to 32 EFPY of facility operation and identified the limiting material for the Waterford 3 RPV as lower shell plate M-1004-2, fabricated from plate heat 57286-1. The licensee calculated the ART values for lower shell plate M-1004-2 for both the 1/4T and 3/4T locations. The critical parameters for the licensees ART determination for each of these locations are shown in the table below.
Material Location Initial RTNDT (in F)
Fluence (in n/cm2)
Chemistry Factor (1)
(in F)
RTNDT (in F)
Margin (2)
(in F)
ART (inF)
Lower Shell Plate M-1004-2 1/4T 22 1.48 x 1019 12.4 14 14
(
I = 0,
= 7) 50 Lower Shell Plate M-1004-2 3/4T 22 0.525 x 1019 12.4 10.1 10.1
(
I = 0,
= 5.05) 42 (1) Credible surveillance specimen test data is available for lower shell plate M-1004-2 from the licensees RPV surveillance program. The chemistry factor for this plate was determined from that surveillance data using Regulatory Guide 1.99, Rev. 2, Position 2.1.
(2) The margin term for each ART calculation was based on the establishment of initial material property uncertainty
(
I) and shift in material property uncertainty (
) consistent with the guidance in Regulatory Guide 1.99, Rev. 2.
Regarding the detailed fracture mechanics evaluation performed to establish the proposed Waterford 3 P/T limits, the licensee submitted information on the through wall temperature gradients resulting from heatup and cooldown transients and their determination of the applied KI at the tip of the postulated 1/4T and 3/4T flaws due to thermal loading in an enclosure to its October 22, 2003, letter. This information, along with knowledge of the applied KI at the tip of the postulated 1/4T and 3/4T flaws due to pressure loads and the material property information cited above, permitted the staff to evaluate the acceptability of the proposed Waterford 3 P/T limit curves.
3.2.2
NRC Staff Evaluation
The NRC 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 NRC staff verified that the licensee's limiting material for the RPV is the lower shell plate M-1004-2. The NRC staffs calculated ART values for the limiting material at the 1/4T and 3/4T locations agreed with licensee's calculated values.
The NRC staff evaluated the licensee's P/T limit curves for acceptability by performing a finite set of check calculations using the methodology referenced in the ASME Code (as indicated by SRP 5.3.2) based on information submitted by the licensee. Further, the NRC staff compared information submitted by the licensee (in particular, the information related to the evaluation of thermal loading conditions) to information submitted previously for other similar RPVs, and determined that the information submitted by the licensee for Waterford 3 was consistent. The NRC staff verified that the licensee's proposed P/T limits satisfy the requirements in Paragraph IV.A.2 of Appendix G of 10 CFR Part 50. Specifically, the NRC staff concluded that the P/T limit curves submitted by the licensee were as conservative as those which would be generated by the NRC staff's application of the methodology specified in Appendix G to Section XI from the 1998 Edition through 2000 Addenda of the ASME Code. Therefore, the NRC staff determined that the licensees proposed P/T limit curves were acceptable for operation of the Waterford 3 RPV though 32 EFPY of operation.
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 most limiting reference temperature for the flange material.Section IV.A.2 of Appendix G states that when the pressure exceeds 20 percent of the pre-service system hydrostatic test pressure, the temperature of the closure flange regions which are 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 limiting flange RTNDT of 20 F for Waterford 3 (which was verified by information in the NRC staffs Reactor Vessel Integrity database), the NRC staff has determined that the proposed P/T limits have satisfied the requirement for the closure flange region during normal operation and inservice leak and hydrostatic testing.
The licensee developed its P/T curves using Code Case N-641 which permits the postulation of a circumferentially-oriented flaw for the evaluation of the circumferential welds and permits the use of an alternate reference fracture toughness (the KIC) fracture toughness curve (instead of the KIa fracture toughness curve) for reactor vessel materials in determining the P/T limit curves. The use of a circumferentially-oriented flaw for the evaluation of a circumferential weld is justified because it would represent the orientation of any significant flaws which may occur as a result of the fabrication process.
The use of the KIC fracture toughness curve is appropriate for the evaluating the potential for crack initiation without imposing unnecessary conservatism. The KIC curve appropriately implements the use of static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a RPV. The NRC staff concluded that P/T limit curves based on the KIC fracture toughness curve referenced by ASME Code Case N-641 will enhance overall plant safety by opening the P/T operating window with the greatest net safety benefit in the region of low-temperature operation. In addition, implementation of the proposed P/T curves, as allowed by ASME Code Case N-641, maintains appropriate margins of safety against brittle failure of the RPV as required by Appendix G to 10 CFR Part 50.
The description for Figures 3.4-2 and 3.4-3 in the index page XIX are revised to reflect the new figures. This change is administrative in nature and the staff finds it acceptable.
The Waterford 3 RPV has a very low copper content, which makes it insensitive to radiation embrittlement as measured by the shift in the ductile-to-brittle temperature.
The calculated enable temperature is lower than the minimum temperature of 200 F required by RG 1.99. The reference to 272 F is changed to 200 F in TS 3.4.1.3 and TS 3.4.1.4 and is found to be consistent with the new analysis associated with the LTOP system enable temperature, and therefore is acceptable.
If a hydrostatic test were required, the LTOP system cannot be in service. The lowest temperature at which a hydrostatic test can be performed is 190 F; the LTOP system can be isolated above 200 F, which allows a 60 F margin to perform the test prior to reaching the TS 3.4.9, ACTION "a" limit. Instrument uncertainty is not included in the enable temperature. Amendment No. 189 dated September 22, 2003, approved the relocation of TS 3.4.9 to the Technical Requirements Manual.
The proposed TS changes include the term of applicability of the new P/T limits curves of 32 EFPYs, the revised maximum heatup of 60 F/hr and cooldown of 100 F/hr, and the new LTOP systems enable temperature of 200 F. These changes correctly reflect the proposed TS changes, and thus, are acceptable.
Note # was required to allow performance of the hydrostatic test to prove restoration of the structural integrity of any ASME Code Class 1 component (TS 3.4.9, Structural Integrity, Action a). With the reduction in the LTOP enable temperature, the note is no longer needed. Based on the proposed P/T curves, in order to satisfy TS 3.4.9, ACTION a, the structural integrity of the ASME Code Class 1 components must be restored prior to reaching 260 F (based on the lowest service temperature of 190 F plus the TS Action requirement of 70 F), therefore deletion of Note # is acceptable.
Changes to the Bases portion of the TSs are consistent with the changes made to the TS. This does not require either a review or acceptance by the NRC staff.
The level of fracture toughness detail in Table B 3/4 4-1 is contained in the summary analysis of the surveillance capsule (Reference 1) that was provided to the NRC by letter dated March 28, 2003. Therefore, the level of detail is excessive and need not be repeated in the Bases portion of the TS. Further, it is not consistent with the level of detail contained in the Bases of NUREG-1432, Rev. 2, Standard Technical Specifications Combustion Engineering Plants. This does not require either a review or acceptance by the NRC staff.
In addition to the licensees request to its revision of the P/T limit curves, in a separate letter dated October 13, 2003, the licensee submitted License Amendment Request No. NPF-38-249 for Waterford 3. In this license amendment, the licensee proposed to increase the rated core thermal power level 3441 MWt to 3716 MWt. This EPU represents an 8 percent increase over the present licensed power level. The NRC staffs safety evaluation for the proposed 8 percent increase over the present licensed power level for Waterford Unit 3 will be issued under the scope of the NRC staffs review of License Amendment Request No. NPF-38-249, which is being separately evaluated.
3.2.2 Summary The staff concludes that the proposed P/T limit curves for the pressure test, core not critical, and core critical conditions satisfy the requirements in Appendix G to 10 CFR Part 50 and Appendix G to Section XI of the ASME Code. Therefore, the proposed P/T limit curves are approved for incorporation into the Waterford 3 TSs and shall be valid until 32 EFPY of facility operation.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 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. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (68 FR 68667 dated December 9, 2003). 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.
6.0 CONCLUSION
The Commission has concluded, based on the considerations 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 Commissions 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.
7.0 REFERENCES
1.
WCAP-16002 NP, Analysis of Capsule 263 from the Entergy Operations Waterford Unit 3 Reactor Vessel Radiation Surveillance Program by S.T. Byrne, et al Westinghouse Electric Company LLC, March 2003.
2.
WCAP-16088 NP, Rev. 1, Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation, by S.T.Byrne et al., Westinghouse Electric Company LLC, September 2004.
3.
DOORS 3.1 One, Two-and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System Radiation Safety Information Computational Center (RSICC)
Computer Code Collection CCC-650, Oak Ridge National Laboratory, Oak Ridge Tennessee, August 1996.
4.
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 Radiation Safety Information Computation Center (RSICC) Data Library Collection (DLC) 185, Oak Ridge National Laboratory, Oak Ridge Tennessee, March 1996.
Principal Contributors: L. Lois T. McLellan Date: June 16, 2004
July 2003 Waterford Steam Electric Station, Unit 3 cc:
Mr. Michael E. Henry, State Liaison Officer Department of Environmental Quality Permits Division P.O. Box 4313 Baton Rouge, Louisiana 70821-4313 Vice President, Operations Support Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Director Nuclear Safety Assurance Entergy Operations, Inc.
17265 River Road Killona, LA 70066-0751 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 General Manager Plant Operations Waterford 3 SES Entergy Operations, Inc.
17265 River Road Killona, LA 70066-0751 Licensing Manager Entergy Operations, Inc.
17265 River Road Killona, LA 70066-0751 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Resident Inspector/Waterford NPS P. O. Box 822 Killona, LA 70066-0751 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Parish President Council St. Charles Parish P. O. Box 302 Hahnville, LA 70057 Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Chairman Louisiana Public Services Commission P. O. Box 91154 Baton Rouge, LA 70825-1697