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| {{#Wiki_filter:Pilgrim Nuclear Power Station License Renewal Application Technical Information APPENDIX A UPDATED FINAL SAFETY ANALYSIS REPORT SUPPLEMENT TABLE OF CONTENTS A.O INTRODUCTION .A-1 A.1 CHANGES TO EXISTING UFSAR INFORMATION .A-2 A.1.1 UFSAR Chapter 3 Changes ..A-2 A.1.2 UFSAR Chapter 4 Changes ......................... | | {{#Wiki_filter:}} |
| A-3 A. 1.3 UFSAR Chapter 6 Changes ...... ..............
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| A-7 A.1.4 UFSAR Chapter 7 Changes .......................
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| A-7 A.1.5 UFSAR Appendix C Changes .. A-8 A.1.6 UFSAR Appendix M Changes ..A-12 A.2 NEW UFSAR SECTION .................
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| A-13 A.2.0 Supplement for Renewed Operating License ..A-1 3 A.2.1 Aging Management Programs and Activities
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| I ............
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| A-13 A.2.1.1 Boraflex Monitoring Program ............................
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| A-13 A.2.1.2 Buried Piping and Tanks Inspection Program ...............
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| A-14 A.2.1.3 BWR CRD-Return Line Nozzle Program .A-14 A.2.1.4 BWR Feedwater Nozzle Program .A-14 A.2.1.5 BWR Penetrations Program .A-14 A.2.1.6 BWR Stress Corrosion Cracking Program. ..........
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| I ....... A-15 A.2.1.7 BWR Vessel ID Attachment Welds Program .A-15 A.2.1.8 BWR Vessel Internals Program .......................
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| A-15 A.2.1.9 Containment Leak Rate Program .A-15 A.2.1.10 Diesel Fuel Monitoring Program .A-16 A.2.1.11 Environmental Qualification (EQ) of Electric Components Program A-16 A.2.1.12 Fatigue Monitoring Program ....................
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| ... A-16 A.2.1.13 Fire Protection Program. ..........................
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| A-16 A.2.1.14 Fire Water System Program .............................
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| A-17 Appendix A Updated Final Safety Analysis Report Supplement Page A-1 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.15 Flow-Accelerated Corrosion Program ....... ...............
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| A-17 A.2. 1.16 Heat Exchanger Monitoring Program ................
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| A-1 7 A.2.1.17 Inservice Inspection
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| -Containment Inservice Inspection (CII)Program A-18 A.2.1.18 Inservice Inspection
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| -Inservice Inspection (ISI) Program ....... A-18 A.2.1.19 Instrument Air Quality Program ...... ....................
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| A-18 A.2.1.20 Metal-Enclosed Bus Inspection Program ....................
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| A-19 A.2.1.21 Non-EQ Inaccessible Medium-Voltage Cable Program ' ...... A-19 A.2.1.22 Non-EQ Instrumentation Circuits Test Review Program....
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| A-1 9 A.2.1.23 Non-EQ Insulated Cables and Connections Program .... ...... A-20 A.2.1.24 Oil Analysis Program ........ ............................
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| A-20 A.2.1.25 One-Time Inspection Program ..... I........................
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| A-20 A.2.1.26 Periodic Surveillance and Preventive Maintenance Program ...A-21 A.2.1.27 Reactor Head Closure Studs Program ............... -A-22 A.2.1.28 Reactor Vessel Surveillance Program .. .A-22 A.2.1.29 Selective Leaching Program .A-23 A.2.1.30 Service Water Integrity Program .A-23 A.2.1.31 Structures Monitoring
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| -Masonry Wall Program .A-23 A.2.1.32 Structures Monitoring
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| -Structures Monitoring Program .A-23 A.2.1.33 Structures Monitoring
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| -Water Control Structures Monitoring Program A-24 A.2.1.34 System Walkdown Program ............................
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| A-24 A.2.1.35 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program A-24 A.2.1.36 Water Chemistry Control -Auxiliary Systems Program .........
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| A-24 A.2.1.37 Water Chemistry Control -BWR Program ..... ............
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| A-24 A.2.1.38 Water Chemistry Control -Closed Cooling Water Program ..... A-25 A.2.2 Evaluation of Time-Limited Aging Analyses .................
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| A-26 A.2.2.1 Reactor Vessel Neutron Embrittlement
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| ....................
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| A-26 A.2.2.1.1 Reactor Vessel Fluence .. .........
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| A-26 A.2.2.1.2 Pressure-Temperature Limits .........................
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| A-26 A.2.2.1.3 Charpy Upper-Shelf Energy .....................
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| I...... A-26 Appendix A Updated Final Safety Analysis Report Supplement Page A-2 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.2.1.4 Adjusted Reference Temperature
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| ......................
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| A-27 A.2.2.1.5 Reactor Vessel Circumferential Weld Inspection Relief ..... A-27 A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability
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| ..... ....... A-28 A.2.2.2 Metal Fatigue ...............
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| ..........................
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| A-28 A.2.2.2.1 Class I Metal Fatigue ...............................
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| A-28 A.2.2.2.2 Non-Class 1 Metal Fatigue ...........................
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| A-29 A.2.2.2.3 Environmental Effects on Fatigue ......................
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| A-29 A.2.2.3 Environmental Qualification of Electrical Components .A-30 A.2.2.4 Fatigue of Primary Containment, Attached Piping, and Components A-30 A.2.2.5 Vessel ID Attachment Welds Fatigue Analysis .A-30 A.2.2.6 Instrument Penetrations Fatigue Analysis ..................
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| A-30 A.2.3 References
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| ..............
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| 4 A-31 Appendix A Updated Final Safety Analysis Report Supplement Page A-3 Pilgrim Nuclear Power Station License Renewal Application Sk Technical Information A.O INTRODUCTION This appendix provides the information to be submitted in an Updated Final Safety Analysis Report Supplement as required by 10 CFR 54.21 (d) for the Pilgrim Nuclear Power Station (PNPS) License Renewal Application (LRA). The LRA contains the technical information required by 10 CFR 54.21(a) and (c). Appendix B of the PNPS LRA provides descriptions of the programs and activities that manage the effects of aging for the period of extended operation.
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| Section 4 of the LRA documents the evaluations of time-limited aging analyses for the period of extended operation.
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| Appendix B and Section 4 have been used to prepare the program and activity descriptions for the PNPS Updated Final Safety Analysis Report (UFSAR) Supplement information in this appendix.This appendix is divided into two parts. The first part identifies changes to the existing sections of the UFSAR related to license renewal. The second part provides new information to be incorporated into the UFSAR. The information presented in both parts will be incorporated into the UFSAR following issuance of the renewed operating license. Upon inclusion of the UFSAR Supplement in the PNPS UFSAR, future changes to the descriptions of the programs and activities will be made in accordance with 10 CFR 50.59.Appendix A Updated Final Safety Analysis Report Supplement Page A-1 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.1 CHANGES TO EXISTING UFSAR INFORMATION This section identifies changes to existing sections of the UFSAR that reflect a renewed operating license. Proposed text deletions are indicated by a strike-through and proposed text additions are indicated by underline.
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| A.1.1 UFSAR Chanter 3 Changes Section 3.3.4.4 Jet Pump Assemblies (6th paragraph)
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| Beams reflecting these design changes are not expected to crack for more than 40 yFs efsevee.Section 3.3.6.8 -Thermal Shock (beginning in the 2nd paragraph)
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| The locations are as follows: 1. Shroud support plate 2, Shroud to shroud support plate discontinuity 3& Shroud inner surface at highest irradiation Fon The peak strain resulting in the shroud support plate is about 6.5 percent. This strain is higher than the 5.0 percent strain permitted by the ASME Code, Section III, for 10 cycles, but the 1 cycle, peak strain corresponds to about 6 allowable cycles of an extended ASME Code curve as applied to less than 10 cycles.Figure 3.3-9 illustrates both the ASME Code curve and the basic material curves from which it was established (with the safety factor of 2 on strain or 20 on cycles, whichever is more conservative).
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| The extension of the ASME Code curve represents a similar criteria to that used in the ASME Code, Section III, but applied to fewer than 10 cycles of loading. For this Type 304 stainless steel material, a 10 percent peak strain corresponds to one allowable cycle of loading. Even a 10 percent strain for a single cycle loading represents a very conservative suggested limit because this has a large safety margin below the point at which even minor cracking is expected to begin.Because the conditions which lead to the calculated peak strain of 6.5 percent are not expected to occur even once during the entire reactor lifetime, the peak strain is considered tolerable.
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| Appendix A Updated Final Safety Analysis Report Supplement Page A-2 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 2. Shroud to shroud support plate discontinuity The results of the analysis of the shroud to shroud support plate discontinuity region are as follows: Amplitude of alternating stress. 180,000 psi Peak strain. 1.34 percent The ASME Code, Section 1I1, allows 220 cycles of this loading, thus no significant deformations result.0 &Shroud inner surfaces at highest irradiation zone The most irradiated point on the inner surface of the shroud is subjected to an estimated total integrated neutron flux of 241.84 x 10291 nvt (>1 MeV) by-the-endof-station lifefor 60 years (54 EFPY) of operation.
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| The peak thermal shock stress is 155,700 psi, corresponding to a peak strain of 0.57 percent. The shroud material is Type 304 stainless steel, which is not significantly affected by irradiation.
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| The material does experience a loss in reduction of area. Because reduction of area is the property which determines tolerable local strain, irradiation effects can be neglected.
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| The peak strain resulting from thermal shock at the inside of the shroud represents no loss of integrity of the reactor vessel inner volume. The service limit of Type 304 stainless steel is approached at a fluence of 8 x 1 0 21 n/cm 2 (BWRVIP-35).
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| As the PNPS shroud.will remain below that fluence level for the Deriod of extended operation.
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| the shroud will remain serviceable.
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| A.1.2 UFSAR Chapter 4 Changes Section 4.2.4 -Power Generation Design Bases 1. The location and design of the external and internal supports provided as an integral part of the reactor vessel shall be such that stresses in the reactor vessel and supports due to reactions at these supports are within ASME Code limits.2. The original reactor vessel design lifetime was shall be 40 years Subsequent evaluation of the vessel determined it acceptable for 60 years (54 EFPY) of operation.
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| Section 4.2.5.1 -Reactor Vessel (1st paragraph)
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| The reactor vessel is a vertical cylindrical pressure vessel with hemispherical heads of welded construction.
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| The reactor vessel was is designed and fabricated for a useful life Appendix A Updated Final Safety Analysis Report Supplement Page A-3 Pilgrim Nuclear Power Station License Renewal Application Technical Information of 40 years based upon the specified design and operating conditions.
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| Subseguent evaluation of the vessel determined it acceptable for 60 years (54 EFPY) of operation.
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| The vessel is designed, fabricated, inspected, tested, and stamped in accordance with the ASME Boiler and Pressure Vessel Code, Section III (1965 Edition and January 1966 addenda), its interpretations, and applicable requirements for Class A Vessels as defined therein. The reactor vessel and its supports are designed in accordance with the loading criteria of Appendix C. The materials used in the design and fabrication of the reactor pressure vessel are shown on Table 4.2-1. Reactor vessel data is shown on Table 4.2-2.(5 th paragraph)
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| Another way of minimizing any changes (clevating)increases to the NDTT is by reducing the integrated neutron exposure at the inner surface of the reactor vessel.Th maximum neutron fluents for this reactor is calculated to be 2.5 x 1018 nt. This numbcr is calculated baed on the assumption of operational design power for 40 yr at 100 perrent availability for neutern cncrgjcF g eroatr than 1 MleV Section 4.2.6 -Reactor Vessel -Safety Evaluation (3 rd paragraph)
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| Stress analysis and load combinations for the reactor vessel have been evaluated for the cycles expected throughout the original 40 year design life, with the conclusion that ASME Code limits are satisfied.
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| The details of assumed loading combinations are described in Appendix C for Class I equipment.
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| (4 th paragraph)
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| The reactor vessel was originally i-designed for a 40=ygar life and will not bo cxposod te-with exposure of not more than 1 x 1019 nvt of neutrons with energies exceeding 1 MeV. Extensive tests have established the magnitude of changes in the NDTT as a function of the integrated neutron dosage. Figure 4.2- 5 presents pertinent test data for SA302B steel and plots the change in ductile to brittle transition temperature as a function of integrated neutron flux (nvt). Because SA533 is the same as 302B, all test data on SA302B is applicable to SA533 used in the vessel. The 30 ft lb refers to the energy absorbed by the Charpy V-Notch sample at the test (transition) temperature.
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| The upper two curves apply to thick walled pressure vessels and the lower curve is for the wall thickness range representative of this reactor vessel. The SA302B steel with the fabrication procedures specified for the reactor vessel is relatively insensitive to neutron irradiation.
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| (6 th paragraph)
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| Appendix A Updated Final Safety Analysis Report Supplement Page Ad4 Pilgrim Nuclear Power Station License Renewal Application Technical Information Surveillance specimens were extracted from the vessel in the 1980 outage and tested by Southwest Research.(2)
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| These results showed a slightly higher than predicted shift in NDT with increasing fluence. The corresponding allowable pressure/temperature boundaries were revised to account for the accelerated shift and projected operation through the 1983 refueling outage. This evaluation (3), which formed the basis for Amendment 82 to Technical Specification 3.6/4.6, relied entirely on the information obtained from Southwest Research Report 02-5951(2) and overestimated the cumulative effects of vessel neutron exposure for subsequent cycles. A more recent and rigorous radiation transport analysis, prepared by General Electric Company (MDE 277-1285)(6) reflects a more refined model of the low-leakage core loading scheme adopted subsequent to Cycle 5 and incorporates technical improvements in determining neutron fluence over the period extending from Cycle 4 through the end ef the present license duration.
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| 32 EFPY of operation.
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| Technical Specification Section 3.6/4.6 has been revised to reflect this correction and Figures 3.6.1/3.6.2, relating RTNDT shift to neutron fluence, were derived from this report. Additionally, Teledyne Engineering Services (TES) was commissioned to prepare a technical report (TR-6052-B-1)(7), supplemented by TES Report TR-7487(9), that provided revised parametric pressure/temperature curves as a function of effective full power years (EFPY) which incorporated the recommendations in Regulatory Guide 1.99, Revision 2, dated May 1988(8).(7th paragraph)
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| The reactor assembly is designed such that the average annular distance from the outermost fuel assemblies to the inner surface of the reactor vessel is approximately 80 cm. This annular volume, which contains the core shroud, the jet pump assemblies, and reactor coolant, serves to attenuate the fast flux incident upon the reactor vessel wall. Assuming plant operation at 1,998 MWt, 80 percent station availability, and 40yar station life, the neutron fluence at the inner surface of the vessel was calculated to be 1.5 x 1018 nvt for neutrons having energies greater than 1 MeV. Initially the "worst case' curve from Figure 4.2-5 would produce an NDTT shift of less than 50 0 F. This figure is retained for historical purposes.
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| With an initial NDTT in the vessel plate material of 40 0 F, the resulting maximum NDTT of the vessel wall at the end of 40 yparp would be less than 90 0 F. This end of life NDTT provides a substantial margin for brittle fracture prevention, since the vessel cannot be pressurized until coolant temperatures in excess of 212 0 F are reached. Vessel operation up to 60 years (54 EFPY) was projected using the methods of Regulatory Guide 1.99. Revision 2. This projection resulted in a maximum fluence to the vessel inner wall of 1.28 x 1018 n/cm2. The limiting CvUSE for the lower shell welds and lower intermediate shell welds remain above the 50 ft-lb minimum required.
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| The lower intermediate shell welds remain limiting for RT;L. with an adiusted RTl of 92.70 F.Appendix A Updated Final Safety Analysis Report Supplement Page A-5 Pilgrim Nuclear Power Station License Renewal Application Technical Information Section 4.2.8.3 -Proposed Limiting Conditions for Initial Plant Operation (Item 4, 2 nd paragraph)
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| The NDTT is defined as the temperature below which ferritic steel breaks in a brittle rather than a ductile manner. Radiation exposure from fast neutrons (>1 MeV) above about 1 0 4.71 nvt may increase the NDTT of the vessel base metal. Extensive tests have established the magnitude of changes in the NDTT as a function of integrated neutron exposure.
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| The initial maximum NDTT of the reactor vessel is not greater than 40 0 F. The original design life of the reactor vessel wajse-40 ykar§ and the maximum fast neutron fluentsce calculated for 40 years was Calculated to be 2.5 x 1 0 4.glfl nvt.The fluence calculated for 60 years (54 EFPY) is still below this estimated bounding value. See Section 4.2.6 for details.Section 4.2.8.4 -Proposed Surveillance Requirements for Initial Plant Operation (Item 1, 4th paragraph)
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| It is not planned that any vessel material, other than that already in the surveillance program described above, will be retained for preparing Charpy V-Notch test specimens for the purpose of additional irradiation monitoring of vessel material, or the monitoring of thermal annealing treatments if required to recover fracture toughness in the later years of vessel service. Refer to the discussion of neutron fluentspe expected during the reactor vessel's 40 yF life in Sections 4.2.5.1 and 4.2.6.Section 4.3.4 -Description (22nd paragraph)
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| The design objective for the recirculation pump casing was i a useful life of 40 years, accounting for corrosion, erosion, and material fatigue. The pump drive motor, impeller, wear rings, and seals are designed for as long a life as is practical.
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| The design provides a unit which should not require removal from the system for rework or overhaul at intervals of less than 5 year_. The pump casing was reviewed for license renewal and loss of material due to corrosion, erosion, and cracking due to material fatigue are managed such that the casing will continue to perform its intended function consistent with the current licensing basis for the period of extended operation.
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| Section 4.6.3 -MSIV Description (2nd paragraph)
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| The design objective for the valve is-was a minimum of 40 yearsQf service at the specified operating conditions.
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| The estimated operating cycles/yhfis were 100 cycles Appendix A Updated Final Safety Analysis Report Supplement Page A-6 Pilgrim Nuclear Power Station License Renewal Application Technical Information during the first year and 50 cycles/year thereafter.
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| In addition to minimum wall thickness required by applicable codes, a corrosion allowance of 0.120 inches minimum is was added. For license renewal. aging management programs were identified, as necessary, to address the effects of aging. including loss of material due to corrosion.
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| for the MSIV through the term of the renewed license. Projected operating cycles through the term of the renewed license are less than the total operating cycles estimated during design.A.1.3 UFSAR Chapter 6 Changes Section 6.4.1 -High Pressure Coolant Injection System (18th paragraph)
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| The system was is designed for an original service life of 40 years, accounting for corrosion, erosion, and material fatigue. The HPCI system was reviewed for license renewal and loss of material due to corrosion.
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| erosion, and cracking due to material fatigue are managed such that the system will continue to perform its intended function consistent with the current licensing basis for the period of extended ooeration.
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| Section 6.5.2.3 -High Pressure Coolant Injection System (HPCIS)(6th paragraph)
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| The HPCIS turbine is designed to accommodate dry and saturated steam. The design objective for the turbine casing Wae s a useful life of 40 years accounting for corrosion, erosion, and material fatigue. The HPCI system was reviewed for license renewal and loss of material due to corrosion, erosion, and cracking due to material fatigue are managed for the period of extended operation.
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| Condensate and moisture carryover are prevented from accumulating by a drain pot and steam traps located immediately upstream of the turbine inlet valve. When the turbine is shutdown, the inlet line is kept at an elevated temperature and the condensate is continuously drained.A.1.4 UFSAR Chapter 7 Changes Section 7.1.6 -Radiation Design Criteria (Control & Instrumentation)
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| : 3. Comparison of the potential exposures which equipment within the primary containment could experience from the design basis LOCA source terms (derived from GE-APED-5756) to the expected original 40-year lifetime exposures shows that the expected 40yr--doses are usually greater than the potential APED accident doses. The safety system equipment specification for components inside the primary containment require that materials used in the component's fabrication are able to withstand a specified total integrated dose which is based upon the component's expected original 40:year lifetime dose plus a LOCA. The Appendix A Updated Final Safety Analysis Report Supplementa" I Page A-7 Pilgrim Nuclear Power Station License Renewal Application Technical Information environmental qualification (EQ) of electrical components program ensures that EQ components are maintained in accordance with their qualification bases.6. The electrical power and control cabling for safety system equipment which must function in a radiation environment is not discussed in Section 14.9, but it has been tested under simulated post accident radiation environment.
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| The cabling has been irradiated with a CO-60 source to a dose of at least 5 X 107 rads which is far in excess of that which safety system cabling inside the primary containment would experience during 40-yf-normal operations plus that which would be experienced over a 30 day period from the release into the primary containment according to the assumptions stated in Chapter 14. The results of the test indicate that the power and control cabling run to Pilgrim Station safety systems is capable of satisfactory performance in a boiling water reactor (BWR) primary containment environment.
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| : 7. The individual components and greases of Limitorque operators have been reviewed by the manufacturer for their ability to withstand the design basis radiation environment; i.e., that experienced during 40 yF ef normal operation plus that radiation which would be experienced resulting from a fission product release into the primary containment according to the assumptions stated in Chapter 14.During that portion of a LOCA in which valve operation would be required.
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| The manufacturer's review indicates that the Limitorque operators are capable of proper operation after irradiation in excess of the design basis radiation environment.
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| In fact, the manufacturer expects proper operation after irradiation up to approximately 1.5 X 108 rads.A.1.5 UFSAR Apiendix C Changes Section C.3.2.2 -Allowable Limits (page C.3-3, second paragraph)
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| The term SF min is defined as the minimum safety factor on load or deflection and is related to the event probability by the following equation: 9 SF min =3 -log 1 0 f where: 10-1 >pfQ > 10-5 Appendix A Updated Final Safety Analysis Report Supplement Page A-8 Pilgrim Nuclear Power Station License Renewal Application Q Technical Information For event probabilities smaller than 10-5 or greater than event probabilities'o0 1 the following apply: 10-1 > Ptw > 106 (SFn = 1.125)1.0 > P§W > 10-1 (SFmin = 2.25)Section C.3.4.1 -Reactor Vessel (2nd paragraph)
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| Stress analysis requirements and load combinations for the reactor vessel have been evaluated for an assumed number of primary loading and cyclic conditions expected throughout the 44-yr-vessel life, with the conclusions that ASME code limits are satisfied.
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| Table C.3-1.Revise table as shown below.TABLE C.3-1 LOADING CONDITION PROBABILITIES P§g4o= §Q40 year event encounter probability Upset (likely) 1.0 > P§_Q40 > l0o-Emergency (low probability) 10-1' > 1 0-3 Faulted (extremely low probability) 10-3 > Pi > 106 Table C.3-6 Revise table heading as shown below.Table C.3-6 MINIMUM SAFETY FACTOR Loading Loads -PfiQ4 SFmin Conditions Revise the table text as follows.Appendix A Updated Final Safety Analysis Report Supplement Page A-9 Pilgrim Nuclear Power Station License Renewal Application Technical Information The minimum safety factor decreases as the event probability diminishes and if the event is too improbable (incredible:
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| Pi < 10-6) then no safety factor is appropriate or required.Table C.3-8 Revise table as shown below.Appendix A Updated Final Safety Analysis Report Supplement Page A-10 Pilgrim Nuclear Power Station License Renewal Application Technical Information TABLE C.3-8 RESULTS OF VESSEL FATIGUE AND STRESS ANALYSIS-
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| ==SUMMARY==
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| OF CRITICAL COMPONENTS Component Galclated Allewable Usage Factor(1)Vessel Shell in Geie 484 80 .RegR 00 2 Closure Studs 202.9 0.07 Closure Flanges- 80 Z 0 Region 0-049 GOottom Head- 71.4 89-0.Support Skirt Junction 0.044 Ghroud Support 1t20-.:7_
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| 60.3 .744 Feedwater Nozzle 7-0. 8" System Transients 0.637 (_Combined Rapid 4-080 and System < ,020 Recirc. Inlet Nozzle go}*2- 474 .01 2 Thermal Sleeve Q.ai CRD Housing to Stub ;62 47.4 76 Tubc Junction Core Spray Nozzle 0.01 Shroud Stabilizer 0.330 RRS Pioina-Looi A 0.110 RRS Ppling-Loop B 0.094 1. Based on the values in 'Pilgrim Reactor Vessel Cyclic Load Analysis." Tech Report 93177-TR-03.
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| August 1994.2- Theze components were justified by a simplified elastic plastic analysis per N 4'117.6 bcauase the prmr4pus secendar; ctress exceeded 38M.2. An additional 20 years of operation will add approximately 0.099 to the usage factor of <0.80 due to rapid thermal cycling.Appendix A Updated Final Safety Analysis Report Supplement Page A-11 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.1.6 UFSAR Apoendix M Changes Section M.1, Introduction to the Report (4th paragraph)
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| A series of exhibits are referenced in the Summary Section. These exhibits present the purchase specifications, inspection report, fabrication test program, summary of tensile tests of special steels, and earthquake analysis of the reactor pressure vessel. These exhibits support the statement made in the opening paragraph.
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| Reactor vessel stresses and analyses are summarized in Section C.3.4.1 of Appendix C. Stress analysis requirements and load combinations for the reactor vessel have been evaluated for the primary loading and cyclic conditions expected throughout the 40 vessel life, with the conclusion that ASME code limits are satisfied.
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| Appendix A Updated Final Safety Analysis Report Supplement P Page A-12 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2 NEW UFSAR SECTION The following information will be integrated into the UFSAR to document aging management programs and activities credited in the PNPS license renewal review and time-limited aging analyses evaluated for the period of extended operation.
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| References to other sections are to UFSAR sections, not to sections in the LRA.A.2.0 Sulolement for Renewed Operatina License The Pilgrim Nuclear Power Station license renewal application (Reference A.2-1) and information in subsequent related correspondence provided sufficient basis for the NRC to make the findings required by 10 CFR 54.29 (Final Safety Evaluation Report) (Reference A.2-2). As required by 10 CFR 54.21(d), this UFSAR supplement contains a summary description of the programs and activities for managing the effects of aging (Section A.2.1) and a description of the evaluation of time-limited aging analyses for the period of extended operation (Section A.2.2). The period of extended operation is the 20 years after the expiration date of the original operating license.A.2.1 Aging Management Proarams and Activities The integrated plant assessment for license renewal identified aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB)for the period of extended operation.
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| This section describes the aging management programs and activities required during the period of extended operation.
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| All aging management programs will be implemented prior to entering the period of extended operation.
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| PNPS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. The Entergy Quality Assurance Program applies to safety-related structures and components.
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| Corrective actions and administrative (document) control for both safety-related and nonsafety-related structures and components are accomplished per the existing PNPS corrective action program and document control program and are applicable to all aging management programs and activities that will be required during the period of extended operation.
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| The confirmation process is part of the corrective action program and includes reviews to assure that proposed actions are adequate, tracking and reporting of open corrective actions, and review of corrective action effectiveness.
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| Any follow-up inspection required by the confirmation process is documented in accordance with the corrective action program.A.2.1.1 Boraflex Monitoring Program The Boraflex Monitoring Program assures that degradation of the Boraflex panels in the spent fuel racks does not compromise the criticality analysis in support of the design of the spent fuel storage racks. The program relies on (1) neutron attenuation testing, (2) determination of boron loss through correlation of silica levels in spent fuel pool Appendix A Updated Final Safety Analysis Report Supplement Page A-11 3 Pilgrim Nuclear Power Station License Renewal Application Technical Information water samples and periodic areal density measurements, and (3) analysis of criticality to assure that the required 5% subcriticality margin is maintained.
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| A.2.1.2 Buried Piping and Tanks Inspection Program The Buried Piping and Tanks Inspection Program includes (a) preventive measures to mitigate corrosion and (b) inspections to manage the effects of corrosion on the pressure-retaining capability of buried carbon steel, stainless steel, and titanium components.
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| Preventive measures are In accordance with standard Industry practice for maintaining external coatings and wrappings.
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| Buried components are inspected when excavated during maintenance.
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| If trending within the corrective action program identifies susceptible locations, the areas with a history of corrosion problems are evaluated for the need for additional inspection, alternate coating, or replacement.
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| A focused inspection will be performed within the first 10 years of the period of extended operation, unless an opportunistic inspection (or an Inspection via a method that allows assessment of pipe condition without excavation) occurs within this ten-year period.A.2.1.3 BWR CRD Return Line Nozzle Program Under the BWR CRD Return Line Nozzle Program, PNPS has cut and capped the CRD return line nozzle to mitigate cracking and continues inservice inspection (ISI)examinations to monitor the effects of crack initiation and growth on the intended function of the control rod drive return line nozzle and cap. ISI examinations include ultrasonic inspection of the nozzle-to-vessel weld and ultrasonic inspection of the dissimilar metal weld overlay at the nozzle.A.2.1.4 BWR Feedwater Nozzle Program Under the BWR Feedwater Nozzle Program, PNPS has removed feedwater blend radii flaws, removed feedwater nozzle cladding, and installed a triple-sleeve-double-piston sparger to mitigate cracking.
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| This program continues enhanced inservice inspection (ISI) of the feedwater nozzles in accordance with the requirements of ASME Section Xl, Subsection IWB and the recommendation of General Electric (GE) NE-523-A71-0594 to monitor the effects of cracking on the intended function of the feedwater nozzles.A.2.1.5 BWR Penetrations Program The BWR Penetrations Program includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) documents BWRVIP-27 and BWRVIP-49 and (b)monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 to ensure the long-term integrity of vessel penetrations and nozzles.Appendix A Updated Final Safety Analysis Report Supplement Page A-14 Pilgrim Nuclear Power Station License Renewal Application i Technical Information A.2.1.6 BWR Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking Program includes (1) preventive measures to mitigate intergranular stress corrosion cracking (IGSCC), and (2) inspection and flaw evaluation to monitor IGSCC and its effects on reactor coolant pressure boundary components made of stainless steel or CASS.PNPS has taken actions to prevent IGSCC and will continue to use materials resistant to IGSCC for component replacements and repairs following the recommendations delineated in NUREG-0313, Generic Letter 88-01, and the staff-approved BWRVIP-75 report. Inspection of piping identified in NRC Generic Letter 88-01 to detect and size cracks is performed in accordance with the staff positions on schedule, method, personnel qualification and sample expansion included in the generic letter and the staff-approved BWRVIP-75 report.A.2.1.7 BWR Vessel ID Attachment Welds Program The BWR Vessel ID Attachment Welds Program includes (1) inspection and flaw evaluation in accordance with the guidelines of staff-approved BWR Vessel and Intermals Project (BWRVIP) BWRVIP-48, and '(2) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 to ensure the long-term integrity and safe operation of reactor vessel inside diameter (ID)attachment welds and support pads.A.2.1.8 BWR Vessel Internals Program The BWR Vessel Internals Program includes (a) inspection, flaw evaluation, and repair in conformance with the applicable, staff-approved BWR Vessel and Intemals Project (BWRVIP) documents, and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-1 30 to ensure the long-term integrity of vessel intermals components.
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| A.2.1.9 Containment Leak Rate Program As described in 10 CFR 50, Appendix J, containment leak rate tests are required to assure that (a) leakage through primary reactor containment and systems and components penetrating primary containment shall not exceed allowable values specified in technical specifications or associated bases and (b) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of containment, and systems and components penetrating primary containment.
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| Corrective actions are taken if leakage rates exceed acceptance criteria.Appendix A Updated Final Safety Analysis Report Supplement Page A-15 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.10 Diesel Fuel Monitoring Program The Diesel Fuel Monitoring Program entails sampling to ensure that adequate diesel fuel quality is maintained to prevent plugging of filters, fouling of injectors, and corrosion of fuel systems. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by periodic draining and cleaning of tanks and by verifying the quality of new oil before its introduction into the storage tanks.A.2.1.11 Environmental Qualification (EQ) of Electric Components Program The PNPS EQ of Electric Components program manages the effects of thermal, radiation, and cyclic aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current license term are refurbished, replaced, or their qualification extended prior to reaching the aging limits established in the evaluations.
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| Aging evaluations for EQ components are considered time-limited aging analyses (TLAAs) for license renewal.A.2.1.12 Fatigue Monitoring Program In order not to exceed design limits on fatigue usage, the Fatigue Monitoring Program tracks the number of critical thermal and pressure transients for selected reactor coolant system components.
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| The program ensures the validity of analyses that explicitly assumed a fixed number of thermal and pressure fatigue transients by assuring that the actual effective number of transients does not exceed the assumed limit.The transient cycles tracked by this program are referenced in Section 4.2.6.A.2.1.13 Fire Protection Program The Fire Protection Program includes a fire barrier inspection and a diesel-driven fire pump inspection.
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| The fire barrier inspection requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained.
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| The diesel-driven fire pump inspection requires that the pump be periodically tested to ensure that the fuel supply line can perform its intended function.The program also includes periodic inspection and testing of the Halon fire suppression system.Corrective actions, confirmation process, and administrative controls in accordance with the requirements of 10 CFR 50 Appendix B are applied to the Fire Protection Program.Appendix A Updated Final Safety Analysis Report Supplement Page A-16 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.14 Fire Water System Program The Fire Water System Program applies to water-based fire protection systems that consist of sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes, and aboveground and underground piping and components that are tested in accordance with applicable National Fire' Protection Association (NFPA) codes and standards.
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| Such testing assures functionality of systems. To determine if significant corrosion has occurred in water-based fire protection systems, periodic flushing, system performance testing and inspections are conducted.
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| Also, many of these systems are normally maintained at required operating pressure and monitored such that leakage resulting in loss of system pressure is immediately detected and corrective actions initiated.
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| In addition, wall thickness evaluations of fire protection piping are periodically performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion.
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| A sample of sprinkler heads will be inspected using the guidance of NFPA 25 (2002 Edition) Section 5.3.1.1.1, which states, 'Where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." This sampling will be repeated every 10 years after initial field service testing.A.2.1.15 Flow-Accelerated Corrosion Program The Flow-Accelerated Corrosion Program applies to safety-related and nonsafety-related carbon steel components in systems containing high-energy fluids carrying two-phase or single-phase high-energy fluid > 2% of plant operating time.The program, based on EPRI recommendations for an effective flow-accelerated corrosion program, predicts, detects, and monitors FAC in plant piping and other pressure retaining components.
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| This program includes (a) an evaluation to determine critical locations, (b) initial operational inspections to determine the extent of thinning at these locations, and (c) follow-up inspections to confirm predictions.
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| The program specifies repair or replacement of components as necessary.
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| A.2.1.16 Heat Exchanger Monitoring Program The Heat Exchanger Monitoring Program inspects heat exchangers for degradation.
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| If degradation is found, then an evaluation is performed t6 evaluate its effects on the heat exchanger's design functions including its ability to withstand a seismic event.Representative tubes within the population of heat exchangers are eddy current tested at a frequency determined by internal and external operating experience to ensure that effects of aging are identified prior to loss of intended function.
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| Along with each eddy Appendix A Updated Final Safety Analysis Report Supplement Page A-1 7 Pilgrim Nuclear Power Station License Renewal Application Technical Information current test, visual inspections are performed on accessible heat exchanger heads, covers and tube sheets to monitor surface condition for indications of loss of material.The population of heat exchangers includes the RHR heat exchangers, core spray pump motor thrust bearing lube oil coolers, HPCI gland seal condenser, HPCI turbine lube oil cooler, RCIC lube oil cooler, recirculation pump motor generator set fluid coupling oil and bearing coolers, CRD pump oil coolers, recirculation pump motor lube oil coolers, clean up recirculation pump lube oil coolers and stuffing box cooler, and EDG lube oil coolers.A.2.1.17 Inservice Inspection
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| -Containment Inservice Inspection (CII) Program The Containment Inservice Inspection Program outlines the requirements for the inspection of Class MC pressure-retaining components (primary containment) and their integral attachments in accordance with the requirements of 10 CFR 50.55a(b)(2) and the 1998 Edition of ASME Section Xl with 2000 Addenda, Inspection Program B.The primary inspection method for the primary containment and its integral attachments is visual examination.
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| Visual examinations are performed either directly or remotely with illumination and resolution suitable for the local environment to assess general conditions that may affect either the containment structural integrity or leak tightness of the pressure retaining component.
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| The program includes augmented ultrasonic exams to measure wall thickness of the containment drywell structure.
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| A.2.1.18 Inservice Inspection
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| -Inservice Inspection (ISI) Program The ISI Program is based on ASME Inspection Program B (Section XI, IWA-2432), which has 10-year inspection intervals.
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| Every 10 years the program is updated to the latest ASME Section Xl code edition and addendum approved in 10 CFR 50.55a. On July 1, 2005 PNPS entered the fourth ISI interval.
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| The code edition and addenda used for the fourth interval is the 1998 Edition with 2000 Addenda.The program consists of periodic volumetric, surface, and visual examination of components and their supports for assessment, signs of degradation, flaw evaluation, and corrective actions.A.2.1.19 Instrument Air Quality Program The Instrument Air Quality Program ensures that instrument air supplied to components is maintained free of water and significant contaminants, thereby preserving an environment that is not conducive to loss of material.
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| Dewpoint, particulate contamination, and hydrocarbon concentration are periodically checked to verify the instrument air quality is maintained.
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| Appendix A Updated Final Safety Analysis Report Supplement Page A-18 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.20 Metal-Enclosed Bus Inspection Program Under the Metal-Enclosed Bus Inspection Program, internal portions of the non-segregated phase bus which connects the 4.16kV switchgear (A3 through A6) are inspected for cracks, corrosion, foreign debris, excessive dust buildup, and evidence of water intrusion.
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| Bus insulation is inspected for signs of embrittlement, cracking, melting, swelling, or discoloration, which may indicate overheating or aging degradation.
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| Internal bus supports are inspected for structural integrity and signs of cracks. Since bolted connections are covered with heat shrink tape or insulating boots per manufacturers recommendations, a sample of accessible bolted connections is visually inspected for insulation material surface anomalies.
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| Enclosure assemblies are visually inspected for evidence of loss of material and, where applicable, enclosure assembly elastomers are visually inspected and manually flexed to manage cracking and change in material properties.
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| These inspections are performed at least once every 10 years.A.2.1.21 Non-EQ Inaccessible Medium-Voltage Cable Program In the Non-EQ Inaccessible Medium-Voltage Cable Program, in scope medium-voltage cables, not designed for, but exposed to significant moisture and voltage are tested at least once every ten years to provide an indication of the condition of the conductor insulation.
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| The specific test performed is a proven test for detecting deterioration of the insulation system due to wetting, such as power factor, partial discharge, polarization index, or other testing that is state-of-the-art at the time the test is performed.
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| Significant moisture is defined as periodic exposures that last more than a few days.Significant voltage exposure is defined as being subjected to system voltage for more than 25% of the time.Inspections for water collection in cable manholes and conduit occur at least once every two years.A.2.1.22 Non-EQ Instrumentation Circuits Test Review Program Under the Non-EQ Instrumentation Circuits Test Review Program, calibration or surveillance results for non-EQ electrical cables in circuits with sensitive, high voltage, low-level signals; (i.e., neutron flux monitoring instrumentation);
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| are reviewed.
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| Most neutron flux monitoring system cables and connections are calibrated as part of the instrumentation loop calibration at the normal calibration frequency, which provides sufficient indication of the need for corrective actions based on acceptance criteria related to instrumentation loop performance.
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| The review of calibration results is performed once every 10 years.Appendix A Updated Final Safety Analysis Report Supplement Page A-1 9 Pilgrim Nuclear Power Station License Renewal Application Technical Information For neutron flux monitoring system cables that are disconnected during instrument calibrations, testing is performed at least once every 10 years using a proven method for detecting deterioration for the insulation system (such as insulation resistance tests, or time domain reflectometry).
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| A.2.1.23 Non-EQ Insulated Cables and Connections Program The Non-EQ Insulated Cables and Connections Program provides reasonable assur-ance that intended functions of insulated cables and connections exposed to adverse localized environments caused by heat, radiation and moisture can be maintained con-sistent with the current licensing basis through the period of extended operation.
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| An adverse localized environment is significantly more severe than the specified service condition for the insulated cable or connection.
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| A representative sample of accessible insulated cables and connections in adverse localized environments is visually inspected at least once every 10 years for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination.
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| A.2.1.24 Oil Analysis Program The Oil Analysis Program maintains oil systems free of contaminants (primarily water and particulates) thereby preserving an environment that is not conducive to loss of material, cracking, or fouling. Activities include sampling and analysis of lubricating oil for detrimental contaminants, water, and particulates.
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| Sampling frequencies are based on vendor recommendations, accessibility during plant operation, equipment importance to plant operation, and previous test results.A.2.1.25 One-Time Inspection Program The elements of the One-Time Inspection Program include (a) determination of the sample size based on an assessment of materials of fabrication, environment, plausible aging effects, and operating experience; (b) identification of the inspection locations in the system or component based on the aging effect; (c) determination of the examination technique, including acceptance criteria that would be effective in managing the aging effect for which the' component is examined; and (d) evaluation of the need for follow-up examinations to monitor the progression of any aging degradation.
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| A one-time inspection activity is used to verify the effectiveness of the water chemistry control programs by confirming that unacceptable cracking, loss of material, and fouling is not occurring on components within systems covered by water chemistry control programs [Sections A.2.1.36, A.2.1.37, and A.2.1.38].
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| Appendix A Updated Final Safety Analysis Report Supplement PaeA2 Page A-20 Pilgrim Nuclear Power Station License Renewal Application
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| (_)Technical Information One-time inspection activities on* internal surfaces of buried carbon steel pipe on the standby gas treatment system discharge to the stack,* internal surfaces of compressed air and EDG system components containing untreated air,* internal surfaces of stainless steel radioactive waste and sanitary soiled waste and vent system components containing untreated water,* small bore piping in the reactor coolant system and associated systems that form the reactor coolant pressure boundary,* reactor vessel flange leak-off line, and* main steam flow restrictors are used to confirm that loss of material, cracking, and reduction of fracture toughness, as applicable, are not occurring or are so insignificant that an aging management program is not warranted.
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| When evidence of an aging effect is revealed by a one-time inspection, routine evaluation of the inspection results will identify appropriate corrective actions.A.2.1.26 Periodic Surveillance and Preventive Maintenance Program The Periodic Surveillance and Preventive Maintenance Program includes periodic inspections and tests that manage aging effects not managed by other aging management programs.
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| The preventive maintenance and surveillance testing activities are generally implemented through repetitive tasks or routine monitoring of plant operations.
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| Temperatures are monitored during periodic emergency diesel generator (EDG), station blackout diesel, and security diesel surveillance tests to verify that associated heat exchangers are capable of removing the required amount of heat, thereby managing fouling of the heat exchanger tubes.Periodic inspections using visual or other non-destructive examination techniques verify that the following components are capable of performing their intended function.* reactor building crane, rails, and girders* refueling platform carbon steel components
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| * main stack components
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| * standby liquid control system discharge accumulators
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| * carbon steel piping in the waterline region of the torus* HPCI gland seal condenser blower and suction piping* RCIC steam supply and exhaust piping downstream of the strainers and steam traps Appendix A Updated Final Safety Analysis Report Supplement Page A-21 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| * standby gas treatment system expansion joints, demister drain valves and demister drain piping* drain lines from each reactor building auxiliary bay passing into the water trough in the torus* clean-up recirculation pump P-204B stuffing box cooler* RBCCW copper alloy cooling coils* EDG, station blackout diesel, and security diesel intake air, air start, and exhaust components
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| * EDG, station blackout diesel, and security diesel jacket water radiators* security diesel oil cooler and aftercooler
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| * area coolers VAC-21 OANB, VAC-202A/B, and VAC-204ANB/C/D
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| * VSF-103A/B, VAC-202A/B, VAC-204A/B/C/D, and EDG engine driven fan duct flexible connections
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| * condensate storage tanks* circulating water, potable & sanitary water, radioactive waste, sanitary soiled waste & vent, plumbing and drains and screen wash system components
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| -flex/expansion joints in the circulating water, HVAC/chilled water, and radioactive waste systems A.2.1.27 Reactor Head Closure Studs Program The Reactor Head Closure Studs Program includes Inservice inspection (ISI) in conformance with the requirements of the ASME Code, Section Xi, Subsection IWB, and preventive measures (e.g. rust inhibitors, stable lubricants, appropriate materials) to mitigate cracking and loss of material of reactor head closure studs, nuts, washers, and bushings.A.2.1.28 Reactor Vessel Surveillance Program PNPS is a participant In the BWR vessel and internals project (BWRVIP) Integrated Surveillance Program (ISP) as incorporated into the plant Technical Specifications by License Amendment 209. The Reactor Vessel Surveillance Program monitors changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. As BWRVIP-ISP capsule test reports become available for RPV materials representative of PNPS, the actual shift in the reference temperature for nil-ductility transition of the vessel material may be updated. In accordance with 10 CFR 50 Appendices G and H, PNPS reviews relevant test reports to assure compliance with fracture toughness requirements and P-T limits.BWRVIP-116, 'BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Implementation for License Renewal," describes the design and implementation of the ISP during the period of extended operation.
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| BWRVIP-116 identifies additional capsules, their withdrawal schedule, and contingencies to ensure that the requirements of 10 CFR 50 Appendix H are met for the period of extended operation.
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| Appendix A Updated Final Safety Analysis Report Supplement Page A-22 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.29 Selective Leaching Program The Selective Leaching Program ensures the integrity of components made of cast iron, bronze, brass, and other alloys exposed to raw water, treated water, or groundwater that may lead to selective leaching.
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| The program includes a one-time visual inspection and hardness measurement of selected components that may be susceptible to selective leaching to determine whether loss of material due to selective leaching is occurring, and whether the process will affect the ability of the components to perform their intended function for the period of extended operation.
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| A.2.1.30 Service Water Integrity Program The Service Water Integrity Program relies on implementation of the recommendations of NRC GL 89-13 to ensure that the effects of aging on the salt service water (SSW)system are managed for the period of extended operation.
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| The program includes component inspections for erosion, corrosion, and blockage and performance monitoring to verify the heat transfer capability of the safety-related heat exchangers cooled by SSW. Chemical treatment using biocides and chlorine and periodic cleaning and flushing of redundant or infrequently used loops are the methods used to control or prevent fouling within the heat exchangers and loss of material in SSW components.
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| A.2.1.31 Structures Monitoring
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| -Masonry Wall Program i_)The objective of the Masonry Wall Program is to manage cracking so that the evaluation basis established for each masonry wall within the scope of license renewal remains valid through the period of extended operation.
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| The program includes all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4. Included components are the 10 CFR 50.48-required masonry walls, radiation shielding masonry walls, masonry walls with the potential to affect safety-related components, and the torus compartment water trough.Masonry walls are visually examined at a frequency selected to ensure there is no loss.of intended function between inspections.
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| A.2.1.32 Structures Monitoring
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| -Structures Monitoring Program Structures monitoring is in accordance with 10 CFR 50.65 (Maintenance Rule) as addressed in Regulatory Guide 1.160 and NUMARC 93-01. Periodic inspections are used to monitor the condition of structures and structural components to ensure there is no loss of structure or structural component intended function.Appendix A Updated Final Safety Analysis Report Supplement Page A-23 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.1.33 Structures Monitoring
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| -Water Control Structures Monitoring Program The Water Control Structures Monitoring Program includes visual inspections to manage loss of material and loss of form for water-control structures (breakwaters, jetties, and revetments).
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| The water-control structures are of rubble mound construction with the outer layer protected by heavy capstone.
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| Parameters monitored include settlement (vertical displacement) and rock displacement.
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| These parameters are consistent with those described in RG 1.127.A.2.1.34 System Walkdown Program The System Walkdown Program entails inspections of external surfaces of components subject to aging management review. The program is also credited with managing loss of material from internal surfaces, for situations in which internal and external material and environment combinations are the same such that external surface condition is representative of internal surface condition.
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| Surfaces that are inaccessible during plant operations are inspected during refueling outages. Surfaces are inspected at frequencies to provide reasonable assurance that effect of aging will be managed such that applicable components will perform their intended function during the period of extended operation.
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| A.2.1.35 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program The purpose of the Thermal Aging and Neutron Irradiation Embrittlement of CASS Program is to assure that reduction of fracture toughness due to thermal aging and reduction of fracture toughness due to radiation embrittlement will n 6 t result in loss of intended function during the period of extended operation.
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| This program evaluates CASS components in the reactor vessel internals and requires non-destructive examinations as appropriate.
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| A.2.1.36 Water Chemistry Control -Auxiliary Systems Program The purpose of the Water Chemistry Control -Auxiliary Systems Program is to manage loss of material for components exposed to treated water.Program activities include sampling and analysis of the stator cooling water system to minimize component exposure to aggressive environments.
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| A.2.1.37 Water Chemistry Control -BWR Program The objective of the Water Chemistry Control -BWR Program is to manage aging effects caused by corrosion and cracking mechanisms.
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| The program relies on monitoring and control of water chemistry based on EPRI Report 1008192 (BWRVIP-Appendix A Updated Final Safety Analysis Report Supplement Page A-24 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| .9*130). BWRVIP-130 has three sets of guidelines:
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| one for primary water, one for condensate and feedwater, and one for control rod drive (CRD) mechanism cooling water. EPRI guidelines in BWRVIP-1 30 also include recommendations for controlling water chemistry in the torus, condensate storage tank, demineralized water storage tanks, and spent fuel pool.The Water Chemistry Control -BWR Program optimizes the primary water chemistry to minimize the potential for loss of material and cracking.
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| This is accomplished by limiting the levels of contaminants in the RCS that could cause loss of material and cracking.
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| Additionally, PNPS has instituted hydrogen water chemistry (HWC) to limit the potential for intergranular SCC (IGSCC) through the reduction of dissolved oxygen in the treated water.A.2.1.38 Water Chemistry Control -Closed Cooling Water Program The Water Chemistry Control -Closed Cooling Water Program includespeventive measures that manage loss of material, cracking, and fouling for components in closed cooling water systems (reactor building closed cooling water, turbine building closed cooling water, emergency diesel generator cooling water, station blackout diesel cooling water, security diesel generator cooling water, and plant heating).
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| These chemistry activities provide for monitoring and controlling closed cooling water chemistry using PNPS procedures and processes based on EPRI guidance for closed cooling water chemistry.
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| Appendix A Updated Final Safety Analysis Report Supplement Page A-25 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.2 Evaluation of Time-Limited Aging Analyses In accordance with 10 CFR 54.21 (c), an application for a renewed license requires an evaluation of time-limited aging analyses (TLAA) for the period of extended operation.
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| The following TLAA have been identified and evaluated to meet this requirement.
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| A.2.2.1 Reactor Vessel Neutron Embrittlement The reactor vessel neutron embrittlement TLAA will either remain valid for the period of extended operation (P-T limits) in accordance with 10 CFR 54.21(c)(1)(i) or have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(ii). Fifty-four EFPY would be the effective full power years at the end of the period of extended operation assuming an average capacity factor of 90% for 60 years.A.2.2. 1.1 Reactor Vessel Fluence Calculated fluence is based on a time-limited assumption defined by the operating term. As such, fluence is the time-limited assumption for the time-limited aging analyses that evaluate reactor vessel embrittlement.
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| Fluence values were calculated using the RAMA fluence calculation method. The RAMA fluence method was developed for the Electric Power Research Institute, Inc.and the Boiling Water Reactor Vessel and Internals Project (BWRVIP) for the purpose of calculating neutron fluence in boiling water reactor components.
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| This method has been approved by the NRC (Reference A.2-9) for application in accordance with Regulatory Guide 1.190.A.2.2.1.2 Pressure-Temperature Limits The P-T limits were derived from calculations made in accordance with the guidance of ASME Appendix G, as modified by Code Cases N-588 and N-640, ASTM Standards, 10 CFR 50 Appendices G and H, Regulatory Guide 1.99 Revision 2, and Generic Letter 88-11.The fluence calculations performed in accordance with RG 1.190 confirm that the fluence for 54 EFPY is less than the fluence used to calculate the P-T limits. The existing Technical Specification P-T limits remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
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| A.2.2.1.3' Charpy Upper-Shelf Energy The predictions for percent drop in CVUSE at 54 EFPY are based on chemistry data and unirradiated CVUSE data submitted to the NRC in the PNPS response to GL 92-01, and %T fluence values..I Appendix A Updated Final Safety Analysis Report Supplement Page A-26 Pilgrim Nuclear Power Station License Renewal Application Technical Information The 54 EFPY CVUSE values were calculated using Regulatory Guide 1.99, Position 1, Figure 2; specifically, the formula for the lines was used to calculate the percent drop in CVUSE.All CVUSE values are predicted to remain well above the requirement of 50 ft-lbs during the period of extended operation.
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| As such, this TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).
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| A.2.2.1.4 Adjusted Reference Temcerature The PNPS reactor vessel was designed for a 40-year life with an assumed exposure of less than 1019 nvt of neutrons with energies exceeding 1 MeV. After approximately 4.17 EFPY, the first surveillance capsule was withdrawn from the vessel and tested.The capsule test report concludes that the shift in RTNDT and upper shelf energy over 32 EFPY will be within 10 CFR 50 Appendix G guidelines.
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| PNPS has projected values for RTNDT and adjusted reference temperature (ART) at 54 EFPY using the methodology of Regulatory Guide 1.99. These values were calculated using the chemistry data, margin values, initial RTNDT values, and chemistry factors (CFs) contained in the PNPS response to GL 92-01 and other licensing correspondence (Reference A.2-1 0). New fluence factors (FFs) were calculated using the expression in Regulatory Guide 1.99, Revision 2, Equation 2 using 54 EFPY fluence values.The RTNDT TLAA has been projected through the period of extended operation, with acceptable results, in accordance with 10 CFR 54.21 (c)(1)(ii).
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| A.2.2.1.5 Reactor Vessel Circumferential Weld Inspection Relief Relief from reactor vessel circumferential weld examination requirements under Generic Letter 98-05 is based on assessments indicating an acceptable probability of failure per reactor operating year. The analysis is based on reactor vessel metallurgical conditions as well as flaw indication sizes and frequencies of occurrence that are expected at the end of a licensed operating period.PNPS received NRC approval for this relief for the remainder of the original 40-year license term (Reference A.2-3). The basis for this relief request is an analysis that satisfied the limiting conditional failure probability for the circumferential welds at the expiration of the current license, based on the NRC SERs for BWRVIP-05 (Reference A.2-6) and BWRVIP-74 (Reference A.2-11) and the extent of neutron embrittlement.
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| The chemistry composition and chemistry factor values for PNPS are slightly higher than those used in the NRC analysis; however, the 54 EFPY fluence value is Appendix A Updated Final Safety Analysis Report Supplement Page A-27 Pilgrim Nuclear Power Station License Renewal Application Technical Information considerably lower than the corresponding 64 EFPY generic value. As a result, the shift in reference temperature is lower than the 64 EFPY shift in the NRC analysis.
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| In addition, the unirradiated reference temperature of the PNPS material is lower than the initial value assumed in the NRC analysis.
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| The combination of a lower'initial reference temperature (RTNDT(U))
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| and a lower shift (RTNDT w/o margin) yields an adjusted reference temperature that is considerably lower than the NRC mean analysis value. Therefore, this TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(ii).
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| A.2.2.1.6 Reactor Vessel Axial Weld Failure Probability The BWRVIP recommendations for inspection of reactor vessel shell welds (BWRVIP-05, Reference A.2-4) are based on generic analyses supporting an NRC SER (References A.2-5, A.2-6). The generic-plant axial weld failure rate is no more than 5 x 10-6 per reactor year as calculated in the BWRVIP-74 SER (Reference A.2-11). BWRVIP-05 showed that this axial weld failure rate is orders of magnitude greater than the 40-year end-of-life circumferential weld failure probability, and used this analysis to justify relief from inspection of the circumferential welds as described above.The basis for this relief request was a plant specific analysis that showed the limiting conditional failure probability for the PNPS circumferential welds at the end of the original operating term were less than the values calculated in the BWRVIP-05 SER.The BWRVIP-05 SER concluded that the reactor vessel failure frequency due to failure of the limiting axial welds in the BWR fleet at the end of 40 years of operation is less than 5x1 04 per reactor year. This failure frequency is dependent upon given assumptions of flaw density, distribution, and location.
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| The failure frequency also assumes that "essentially 100%" of the reactor vessel axial welds will be inspected.
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| The BWRVIP-74 SER states it is acceptable to show that the mean RTNDT of the limiting beltline axial weld at the end of the period of extended operation is less than the limiting value given in the SERs for BWRVIP-74 and BWRVIP-05.
| |
| The projected 54 EFPY mean RTNDT values for PNPS are less than the limiting 64 EFPY RTNDT in the analysis performed by the NRC staff (Table 2.6-5 of the BWRVIP-05 SER). As such, this TLAA has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).
| |
| A.2.2.2 Metal Fatigue A.2.2.2.1 Class I Metal Fatigue Class I components evaluated for fatigue and flaw growth include the reactor pressure vessel (RPV) and appurtenances, certain reactor vessel internals, the reactor recirculation system (RRS), and the reactor coolant system (RCS) pressure Appendix A Updated Final Safety Analysis Report Supplement IIPage A-28 Pilgrim Nuclear Power Station License Renewal Application Technical Information boundary.
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| The PNPS Class 1 systems include components within the ASME Section Xl, IWB inspection boundary.The design of the reactor vessel internals is in accordance with the intent of ASME Section 1I1. A review of design basis documents reveals that the only reactor vessel internals components for which there is a fatigue evaluation are the core shroud tie rods (stabilizer), the result of a repair to structurally replace circumferential shroud welds.The PNPS fatigue monitoring program will assure that the allowed number of transient cycles is not exceeded.
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| The program requires corrective action if transient cycle limits are approached.
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| Consequently, the TLAA (fatigue analyses) based on those transients will remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(i) or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(ifl).A.2.2.2.2 Norn-Class 1 Metal Fatiaue For non-Class 1 components identified as subject to cracking due to fatigue, a review of system operating characteristics was conducted to determine the approximate frequency of any significant thermal cycling. If the number of equivalent full Q temperature cycles is below the limit used for the original design (usually 7000 cycles), the component is suitable for extended operation.
| |
| If the number of equivalent full temperature cycles exceeds the limit, evaluation of the individual stress calculations require evaluation.
| |
| No components were identified with projected cycles exceeding 7000. Therefore, the TLAA for non-Class I piping and components remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(i).A.2.2.2.3 Environmental Effects on Fatigue The effects of reactor water environment on fatigue were evaluated for license renewal. Projected cumulative usage factors (CUFs) were calculated for the limiting locations identified in NUREG/CR-6260.
| |
| Several locations may exceed a CUF of 1.0 with consideration of environmental effects during the period of extended operation.
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| For these locations, prior to the period of extended operation, PNPS will (1) refine the fatigue analysis to lower the predicted CUF to less than 1.0; (2) manage fatigue at the affected locations with an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC); or (3)repair or replace the affected locations.
| |
| Appendix A Updated Final Safety Analysis Report Supplement
| |
| .'' Page A-29 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.2.3 Environmental Qualification of Electrical Components The PNPS EQ Program implements the requirements of 10 CFR 50.49 (as further defined by the Division of Operating Reactors Guidelines, NUREG-0588, and Reg.Guide 1.89). The program requires action before individual components exceed their qualified life. In accordance with 10 CFR 54.21(c)(1)(iii), implementation of the EQ Program provides reasonable assurance that the effects of aging on components associated with EQ TLAAs will be adequately managed such that the intended functions can be maintained for the period of extended operation.
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| A.2.2.4 Fatigue of Primary Containment, Attached Piping, and Components In conjunction with the Mark I Containment Long-Term Program, the torus and attached piping systems were analyzed for fatigue due to mechanical loadings as well as thermal and anchor motion. This analysis was based on assumptions of the number of SRV actuations, operating basis earthquakes, and accident conditions during the life of the plant.The fatigue usage calculated for PNPS is zero. However, the analysis considered all BWR plants which utilize the Mark I containment design. The analysis concluded that for all plants and piping systems considered, the fatigue usage factor for an assumed 40-year plant life was less than 0.5. Extending plant life by an additional 20 years would produce a usage factor below 0.75. Since this is less than 1.0, the fatigue criteria are satisfied.
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| This TLAA has been projected through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).
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| A.2.2.5 Vessel ID Attachment Welds Fatigue Analysis The BWRVIP-48 fatigue analyses for various configurations of different vessel ID bracket attachments are considered TLAA. The PNPS bracket configurations were included in the analysis.
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| Analysis of fatigue for 60 years showed that no CUFs are above 0.4. This analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
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| A.2.2.6 Instrument Penetrations Fatigue Analysis The BWRVIP-49 fatigue analysis of several configurations of instrumentation penetrations, including the PNPS configuration, is considered a TLAA. Analysis of fatigue for 60 years showed that all CUFs are below 0.4. This analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).
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| Appendix A Updated Final Safety Analysis Report Supplement Page A-30 Pilgrim Nuclear Power Station License Renewal Application Technical Information A.2.3 References A.2-1 PNPS License Renewal Application A.2-2 (NRC SER for PNPS License Renewal -later)A.2-3 Boska, J. (NRC), to M. Bellamy (ENGC), "Pilgrim Nuclear Power Station-Pilgrim Relief Request No. 28, Relief from ASME Code, Section Xl, Examinations of Reactor Pressure Vessel Circumferential Shell Welds (TAC No. MB6074)," letter dated April 11, 2003.A.2-4 BWRVIP-05, EPRI Report TR-1 05697, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," for the Boiling Water Reactor Owners Group (Proprietary), September 28, 1995, with supplementing letters of June 24 and October 29, 1996; May 16, June 4, June 13, and December 18, 1997; and January 13, 1998.A.2-5 Lainas, G. C. (NRC), to C. Terry (Niagara Mohawk Power Company, BWRVIP Chairman), BWRVIP-05 SER (Final), Final Safety Evaluation of the BWRVIP Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925), letter dated July 28, 1998.A.2-6 Strosnider, J. R., Jr., (NRC) to C. Terry (BWRVIP Chairman), BWRVIP-05SER, Supplement to Final Safety Evaluation of the BWRVIP Vessel and Internals Project BWRVIP-05 Report (TAC No. MA3395), letter dated March 7, 2000.A.2-7 Riggs, W. J. (ENGC) to NRC, uAmendment 01-01 to the Third Ten-Year Interval Inservice Inspection Program," letter dated November 20, 2001.A.2-8 Milano, P. D. (NRC), to E. T. Boulette (BECo), "Evaluation of the Third Ten-Year Interval Inspection Program Plan, and Associated Requests for Relief for Pilgrim Nuclear Power Station (TAC No. M93398)," letter dated March 20, 1997.A.2-9 Bateman, W. H. (NRC), to Eaton, W. (BWRVIP), "Safety Evaluation of Proprietary EPRI Reports BWRVIP-114, -115, -117, and -121 and TWE-PSE-001-R-001,m letter dated May 13, 2005.A.2-10 Bellamy, M. (ENGC), to Document Control Desk (NRC), "Additional Information Related to Pilgrim Technical Specification Change Concerning Pressure-Temperature Limit Curves of Figure 3.6.1.2 and 3," letter 2.01.014 dated January 30, 2001.A.2-11 Grimes, C. l. (NRC), to C. Terry (BWRVIP Chairman), Acceptance for referencing of EPRI Proprietary Report TR-113596, 'BWR Vessel and Internals Project, BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74) and Appendix A,'Demonstration of Compliance with the Technical Information requirements of the License Renewal Rule (10CRF54.21)', letter dated October 18, 2001.Appendix A Updated Final Safety Analysis Report Supplement Page A-31 Pilgrim Nuclear Power Station License Renewal Application Technical Information APPENDIX B AGING MANAGEMENT PROGRAMS AND ACTIVITIES TABLE OF CONTENTS B.0 INTRODUCTION
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| ..............
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| -B-1 B.0.1 Overview .B-1 8.0.2 Format of Presentation .B-1 B.0.3 PNPS Corrective Actions, Confirmation Process and Administrative Controls ... B-2 B.0.4 Operating Experience .B-3 B.0.5 Aging Management Programs .B-3 B.O.6 Correlation with NUREG-1801 Aging Management Programs ..........
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| B-6 B.1 AGING MANAGEMENT PROGRAMS AND ACTIVITIES
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| ..................
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| .. B-15 B.1.1 Boraflex Monitoring
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| ... B-1 5 B.1.2 Buried Piping and Tanks Inspection .B-17 B.1.3 BWR CRD Return Line Nozzle .... B-19 B.1.4 BWR Feedwater Nozzle .B-22 B.1.5 BWR Penetrations
| |
| ..;.. .B-24 B. 1.6 BWR Stress Corrosion Cracking .........8. B-27 B.1.7 B.1.8 B.1.9 B.1.10 B.1.11 B.1.12 B.1.13 B.1.14 B.1.15 BWR Vessel ID Attachment Welds .........................
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| B-29 BWR Vessel Internals
| |
| .....................................
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| B-31 Containment Leak Rate ..........................
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| ...........
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| B-35 Diesel Fuel Monitoring
| |
| .....................................
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| B-36 Environmental Qualification of Electric Components
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| ..................
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| B-39 Fatigue Monitoring
| |
| ...B-41 Fire Protection
| |
| ...8.-B43 B.1.13.1 Fire Protection
| |
| .........................
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| B-43 B.1.13.2 Fire Water System .........................
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| B-46 Flow-Accelerated Corrosion
| |
| ..................
| |
| B-50 Heat Exchanger Monitoring
| |
| ..................
| |
| B-52 Appendix B Aging Management Programs and Activities Page i Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.16 Inservice Inspection
| |
| ...........................................
| |
| B-55 B.1.16.1 Containment Inservice Inspection (CII) ...... ..............
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| B-56 B.1.16.2 Inservice Inspection (ISI) ................................
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| B-59 B.1.17 Instrument Air Quality ..........................................
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| B-63 B.1.18 Metal-Enclosed Bus Inspection
| |
| ...................................
| |
| B-66 B.1.19 Non-EQ Inaccessible Medium-Voltage Cable ..................
| |
| ..... B-68 B.1.20 Non-EQ Instrumentation Circuits Test Review ...... B-69 B.1.21 Non-EQ Insulated Cables and Connections
| |
| ..................
| |
| B-71 B.1.22 Oil Analysis .. ............
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| B-73 B.1.23 One-Time Inspection
| |
| .........................................
| |
| B-76 B.1.24 Periodic Surveillance and Preventive Maintenance
| |
| ............
| |
| 8I ....... B-79 B.1.25 Reactor Head Closure Studs .............................
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| B-87 B. 1.26 Reactor Vessel Surveillance
| |
| ......................
| |
| B-89 B.1.27 Selective Leaching ............................
| |
| B-91 B.1.28 Service Water Integrity
| |
| .............
| |
| ............
| |
| B-92 B.1.29 Structures Monitoring.
| |
| ...........................
| |
| B-94 ( )B.1.29.1 Masonry Wall .........................
| |
| B-94 B.1.29.2 Structures Monitoring
| |
| ....................
| |
| ..... B-95 B.1.29.3 Water Control Structures Monitoring
| |
| ...........
| |
| 8...........
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| B-97 B.1.30 System Walkdown .............................................
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| B-99 B.1.31 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) ........................................
| |
| B-100 B.1.32 Water Chemistry Control .......................-
| |
| B-102 B.1.32.1 Water Chemistry Control -Auxiliary Systems ................
| |
| B-102 B.1.32.2 Water Chemistry Control -BWR .........................
| |
| B-105 B.1.32.3 Water Chemistry Control -Closed Cooling Water ...... ...... B-107 B.2 REFERENCES
| |
| ..................................................
| |
| B-110 Appendix B Aging Management Programs and Activities Page ii Pilgrim Nuclear Power Station License Renewal Application Technical Information B.0 INTRODUCTION B.O.1 OVERVIEW The aging management review results for the integrated plant assessment of Pilgrim Nuclear Power Station (PNPS) are presented in Sections 3.1 through 3.6 of this application.
| |
| The programs credited in the integrated plant assessment for managing aging effects are described in this appendix.Each aging management program described in this appendix has ten elements in accordance with the guidance in NUREG-1800 (Reference B.2-1) Appendix A.1, "Aging Management Review-Generic," Table A.1-1, "Elements of an Aging Management Program for License Renewal." For aging management programs that are comparable to the programs described in Sections X and Xl of NUREG-1801 (Reference B.2-2), "Generic Aging Lessons Learned (GALL) Report," the ten elements have been compared to the elements of the NUREG-1801 program. For plant-specific programs which do not correlate with NUREG-1801, the ten elements are addressed in the program description.
| |
| B.0.2 FORMAT OF PRESENTATION For those aging management programs that are comparable to the programs described in Sections X and Xl of NUREG-1801, the program discussion is presented in the following format:* Program Description -abstract of the overall program.* NUREG-1801 Consistency
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| -summary of the degree of consistency between the PNPS program and the corresponding NUREG-1801 program, when applicable (i.e., degree of similarity, etc.).* Exceptions to NUREG-1801
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| -exceptions to the NUREG-1 801 program, including a justification for the exceptions (when applicable).
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| * Enhancements
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| -future program enhancements with a proposed schedule for their completion (when applicable), including additional program features to manage aging effects not addressed by the NUREG-1801 program.* Operating Experience -discussion of operating experience information specific to the program.* Conclusion -statement of reasonable assurance that the program is effective, or will be effective, once implemented with necessary enhancements.
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| For plant-specific programs, the above format is generally followed, with additional discussion of each of the ten elements.Appendix B Aging Management Programs and Activities Page B-1 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.O.3 PNPS CORRECTIVE ACTIONS, CONFIRMATION PROCESS AND ADMINISTRATIVE CONTROLS Three attributes common to all aging management programs are corrective actions, confirmation process and administrative controls.
| |
| Discussion of these attributes is presented below.Corrective actions have program-specific details which are included in the descriptions of the individual programs in this report, but further discussion of the confirmation process and administrative controls is not necessary and is not included in the descriptions of the individual programs.Corrective Actions PNPS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B.Conditions adverse to quality, such as failures, malfunctions, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected.
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| In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence.
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| In addition, the root cause of the significant condition adverse to quality and the corrective action implemented are documented and reported to appropriate levels of management.
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| Confirmation Process PNPS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B.The Entergy Quality Assurance Program applies to PNPS safety-related structures and components.
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| Corrective actions and administrative (document) control for both safety-related and nonsafety-related structures and components are accomplished per the existing PNPS corrective action program and document control program. The confirmation process is part of the corrective action program and includes* reviews to assure that proposed actions are adequate,* tracking and reporting of open corrective actions, and* review of corrective action effectiveness.
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| Any follow-up inspection required by the confirmation process is documented in accordance with the corrective action program. The corrective action program constitutes the confirmation process for aging management programs and activities.
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| The PNPS confirmation process is consistent with NUREG-1 801.Appendix B Aging Management Programs and Activities Page B-2 Pilgrim Nuclear Power Station License Renewal Application Technical Information Administrative Controls PNPS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B.The Entergy Quality Assurance Program applies to PNPS safety-related structures and components.
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| Administrative (document) control for both safety-related and nonsafety-related structures and components is accomplished per the existing document control program. The PNPS administrative controls are consistent with NUREG-1801.
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| B.O.4 OPERATING EXPERIENCE Operating experience for the programs and activities credited with managing the effects of aging was reviewed.
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| The operating experience review included a review of corrective actions resulting in program enhancements.
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| For inspection programs, reports of recent inspections, examinations, or tests were reviewed to determine if aging effects have been identified on applicable components.
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| For monitoring programs, reports of sample results were reviewed to determine if parameters are being maintained as required by the program. Also, program owners contributed evidence of program success or weakness and identified applicable self-assessments, QA audits, peer evaluations, and NRC reviews.B.O.5 AGING MANAGEMENT PROGRAMS The following aging management programs are described in the sections listed of this appendix.Programs are identified as either existing or new. The programs are either comparable to programs described in NUREG-1801 or are plant-specific.
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| The correlation between NUREG-1801 programs and PNPS programs is shown in Table B-2, with plant-specific programs listed near the end.Table B-1 Aging Management Programs 1) Boraflex Monitoring Program B.1.1 existing 2) Buried Piping and Tanks Inspection B.1.2 new Program 3) BWR CRD Return Line Nozzle Program B.1.3 existing 4) BWR Feedwater Nozzle Program B.1.4 existing 5) BWR Penetrations Program l B.1.5 existing 6) BWR Stress Corrosion Cracking Program B.1.6 existing 7) BWR Vessel ID Attachment Welds B. 1.7 existing Program Appendix B Aging Management Programs and Activities Page B-3 Pilgrim Nuclear Power Station License Renewal Application Technical Information Q-}Table B-1 Aging Management Programs (Continued)
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| : 8) BWR Vessel Internals Program B.1.8 existing 9) Containment Leak Rate Program B. 1.9 existing 10) Diesel Fuel Monitoring Program B.1.10 existing 11) Environmental Qualification (EQ) of B.1.11 existing Electric Components Program 12) Fatigue Monitoring Program B.1.12 existing 13) Fire Protection
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| -Fire Protection Program B.1.13.1 existing 14) Fire Protection
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| -Fire Water System B.1.13.2 existing Program 15) Flow-Accelerated Corrosion Program B.1.14 existing 16) Heat Exchanger Monitoring Program B.1.15 new 17) Inservice Inspection
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| -Containment B.1.16.1 existing Inservice Inspection (CII) Program 18) Inservice Inspection
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| -Inservice Inspection B.1 .16.2 existing (ISI) Program 19) Instrument Air Quality Program B.1.17 existing 20) Metal-Enclosed Bus Inspection Program B.1.18 new 21) Non-EQ Inaccessible Medium-Voltage B.1.19 new Cable Program 22) Non-EQ Instrumentation Circuits Test B.1.20 new Review Program 23) Non-EQ Insulated Cables and B.1.21 new Connections Program 24) Oil Analysis Program B.1.22 existing 25) One-Tlime Inspection Program B.1.23 new 26) Periodic Surveillance and Preventive B.1.24 existing Maintenance Program Appendix B Aging Management Programs and Activities Page B-4 Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-1i Aging Management Programs (Continued)
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| : 27) Reactor Head Closure Studs Program B.1.25 existing 28) Reactor Vessel Surveillance Program B.1.26 existing 29) Selective Leaching Program B.1.27 new 30) Service Water Integrity Program B.1.28 existing 31) Structures Monitoring
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| -Masonry Wall B.1.29.1 existing Program 32) Structures Monitoring
| |
| -Structures B. 1.29.2 existing Monitoring Program 33) Structures Monitoring
| |
| -Water Control B.1.29.3 existing Structures Monitoring Program 34) System Walkdown Program B.1.30 existing 35) Thermal Aging and Neutron Irradiation B.1.31 new Embrittlement of Cast Austenitic Stainless Steel (CASS) Program 36)' Water Chemistry Control -Auxiliary B.1.32.1 existing Systems Program 37) Water Chemistry Control -BWR Program B.1.32.2 existing 38) Water Chemistry Control -Closed Cooling B.1.32.3 existing Water Program Appendix B Aging Management Programs and Activities Page B-6 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.O.6 CORRELATION WITH NUREG-1801 AGING MANAGEMENT PROGRAMS The correlation between NUREG-1801 programs and PNPS programs is shown below. For the PNPS programs, links to appropriate sections of this appendix are provided.Table B-2 PNPS AMP Correlation with NUREG-1801 Programs NUREG-1801 NUREG-1801 Program PNPS Program Number Environmental Qualification (EQ) Environmental Qualification (EQ) of X.El of Electric Components Electric Components Program [B.1.11]X.MI Metal Fatigue of Reactor Coolant Fatigue Monitoring Program [B.1.12]Pressure Boundary Concrete Containment Tendon X.S1 Pets Not applicable ASME Section'XI Inservice See plant-specific inservice Inspection XI.MI Inspection, Subsections IWB, -Inservice Inspection (ISI) Program IWC, and IWD [B.1.16.2]
| |
| Water Chemistry Control -BWR XL.M2 Water Chemistry Program {B. 1.32.2]XI.M3 Reactor Head Closure Studs 0Reactor Head Closure Studs Program XLM3Reacor ead losre Suds[B.1 .25]XLM4BWRVessl I Attchmnt WldsBWR Vessel ID Attachment Welds XI.M4 BWR Vessel ID Attachment Welds Program [B.1.7]XL.M5 BWR Feedwater Nozzle BWR Feedwater Nozzle Program[B.1.4]BWR Control Rod Drive Return BWR CRD Return Line Nozzle XI.M6 Line Nozzle Program [B.1.3]XI.M7 BWR Stress Corrosion Cracking BWR Stress Corrosion Cracking Program [B.1.6]XL.M8 BWR Penetrations BWR Penetrations Program [B.1.5]XL.M9 BWR Vessel Internals BWR Vessel Intemals Program [BA1.8]C.'Appendix B Aging Management Programs and Activities
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| 'Page B-6 Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-2 PNPS AMP Correlation with NUREG-1801 Programs (Continued)
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| NUREG-1801 NUREG-1801 Program PNPS Program Number XI.M10 Boric Acid Corrosion Not applicable Nickel-Alloy Nozzles and Xl M11 Penetrations Not applicable Nickel-Alloy Penetration Nozzles XI.M11A Welded to the Upper Reactor Not applicable Vessel Closure Heads of Pressurized Water Reactors Thermal Aging Embrittlement of XI.M12 Cast Austenitic Stainless Steel Not applicable (CASS)m AThermal Aging and Neutron Thermal Aging and Neutron Ebiteeto atAseii XI.M13 Irradiation Embrittlement of Cast ESbrittlement of Cast Austenitic Austenitic Stainless Steel (CASS) Stain[ess Steel (CASS) Program XI.M14 Loose Part Monitoring Not applicable XI.M15 Neutron Noise Monitoring Not applicable XI.M16 PWR Vessel Internals Not applicable X.M17 -Flow-Accelerated Corrosion X Flow-Accelerated Corrosion Program XI.M7 Flw-Acelerted orroion[B.1
| |
| .14]XI.M18 Bolting Integrity Not applicable XI.M19 Steam Generator Tube Integrity Not applicable XL.M20 Open-Cycle Cooling Water Service Water Integrity Program System [B.1 .281 XlM21 Closed-Cycle Cooling Water Water Chemistry Control -Closed System Cooling Water Program [B. 1.32.31 XL.M22 Boraflex Monitoring Boraflex Monitoring Program [B.1.1]Appendix B Aging Management Programs and Activities Page B-7 Pilgrim Nuclear Power Station License Renewal Application Technical Information (h- I Table B-2 PNPS AMP Correlation with NUREG-1801 Programs (Continued)'
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| NUREG-1801 NUREG-1801 Program PNPS Program Number Inspection of Overhead Heavy XI.M23 Load and Light Load (Related to Not applicable Refueling)
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| Handling Systems XI.M24 Compressed Air Monitoring Not applicable BWR Reactor Water Cleanup XL.M25 SytmNot applicable System XI.M26 Fire Protection Fire Protection Program [B.1.13.1]
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| XI.M27 Fire Water System Fire Water System Program [B.1.13.2]
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| XI.M28 Buried Piping and Tanks Not applicable Surveillance XI.M29 Aboveground Steel Tanks Not applicable XI.M30 Fuel Oil Chemistry Diesel Fuel Monitoring Program[B. 1. 1O]_ RReactor Vessel Surveillance Program XI.M31 Reactor Vessel Surveillance
| |
| [B.1.26]XI.M32 One-Time Inspection One-Time Inspection Program [B.1.23]XL.M33 Selective Leaching of Materials Selective Leaching Program [B.1.27]Xl.M34 Buried Piping and Tanks Buried Piping and Tanks Inspection Inspection Program [B.1.2]One-time Inspection of ASME XI.M35 One-Time Inspection Program [B.1.23]XI.M36 External Surfaces Monitoring System Walkdown Program [B.1.30]XI.M37 Flux Thimble Tube Inspection Not applicable i Page B-8 , Appendix B Aging Management Programs and Activities Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-2 PNPS AMP Correlation with NUREG-1801 Programs (Continued)
| |
| NUREG-1801 NUREG-1801 Program PNPS Program Number Inspection of Internal Surfaces in XI.M38 Miscellaneous Piping and Ducting Not applicable Components XI.M39 Lubricating Oil Analysis Oil Analysis Program [B.1.22]Electrical Cables and Connections Xl E1Not Subject to 10 CFR 50.49 Non-EQ Insulated Cables and.Environmental Qualification Connections Program [B.1.21)Requirements Electrical Cables and Connections Not Subject to 10 CFR 50.49 Non-EQ Instrumentation Circuits Test XI.E2 Environmental Qualification Review Program eB.1 .20]Requirements Used in Instrumentation Circuits Inaccessible Medium-Voltage Xl E3Cables Not Subject to 10 CFR Non-EQ Inaccessible Medium-Voltage
| |
| .50.49 Environmental Qualification Cable Program [B.1.19]Requirements Metal-Enclosed Bus Inspection XI.E4 Metal Enclosed Bus Program [B.1.18]XL.E5 Fuse Holders Not applicable Electrical Cable Connections Not Xl E6 Subject to 10 CFR 50.49 Not applicable
| |
| : l. Environmental Qualification Requirements See plant-specific Inservice Inspection XI.S1 ASME Section Xl, Subsection IWE -Containment Inservice Inspection (CII) Program [B.1.16.1]
| |
| XI.S2 ASME Section Xl, Subsection IWL Not applicable Appendix B Aging Management Programs and Activities Page B-9 Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-2 PNPS AMP Correlation with NUREG-1801 Programs (Continued)
| |
| NUREG-1801 NUREG-1801 Program PNPS Program Number See plant-specific Inservice Inspection XL.S3 ASME Section Xl, Subsection IWF -Inservice Inspection (ISI) Program[B.1.16.2]
| |
| 10 50, Appendix Containment Leak Rate Program XI.S4 1 0 CFR 50 pedxJ[B.1.9]
| |
| XL.S5 Masonry Wall Program Structures Monitoring
| |
| -Masonry Wall Program [B.1.29.1]
| |
| ..e Structures Monitoring
| |
| -Structures XI.56 Structures Monitoring Program Monitoring Program [B.1 .29.2]RG 1.127, Inspection of Water- Structures Monitoring
| |
| -Water Control XI.S7 Control Structures Associated with Structures Monitoring Program Nuclear Power Plants [B.1.29.3]
| |
| XI.S8 Protective Coating Monitoring and Not applicable Maintenance Program Plant-Specific Programs NA Plant-specific program Heat Exchanger Monitoring Program[B.1.15]Inservice Inspection
| |
| -Containment NA Plant-specific program Inservice Inspection (CII) Program[B.1.16.11 NA Plant-specific program .Inservice Inspection
| |
| -Inservice NA Plnt-secifc prgramInspection (151) Program [B.1 .16.2]NA .Plant-specific program Instrument Air Quality Program[B. 1 .17]Periodic Surveillance and Preventive NA Plant-specific program Maintenance Program [B.1.241 NA Plant-specific program Water Chemistry Control -Auxiliary Systems Program [B.1.32.1]
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| __p1, Appendix B Aging Management Programs and Activities II Page B-10 Pilgrim Nuclear Power Station License Renewal Application Technical Information PNPS programs have been compared to the NUREG-1 801 programs with the results being shown in Table B-3 as* programs consistent with NUREG-1801;
| |
| * programs with enhancements;
| |
| * programs with exception to NUREG-1801;
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| * not comparable to NUREG-1801 (plant-specific)
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| Table B-3 PNPS Program Consistency with NUREG-1801 NUREG-1801 Comparison Programs Programs Plant Consistent Programs with with Program Name Specific with Enhancements Exceptions to 1801 NUREG-1801 Boraflex Monitoring Program X Buried Piping and Tanks X Inspection Program BWR CRD Return Line Nozzle x Program BWR Feedwater Nozzle x Program BWR Penetrations Program x BWR Stress Corrosion Cracking X X Program BWR Vessel ID Attachment X Welds Program BWR Vessel Intemals Program X X Containment Leak Rate X Program Diesel Fuel Monitoring Program X X Environmental Qualification (EQ) X of Electric Components Program AppedixB Agng anaemen Prgras an Aciviies ageB-i Appendix B Aging Management Programs and Activities Page B-11 Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-3 PNPS Program Consistency with NUREG-1801 (Continued):
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| NUREG-1801 Comparison Programs Plant Consistent Programs Program Name Programs with with Prga aeSpecific wih Enhancements' Exceptions to NUREG- NUREG-1801 1801 Fatigue Monitoring Program X Fire Protection
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| -Fire Protection X X Program Fire Protection
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| -Fire Water X X System Program Flow-Accelerated Corrosion X Program Heat Exchanger Monitoring X Program Inservice Inspection
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| -X Containment Inservice Inspection (CII) Program Inservice Inspection
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| -Inservice X Inspection (ISI) Program Instrument Air Quality Program X Metal-Enclosed Bus Inspection X Program Non-EQ Inaccessible Medium- X Voltage Cable Program Non-EQ Instrumentation Circuits X Test Review Program Non-EQ Insulated Cables and X Connections Program Oil Analysis Program X X One-Time Inspection Program X 01 Appendix B Aging Management Programs and Activities Page B-12 v Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-3 PNPS Program Consistency with NUREG-1801 (Continued)
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| NUREG-1801 Comparison Programs Programs Plant Consistent Programs with with Program Name Specific with Enhancements Exceptions to NUREG NUREG-1 801 1801 Periodic Surveillance and X Preventive Maintenance Program Reactor Head Closure Studs X Program Reactor Vessel Surveillance X X Program Selective Leaching Program X Service Water Integrity Program X Structures Monitoring
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| -Masonry X Wall Program Structures Monitoring
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| -X X Structures Monitoring Program Structures Monitoring
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| -Water X X Control Structures Monitoring Program System Walkdown Program X Thermal Aging and Neutron X Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program Water Chemistry Control -X Auxiliary Systems Program Water Chemistry Control -BWR X Program Appendix B Aging Management Programs and Activities Page B-13 Pilgrim Nuclear Power Station License Renewal Application Technical Information Table B-3 PNPS Program Consistency with NUREG-1801 (Continued)
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| NUREG-1801 Comparison Programs P Plant Consistent Programs with with Program Name Specific with Enhancements Exceptions to NUREG- NUREG-1 801 1801 Water Chemistry Control -X Closed Cooling Water Program Appendix B Aging Management Programs and Activities Page B-14 Pilgrim Nuclear Power Station License Renewal Application Technical Information AGING MANAGEMENT PROGRAMS AND ACTIVITIES B.1.1 BORAFLEX MONITORING Program Descriition The Boraflex Monitoring Program at PNPS is comparable to the program described in NUREG-1801, Section XL.M22, Boraflex Monitoring.
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| The Boraflex Monitoring Program assures that degradation of the Boraflex panels in the spent fuel racks does not compromise the criticality analysis in support of the design of the spent fuel storage racks. The program relies on periodic inspection of the Boraflex, monitoring of silica levels in the spent fuel pool water, and analysis of criticality to assure that the required 5%subcriticality margin is maintained.
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| NUREG-1801 Consistency The Boraflex Monitoring Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M22, Boraflex Monitoring.
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| Exceptions to NUREG-1801 None Enhancements None Oneratina Experience Blackness testing was performed on Boraflex panels in the spent fuel storage racks during 1996 and 1998 to provide a baseline for development of the monitoring program and assure that the required 5% subcriticality margin is maintained.
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| Results of the 1996 testing showed shrinkage and gapping in the Boraflex, but did not indicate erosion of the Boraflex was occurring.
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| Analysis of the criticality design of the fuel pool based on the observed gap sizes and locations showed a very minor and negligible effect of the gaps on rack reactivity.
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| Therefore, the pool subcriticality margin was greater than 5%. Results of the 1998 testing showed about a 20% increase in average gap size, but overall shrinkage (gaps and end shortening) of the material was much less on a percentage change basis. There were no very large gaps, and the report concluded that the Boraflex poison material in the spent fuel storage racks continues to perform its intended function.The Boraflex Monitoring Program at PNPS has been instituted recently.
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| Therefore, there is no additional plant-specific operating experience.
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| Appendix B Aging Management Programs and Activities Page B-1 5 Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Boraflex Monitoring Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| .:Q Appendix B Aging Management Programs and Activities
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| -Page B-16 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.2 BURIED PIPING AND TANKS INSPECTION Program Descrintion The Buried Piping and Tanks Inspection Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M34, Buried Piping and Tanks Inspection.
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| This program includes (a) preventive measures to mitigate corrosion and (b) inspections to manage the effects of corrosion on the pressure-retaining capability of buried carbon steel, stainless steel, and titanium components.
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| Preventive measures are in accordance with standard industry practice for maintaining external coatings and wrappings.
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| Buried components are inspected when excavated during maintenance.
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| A focused inspection will be performed within the first 10 years of the period of extended operation, unless an opportunistic inspection (or an inspection via a method that allows assessment of pipe condition without excavation) occurs within this ten-year period.NUREG-1801 Consistency The Buried Piping and Tanks Inspection Program at PNPS will be consistent with program attributes described in NUREG-1801, Section XI.M34, Buried Piping and Tanks Inspection, with one exception.
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| Exceptions to NUREG-1801 The Buried Piping and Tanks Inspection Program at PNPS will be consistent with program attributes described in NUREG-1801, Section XI.M34, Buried Piping and Tanks Inspection, with the following exception.
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| Attributes Affected Exception 4. Detection of Aging Effects Inspections via methods that allow assessment of pipe condition without excavation may be substituted for inspections requiring excavation solely for the purpose of inspections Exception Note 1. Methods such as phased array UT technology provide indication of wall thickness for buried piping without excavation.
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| Use of such methods to identify the effects of aging is preferable to excavation for visual inspection, which could result in damage to coating or wrappings.
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| Appendix B Aging Management Programs and Activities Page B-17 Pilgrim Nuclear Power Station License Renewal Application Technical InformationEnhancements None Operating Experience The Buried Piping and Tanks Inspection Program at PNPS is a new program for which there is no operating experience.
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| Conclusion Implementation of the Buried Piping and Tanks Inspection Program will provide reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Page B-18 Appendix B Aging Management Programs and Activities Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.3 BWR CRD RETURN LINE NOZZLE Proaram Description The BWR Control Rod Drive (CRD) Return Line Nozzle Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M6, BWR Control Rod Drive Return Line Nozzle.Under this program, PNPS has cut and capped the CRD return line nozzle to mitigate cracking, and continues Inservice Inspection (ISI) examinations to monitor the effects of crack initiation and growth on the intended function of the control rod drive return line nozzle and cap.In 2003, a structural weld overlay was installed over a crack in the CRD return line nozzle-to-cap weld. The Inconel 52 weld metal used in the overlay is highly resistant to stress corrosion cracking.NUREG-1801 Consistency The BWR CRD Return Line Nozzle Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M6, BWR Control Rod Drive Return Line Nozzle, with exceptions.
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| Exceptions to NUREG-1801 The BWR CRD Return Line Nozzle Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M6, BWR Control Rod Drive Return Line Nozzle, with the following exceptions.
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| Attributes Affected Exceptions
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| : 3. Parameters Monitored/
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| PNPS examines 1/2 inch of the volume next Inspected to the widest part of the N10 nozzle-to-vessel weld, rather than half of the vessel wall thickness.
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| 1 4. Detection of Aging Effects The extent and schedule of inspection, as 5. Monitoring and Trending delineated in NUREG 0619, are not followed.
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| Specifically, liquid penetrant testing (PT) of CRDRL nozzle blend radius and bore regions is not performed.
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| 2 6. Acceptance Criteria PNPS repaired the CRDRL nozzle by weld overlay rather than removing the crack by grinding and examines the overlay using UT in lieu of RT.3 Exception Notes Appendix B Aging Management Programs and Activities Page B-19 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 1. Extending the examination volume into the base metal as required by ASME Section Xl, 1998 Edition, 2000 Addenda, Figure IWB-2500-7(b) prolongs the examination time significantly and results in no net increase in safety. The extra volume is base metal region which is not prone to in-service cracking and has been extensively examined before the vessel was put into service and during the first, second and third interval examinations.
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| : 2. The weld overlay installed over a crack in the CRD return line nozzle-to-cap weld covers the nozzle, the nozzle-to-cap weld, and part of the cap. The Inconel 52 weld overlay, which is highly resistant to stress corrosion cracking, is ultrasonically inspected in accordance with GL 88-01 and BWRVIP-75.
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| The weld overlay provides reasonable assurance of structural and pressure boundary integrity of the RPV capped N10 nozzle and, thus, provides an acceptable level of quality and safety.Since the nozzle and original nozzle-to-cap weld are covered by the overlay, and the overlay is examined, examination of the nozzle and original nozzle-to-cap weld is not required.3. In its letter of February 25, 2005, the NRC concluded that the proposed alternative provides reasonable assurance of structural and pressure boundary integrity of the RPV capped NIO nozzle and, thus, provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorized the use of ASME Code Case N-504-2, as modified, and the use of UT in lieu of RT, to perform a weld overlay repair of the CRD return line nozzle-to-cap weld (N 10).Enhancements None Operating Exnerience On October 1, 2003, a reactor coolant pressure boundary leak from the NIO nozzle-to-cap weld area was identified during a planned visual inspection of the drywell. Through-wall leakage from the N1O nozzle-to-cap butt weld was caused by an incipient crack or crevice condition remaining in the weld after repair welding performed as part of the nozzle-to-cap fabrication welding in 1977. Subsequent crack propagation continued through-wall by an interdendritic stress corrosion cracking mechanism due to high residual weld stresses in the Inconel 82/182 weld metal as a result of the repair. A structural weld overlay was installed with Inconel 52 weld metal, which is highly resistant to stress corrosion cracking.
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| The weld overlay process also imparts a compressive residual stress due to the welding process, which prevents further crack growth.The N10 nozzle-to-cap weld received all code-required preservice NDE examinations and was pressure tested prior to returning to service. Ultrasonic examinations have the capability to detect incipient cracking including hard-to-detect flaws related to stress corrosion cracking mechanisms and flaws that occur entirely within the weld metal. Thus, the examinations would have detected weld cracking.
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| Since the weld overlay is highly resistant to cracking, and will Appendix B Aging Management Programs and Activities Page B-20 Pilgrim Nuclear Power Station License Renewaf Application Technical Information continue to be examined as required, the BWR CRD Return Line Nozzle Program remains effective for managing the effect of cracking on the intended function of the CRD return line nozzle.Conclusion The BWR CRD Return Line Nozzle Program has been effective at managing aging effects. The BWR CRD Return Line Nozzle Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-21 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.4 BWR FEEDWATER NOZZLE Proaram Description The BWR Feedwater Nozzle Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M5, BWR Feedwater Nozzle.Under this program, PNPS has removed feedwater blend radii flaws, removed feedwater nozzle cladding, and installed a triple-sleeve-double-piston sparger to mitigate cracking.
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| This program continues enhanced inservice inspection (ISI) of the feedwater nozzles in accordance with the requirements of ASME Section Xl, Subsection IWB and the recommendation of General Electric (GE) NE-523-A71-0594 to monitor the effects of cracking on the intended function of the feedwater nozzles.NUREG-1801 Consistency The BWR Feedwater Nozzle Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M5, BWR Feedwater Nozzle, with exceptions.
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| Excentions to NUREG-1801 The BWR Feedwater Nozzle Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M5, BWR Feedwater Nozzle, with the following exceptions.
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| Attributes Affected Exceptions
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| : 2. Preventive Actions A low-flow controller was not installed and the reactor water cleanup system was not rerouted.1 3. Parameters/Monitored PNPS reduced the examination volume Inspected next to the widest part of the feedwater nozzle-to-vessel welds from half of the vessel wall thickness to 1/2"1.2 Exception Notes Appendix B Aging Management Programs and Activities Page B-22 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 1. In its safety evaluation of BWR feedwater and CRD return line modifications at PNPS, NRC noted that the intent of the requirements of NUREG-0619 and NEDE-21821-A had been satisfied with the PNPS modifications.
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| Since the stainless steel cladding has been removed and the improved spargers have been installed, an adequate margin of safety against feedwater nozzle crack growth exists. Therefore, NRC concluded that, with continued inspections to monitor for crack initiation and growth, PNPS can operate without rerouting the RWCU and without installing a low-flow controller for the feedwater system. Since inspections to monitor for crack initiation and growth will continue, this conclusion remains valid for the period of extended operation.
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| : 2. Extending the examination volume into the base metal as required by ASME Section Xl, 1998 Edition, 2000 Addenda, Figure IWB-2500-7(b) prolongs the examination time significantly and results in no net increase in safety. The extra volume is base metal region which is not prone to in-service cracking and has been extensively examined before the vessel was put into service and during the first, second and third interval examinations.
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| Enhancements None Operating Experience In October, 1989 it was discovered that feedwater nozzles were not being examined with scans designed for the bore. Procedures were revised and subsequent examinations were performed in accordance with NUREG-0619.
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| Since feedwater nozzle bores have subsequently been examined without recordable indications, and will continue to be examined as required, this programmatic error did not impact the ability of the BWR Feedwater Nozzle Program to manage the effect of cracking on the intended function of the feedwater nozzles.Ultrasonic testing of the feedwater nozzles during RFO14 (April, 2003) resulted in no recordable indications.
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| Absence of recordable indications on the feedwater nozzles provides evidence that the program is effective for managing cracking of the nozzles.Conclusion The BWR Feedwater Nozzle Program has been effective at managing aging effects. The BWR Feedwater Nozzle Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-23 .
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.5 BWR PENETRATIONS Program Description The BWR Penetrations Program at PNPS is comparable to the program described in NUREG-1 801, Section XL.M8, BWR Penetrations.
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| The program includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) documents BWRVIP-27 and BWRVIP-49 and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 to ensure the long-term integrity of vessel penetrations and nozzles.NUREG-1801 Consistency The BWR Penetrations Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M8, BWR Penetrations, with exceptions.
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| Exceptions to NUREG-1801 The BWR Penetrations Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M8, BWR Penetrations with the following exceptions.(up Attributes Affected Exceptions
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| : 1. Scope of Program Surface examinations are not performed on 3. Parameters Monitored/Inspected instrument penetration nozzle welds. In 4. Detection of Aging Effects accordance with ASME Section Xl, Code Case N-578 for elements classified as low risk, inspections to monitor the effects of cracking on the intended function of instrument penetration nozzles (N15A/B and N16A/B) include enhanced visual (VT-2 with insulation removed) examinations during system pressure testing. Also, a UT exam of the N1 6B safe end-to-reducer weld is performed once every 10 years.However, ASME Section Xl, Table IWB-2500-1 and BWRVIP-49 (by reference) also recommend surface examinationrs.
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| Appendix B Aging Management Programs and Activities P Page B-24 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 3. Parameters Monitored/
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| Table IWB-2500-1 from the 1998 edition Inspected with 2000 addenda of ASME Section Xt is used, while NUREG-1801 specifies the 2001 edition with 2002 and 2003 addenda.2 Exception Notes 1. PNPS has implemented risk informed ISI (RI-ISI) in accordance with ASME Section Xl, Code Case N-578. The overall risk to the plant is reduced when RI-ISI is applied because the process concentrates on examining welds that have the greatest risk in terms of consequences of failure and potential degradation.
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| In addition, RI-ISI examinations are focused on those examination volumes where flaws are most likely to be located. As such, RI-ISI does a better job in capturing risk than existing ASME Section Xl requirements, which are based on design stresses and random selection.
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| Also, PNPS replaced the original IGSCC-susceptible 304 stainless steel safe end extensions for the N15 and N16 nozzles with more IGSCC-resistant Inconel material.2. Since ASME Section Xl through the 2003 Addenda has been accept by reference in 10 CFR 50.55a paragraph (b) (2) without modification or limitation on use of Table IWB-2500-1 from the 1998 edition with 2000 addenda for BWR components, use of this version is appropriate to assure that components crediting this program can perform their intended function consistent with the current licensing basis during the period of extended operation.
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| Enhancements None Operating Experience In January 2005 three 2Y2" piping butt welds in SLC system piping adjacent to nozzle N14 were found to be unidentified on inspection drawings and not included in ISI weld population totals.Two of the welds (RPV-N14-T1 and RPV-N14-T2) are shop welds in a vendor supplied tee. The third weld (RPV-14-2) is the connection field weld between the tee and the SLC nozzle (N14)safe end extension piece. This weld was included in surface examinations of the N14 nozzle safe end weld and safe end extension piece performed in RF11. Corrective actions included adding the welds to ISI weld population totals and performing a nozzle surface examination of weld RPV-N14-2 during RFO15. Since RPV-N14-2 has been examined without recordable indications, and will continue to be examined as required, this programmatic error did not impact the ability of the BWR Penetrations Program to manage the effect of cracking on the intended function of the SLC nozzle.Inservice examination of the SLC nozzle, (including weld RPV-N14-2 as discussed above), during RFO15 (April, 2005) resulted in no recordable indications.
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| Absence of recordable Appendix B Aging Management Programs and Activities Page B-25 Pilgrim Nuclear Power Station License Renewal Application Technical Information indications on the SLC nozzle and adjacent welds provides evidence that the program is effective for managing cracking of the nozzle.(%-Liquid penetrant examination of instrument penetration nozzle N15A in 1990 resulted in no recordable indications.
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| Absence of recordable indications on the instrument nozzles provides evidence that the program is effective for managing cracking of the instrument penetration nozzles.Inservice examination of instrument penetration nozzles during RFO15 (April, 2005) resulted in no recordable indications.
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| Absence of recordable indications on the instrument nozzles provides evidence that the program is effective for managing cracking of the nozzles.Conclusion The BWR Penetrations Program has been effective at managing aging effects. The BWR Penetrations Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| (4i0 Appendix B Aging Management Programs and Activities Page B-26 W Pilgrim Nuclear Power Station License Renewal Application
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| ' ITechnical Information B.1.6 BWR STRESS CORROSION CRACKING Program Descrintion The BWR Stress Corrosion Cracking Program at PNPS is comparable to the program described in NUREG-1801, Section Xl.M7, BWR Stress Corrosion Cracking.The program includes (a) preventive measures to mitigate intergranular stress corrosion cracking (IGSCC), and (b) inspection and flaw evaluation to monitor IGSCC and its effects on reactor coolant pressure boundary components made of stainless steel or CASS.NUREG-1801 Consistency The BWR Stress Corrosion Cracking Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M7, BWR Stress Corrosion Cracking, with an exception and an enhancement.
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| Excentions to NUREG-1801 The BWR Stress Corrosion Cracking Program at PNPS is consistent with the program described in NUREG-1801, Section XLI.M7, BWR Stress Corrosion Cracking with the following exception.
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| Attributes Affected Exception 6. Acceptance Criteria The 1998 edition with 2000 addenda of ASME Section XI, Subsection IWB-3600 is used for flaw evaluation, while NUREG-1801 specifies the 1986 edition of ASME Section XI, Subsection IWB-3600 for flaw evaluation.
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| 1 Exception Note 1. Since ASME Section Xl through the 2003 Addenda has been accept by NRC in 10 CFR 50.55a paragraph (b) (2) without modification or limitation on use of subsection IWB-3600 from the 1998 edition with 2000 addenda, use of this version for flaw evaluation is appropriate to assure that components crediting this program can perform their intended function consistent with the current licensing basis during the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-27 Pilgrim Nuclear Power Station License Renewal Application Technical Information Enhancements The following enhancement will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancement
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| : 5. Monitoring and Trending The implementing procedure for ASME Section Xl inservice inspection and testing will be enhanced to specify that the guidelines in Generic Letter 88-01 or approved BWRVIP-75 shall be considered in determining sample expansion if indications are found in Generic Letter 88-01 welds.Operating Experience Ultrasonic examinations of GL 88-01 nozzle safe end welds and austenitic stainless steel reactor coolant piping with 4" and greater nominal diameter and operating temperature greater than 200*F during RFO14 (April, 2003) resulted in no recordable indications.
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| Absence of recordable indications on the nozzles and piping provides evidence that the program is effective for managing cracking of austenitic stainless steel components.
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| Ultrasonic examinations of nozzle safe end welds and austenitic stainless steel reactor coolant piping with 4" and greater nominal diameter and operating temperature greater than 200"F during RFO15 (April 2005) resulted in no recordable indications.
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| Absence of recordable indications on the nozzles and piping provides evidence that the program is effective for managing cracking of the nozzles and piping.Conclusion The BWR Stress Corrosion Cracking Program has been effective at managing aging effects.The BWR Stress Corrosion Cracking Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-28 Q."'
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.7 BWR VESSEL ID ATTACHMENT WELDS Program Description The BWR Vessel ID Attachment Welds Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M4, BWR Vessel ID Attachment Welds.The program includes (a) inspection and flaw evaluation in accordance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) BWRVIP-48 and (b)monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 (EPRI Report 1008192) to ensure the long-term integrity and safe operation of reactor vessel inside diameter (ID) attachment welds and support pads.NUREG-1801 Consistency The BWR Vessel ID Attachment Welds Program at PNPS is consistent with the program described in NUREG-1 801, Section XI.M4, BWR Vessel ID Attachment Welds with one exception.
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| Exceptions to NUREG-1801 The BWR Vessel ID Attachment Welds Program at PNPS is consistent with the program i ) described in NUREG-1 801, Section XI.M4, BWR Vessel ID Attachment Welds with the following exception.
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| Attributes Affected Exception 3. Parameters Monitored/
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| Table IWB-2500-1 from the 1998 edition Inspected with 2000 addenda of ASME Section Xl is used, while NUREG-1 801 specifies the 2001 edition with 2002 and 2003 addenda.1 Exception Note 1. Since ASME Section Xi through the 2003 Addenda has been accept by reference in 10 CFR 50.55a paragraph (b) (2) without modification or limitation on use of Table IWB-2500-1 from the 1998 edition with 2000 addenda for BWR components, use of this version is appropriate to assure that components crediting this program can perform their intended function consistent with the current licensing basis during the period of extended operation.
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| Enhancements None Appendix B Aging Management Programs and Activities Page B-29 Pilgrim Nuclear Power Station License Renewal Application Technical Information/Operating Experience Visual and enhanced visual examinations of vessel attachment welds (feedwater bracket attachment and jet pump riser braces) during RFO14 (April, 2003) resulted in no recordable indications.
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| Previous visual and enhanced visual examinations of vessel attachment welds resulted in no recordable indications.
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| Absence of recordable indications on the vessel attachment welds provides evidence that the program is effective for managing cracking of the welds.Visual and enhanced visual examinations of vessel attachment welds (core spray piping bracket, guide rod bracket attachment, steam dryer support brackets, steam dryer hold-down brackets, and surveillance specimen holder brackets) during RFO15 (April, 2005) resulted in no recordable indications.
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| Absence of recordable indications on the vessel attachment welds provides evidence that the program is effective for managing cracking of the welds.Conclusion The BWR Vessel ID Attachment Welds Program has been effective at managing aging effects.The BWR Vessel ID Attachment Welds Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-30 i Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.8 BWR VESSEL INTERNALS Program Description The BWR Vessel Internals Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M9, BWR Vessel Internals.
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| The program includes (a) inspection, flaw evaluation, and repair in conformance with the applicable, staff-approved BWR reactor vessel and internals project (BWRVIP) documents, and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-1 30 to ensure the long-term integrity of vessel internals components.
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| NUREG-1801 Consistency The BWR Vessel Internals Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M9, BWR Vessel Internals, with exceptions and an enhancement.
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| Exceptions to NUREG-1801 The BWR Vessel Internals Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M9, BWR Vessel Internals, with the following exceptions.
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| Attributes Affected Exceptions
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| : 1. Scope of Program Low-pressure Coolant Injection (LPCI)4. Detection of Aging Effects Coupling BWRVIP-42 guidelines are not applicable to PNPS.1 1. Scope of Program Top Guide 4. Detection of Aging Effects Inspection of the four top guide hold-down assemblies and four top guide aligner assemblies Is not performed at PNPS.2 The top guide rim weld does not exist at PNPS and is therefore exempt.1. Scope of Program Core Spray 4. Detection of Aging Effects PNPS defers inspection of three inaccessible welds inside each of the two core spray nozzles until a delivery system for ultrasonic testing of the hidden welds is developed.
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| Thus, PNPS does not meet the BWRVIP-18 requirement to perform an ultrasonic inspection of a full target weld set every other refueling outage.3 Appendix B Aging Management Programs and Activities Page B-31 Pilgrim Nuclear Power Station License Renewal Application Technical Information ( i 1. Scope of Program Jet Pump Assembly 4. Detection of Aging Effects PNPS defers inspection of jet pump inaccessible welds until a delivery system for ultrasonic testing of the hidden welds is developed.
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| Thus, PNPS does not meet the BWRVIP-41 requirement to perform a modified VT-1 of 100% of these welds over two 6-year inspection cycles and 25% per inspection cycle thereafter.
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| 4 3. Parameters Monitored/
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| Table IWB-2500-1 from the 1998 edition Inspected with 2000 addenda of ASME Section Xl is used, while NUREG-1801 specifies the 2001 edition with 2002 and 2003 addenda.5 Exception Notes 1. BWRVIP-42 provides guidelines for inspection and evaluation of the low-pressure coolant injection (LPCI) coupling.
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| PNPS has no LPCI coupling.2. PNPS has a plant-specific analysis to account for plant-specific dynamic loading of the top guide hold-down and aligner assemblies, which concludes that less than 20%of the weld area on the top guide hold-down and aligner assemblies is needed to resist load. Therefore, in accordance with Table 3-2 of BWRVIP-26, inspection of the four top guide hold-down assemblies and four top guide aligner assemblies is not performed at PNPS.CW 3. Inspection of similar creviced and uncreviced welds; including junction box-to-pipe welds, upper elbow welds, junction box cover plate weld, P1 weld, and downcomer sleeve welds; showed no indication of cracking.
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| Therefore, deferral of inspection of the inaccessible welds is justified.
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| : 4. The hidden jet pump welds are far enough into the nozzle that failure at these welds would not result in the thermal sleeve disengaging from the nozzle before the riser contacted the shroud. If the jet pump thermal sleeve severed, the riser brace would maintain the geometry of the jet pump well past the time that leakage would be detected through operational parameters and the plant could be safely shut down.In addition, PNPS instituted hydrogen water chemistry in 1991 to mitigate cracking in the reactor intemals, and to address crack growth in the jet pump thermal sleeve welds in particular.
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| Therefore, deferral of inspection of the inaccessible welds is justified.
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| Appendix B Aging Management Programs and Activities Page B-32 Q.." i Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 5. Since ASME Section Xl through the 2003 Addenda has been accept by reference in 10 CFR 50.55a paragraph (b) (2) without modification or limitation on use of Table IWB-2500-1 from the 1998 edition with 2000 addenda for BWR components, use of this version is appropriate to assure that components crediting this program can perform their intended function consistent with the current licensing basis during the period of extended operation.
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| Enhancements The following enhancement will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancement
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| : 1. Scope of Program The PNPS top guide fluence is projected to exceed the threshold for IASCC (5x10 2 0 n/cm 2) prior to the period of extended operation.
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| Therefore, ten (10) percent of the top guide locations will be inspected using enhanced visual inspection technique, EVT-1, within the first 12 years of the period of extended operation, with one-half of the inspections (50 percent of locations) to be completed within the first 6 years of the period of extended operation.
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| Locations selected for examination will be areas that have exceeded the neutron fluence threshold.
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| Operating Experience Visual and enhanced visual examinations of vessel internals (shroud support plate gusset welds, core spray piping, jet pump riser braces, jet pump diffusers, CRD guide tube handle attachment, steam dryer, and feedwater spargers) during RFO14 (April, 2003) resulted in no new recordable indications.
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| Previous visual and enhanced visual examinations of vessel internals revealed indications on core spray piping welds, and steam dryer leveling screw tack welds. Absence of new recordable indications on the vessel internals provides evidence that the program is effective for managing cracking of the welds.Visual and enhanced visual examinations of vessel internals (core spray piping welds, core spray spargers, integrally welded core support structures, jet pump restrainer wedges, shroud vertical welds, shroud top guide ring, shroud support, steam dryer, steam dryer level screw tack weld cracks, steam separator/shroud head, and top guide grid beams) during RFO15 (April, 2005)resulted in no new recordable indications.
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| Absence of new recordable indications on the vessel internals provides evidence that the program is effective for managing cracking of the welds.Appendix B Aging Management Programs and Activities Page B-33 Pilgrim Nuclear Power Station License Renewal Application Technical Information The core shroud provides 2/3-core coverage in case of a LOCA. Because IGSCC cracking of sensitized shroud welds was an industry issue, PNPS implemented a preemptive shroud hold-down modification during RFO10 in 1995.Conclusion The BWR Vessel Internals Program has been effective at managing aging effects. The BWR Vessel Internals Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-34 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.9 CONTAINMENT LEAK RATE Proaram Description The Containment Leak Rate Program at PNPS is comparable to the program described in NUREG-1801, Section XI.S4, 10 CFR 50, Appendix J.As described in 10 CFR Part 50, Appendix J, containment leak rate tests are required to assure that (a) leakage through primary reactor containment and systems and components penetrating primary containment shall not exceed allowable values specified in technical specifications or associated bases and (b) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of containment, and systems and components penetrating primary containment.
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| NUREG-1801 Consistency The Containment Leak Rate Program at PNPS is consistent with the program described in NUREG-1801, Section XI.S4, 10 CFR Part 50, Appendix J.Exceptions to NUREG-1801 None Enhancements None Operating Experience During the most recent integrated leakage testing of primary containment, as-found and as-left test data met all applicable test acceptance criteria, indicating that the program is effective at managing the effects of loss of material and cracking on primary containment components.
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| QA audits in 2000 and 2005 revealed no issues or findings that could impact effectiveness of the program.Conclusion The Containment Leak Rate Program has been effective at managing aging effects. The Containment Leak Rate Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-35 Pilgrim Nuclear Power Station License Renewal Application Technical InformationB.1.10 DIESEL FUEL MONITORING Program Description The Diesel Fuel Monitoring Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M30, Fuel Oil Chemistry Program.The program entails sampling to ensure that adequate diesel fuel quality is maintained to prevent plugging of filters, fouling of injectors, and corrosion of fuel systems. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by periodic draining and cleaning of tanks and by verifying the quality of new oil before its introduction into the storage tanks. Sampling and analysis activities are in accordance with technical specifications on fuel oil purity and the guidelines of ASTM Standards D4057-81 and D975-81 (or later revisions of these standards).
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| NUREG-1801 Consistency The Diesel Fuel Monitoring Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M30, Fuel Oil Chemistry Program, with exceptions and enhancements.
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| Exceptions to NUREG-1801 The Diesel Fuel Monitoring Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M30, Fuel Oil Chemistry Program, with the following exceptions.
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| C Attributes Affected Exceptions
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| : 1. Scope of Program The guidelines of ASTM Standard D6217 6. Acceptance Criteria are not used along with those of D2276 for determination of particulates.
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| : 2. Preventive Actions No additives are used beyond what the refiner adds during production.
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| 2 2. Preventive Actions The security diesel generator fuel storage tank is not periodically cleaned and inspected because the internals are inaccessible.
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| 3 3. Parameters Monitored/
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| Determination of particulates maybe Inspected according to ASTM Standard D2276, rather 6. Acceptance Criteria than modified ASTM D2276 Method A.4 Exception Notes Appendix B Aging Management Programs and Activities
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| .Page B-36 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 1. PNPS technical specifications specify use of ASTM D975-81, which recommends use of ASTM D2276. Therefore, the guidelines of D2276 are appropriate for determination of particulates.
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| : 2. PNPS does not add biocides, stabilizers, or corrosion inhibitors to the diesel fuel.Plant-specific operating experience has not indicated significant problems related to MIC. Since water contamination in the diesel fuel storage tanks is minimized, the potential for MIC is limited.3. The security diesel fuel storage tank does not have manways or other means of access to the internals.
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| : 4. Determination of particulates maybe according to ASTM Standard D2276 which conducts particulate analysis using a 0.8 micron filter, rather than the 3.0 micron filter specified in NUREG-1801.
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| Use of a filter with a smaller pore size results in a larger sample of particulates since smaller particles are retained.
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| Thus, use of a 0.8 micron filter is more conservative than use of the 3.0 micron filter specified in NUREG-1801.
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| Enhancements The following enhancements will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancements
| |
| : 1. Scope of Program The Diesel Fuel Monitoring Program will be enhanced to include periodic sampling of the security diesel generator fuel storage tank, near the bottom, to determine water content.4. Detection of Aging Effects The Diesel Fuel Monitoring Program will be enhanced to include periodic ultrasonic measurement of the bottom surface of the security diesel generator fuel storage tank to ensure that significant degradation is not occurring.
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| : 6. Acceptance Criteria UT measurements of tank bottom surfaces will have acceptance criterion
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| > 60% Tnom.Operating Experience In 2001, two diesel fuel oil deliveries were rejected; one because the oil viscosity was too low and one because the oil had detectable visible particulate contamination.
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| Rejection of inferior fuel shipments maintains diesel fuel quality to prevent loss of material and cracking of fuel system components.
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| Appendix B Aging Management Programs and Activities Page B-37 Pilgrim Nuclear Power Station License Renewal Application Technical Information Monthly sampling of the B EDG fuel oil tank and the B SBO fuel oil tank in August, 2003 indicated a small amount of water was in the tanks. Gaskets were replaced although the indication of water was determined to be a false positive.
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| The tanks were confirmed to be water-free during subsequent testing. Sampling of the B EDG fuel oil tank in January 2005 indicated a small amount of water was in the tank. However, subsequent testing confirmed the tank to be water-free. Other fuel oil sampling results from 2000 through August 2005 reveal that fuel oil quality is being maintained in compliance with acceptance criteria.
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| A 1998 visual and ultrasonic inspection of A and B diesel fuel oil storage tank internals revealed no degradation.
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| A 2002 visual inspection of A and B SBO fuel oil storage tank internals revealed no degradation.
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| Continuous confirmation of diesel fuel quality, timely corrective actions, and absence of degradation in the fuel oil storage tanks provide evidence that the program is effective in managing loss of material and cracking of fuel system components.
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| Conclusion The Diesel Fuel Monitoring Program has been effective at managing aging effects. The Diesel Fuel Monitoring Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Q Appendix B Aging Management Programs and Activities Page B-38 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.11 ENVIRONMENTAL QUALIFICATION OF ELECTRIC COMPONENTS Proaram Description The Environmental Qualification (EQ) of Electric Components Program at PNPS is comparable to the program described in NUREG-1801, Section X.E1, Environmental Qualification (EQ) of Electric Components.
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| The Nuclear Regulatory Commission (NRC) has established nuclear station environmental qualification (EQ) requirements in 10 CFR Part 50, Appendix A, Criterion 4, and 10 CFR 50.49.10 CFR 50.49 specifically requires that an EQ program be established to demonstrate that certain electrical components located in harsh plant environments (that is, those areas of the plant that could be subject to the harsh environmental effects of a loss of coolant accident[LOCA], high energy line breaks [HELBs] or post-LOCA radiation) are qualified to perform their safety function in those harsh environments.
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| 10 CFR 50.49 requires that the effects of significant aging mechanisms be addressed as part of environmental qualification.
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| The PNPS EQ program manages the effects of thermal, radiation, and cyclic aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current license term are refurbished, replaced, or their qualification is extended prior to reaching the aging limits established in the evaluation.
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| Aging evaluations for EQ components are considered time-limited aging analyses (TLAAs) for license renewal.NUREG-1801 Consistency The Environmental Qualification (EQ) of Electric Components Program at PNPS is consistent with the program described in NUREG-1801, Section X.E1, Environmental Qualification (EQ) of Electric Components.
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| Exceotions to NUREG-1801 None Enhancements None Or1eratina ExRerience The overall effectiveness of the Environmental Qualification (EQ) of Electric Components Program is demonstrated by the excellent operating experience for systems, structures, and components in the program. The program has been subject to periodic internal and external assessments that have resulted in program improvement.
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| Appendix B Aging Management Programs and Activities Page B-39 Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Environmental Qualification (EQ) of Electric Components Program has been effective at-managing aging effects. The Environmental Qualification (EQ) of Electric Components Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| (_,-Page B-40 1 Appendix B Aging Management Programs and Activities Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.12 FATIGUE MONITORING Program Description The Fatigue Monitoring Program at PNPS is comparable to the program described in NUREG-1801, Section X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary.In order not to exceed design limits on fatigue usage, the Fatigue Monitoring Program tracks the number of critical thermal and pressure transients for selected reactor coolant system components.
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| The program ensures the validity of analyses that explicitly assumed a specified number of thermal and pressure fatigue transients by assuring that the actual effective number of transients is not exceeded.NUREG-1801 Consistency The Fatigue Monitoring Program at PNPS is consistent with the program described in NUREG-1801, Section X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary, with exceptions.
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| Exceptions to NUREG-1801 The Fatigue Monitoring Program at PNPS is consistent with the program described in NUREG-1801, Section X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary, with the following Li exceptions.
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| Attributes Affected Exceptions
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| : 2. Preventive Actions The Fatigue Monitoring Program only involves tracking the number of transient cycles and does not include assessment of the impact of the reactor water environment on critical components.
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| 1 4. Detection of Aging Effects The PNPS program does not provide for periodic update of the fatigue usage calculations.
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| 2 Exception Notes 1. The effect of the reactor water environment on fatigue is addressed as described in Section 4.3.3.2. Updates of fatigue usage calculations are not necessary unless the number of accumulated fatigue cycles approaches the number of assumed design cycles. The PNPS program provides for periodic assessment of the number of accumulated cycles. If a design cycle assumption is approached, corrective action is taken which may include update of the fatigue usage calculation.
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| Appendix B Aging Management Programs and Activities Page B-41 Pilgrim Nuclear Power Station License Renewal Application Technical Information Enhancements None Operating Experience:
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| Industry experience has been factored into the PNPS fatigue monitoring program through incorporation of Regulatory Guides and BWRVIP documents.
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| The locations at which CUFs are calculated include those identified in NUREG/CR-6260.
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| Industry experience has identified thermal stresses that were not considered in the original design of PNPS. These thermal stresses have been evaluated.
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| PNPS will continue to evaluate future industry experience on fatigue of Class 1 components.
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| For recent reactor shutdowns and startups, cycle limitations did not trend toward exceeding the allowable number of cycles. This demonstrates that the program continues to monitor plant transients and track the accumulation of these transients.
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| Conclusion The Fatigue Monitoring Program has been demonstrated to maintain the validity of the fatigue design basis for reactor coolant system components designed to withstand the effects of cyclic loads due to reactor system temperature and pressure changes.The Fatigue Monitoring Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-42 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.13 FIRE PROTECTION The fire protection programs for PNPS include the Fire Protection Program and the Fire Water System Program. These two programs are comparable to NUREG-1801, Section XL.M26, Fire Protection and NUREG-1 801, Section XI.M27, Fire Water System, respectively.
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| The Fire Protection programs are discussed in more detail in the following subsections
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| * Fire Protection
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| * Fire Water System B.1.13.1 FIRE PROTECTION Proaram Description The Fire Protection Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M26, Fire Protection.
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| The fire protection program includes a fire barrier inspection and a diesel-driven fire pump inspection.
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| The fire barrier inspection requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained.
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| The diesel-driven fire pump inspection requires that the pump be periodically tested to ensure that the fuel supply line can perform its intended function.
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| The program also includes periodic inspection and testing of the Halon fire suppression system.Corrective actions, confirmation process, and administrative controls in accordance with the requirements of 10 CFR 50 Appendix B are applied to the Fire Protection Program.NUREG-1801 Consistency The Fire Protection Program at PNPS is consistent with the program described in NUREG-1 801, Section XLI.M26, Fire Protection, with exceptions and enhancements.
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| Exceptions to NUREG-1801 The Fire Protection Program at PNPS is consistent with the program described in NUREG-1 801, Section XI.M26, Fire Protection with the following exceptions.
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| Attributes Affected Exceptions
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| : 1. Scope of Program This program is not necessary to manage aging effects for carbon dioxide fire, protection system components.'
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| Appendix B Aging Management Programs and Activities IPage B-43 Pilgrim Nuclear Power Station License Renewal Application Technical Information Q.." 4. Detection of Aging Effects The NUREG-1801 program states that approximately 10% of each type of penetration seal should be visually inspected at least once every refueling outage. The PNPS program specifies inspection of approximately 20% of the seals each operating cycle, with all accessible fire barrier penetration seals being inspected at least once every five operating cycles.2 Exception Notes 1. The carbon dioxide fire suppression system is not subject to aging management review.2. Since aging effects are typically manifested over several years, this variation in inspection frequency is insignificant.
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| Enhancements The following enhancements will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancements
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| : 3. Parameters Monitored/
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| Procedures will be enhanced to state that Inspected the diesel engine sub-systems (including
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| : 6. Acceptance Criteria the fuel supply line) shall be observed while the pump is running. Acceptance criteria will be enhanced to verify that the diesel engine did not exhibit signs of degradation while it was running; such as fuel oil, lube oil, coolant, or exhaust gas leakage.3. Parameters Monitored/
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| The procedure for Halon system functional Inspected testing, will be enhanced to state that the 6. Acceptance Criteria Halon 1301 flex hoses shall be replaced if leakage occurs during the system functional test.C)Operating Experience Inspections of fire stops, fire barrier penetration seals, fire barrier walls, ceilings, and floors from 1998 through 2004, revealed signs of degradation such as cracks, gaps, voids, holes or missing material.
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| Identification of degradation and corrective action prior to loss of intended function Appendix B Aging Management Programs and Activities Page B-44 i 4J Pilgrim Nuclear Power Station License Renewal Application Technical Information provide evidence that the program is effective for managing aging effects for fire barrier components.
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| Visual inspections and functional tests of fire doors, from 1998 through 2004, detected degradation of fire doors, such as corrosion, wear and missing parts. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for fire doors.Observation of the diesel-driven fire pump during a performance test in 2000 revealed leakage from the cooling system. The cause was determined to be corrosion of the heat exchanger shell, which was repaired.
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| Observation of the diesel-driven fire pump during performance tests in 2001 revealed degradation of several components in the engine oil and coolant systems. The pump also failed a flow test. Therefore, the entire assembly (engine, controller, and pump) was replaced in 2002. Identification of degradation and corrective action provide evidence that the program is effective for managing aging of diesel-driven fire pump subsystem components.
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| Recent (2002 and 2003) visual inspections of cable spreading room Halon cylinders, associated hoses, valves and piping, detected no evidence of damage or corrosion.
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| Absence of cracks or corrosion provides evidence that the program is effective for managing aging effects for cable spreading room Halon system components.
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| On July 31, 2003, NRC completed a triennial fire protection team inspection to assess whether PNPS has implemented an adequate fire protection program and that post-fire safe shutdown capabilities have been established and are being properly maintained at PNPS. Results confirmed that PNPS was maintaining the fire protection systems in accordance with their fire protection program and that PNPS was identifying program deficiencies and implementing appropriate corrective actions. The team also evaluated the material condition of fire walls, fire doors, fire dampers and fire barrier penetration seals and concluded that PNPS was maintaining passive features in a state of readiness.
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| A QA audit in May 2004 and an NRC inspection in June 2005 revealed no issues or findings that could impact effectiveness of the program to manage aging effects for fire protection components.
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| Conclusion The Fire Protection Program has been effective at managing aging effects. The Fire Protection Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B45 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.13.2 FIRE WATER SYSTEM Program Description The Fire Water System Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M27, Fire Water System.This aging management program applies to water-based fire protection systems that consist of sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes, and aboveground and underground piping and components that are tested in accordance with applicable National Fire Protection Association (NFPA) codes and standards.
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| Such testing assures functionality of systems. Also, many of these systems are normally maintained at required operating pressure and monitored such that leakage resulting in loss of system pressure is immediately detected and corrective actions initiated.
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| In addition, a sample of sprinkler heads will be inspected using the guidance of NFPA 25 (2002 Edition) Section 5.3.1.1.1.
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| NFPA 25 states that, "where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." NFPA 25 also contains guidance to perform this sampling every 10 years after initial field service testing.NUREG-1801 Consistency The Fire Water System Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M27, Fire Water System, with an exception and enhancements.
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| Appendix B Aging Management Programs and Activities Page B-46 Pilgrim Nuclear Power Station License Renewal Application Technical Information Exceptions to NUREG-1801 The Fire Water System Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M27, Fire Water System, with the following exception.
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| Attributes Affected Exception 4. Detection of Aging Effects NUREG-1801 specifies annual fire hydrant hose hydrostatic tests. Under the PNPS program, hydrostatic test of hoses occurs once per 3 years.NUREG-1 801 specifies annual gasket inspections.
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| Under the PNPS program, visual inspection, re-racking and replacement of gaskets in couplings occurs at least once per operating cycle.NUREG-1801 specifies annual fire hydrant flow tests. Under the PNPS program, verification of operability and no flow blockage occurs at least once every 2 fuel cycles.Exception Note 1. Since aging effects are typically manifested over several years, differences in inspection and testing frequencies are insignificant.
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| Enhancements The following enhancements will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancements
| |
| : 3. Parameters Monitored/
| |
| Procedures will be enhanced to include Inspected inspection of hose reels for corrosion.
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| : 6. Acceptance Criteria Acceptance criteria will be enhanced to verify no significant corrosion.
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| : 4. Detection of Aging Effects A sample of sprinkler heads will be inspected using guidance of NFPA 25 (2002 Edition) Section 5.3.1.1.1.
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| NFPA 25 also contains guidance to repeat this sampling every 10 years after initial field service testing.Appendix B Aging Management Programs and Activities Page B-47 Pilgrim Nuclear Power Station License Renewal Application Technical Information i Attributes Affected Enhancements
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| : 4. Detection of Aging Effects Wall thickness evaluations of fire protection piping will be performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion.
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| These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation.
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| Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.Operating Experience A fire hose station inspection in 1999 identified a degraded hose station. The hose reel was replaced.
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| Hydrostatic testing and visual inspections of fire hose station equipment in 2004 and 2005 revealed no loss of material.
| |
| Absence of significant corrosion provides evidence that the program is effective for managing loss of material for fire water system components.
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| Inspection of fire water storage tank, T-107A, in 2001 revealed minimal localized leakage, (a probably due to loss of material on the tank bottom. The leakage is being monitored and repair is scheduled.
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| Also, inspection of fire water storage tank, T-107B, in 2003 revealed that microbiologically influenced corrosion (MIC) is occurring at spots (<1/16" in diameter) on internal surfaces.
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| Similar corrosion was seen prior to tank recoating in 1993. Results of the next inspection (2008) will be compared with 2003 results to determine the need for repair of the tank.Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for fire water system components.
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| Full flow tests of fire main segments and hydrant inspections from 2001 through 2004 found no evidence of obstruction or loss of material.
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| Spray and sprinkler system functional tests, and visual inspections of piping and nozzles, in 2003 found no evidence of blockage or loss of material.
| |
| Confirmation of absence of degradation provides evidence that the program is effective for managing loss of material for fire water system components.
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| In 2001, an underground fire main broke due to fabrication and installation anomalies.
| |
| A 16'section of the pipe was replaced.
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| Inspection of internal and external surfaces of the removed pipe section revealed only one small spot of corrosion on the external surface where the coating was cracked. Confirmation of absence of degradation provides evidence that the program is effective for managing loss of material for fire water system components.
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| Appendix B Aging Management Programs and Activities Page B-48 Pilgrim Nuclear Power Station License Renewal Application Technical Information On July 31, 2003, NRC completed a triennial fire protection team inspection to assess whether PNPS has implemented an adequate fire protection program and that post-fire safe shutdown capabilities have been established and are being properly maintained at PNPS. Results confirmed that PNPS was maintaining the fire protection systems in accordance with their fire protection program and that PNPS was identifying program deficiencies and implementing appropriate corrective actions. The team also evaluated the material condition of selected wet pipe sprinkler systems, standpipe systems, and hose reels and concluded that PNPS was maintaining passive features in a state of readiness.
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| A QA audit in May 2004 revealed no issues or findings that could impact effectiveness of the program to manage loss of material for fire water system components.
| |
| Conclusion The Fire Water System Program has been effective at managing aging effects. The Fire Water System Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-49 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.14 FLOW-ACCELERATED CORROSION Proaram Descrintion The Flow-Accelerated Corrosion (FAC) Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M17, Flow-Accelerated Corrosion.
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| This program applies to safety-related and nonsafety-related carbon steel components in systems containing high-energy fluids carrying two-phase or single-phase high-energy fluid > 2%of plant operating time.The program, based on EPRI Report NSAC-202L-R2 recommendations for an effective flow-accelerated corrosion program, predicts, detects, and monitors FAC in plant piping and other pressure retaining components.
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| This program includes (a) an evaluation to determine critical locations, (b) initial operational inspections to determine the extent of thinning at these locations, and (c) follow-up inspections to confirm predictions, or repair or replace components as necessary.
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| NUREG-1801 Consistency The FAC Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M17, Flow-Accelerated Corrosion.
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| Exceptions to NUREG-1801 (s None Enhancements None Onerating Experience Sixty-five FAC UT examinations were performed on-line (between RFO13 and RFO14) and during RFO14 (April, 2003). The examinations included components in the condensate, extraction steam, feedwater, heater vents and drains, main steam, reactor core isolation cooling, and reactor water cleanup systems. Five of the examinations detected decreased wall thickness.
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| Two of the components were accepted after re-evaluation and the other three components were replaced.
| |
| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material in carbon steel components.
| |
| Ninety-seven FAC UT examinations were performed on-line (between RFO14 and RFO15) and during RFO15 (April, 2005). The examinations included components in the condensate, extraction steam, feedwater, heater vents and drains, main steam, reactor core isolation cooling, Appendix B Aging Management Programs and Activities Page B-50 Pilgrim Nuclear Power Station License Renewal Application Technical Information and reactor water cleanup systems. Three of the examinations detected decreased wall thickness.
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| Two of the components were accepted after re-evaluation and the other component was repaired.
| |
| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material in carbon steel components.
| |
| During RFO15 (April, 2005), five piping upgrades to FAC resistant material (ASTM A335 GR P11) were performed.
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| The FAC program document was developed with input from each of the Entergy Nuclear Northeast (ENN) FAC engineers as a standardized ENN procedure.
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| Therefore, it includes improvements based on industry and other ENN plant OE. For example, skid mounted piping is now included in the enhanced system susceptibility evaluation.
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| During RFO15, several FAC points were added to inspections, or re-inspected, in response to industry OE and the MIHAMA Japan failure.A self-assessment in January 2005 revealed no issues or findings that could impact effectiveness of the program to manage FAC in carbon steel components in systems containing high-energy fluids > 2% of plant operating time.Conclusion The FAC Program has been effective at managing aging effects. The FAC Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-5I Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.15 HEAT EXCHANGER MONITORING Proaram Description There is no corresponding NUREG-1801 program.The Heat Exchanger Monitoring Program will inspect heat exchangers for degradation.
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| If degradation is found, then an evaluation will be performed to evaluate its effects on the heat exchanger's design functions including its ability to withstand a seismic event.Representative tubes within the sample population of heat exchangers will be eddy current tested at a frequency determined by internal and external operating experience to ensure that effects of aging are identified prior to loss of intended function.
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| Along with each eddy current test, visual inspections will be performed on accessible heat exchanger heads, covers and tube sheets to monitor surface condition for indications of loss of material.
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| The sample population of heat exchangers includes the RHR heat exchangers, core spray pump motor thrust bearing lube oil coolers, HPCI gland seal condenser, HPCI turbine lube oil cooler, RCIC lube oil cooler, recirculation pump motor generator set fluid coupling oil and bearing coolers, CRD pump oil coolers, recirculation pump motor lube oil coolers, clean up recirculation pump lube oil coolers and stuffing box cooler, and EDG lube oil coolers.The program will be initiated prior to the period of extended operation.
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| Evaluation
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| : 1. Scope of Program The Heat Exchanger Monitoring Program will manage aging effects on selected heat exchangers in various systems as identified in aging management reviews.2. Preventive Actions This is an inspection program and no actions are taken as part of this program to prevent degradation.
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| : 3. Parameters Monitoredllnspected Where practical, eddy current inspections of shell-and-tube heat exchanger tubes will be performed to determine tube wall thickness.
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| Visual inspections will be performed on heat exchanger heads, covers and tube sheets where accessible to monitor surface condition for indications of loss of material.Appendix B Aging Management Programs and Activities Page B-52 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 4. Detection of Aging Effects Loss of material is the aging effect managed by this program. Representative tubes within the sample population of heat exchangers will be eddy current tested at a frequency determined by internal and external operating experience to ensure that effects of aging are identified prior to loss of intended function.
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| Visual inspections of accessible heat exchangers will be performed on the same frequency as eddy current inspections.
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| An appropriate sample population of heat exchangers will be determined based on operating experience prior to inspections.
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| Inspection can reveal loss of material that could result in degradation of the heat exchangers.
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| Fouling is not addressed by this program.5. Monitoring and Trending Results will be evaluated against established acceptance criteria and an assessment will be made regarding the applicable degradation mechanism, degradation rate and allowable degradation level. This information will be used to develop future inspection scope and to modify inspection frequency, if appropriate.
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| Wall thickness will be trended and projected to the next inspection.
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| Corrective actions will be taken if projections indicate that the acceptance criteria may not be met at the next inspection.
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| : 6. Acceptance Criteria The minimum acceptable tube wall thickness for each heat exchanger to be eddy current inspected will be established based upon a component-specific engineering evaluation.
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| Wall thickness will be acceptable if greater than the minimum wall thickness for the component.
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| The acceptance criterion for visual inspections of heat exchanger heads, covers and tubesheets will be no evidence of degradation that could lead to loss of function.
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| If degradation that could lead to loss of intended function is detected, a condition report will be written and the issue resolved in accordance with the site corrective action program.7. Corrective Actions This program will be administered under the site QA program which meets requirements of 10 CFR Part 50, Appendix B.8. Confirmation Process This attribute is discussed in Section B.O.3.Appendix B Aging Management Programs and Activities Page B-53 Appendix B Aging Management Programs and Activities Page B-53.
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 9. Administrative Controls This attribute is discussed in Section B.O.3.10. Operating Experience The Heat Exchanger Monitoring Program at PNPS is a new program for which there is no operating experience.
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| Conclusion The Heat Exchanger Monitoring Program will be effective for managing aging effects since it will incorporate proven monitoring techniques and conservative acceptance criteria.
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| The Heat Exchanger Monitoring Program will provide reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-54 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.16 INSERVICE INSPECTION Regulation 10 CFR 50.55a, imposes inservice inspection (ISI) requirements of ASME Code, Section Xl, for Class 1, 2, and 3 pressure-retaining components, their integral attachments, and supports in light-water cooled power plants. Inspection, repair, and replacement of these components are covered in Subsections IWB, IWC, IWD, and IWF respectively.
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| The program includes periodic visual, surface, and volumetric examination and leakage tests of Class 1, 2, and 3 pressure-retaining components, their integral attachments and supports.Inservice inspection of supports for ASME piping and components is addressed in Section Xl, Subsection IWF. ASME Code Section Xl, Subsection IWF constitutes an existing mandated program applicable to managing aging of ASME Class 1, 2, 3, and MC supports for license renewal.Additionally, 10 CFR 50.55a imposes inservice inspection requirements of ASME Code Section Xl for class MC and class CC containment structures.
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| Subsection IWE contains inspection requirements for class MC metal containments and class CC concrete containments.
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| The scope of IWE includes steel liners for concrete containment and their integral attachments; containment hatches and airlocks; moisture barriers; and pressure-retaining bolting.The program uses nondestructive examination (NDE) techniques to detect and characterize flaws. Three different types of examinations are volumetric, surface, and visual. Volumetric examinations are the most extensive, using methods such as radiographic, ultrasonic or eddy current examinations to locate surface and subsurface flaws. Surface examinations, such as magnetic particle or dye penetrant testing, are used to locate surface flaws.Three levels of visual examinations are specified.
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| VT-1 visual examination is conducted to assess condition of the surface of the part being examined, looking for cracks and symptoms of wear, corrosion, erosion or physical damage. It can be done with either direct visual observation or with remote examination using various optical/video devices. The VT-2 examination is conducted specifically to locate evidence of leakage from pressure retaining components (period pressure tests). While the system is under pressure for a leakage test, visual examinations are conducted to detect direct or indirect indication of leakage. The VT-3 examination is conducted to determine the general mechanical and structural condition of components and supports and to detect discontinuities and imperfections.
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| For containment inservice inspection, general visual and detailed visual examinations are used in addition to VT examinations as allowed by 10 CFR 50.55a to include applicable relief requests.The inservice inspection programs are discussed in more detail in the following subsections
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| * Containment Inservice Inspection (CII)e Inservice Inspection (ISI)Appendix B Aging Management Programs and Activities P Page B-55 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.16.1 CONTAINMENT INSERVICE INSPECTION (CII)Proaram Description The Containment Inservice Inspection (CII) Program is a plant-specific program encompassing the requirements for the inspection of Class MC pressure-retaining components (Primary Containment) and their integral attachments in accordance with the requirements of 10 CFR 50.55a(b)(2) and the 1998 Edition of ASME Section Xl with 2000 Addenda, Inspection Program B.Evaluation
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| : 1. Scope of Program The ClI Program, under ASME Section Xl Subsection IWE, manages loss of material for the primary containment and its integral attachments.
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| The primary containment is a General Electric Mark I pressure suppression containment system. The system consists of a drywell (housing the reactor vessel and reactor coolant recirculation loops), a pressure suppression chamber (housing a water pool), and the connecting vent system between the drywell and the water pool, isolation valves, and containment cooling systems. The code of construction for the containment structure-is the ASME Section III, 1965 Edition and the latest addenda as of June 9, 1969, including Code Cases 1330-1 and 1177-5. (m 2. Preventive Actions The CII Program is a monitoring program that does not include preventive actions.3. Parameters Monitored/Inspected The primary containment and its attachments are inspected for evidence of cracks, wear, and corrosion.
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| : 4. Detection of Aging Effects The CHI Program manages loss of material for the primary containment and its integral attachments.
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| The primary inspection method for the primary containment and its integral attachments is visual examination.
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| Visual examinations are performed either directly or remotely with sufficient illumination and resolution suitable for the local environment to assess general conditions that may affect either the containment structural integrity or leak tightness of the pressure retaining component.
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| The program includes augmented ultrasonic exams to measure wall thickness of the containment structure.
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| Appendix B Aging Management Programs and Activities Page B-56 Pilgrim Nuclear Power Station License Renewal Application Technical Information For steel, the CHI Program manages loss of material and cracking for ASME Code Class MC pressure-retaining steel components and their integral attachments.
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| This aging effect is managed by visual inspections required by ASME Section Xl, Subsection IWE.5. Monitoring and Trending Results are compared, as appropriate, to baseline data and other previous test results. If indications are accepted for continued use by analytical evaluation, the areas containing such flaws are monitored during successive inspection periods.6. Acceptance Criteria Results are compared, as appropriate, to baseline data, other previous test results, and acceptance criteria of the ASME Section Xl, Subsection IWE for evaluation of any evidence of degradation.
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| : 7. Corrective Actions Subsection IWE states that components whose examination results indicate flaws or areas of degradation that do not meet the acceptance standards are acceptable if an engineering evaluation indicates that the flaw or area of degradation is nonstructural in nature or has no effect on the structural integrity of the containment.
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| Except as permitted by 10 CFR 50.55a(b)(ix)(D), components that do not meet the acceptance standards are subject to additional examination requirements, and the components are repaired or replaced to the extent necessary to meet the acceptance standards.
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| : 8. Confirmation Process This attribute is discussed in Section B.0.3.9. Administrative Controls This attribute is discussed in Section B.O.3.10. Operating Experience In 1999, the below-water regions of all 16 torus bays as well as the drywell to torus vent areas with water accumulation were inspected.
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| Results revealed areas of defects such as depleted zinc, localized pitting corrosion, and minor surface rusting.Degraded areas were re-coated to prevent further corrosion and re-examined.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.Appendix B Aging Management Programs and Activities Page B-57 Pilgrim Nuclear Power Station License Renewal Application Technical Information An IWE visual exam in 1999 detected loose torus anchor bolt extensions and baseplate corrosion exceeding acceptance criteria.
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| Bolt extensions were tightened.
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| Corrosion was accepted by evaluation.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.During RFO14 (April, 2003) ultrasonic thickness examination of the torus shell, several measurements were below the nominal wall thickness of 0.629". Since the measurements were all greater than the minimum allowable thickness of 0.563", no further action was taken. CII examinations will continue to monitor thickness of the torus shell. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.Results of the CHI general visual walkdown of primary containment during RFO14 (April, 2003) were compared with those from the previous inspection.
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| The only new indication was in the CRD penetration area, where there is some surface corrosion but it is not significant and is structurally acceptable.
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| No significant corrosion was found in other areas. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.CIH inspections during RFO1 5 (April 2005) did not reveal evidence of loss of material.Absence of degradation provides evidence that the program is effective for managing t aging effects.Oyster Creek experienced drywell corrosion due to salt water intrusion.
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| To ensure the same problem did not exist at PNPS, augmented IWE UT inspections were performed.
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| A QA audit and an NRC inspection in spring 2005 revealed no issues or findings that could impact effectiveness of the program.Conclusion The CII Program has been effective at managing aging effects. The CII Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-58 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.16.2 INSERVICE INSPECTION Program Description The PNPS Inservice Inspection (ISI) Program is a plant-specific program encompassing ASME Section Xl, Subsections IWA, IWB, IWC, IWD and IWF requirements.
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| The ISI Program is based on ASME Inspection Program B (IWA-2432), which has 10-year inspection intervals.
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| Every 10 years the program is updated to the latest ASME Section Xl code edition and addendum approved by the NRC in 10 CFR 50.55a. On July 1, 2005 PNPS entered the fourth ISI interval.
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| The ASME code edition and addenda used for the fourth interval is the 1998 Edition with 2000 Addenda. The current program ensures that the structural integrity of Class 1, 2, and 3 systems and associated supports is maintained at the level required by 10 CFR 50.55a.Evaluation
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| : 1. Scope of Program The ISI Program manages cracking, loss of material, and reduction of fracture toughness of reactor coolant system piping, components, and supports.
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| The program implements applicable requirements of ASME Section Xl, Subsections IWA, IWB, IWC, IWD and IWF, and other requirements specified in 10 CFR 50.55a with approved NRC alternatives and relief requests.
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| Every 10 years the ISI Program is updated to the latest ASME Section Xl code edition and addendum approved by the NRC in 10 CFR 50.55a.ASME Section Xl inspection requirements for Reactor Vessel Intemals (Subsection IWB, Categories B-N-1 and B-N-2) are not in the ISI Program, but are included in the BWR Vessel Internals Program.2. Preventive Actions The ISI Program is a condition monitoring program that does not include preventive actions.3. Parameters Monitored/linspected The program uses nondestructive examination (NDE) techniques to detect and characterize flaws. Volumetric exarqinations such as radiographic, ultrasonic or eddy current examinations are used to 1o6ate surface and subsurface flaws. Surface examinations, such as magnetic particle or dye penetrant testing, are used to locate surface flaws.Appendix B Aging Management Programs and Activities Page B-59 Pilgrim Nuclear Power Station License Renewal Application Technical Information Three levels of visual examinations are specified.
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| VT-1 visual examination is conducted to assess the condition of the surface of the part being examined, looking for cracks and symptoms of wear, corrosion, erosion or physical damage. It can be done with either direct visual observation or with remote examination using various optical and video devices. VT-2 visual examination is conducted specifically to locate evidence of leakage from pressure retaining components (period pressure tests).While the system is under pressure for a leakage test, visual examinations are conducted to detect direct or indirect indication of leakage. VT-3 visual examination is conducted to determine general mechanical and structural condition of components and supports and to detect discontinuities and imperfections.
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| : 4. Detection of Aging Effects The ISI Program manages cracking and loss of material, as applicable, for carbon steel, low alloy steel and stainless steel/nickel based alloy subcomponents of the reactor pressure vessel using NDE techniques specified in ASME Section Xl, Subsections IWB, IWC, and IWD examination categories.
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| The ISI Program manages cracking, loss of material, and reduction of fracture toughness, as applicable, of reactor coolant system components using NDE techniques specified in ASME Section Xl, Subsections IWB, IWC and IWD examination categories.
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| The ISI Program manages loss of material for ASME Class MC and Class 1, 2, and 3 piping and component supports and their anchorages by visual examination of components using NDE techniques specified in ASME Section Xl, Subsection IWF examination categories.
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| No aging effects requiring management are identified for lubrite sliding supports.However, the ISI Program will confirm the absence of aging effects for the period of extended operation.
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| : 5. Monitoring and Trending Results are compared, as appropriate, to baseline data and other previous test results. If indications are accepted for continued use by analytical evaluation, the areas containing such flaws are monitored during successive inspection periods.ISI results are recorded every operating cycle and provided to the NRC after each refueling outage via Owner's Activity Reports prepared by the ISI Program Coordinator.
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| These detailed reports include scope of inspection and significant inspection results.Appendix B Aging Management Programs and Activities Page B-60 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 6. Acceptance Criteria A preservice, or baseline, inspection of program components was performed prior to startup to assure freedom from defects greater than code-allowable.
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| This baseline data also provides a basis for evaluating subsequent inservice inspection results.Since plant startup, additional inspection criteria for Class 2 and 3 components have been imposed by 10 CFR 50.55a for which baseline and inservice data has also been obtained.
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| Results of inservice inspections are compared, as appropriate, to baseline data, other previous test results, and acceptance criteria of the ASME Section Xl, 1998 Edition, 2000 Addenda, for evaluation of any evidence of degradation.
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| : 7. Corrective Actions If a flaw is discovered during an ISI examination, an evaluation is conducted in accordance with articles IWA-3000 and IWB-3000, IWC-3000, IWD-3000 or IWF-3000 as appropriate.
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| If flaws exceed acceptance standards, such flaws are removed, repaired, or the component is replaced prior to its return to service. For Class 1, 2, and 3, repair and replacement is in conformance with IWA-4000.
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| Acceptance of flaws which exceed acceptance criteria may be accomplished through analytical evaluation without repair, removal or replacement of the flawed component if the evaluation meets the criteria specified in the applicable article of the code.8. Confirmation Process This attribute is discussed in Section B.O.3.9. Administrative Controls This attribute is discussed in Section B.0.3.10. Operating Experience Intergranular stress corrosion cracking was discovered during RFOO6 in the thermal sleeve at nine of the ten recirculation supply nozzles. GE has performed an evaluation to demonstrate no further crack growth with hydrogen water chemistry protection.
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| A scheduled ISI surface examination in 1997 detected an indication adjacent to a welded pipe support lug. The lug was removed and the indication was repaired by welding. A scheduled ISI visual examination in 1999 detected a snubber with restricted movement and cold piston setting out of tolerance.
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| The restriction was re-worked and the cold piston setting was accepted by evaluation.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.Appendix B Aging Management Programs and Activities Page B61.
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information 142 scheduled ISI (ASME Section Xl Subsections IWB, IWC, IWD, and IWF)examinations were performed on-line (between RFO13 and RFO14) and during RFO14 (April 2003). Results show that one spring hanger support in the residual heat removal system required rework because ISI visual inspection determined that bolting was loose. Identification of degradation and corrective action prior to loss of Intended function provide evidence that the program is effective for managing aging effects.194 scheduled ISI (ASME Section XI Subsections IWB, IWC, IWD, and IWF)examinations were performed on-line (between RFO14 and RF015) and during RFO1 5 (April 2005). Results show that cracked welds on four steam dryer tie-bars were repaired, loose bolting on a hanger was reworked, a UT exam indication on a standby liquid control system weld was repaired, and a number of RPV safe-end welds were accepted by evaluation because they had wall thickness less than the screening criteria, but not less than design minimums.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects.A QA audit and an NRC inspection in spring 2005 revealed no issues or findings that could impact effectiveness of the program.Conclusion The ISI Program has been effective at managing aging effects. The ISI Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-62 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.17 INSTRUMENT AIR QUALITY Program Description The Instrument Air Quality Program is a plant-specific program which ensures that instrument air supplied to components is maintained free of water and significant contaminants, thereby preserving an environment that is not conducive to loss of material.
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| Dewpoint, particulate contamination, and hydrocarbon concentration are periodically checked to verify the instrument air quality is maintained.
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| Evaluation
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| : 1. Scope of Program This program applies to components within the scope of license renewal and subject to aging management review that are supplied with instrument air, for which pressure boundary integrity is required for the component to perform its intended function.2. Preventive Actions System air quality is monitored and maintained within specified limits to ensure that instrument air supplied to components is maintained free of water and significant contaminants, thereby preventing loss of material.3. Parameters Monitored/inspected Dewpoint, particulate contamination, and hydrocarbon concentration (oil mist) are periodically checked to verify instrument air quality is maintained.
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| : 4. Detection of Aging Effects Dewpoirit, particulate contamination and hydrocarbon concentration are periodically checked to verify instrument air quality is maintained, thereby preventing loss of material.
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| At least once per 18 months, dew point,; particulate contamination, and hydrocarbon concentration are monitored at several locations in the instrument air system.5. Monitoring and Trending Results of sample analyses are maintained in the chemistry log. A condition report is issued if data indicates deteriorating instrument air quality.Appendix B Aging Management Programs and Activities Page B-63 Appendix B Aging Management Programs and Activities Page B-63 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 6. Acceptance Criteria* dew point < -20'F* oil mist and particulate
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| < 1.2 mg/rn 3 7. Corrective Actions Corrective actions are carried out in accordance with the PNPS 10 CFR Part 50, Appendix B, Corrective Action Program.8. Confirmation Process This attribute is discussed in Section B.0.3.9. Administrative Controls This attribute is discussed in Section B.0.3.10. Operating Experience In 1999, an instrument air dryer dewpoint reading was greater than the acceptance criterion of < -20'F. A faulty solenoid valve was replaced and dewpoint was confirmed
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| < -20'F. Monitoring of instrument air quality and subsequent corrective actions provide evidence that the program is effective in managing loss of material and cracking of instrument air system components.
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| For a period of time (October 2001 through March 2005), dew point, particulate contamination, and hydrocarbon concentration (oil mist) were not sampled in the instrument air system. Procedures were corrected in March 2005 to require dew point, particulate contamination, and hydrocarbon concentration (oil mist) sampling at several locations in the instrument air system. Sample results for the service air system, which supplies the instrument air system, show that dewpoint, oil mist and particulates were within acceptance criteria.
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| Instrument air header moisture checks during the same period found little or no moisture.
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| Therefore, instrument air quality is assumed to have been maintained and will be maintained from now on by sampling in accordance with the Instrument Air Quality Program. Continuous confirmation of instrument air quality and subsequent corrective actions provide evidence that the program is effective in managing loss of material and cracking of instrument air system components.
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| Appendix B Aging Management Programs and Activities Page B-64 )
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information Enhancements The following enhancement will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancement
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| : 1. Scope of Program The Instrument Air Quality Program will be enhanced to include a sample point in the standby gas treatment and torus vacuum breaker instrument air subsystem in addition to the instrument air header sample points.Conclusion The Instrument Air Quality Program has been effective at managing aging effects. The Instrument Air Quality Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-65 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.18 METAL-ENCLOSED BUS INSPECTION Proaram Descrintion The Metal-Enclosed Bus Inspection Program at PNPS will be comparable to the program described in NUREG-1801, Section XI.E4, Metal-Enclosed Bus.The program will manage the effects of aging on non-segregated phase bus which connects the 4.16 kV switchgear (A3 through A6) through visual inspection of enclosure assemblies and interior portions of the bus. This inspection will also verify the absence of water or debris.The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The program attributes of the Metal-Enclosed Bus Inspection Program at PNPS will be consistent with the program attributes described in NUREG-1 801, Section XI.E4, Metal-Enclosed Bus, with exceptions.
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| Excentions to NUREG-1801 The program attributes of the Metal-Enclosed Bus (MEB) Inspection Program at PNPS will be consistent with the program attributes described in NUREG-1801, Section Xl.E4, Metal-Enclosed Bus Aging Management Program, with the following exceptions.
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| Attributes Affected Exception 3. Parameters Monitored/
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| MEB enclosure assemblies will be Inspected inspected in addition to internal surfaces.4. Detection of Aging Effects 4. Detection of Aging Effects MEB bolted connections will be visually inspected every 10 years, rather than every five years as stated in NUREG-1801.
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| 2 Exception Notes 1. Inspection of MEB enclosure assemblies under the Metal-Enclosed Bus Inspection Program assures that effects of aging will be identified prior to loss of intended function.2. As stated in NUREG-1801 for the other inspections in this program, a 10 year inspection interval will provide two data points during a 20-year period, which can be used to characterize the degradation rate. This is an adequate period to preclude failures of the MEBs since experience has shown that aging degradation is a slow process.Appendix B Aging Management Programs and Activities Page B-66 Pilgrim Nuclear Power Station License Renewal Application Technical Information Enhancements None Operating Experience The Metal-Enclosed Bus Inspection Program at PNPS is a new program for which there is no operating experience.
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| Conclusion The Metal-Enclosed Bus Inspection Program will be effective for managing aging effects since it will incorporate appropriate monitoring techniques.
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| The Metal-Enclosed Bus Inspection Program will provide reasonable assurance that the effects of aging will be managed such that the applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation Appendix B Aging Management Programs and Activities Page B-67 Pilgrim Nuclear Power Station License Renewal Application Technical Information 0 B.1.19 NON-EQ INACCESSIBLE MEDIUM-VOLTAGE CABLE Program Description The Non-EQ Inaccessible Medium-Voltage Cable Program at PNPS will be comparable to the program described in NUREG-1801, Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
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| In this program, periodic actions will be taken to prevent cables from being exposed to significant moisture, such as inspecting for water collection in cable manholes and conduit, and draining water, as needed. In scope medium-voltage cables exposed to significant moisture and voltage will be tested at least once every ten years to provide an indication of the condition of the conductor insulation.
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| The specific type of test performed will be determined prior to the initial test.The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The program attributes of the Non-EQ Inaccessible Medium-Voltage Cable Program at PNPS will be consistent with the program attributes described in NUREG-1 801, Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject To 10 CFR 50.49 Environmental Qualification Requirements.
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| (Exceptions to NUREG-1801 None Enhancements None Operating Experience The Non-EQ Inaccessible Medium-Voltage Cable Program at PNPS is a new program for which there is no operating experience.
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| Conclusion The Non-EQ Inaccessible Medium-Voltage Cable Program will be effective for managing aging effects since it will incorporate appropriate monitoring techniques.
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| The Non-EQ Inaccessible Medium-Voltage Cable Program will provide reasonable assurance that the effects of aging will be managed such that the applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-68 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.20 NON-EQ INSTRUMENTATION CIRCUITS TEST REVIEW Program Descrintion The Non-EQ Instrumentation Circuits Test Review Program at PNPS will be comparable to the program described in NUREG-1801, Section XL.E2, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.The Non-EQ Instrumentation Circuits Test Review Program will provide reasonable assurance that the intended functions of instrument cables exposed to adverse localized equipment environments caused by heat, radiation and moisture can be maintained consistent with the current licensing basis through the period of extended operation.
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| An adverse localized environment is significantly more severe than the specified service environment for the cable.This program will consider the technical information and guidance provided in NUREG/CR-5643, IEEE Std. P1205, SAND96-0344, and EPRI TR-109619.
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| The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The program will be consistent with NUREG-1801, Section XL.E2, Electrical Cables and 4I Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.Exceptions to NUREG-1801 None Enhancements None Operating Exnerience The Non-EQ Instrumentation Circuits Test Review Program at PNPS is a new program for which there is no operating experience.
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| Industry and plant-specific operating experience will be considered in the development of this program, and future operating experience will be appropriately incorporated into the program.Appendix B Aging Management Programs and Activities Page B-69 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| (.Conclusion The Non-EQ Instrumentation Circuits Test Review Program will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls.
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| Implementation of the Non-EQ Instrumentation Circuits Test Review Program will provide reasonable assurance that the effects of aging will be managed so that the components within the scope of this program will perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Page B-70 Appendix B Aging Management Programs and Activities Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.21 NON-EQ INSULATED CABLES AND CONNECTIONS Program Description The Non-EQ Insulated Cables and Connections Program at PNPS will be comparable to the program described in NUREG-1801, Section XI.E1, Electrical Cables and Connections Not Subjectto 10 CFR 50.49 Environmental Qualification Requirements.
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| The Non-EQ Insulated Cables and Connections Program will provide reasonable assurance that intended functions of insulated cables and connections exposed to adverse localized environments caused by heat, radiation and moisture can be maintained consistent with the current licensing basis through the period of extended operation.
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| An adverse localized environment is significantly more severe than the specified service condition for the insulated cable or connection.
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| A representative sample of accessible insulated cables and connections within the scope of license renewal will be visually inspected for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination.
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| The technical basis for sampling will be determined using EPRI document TR-1 09619, "Guideline for the Management of Adverse Localized Equipment Environments." The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The Non-EQ Insulated Cables and Connections Program at PNPS will be consistent with the program described in NUREG-1801, Section XI.E1, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.
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| Exceptions to NUREG-1801 None Enhancements None Operating Experience The Non-EQ Insulated Cables and Connections Program at PNPS is a new program for which there is no operating experience.
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| Appendix B Aging Management Programs and Activities Page B-71 Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Non-EQ Insulated Cables and Connections Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls.
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| The Non-EQ Insulated Cables and Connections Program will provide reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities.I Page B-72 (1-.W Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.22 OIL ANALYSIS Proaram Description The Oil Analysis Program at PNPS is comparable to the program described in NUREG-1801, Section Xl.M39, Lubricating Oil Analysis.The Oil Analysis Program maintains oil systems free of contaminants (primarily water and particulates) thereby preserving an environment that is not conducive to loss of material, cracking, or fouling.Sampling frequencies are based on vendor recommendations, accessibility during plant operation, equipment importance to plant operation, and previous test results.NUREG-1801 Consistency The Oil Analysis Program at PNPS is consistent with the program described in NUREG-1 801, Section XI.M39, Lubricating Oil Analysis, with an exception and enhancements.
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| Exceptions to NUREG-1801 The Oil Analysis Program at PNPS is consistent with the program described in NUREG-1 801, Section XI.M39, Lubricating Oil Analysis with the following exception.
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| Attributes Affected Exception 3. Parameters Monitored/
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| Flash point is not determined for sampled Inspected oil.1 Exception Note 1. Analyses of filter residue or particle count, viscosity, total acid/base (neutralization number), water content, and metals content provide sufficient information to verify the oil is suitable for continued use.Appendix B Aging Management Programs and Activities Page B-73 Pilgrim Nuclear Power Station License Renewal Application Technical Information rw Enhancements The following enhancements will be initiated prior to the period of extended operation.
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| <Attributes Affected Enhancements
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| : 1. Scope of Program The Oil Analysis Program will be enhanced to periodically change CRD pump lubricating oil. A particle count and check for water will be performed on the drained oil to detect evidence of abnormal wear rates, contamination by moisture, or excessive corrosion.
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| : 3. Parameters Monitored/
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| Procedures for security diesel and reactor Inspected water cleanup pump oil changes will be enhanced to obtain oil samples from the drained oil. Procedures for lubricating oil analysis will be enhanced to specify that a particle count and check for water are performed on oil samples from the fire water pump diesel, security diesel, and reactor water cleanup pumps.Operating Exnerience Lube oil analysis for residual heat removal pump B in July, 2003 showed viscosity slightly outside of the acceptable range. No other problems were noted with the oil. Retest confirmed viscosity condition.
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| Oil was changed at next system window. Continuous confirmation of oil quality and timely corrective actions provide evidence that the program is effective in managing aging effects for lube oil components.
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| Lube oil testing of the A diesel generator in December, 2004 and of the B diesel generator in January, 2005 indicated a step change in the wear particle count. The increase in iron and aluminum was very minor and levels remained well below those at which corrective action is necessary.
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| The analysis laboratory indicated that the increases may be the result of new analysis equipment that has a higher resolution.
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| Quarterly trending will continue for wear products and appropriate action will be taken if required.
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| Continuous confirmation of oil quality and timely corrective actions provide evidence that the program is effective in managing aging effects for lube oil components.I Appendix B Aging Management Programs and Activities Page B-74 Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Oil Analysis Program has been effective at managing aging effects. The Oil Analysis Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-75 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| (%wp~B.1.23 ONE-TIME INSPECTION Program Description The One-Time Inspection Program at PNPS is a new program that wilI be implemented prior to the period of extended operation.
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| The program will be comparable to the program described in NUREG-1801, Section XL.M32, One-Time Inspection.
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| The one-time inspection activity for small bore piping in the reactor coolant system and associated systems that form the reactor coolant pressure boundary, will also be comparable to the program described in NUREG-1801, Section Xl.M35, One-Time Inspection of ASME Code Class I Small-Bore Piping. The PNPS program will be consistent with the program elements described in NUREG-1 801.The program will include one activity to verify effectiveness of an aging management program and activities to confirm the absence of aging effects as described below.Water chemistry control programs One-time inspection activity will verify the effectiveness of the water chemistry control aging management programs by confirming that unacceptable cracking, loss of material, and fouling is not occurring.
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| Internal surfaces of buried carbon steel One-time inspection activity will confirm pipe on the standby gas treatment system that loss of material is not occurring or is so discharge to the stack insignificant that an aging management program is not warranted.
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| Internal surfaces of compressed air and One-time inspection activity will confirm emergency diesel generator system that cracking (EDG system) and loss of components containing untreated air material (compressed air and EDG systems) are not occurring or are so insignificant that an aging management program is not warranted.
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| Internal surfaces of stainless steel One-time inspection activity will confirm radioactive waste and sanitary soiled that loss of material is not occurring or is so waste and vent system components insignificant that an aging management containing untreated water program is not warranted.
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| Small bore piping in the reactor coolant One-time inspection activity will confirm system and associated systems that form that cracking and reduction of fracture the reactor coolant pressure boundary toughness are not occurring or are so insignificant that an aging management program is not warranted.
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| CJ Appendix B Aging Management Programs and Activities Page B-76 Aw' Pilgrim Nuclear Power Station License Renewal Application Technical Information RV flange leakoff line One-time inspection activity will confirm that cracking is not occurring or is so insignificant that an aging management program is not warranted.
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| Main steam flow restrictors (CASS) One-time inspection activity will confirm that loss of material, cracking, and reduction of fracture toughness are not occurring or are so insignificant that an aging management program is not warranted.
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| The elements of the program include (a) determination of the sample size based on an assessment of materials of fabrication, environment, plausible aging effects, and operating experience; (b) identification of the inspection locations in the system or component based on the aging effect; (c) determination of the examination technique, including acceptance criteria that would be effective in managing the aging effect for which the component is examined; and (d)evaluation of the need for follow-up examinations to monitor the progression of any aging degradation.
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| When evidence of an aging effect is revealed by a one-time inspection, routine evaluation of the inspection results will identify appropriate corrective actions.The inspection will be performed within the 10 years prior to the period of extended operation.
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| NUREG-1801 Consistency The One-Time Inspection Program will be consistent with the program described in NUREG-1801, Section XI.M32, One-lime Inspection.
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| The one-time inspection activity for small bore piping in the reactor coolant system and associated systems that form the reactor coolant pressure boundary, will also be consistent with the program described in NUREG-1801, Section XL.M35, One-Time Inspection of ASME Code Class I Small-Bore Piping.Exceotions to NUREG-1801 None Enhancements None Appendix B Aging Management Programs and Activities Page B-77 Pilgrim Nuclear Power Station License Renewal Application Technical Information Operating Experience The One-Time Inspection Program is a new program for which there is no operating experience.
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| Industry and plant-specific operating experience will be considered in development of this program, as appropriate.
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| Conclusion Verification of the effectiveness of the Water Chemistry Control programs and confirmation of the absence of aging effects on specific standby gas treatment, compressed air, emergency diesel generator, radioactive waste, sanitary soiled waste and vent, and reactor coolant system components will be undertaken in the One-Time Inspection Program to ensure component intended functions can be maintained in accordance with the current licensing basis (CLB) during the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-78 i Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.24 PERIODIC SURVEILLANCE AND PREVENTIVE MAINTENANCE Program Descrintion There is no corresponding NUREG-1801 program.The PNPS Periodic Surveillance and Preventive Maintenance Program includes periodic inspections and tests that manage aging effects not managed by other aging management programs.
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| The preventive maintenance and surveillance testing activities are generally implemented through repetitive tasks or routine monitoring of plant operations.
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| Credit for program activities has been taken in the aging management review of the following systems and structures.
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| reactor building process facilities standby liquid control system automatic depressurization system high pressure coolant injection system reactor core isolation cooling system Perform visual or other non-destructive examination to manage loss of material for the reactor building crane, rails, and girders and refueling platform carbon steel components.
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| Visually inspect the main stack components to manage loss of material for carbon steel and cracking, spalling, or loss of material for concrete.Use UT or other NDE techniques to verify remaining wall thickness to manage loss of material from internal surfaces of the carbon steel discharge accumulators.
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| Use visual or other NDE techniques to inspect torus to manage loss of material for carbon steel piping in the waterline region of the torus.Use visual or other NDE techniques to inspect a representative sample of the internals of gland seal condenser blower (P-223) and suction piping to manage loss of material.Use visual or other NDE techniques to inspect a representative sample of RCIC steam supply and exhaust piping downstream of the strainers and steam traps to manage loss of material.Appendix B Aging Management Programs and Activities Page B-79 Pilgrim Nuclear Power Station License Renewal Application Technical Information standby gas treatment system Perform a visual inspection of accessible expansion joints for cracks. Also perform manual flexing (manipulation) of the expansion joints to determine if they have become brittle. These inspections will verify the absence of significant change in material properties.
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| Use visual or other NDE techniques to inspect internal surfaces of the valve bodies and piping in the demister drains to manage loss of material.Use visual or other NDE techniques to inspect a representative sample of the internal and external surfaces of the drain lines from each reactor building auxiliary bay passing into the water trough in the torus room to manage loss of material.reactor building closed cooling water system Use visual or other NDE techniques to inspect clean-up recirc pump P-204B stuffing box cooler to manage loss of material due to wear.Use visual or other NDE techniques to inspect a representative sample of the in-scope RBCCW copper alloy cooling coils to manage loss of material.(.emergency diesel generator system Use visual or other NDE techniques to inspect a representative sample of EDG intake air, air start, and exhaust components to manage loss of material and fouling.Visually inspect A/B EDG jacket water radiators to manage loss of material and fouling.Perform EDG surveillance test (loaded) to manage fouling for heat exchanger tubes.Appendix B Aging Management Programs and Activities Page B-80 (
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information station blackout diesel generator system Use visual or other NDE techniques to inspect a representative sample of station blackout diesel intake air, air start, and exhaust components to manage loss of material, cracking, and fouling.Visually inspect station blackout jacket water radiator to manage loss of material and fouling.Perform station blackout diesel surveillance test to manage fouling for heat exchanger tubes.heating, ventilation, and air conditioning systems Use visual or other NDE techniques to inspect the air side of the copper alloy tubes of heat exchangers VAC-2OIA/B, VAC-202AIB, and VAC-204A/B/CID to manage loss of material and fouling.Visually inspect and manually flex VSF-103ANB, VAC-202A1B, VAC-204A1B/C/D, and EDG engine-driven fan duct flexible connections to manage cracking and change in material properties.
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| security diesel Perform security diesel generator surveillance test (loaded) to manage fouling for heat exchanger tubes.Use visual or other NDE techniques to inspect a representative sample of security diesel oil cooler, aftercooler, and radiator tubes to manage loss of material.Use visual or other NDE techniques to inspect a representative sample of security diesel intake air and exhaust components to manage cracking and loss of material on internal surfaces.condensate storage system Use visual or other NDE techniques to inspect a representative sample of the internal and external surfaces of the condensate storage tanks to manage loss of material.Appendix B Aging Management Programs and Activities Page B-81 Pilgrim Nuclear Power Station License Renewal Application Technical Information ( ;nonsafety-related systems Use visual or other NDE techniques to inspect a affecting safety-related representative sample of circulating water, potable &systems sanitary water, radioactive waste, sanitary soiled waste &vent, plumbing and drains and screen wash system components to manage internal loss of material.Visually inspect and manually flex a representative sample of the flex/expansion joints in the circulating water, HVAC/chilled water, and radioactive waste systems to manage cracking and change in material properties.
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| Evaluation
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| : 1. Scope of Program The PNPS Periodic Surveillance and Preventive Maintenance Program, with regard to license renewal, includes those tasks credited with managing aging effects identified in aging management reviews.2. Preventive Actions Inspection and testing activities used to identify component aging effects do not prevent aging effects. However, activities are intended to prevent failures of components that might be caused by aging effects.3. Parameters Monitored/lnspected This program provides instructions for monitoring structures, systems, and components to detect degradation.
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| Inspection and testing activities monitor various parameters including system flow, system pressure, surface condition, loss of material, presence of corrosion products, and signs of cracking.4. Detection of Aging Effects Preventive maintenance activities and periodic surveillances provide for periodic component inspections and testing to detect aging effects. Inspection intervals are established such that they provide timely detection of degradation.
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| Inspection intervals are dependent on component material and environment and take into consideration industry and plant-specific operating experience and manufacturers' recommendations.
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| Each inspection or test occurs at least once every ten years.Appendix B Aging Management Programs and Activities Page B-82 Pilgrim Nuclear Power Station License Renewal Application Technical Information The extent and schedule of inspections and testing assure detection of component degradation prior to loss of intended functions.
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| Established techniques such as visual inspections are used.5. Monitoring and Trending Preventive maintenance and surveillance testing activities provide for monitoring and trending of aging degradation.
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| Inspection and testing intervals are established such that they provide for timely detection of component degradation.
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| Inspection and testing intervals are dependent on component material and environment and take into consideration industry and plant-specific operating experience and manufacturers' recommendations.
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| : 6. Acceptance Criteria Periodic Surveillance and Preventive Maintenance Program acceptance criteria are defined in specific inspection and testing procedures.
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| The procedures confirm component integrity by verifying the absence of aging effects or by comparing applicable parameters to limits based on applicable intended functions established by plant design basis.7. Corrective Actions The PNPS Corrective Action Program, quality assurance procedures, site review and approval process, and administrative controls are implemented in accordance with requirements of 10 CFR Part 50, Appendix B.8. Confirmation Process This attribute is discussed in Section B.0.3.9. Administrative Controls This attribute Is discussed in Section B.0.3.10. Operating Experience Inspection of the reactor building crane in 2000 and of the refueling platform in March, 2003 found no significant corrosion or wear. Absence of significant corrosion and wear provides evidence that the program is effective for managing loss of material for the reactor building crane, rails, and girders and refueling platform carbon steel components.
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| Visual inspection of the main stack and guy wires in June, 2004 revealed no significant corrosion of steel structures and components.
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| Similarly, inspection of the Appendix B Aging Management Programs and Activities Page B-83 Pilgrim Nuclear Power Station License Renewal Application Technical Information concrete anchor blocks revealed no cracking, spalling, or other loss of material.Absence of steel corrosion and concrete cracking, spalling, and loss of material provides evidence that the program is effective for managing aging effects for components of the main stack.In 1999, visual inspection of the drywell spray header revealed no significant corrosion.
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| Absence of significant'corrosion provides evidence that the program is effective for managing loss of material for the drywell spray header.In 1999, the below-water regions of all 16 torus bays as well as the drywell to torus vent areas with water accumulation were inspected.
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| The condition of other submerged structures and components was also reported.
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| Results revealed no significant corrosion on submerged structures and components within the torus.Absence of significant corrosion provides evidence that the program is effective for managing loss of material for carbon steel SRV tailpipes in the waterline region of the torus.During visual inspection of standby gas treatment system exhaust fans in 2000 and 2001, the expansion joints which connect the fans to ductwork were disconnected from the fans to facilitate fan inspection.
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| Inspection of the expansion joints after this evolution revealed no cracking.
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| Absence of cracking provides evidence that the program is effective for managing cracking and change in material properties for the expansion joints.No significant corrosion or wear was found on the reactor recirculation system MG sets area cooling coils during an inspection in 2000. Absence of significant corrosion or wear provides evidence that the program is effective for managing loss of material for RBCCW copper alloy cooling coils.During a 2002 run of the A EDG, soot buildup was noticed on the turbo charger.Although no obvious leakage was noted, soot buildup may indicate existence of a small exhaust leak. Thermography was performed during the next diesel run to determine if and where leakage was occurring, but no leakage was found.Identification of possible degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for EDG exhaust components.
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| Inspections of EDG air intake and jacket water radiator components in 1999 and 2004 revealed no significant corrosion, wear, or fouling. Also, no significant corrosion was found on air start components or exhaust components during the inspections.
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| Absence of aging effects provides evidence that the program is effective for managing aging effects for EDG components.
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| Appendix B Aging Management Programs and Activities Page B-84 Pilgrim Nuclear Power Station License Renewal Application Technical Information EDG surveillance tests were performed in April 2005. Results for both generators show that air manifold temperature did not fluctuate significantly during the loaded run, providing evidence that the program is effective for managing fouling of EDG intake air cooler tubes.Inspections of station blackout (SBO) diesel jacket water radiator components in 2001 revealed no significant corrosion, wear, or fouling. Also, no significant corrosion was found on air start components or exhaust components during the inspections.
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| Minor corrosion on inside surface of the air intake silencer housing was determined to not affect the ability of the silencer to perform its intended function.
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| Absence of significant aging effects provides evidence that the program is effective for managing aging effects for SBO diesel components.
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| SBO diesel generator surveillance tests were performed in May 2005. Results show that air manifold temperature did not fluctuate significantly during the loaded run, providing evidence that the program is effective for managing fouling of SBO diesel intake air cooler tubes.Visual inspection of the control room emergency air supply system blowers in 1999 revealed no cracking of the flexible connectors on these components.
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| Absence of cracking provides evidence that the program is effective for managing cracking for flexible connectors.
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| A thorough inspection of the security diesel intake air components, exhaust components, and the jacket water radiator in 1998 revealed no significant corrosion, cracking, wear, or fouling. Absence of aging effects provides evidence that the program is effective for managing aging effects for security diesel system components.
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| Security diesel generator surveillance tests were performed in 2002, 2003, and 2004.Results show that air manifold temperature did not fluctuate significantly during the loaded run, providing evidence that the program is effective for managing fouling of security diesel intake air cooler tubes.An inspection of the 'A' condensate storage tank in April, 2003 noted paint flaking off the interior of the tank, corrosion nodules on the sidewall and floor, and a 2"-3" diameter by t2" deep depression in the tank floor. The 'B' condensate storage tank was also inspected and no corrosion or coating degradation was observed.
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| A long-term corrective action was initiated to assess the interior condition of the tank, review the existing coating system, select an appropriate recoating system, and repair and recoat the 'A' condensate storage tank. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for the condensate storage tanks.Appendix B Aging Management Programs and Activities Page B-85 Pilgrim Nuclear Power Station License Renewal Application Technical Information Qf )Enhancements Prior to the period of extended operation, program activity implementing documents will be enhanced as necessary to assure that the effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Conclusion The Periodic Surveillance and Preventive Maintenance Program has been effective at managing aging effects. The Periodic Surveillance and Preventive Maintenance Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-86 _I Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.25 REACTOR HEAD CLOSURE STUDS Program Description The Reactor Head Closure Studs Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M3, Reactor Head Closure Studs.This program includes inservice inspection (ISI) in conformance with the requirements of ASME Section XI, Subsection IWB, and preventive measures (e.g. rust inhibitors, stable lubricants, appropriate materials) to mitigate cracking and loss of material of reactor head closure studs, nuts, washers, and bushings.NUREG-1801 Consistency The Reactor Head Closure Studs Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M3, Reactor Head Closure Studs, with one exception.
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| Excentions to NUREG-1801 The Reactor Head Closure Studs Program at PNPS is consistent with the program described in NUREG-1 801, Section XI.M3, Reactor Head Closure Studs, with the following exception.
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| Attributes Affected Exception 4. Detection of When reactor head closure studs are Aging Effects removed for examination, either a surface or volumetric examination is allowed.'Exception Note 1. Cracking initiates on the outside surfaces of bolts and studs. Therefore, a qualified surface examination meeting the acceptance standards of ASME Section Xl, Subsection IWB-3515 provides at least the sensitivity for flaw detection that an end shot ultrasonic examination provides on bolts or studs. Thus, when reactor head closure studs are removed for examination, either a surface or volumetric examination is allowed.Enhancements None Operating Experience Volumetric examination of 18 reactor head closure studs and visual examination of 18 nuts and 18 washers during RFO15 (April, 2005) resulted in no new recordable indications.
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| Absence of Appendix B Aging Management Programs and Activities Page B87.
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| Pilgrim Nuclear Power Station License Renewal Application Technical Information new recordable indications provides evidence that the program is effective for managing loss of material and cracking of the reactor head closure studs, nuts, washers, and bushings.Conclusion The Reactor Head Closure Studs Program has been effective at managing aging effects. The Reactor Head Closure Studs Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-88 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.26 REACTOR VESSEL SURVEILLANCE Program Description The Reactor Vessel Surveillance Program complies with the guidelines for an acceptable Integrated Surveillance Program as described in NUREG-1 801, Section XI.M31, Reactor Vessel Surveillance.
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| This program manages reduction in fracture toughness of reactor vessel beltline materials to assure that the pressure boundary function of the reactor pressure vessel is maintained for the period of extended operation.
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| PNPS is a participant in the Boiling Water Reactor Vessel and Internals Project (BWRVIP)Integrated Surveillance Program (ISP) as approved by License Amendment 209. This program monitors changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. As BWRVIP-ISP capsule test reports become available for RPV materials representative of PNPS, the actual shift in the reference temperature for nil-ductility transition of the vessel material may be updated. In accordance with 10 CFR 50 Appendices G and H, PNPS reviews relevant test reports to assure compliance with fracture toughness requirements and P-T limits.BWRVIP-116, 'BWR Vessel and Internals Project Integrated Surveillance Program (ISP)Implementation for License Renewal," describes the design and implementation of the ISP during the period of extended operation.
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| BWRVIP-116 identifies additional capsules, their withdrawal schedule, and contingencies to ensure that the requirements of 10 CFR 50 Appendix H are met for the period of extended operation.
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| NUREG-1801 Consistency The Reactor Vessel Surveillance Program at PNPS is consistent with the program described in NUREG-1801, Section XI.M31, Reactor Vessel Surveillance, with one enhancement.
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| Exceptions to NUREG-1801 None Enhancements The following enhancement will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancement
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| : 5. Monitoring and Trending Actions The Reactor Vessel Surveillance Program 6. Acceptance Criteria will be enhanced to proceduralize the data 6. Cctane ctia analysis, acceptance criteria, and corrective
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| .actions described in this program description.
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| Appendix B Aging Management Programs and Activities Page B-89 Pilgrim Nuclear Power Station License Renewal Application Technical Information Operating Experience PNPS is a participant in the Boiling Water Reactor Vessel and Internals Project (BWRVIP)Integrated Surveillance Program (ISP) as approved by Amendment 209 to the operating License.The fact that PNPS participates in the BWRVIP ISP ensures that future operating experience from all participating BWRs will be factored into this program.Conclusion The Reactor Vessel Surveillance Program provides reasonable assurance that aging effects will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-90 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.27 SELECTIVE LEACHING Proaram Descriotion The Selective Leaching Program at PNPS will be comparable to the program described in NUREG-1801, Section XL.M33 Selective Leaching of Materials.
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| The Selective Leaching Program will ensure the integrity of components made of cast iron, bronze, brass, and other alloys exposed to raw water, treated water, or groundwater that may lead to selective leaching.
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| The program will include a one-time visual inspection and hardness measurement of selected components that may be susceptible to selective leaching to determine whether loss of material due to selective leaching is occurring, and whether the process will affect the ability of the components to perform their intended function for the period of extended operation.
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| The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The Selective Leaching Program at PNPS will be consistent with the program described in NUREG-1801, Section XL.M33 Selective Leaching of Materials.
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| Excentions to NUREG-1801 None Enhancements None Onerating Exoerience The Selective Leaching Program is a new program for which there is no operating experience.
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| Conclusion The Selective Leaching Program will be effective for managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls.
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| The Selective Leaching Program will provide reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-91 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.28 SERVICE WATER INTEGRITY Program Descrintion The Service Water Integrity Program at PNPS is comparable to the program described in NUREG-1801, Section XL.M20, Open-Cycle Cooling Water System.This program relies on implementation of the recommendations of GL 89-13 to ensure that the effects of aging on the salt service water (SSW) system are managed for the period of extended operation.
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| The program includes surveillance and control techniques to manage aging effects caused by biofouling, corrosion, erosion, protective coating failures, and silting in the SSW system or structures and components serviced by the SSW system.NUREG-1801 Consistency The Service Water Integrity Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M20, Open-Cycle Cooling Water System with exceptions.
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| Exceptions to NUREG-1801 The Service Water Integrity Program at PNPS is consistent with the program described in NUREG-1801, Section XL.M20, Open-Cycle Cooling Water System with the following exceptions.
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| (4f, Attributes Affected Exceptions
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| : 2. Preventive Actions NUREG-1801 states that system components are lined or coated.Components are lined or coated only where necessary to protect the underlying metal surfaces.1 5. Monitoring and Trending NUREG-1801 states that testing and inspections are performed annually and during refueling outages. The PNPS program requires tests and inspections each refueling outage.2 Exception Notes 1. NUREG-1801 states that system components are constructed of appropriate materials and lined or coated to protect the underlying metal surfaces from being exposed to aggressive cooling water environments.
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| Not all PNPS system components are lined or coated. Components are lined or coated only where necessary to protect the underlying metal surfaces.Appendix B Aging Management Programs and Activities Page B-92 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 2. NUREG-1801 program entails testing and inspections performed annually and during refueling outages. The PNPS program requires tests and inspections each refueling outage, but not annually.
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| Since aging effects are typically manifested over several years, the difference in inspection and testing frequency is insignificant.
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| Enhancements None Oreratina Experience Results of heat transfer capability testing of the reactor building closed cooling water (RBCCW)heat exchangers from 2001 through 2004 show that the heat exchangers are capable of removing the required amount of heat. Confirmation of adequate thermal performance provides evidence that the program is effective for managing fouling of SSW cooled heat exchangers.
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| Results of SSW visual inspections, eddy current testing, ultrasonic testing, and radiography testing from 1998 through 2004 revealed areas of erosion and areas of corrosion on internal and external surfaces.
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| SSW butterfly valves, pump discharge check valves, air removal valves, and pipe spools have been replaced with components made of corrosion resistant materials.
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| Also, RBCCW heat exchanger channel assemblies have been replaced and tubes have been sleeved to address erosion and corrosion.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for SSW system components.
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| Visual inspections of SSW piping revealed degradation of the lining in original SSW carbon steel rubber lined piping. Pipe lining is intended to protect pipe internal surfaces from erosion and corrosion.
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| Therefore, SSW piping has been replaced with carbon steel pipe with cured-in-place rubber lining, relined with a ceramic epoxy compound, or replaced with titanium pipe.Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material for SSW system components.
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| Conclusion The Service Water Integrity Program has been effective at managing aging effects. The Service Water Integrity Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended function consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-93 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.29 STRUCTURES MONITORING The Structures Monitoring programs are discussed in more detail in the following subsections
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| * Masonry Wall* Structures Monitoring
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| * Water Control Structures Monitoring B.1.29.1 MASONRY WALL Program Descrintion The Masonry Wall Program at PNPS is comparable to the program described in NUREG-1801, Section Xl.S5, Masonry Wall Program.The objective of the Masonry Wall Program is to manage aging effects so that the evaluation basis established for each masonry wall within the scope of license renewal remains valid through the period of extended operation.
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| The program includes all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4. Included components are the 10 CFR 50.48-required masonry walls, radiation shielding masonry walls, masonry walls with the potential to affect safety-related components, and the torus compartment water trough.Masonry walls are visually examined at a frequency selected to ensure there is no loss of intended function between inspections.
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| NUREG-1801 Consistency The Masonry Wall Program is consistent with the program described in NUREG-1801, Section Xl.S5, Masonry Wall Program.Exceptions to NUREG-1801 None Enhancements None Appendix B Aging Management Programs and Activities Page B-94 Pilgrim Nuclear Power Station License Renewal Application Technical Information Oneratina Experience Examinations of masonry walls within the scope of license renewal in 2002 did not find evidence of cracking.
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| A review of condition reports from 1998 through 2004 did not reveal any instances of cracked masonry walls. Absence of cracking provides evidence that the program is effective for managing cracking of masonry walls.Conclusion The Masonry Wall Program has been effective at managing aging effects. The Masonry Wall Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| B.1.29.2 STRUCTURES MONITORING Program Descrintion The Structures Monitoring Program at PNPS is comparable to the program described in NUREG-1801, Section Xl.S6, Structures Monitoring Program.Structures monitoring in accordance with 10 CFR 50.65 (Maintenance Rule) is addressed in Regulatory Guide 1.160 and NUMARC 93-01. These two documents provide guidance for development of licensee-specific programs to monitor the condition of structures and structural components within the scope of the Maintenance Rule, such that there is no loss of structure or structural component intended function.Since protective coatings are not relied upon to manage the effects of aging for structures included in the Structures Monitoring Program, the program does not address protective coating monitoring and maintenance.
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| NUREG-1801 Consistency The Structures Monitoring Program is consistent with the program described in NUREG-1801, Section XL.S6, Structures Monitoring Program.Exceptions to NUREG-1801 None Appendix B Aging Management Programs and Activities Page B-95 Pilgrim Nuclear Power Station License Renewal Application Technical InformationEnhancements The following enhancements will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancements
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| : 1. Scope of Program The Structures Monitoring Program procedure will be enhanced to clarify that the discharge structure, security diesel generator building, trenches, valve pits, manholes, duct banks, underground fuel oil tank foundations, manway seals and gaskets, hatch seals and gaskets, underwater concrete in the intake structure, and crane rails and girders are included in the program.4. Detection of Aging Effects Guidance for performing structural examinations of elastomers (seals, gaskets, seismic joint filler, and roof elastomers) to identify cracking and change in material properties will be added to the Structures Monitoring Program procedure.
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| Q.,I1 Oerating Exnerience Inspections of structural steel, concrete exposed to fluid, and structural elastomers from 1998 through 2004 revealed signs of degradation such as cracks, gaps, corrosion (rust), and flaking coatings.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for structural components.
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| Structural inspection of pipe supports and cable trays in November 2004 revealed numerous minor signs of degradation which were repaired.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for structural components.
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| A self-assessment in July 2005 revealed no issues or findings that could impact effectiveness of the program.Appendix B Aging Management Programs and Activities P --9 Page B-96 i 0i Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Structures Monitoring Program has been effective at managing aging effects. The Structures Monitoring Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| B.1.29.3 WATER CONTROL STRUCTURES MONITORING Program Description The Water Control Structures Monitoring Program at PNPS is comparable to the program described in NUREG-1801, Section XI.S7, RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants.The program includes visual inspections to manage loss of material and loss of form for water-control structures (breakwaters, jetties, and revetments).
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| The water-control structures are of rubble mound construction with the outer layer protected by heavy capstone.
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| Parameters monitored include settlement (vertical displacement) and rock displacement.
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| These parameters are consistent with those described in RG 1.127.NUREG-1801 Consistency The Water Control Structures Monitoring Program is consistent with the program described in NUREG-1801, Section XI.S7, RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants.Exceptions to NUREG-1801 None Enhancements The following enhancement will be initiated prior to the period of extended operation.
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| Attributes Affected Enhancement
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| : 1. Scope of Program Program scope will be enhanced to include the east breakwater, jetties, and onshore revetments in addition to the main breakwater.
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| Appendix B Aging Management Programs and Activities Page B-97 Pilgrim Nuclear Power Station License Renewal Application Technical Information Operating Exnerience Preliminary results of the 2004 inspection of the main breakwater indicated one area of the breakwater had rock displacement resulting in the complete dislodging of the rocks on the shore side of the main breakwater.
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| Since the discontinuity extended beyond the facade but did not involve the full height or width of the water-control structure, an evaluation was performed to determine if repair was required to restore the designed stability of the structure.
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| Results of the evaluation show that the designed stability of the structure was not impacted, however a work request was issued to repair the structure due to the possibility of future storms extending the damaged areas and restriction to personnel from easily walking on the structure.
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| Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing loss of material and loss of form for water-control structures.
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| Conclusion The Water Control Structures Monitoring Program has been effective at managing aging effects.The Water Control Structures Monitoring Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-98 _11 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.30 SYSTEM WALKDOWN Program Descrintion The System Walkdown Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M36, External Surfaces Monitoring.
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| This program entails inspections of external surfaces of components subject to aging management review. The program is also credited with managing loss of material from internal surfaces, for situations in which internal and external material and environment combinations are the same such that external surface condition is representative of internal surface condition.
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| NUREG-1801 Consistency The System Walkdown Program is consistent with the program described in NUREG-1801, Section XI.M36, External Surfaces Monitoring.
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| Exceptions to NUREG-1801 The System Walkdown Program is consistent with the program described in NUREG-1801, Section XI.M36, External Surfaces Monitoring.
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| Enhancements None Operating Experience System walkdowns between 1998 and 2004 identified evidence of aging effects, including corrosion and leakage. Examples include fire water storage tank and diesel fire pump fuel oil day tank leakage, through-wall leakage on SSW piping, signs of corrosion in fan room and auxiliary bays, and through-wall leakage on a drain line to the aux bay sump. Corrective actions were accomplished in accordance with the site Corrective Action Program. Identification of degradation and corrective action prior to loss of intended function provide evidence that the program is effective for managing aging effects for passive components.
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| Conclusion The System Walkdown Program has been effective at managing aging effects. The System Walkdown Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-99 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.31 THERMAL AGING AND NEUTRON IRRADIATION EMBRITTLEMENT OF CAST AUSTENITIC STAINLESS STEEL (CASS)Program Description The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program at PNPS will be comparable to the program described in NUREG-1801, Section XL.M13, Thermal Aging and Neutron Irradiation Embrittlement of CASS.'The purpose of the Thermal Aging and Neutron Irradiation Embrittlement of CASS Program is to assure that reduction of fracture toughness due to thermal aging and reduction of fracture toughness due to radiation embrittlement will not result in loss of intended function.
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| This program will evaluate CASS components in the reactor vessel internals and require non-destructive examinations as appropriate.
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| EPRI, the BWR Owners Group and other industry groups are focused on reactor vessel internals to ensure a better understanding of aging effects. Future Boiling Water Reactor Vessel Internals Project (BWRVIP) reports, EPRI reports, and other industry operating experience will provide additional bases for evaluations and inspections under this program. This program will supplement reactor vessel internals inspections required by the BWR Vessel Internals Program to assure that aging effects do not result in loss of the intended functions of reactor vessel internals during the period of extended operation.
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| The program will be initiated prior to the period of extended operation.
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| NUREG-1801 Consistency The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program will be consistent with the program described in NUREG-1801, Section XI.M113, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.Exceptions to NUREG-1801 The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program will be consistent with the program described in NUREG-1801, Section X1.M13, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.Enhancements None Operatina Experience The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program is a new program for which there is no operating experience.
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| Appendix B Aging Management Programs and Activities Page B-100 Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Thermal Aging and Neutron Irradiation Embrittlement of CASS Program will use existing techniques with demonstrated capability and a proven industry record to provide reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-101 Pilgrim Nuclear Power Station License Renewal Application Technical Information B.1.32 WATER CHEMISTRY CONTROL The PNPS chemistry program is the personnel, programs, policies and procedures designed to control site water chemistry to maximize plant availability, extend operating lifetime,;
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| and minimize radiation levels. Based on applicable EPRI Guidelines, the program controls contaminants at lowest practical levels and provides corrosion protection for major systems and components.
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| The following subsections address individual PNPS water chemistry control programs in more detail.* Water Chemistry Control -Auxiliary Systems* Water Chemistry Control -BWR* Water Chemistry Control -Closed Cooling Water B.1.32.1 WATER CHEMISTRY CONTROL -AUXILIARY SYSTEMS Proaram Descrintion There is no corresponding NUREG-1801 program.The purpose of the Water Chemistry Control -Auxiliary Systems Program is to manage loss of 0 material for components exposed to treated water.Program activities include sampling and analysis of the stator cooling water system to minimize component exposure to aggressive environments.
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| Evaluation
| |
| : 1. Scope of Program Program activities include sampling and analysis of the stator cooling water system to minimize component exposure to aggressive environments.
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| City water is taken from the Town of Plymouth water main and distributed throughout the potable and sanitary water system at town water pressure.
| |
| City water is monitored and treated by the Town of Plymouth to meet the regulations of the Commonwealth of Massachusetts.
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| : 2. Preventive Actions The program includes monitoring and control of stator cooling water to minimize exposure to aggressive environments.
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| Appendix B Aging Management Programs and Activities Page B-102 Pilgrim Nuclear Power Station License Renewal Application Technical Information City water used in the potable and sanitary water system is monitored and treated by the town of Plymouth to meet the regulations of the Commonwealth of Massachusetts.
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| : 3. Parameters Monitored/lnspected In accordance with industry recommendations, stator cooling water parameters monitored are conductivity, corrosion products, and dissolved oxygen.City water used in the potable and sanitary water system is monitored and treated by the town of Plymouth to meet the regulations of the Commonwealth of Massachusetts.
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| : 4. Detection of Aging Effects The program manages loss of material for stator cooling water system and potable and sanitary water system components.
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| The One-Time Inspection Program describes inspections planned to verify the effectiveness of water chemistry control programs to ensure that significant degradation is not occurring and component intended function is maintained during the period of extended operation.
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| : 5. Monitoring and Trending Values from analyses are archived for long-term trending and review.6. Acceptance Criteria In accordance with industry recommendations, acceptance criteria for the stator cooling water system are as follows.* conductivity
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| < 0.3 S/cm* dissolved oxygen > 2.0 ppm / < 8.Oppm* corrosion products no detectable activity 7. Corrective Actions If acceptance criteria are not met, chemistry parameters are adjusted as appropriate.
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| Additional sampling and verification is performed if necessary.
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| Corrective actions for unacceptable inspection results are identified and implemented in accordance with the Corrective Action Program.Appendix B Aging Management Programs and Activities IPage B-103 Pilgrim Nuclear Power Station License Renewal Application Technical Information
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| : 8. Confirmation Process This attribute is discussed in Section B.0.3.9. Administrative Controls This attribute is discussed in Section B.0.3.10. Operating Experience In spring 2001, a small leak of hydrogen into the stator coolant that caused displacement of oxygen was identified and repaired.
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| Continuous confirmation of stator cooling water quality and timely corrective actions provides evidence that the program is effective in managing loss of material for stator cooling water system components.
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| Stator cooling water sample results between October 2001 and January 2002 revealed oxygen concentrations below the acceptance criterion of 2 ppm. Feed and bleed operations were used to introduce atmospheric oxygen into the cooling water to correct the oxygen level. Oxygen levels did not go below 0.76 ppm and copper concentrations remained normal with no adverse trend. Continuous confirmation of stator cooling water quality and timely corrective actions provides evidence that the program is effective in managing loss of material for stator cooling water system components.
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| Stator cooling water sample results for the period 1/1/2004 through 9/7/2005 revealed only two instances of a parameter outside the acceptance criteria.
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| On 7/1/04, measured dissolved oxygen was 1.84 ppm. The acceptance criterion for dissolved oxygen is > 2.Oppm and < 8.Oppm. Subsequent readings were within the acceptance criterion and corrective action was not required.
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| On 4/7/05, measured dissolved oxygen was 0.90 ppm. In this instance it was determined that the oxygen probe had failed. Grab sample analysis resulted in a dissolved oxygen reading within acceptance criteria.
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| Continuous confirmation of stator cooling water quality provides evidence that the program is effective in managing loss of material for stator cooling water system components.
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| QA audits in 2000, 2002, and 2004 revealed no issues or findings that could impact effectiveness of the program.Appendix B Aging Management Programs and Activities Page B-1 04 i. .WIi Pilgrim Nuclear Power Station License Renewal Application Technical Information Conclusion The Water Chemistry Control -Auxiliary Systems Program has been effective at managing loss of material for components exposed to treated water. The Water Chemistry Control -Auxiliary Systems Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| B.1.32.2 WATER CHEMISTRY CONTROL -BWR Proaram Descriotion The Water Chemistry Control -BWR Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M2, Water Chemistry.
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| The objective of this program is to manage aging effects caused by corrosion and cracking mechanisms.
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| The program relies on monitoring and control of water chemistry based on EPRI Report 1008192 (BWRVIP-130).
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| BWRVIP-130 has three sets of guidelines:
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| one for primary water, one for condensate and feedwater, and one for control rod drive (CRD) mechanism cooling water. EPRI guidelines in BWRVIP-1 30 also include recommendations for controlling water chemistry in the torus, condensate storage tanks, demineralized water storage tanks, and spent fuel pool.The Water Chemistry Control -BWR Program optimizes the primary water chemistry to minimize the potential for loss of material and cracking.
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| This is accomplished by limiting the levels of contaminants in the RCS that could cause loss of material and cracking.
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| Additionally, PNPS has instituted hydrogen water chemistry (HWC) to limit the potential for IGSCC through the reduction of dissolved oxygen in the treated water.NUREG-1801 Consistenc
| |
| .The Water Chemistry Control -BWR Program is consistent with the program described in NUREG-1801, Section XI.M2, Water Chemistry.
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| Exceptions to NUREG-1801 None Enhancements None Appendix B Aging Management Programs and Activities
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| .Page B-105 Pilgrim Nuclear Power Station License Renewal Application Technical Information Onerating Experience During the period from 1998 through 2004, several condition reports were initiated due to adverse trends in parameters monitored by the Water Chemistry Control -BWR Program.Corrective actions were taken within the Corrective Action Program to preclude reaching unacceptable values for the parameters.
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| Continuous confirmation of water quality and corrective action prior to reaching control limits provide evidence that the program is effective in managing aging effects for applicable components.
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| During the period from 1998 through 2004, several condition reports were initiated due to parameters monitored by the Water Chemistry Control -BWR Program outside of administrative limits, but still within EPRI acceptance criteria.
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| Corrective actions were taken within the Corrective Action Program to preclude violating EPRI acceptance criteria.
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| Continuous confirmation of water quality and corrective action prior to reaching control limits provide evidence that the program is effective in managing aging effects for applicable components.
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| During the period from 1998 through 2004, the following two incidents were found in which parameters monitored by the Water Chemistry Control -BWR Program were outside of EPRI acceptance criteria.* Following a downpower on March 29, 2002, dissolved oxygen measurement from the B high pressure feedwater (HPFW) train was -28 ppb, below the minimum required reading of 30 ppb (EPRI action level 1). Dissolved oxygen measured from the A HPFW train and condensate demineralizer effluent (CDE) were acceptable
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| (- 70 to 80 ppb). Root cause was B HPFW sample line contamination, not actual low oxygen in the feedwater.
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| The B HPFW sample line was replaced.* On October 28, 2002, HPFW and CDE dissolved oxygen levels spiked to 400 to 500 ppb for about 15 minutes before returning to normal. EPRI action level I for HPFW dissolved oxygen is 200 ppb. Root cause was determined to be inadequate filling of the D demineralizer prior to its return to service. The procedure states, "It is EXTREMELY important that all air is vented from a Cond Demin before it is placed in service to prevent air injection into the Feedwater System." Procedural steps were emphasized that will insure proper venting and mitigate elevated oxygen levels in the feedwater system.Continuous confirmation of water quality and timely corrective action provide evidence that the program is effective in managing aging effects for applicable components.
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| QA audits in 2000 and 2002 revealed no issues or findings that could impact effectiveness of the program.A QA audit in 2004 revealed that reactor coolant sodium and lithium analyses were not being performed weekly during the first half of 2004. Corrective action was taken to replace the analysis instrument and ensure required analyses are performed.
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| Confirmation of water quality Appendix B Aging Management Programs and Activities Page B-106 Pilgrim Nuclear Power Station License Renewal Application Technical Information and timely corrective actions provide evidence that the program is effective in managing aging effects for applicable components.
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| A corporate assessment in 2003 identified areas for improvement in administrative controls, but revealed no issues or findings that could impact effectiveness of the program.Conclusion The Water Chemistry Control -BWR Program has been effective at managing aging effects.The Water Chemistry Control -BWR Program at PNPS provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| B.1.32.3 WATER CHEMISTRY CONTROL -CLOSED COOLING WATER Proaram Description The Water Chemistry Control -Closed Cooling Water Program at PNPS is comparable to the program described in NUREG-1801, Section XI.M21, Closed-Cycle Cooling Water System.This program includes preventive measures that manage loss of material, cracking, and fouling for components in closed cooling water systems (reactor building closed cooling water, turbine building closed cooling water, emergency diesel generator cooling water, station blackout diesel cooling water, security diesel generator cooling water, and plant heating).
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| These chemistry activities provide for monitoring and controlling closed cooling water chemistry using PNPS procedures and processes based on EPRI guidance for closed cooling water chemistry.
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| NUREG-1801 Consistency The Water Chemistry Control -Closed Cooling Water Program is consistent with the program described in NUREG-1801, Section XI.M21, Closed-Cycle Cooling Water System, with one exception.
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| Exceptions to NUREG-1801 The Water Chemistry Control -Closed Cooling Water Program is consistent with the program described in NUREG-1801, Section XI.M21, Closed-Cycle Cooling Water System, with the following exception.
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| Attributes Affected Exception 4. Detection of Aging Effects The PNPS Water Chemistry Control -, Closed Cooling Water Program does not include performance and functional testing.1 Appendix B Aging Management Programs and Activities.Page B-107 Pilgrim Nuclear Power Station License Renewal Application Technical Information Exception Note 1. While NUREG-1801, Section XL.M21, Closed-Cycle Cooling Water System endorses EPRI report TR-107396 for performance and functional testing guidance, EPRI report TR-107396 does not recommend that equipment performance and functional testing be part of a water chemistry control program. This appears appropriate since monitoring pump performance parameters is of little value in managing effects of aging on long-lived, passive CCW system components.
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| Rather, EPRI report TR-1 07396 states in section 5.7 (Section 8.4 in EPRI report 1007820)that performance monitoring is typically part of an engineering program, which would not be part of water chemistry.
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| In most cases, functional and performance testing verifies that component active functions can be accomplished and as such would be included as part of Maintenance Rule (10 CFR 50.65). Passive intended functions of pumps, heat exchangers and other components will be adequately managed by the closed cooling water chemistry program through monitoring and control of water chemistry parameters.
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| Enhancements None Operating Experience During the period from 1998 through 2004, several condition reports were initiated due to (;adverse trends in parameters (nitrite and tolytriazole) monitored by the Water Chemistry Control-Closed Cooling Water Program. Corrective actions were taken within the Corrective Action Program to preclude reaching unacceptable values. No increases, long or short term, were observed in iron or copper levels. Continuous confirmation of water quality and corrective action prior to reaching control limits provide evidence that the program is effective in managing aging effects for applicable components.
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| During the period from 1998 through 2004, two condition reports were initiated due to parameters monitored by the Water Chemistry Control -Closed Cooling Water Program outside of administrative limits, but still within EPRI acceptance criteria.
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| Corrective actions were taken within the Corrective Action Program to preclude violating EPRI acceptance criteria.
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| Continuous confirmation of water quality and corrective action prior to reaching control limits provide evidence that the program is effective in managing aging effects for applicable components.
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| During the period from 1998 through 2004, a few incidents were found in which station heating system parameters monitored by the Water Chemistry Control -Closed Cooling Water Program were outside of EPRI action level I acceptance criteria.
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| Monitoring frequency was increased and the parameter was returned to within the prescribed normal operating range as soon as possible (well within the 90 days permitted by action level 1). Continuous confirmation of water quality and timely corrective action provide evidence that the program is effective in managing aging effects for applicable components.
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| Appendix B Aging Management Programs and Activities Page B-108 Pilgrim Nuclear Power Station License Renewal Application Technical Information QA audits in 2000 and 2002 revealed no issues or findings that could impact effectiveness of the program.A self-assessment in October 2003 noted that chemistry specifications and methods of control are not clearly established for nonsafety-related diesel jacket coolant systems. This assessment and a QA audit in early 2004 revealed that corrective actions for condition reports addressing closed cooling water (CCW) analyses had not been completed in a timely manner. Specifically, condition reports initiated in early 2003 identified that for RBCCW, TBCCW and plant heating, some chemical analyses are not being performed in the frequencies defined in procedures due to faulty analysis equipment.
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| In June 2004 corrective actions had not been completed.
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| Corrective actions were taken by the end of 2004 to reinstate all analyses and confirm water quality for the RBCCW, TBCCW, and plant heating systems. Completion of corrective actions and confirmation of water quality provide evidence that the program is effective in managing aging effects for applicable components.
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| When the revised EPRI CCW Guidelines were first implemented (January 2005), new jacket coolant chemistry parameters did not meet recommendations for the EDG, SBO, and security diesels. The parameters that did not meet recommendations are indicators that the glycol and corrosion inhibitor products in the jacket cooling water systems are degrading and becoming less effective.
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| Evaluation determined that there were no immediate concerns of corrosion or cooling ability breakdown for the diesels as other parameter routinely analyzed are in specification and had no adverse trend to indicate an immediate need for action. Work requests were issued to i,>/ change the SBO and security diesel cooling water during the next maintenance window.Evaluation determined that EDG jacket coolant change-out was not warranted.
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| Continuous confirmation of water quality and timely corrective action provide evidence that the program is effective in managing aging effects for applicable components.
| |
| A self-assessment of the Water Chemistry Control -Closed Cooling Water Program was performed in August 2005 to assess how well the program is implementing the revised EPRI CCW guidelines.
| |
| The assessment concluded that open issues remain regarding the tolytriazole achievable limit for the security diesel and reserve alkalinity achievable limit for the EDGs and SBO diesel. Resolution of these open issues is scheduled to assure that the program is effective in managing aging effects for applicable components.
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| Conclusion The Water Chemistry Control -Closed Cooling Water Program has been effective at managing aging effects. The Water Chemistry Control -Closed Cooling Water Program provides reasonable assurance that effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.
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| Appendix B Aging Management Programs and Activities Page B-109 Pilgrim Nuclear Power Station License Renewal Application Technical Information ( %B.2 REFERENCES B.2-1 NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, September 2005.B.2-2 NUREG-1 801, Generic Aging Lessons Learned (GALL) Report, U.S. Nuclear Regulatory Commission, September 2005.Appendix B Aging Management Programs and Activities Page B-110 Appendix C Response to BWRVIP Applicant Action Items Pilgrim Nuclear Power Station Pilgrim Nuclear Power Station License Renewal Application Technical Information Of the BWRVIP documents credited for PNPS license renewal, the following have (or are expected to have) NRC safety evaluation (SE) reports for license renewal.BWRVIP-18 BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines BWRVIP-25 BWR Core Plate Inspection and Flaw Evaluation Guidelines BWRVIP-26 BWR Top Guide Inspection and Flaw Evaluation Guidelines BWRVIP-27 BWR Standby Liquid Control System / Core Plate AP Inspection and Flaw Evaluation Guidelines BWRVIP-38 BWR Shroud Support Inspection and Flaw Evaluation Guidelines BWRVIP-41 BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines BWRVIP-47 BWR Lower Plenum Inspection and Flaw Evaluation Guidelines BWRVIP-48 Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines BWRVIP49 Instrument Penetration Inspection and Flaw Evaluation Guidelines BWRVIP-74 BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines BWRVIP-76 BWR Core Shroud Inspection and Flaw Evaluation Guidelines BWRVIP-116 BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Implementation for License Renewal License renewal application action items identified in the corresponding SE report for each of the above reports are addressed in the following table. BWRVIP-76 and BWRVIP-116 are not included in the table because, although they are expected to have SE reports for license renewal, they have not yet been issued. BWRVIP documents without SE reports for license renewal have no applicant action items and are, therefore, not included in the table.The SE reports contain three common applicant action items, which are addressed only once in the table. For SE reports that contain additional applicant action items, the response is provided separately following the responses to the three common action items.Appendix C Response to BWRVIP Applicant Action Items Page C-2 Pilgrim Nuclear Power Station License Renewal Application Technical Information Action Item Description Response Common Action Items from BWRVIP-18, -25, -26, -27, -38, -41, -47, -48, and -49 BWRVIP-AII (1) The BWRVIP reports have been reviewed The license renewal applicant is to verify that its plant and PNPS has been verified to be bounded is bounded by the report. Further, the renewal applicant by the reports. Additionally, PNPS commits is to commit to programs described as necessary in to programs described as necessary in the the BWRVIP reports to manage the effects of aging BWRVIP reports to manage the effects of during the period of extended operation.
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| Applicants for aging during the period of extended license renewal will be responsible for describing any operation.
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| Commitments are such commitments and identifying how such administratively controlled in accordance commitments will be controlled.
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| Any deviations from the with the requirements of 10 CFR 50, aging management programs within these BWRVIP Appendix B. Site procedures require that reports described as necessary to manage the effects of deviation from a BWRVIP report approved aging during the period of extended operation and to by the NRC will be reported to the NRC maintain the functionality of the components or other within 45 days of receipt of NRC final information presented in the report, such as materials approval of the guideline.
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| of construction, will have to be identified by the renewal applicant and evaluated on a plant-specific basis in accordance with 10 CFR 54.21 (a)(3) and (c)(1).BWRVIP-AII (2) The FSAR supplement is included as 10 CFR 54.21 (d) requires that an FSAR supplement Appendix A and includes a summary of the for the facility contain a summary description of programs and activities specified as the programs and activities for managing the effects of necessary for the BWRVIP program.aging and the evaluation of TLAAs for the period of extended operation.
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| Those applicants for license renewal referencing the applicable BWRVIP report shall ensure that the programs and activities specified as necessary in the applicable BWRVIP reports are summarily described in the FSAR supplement.
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| AppedixC Repone t BWRIP p~lcantActon temsPag C.Appendix C Response to BWRVIP Applicant Action Items Page C-3 Pilgrim Nuclear Power Station License Renewal Application Technical Information Action Item Description Response BWRVIP-AII (3) No technical specification changes have 10 CFR 54.22 requires that each application for been identified for PNPS based upon the license renewal include any technical specification BWRVIP reports.changes (and the justification for the changes) or additions necessary to manage the effects of aging during the period of extended operation as part of the renewal application.
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| The applicable BWRVIP reports may state that there are no generic changes or additions to technical specifications associated with the report as a result of its aging management review and that the applicant will provide the justification for plant-specific changes or additions.
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| Those applicants for license renewal referencing the applicable BWRVIP report shall ensure that the inspection strategy described in the reports does not conflict with or result in any changes to their technical specifications.
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| If technical specification changes or additions do result, then the applicant must ensure that those changes are included in its application for license renewal.Additional Action Items BWRVIP-18, Core Spray Internals Inspection and Flaw Eva luation Guidelines BWRVIP-18 (4) There were no TLAA issues identified for Applicants referencing the BWRVIP-18 report for PNPS for BWRVIP-18.
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| license renewal should identify and evaluate any potential TLAA issues which may impact the structural integrity of the subject RPV internal components.
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| BWRVIP-25, Core Plate Inspection and Flaw Evaluation Guidelines BWRVIP-25 (4) PNPS has installed core plate wedges to Due to susceptibility of the rim hold-down bolts to laterally restrain the core plate. These stress relaxation, applicants referencing the BWRVIP-25 wedges perform the lateral support function report for license renewal should identify and evaluate previously performed by the rim hold down the projected stress relaxation as a potential TLAA issue. bolts. With the wedges in place there is no required preload on the bolts, and hence there is no safety determination made based on the remaining preload on the bolts. As such, loss of preload for the core plate rim hold down bolts is not a TLAA for PNPS.0.Appendix C Response to BWRVIP Applicant Action Items Page C-4 i Pilgrim Nuclear Power Station License Renewal Application Technical InformationAction Item Description Response BWRVIP-25 (5) PNPS follows BWRVIP-25 guidelines for Until such time as an expanded technical basis for rim hold-down bolt inspection under the not inspecting the rim hold-down bolts is approved by BWR Vessel Intemals Program.the staff, applicants referencing the BWRVIP-25 report for license renewal should continue to perform inspections of the rim hold-down bolts.BWRVIP-26, Top Guide Inspection and Flaw Evaluation Guidelines BWRVIP-26 (4) Accumulated neutron fluence projected to Due to IASCC susceptibility of the subject safety- 60 years for PNPS exceeds the threshold related components, applicants referencing the BWRVIP- for IASCC susceptibility for the top guide.26 report for license renewal should identify and Since PNPS has implemented the evaluate the projected accumulated neutron fluence as a inspection requirements of BWRVIP-26, the potential TLAA issue. BWR Vessel Internals Program will adequately manage the effects of aging on the top guide for the period of extended operation.
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| BWRVIP-27, Standby Liquid Control System / Core Plate IAP Intemals Inspection and Flaw Evaluation Guidelines BWRVIP-27 (4) BWRVIP-27 fatigue analysis of the standby Due to the susceptibility of the subject components liquid control system I core plate AP line for to fatigue, applicants referencing the BWRVIP-27 report 60 years is a potential TLAA. However, this for license renewal should identify and evaluate fatigue analysis is applicable only to forged the projected fatigue cumulative usage factors as a low alloy steel nozzles. Since PNPS has an potential TLAA issue. Alloy 600 insert for the standby liquid control system / core plate AP connection to the vessel, this potential TLAA is not applicable to PNPS.BWRVIP-47, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines BWRVIP-47 (4) The only fatigue analysis of lower plenum Due to fatigue of the subject safety-related pressure boundary components is one for components, applicants referencing the BWRVIP-47 the shroud stabilizer, which is a new report for LR should identify and evaluate the projected analysis following modification.
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| Therefore, CUF as a potential TLAA issue, PNPS does not have TLAA associated with lower plenum pressure boundary components that need to be evaluated for license renewal.Appendix C Response to BWRVIP Applicant Action Items Page C-S Appendix C Response to BWRVIP Applicant Action Items Page C-5 Pilgrim Nuclear Power Station License Renewal Application Technical Information Action Item Description Response BWRVIP-74, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines BWRVIP-74-A (4) The vessel flange leak detection (VFLD)The staff is concerned that leakage around the line is in scope and has loss of material and reactor vessel seal rings could accumulate in the VFLD cracking identified as aging effects requiring lines, cause an increase in the concentration of management.
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| Aging of the vessel flange contaminants and cause cracking in the VFLD line. The leak detection line is managed by the Water BWRVIP-74 report does not identify this component as Chemistry Control -BWR and One-Time within the scope of the report. However, since the VFLD Inspection Programs.line is attached to the RPV and provides a pressure boundary function, LR applicants should identify an AMP for the VFLD line.BWRVIP-74-A (5) Descriptions of plant-specific aging LR applicants shall describe how each plant- management programs in Appendix B specific aging management program addresses the address the required ten elements.following elements:
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| (1) scope of program, (2) preventative actions, (3) parameters monitored and inspected, (4)detection of aging effects, (5) monitoring and trending, (6) acceptance criteria, (7) corrective actions, (8) confirmation process, (9) administrative controls, and (10) operating experience.
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| BWRVIP-74-A (6) The Water Chemistry Control -BWR The staff believes inspection by itself is not sufficient Program monitors and controls reactor to manage cracking.
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| Cracking can be managed by water chemistry in accordance with the a program that includes inspection and water guidelines of BWRVIP-130, which chemistry.
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| BWRVIP-29 describes a water chemistry supercedes BWRVIP-29.
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| program that contains monitoring and control guidelines for BWR water that is acceptable to the staff. BWRVIP-29 is not discussed in the BWRVIP-74 report. Therefore, in addition to the previously discussed BWRVIP reports, LR applicants shall contain water chemistry programs based on monitoring and control guidelines for reactor water chemistry that are contained in BWRVIP-29.
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| BWRVIP-74-A (7) The Reactor Vessel Surveillance Program LR applicants shall identify their vessel is an ISP program.surveillance program, which is either an ISP or plant-specific-invessel surveillance program, applicable to the LR term.Q,).Appendix C Response to BWRVIP Applicant Action Items Page C-6 Pilgrim Nuclear Power Station License Renewal Application Technical Information Action Item Description Response BWRVIP-74-A (8) Thermal fatigue (including discussion of LR applicants should verify that the number of cycles, projected cumulative usage factors, cycles assumed in the original fatigue design is environmental fatigue, etc.) is evaluated as conservative to assure that the estimated fatigue usage a TLAA in Section 4.3.for 60 years of plant operation is not underestimated.
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| The use of alternative actions for cases where the estimated fatigue usage is projected to exceed 1.0 will require case-by-case staff review and approval.
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| Further, a LR applicant must address environmental fatigue for the components listed in the BWRVIP-74 report for the LR period.BWRVIP-74-A (9) Development of pressure-temperature Appendix A to the BWRVIP-74 report indicates that a limits for the period of extended operation is set of P-T curves should be developed for the heat-up described as a TLAA in Section 4.2.2.and cool-down operating conditions in the plant at a given EFPY in the LR period.BWRVIP-74-A (10) Discussion of Charpy upper-shelf energy To demonstrate that the beltline materials meet for the period of extended operation is the Charpy USE criteria specified in Appendix B of described as a TLAA in Section 4.2.3.the report, the applicant shall demonstrate that the percent reduction in Charpy USE for their beltline materials are less than those specified for the limiting BWR/3-6 plates and the non-Linde 80 submerged arc welds and that the percent reduction in Charpy USE for their surveillance weld and plate are less than or equal to the values projected using the methodology in RG 1.99, Revision 2.BWRVIP-74-A (11 ) Discussion of relief from in-service To obtain relief from the in-service inspection of inspection of the circumferential welds for the circumferential welds during the LR period, the the period of extended operation is included BWRVIP report indicates each licensee will have to in Section 4.2.5.demonstrate that (1) at the end of the renewal period, the circumferential welds will satisfy the limiting conditional failure frequency for circumferential welds In the Appendix E for the staff's July 28, 1998, SER, and (2) that they have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staff's FSER.Appendix C Response to BWRVIP Applicant Action Items Page C-i Appendix C Response to BWRVIP Applicant Action Items Page C-7 Pilgrim Nuclear Power Station License Renewal Application Technical Information Action Item Description Response BWRVIP-74-A (12) Discussion of axial weld failure probability As indicated in the staff's March 7, 2000, letter to during the period of extended operation is Carl Terry, a LR applicant shall monitor axial beltline included in Section 4.2.6.weld embrittlement.
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| One acceptable method is to determine that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in Table 1 of this FSER.BWRVIP-74-A (13) The method used for the neutron flux The Charpy USE, P-T limit, circumferential weld and axial calculation is described in Section 4.2.1.weld RPV integrity evaluations are all dependent upon the neutron fluence. The applicant may perform neutron fluence calculations using staff approved methodology or may submit the methodology for staff review. If the applicant performs the neutron fluence calculation using a methodology previously approved by the staff, the applicant should identify the NRC letter that approved the methodology.
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| BWRVIP-74-A (14) No flaw evaluations were identified.
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| Components that have indications that have been previously analytically evaluated in accordance with subsection IWB-3600 of Section Xl to the ASME Code until the end of the 40-year service period shall be reevaluated for the 60-year service period corresponding to the LR term.AppedixC Repone t BWRIP pplcantActon temsPag C-(I.Appendix C Response to BWRVIP Applicant Action Items Page C-8 e_)
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| Appendix D Technical Specification Changes 10 CFR 54.22 requires that an application for license renewal include any technical specification changes or additions necessary to manage the effects of aging during the period of extended operation.
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| A review of the information in this License Renewal Application and the Pilgrim Nuclear Power Station Technical Specifications determined that no changes to the Technical Specifications are required.
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| -v Appendix E Applicant's Environmental Report Operating License Renewal Stage Pilgrim Nuclear Power Station DRAFT Introduction Entergy Nuclear Generation Company, Inc. (hereafter referred to as "Entergy"), submits this Environmental Report (ER) in conjunction with the application to the U.S. Nuclear Regulatory Commission (NRC) to renew the operating license for Pilgrim Nuclear Power Station (PNPS) for twenty years beyond the end of the current license. In compliance with applicable NRC requirements, this ER analyzes potential environmental impacts associated with renewal of the PNPS operating license. This ER is designed to assist the NRC staff with the preparation of the PNPS specific Supplemental Environmental Impact Statement required for license renewal.The PNPS ER is provided in accordance with 10 CFR 54.23, which requires license renewal applicants to submit a supplement to the ER that complies with the requirements of Subpart A of 10 CFR 51. This report also addresses the more detailed requirements of NRC environmental regulations in 10 CFR 51.45 and 10 CFR 51.53, as well as the underlying intent of the National Environmental Policy Act, 42 USC 4321 et seq. For major federal actions, the NEPA requires federal agencies to prepare a detailed statement that addresses significant environmental impacts, adverse environmental effects that cannot be avoided if the proposal is implemented, alternatives to the proposed action, and irreversible and irretrievable commitments of resources associated with implementation of the proposed action.Supplement 1 to Regulatory Guide 4.2 -Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses was used as guidance on the (_, format and content of this ER. The level of information provided on the various topics and issues in this ER is commensurate with the environmental significance of the topic or issue.Based upon the evaluations discussed in this ER, Entergy concludes that the environmental impacts associated with renewal of the PNPS operating license are small. No major plant refurbishment activities have been identified as necessary to support the continued operation of PNPS beyond the end of the existing operating license term. Although normal plant maintenance activities may later be performed for economic and operational reasons, no significant environmental impacts associated with such refurbishments are expected.The application to renew the operating license of PNPS assumes that licensed activities are now conducted, and will continue to be conducted, in accordance with the facility's current licensing basis (e.g., use of low enriched uranium fuel only). Changes made to the current licensing basis of PNPS during the staff review of this application are to be made in accordance with the Atomic Energy Act of 1954, as amended, and in accordance with Commission regulations.
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| Pilgrim Nuclear Power Station Applicants Environmental Report Operating License Renewal Stage TABLE OF CONTENTS 1.0 PURPOSE AND NEED FOR THE PROPOSED ACTION ......................
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| 2.0 SITE AND ENVIRONMENTAL INTERFACES
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| ...........................
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| 1-1 2-1 2.1 Location and Features ....................
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| 2.2 Aquatic and Riparian Ecological Communities.
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| .2.2.1 Phytoplankton
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| ......................
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| ..............
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| ........ 2-1 2-1 2-3 2.2.2 Zooplankton
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| ....................
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| 2.2.3 Macroinvertebrates/Shellfish.........
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| 2.2.3.1 American Lobster .............
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| 2.2.4 Fish Community
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| ...................
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| 2.2.4.1 Bottom Trawl Sampling.........
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| 2.2.4.2 Gill Net Sampling .............
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| 2.2.4.3 Haul-Seine Sampling ..........
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| 2.2.4.4 Recreational Creel Surveys .....2.2.4.5 Atlantic Menhaden ............
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| 2.2.4.6 Winter Flounder .........2.2.5 Summary.......................
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| Groundwater Resources................
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| Critical and Important Terrestrial Habitats ...Threatened or Endangered Species........
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| Regional Demography
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| ..................
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| 2.6.1 Regional Population
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| ................
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| 2.6.2 Minority and Low-Income Populations..
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| 2.6.2.1 Background................
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| 2.6.2.2 Minority Populations
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| ...........
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| ................................................................I......................I..........2.3 2.4 2.5 2.6.. ............
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| ...........
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| ...........
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| ........................
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| ;...........
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| ............
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| .................
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| 2-3....................
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| 2-3....................
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| 2-4....................
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| 2-4....................
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| 2-4....................
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| 2-5...................
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| 2-5..... 2-6....................
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| 2-6............
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| .. 2-6.... .2-7......................
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| 2-7....................
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| 2-7............
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| 2-9...................
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| 2-15.....................
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| 2-15.............
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| 2-18...................
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| 2-18...........--------
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| 2-18................
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| 2-19..........
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| 2-23..................
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| 2-25............
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| 2-25....................
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| 2-25....................
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| 2-26...................
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| .2-26..........
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| 2-26.............
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| ..... 2-26....................
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| 2-27....................
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| 2-27...................
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| 2-27............
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| 2-28...................
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| 2-29 2.6.2.3 Low-Income Populations
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| ...2.7 Taxes ....... ...2.8 Land Use Planning .................
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| 2.8.1 Plymouth County ..............
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| 2.8.1.1 Existing Land Use Trends...2.8.1.2 Future Land Use Trends....
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| 2.8.2 Barnstable County ....2.8.2.1 Existing Land Use Trends.. .2.8.2.2 Future Land Use Trends....
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| 2.9 Social Services and Public Facilities
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| ...2.9.1 Public Water Supply...........
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| 2.9.1.1 Plymouth County .........2.9.1.2 Barnstable County .....2.9.1.3 Assessment............
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| ......!,.. .. .. .. ....... ..... .. .. .... .. .. .. .. .. ....~~ ~ ~ .-... .. .. .. .. .. ...I Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.9.2 Transportation
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| .................................................
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| 2-30 2.10 Meteorological and Air Quality .........................................
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| 2-31 2.11 Historic and Archaeological Resources
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| ...................................
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| 2-31 2.11.1 Pre- and Post-Construction Historic/Archaeological Analyses .............
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| 2-31 2.11.2 Additional Information Regarding the Plimoth Plantation/Brown University Archaeological Survey ..............
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| e ...........................
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| 2-32 2.11.3 Current Historic/Archaeological Analysis ..... .........
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| .2-33 2.12 Known and Forseeable Federal and Non-Federal Actions ....................
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| 2-34 2.13 References
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| ..... ...............................
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| .........
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| 2-35 3.0 THE PROPOSED ACTION ................................
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| .... ... 3-1 3.1 Description of the Proposed Action ........................................
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| 3-1 3.2 General Plant Information
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| ..............................................
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| 3-1 3.2.1 Reactor and Containment Systems ......... .......................
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| 3-1 3.2.2 Cooling and Auxiliary Water Systems ................................
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| 3-2 3.2.2.1 Surface Water ...........................
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| 3-2 3.2.2.2 Groundwater
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| ................................
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| 3-2 3.2.3 Radioactive Waste Treatment Processes (Gaseous, Liquid, and Solid) ...... 3-3 3.2.3.1 Liquid Waste Processing Systems and Effluent Controls .............
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| 34 3.2.3.1.1 Clean Radwaste ...................
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| -. -34 3.2.3.1.2 Chemical Radwaste ... ........................
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| 3-5 3.2.3.1.3 Miscellaneous Radwaste .................................
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| 3-6 3.2.3.2 Gaseous Waste Processing Systems and Effluent Controls ..........
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| 3-7 3.2.3.2.1 Air Ejector Offgas and Augmented Offgas System ..... ........ 3-7 3.2.3.2.2 Turbine Sealing and Mechanical Vacuum Pump Systems ........ 3-8 3.2.3.2.3 Miscellaneous Gaseous Effluents (Low Release Potential Effluents) 3-8 3.2.3.2.4 Miscellaneous Gaseous Effluents
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| ...........
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| ...............
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| 3-9 3.2.3.3 Solid Waste Processing
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| .......................
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| I..............
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| 3-10 3.2.3.3.1 Reactor Cleanup Sludge ...........
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| 3-10 3.2.3.3.2 Spent Resin and Miscellaneous Solid Waste System ........ .. 3-11 3.2.3.3.3 Trash Compaction Facility ...........
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| ...................
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| 3-12 3.2.3.3.4 Decontamination and Trash and Laundry Processing Facility....
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| 3-12 3.2.4 Transportation of Radioactive Materials
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| .........................
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| ....... 3-13 3.2.5 Nonradioactive Waste Systems .........
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| ...................
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| I.........
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| 3-13 3.2.6 Maintenance, Inspection, and Refueling Activities
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| .......................
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| 3-13 3.2.7 Transmission Facilities
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| .............
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| ...........
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| 3-14 3.3 Refurbishment Activities
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| ...............................................
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| 3-14 3.4 Programs and Activities for Managing the Effects of Aging ..............
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| ..... 3-15 3.5 Employment
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| ........................................................
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| 3-15 3.6 References
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| ........................................................
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| 3-20 4.0 ENVIRONMENTAL CONSEQUENCES OF THE PROPOSED ACTION ... 4-1 4.1 Water Use Conflicts
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| ..................................................
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| 4-6 ii h Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.1.1 Description of Issue .............................................
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| 4-6 4.1.2 Findings from Table B-1, Appendix B to Subpart A ....... ...............
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| 4-7 4.1.3 Requirement
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| [10 CFR 51.53(c)(3)(kk)(A)
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| ...............................
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| 4-7 4.1.4 Analysis of Environmental Impact ...................................
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| 4-7 4.2 Entrainment of Fish and Shellfish in Early Life Stages .........................
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| 4-7 4.2.1 Description of Issue .4-7 4.2.2 Findings from Table B-1, Appendix B to Subpart A ......................
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| 4-7 4.2.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| .4-7 4.2.4 Background
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| .4-8 4.2.5 Analysis of Environmental Impact .4-8 4.2.6 Conclusion...
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| 4-9 4.3 Impingement of Fish and Shellfish
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| .. 4-9 4.3.1 Description of Issue ..................
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| 4-9 4.3.2 Findings from Table B-1, Appendix B to Subpart A .4-9 4.3.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| .4-9 4.3.4 Background
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| .4-10 4.3.5 Analysis of Environmental Impact .4-10 4.3.6 Conclusion
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| ..................
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| I. 4-10 4.4 Heat Shock ......................................................
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| 4-11 4.4.1 Description of Issue ..............................................
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| 4-11 4.4.2 Findings from Table B-1, Appendix B to Subpart A .4-11 4.4.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| .............................
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| 4-11 4.4.4 Background
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| .4-11 4.4.5 Analysis of Environmental Impact .4-11 4.4.6 Conclusion
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| .4-12 4.5 Groundwater Use Conflicts (Plants Using'>100 gpm of Groundwater)
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| ..4-13 4.5.1 Description of Issue .4-13 4.5.2 Findings from Table B-1, Subpart A, Appendix A .4-13 4.5.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(C)
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| .4-13 4.5.4 Analysis of Environmental Impact ..................................
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| 4-13 4.6 Groundwater Use Conflicts (Plants Using Cooling Towers Withdrawing Make-Up Water from a Small River) .. 4-13 4.6.1 Description of Issue .4-13 4.6.2 Findings from Table B-1, Appendix B to Subpart A .4-14 4.6.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(A)
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| .4-14 4.6.4 Analysis of Environmental Impact .4-14 4.7 Groundwater Use Conflicts (Plants Using ,Ranney Wells) ..4-14 4.7.1 Description of Issue .4-14 4.7.2 Findings from Table B-1, Subpart A, Appendix A .4-14 4.7.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(C)]
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| .4-14 4.7.4 Analysis of Environmental Impact .4-14 iIJ iii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.8 Degradation of Groundwater Quality .,.................................
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| 4-15 4.8.1 Description of Issue ...................
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| 4-15 4.8.2 Findings from Table B-1, SubpartA, AppendixA
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| ..........
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| .............
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| 4-15 4.8.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(D)]
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| ... 1............
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| ................
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| 4-15 4.8.4 Analysis of Environmental Impact ..................................
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| 4-15 4.9 Impacts of Refurbishment on Terrestrial Resources
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| .........................
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| 4-15 4.9.1 Description of Issue ..... 4-15 4.9.2 Findings from Table B-1, Subpart A, Appendix A .........
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| .............
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| 4-15 4.9.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(E)]
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| .... .....................
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| ..... 4-15 4.9.4 Analysis of Environmental Impact ...............................
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| 4-16 4.10 Threatened or Endangered Species .................................-.
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| 4-16 4.10.1 Description of Issue ...............
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| .................
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| 4-16 4.10.2 Findings from Table B-1, Appendix B to Subpart A ................
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| I..... 4-16 4.10.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(E)]
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| .............................
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| 4-16 4.10.4 Background
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| ..................
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| .... 4-16 4.10.5 Analysis of Environmental Impacts ................................
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| 4-16 4.10.6 Conclusion
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| ............
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| 4-18 4.11 Air Quality During Refurbishment (Nonattainment and Maintenance Areas) ...... 4-18 4.11.1 Description of Issue ......................
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| ........................
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| 4-18 4.11.2 Findings from Table B-1, Subpart A, Appendix A .........
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| .............
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| 4-18 4.11.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(F)]
| |
| ...........
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| ....... 4-19 4.11.4 Analysis of Environmental Impact ...............
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| ....................
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| 4-19 4.12 Impact on Public Health of Microbiological Organisms
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| ...........
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| ... .........
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| 4-19 4.12.1 Description of Issue ....................
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| 4-19 4.12.2 Finding from Table B-1, Appendix B to Subpart A .........
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| .... .........
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| 4-19 4.12.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(G)]
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| ......................
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| ....... 4-19 4.12.4 Analysis of Environmental Impact .... ........ 4-19 4.13 Electromagnetic Fields-Acute Effects ....... ...........................
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| 4-19 4.13.1 Description of Issue ........................................
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| ... 4-19 4.13.2 Findings from Table B-1, Subpart A, Appendix A ..........
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| ........ I..... 4-20 4.13.3 Requirements
| |
| [10 CFR 51.53(c)(3)(ii)(H)]
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| .............
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| 4-20 4.13.4 Background
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| ...........................................
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| 4-20 4.13.5 Analysis of Environmental Impact .. ................................
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| 4-20 4.13.6 Conclusion
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| .....................
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| 4-22 4.14 Housing Impacts ..... 4-22 4.14.1 Description of Issue .......................................... 22 4.14.2 Findings from Table B-1, Appendix B to Subpart A .....................
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| 4-22 4.14.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(1)]
| |
| ...............
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| I ...............
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| 4-22 4.14.4 Background
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| .......................
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| 4-22 4.14.5 Analysis of Environmental Impact .............................
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| 4-23 4.14.6 Conclusion
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| ......................
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| 4-23 iv 0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage A,,-Q 4.15 Public Utilities:
| |
| Public Water Supply Availability.
| |
| .........................
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| .4-24 4.15.1 Description of Issue .............
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| 4.15.2 Findings from Table B-1, Appendix B to Subpart A ..............
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| 4.15.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(1)
| |
| ................
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| 4.15.4 Public Water Supply -Background
| |
| ..........................
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| 4.15.5 Analysis of Environmental Impact .......................
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| 4.15.6 Conclusion
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| .....4~~~~.1. 6 o cu in... .. .. ... .. .. ... .... ... ... ... ... ..4.16 Education Impacts from Refurbishment
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| ...........................
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| 4.16.1 Description of Issue ..................................
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| 4.16.2 Findings from Table B-1, Appendix B to Subpart A ..............
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| 4.16.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(1)]
| |
| .......................
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| 4.16.4 Analysis of Environmental Impact ...........................
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| 4.17 Offsite Land Use-Refurbishment
| |
| ..............................
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| 4.17.1 Description of Issue ......................................
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| 4.17.2 Findings from Table B-1, Appendix B to Subpart A .............
| |
| 4.17.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(1)]
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| ..................
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| 4.17.4 Analysis of Environmental Impact ..........................
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| 4.18 Offsite Land Use-License Renewal Term ..............
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| i 4.18.1 Description of Issue ......................................
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| 4.18.2 Findings from Table B-1, Appendix B to Subpart A ..............
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| 4.18.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(1)]
| |
| ......................
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| 4.18.4 Background
| |
| ..........................................
| |
| ..4-24..4-24..4-24..4-24..4-24..4-25.... 4-25..4-25..4-25..4-25..4-25..4-25..4-25..4-26 ,.4-26-..4-26..4-26..4-26..4-26..4-26..4-26 4.18.5 Analysis of Environmental Impact .4.18.5.1 Population-Driven Land Use Changes.....
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| 4.18.5.2 Tax-Driven Land Use Changes ...........
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| 4.18.6 Conclusion.
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| ........4.19 Transportation
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| .................................
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| ....................
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| 4-27....I...................
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| ............
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| ........ 4-27........ 4-28..........
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| 4-28..........
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| 4-29 4.19.1 Description of Issue. ................
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| 4-29 4.19.2 Finding from Table B-1, Appendix B to Subpart A .......................
| |
| .4-29 4.19.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(J)]
| |
| ................
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| 4-29 4.19.4 Background
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| .4-29 4.19.5 Analysis of Environmental Impact ................................
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| 4-29 4.19.6 Conclusion
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| ................................
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| .l 4-30 4.20 Historic and Archaeological Properties
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| ......................
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| .... 4-30 4.20.1 Description of Issue .4-30 4.20.2 Finding from Table B-1, Appendix B to Subpart A ...! .4-30 4.20.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(K)]
| |
| ..4-30 4.20.4 Background
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| ..4-30 4.20.5 Analysis of Environmental Impact .........................
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| 4-30 4.20.6 Conclusion
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| ..4-31 4.21 Severe Accident Mitigation Alternatives
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| ...4-31 4.21.1 Description of Issue.. 4-31 v Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage H%-i 4.21.2 Finding from Table B-1, Appendix B to Subpart A ............
| |
| ...I...........
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| 4-31 4.21.3 Requirement
| |
| [10 CFR 51.53(c)(3)(ii)(L)]
| |
| .... ...........
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| 4-31 4.21.4 Background
| |
| .............
| |
| ....... 4-31 4.21.5 Analysis of Environmental Impact ..............
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| .I 4-32 4.21.5.1 Establish the Baseline Impacts of a Severe Accident .... 4-35 4.21.5.1.1 The PSA Model-Level I and Level 2 Analysis .4-35 4.21.5.1.2 The PSA External Events Model -Individual Plant Examination of External Events (IPEEE) Model .4-35 4.21.5.1.3 The MACCS2 Model -Level 3 Analysis .4-36 4.21.5.1.4 Evaluation of Baseline Severe Accident Impacts Using the Regulatory Analysis Technical Evaluation Handbook Method .... 4-36 4.21.5.2 Identify SAMA Candidates
| |
| .4-44 4.21.5.3 Preliminary Screening (Phase I) .4-44 4.21.5.4 Final Screening and Cost Benefit Evaluation (Phase II) .4-44 4.21.5.5 Sensitivity Analysis ............
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| .4-48 4.21.6 Conclusion
| |
| .............
| |
| 4......- 449 4.22 Environmental Justice .. .4-52 4.22.1 Description of Issue ..............................................
| |
| 4-52 4.22.2 Finding from Table B-1, Appendix B to Subpart A. 4-52 4.22.3 Requirement
| |
| ..................
| |
| 4-52 4.22.4 Background
| |
| .. 52 4.22.5 Analysis .. 4-52 4.22.6 Conclusion
| |
| .. 4-53 4.23 References.
| |
| ................
| |
| i .........................
| |
| 5.0 ASSESSMENT OF NEW AND SIGNIFICANT INFORMATION..............
| |
| 6.0
| |
| | |
| ==SUMMARY==
| |
| OF LICENSE RENEWAL IMPACTS AND MITIGATING ACTIONS ...4-54.. 5-1.. 6-1 6.1 License Renewal Impacts .........................
| |
| 6.2 Mitigation.
| |
| .....................................
| |
| 6.2.1 Requirement
| |
| [10 CFR 51.53(c)(3)(iii)]
| |
| ............
| |
| 6.2.2 Entergy Response ...........................
| |
| 6.3 Unavoidable Adverse Impacts ......................
| |
| 6.3.1 Requirement
| |
| [10 CFR 51.45(b)(2)]
| |
| ..............
| |
| 6.3.2 Entergy Response ...........................
| |
| 6.4 Irreversible or Irretrievable Resource Commitments.
| |
| 6.4.1 Requirement
| |
| [10 CFR 51.45(b)(5)]
| |
| ..............
| |
| 6.4.2 Entergy Response ...........................
| |
| 6.5 Short-Term Use Versus Long-Term Productivity
| |
| ........6.5.1 Requirement
| |
| [10 CFR 51.45(b)(4)]
| |
| ..............
| |
| 6.5.2 Entergy Response ...........................
| |
| 6.6 References
| |
| ...................................
| |
| ..............
| |
| ..............
| |
| ..............
| |
| ...............
| |
| .............................. 6-1...... 6-1...... 6-1...... 6-1...... 6-4...... 6-4...... 6-4...... 6-5...... 6-5...... 6-5...... 6-5...... 6-5...... 6-5...... 6-6.... .. ....... .. ....... .. ...vi QW Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 7.0 ALTERNATIVES CONSIDERED
| |
| ................
| |
| 7-1 7.1 Introduction
| |
| .. ........................
| |
| 7-1 7.2 Proposed Action. ..............................-...........
| |
| ..........
| |
| 7-1 7.3 No-Action Alternative
| |
| .......................
| |
| ........ ................
| |
| 7-1 7.4 Decommissioning Impacts ........................
| |
| 7-2 7.5 Alternative Energy Sources ........................
| |
| 7-3 7.6 References...................................
| |
| 7-5 8.0 COMPARISON OF IMPACTS ..............
| |
| .... 8-1 8.1 Comparison of Environmental Impacts for Reasonable Alternatives
| |
| .........
| |
| .... 8-1 8.1.1 Coal-Fired Generation
| |
| .............................. 2 8.1.1.1 Closed-Cycle Cooling System ...............................
| |
| 8-7 8.1.1.1.1 Land Use ..... .8-7 8.1.1.1.2 Ecology..............................
| |
| ................
| |
| 8-7 8.1.1.1.3 Water Use and Quality ............................
| |
| 8-8 8.1.1.1.4 Air Quality ...........................
| |
| 8-8 8.1.1.1.5 Waste ............................
| |
| 8-9 8.1.1.1.6 Human Health ............................
| |
| 8-10 8.1.1.1.7 Socioeconomics
| |
| ........ ...................
| |
| 8-10 8.1.1.1.8 Aesthetics
| |
| ............................
| |
| 8-11-8.1.1.1.9 Historic and Archaeological Resources
| |
| .....................
| |
| 8-11 8.1.1.2 Once-Through Cooling System .............................
| |
| 8-14 8.1.2 Gas-Fired Generation
| |
| ...............................
| |
| 8-15 8.1.2.1 Closed-Cycle Cooling System ...............................
| |
| 8-18 8.1.2.1.1 Land Use ............................
| |
| 8-18 8.1.2.1.2 Ecology ............................
| |
| 8-18 8.1.2.1.3 Water Use and Quality .............................
| |
| 8-19 8.1.2.1.4 Air Quality ............................
| |
| 8-19 8.1.2.1.5 Waste ............................
| |
| 8-20 8.1.2.1.6 Human Health ............................
| |
| 8-20 8.1.2.1.7 Socioeconomics..
| |
| ...... 8-20 8.1.2.1.8 Aesthetics
| |
| ............................
| |
| 8-21 8.1.2.1.9 Historic and Archaeological Resources
| |
| ............
| |
| .........
| |
| 8-22 8.1.2.2 Once-Through Cooling System ...............................
| |
| 8-24 8.1.3 Nuclear Power Generation
| |
| ...............................
| |
| 8-26 8.1.3.1 Closed-Cycle Cooling System ............................
| |
| 8-26 8.1.3.1.1 Land Use .8-26 8.1.3.1.2 Ecology .8-26 8.1.3.1.3 Water Use and Quality .8-27 8.1.3.1.4 Air Quality .8-27 8.1.3.1.5 Waste .8-27 8.1.3.1.6 Human Health .8-27 8.1.3.1.7 Socioeconomics
| |
| .8-28 vii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.3.1.8 Aesthetics
| |
| ..............
| |
| 8.1.3.1.9 Historic and Archeological R.828... .8-resources......................
| |
| 8-28 8.1.3.2 Once-Through Cooling System .8.1.4 Purchased Electrical Power.8.2 Alternatives Not Within the Range of Reasonable Alternatives.
| |
| .8.2.1 Wind .8.2.2 Solar.8.2.3 Hydropower.
| |
| 8.2.4 Geothermal
| |
| .8.2.5 Wood Energy .8.2.6 Municipal Solid Waste .8.2.7 Other Biomass-Derived Fuels .8.2.8 Oil.8.2.9 Fuel Cells .8.2.10 Delayed Retirement
| |
| .8.2.11 Utility-Sponsored Conservation 8.2.12 Combination of Alternatives
| |
| ...Proposed Action vs. No-Action.
| |
| Summary.References.
| |
| ..................................................
| |
| .................
| |
| ..................
| |
| ..................
| |
| ...........
| |
| ... ...................
| |
| ...................
| |
| ..... .. .. .. .. ..;... .. .. .. .. .. .. .. ...................
| |
| ...............
| |
| ..................
| |
| ..............
| |
| 8-30..............
| |
| 8-31..............
| |
| 8-32................
| |
| 8-32..............
| |
| 8-33..............
| |
| 8-34................
| |
| 8-34..............
| |
| 8-34..............
| |
| 8-35............
| |
| 8-35..............
| |
| 8-35..............
| |
| 8-36.............. 36................
| |
| 8-37... .8-37..............
| |
| .8-38..............
| |
| .8-38..8-41................
| |
| 9-1...............
| |
| 9-1...............
| |
| .9-1........ 9-1................
| |
| 9-1....... 9-2................
| |
| 9-7 8.3 8.4 8.5 9.0 STATUS OF COMPLIANCE
| |
| ...........................
| |
| 9.1 Requirement
| |
| [10 CFR 51.45(d)]
| |
| ........................
| |
| 9.2 Environmental Permits ...............................
| |
| 9.2.1 Coastal Zone Management Program Compliance.
| |
| 9.2.2 Water Quality (401) Certification
| |
| ...................
| |
| 9.3 Environmental Permits -Discussion of Compliance.........
| |
| 9.4 References
| |
| .....................................
| |
| viii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage LIST OF TABLES Table 2-1 Endangered and Threatened Species that Occur in the Vicinity of PNPS or in Plymouth County, MA ...........................................
| |
| 2-11 Table 2-2 Estimated Populations and Annual Growth Rates in Plymouth and Bamstable Counties 1980-2040
| |
| ................................................
| |
| 2-17 Table 2-3 Minority and Low-income Population Information
| |
| ...........................
| |
| 2-21 Table 2-4 Property Taxes .............
| |
| I 2-24 Table 2-5 Selected Plymouth County Public Water Suppliers and Capacities for the Year2003 ...........
| |
| 2 2-28 Table 2-6 Barnstable County Public Water Suppliers and Capacities for the Year 2003 .... 2-29 Table 2-7 Traffic Counts for Roads in the Vicinity of PNPS .......................
| |
| .2-30 Table 2-8 Town of Plymouth, Massachusetts, Sites Listed in the National Register of Historic Places and/or the State Register of Historic Places .................
| |
| .. 2-33 Table 3-1 Employee Residence Information, PNPS, February 2005 ...................
| |
| 3-16 Table 4-1 Category I Issues Not Applicable to PNPS ...............................
| |
| 4-2 Table 4-2 Category I Issues Applicable to PNPS .................
| |
| .................
| |
| 4-3 Table 4-3 Estimated Present Dollar Value Equivalent of Internal Events CDF at PNPS ... 4-43 Table 4-4 Final SAMAs ......................................................
| |
| 4-50 Table 6-1 Environmental Impacts Related to License Renewal at PNPS .................
| |
| 6-2 Table 8-1 Coal-Fired Alternative Emission Control Characteristics
| |
| .....................
| |
| 8-3 ix Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-2 Air Emissions from Coal-Fired Alternative
| |
| .........
| |
| .......................
| |
| 8-4 Table 8-3 Solid Waste from Coal-Fired Alternative
| |
| ..................................
| |
| 8-6 Table 8-4 Summary of Environmental Impacts from Coal-Fired Generation Using Closed-Cycle Cooling at an Alternate Greenfield Site ...............
| |
| I.. 8-12 Table 8-5 Summary of Environmental Impacts from Coal-Fired Generation Using Once-Through Cooling at an Alternate Greenfield Site ...............
| |
| 8-14 Table 8-6 Gas-Fired Alternative Emission Control Characteristics
| |
| .8-16 Table 8-7 Air Emissions from Gas-Fired Alternative
| |
| .8-17 Table 8-8 Summary of Environmental Impacts from Gas-Fired Generation Using Closed-Cycle Cooling at PNPS or at Alternate Greenfield Site .8-22 Table 8-9 Summary of Environmental Impacts from Gas-Fired Generation Using Once-Through Cooling at PNPS or at an Alternate Grebnfield Site .8-24 Table 8-10 Summary of Environmental Impacts from Nuclear Power Generation Closed-Cycle Cooling at Alternate Greenfield Site .8-29 Table 8-11 Summary of Environmental Impacts from Nuclear Power Generation Using Once-Through Cooling at Alternate Greenfield Site .8-30 Table 9-1 Environmental Authorizations for PNPS License Renewal .9-2 Table 9-2 Environmental Authorizations for Current PNPS Operations
| |
| .9-4 x Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage LIST OF FIGURES Figure 2-1 50-Mile Vicinity Map ....................
| |
| 240 Figure 2-2 General Area Near PNPS ....................
| |
| ' 2-41 Figure 2-3 Site Boundary ...............................................
| |
| .242 Figure 24 American Indian or Alaskan Native Minority Population Map ..... ...........
| |
| 243 Figure 2-5 Asian or Pacific Islander Minority Population Map .........................
| |
| 2-44 Figure 2-6 Native Hawaiian or Other Pacific Islander Minority Population Map. 245 Figure 2-7 iiI Black Races Minority Population Map. 246 Figure 2-8 All Other Single Minorities Map .2-47 Figure 2-9 Aggregate of Minority Races Population Map. 248 Figure 2-10 Hispanic Minority Population Map .249 Figure 2-11 Low-income Population Map .2-50 Figure 2-12 State and Federal Lands-50 Mile Radius .2-51 Figure 3-1 Station Layout .3-21 xi Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage LIST OF ATTACHMENTS Attachment A Attachment B Attachment C Attachment D Attachment E NPDES Permit and Water Quality Certification Special Status Species Correspondence Massachusetts Historical Commission Correspondence Coastal Zone Management Consistency Certification Severe Accident Mitigation Alternatives (SAMA)xii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ACRONYMS AND ABBREVIATIONS ABWR AC ADS AEC ALARA AOG AQCR ASOS ATWS Btu BWR BWROG advanced boiling water reactor alternating current automatic depressurization system Atomic Energy Commission as low as reasonably achievable augmented off-gas Air Quality Control Region automated surface observatory system anticipated transient without scram British thermal unit boiling water reactor Boiling Water Reactor Owners Group CaO CAPB CaSO 4 2H 2 0 CDF CEQ CET CFR CMR CO calcium oxide (lime)collapsed accident progression bins calcium sulfate dihydrate core damage frequency Council on Environmental Quality containment event tree Code of Federal Regulations Code of Massachusetts Regulations carbon monoxide xiii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage CPUE catch per unit effort Acronyms and Abbreviations (continued)
| |
| Csl cesium iodide CST condensate storage tank CWA Clean Water Act DC direct current DCH direct containment heating DECON decontamination and dismantlement DOE United States Department of Energy DOT U. S. Department of Transportation DSM demand side management DTV direct torus vent ECCS emergency core cooling system EDG emergency diesel generator EIA Energy Information Administration ENSR ENSR Corporation EPA U.S. Environmental Protection Agency EPG emergency plant guidelines EPRI Electric Power Research Institute ER environmental report EREN Energy Efficiency and Renewable Energy Network xiv Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Acronyms and Abbreviations (continued)
| |
| FES Final Environmental Statement FHA Federal Highway Administration FIVE fire induced vulnerability evaluation ft 3 cubic feet FWS U.S. Fish and Wildlife Service gal gallon GE General Electric GEIS Generic Environmental Impact Statement GIS geographic information system gpm gallons per minute HEP human error probability HIC high integrity container HPCI high pressure coolant injection HRA human reliability analysis IDCOR Industrial Degraded Core Rulemaking INEL Idaho National Engineering Laboratory IPA integrated plant assessment IPE individual plant examination IPEEE individual plant examination of external events ISLOCA interface system loss of coolant accident ISO International Standards Organization xv Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Acronyms and Abbreviations (continued)
| |
| IORV inadvertent stuck open relief valve KM kilometer kV kilovolts kWh kilowatt-hour lb pound LERF large early release frequency LLRWSF low-level radwaste storage facility LOCA loss of coolant accident LOOP loss of offsite power LPCI low pressure core injection MACCS2 Melcor Accident Consequences Code System 2 MAPC Metropolitan Area Planning Council MCC motor control center MCZM Massachusetts Coastal Zone Management MDEP Massachusetts Department of Environmental Protection MDFW Massachusetts Division of Fisheries and Wildlife MDTE Massachusetts Department of Telecommunications and Energy MG million gallons MGD million gallons per day MGL Massachusetts General Laws MISER Massachusetts Institute for Social and Economic Research MM million xvi Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Acronyms and Abbreviations (continued)
| |
| MOV motor-operated valve mrad millirad mrem millirem MSIV main steam isolation valve MW megawatt MWe megawatts, electric MWt megawatts, thermal NA not applicable NEI Nuclear Energy Institute NEPA National Environmental Policy Act NESC National Electric Safety Code NHESP Natural Heritage and Endangered Species Program NMFS National Marine Fisheries Service NOx oxides of nitrogen NPDES National Pollutant Discharge Elimination System NRC U.S. Nuclear Regulatory Commission NREL National Renewable Energy Laboratory NSPS New Source Performance Standard ODCM Offsite Dose Calculation Manual OECR offsite economic cost risk PCS primary containment system xvii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Acronyms and Abbreviations (continued)
| |
| PDR population dose risk PDS plant damage states PM 1 0 particulate matter with diameter less than 10 microns PNPS Pilgrim Nuclear Power Station ppm parts per million PRA probabilistic risk assessment PSA probabilistic safety analysis RAI RBCCW RCIC RHR RPS RPV RRW RWCU Request for Additional Information reactor building closed cooling water reactor core isolation cooling residual heat removal reactor protection system reactor pressure vessel risk reduction worth reactor water cleanup IC SAFSTOR SAMA SAMDA SBO SCR SGTS SHPO safe storage severe accident mitigation alternatives severe accident mitigation design alternatives station blackout selective catalytic reduction standby gas treatment system State Historic Preservation Officer xviii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Acronyms and Abbreviations (continued)
| |
| SLC standby liquid control Sox oxides of sulfur SQUG Seismic Qualification Utility Group SRV safety relief valve SSCs systems, structures, and components SSW salt service water TCA Tennessee Code Annotated TCF trash compaction facility T-H thermal-hydraulic THERP technique for human error rate probability TSP total suspended particulates TtNUS Tetratech NUS TVA Tennessee Valley Authority UFSAR Updated Final Safety Analysis Report URC ultrasonic resin cleaner USC United States Code USCB U. S. Census Bureau WMS waste management system yr year xix Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 1.0 PURPOSE AND NEED FOR THE PROPOSED ACTION For license renewal, the NRC has adopted the following definition of purpose and need, stated in Section 1.3 of NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants: "The purpose and need for the proposed action (renewal of an operating license) is to provide an option that allows for power generation capability beyond the term of a current nuclear power plant operating license to meet future system generating needs, as such needs may be determined by State, utility, and, where authorized Federal (other than NRC)decision makers.'Nuclear power plants are licensed by the NRC to operate up to 40 years, and the licenses may be renewed [10 CFR 50.511 for periods up to 20 years. As stated in 10 CFR 54.17(c), "[a]n application for a renewed license may not be submitted to the Commission earlier than 20 years before the expiration of the operating license currently in effect." The proposed action is to extend the operating license for PNPS for a period of 20 years beyond the current operating license expiration date. For PNPS (Facility Operating License DPR-35), the requested renewal would extend the existing license expiration date from midnight June 8, 2012, until midnight June 8, 2032.1-1 (
| |
| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.0 SITE AND ENVIRONMENTAL INTERFACES 2.1 Location and Features PNPS is located on the western shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts.
| |
| It is 38 miles southwest of Boston, Massachusetts, and 44 miles east of Providence, Rhode Island. Approximately 60% of the area within a 50-mile radius is open water.Figure 2-1 and Figure 2-2 are PNPS 50-mile and 6-mile vicinity maps, respectively.
| |
| Access to the site is available by road or from Cape Cod Bay. Land access is provided by a private two-lane paved road, which connects PNPS with Route 3A, which leads to Plymouth, White Horse Beach, and nearby Route 3. Alternate access to Plymouth and Route 3, via Route 3A, is provided by Rocky Hill Road. Immediately south of the intake is a boat landing providing sea access to the site. The landing is used for off-loading large equipment or large structural assemblies from barges.The industrial facility encompasses approximately 140 acres (Figure 2-3). In addition, approximately 1,500 acres owned by Entergy is in a forest management trust. The nearest residences lie outside the site boundary to the northwest.
| |
| The nearest residence is 2395 feet (0.45 mile) from the reactor. A single tract of land within Entergy's property is still owned by a private party. Entergy has made no arrangements with the current owner regarding future use or occupancy of the property.
| |
| The tract is outside the NRC-mandated 1,800-foot buffer between z ;the reactor and the nearest residence.
| |
| The site boundary (Figure 2-3) is posted and a perimeter security fence surrounds the protected area of the station.The principal structures at PNPS consist of the reactor and turbine buildings (each with auxiliary bays), the offgas retention building, the radwaste building, the diesel generator building, the administration building, the intake structure, and the main stack [Reference 2-37, Section 12.1].The reactor and nuclear steam supply system for PNPS, along with the mechanical and electrical systems required for the safe operation of PNPS, are primarily located in the reactor building.Figure 3-1 shows the general features of PNPS and the station layout. Figure 2-3 shows the site boundaries.
| |
| No residences are permitted within this exclusion zone.State and Federal lands within a 50-mile radius are shown in Figure 2-12.The nearest population centers are Boston, Massachusetts, and Providence, Rhode Island. The region within 6 miles of the site (Figure 2-2) is completely within Plymouth County and includes part of the Town of Plymouth, the nearest urbanized area. Topography consists of rolling forested hills interspersed with urban areas and a small number of agricultural areas, the majority of which are cranberry bogs. The area within 2 miles of PNPS is developed with permanent and seasonal residences in Plymouth, Priscilla Beach, and White Horse Beach.Section 3.2 describes key features of PNPS, including reactor and containment systems, cooling and auxiliary water systems, radwaste system, and transmission facilities.
| |
| 2-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.2 Aquatic and Riparian Ecoloaical Communities PNPS lies on the western shore of Cape Cod Bay near Plymouth, Massachusetts (Figure 2-1).Cape Cod Bay has a surface area of approximately 430 square nautical miles, or about 365,000 acres [Reference 2-1, Section 2.D0. Water in Cape Cod Bay tends to circulate counterclockwise; as a result, there is a consistent net flow of water to the south along the coast in the general vicinity of PNPS. However, this circulation pattern is less evident in the shallow waters-(<
| |
| 30 feet deep) immediately offshore of PNPS, where submarine ledges disrupt the typical north-to-south movement of water. Water of the bay is exchanged by at least three processes:
| |
| (1) tidal exchange, (2) the general counter-clockwise circulation, and (3) wind-induced motion.Approximately 10% of the total volume of water in the bay is exchanged daily by these processes[Reference 2-1, Section 2.D].Water temperatures in the vicinity of the station show typical annual cycles. Highest surface temperatures typically occur in August, when temperatures average around 65 0 F and are as high as approximately 73 0 F [Reference 2-1, Section 2.D]. Summer water temperatures tend to fluctuate dramatically, however, and may dip into the low 40s. Lowest surface water temperatures occur between December and March, when mean temperatures range between 30'F and 40 0 F [Reference 2-1, Section 2.D]. In summer and early fall, surface water temperature may be up to 100 warmer than bottom temperatures
| |
| [Reference 2-1, Section 2.D]. A weak thermocline may be present at these times of year.The Final Environmental Statement (FES) [Reference 2-1] briefly describes the biological communities of the PNPS area, focusing on two species of commercial importance, the 0 American lobster (Homarus americanus) and the marine alga Irish moss (Chrondrus crispus).
| |
| At the time the FES was written, as many as 10,000 lobster pots were fished between the two submarine ledges, Rocky Point and White Horse, that bracket the site, and the 50-foot contour, an area of roughly one square mile [Reference 2-1, Section 2.E]. In 1970, roughly half of the lobsters brought ashore at Plymouth were captured in this general area. Irish moss is a periphytic marine alga that contains carrageenan, which is used as a stabilizing agent in paints, medicines, and foods. It was harvested in the area of PNPS until the 1990s.At the time the FES was written, mollusks were not found in large numbers in the vicinity of the station. This was attributed to the absence of suitable substrate.
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| Groundfish (e.g., cod, haddock, winter flounder, and hake) were not sought by commercial fishermen in the vicinity of PNPS in the early 1970s as regulations restricted commercial fishing in Cape Cod Bay to areas at least 3 miles from shore between April 1 and November 1. Inshore trawling for winter flounder was permitted from November to March, with an annual catch of approximately 115,000 pounds[Reference 2-1, Section 2.E]. Sport fishing for inshore species such as tautog, bluefish, and flounder was relatively unimportant in the vicinity of the station and sport fishing for pelagic species such as tunas, striped bass, and mackerel was difficult because of the many lobster pots and their floats.The March 2000 316 Demonstration Report -Pilgrim Nuclear Power Station [Reference 2-121 is an up-to-date source of information on the aquatic communities of western Cape Cod Bay, including those in the vicinity of PNPS. This report summarizes research and monitoring studies 2-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage conducted since the late 1960s by Boston Edison Company and its contractors, Entergy and its contractors, university researchers, and state and federal resource agencies.
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| Although focused on the potential impacts of PNPS operations, it contains a wealth of baseline information on the marine life of Plymouth Bay, Cape Cod Bay, and the Gulf of Maine.2.2.1 Phytoplankton The phytoplankton community of western Cape Cod Bay, including the vicinity of PNPS, appeared to be more similar to the Gulf of Maine (the area north of Cape Cod) than to the community to the south of Cape Cod [Reference 2-12, Section 4.2.1]. In the vicinity of PNPS, phytoplankton density showed two annual peaks, one in early spring and another in mid-summer. Lowest densities were observed in mid-winter.
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| Diatoms dominated collections in the 1970s. Monitoring studies of Massachusetts Bay and Cape Cod Bay in the 1990s to assess impacts of an offshore sewage outfall in Boston Harbor showed the nuisance phytoflagellate Phaetocystis poucheti dominating collections in early spring and microflagellates and diatoms dominating collections in the fall [Reference 2-12, Section 4.2.1]. The increased abundance of nuisance phytoplankton species in Cape Cod Bay may be related to water quality degradation.
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| Spring blooms of Phaetocystis pouchetii are a regular occurrence in coastal portions of the Gulf of Maine, and are associated with eutrophication in coastal waters [Reference 2-16].2.2.2 Zooplankton Zooplankton abundance showed seasonal cycles, with highest densities in late summer and lowest densities in late winter [Reference 2-12, Section 4.2.2]. Copepods, especially Acartia clausi and A. tonsi, dominated samples, with two distinct species aggregations, inshore and offshore [Reference 2-12, Section 4.2.2]. Differences in species composition were attributed to higher nutrient levels in inshore areas.2.2.3 Macroinvertebrates/Shellfish Macroinvertebrates are found in four kinds of habitats near PNPS: rocky intertidal, rocky subtidal, sandy intertidal, and sandy subtidal.
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| The common barnacle, Balanus balanus, is ubiquitous in rocky intertidal areas near PNPS and is the dominant macrofaunal organism in the upper rocky intertidal zone [Reference 2-12, Section 4.2.4.1].
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| The marine gastropods Littorira Iittorea and Littorira obtusata are also common In this zone. In the middle and lower intertidal zones, Balanus is often replaced by the blue mussel (Mytilus edulis) and macroalgae.
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| Sessile species in the rocky intertidal zone are subject to predation by Asterias spp. and the carnivorous gastropod Nucella lapillus [Reference 2-12, Section 4.2.4.1].
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| The benthic fauna of the rocky subtidal zone were dominated by amphipods (34 species collected), polychaetes (30 species collected), and molluscs (30 species collected).
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| Species representing other groups such as nemertea, echinoderms, and anemones were collected less frequently.
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| Measures of species richness (total number of species collected) varied considerably from year to year, and appeared to be independent of PNPS operations (capacity factors) [Reference 2-12, Figure 4.2-13]. Total faunal densities also varied widely, due in part to annual fluctuations in numbers of the blue mussel [Reference 2-12, Section 4.2.4.2].
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| The two most common species In sandy subtidal areas were the marine amphipods Acanthohaustorius millsi and Protohaustorius deichmannae 2-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage[Reference 2-12, Section 4.2.4.3].
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| Other species found in the sandy subtidal areas included the common sand shrimp (Crangon septemspinosus), the moon snail (Lunatia heros), and the sand dollar (Echinarachnius parma). No differences were seen between the station near the PNPS discharge canal and the White Horse Beach (control) station, approximately 1.3 miles from PNPS, in terms of species richness (number of species observed), except where there were obvious differences in substrate type [Reference 2-12, Section 4.2.4.3].2.2.3.1 American Lobster The American lobster is common in western Cape Cod Bay and supports a valuable commercial fishery in the PNPS area, primarily between March and November.
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| Because of the commercial importance of this species, a number of special studies have been conducted in the vicinity of PNPS. Studies suggest that a significant percentage of larval lobsters in Cape Cod Bay in June may have come through the Cape Cod Canal, having been spawned in the eastern end of the canal or even points south (Buzzard's Bay, south of Cape Cod). A study of sublegal, sexually immature lobsters captured and released in the vicinity of PNPS indicated that movement of sub-adults was limited: 71 % were recaptured on the rocky ledges where they had been captured previously
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| [Reference 2-12, Section 4.2.4.3].
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| An evaluation of lobster harvest in the PNPS area, reference areas, and the Gulf of Maine showed that catch rates in the PNPS area (and reference areas) tracked those in the Gulf of Maine and appeared to be unaffected by PNPS operations
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| [Reference 2-12, Section 4.2.4.3].2.2.4 Fish Community The species composition of finfish in western Cape Cod Bay reflects a transition between the Gulf of Maine and the Mid-Atlantic Bight [Reference 2-12, Section 4.2.5]. Cape Cod serves as the southern-most boundary for several northern Atlantic fish species and the northern-most boundary for several fish species that inhabit the warmer waters south of Cape Cod, an overlap that results in high species richness and diversity.
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| Fish move freely through the Cape Cod Canal, a 17.5-mile long man-made waterway that connects Cape Cod Bay (on the north) and Buzzards Bay (on the south).Marine finfish were monitored in the vicinity of PNPS from 1970 to 1994 to assess possible effects of station operations on local populations.
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| Bottom trawling gear was used to collect bottom-dwelling fish species inhabiting inshore bottom waters. Gill nets were used to collect pelagic species inhabiting open waters (higher in the water column). Haul seines were used to collect inshore species in relatively shallow waters.2.2.4.1 Bottom Trawl Samplina Bottom trawling was carried out at stations at the entrance to Plymouth Bay (west of PNPS) and within a 2-mile radius of the station. A total of 50 species were collected over a 13-year (1970-1982) period [Reference 2-12, Table 4.2-7]. Six species accounted for 92% of all fish collected.
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| In order of abundance, these species were winter flounder (Pseudopleuronectes americanus; 44.2% of total catch), yellowtail flounder (Pleuronectes ferrugineus; 13.2%), skates (Raja spp.;10.3%), ocean pout (Macrozoarces americanus; 9.1%), longhorn sculpin (Myoxocephalus 24 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage octodecemspinosus; 8.9%), and windowpane flounder (Scopthalmus aquosus; 6.4%). Winter flounder ranked first in abundance in each of the 13 years, with the other species' rankings changing over time. Relative abundance of ocean pout decreased over the course of the study, while relative abundance of skates increased
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| [Reference 2-12, Section 4.2.5.21.Trawling continued through 1993, but the analysis focused on 3 common species: winter flounder, little skate (Raja erinacea), and windowpane flounder.
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| These 3 species comprised between 75 and 91% of the total bottom trawl catch between ~1989 and 1993 [Reference 2-12, Section 4.2.5.2].
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| Winter flounder numbers decreased steadily from 1983 to 1991, then rebounded in 1992 and 1993 [Reference 2-12, Figure 4.2-25]. Little skate and windowpane flounder showed declines over the same period, but the declines occurred later (1987-1988) and were more precipitous
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| [Reference 2-12, Figures 4.2-27 and 4.2-28]. Like the winter flounder, windowpane and little skate showed an increase from 1991 to 1992 and 1993.2.2.4.2 Gill Net Sampling Pelagic fish were collected from 1971-1992 at a site just north of the station, partially within the thermal plume. Abundance of these pelagic species (indicated by pooled catch-per-unit-effort, or CPUE) was highest in 1977, declined from 1977 to 1985, increased from 1985 to 1988, then declined from 1988 to 1992 (1992 had the lowest CPUE of the study) [Reference 2-12, Figure 4.2-31].Pollock (Pollachius virens) dominated gill-net collections over the 22-year study period, and comprised 40% of the total gill net catch in 1992 [Reference 2-12, Section 4.2.5.21.
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| Pollock abundance declined from 1977-1981 (CPUE of 85 to 145 fish per gill net set) to 1990-1992 (CPUE of 15 to 45 fish per gill net set) [Reference 2-12, Figure 4.2-32]. Striped bass (Morone saxatilis), unlike other pelagic species, increased in abundance from the late 1970s to the early 1990s, apparently responding to restrictions on commercial and recreational fishing and other initiatives intended to restore this species along the Atlantic Coast. Atlantic herring (Clupea harengus) abundance increased from the late 1970s until the mid-1 980s, fluctuated through the late 1980s, increased greatly in 1990, then plunged to low levels in 1991 and 1992. Population trends of pelagic fishes in the vicinity of PNPS appeared tied to population trends in the Gulf of Maine and the western North Atlantic and are unaffected by station operations.
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| 2.2.4.3 Haul-Seine Sampling Haul seines were used to collect fish from shallow inshore habitats in the area of PNPS from 1981 to 1991. Three stations were west of PNPS in Plymouth Harbor (Gray's Beach, Long Point, and Warren's Cove), two stations were east of PNPS (White Horse Beach and Manomet Beach), and one station was near the PNPS intake. These haul-seine samples yielded 185,000 fish representing 46 species, with the Atlantic silverside (Menidia menidia) dominating collections (67% of the 11 -year total) [Reference 2-12, Section 4.2.5.2].
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| The greatest number of species was observed at the intake station, followed by Long Point, Warren's Cove, and Manomet Beach.Numbers of fish collected tended to fluctuate dramatically from year to year, probably due to the schooling nature of several common species. Although statistical variances were large, some trends were apparent.
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| For example, catch rates of the most abundant shallow-water species, the 2-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Atlantic silverside, showed no statistically significant downward trend in the intake area over the 1981-1991 period. There was no discernible trend in winter flounder catch rates in the vicinity of PNPS during the 11-year study period.2.2.4.4 Recreational Creel Surveys Recreational creel surveys were conducted (1973 to 1975, 1983, and 1985) to determine the extent of the shore-based recreational fishery in the area of PNPS. Cunner (Tautogolabrus adspersus; 45.7%), bluefish (Pomatomus saltatrix; 29.7%), pollock (9.3%), striped bass (6.0%), and winter flounder (4.8%) were the species caught most often by surf fishermen
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| [Reference 2-12, Section 4.2.5.2].
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| Between 1990 and 1998 bluefish and striped bass were the species most often caught by shore anglers in the area of PNPS. Creel data are an indirect measure of abundance and depend on angler effort, the state of the local economy, and even changing trends in "desirable' species. Nevertheless, these creel data provide additional evidence of a recovering striped bass fishery in the Cape Cod area.2.2.4.5 Atlantic Menhaden In the early years of PNPS operation, substantial numbers of Atlantic menhaden (Brevoortia tyrannus) died in the vicinity of the PNPS discharge canal from gas bubble disease. Gas bubble disease occurs when the dissolved gases in a fish's tissues and blood come out of solution and form bubbles, interfering with normal blood flow and respiration.
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| This is normally caused by a change in temperature or pressure, or by supersaturated conditions that sometimes occur in the heated discharge areas of power plants. In 1973, a total of 43,000 Atlanta menhaden succumbed to gas bubble disease in the area of the PNPS discharge canal. Another 5,000 menhaden were lost in 1976 [Reference 2-12, Section 4.2.6.1, and Reference 2-32, page 4-22].Following the 1976 fish kill, a barrier net was placed across the mouth of the discharge canal from April 1 to November 1 to prevent fish from moving into the canal. Because no outbreaks of gas bubble disease and no significant fish kills were observed in the discharge canal from 1976 through the early 1990s, Boston Edison sought approval from EPA to discontinue deployment of the barrier net. Boston Edison received approval from EPA in November 1994 to discontinue regular use of the barrier net in the discharge canal, provided the net is kept nearby in serviceable condition should a recurrence require its use in the future.2.2.4.6 Winter Flounder The local population of winter flounder is of special concern because it provides an important commercial and recreational fishery and because the area around PNPS serves as spawning, nursery, and feeding grounds for the species. As noted previously, this species dominated bottom trawl collections from 1970-1982 in the vicinity of PNPS. Since 1993, trawl surveys and mark-and-recapture studies have been carried out to determine distribution, abundance, and movement pattems of the local winter flounder population
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| [Reference 2-12, Section 4.2.5.2]., These trawl surveys indicated that annual mean CPUE increased until 1996, peaked in 1997, and declined in 1998 and 1999 [Reference 2-12, Table 4.2-9]. Measures of adult abundance also peaked in 1996 and 1997 and declined in 1998 and 1999.2-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Spring 2000 surveys yielded higher CPUEs and markedly higher measures of abundance[Reference 2-13, Section 3.1]. Unadjusted estimates of winterflounder abundance in the study area were 232,087 adults and 422,572 total winter flounder; adjusted numbers (assuming a trawl efficiency of 50%) were 464,172 and 826,548 respectively
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| [Reference 2-13, Section 3.1]. Winter flounder absolute abundance estimates for adults and total flounder (adults and sub-adults) were 1.8 and 1.5 times their respective 1995-1999 means, suggesting that abundance was substantially higher in 2000 than in the previous 5 years. This increase In abundance of sub-adults and adults was consistent with the apparent high abundance of larval winter flounder in 1997 and 1998 [Reference 2-13, Section 3.1].2.2.5 Summary The aquatic communities of western Cape Cod Bay have been monitored by Boston Edison and Entergy since 1969 to assess potential impacts of PNPS operations.
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| These monitoring studies suggest that PNPS operations have not had a significant effect on local and regional populations of fish and shellfish.
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| Trends in abundance of groundfish, pelagic fish, and shellfish (lobsters in particular) in western Cape Cod Bay mirror population trends in the larger Gulf of Maine and the western North Atlantic and do not appear to be influenced by PNPS operations.
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| 2.3 Groundwater Resources PNPS is located on the shore of Cape Cod Bay within the Northeast Uplands Physiographic
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| ; Province of the Appalachian Mountains.
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| The rocks and sediment in the region range in age from Precambrian to Recent. Pleistocene Glacial till and outwash of variable thickness generally mantles bedrock in the area. Bedrock at the site is approximately 65 feet below ground surface.Groundwater in the area generally occurs in the glacial soils [Reference 2-37]. Most of the residences in the area receive their water from the Town of Plymouth, as does PNPS. The source of Plymouth's water is 11 groundwater wells [Reference 2-41]. Groundwater use is limited to a few locations because the Town of Plymouth supplies most of the residences in the area. There is no current or proposed major groundwater use in the vicinity of the site. The groundwater at the site generally follows the site surface topography.
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| As a result, moderately steep groundwater gradients are present with flow toward Cape Cod Bay [Reference 2-37, Section 1.6].2.4 Critical and Important Terrestrial Habitats The 140-acre PNPS site that contains the major generating facilities, office buildings, warehouses, parking lots, and switchyard, is industrial in character, and provides some limited wildlife habitat (lawns, shrubs, and flowerbeds around buildings) for species that tolerate high levels of human activity.
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| Wooded areas immediately north, south, and west of the developed portion of the site offer higher-quality wildlife habitat, but the value of these areas is diminished by proximity to PNPS and to Rocky Hill Road. Cape Cod Bay lies to the east of the site.In addition, Entergy owns approximately 1,500 acres south and west of Rocky Hill Road. These Entergy-owned lands are managed in a forest trust and are not considered part of the PNPS site proper. This acreage has been designated "Forest Land" under Chapter 61/Chapter 61A of the 2-7 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage General Laws of the Commonwealth, meaning that the State Forester has certified that the land is being managed under an approved Forest Management Plan to "...improve the quality and quantity of a continuous forest crop" (from Certificate for Chapter 61/Chapter 61A Forest Lands, dated September 16, 2002, and signed by the State Forester).
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| The Forest Management Plan [Reference 2-14J prepared for the Massachusetts Department of Environmental Management provides a history of forest management on the property, descriptions of each timber stand (dominant species, age/size of trees, soils, topography), and future plans for each stand (i.e., planting, fertilizing, weeding, thinning, or harvesting).
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| This forestland, which is dominated by second-growth mixed hardwoods (mostly oaks) and pines (mostly white pine and pitch pine), also contains some small wetland areas and abandoned fields in varying stages of succession.
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| These natural areas provide habitat for a variety of wildlife including amphibians (e.g., spotted salamander, redback salamander), reptiles (e.g., Eastern box turtle, Eastern painted turtle), small mammals (e.g., white-footed mouse, gray squirrel, Eastern cottontail rabbit), white-tailed deer, upland game birds (e.g., ruffed grouse, turkey), songbirds (e.g., warblers, sparrows, flycatchers), and birds of prey (e.g., red-tailed hawk, great homed owl)[Reference 2-1; Reference 2-7; Reference 2-11; Reference 2-35].To determine if sensitive or ecologically-significant habitats were present in the vicinity of the PNPS site, Entergy reviewed Massachusetts Geographic Information System (GIS) data layers for "priority habitat" (known habitats of state-protected plants and animals), "estimated habitat" (known habitats of state-protected wildlife occurring in wetland areas), and certified vernal pools (vernal pools are afforded protection under the Massachusetts Wetlands Protection Act when they satisfy specific criteria with regard to hydrology and indicator species).
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| These data layers are derived from databases maintained by the Massachusetts Division of Fisheries and Wildlife's (MDFW) Natural Heritage & Endangered Species Program (NHESP). Entergy also reviewed lists of threatened and endangered species known to occur in Massachusetts to determine if critical habitat had been identified in the PNPS vicinity for any of these species.Based on this investigation and correspondence with the MDFW, there is one site of both priority and estimated habitat for the spotted turtle (Clemmys guttata), which is a state species of special concern, on the 140-acre PNPS site. Two NHESP priority sites of rare species habitats lie within several hundred yards of the PNPS-to-Snake Hill Road transmission corridor.
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| NHESP prefers not to reveal the sensitive species found or potentially found in these areas. The PNPS-to-Snake Hill Road corridor does not actually cross these significant areas, nor does it encroach or impinge upon them in any way. The closest certified vernal pool is approximately one mile away from the transmission corridor.A 0.5-mile-long segment of the PNPS-to-Snake Hill Road transmission corridor passes through an area designated critical habitat (at 50 CFR 17.95) for the northern red-bellied cooter (Pseudemys rubriventris).
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| Critical habitat is defined and used in the Endangered Species Act to describe specific geographic areas essential to the conservation of a threatened or endangered species that may require special management and protection.
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| Federal agencies are required to consult with the U.S. Fish and Wildlife Service (FWS) on activities they carry out, fund, or authorize to ensure that these activities will not destroy or adversely modify critical habitats.
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| As 2-8 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage noted elsewhere in this document, Entergy does not own, operate, or maintain the PNPS-to-Snake Hill Road transmission corridor.Section 3.2.7 describes the transmission lines that Boston Edison built to connect PNPS to the transmission system. Two 345-kilovolt (kV) transmission lines leave the PNPS switchyard, but these transmission lines merge and share a single, 300-foot-wide corridor from the PNPS site to the Snake Hill Road substation.;
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| These transmission lines are owned and maintained by NSTAR, which transmits and delivers electricity to homes and businesses in eastern Massachusetts.
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| NSTAR normally controls woody vegetation In transmission corridors in accessible upland areas by mowing. NSTAR's corridor vegetation maintenance program is an integrated one that uses a combination of mechanical, chemical, and biological control methods. This methodology creates stable communities of native plants that are not capable of growing into electric conductors, provides excellent habitat for wildlife, and supports biodiversity.
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| NSTAR's vegetation program complies with all state and federal regulations.
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| Prior to carrying out vegetation management in rights-of-way, NSTAR environmental personnel review work plans with maintenance crews and consult with local town conservation committees when necessary to ensure that wetland areas and sensitive plant communities are protected.
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| NSTAR also schedules vegetation management practices in consideration of species life cycles in the areas to be maintained.
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| No additional areas designated by FWS as critical habitat for listed species occur at PNPS or occur within or adjacent to associated transmission lines. In addition, the transmission corridors do not cross any state or federal parks, wildlife refuges, or wildlife management areas.2.5 Threatened or Endangered Snecies More than 80 state- and federally-listed species could occur in Plymouth County, a relatively large county that encompasses a variety of habitats ranging from upland forests to farmlands to bogs to marshlands
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| [Reference 2-27; Reference 2-28] (Table 2-1). Another 10 marine species listed by the FWS and National Marine Fisheries Service (NMFS) could occur in Cape Cod Bay[Reference 2-15; Reference 2-28]: 5 species of whale (sei, right, blue, finback, and humpback)and 5 species of sea turtle (loggerhead, leatherback, hawksbill, green, and Kemp's ridley).No state- or federally-listed endangered or threatened species is known or believed to occur on the PNPS site. The PNPS-to-Snake Hill Road transmission corridor crosses habitat designated critical for the endangered northern red-bellied cooter (see Section 2.4 for a discussion of this critical habitat), but the part of the critical habitat crossed by the transmission corridor appears to be a buffer area for the population rather than high-quality turtle habitat. Northern red-bellied cooters have never been observed by Boston Edison, Entergy, or NSTAR biologists in this transmission corridor.
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| No other state- or federally-listed endangered or threatened species is known or believed to occur in this transmission corridor.
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| A state-listed species of special concern, the spotted turtle, does have a priority habitat area on the PNPS site property.
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| Spotted turtles have not been observed by Entergy personnel or contractors on the PNPS site.Several listed species are known to occur in the general vicinity of the PNPS site, however, and cannot be ruled out as occasional visitors to the PNPS site and environs.
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| These include the bald eagle, piping plover, and roseate tern. Bald eagles are present year-round in Massachusetts and 2-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ij congregate in significant numbers in wintering areas along the coast of Cape Cod and Buzzards Bay [Reference 2-28]. PNPS environmental personnel have never observed bald eagles foraging in the vicinity of the PNPS site. In March 2005, juvenile and adult bald eagles were observed at Plimoth Plantation in Plymouth, Massachusetts, which is approximately four miles from PNPS. Piping plovers nest in summer on sandy coastal beaches along the Massachusetts coast, preferring the dry, light-colored sand found along the outer shores [Reference 2-28].Although piping plover nesting has not been documented on the PNPS site, individual birds almost certainly move through the PNPS area when migrating to breeding areas farther north of Plymouth Bay and returning to wintering areas along the south Atlantic and Gulf coasts. Like the piping plover, the roseate tern nests in colonies along the Massachusetts coast in summer[Reference 2-28]. The roseate tern nests in areas with thick vegetative cover, always in association with the common tern. Although suitable nesting habitat has not been identified at PNPS, migrating terns may move through the site in late spring (en route to nesting areas in Maine and Nova Scotia) and late summer (en route to wintering areas in the West Indies and Latin America).Six great whale species migrate along the coast of Massachusetts, with concentrations occurring in spring in the plankton-rich and fish-filled waters of Stellwagen Bank, an 800-square-mile area of shallow water just off the tip of Cape Cod. The whale species seen most frequently off the coast of Massachusetts are minke, finback, and humpback whales. The minke whale is the most abundant of the baleen whales and is not a listed or candidate species at present. The finback and humpback are listed as federally endangered.
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| The northern right whale, rarest of the great whales, is occasionally observed in Cape Cod Bay in spring and summer months. The western North Atlantic population is believed to number between 290 and 350 individuals
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| [Reference 2-8;Reference 2-30]. Critical habitat has been designated for the endangered northern right whale in Cape Cod Bay (50 CFR 226). No whales have been observed in the shallow waters off PNPS (or in the intake and discharge canal areas) by Boston Edison or Entergy biologists since biological monitoring began in the late 1960s.Five species of sea turtle occur along the Massachusetts coast, but sightings are uncommon and limited for the most part to sub-adult "wanderers" [Reference 2-39]. Young sea turtles often migrate" north (float with Gulf stream currents) and feed in Cape Cod Bay during the warm summer months. When water temperatures drop suddenly in late fall/early winter, turtles still in Cape Cod waters are sometime cold-stunned and washed ashore on area beaches. In most years, fewer than 20 sea turtles are stranded, but in the winter of 1999-2000, a total of 277 sea turtles were found on Cape Cod beaches. Slightly more than half (144) of the turtles were transported alive to Boston's New England Aquarium for treatment and subsequently relocated to Florida. In 2003, 89 sea turtles were found stranded on Cape Cod beaches [Reference 2-24].Forty-four of these turtles survived [Reference 2-241. In the twenty-five years that records have been kept documenting the numbers of cold-stunned sea turtle strandings in Massachusetts, only one sea turtle has stranded in Plymouth.
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| In November 2003, a small (approximately 50 pounds) loggerhead sea turtle stranded on Priscilla Beach, which is approximately 0.63 miles from PNPS [Reference 2-40]. However, no sea turtles have ever been observed in the intake or discharge canals or along the PNPS waterfront.
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| 2-10 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-1 Endangered and Threatened Species that Occur In the Vicinity of PNPS or in Plymouth County, MA 1Fede1a Statue Scientific Name Common Name State Status1 Status1 Mammals Balaenoptera borealis Sei whale E E Balaena glacialis Right whale E E Balaenoptera musculus Blue whale E E Balaenoptera physalus Finback whale E E Megaptera novaeangliae Humpback whale E E Birds Ammodramus savannarum Grasshopper sparrow T Bartramia longicauda Upland sandpiper E Botaurus lentiginosus American bittern E Charadrius melodus2 Piping plover T T Circus cyaneus Northem harrier T Haliaeetus leucocephalus Bald eagle T E lxobrychus exilis 2 Least bittern E Parula americana Northem parula T Podilymbus podiceps Pied-billed grebe E Rallus elegans King rail -.T Sterna dougallii dougalli 2 Roseate tem E E Reptiles Caretta caretta Loggerhead sea turtle T T Chelonia mydas Green sea turtle T T Dermochelys coracea Leatherback sea turtle E E Emydoidea blandingfi Blanding's turtle -T Eretmochelys imbricata Hawksbill sea turtle E E Lepidochelys kempfi Kemp's Ridley sea turtle E E Malaclemys terrapin Diamondback terrapin T Pseudemys rubriventris bangsiP Northern red-bellied cooter E E 2-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-1 Endangered and Threatened Species that Occur in the Vicinity of PNPS or in Plymouth County, MA (Continued)
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| Scientific Name Common Name Federal State.__Status 1 Status'Amphibians Ambystoma opacum Marbled salamander T Scaphiopus holbrookii Eastern spadefoot toad T Invertebrates Acronicta albarufa Barrens daggermoth T Alasmidonta heterodon Dwarf wedgemussel E E Cicinnus melsheimen Melsheimer's sack bearer T Cycnia inopinatus Unexpected cycnia T Enallagma recurvatum 2 Pine barrens bluet T Erynnis persius persius2 Persius duskywing E Hypomecis buchholzaria Buchholz's gray .E Lampsilis cariosa Yellow lampmussel E Metarranthis apiciana Barrens metarranthis moth E Nicrophorus americanus American burying beetle E Papaipema appassionata Pitcher plant borer moth T Papaipema stenocelis Chain fern borer moth T Papaipema sulphurata2 Water-willow stem borer T Somatochlora kennedyi Kennedy's emerald E Zanclognatha martha Pine barrens zanclognatha T Vascular Plants Agalinis acuta Sandplain gerardia E Aristida purpurascens Purple needlegrass
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| -T Asclepias verticillata Linear-leaved milkweed -T Bidens hyperborea var. hyperborea Estuary beggarticks
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| -E Calamagrostis pickeringfi Reed bentgrass
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| -E Cardamine longii Long's bittercress
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| .E Carex polymorpha Variable sedge E C4. i 2-12 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-1 Endangered and Threatened Species that Occur in the Vicinity of PNPS or in Plymouth County, MA (Continued)
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| Scientific Name Common Name Federal State Status' Status 1 Carex striata var. brevis Walter's sedge -E Crassula aquatica Pygmyweed
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| -T Cyperus houghtonil Houghton's flatsedge
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| -E Dichanthelium mattamuskeetense Mattamuskeet panic-grass
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| -E Elatine americana American waterwort
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| -E Eriocaulon parkeri Estuary pipewort E Eupatorium aromaticum Lesser snakeroot
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| .E Eupatorium leucolepis var. novae- New England boneset E angliae 2 Isoetes acadiensis Acadian quillwort E Isotria medeoloides Small whorled pogonia T Linum medium var. texanum Rigid flax -T Lipocarpha micrantha Dwarf bulrush -T Ludwigia sphaerocarpa Round-fruited false-loosestrife
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| -E Lycopus rubellus Gypsywort
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| -E Mertensia maritima Oysterleaf
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| -E Ophioglossum pusillum Northern adder's-tongue
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| -T Panicum rigiduium var. Pubescens Long-leaved panic-grass
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| -T Platanthera flava var. herbiola Pale green orchid -T Polygonum setaceum var. interiectum Strigose knotweed -T Prenanthes serpentaria Lion's foot -E Ranunculus micranthus Tiny-flowered buttercup
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| -E Ranunculus pensylvanicus Bristly buttercup
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| -T Rhynchospora inundata2 Inundated homed-sedge
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| -T Rhynchospora nitens 2 Short-beaked bald-sedge
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| -T Rhynchospora torreyana 2 Torrey's beak-sedge
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| -E Rumex pallidus Seabeach dock -T 2-13 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-1 Endangered and Threatened Species that Occur in the Vicinity of PNPS or in Plymouth County, MA (Continued)
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| Federal State Scientific Name Common Name Status 1 Status 1 Sabatia campanulata Slender marsh pink -E Sagittaria subulata var. subulata River arrowhead
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| -E Sanicula canadensis Canadian sanicle -T Scirpus longii Long's bulrush -T Senna hebecarpa Wild senna- E Spartina cynosuroides Salt reedgrass -T Sphenopholis pensylvanica Swamp oats -T Symphyotrichum concolor Eastern silvery aster -E Triosteum perfoliatum Broad tinker's weed -E Viola brittoniana Britton's violet T 1. E = Endangered; T = Threatened;-
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| = Not listed.2. Species reported by the Massachusetts NHESP as occurring within six miles of PNPS.Source: References 2-15, 2-27, 2-28 and 2-51.(I , 2-14 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.6 Regional Demographv 2.6.1 Regional Population The Generic Environmental Impact Statement for License Renewal of Nuclear Plants presents a population characterization method that is based on two factors: "sparseness" and "proximity"[Reference 2-32, Section C.1.4]. "Sparseness." measures population density and city size within 20 miles of a site and categorizes the demographic information as follows.Demographic Categories Based on Sparseness Category Most sparse 1. Less than 40 persons per square mile and no community with 25,000 or more persons within 20 miles 2. 40 to 60 persons per square mile and no community with 25,000 or more persons within 20 miles 3. 60 to 120 persons per square mile or less than 60 persons per square mile with at least one community with 25,000 or more persons within 20 miles Least sparse 4. Greater than or equal to 120 persons per square mile within 20 miles Source: Reference 2-32"Proximity" measures population density and city size within 50 miles and categorizes the demographic information as follows.Demographic Categories Based on Proximity Category Not in close proximity
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| : 1. No city with 100,000 or more persons and less than 50 persons per square mile within 50 miles 2. No city with 100,000 or more persons and between 50 and 190 persons per square mile within 50 miles 3. One or more cities with 100,000 or more persons and less than 190 persons per square mile within 50 miles In close proximity
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| : 4. Greater than or equal to 190 persons per square mile within 50 miles Source: Reference 2-32 2-15 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The GEIS then uses the following matrix to rank the population in the vicinity of the plant as low, medium, or high.GEIS Sparseness and Proximity Matrix Proximity 1 2 3 4 o 1 1.1 i.2 1.3 1.4 c02 -2.1 2.2 2.3 2.4 3 3 3.1 3.2 3.3 4 4.1 4.2 Low Medium High Population Population Population Area Area Area Source: Reference 2-32 Entergy used 2000 census data from the U.S. Census Bureau (USCB) website [Reference 2-43]and GIS software (ArcView@)
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| to determine demographic characteristics in the PNPS vicinity.As derived from USCB information, approximately 285,547 people live within 20 miles of PNPS.Massachusetts has a population density of 422 persons per square mile within 20 miles of PNPS and, applying the GEIS sparseness index, falls into the least sparse category, Category 4 (having greater than or equal to 120 persons per square mile within 20 miles). This calculation and the one for the population within 50 miles corrects for the area within the radius that is water.As estimated from USCB information, approximately 4,629,116 people live within 50 miles of PNPS. This equates to a population density of 1,167 persons per square mile within 50 miles.Applying the GEIS proximity index, PNPS is classified as Category 4 proximity (having greater than or equal to 190 persons per square mile within 50 miles). According to the GEIS sparseness and proximity matrix, the PNPS ranks of sparseness Category 4 and proximity Category 4 result in the conclusion that PNPS is located in a "high" population area.All or parts of 15 counties (Figure 2-1) and the cities of Boston, Massachusetts, and Providence, Rhode Island, are located within 50 miles of PNPS.L-*v Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Plymouth and Barnstable Counties are largely residential and have a combined total population of approximately 700,000 [References 2-46 and 2-47]. Plymouth County extends to metropolitan Boston and is primarily made of small towns, such as the coastal towns along Cape Cod Bay.Barnstable County is made up of 15 small towns and is bordered by Cape Cod Bay, the Atlantic Ocean, Nantucket Sound, and Plymouth County. From 1970 to 2000, Plymouth County had an average annual growth rate of 1.4% and Barnstable County had an average annual growth rate of 4.3%. Both Plymouth and Barnstable Counties have been growing at a rate faster than that of Massachusetts as a whole. From 1970 to 2000, Massachusetts's average annual population growth rate was 0.39% [adapted from Reference 2-43].Table 2-2 shows estimated populations and annual growth rates through 2040 for the two counties with the greatest potential to be socioeconomically affected by license renewal activities.
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| The license renewal term is through 2032.Table 2-2 Estimated Populations and Annual Growth Rates in Plymouth and Barnstable Counties 1980-2040 Plymouth County Barnstable County Percent Annual Percent Annual Year Population Growth Population Growth 1980 405,4371 147,9251 1990 435,2761 0.7 186,6051 2.6 2000 472,8222 0.9 222,2302 1.9 2010 496,0533 0.5 257,8443 1.6 2020 517,6443 0.4 299,0354 1.6 2030 551,0054 0.6 334,7664 1.2 2040 579,5294 0.5 368,7204 1.0 1. Reference 2-42 2. References 2-46 and 2-47 3. Reference 2-29 4. Reference 2-38 2-17 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.6.2 Minority and Low-income Populations 2.6.2.1 Background The NRC performs environmental justice analyses utilizing a 50-mile radius around the plant as the environmental impact site and the state as the geographic area for comparative analysis.Entergy has adopted this approach for identifying the minority and low-income populations that could be affected by PNPS operations.
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| Entergy used ArcView geographic information system software to combine U.S. Census Bureau (USCB) TIGER line data with USCB 2000 census data to determine minority characteristics on a block-group level and low-income characteristics on a census tract. Entergy included all census tracts/block groups if any of their area lay within 50 miles of PNPS. The 50-mile radius includes 3,845 block groups and 1,034 census tracts. Entergy defines the geographic area for PNPS as a two-state area, with the largest portion of that area located in Massachusetts and a smaller portion in Rhode Island.2.6.2.2 Minority Populations The NRC procedural guidance for performing environmental assessments and considering environmental issues defines a "minority" population as American Indian or Alaskan Native;Asian; Native Hawaiian or Pacific Islander; Black races; other; multi-racial; the aggregate of all minority races; or Hispanic ethnicity
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| [Reference 2-33]. The guidance indicates that a minority population exists if either of the two following conditions exists: Exceeds 50 Percent -the minority population of the environmental impact site exceeds 50 percent, or More than 20 Percentage Points Greater -the minority population percentage of the environmental impact site is significantly greater (typically at least 20 percentage points)than the minority population percentage in the geographic area chosen for comparative analysis.NRC guidance calls for use of the most recent USCB decennial census data. Entergy used 2000 census data [References 2-43 and 2-44] to determine the percentage of the total populations in the two states that belong to each minority group, and to identify minority populations within 50 miles of PNPS.For each minority, Entergy divided USCB minority population numbers for each block group by the total population within that block group to obtain the percent of the block group's population that belonged to the minority.
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| For each of the 3,845 block groups within 50 miles of PNPS, Entergy calculated the percent of the population in each minority category and compared the result to the corresponding geographic area's minority threshold percentages to determine whether minority populations exist.2-18 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Massachusetts had approximately 83% of the block groups with the remaining 17% in Rhode Island. USCB data [Reference 2-43] for Massachusetts characterize 0.2% of the state's population as American Indian or Alaskan Native; 3.8% Asian; 0.0% Native Hawaiian or other Pacific Islander; 5.4% Black races; 3.7% all other single minorities; 2.3% multi-racial; 15.5%aggregate of minority races; and 6.8% Hispanic ethnicity.
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| USCB data [Reference 2-44] for Rhode Island characterizes 0.5% of the state's population as American Indian or Alaskan Native;2.3% Asian; 0.1% Native Hawaiian or other Pacific Islander; 4.5% Black races; 5.0% all other single minorities; 2.7% multi-racial; 15% aggregate of minority races; and 8.7% Hispanic ethnicity.
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| Based on either the "more than 20 percent" or the "exceeds 50 percent" criteria, no multi-racial block groups exist in the geographic area.Based on the "more than 20 percent" criterion, an American Indian or Alaskan Native minority population exists in one block group, in Dukes County, Massachusetts (Table 2-3, Figure 2-4).Based on the "more than 20 percent" criterion, Asian minority populations exist in 57 block groups; 54 in Massachusetts and 3 in Rhode Island (Table 2-3, Figure 2-5).Based on the "more than 20 percent" criterion, a Native Hawaiian or other Pacific Islander minority population exists in one block group in Suffolk County, Massachusetts (Table 2-3, Figure 2-6).Based on the "more than 20 percent" criterion, Black Races minority populations exist in 261 block groups (Table 2-3, Figure 2-7) with 233 of the block groups in Massachusetts and the remaining 28 in Rhode Island.Based on the "more than 20 percent" criterion, All Other Single Minority Races populations exist in 135 block groups (Table 2-3, Figure 2-8). Seventy-seven of the block groups are in Massachusetts and 58 are in Providence County, Rhode Island.Based on the "more than 20 percent" criterion, Aggregate of Minority Races populations exist in 597 block groups (Table 2-3, Figure 2-9) with 477 of the block groups in Massachusetts and 120 in Rhode Island.Based on the "more than 20 percent" criterion, Hispanic Ethnicity minority populations exist in 240 block groups (Table 2-3, Figure 2-10) with 145 of them in Massachusetts and the other 95 in Providence County, Rhode Island.As a general matter, there are relatively few block groups in the geographic areas that constitute minority populations, and these are generally in towns or urban areas more than 20 miles from the site.2-19 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.6.2.3 Low-Income Populations NRC guidance defines "low-income' by using USCB statistical poverty thresholds
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| [Reference 2-33, Appendix D]. The USCB characterizes 9.9% of Massachusetts and 12.4% of Rhode Island households as low-income
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| [Reference 2-45].For each census tract within the 50-mile radius of PNPS (see Section 2.6.2.1 for a discussion of how census tracts were selected and population percentages were calculated), the number of low-income households was divided by the number of total households in that tract to obtain the percent of low-income households for that tract. A low-income population is considered to be present if (1) the low-income population of the census tract or environmental impact site exceeds 50%, or (2) the percentage of households below the poverty level in a census tract is significantly greater (typically at least 20 points) than the low-income population percentage in the geographic area chosen for comparative analysis.Based on the "more than 20 percent" criterion, low-income populations exist in 69 census tracts (Table 2-3, Figure 2-11), 48 in Massachusetts and 21 in Providence County, Rhode Island.As a general matter, there are relatively few low income populations in the geographic areas, and these are generally in towns or urban areas more than 20 miles from the site.2-20 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-3 Minority and Low-Income Population Information Native Census American Hawaiian Aggregate 2000 2000 Indian or or Other All Other Mufti. Of 2000 Tracts Block Alaskan Pacific Black Single Racial Minority Hispanic Census Low County State Groups, Native Asian Islander Races Minorities Minorities Races Ethnicity Tracts Income Barnstable MA 199 0 0 0 0 0 0 0 0 51 0 Bristol MA 417 0 1 0 0 11 0 22 6 117 9 Dukes MA 20 1 0 0 0 0 0 1 0 4 0 Essex MA 311 0 0 0 1 5 0 33 25 81 2 Middlesex MA 753 0 11 0 14 2 0 53 8 194 0 Nantucket MA 5 0 0 0 0 0 0 0 0 3 0 Norfolk MA 473 0 14 0 5 0 0 21 0 121 0 Plymouth MA 366 0 0 0 17 8 0 43 0 92 1 Suffolk MA 631 0 28 1 196 51 0 304 106 177 36 Worchester MA 14 0 0 0 0 0 0 0 0 6 0 Bristol RI 41 0 0 0 0 0 0 0 0 11 0 Kent RI 83 0 0 0 0 0 0 0 0 23 0 Newport RI 60 0 0 0 1 0 0 2 0 22 0 Providence RI 468 0 3 0 27 58 0 118 95 130 21 Washington RI 4 0 0 0 0 0 0 0 0 2 0 2-21 J 3 Ji Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-3 Minority and Low-income Population Information (Continued)
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| Totals 3845 1 57 1 261 135 0 597 240 1034 69 Native tAmericanHawaiian B All Other Multi- Aggregate State Idaor Asian or Other Single Racial oHipnc Low Income AlaskanPaic Races Minortes Mnris Minrty Ethnicity Native ~ IslanderRae State Averages Massachusetts 0.2% 3.8% 0.0% 5.4% 3.7% 2.3% 15.5% 6.8% 9.9%Rhode Island 0.5% 2.3% 0.1% 4.5% / 5.0% 2.7% 15% 8.7% 12.4%Percentage that Identifies a Minority Block on Low-income Tract Massachusetts 20.2% 23.8% 20% 25.4% 27.3% 22.3% 35.5% 26.8% 29.9%Rhode Island 20.5% 22.3% 20.1% 24.5% 25% 22.7% 35% 28.7% 32.4%2-22 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.7 Taxes PNPS pays annual property taxes to the Town of Plymouth.
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| Taxes fund the Town of Plymouth's operations, the school system, public works, the Town General Fund, and the police and fire departments
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| [Reference 2-19].In 1998, the Commonwealth of Massachusetts deregulated its utility industry.
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| As a result, the Massachusetts legislature changed property tax assessment methodologies for utilities from net book value to fair market value. In 1999, Boston Edison Company sold PNPS to Entergy Corporation for roughly an order of magnitude less than the value being carried on the books at that time. Therefore, the property taxes being paid to the Town of Plymouth for PNPS have been reduced from pre-1999 payments.
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| Entergy paid $1.6 million in property taxes for the Town's 1999-2000 fiscal year. For the fiscal year 2004, Entergy's property tax bill was $1.6 million. The Town of Plymouth and Entergy have negotiated payment in lieu of taxes of $1 million annually with the potential for payments to increase should Entergy make capital improvements or substantial additions to the plant. The agreement is through 2012, and would be renegotiated in the event of license renewal. Boston Edison's parent, NSTAR, retained ownership of all transmission functions and facilities and will continue to pay property taxes to the Town of Plymouth for those facilities.
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| Because the transmission facilities are part of the utility industry and also subject to the new property tax assessment methodologies, NSTAR will pay reduced property taxes to the Town of Plymouth.
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| In order to ease deregulation impacts to the Town of Plymouth, the Massachusetts legislature has required NSTAR to make payments to the Town of Plymouth until the end of PNPS' current license in 2012. Those payments are gradually being reduced until they reach $1 million in 2007. From 2007 to 2012, NSTAR will pay the Town of Plymouth $1 million annually.
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| This is a significant reduction from the $15 million in tax revenues previously received by the Town from Boston Edison Company. Table 2-4 lists the tax payments for the years 1997 through 2012.Until 1999, PNPS' property taxes provided approximately 24% of the Town of Plymouth's total property tax revenues.
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| Currently, PNPS pays approximately 2 to 3% of the total property taxes received by the Town of Plymouth.2-23 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-4 Property Taxes Town of Plymouth Boston Edison or Property Tax Paid by Property Tax Property Tax Paid PNPS % of Total Boston Edison or Year Revenues by PNPS Property Taxes NSTAR 19972 $63,082,5791 NA 24 $15,000,000 19982 $64,415,1021 NA 24 $15,187,000 1999 $67,179,6361
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| $800,000 prorated $15,187,000 2000 $71,834,404'
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| $1,600,000 2 $15,187,000 2001 $75,157,4983
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| $2,500,000 3 $15,187,000 2002 $76,393,5223
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| $2,011,445 3 $13,000,000 2003 $78,703,1113
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| $1,617,779 2 $13,000,000 2004 $86,587,2053
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| $1,600,000 2 $13,000,000 2005 $1,400,0000,000 2006 _ $1,000,000
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| $11,000,000 2007 -$1,000,000
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| .$1,000,000 2008 -$1,000,000
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| $1,000,000 2009 -$1,000,000
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| $1,000,000 2010 -$1,000,000
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| $1,000,000 2011 -$1,000,000
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| $1,000,000 2012 -$1,000,000
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| $1,000,000
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| : 1. Reference 2-17 2. Boston Edison owned PNPS until 1999 and paid taxes to the Town of Plymouth on the plant and transmission facilities.
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| : 3. Reference 2-20 NA = Not applicable.
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| 2-24 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.8 Land Use Planning Localities in southeastern Massachusetts have united to develop a regional growth management project called the Southeastern Massachusetis Vision 2020 Project, which has been designed to address the rapid growth and change occurring in the area of Massachusetts between Boston, Cape Cod, and Rhode Island. The project includes 51 cities and towns, Including all communities in Plymouth and Bristol Counties and four communities in Norfolk County. Three regional planning agencies in southeastern Massachusetts are overseeing the project: the Old Colony Planning Council, the Southeastern Regional Planning and Economic Development District, and the Metropolitan Area Planning Council [Reference 2-34, Chapter 1].This section focuses on Plymouth and Barnstable Counties because most of the permanent PNPS workforce live in these counties (see Section 3.5) and Entergy pays property taxes in the Town of Plymouth.
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| The planning commissions for the areas of Plymouth County where most, Pilgrim employees reside are the Old Colony Planning Council, the Metropolitan Area Planning Council, and the Southeastern Regional Planning and Economic Development District.Bamnstable County has its own regional planning organization, the Cape Cod Commission
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| [Reference 2-4, Section 1].Both counties have experienced growth over the last several decades (Table 2-2) and their regional policy plans reflect planning efforts and public involvement in the planning process.Land use planning tools, such as zoning, historic districts, and incentives for redevelopment guide, but do not restrict, future growth and development.
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| All plans share the goals of managing growth and development, protecting public drinking water supplies, reducing traffic congestion, and controlling sprawl. As demonstrated below, the land use plans for the two counties guide development, but do not contain strict growth control measures that limit overall housing development
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| [Reference 2-9].2.8.1 Plymouth County Plymouth County occupies roughly 661 squared miles of land area [Reference 2-46]. Over 59,000 acres of farmland are in Plymouth County and it is ranked third of 14 counties in Massachusetts in agricultural sales [Reference 2-48].2.8.1.1 Existing Land Use Trends As of 1991, 22 to 47% of the land within the Old Colony Planning Council portion of Plymouth County was potentially "developable' (i.e., agricultural, forest,: and open space) [Reference 2-34, Figure 4.7]. The developed land is primarily residential
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| [Reference 2-34, Chapter 41; however, Plymouth County is also home to industry, wholesale and retail businesses, and service-based businesses
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| [Reference 2-36]. The South Shore subregion (Rockland, Norwell, Scituate, Marshfield, Hanover, and Duxbury) is classified as suburban/rural.
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| Because of the limited public sewerage and public transit in the South Shore subregion, the Metropolitan Area Planning Council designates this area as appropriate for very limited new growth [Reference 2-18, page 12].2-25 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ( y The land within the Town of Plymouth, where PNPS is located, and where roughly 30% of the employees reside, was classified in 1999 as follows: 15.8% residential, 0.9% commercial, 3.0%industrial, 4.2% agriculture, 3.44% urban open land, 6.4% water, 3.1% open land, and 63.3%natural land/ undisturbed vegetation
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| [Reference 2-211. The Town of Plymouth has zoning districts for a range of residential, commercial, and industrial development, and regulations that guide that development
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| [Reference 2-34, Appendix].
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| 2.8.1.2 Future Land Use Trends The Old Colony Planning Council guides much of the land development in Plymouth County.The Council is charged with designating priority development areas that have combinations of land, infrastructure, services, accessibility, and amenities suited to accommodate a significant portion of the region's anticipated growth. Growth will be encouraged within the boundaries of the priority development areas. The region's desired pattern for new growth is the compact, mixed-use community center. Communities will allocate land for future residential development with guidance from the Council. This future residential development will occur in areas which are designated for growth, are compatible with adjoining uses, and where there will be no significant adverse or unmitigated impacts to environmental resources.
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| Build-out and site-suitability analyses will be conducted throughout the region to assist in identifying areas for future development
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| [Reference 2-34, Chapter 31.The Town of Plymouth conducted a build-out study considering local zoning requirements, geographic limitations, transportation, and water supply constraints in 1999. The study identified l 29,000 acres that would be appropriate for residential development and approximately 375 acres for commercial and industrial development.
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| 2.8.2 Barnstable County Bamstable County encompasses approximately 396 square miles [Reference 2-47]. According to the Bamstable Regional Policy Plan, Barnstable County is treasured for the distinctive historic and small town character of its communities and its open landscapes
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| [Reference 2-6, Section 11.6.2.8.2.1 Existing Land Use Trends Every Bamstable County community is struggling to manage growth, preserve historic resources, and maintain town character, often without adequate growth controls and zoning standards.
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| In 1990, land use classifications in Barnstable County were as follows: 30% residential, 0.8% crop land and pasture, 47% forest, 8.3% open land, 1.9% commercial, 0.50% industrial, and 4.6%water [adapted from Reference 2-5]. Recent land development in Barnstable County has been primarily residential.
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| In 1996, developed land represented more than 33% of Bamstable County's total land area [Reference 2-6, Section 11.11.2.8.2.2 Future Land Use Trends Barnstable County, through its regional planning organization, the Cape Cod Commission, has developed land use and growth policies.
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| The Cape Cod Commission's goal for future land use 2-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage and growth has been "to encourage growth and development consistent with the carrying capacity of Cape Cod's natural environment in order to maintain the Cape's economic health and quality of life, and to encourage the preservation and creation of village centers and downtown areas that provide a pleasant environment for living, working, and shopping for residents and visitors" [Reference 2-4]. To achieve this goal, Barnstable has the following requirements
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| [Reference 2-4]:* Compact forms of development such as cluster development, redevelopment within certified growth/activity centers, and, v/here appropriate, mixed-use residential/
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| commercial development shall be encouraged in order to minimize further land consumption and protect open space.* All residential subdivisions of five or more lots shall submit a cluster development preliminary plan for consideration by towns or the Commission as appropriate during the development review process.* Extension or creation of new roadside "strip" commercial development outside of certified growth/activity centers shall be prohibited.
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| * Development and redevelopment shall be directed away from Significant Natural Resource Areas as illustrated on the Cape Cod Significant Natural Resource Area Map dated September 5, 1996, as amended.2.9 Social Services and Public Facilities 2.9.1 Public Water Supply Because PNPS is located in Plymouth County and most of the PNPS employees reside in Plymouth or Barnstable Counties, the discussion of public water supply systems will focus on towns within these counties (Table 2-5 and Table 2-6). County-level data is not available.
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| 2.9.1.1 Plymouth County Groundwater is the primary source of potable water for the communities in Plymouth County.However, the Scituate and Abington-Rockland drinking water systems are supplied from both groundwater and surface water. The Brockton water system is supplied by surface water only.The various water systems buy from or sell to other nearby water systems, depending on demand. System water demand for the communities in Plymouth County which make up a large percentage of the PNPS employment population in 2003 ranged from a low of 0.26 million gallons per day (MGD) to a high of 4.61 MGD. Average daily consumption among these towns is approximately 1.73 MGD (calculated from data provided by Reference 2-25). There are several towns where a number of PNPS employees reside which do not have municipal water, but rather individual private wells.Table 2-5 compares average daily use and authorized withdrawal volumes (capacities) for selected Plymouth County water systems.2-27 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-5 Selected Plymouth County Public Water Suppliers and Capacities for the Year 2003 Average Consupto Authorized Withdrawal
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| ..Consumption
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| )(MGD)1 Volume (Capacity MGD)2 Duxbury Water Department 1.35 1.85 Halifax Water Department 0.49 0.68 Kingston Water Department 1.39 1.56 Marshfield Water Department 2.90 3.3 Middleborough Water Department 1.53 3.03 broke WaterDivision 13 1i26 Plymouth Water Division 4.61 6.36 PlymouthWaterC.fF--i';
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| 0 26 022i 1. Reference 2-25 2. Reference 2-26 Because no county-level data were available, Entergy evaluated the water systems in the Plymouth and Bamstable Counties towns where approximately 70% of the Pilgrim workforce reside. The remaining 30% of the workforce was scattered among numerous towns and few employees lived in any single town.Shading indicates communities where consumption exceeds capacity and shortfalls are made up by purchase.2.9.1.2 Bamstable County A network of 145 groundwater wells supported by the Cape Cod Aquifer supplies Barnstable County's potable water. A 1994 U.S. Geological Survey study indicated that approximately 5.6%of Bamstable County's land area would be suitable for new well sites [Reference 2-4, Section 2.1]. The average daily water demand for 2003 for the water systems serving the areas of Bamstable County where the majority of PNPS employees reside is 1.15 MGD. The water demand ranged from a low of 0.10 MGD to a high of 2.74 MGD (calculated from data provided by Reference 2-25].Table 2-6 compares average daily use and authorized withdrawal volumes (capacities) for the Bamstable County water systems.2-28 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-6 Barnstable County Public Water Suppliers and Capacities for the Year 2003.Average Authorized Withdrawal Consumption Volume (Capacity MGD)(MGD)3 Barnstable Fire Districtl 0.54 0.66 Barnstable Water Company 1 2.57 3.42 Bourne Water District 2 1.17 1.40 Buzzards Bay Water District 2 0.46 0.53-COMM Water Department 1 2.74 3.57 Cotuit WtrDepartment 1 Ad,:C ,;
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| * 0.49 j 0 , ,,,, ,-. 48,,;Adi-
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| , Mashpee Water Department 1.26 1.30 North Saaore Water Disrd2.5 O8 Sandwich Water District 1.67 2.64 SothSgamore ,Water itri , 0.1 0-.0<n -t 1. The Town of Bamstable is composed of 7 villages and is serviced by 4 water suppliers.
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| : 2. The Town of Boume Is composed of 7 villages and is serviced by 4 water suppliers.
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| : 3. Reference 2-25 4. Reference 2-26 Because no county-level data were available, Entergy evaluated the water systems in the Plymouth and Barnstable County towns where approximately 70% of the Pilgrim workforce reside. The remaining 30%of the workforce was scattered among numerous towns and few employees lived in any single town.Shading indicates communities where consumption exceeds capacity and shortfalls are made up by purchase.2.9.1.3 Assessment As presented in Table 2-5 and Table 2-6, average daily consumption rates exceed the authorized withdrawal limits (capacities) in several communities.
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| Those communities purchase water from communities with excess capacity to meet the residual demand. Overall, the region has excess capacity and has been able to meet total demand. The Town of Plymouth is reviewing options for meeting future demand [Reference 2-10].2-29 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (_, 2.9.2 Transportation Road access to PNPS is via Rocky Hill Road or Power House Road (formerly known as Edison Access Road). Both are two-lane paved roads, the second of which is privately owned by Entergy (see Figure 2-2 and Figure 2-3). Rocky Hill Road intersects with State Route 3A approximately 1.5 miles west of the station, and Power House Road intersects with State Route 3A, approximately 1.5 miles south of the station and 2.5 miles east of the Rocky Hill/3A intersection.
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| State Route 3A runs north-south through the Town of Plymouth, providing access to Rocky Hill Road and Power House Roads from Plymouth.
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| State Route 3A provides access to the major north-south highway in the vicinity of the Town of Plymouth, State Route 3. State Route 3 is used by employees traveling south from the towns of Marshfield, Duxbury, Kingston, and Pembroke.Employees traveling north would use either State Route 3A or 3 to Beaver Dam Road, which intersects State Route 3A south of Power House Road. Employees traveling east to PNPS would use State Route 44 to State Route 3A or 3. The level of service determination for the State Route 3A intersection with Beaver Dam Road (southeast of PNPS) and White Horse Road (the eastern extension of Beaver Dam Road) is C [Reference 2-50]. Table 2-7 provides daily traffic counts for roads in the vicinity of PNPS. The Massachusetts Highway Department does not have level-of-service data for those roads.Table 2-7 Traffic Counts for Roads in the Vicinity of PNPS RRoute otLoainEstimated Average Daily Ya No. Route Location Traffic Volume Year 3 North of Clark Road 1 30,500 1992 3A North of Beaver Dam Road 14,400 2003 3A South of Rocky Hill Road 13,000 1995 3A South of Route 44 12,700 1998 44 East of Route 3 17,677 1990 Source: Reference 2-22.1. Beaver Dam Road is known as Clark Road south of the intersection with Sandwich Road (see Figure 2-2)2-30 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.10 Meteoroloaical and Air Quality PNPS is located along the rocky western shoreline of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts.
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| The station proper is on the Bay side of the northeast end of Pine Hills, a ridge of low hills about four miles long and tending in a north-south direction[Reference 2-1, Section I.D]. These hills reach a maximum height of 395 feet and form the major drainage divide in the area [Reference 2-1, Section II.D1. Since the site is located along the coast, approximately 60% of the area within a 50-mile radius is open water [Reference 2-1, Section ll.B].The temperature regime of the region is influenced by the proximity of the adjacent waters and as such does not exhibit the wider diurnal and seasonal variations of nearby inland locations.
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| The average annual temperature at Plymouth is 50 0 F with a high monthly average of 71'F in July and low monthly average of 29 0 F in February [Reference 2-37, Section 2.3.5]. Monthly averages for precipitation at Plymouth vary from about 3 inches to 4.5 inches. Although snowfall amounts typically average 42 inches per year, the Plymouth area is subjected to a wide range of snowfall since it is located in the northeastern part of the United States. The storm cycle consists generally of northeasters in the winter and spring, and thunderstorms in late spring and summer.Hurricanes sometimes occur in the late summer and fall, with tornado activity in eastern Massachusetts being uncommon.Plymouth County is part of the Metropolitan Providence Interstate Air Quality Control Region (AQCR). This AQCR is composed of part of Massachusetts and all of Rhode Island. Based on 40 CFR 81 and the EPA's 2003 Annual Report on Air Quality in New England, PNPS is located in a non-attainment area for ozone that is classified as serious for the 1-hour standard and moderate for the 8-hour standard.
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| For particulate matter (PM 1 0), sulfur dioxide, carbon monoxide, nitrogen dioxide, and lead, the area is either in attainment or designated as unclassifiable.
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| The closest non-attainment area for particulate matter is New Haven, Connecticut, approximately 135 miles from PNPS. The closest non-attainment area for sulfur dioxide is Mansfield, New Jersey, approximately 250 miles from PNPS. There are no designated Class I Federal areas listed in 40CFR81.41 within a 50-mile radius of PNPS.PNPS has house heating boilers and diesel generators located on-site. Emissions from these sources are regulated by an emissions cap approved by the MDEP in July 2005. This cap limits facility emissions to less than 50% of the major source category emissions.
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| This permit limits the fuel usage and hours of operation of these emission sources.2.11 Historic and Archaeological Resources 2.11.1 Pre- and Post-Construction Historic/Archaeological Analyses The FES for construction of PNPS, published in 1972, states that the Atomic Energy Commission (AEC) consulted with the Department of the Interiors Advisory Council on Historic Preservation regarding the potential impacts of PNPS on local historic landmarks
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| [Reference 2-1]. The 2-31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Council concluded that the probable effect on these properties cannot be judged to be sufficiently adverse to warrant Council comment [Reference 2-1]. The FES also stated that there is no evidence that the site has any specific historical significance
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| [Reference 2-1].The FES for construction of the proposed PNPS Unit 21, published in 1974, indicated that an extensive archaeological survey was conducted in October 1972 on the original 517-acre station site plus the transmission corridor extending southwest to Jordan Road [Reference 2-2].Archaeologists and students from the Archaeological Research Department of Plimoth Plantation and the Brown University Department of Anthropology conducted the survey. Twenty-four historic sites were discovered and determined to be insignificant
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| [Reference 2-2]. One pre-historic site (located in the southwest corner of the original PNPS property) was considered to be significant
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| [Reference 2-2]. A second more extensive examination, conducted with the assistance of the Massachusetts Archaeological Society, resulted in the conclusion that there was "no evidence of Indian occupation" in the area of the station [Reference 2-2]. The Massachusetts Archaeological Society report also concluded that the onsite pre-historic site was not significant
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| [Reference 2-3]. Therefore, Boston Edison concluded that there were no historical, cultural, archaeological, or architectural resources that would be affected by the construction or operation of Unit 2 [Reference 2-2]. This conclusion was supported by the Massachusetts Historical Commission in a letter dated April 24, 1974 [Reference 2-2].On November 27, 1990, the NRC issued an Environmental Assessment for the extension of the PNPS operating license from August 26, 2008, to June 9, 2012. In the environmental assessment, the NRC reported that the continued operation of PNPS would meet 36 CFR 800 0"Protection of Historic Properties" requirements
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| [Reference 2-31]. After researching the National Historic Register files through the Massachusetts Historical Commission and consulting with a number of local and national historical organizations, NRC concluded that there had been no evidence of local historic site deterioration due to plant operations
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| [Reference 2-31]. Therefore, the NRC concluded that "the operation of Pilgrim Nuclear Power Station.. .will cause no adverse effect or induce any detrimental impact on the historic sites located in Plymouth" [Reference 2-31].2.11.2 Additional Information Regarding the Plimoth Plantation/Brown University Archaeological Survey The October 1972 survey reported that, because pre-historic archaeological sites in the general locale were of a very low profile, they would be difficult to discover in the rugged terrain of the survey area. None of the areas surveyed was heavily populated during the historic period (Colonial or European settlements).
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| Nearby Plymouth was sparsely settled in 1620. Most of the Rocky Hill area was considered too rugged for settlers' habitation or agricultural production.
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| Seventeenth and eighteenth century sites may have existed in the well-drained land and oceanfront areas. Local informants recall an early cellar that may have been destroyed in the construction of Power House Road. An indication of this particular habitation appears on a late nineteenth century map of the area. However, the same map reinforces the observation that few sites of early habitation would be found in the Rocky Hill area [Reference 2-3, Amendment 6].1. Unit 2 was never built.2-32 Pilgrim Nuclear Power Station I IApplicant's Environmental Report Operating License Renewal Stage 2.11.3 Current Historic/Archaeological Analysis An examination of the archaeological site files and maps maintained by the Office of the State Archaeologist at the Massachusetts Historical Commission revealed approximately 130 archaeological (pre-historic and historic) sites within a 6-mile radius of the station. Five sites (84, 813, 815, 816, and 19-68) appear to fall within or near the Jordan Road transmission corridor.Beyond the Jordan Road tap, site 361 appears to fall near the corridor.
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| Protective measures for such sites can include signage warning against ground disturbance without proper authorization and supportive procedures for protecting the resource in place or, in the extreme, relocating the resource.
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| However, Entergy does not own or manage these rights of way and has no authority to implement protective measures.Currently, 109 "above-ground" locations are listed in the National Register of Historic Places for Plymouth County [Reference 2-49]. Twenty of these locations are within the Town of Plymouth.The State Register of Historic Places 2003, a -report published by the Massachusetts Historical Commission, states that the Town of Plymouth is home to 21 sites or areas of historic significance
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| [Reference 2-23]. Table 2-8 lists the 21 sites, recognized by either one or both of the two agencies, which are located within the Town of Plymouth.Table 2-8 Town of Plymouth, Massachusetts, Sites Listed in the National Register of Historic Places and/or the State Register of Historic Places Site Name Location Bartlett-Russell-Hedge House 32 Court Street Bradford-Union Street Historic District Bradford, Union, Emerald, Water Cure, and Freedom Streets Clifford-Warren House East of Plymouth at 3 Clifford Road Cole's Hill -Carver Street Harlow Old Fort House 119 Sandwich Street Sgt. William Harlow Family Homestead 8 Winter Street Hillside 230 Summer Street Jabez Howland House 33 Sandwich Street Light Houses of Massachusetts (Thematic Group 42 properties in 23 towns Nomination)
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| National Monument to the Forefathers Allerton Street Old County Courthouse Leyden and Market Streets Parting Ways Archaeological District Address Restricted 2-33 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 2-8 Town of Plymouth, Massachusetts, Sites Listed in the National Register of Historic Places and/or the State Register of Historic Places (Continued)
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| Site Name Location Pilgrim Hall 75 Court Street Plymouth Antiquarian House 126 Water Street Plymouth Historic District 1 Roughly bounded by Town Square, Town Brook, Court, Main, and Water Streets from Samoset to Sandwich Streets Plymouth Light Station; Gurnet Point Plymouth Post Office Building 5 Main Street Plymouth Rock Water Street Plymouth Village Historic District Roughly bounded by Water, Main, and Brewster Streets Richard Sparrow House 42 Summer Street Town Brook Historic and Archaeological District Address Restricted Source: Reference 2-49 1. Not listed in the National Register of Historic Places, but listed in the State Register of Historic Places 2003 [Reference 2-23].(Q'2.12 Known and Forseeable Federal and Non-Federal Actions Entergy did not identify any known or reasonably foreseeable federal or non-federal projects or other activities that may contribute to the cumulative environmental impacts of license renewal.(4-0 J 2-34 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2.13 References Note to reader: Some web pages cited in this document are no longer available, or are no longer available through the original URL addresses.
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| Hard copies of all cited web pages are available in Entergy files. Some sites (e.g., the census data) cannot be accessed through their URLs. The only way to access these pages is to follow queries on previous web pages. The complete URLs used by Entergy have been cited for these pages, even though they may not be directly accessible.
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| 2-1 U.S. Atomic Energy Commission, Division of Radiological and Environmental Protection, Final Environmental Statement Related to Operation of Pilgrim Nuclear Power Station, Docket No. 50-293, Washington, DC, May 1972.2-2 U.S. Atomic Energy Commission, Directorate of Licensing, Final Environmental Statement Related to the Proposed Pilgrim Nuclear Power Station, Unit 2, Docket No.50-471, Washington, DC, September 1974.2-3 Boston Edison Company, Pilgrim Nuclear Power Station-Unit 2 Environmental Report, Amendment 1, Plymouth, MA, 1976.2-4 Cape Cod Commission, Regional Policy Plan, Barnstable, MA, November 1996.2-5 Cape Cod Commission, Cape Trends: Demographic and Economic Characteristics and Trends, Bamstable County -Cape Cod, 5th Edition, Barnstable, MA, 1998.2-6 Cape Cod Commission, Regional Policy Plan: County of Barnstable, Massachusetts, Barnstable, MA, September 10, 2003.2-7 Conant, R., A Field Guide to Reptiles and Amphibians of Eastern/Central North America, 2nd edition, Houghton Mifflin Company, Boston, MA, 1975.2-8 Corn, M. L., "The Northern Right Whale," CRS Report for Congress, April 14, 1995, available at http://www.cnie.orgfnle/crsreports/biodiversity/biodiv-12.cfm, accessed December 4, 2001.2-9 Daniels, T. L., J. W. Keller, and M. B. Lapping, The Small Town Planning Handbook, Chapter 16: 'The Zoning Ordinance," 2nd edition, American Planning Association, Chicago, IL, 1995.2-10 Dayian, L., Water System Data Provided by the Massachusetts Environmental Protection Department, personal communication with M. Hoganson, TtNUS, June 4, 2001.2-11 DeGraaf, R. M. and D. D. Rudis, New England Wildlife:
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| Habitat, Natural History, and Distribution, General Technical Report NE-1 08, U.S. Department of Agriculture, Forest Service, Northeastern Forest Experiment Station, 1986.2-35 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2-12 ENSR Corporation, Redacted Version 316 Demonstration Report -Pilgrim Nuclear Power Station, Document Number 0970-021-200, prepared for Entergy Nuclear Generation Company, Plymouth, MA, March 2000.2-13 Entergy Nuclear Generation Company, Environmental Protection Department, Pilgrim Nuclear Power Station, Marine Ecology Studies Related to Operation of Pilgrim Station, Semi-Annual Report #,56 (January -June 2000), Plymouth, MA, October 31, 2000.2-14 Entergy Nuclear Generation Company, Forest Management Plan, prepared for Entergy by Benjamin Forestry Services, South Easton, Massachusetts, and submitted to Massachusetts Department of Environmental Management, Division of Forests & Parks, Boston, MA, September 12, 2002.2-15 U.S. Fish & Wildlife Service, Threatened and Endangered Species System (TESS);Listings by State and Territory as of 02/23/2005:
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| Massachusetts, February 23, 2005, available at http://ecos.fws.gov/tesspublic/TESSWebpageUsaLists?state=MA.
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| 2-16 Keller, M. D. and M. E. Sieracki, 'Abstract:
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| Spring bloom dynamics in the Gulf of Maine, with emphasis on the noxious indicator phytoplankton species, Phaeocystis pouchetii," Gulf of Maine Information System, undated, available at http://woodshole.er.usgs.gov/
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| project-pages/oracle/GoMaine/spring.htm, accessed December 4, 2001.'2-17 Maccaferri, E., Jr., Town of Plymouth Treasurer, "Total revenues and operating budget information," Facsimile transmission to E. N. Hill, TtNUS, Plymouth, MA, May 22, 2001.2-18 Metropolitan Area Planning Council, Metro Plan: Summary of the Regional Plan for the Boston Metropolitan Area, Boston, MA, 2001.2-19 Massachusetts Department of Revenue Division of Local Services, "At a Glance Report for Plymouth (As of 4/16/02)," Boston, MA, April 2002, available at http://dorapps.dor.state.ma.us/ataglance/home/communitylist.ASP?report3, accessed January 12, 2005.2-20 Massachusetts Department of Revenue Division of Local Services, "Municipal Budgeted Revenues:
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| Revenues by Source FY2000 through 2004," Boston, MA, 2004, available at http://www.dis.state.ma.us/mdmstuf/MunicipalBudgetedRevenues/
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| RevsO004.xIs, accessed January 12, 2005.2-21 Massachusetts Executive Office of Environmental Affairs, "Massachusetts Geographic Information System -Land Use Summary Statistics
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| -September 2003," Boston, MA, 2003, available at http://www.state.ma.us/mgis/landusestats.htm, accessed January 18, 2005.2-22 Massachusetts Highway Department, Traffic Data for Plymouth, Boston, MA, 2004, available at http://www.state.ma.us/mhd/trafficc/bytown/plymouth.htm, accessed January 10, 2005.2-36 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2-23 Massachusetts Historical Commission, State Register of Historic Places 2003, Boston, MA, 2004.2-24 Mass Audubon, "Natural History: Sea Turtles on Cape Cod," 2003, available at http://www.massaudobon.org/NatureConnection/SanctuarieslWellfleet/seaturtles.php, accessed January 11, 2005.2-25 Massachusetts Department of Environmental Protection, Bureau of Resource Protection
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| -Drinking Water Program, 2003 Public Water Supply Annual Statistical Report, Boston, MA, 2004.2-26 Massachusetts Department of Environmental Protection, Email correspondence between J. Drake (MDEP) and J. Brochu, Entergy, February 25, 2005.2-27 Massachusetts Division of Fisheries and Wildlife, "Rare Species by County: Plymouth," Boston, MA, March 1, 2003, available at http://www.mass.gov/dfwele/dfw/nhesp/
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| plym.htm, accessed January 11, 2005.2-28 Massachusetts Division of Fisheries and Wildlife, "Massachusetts List of Endangered, Threatened and Special Concern Species," Boston, MA, June 18, 2004, available at http://www.mass.gov/dfwele/dfw/nhesp/nhrare.htm, accessed February 23, 2005.2-29 Massachusetts Institute for Social and Economic Research, "MISER Population Projections for Massachusetts, 2000-2020," Boston, MA, 2003, available at http:/l www.umass.edu/miser/population/miserproj.htm, accessed January 28, 2005.2-30 National Marine Fisheries Service, Office of Protected Resources, "Northem Right Whale (Eubalaena glacialis):
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| Western North Atlantic Stock," Stock Assessment Report (2001), 2001, available at http://www.nmfs.noaa.gov/protres/PR2I StockAssessment Program/individualsars.html, accessed November 6, 2001.2-31 U.S. Nuclear Regulatory Commission, Environmental Assessment by the Office of Nuclear Reactor Regulation Relating to the Change in Expiration Date of Facility Operating License No. DPR-35 Boston Edison Company for the Pilgrim Nuclear Power Station, Docket No. 50-293, Washington, DC, November 27, 1990.2-32 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes 1 and 2, Washington, DC, May 1996.2-33 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,'Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues," NRR Office Instruction No. LIC-203, Revision 1, May 24, 2004.2-34 Old Colony Planning Council, Regional Policy Plan: A Guide for Shaping Our Communities and the OCPC Region, October 20, 2000.2-37 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2-35 Peterson, R. T., Eastem Birds, Peterson Field Guides, 4th edition, Houghton Mifflin Company, Boston, MA, 1980.2-36 Plymouth County Development Council, 'Plymouth County Business Information," Plymouth, MA, 2001, available at http://www.plymouth-1620.comlWeb/My%2OWebs/
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| business/statistics.htm, accessed April 30, 2001.2-37 Pilgrim Nuclear Power Station, Updated Final Safety Analysis Report, Plymouth, MA.1 2-38 Pilgrim Nuclear Power Station, Population Projection Calculations for Pilgrim Nuclear Power Station, Plymouth, MA, 2005.2-39 Prescott, R., "Sea Turtles in New England Waters," in Conservation Perspectives, the on-line journal of the Massachusetts Chapter of the Society for Conservation Biology, Inc., October 2000, available at http://www.massscb.org/epublications/october2000/
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| seaturtle.html, accessed December 4, 2001.2-40 Prescott, R., Email correspondence with J. Brochu, Entergy, January 15, 2005.2-41 Town of Plymouth, Annual Report of the Town of Plymouth, Massachusetts for the Year Ending December31, 2003, Plymouth, MA, 2004.2-42 U.S. Census Bureau, "Massachusetts:
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| Population of Counties by Decennial Census: 1900 to 1990," 1995, available at http:/lwww.cache.census.gov/population/cencountsl
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| /ma190090.txt, accessed April 18, 2001.2-43 U.S. Census Bureau, American Factfinder, "QT-PL. Race, Hispanic or Latino, and Age: 2000 Data Set: Census 2000 Redistricting Data (Public Law 94-171) Summary File, Massachusetts," 2000, available online at http://factfinder.census.gov/, accessed December 4, 2002.2-44 U.S. Census Bureau, American Factfinder, "QT-PL. Race, Hispanic or Latino, and Age: 2000. Data Set: Census 2000 Redistricting Data (Public Law 94-171) Summary File, Rhode Island," 2000, available online at http://factfinder.census.gov/, accessed December 4, 2002.2-45 U.S. Census Bureau, American Factfinder, "HCT24: Tenure by Poverty Status in 1999 by Age of Householder
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| [431 -Universe:
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| Occupied Housing Units: 1999, Data Set: Census 2000 Summary File 3 (SF3) -Sample Data," 2000, available online at http://factfinder.census.gov, accessed December 4, 2002.SW 1. Pilgrim's UFSAR update is done on a page-by-page basis, rather than by entire section or volume. Therefore, several different revisions (up to Revision 24) of the UFSAR update have been used in this Environmental Report.2-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 2-46 U.S. Census Bureau, "State and County QuickFacts:
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| Plymouth County, Massachusetts," 2001, available online at http://quickfacts.census.gov/qfd/states/25/
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| 25023.html, accessed April 30, 2001.2-47 U.S. Census Bureau, "State and County QuickFacts:
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| Barnstable County, Massachusetts," 2001, available online at http://quickfacts.census.gov/qfd/states/25/
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| 25001.html, accessed April 30, 2001.2-48 U.S. Department of Agriculture, National Agricultural Statistics Service, 2002 Census of Agriculture, Massachusetts State and County Profiles, 2004, available online at http://www.nassausda.gov/census/censusO2/profiles/ma/index.htm, accessed January 18, 2005.2-49 U.S. Department of the Interior, National Park Service, "Plymouth County, Massachusetts, Listing of Sites on the National Register of Historic Places", 2004, Available at http://www.nr.nps.gov/, accessed January 17, 2005.2-50 Vanasse & Associates, Inc., Traffic Data Technical Data Phased Review Document: Phase 1ll, The Pinehills, Plymouth, Massachusetts, Volume II, Appendix E, Andover, MA, 2001.2-51 Massachusetts Division of Fisheries
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| & Wildlife, National Heritage & Endangered Species Program, BioMap and Living Waters: Guiding Land Conservation for Biodiversity in Massachusetts, Core Habitats of Plymouth, Westborough, MA, 2004.2-39 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage I)Figure 2-1 50-Mile Vicinity Map 2-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Figure 2-2 General Area Near PNPS 2-41 CO) I ' --' -- -' '00 q~~~~ i. .. ..1 .. .... ,\0 >co Ca.TM~ro~ftUnas(DO co 0 0 C. CC.
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage*PIPS*#Jr I' and mf wirican Indian, or Alaskan Native B C 10 iS 20 Wim Rbbd& 8 V o5 20 25 035asKbnfto Figure 2-4 American Indian or Alaskan Native Minority Population Map 2-43 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage pa 1 Co: .-*PNP$-.2,y Affierican kam o Amea\ikan Nivran e ic. , Counitys o IriI Aan.5 20 25 32 5&-Figure 2-5 Asian or Pacific Islander Minority Population Map 2-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Et-e/- W '0 .X-- ''oys : h&faad; VI
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| * N ~ef; C --y i ~.E f ' -' *0-\i LIEVEND-Native Haw~aun crMP ofter Padific 1 IWandr Minority County Boundtaries__
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| _ _ _ _ _ _ _ _ _ _ _8 c 5 c0co 20 M1 i 6t Is 'iS63 3 ixb Figure 2-6 Native Hawaiian or Other Pacific Islander Minority Population Map 2-45 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage a'~.Cap& Codi T :V Sotmd-., ~~~~Mads .Eitf:VD:0:j E;: IWiSiT?:EyEC:0:iet0 YEi':it:if;Lii:;f ffi?: inf;;, i~ f~~t:;ii::i~g
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| -, i i t S i .etE:!diS f;2: ~ifiE :: S ty if:, S, ~if g0 ER! L E / E 7, , a:rdi!l--f4::NE 7 L;:, i : 5 : , : f ::;L , hi? i~Ei: 7 A W7:] !g: LEG&ND-Black k~riocity
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| ,,N!PS.- -County Bourodanies
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| % Is 0-C.. I 10 is 2t0&,sdI hUid 3 0 5 ¶O 15 20 26 303ti5 K13mwz Figure 2-7 13Black Races Minority Population Map 2-46 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage yxay LEGEND-.l' Oeer SingI' Minorities P=P County Boundaries J WN q Nt is la is z- -W2S 33 -13v;Figure 2-8 All Other Single Minorities Map 2-47 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage At a~t I ' t ,i-n PAf* ot'.MAsi i*.:._j.iod X ,, , , f Aa-! .Y;-n -- -t -i -fE -A~i 10 C-Aggregate of Mintority RacesIiI* Courdy Soundaries]
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| blind I 0 5 1.0 15 20 25 30 35 KS~met. I Figure 2-9 Aggregate of Minority Races Population Map 2-48 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage IsC frig2t *PNPS Cepo Ca'd* BSay 4'7''ndtuket... , j ;;;; 0 tdS;0;}$00 0;0 ;;3-0f0S ;;.. ~ ~ ~ ~ ~ Ua~n EWn eywdlt5 00020000E; 0::~;0 ::ig 000ii;0;00t~03 t0tSt 0@;s~~~~~~~~~~.
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| ji'''St' ';t 'iEE:SC ,"J'%/ .a ,0,-~f-ni"0ydDf~t:.eii~,,00ttSifi!if~i~i000F,00s
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| ;i;000;t;0
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| *;;Td ; i;i LEGEND County Soundaries 10*20 25 20 MSKlntu_LJWJA1s 0 M5 S 02 3 5kimbs Figure 2-10 Hispanic Minority Population Map 2-49 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (_m St4 W Pi rwl.I*PNPI.C"p Cod 7n....I LI-7: C'.E f F-,; ..i i! : i F E r; YES;Mikemd IljrtbavgWaw
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| .... --.Wm i:: E i 4i: ni S B: LEGEND County Boundaries I 0 10 20 6 30IS metwz Figure 2-11 Low-Income Population Map 2-50 (C4'
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| .; Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Figure 2-12 State and Federal Lands-50 Mile Radius 2-51 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 3.0 THE PROPOSED ACTION 3.1 Descriotion of the Proposed Action The proposed action is to renew the facility operating license for PNPS for an additional 20 years beyond the expiration of the current operating license. For PNPS (Facility Operating License DPR-35), the requested renewal would extend the license expiration date from midnight June 8, 2012, to midnight June 8, 2032.There are no changes related to license renewal with respect to operation of PNPS that would significantly affect the environment during the period of extended operation.
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| The application to renew the operating license of PNPS assumes that licensed activities are now conducted, and would continue to be conducted, in accordance with the facility's current licensing bases (e.g., use of low enriched uranium fuel only). Changes made to the current licensing basis of PNPS during the staff review of this application would be made in accordance with the Atomic Energy Act of 1954, as amended, and in accordance with Commission regulations.
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| 3.2 General Plant Information The principal structures at PNPS consist of the reactor and turbine buildings (each with auxiliary bays), the offgas retention building, the radwaste building, the diesel generator building, the administration building, the intake structure, and the main stack [Reference 3-6, Section 12.1].The reactor and nuclear steam supply system for PNPS, along with the mechanical and electrical systems required for the safe operation of PNPS, are primarily located in the reactor building.
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| C;Figure 3-1 shows the general features of PNPS and the station layout. Figure 2-3 shows the site boundaries.
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| No residences are permitted within the site boundaries, with the nearest residence being outside of the NRC-mandated 1800-foot exclusion zone.3.2.1 Reactor and Containment Systems PNPS is a single-unit plant with a boiling water reactor design and a turbine generator manufactured by General Electric Company. The architect/engineer and constructor was Bechtel. The unit was initially licensed for an output of 1,998 megawatts-thermal (MWt), and an electric rating of 687 megawatts-electric (MWe) [Reference 3-6, Section 1.1]. PNPS achieved commercial operation in December 1972. In 2003, PNPS implemented a Thermal Power Optimization of 1.5% to achieve the current electrical rating of 715 MWe.The reactor's primary containment is a pressure suppression system consisting of a drywell, pressure suppression chamber, vent system, isolation valves, containment cooling system, and other service equipment.
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| The containment is designed to withstand an internal pressure of 62 pounds per square inch above atmospheric pressure and act as a radioactive materials barrier [Reference 3-6, Section 5.2.3.21.
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| A secondary containment completely encloses both primary containment and fuel storage areas and acts as a radioactive materials barrier.Together with their engineered safety features, each containment is designed to provide adequate radiation protection for both normal operation and postulated design-basis events or accidents, such as earthquakes or loss of coolant. 1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage PNPS fuel is low-enriched uranium dioxide with maximum enrichments of 4.6% by weight uranium-235 and fuel burnup levels of 48,000 megawatt-days per metric ton uranium.3.2.2 Cooling and Auxiliary Water Systems 3.2.2.1 Surface Water PNPS is equipped with a once-through heat dissipation system that withdraws cooling water from and discharges it to Cape Cod Bay (Figure 3-1). The principal components of the circulating water system are the intake canal, intake structure or "screen house" with the intake pumps, condenser and service water systems, and discharge canal (Figure 3-1).Two pumps in the intake structure provide a continuous supply (311,000 gallons per minute[gpm]) of condenser cooling water. Also housed in the intake structure are five service water pumps (four running and one on standby) that can supply 13,500 gpm of cooling water to the service water system. Seawater for cooling and service water is withdrawn from Cape Cod Bay via an embayment formed by two breakwaters.
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| The intake structure consists of wing walls, a skimmer wall which functions as a submerged baffle, vertical bar racks that capture large debris, and vertical traveling screens. The four traveling screens (two per condenser cooling water pump) prevent small debris and small aquatic organisms from being entrained into the cooling water or service water systems. Each screen is made up of 53 basket segments with 1/ inch by/2 inch stainless steel mesh. The screens are washed when they are operating.
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| The wash normally is discharged via a sluice to the intake embayment approximately 300 feet from the intake structure.
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| During storms, the wash is discharged to the discharge canal [Reference 3-61.During spring, summer, and fall, the circulating water system is chlorinated for up to two hours per day, one hour each pump, to control nuisance biological growth. Total residual chlorine cannot exceed 0.10 parts per million (ppm) in the cooling water discharge
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| [Reference 3-3].Continuous chlorination of the service water system can be used to control nuisance biological organisms with a maximum daily concentration of 1.0 ppm and an average monthly concentration of 0.5 ppm [Reference 3-3] in the service water discharge.
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| During chlorination, the screens are operated, and sodium thiosulfate is added to the wash water to remove chlorine and protect organisms returned to the intake canal. Molluscicides are not permitted without the prior approval of the EPA and the Commonwealth
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| [Reference 3-3].After moving through the condensers, cooling water is discharged into a 900-foot-long discharge channel immediately adjacent to the intake embayment.
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| The discharge channel is created by two breakwaters, one of which is shared with the intake embayment.
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| At low tide, the water in the discharge channel is several feet higher than sea level and the discharge is rapid and turbulent.
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| At high tide, the velocity is much lower. The increase in water temperature across the condensers ranges from 27 to 30'F [Reference 3-2]; the plant is permitted for as much as a 32 0 F temperature change [Reference 3-3].3.2.2.2 Groundwater The Town of Plymouth gets its water from groundwater (see Section 2.9.1) and supplies potable and reactor makeup water to PNPS via the town's municipal water system. PNPS's estimated 3-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage annual water consumption for a non-outage year (based on May 2003 through April 2004 actual consumption) is approximately 39.1 million gallons.PNPS has an onsite sewage treatment and disposal facility.
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| Wastewater is processed in the wastewater treatment facility and ultimately discharged to a leach field (Reference 3-4).Because the groundwater flow at the site is toward Cape Cod Bay, any treated discharge that may reach the groundwater does not enter a drinking water source.The site has one groundwater well, which has been used in the past for irrigation purposes only.The well is capable of producing at a rate of 20 gpm. This well was installed in 2000; however, it is no longer in use for irrigation purposes, and it is not anticipated that the well will be returned to service at anytime in the future.3.2.3 Radioactive Waste Treatment Processes (Gaseous, Liquid, and Solid)PNPS uses liquid, gaseous, and solid waste processing systems to collect and treat, as needed, radioactive materials that are produced as a by-product of plant operations.
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| Radioactive materials in liquid and gaseous effluents are reduced to levels as low as reasonably achievable.
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| Radionuclides removed from the liquid and gaseous processing systems are converted to a solid waste form for eventual disposal with other solid radioactive wastes at a licensed disposal facility.The PNPS waste processing systems meet the design objectives of 10 CFR 50, Appendix I, and control the processing, disposal, and release of radioactive liquid, gaseous, and solid wastes.Radioactive material in the reactor coolant is the source of most gaseous, liquid, and solid radioactive wastes in light water reactors.
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| Radioactive fission products build up within the fuel as a consequence of the fission process. The fission products are contained within the sealed fuel rods; however, small quantities of radioactive materials may be transferred from the fuel elements to the reactor coolant under normal operating conditions.
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| Neutron activation of materials in the primary coolant system also contributes to radionuclides in the coolant.Radioactive wastes resulting from station operation are classified as liquid, gaseous, and solid.The following definitions apply to radioactive wastes [Reference 3-6, Section 9.1].(1) Liquid Radioactive Wastes -Liquids directly from the reactor process and auxiliary systems or liquids which can become contaminated due to contact with these liquids from reactor process systems (2) Gaseous Radioactive Wastes -Gases or airborne particulates vented directly from reactor and turbine equipment containing radioactive material or indirectly from the main stack (3) Solid Radioactive Wastes -Solids from the reactor primary or auxiliary systems, solids in contact with reactor primary system liquids or gases, and solids (such as cleaning materials), used in reactor primary, turbine systems, and auxiliary systems operations.
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| 3-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Reactor fuel assemblies that have exhausted a certain percentage of their fissile uranium content are referred to as spent fuel. Spent fuel assemblies are removed from the reactor core and replaced by fresh fuel during routine refueling outages, typically every 24 months. The spent fuel assemblies are then stored for a period of time in the spent fuel pool in the reactor building and may later be transferred to dry storage, if needed, at an onsite interim spent fuel storage installation provided necessary regulatory approvals are obtained.
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| PNPS also provides for onsite storage of mixed wastes, which contain both radioactive and chemically hazardous materials.
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| Storage of radioactive materials is regulated by the NRC under the Atomic Energy Act of 1954, as amended, and storage of hazardous wastes is regulated by the EPA under the Resource Conservation and Recovery Act of 1976.Systems used at PNPS to process liquid, gaseous, and solid radioactive wastes are described in the following sections.3.2.3.1 Liquid Waste Processing Systems and Effluent Controls The Liquid Radwaste System collects, processes, stores, and disposes of all radioactive liquid wastes. Equipment is selected, arranged, and shielded to permit operation, inspection, and maintenance within personnel radiation exposure limits. Sumps, pumps, valves, and instruments are located in controlled access areas. Tanks and processing equipment which may contain i quantities of liquid radwastes are shielded.
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| In addition, equipment is selected for a minimum of maintenance.
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| [Reference 3-6, Section 9.2.4]The system is divided into several subsystems so that the liquid wastes from various sources can be segregated and processed separately.
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| Cross connections between the subsystems provide additional flexibility for processing of the wastes by alternate methods. The liquid radwastes are classified, collected, and treated in subsystems as either clean, chemical, or miscellaneous radwastes.
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| [Reference 3-6, Section 9.2.4]Very lows levels of radioactivity may be released in plant effluents if they meet the limits specified in the NRC's regulations.
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| These releases are closely monitored and evaluated for compliance with NRC restrictions in accordance with the PNPS Offsite Dose Calculation Manual.3.2.3.1.1 Clean Radwaste Clean radwastes are liquids having a varying amount of radioactivity and are expected to have low conductivity.
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| Clean radwaste is collected in the following sumps. [Reference 3-6, Section 9.2.4.11* drywell equipment drain sump,* reactor building equipment drain sump* turbine building equipment drain sump* radwaste building equipment drain sump* retention building equipment drain sump 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage From these sumps, the wastes are transferred to the clean waste receiver tanks for processing.
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| The drywell and turbine equipment drain sump discharge may be directed to the main condenser in order to provide operating flexibility and reduce water inventory delivered to radwaste for processing.
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| Resin transfer water, ultrasonic resin cleaner (URC) flushwater, and drains are routed to the clean waste receiver tank. [Reference 3-6, Section 9.2.4.1]The clean radwaste system also receives liquid from the URC. The URC is designed to remove suspended solids from condensate demineralizer resins without requiring chemical regeneration.
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| The major components of the URC are the cleaning column, flow adjustment panel, and control panel. Resin enters the cleaning column and falls through an ultrasonic field where the solids are removed. A countercurrent flow of water removes the solids and resin fines and transfers them to a holding tank. The wastewater containing the solids is then pumped to the clean radwaste system and/or chemical waste system. The cleaned resin is then transferred back to the condensate demineralizer system for reuse. [Reference 3-6, Section 9.2.4.1]Wastes from the receiver tanks are processed through flat bed filters and/or a mixed bed demineralizer, thermex, and/or radwaste filter demineralizer, or other water processing equipment before collection in the treated water holdup tanks. After the liquid wastes in the treated water holdup tanks have been sampled and analyzed, they are normally returned to the condensate storage tanks (CST) for reuse within the plant or sent to the main condenser hotwell.If the analysis of the sample reveals water of high contaminants or high radioactivity concentration, it may be reprocessed.
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| Abnormally high conductivity water may either be reprocessed in the chemical waste system or be discharged at a controlled rate through the liquid radwaste discharge header to the circulating water discharge canal. [Reference 3-6, Section 9.2.4.1]3.2.3.1.2 Chemical Radwaste Chemical radwastes are liquid wastes which generally have low concentrations of radioactive impurities and rather high conductivities.
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| [Reference 3-6, Section 9.2.4.2.1]
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| Chemical radwastes are collected in the following sumps [Reference 3-6, Section 9.2.4.2.1].
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| * drywell floor drain sump* reactor building floor drain sump* turbine building floor drain sump* radwaste building floor drain sump* retention building floor drain sump The sump wastes are primarily minor equipment leakages, tank overflows, equipment drains, and floor drainage.
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| When a sump has filled to a preset liquid level, the wastes are automatically pumped to the chemical waste receiver tank. Floor drain sump wastes may also be processed through the clean radwaste system if the wastes are relatively low in conductivity.
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| Laboratory wastes are routed directly to the chemical waste receiving tank. [Reference 3-6, Section 9.2.5.2.1]
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| 3-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The chemical waste receiver and monitor tanks are atmospheric tanks with a capacity of 15,000 gallons and 20,000 gallons, respectively.
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| The receiver tanks have level indicators and annunciators which will be used in monitoring the waste inventory.
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| The monitor tanks have level and temperature indicators and annunciators.
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| Depending on the activity level, the wastes after storage and decay may be released on a controlled basis through the liquid radwaste discharge header to the circulating water discharge canal or further processed.
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| Both the chemical waste receiver and monitor tanks are located in shielded cells to maintain safe operating conditions and minimize radiation exposure to station personnel.
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| [Reference 3-6, Section 9.2.4.2.1]
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| During operation it is expected that the daily flow from the floor drain sumps will be approximately 5,000 gallons. The drywell floor sump wastes will normally be transferred to the clean radwaste system. The chemical wastes can be pumped through the Thermex filter to remove suspended solids. [Reference 3-6, Section 9.2.4.2.21 3.2.3.1.3 Miscellaneous Radwaste Miscellaneous radwastes are those wastes which potentially have high detergent or contaminant level, but are of low radioactivity concentration.
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| [Reference 3-6, Section 9.2.4.3.2]
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| The miscellaneous waste system collects equipment washdown and decontamination solution wastes, radiochemistry laboratory solution wastes, miscellaneous water waste, and personnel decontamination wastes. The miscellaneous waste system processes and strains these liquid wastes before discharge through the radwaste discharge header into the circulating water discharge canal. The liquid wastes are sampled and analyzed before release and continually monitored during release. [Reference 3-6, Section 9.2.4.3.2]
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| The miscellaneous waste drain tank collects drainage from floor drains originating in the following areas [Reference 3-6, Section 9.2.4.3.21.
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| * turbine washdown area* personnel decontamination areas* fuel cask decontamination area* reactor head washdown area* truck decontamination area* machine shop wastes* retube building decontamination area During normal operation it is expected that the monthly volume of miscellaneous wastes will be approximately 1,000 gallons. When one section of the miscellaneous waste tank is filled, the wastes are sampled and analyzed for radioactivity.
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| The wastes are pumped through a strainer and discharged at a controlled rate through the liquid radwaste discharge header into the circulating water discharge canal. The miscellaneous waste is continuously monitored for activity as it passes through the radwaste discharge header. If necessary, miscellaneous wastes of high radioactivity concentrations and low detergent levels may be transferred to the chemical waste receiver tank for further processing.
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| [Reference 3-6, Section 9.2.4.3.3]
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| 3-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Controls for limiting the release of radiological liquid effluents are described in the ODCM.Controls are based on (1) concentrations of radioactive materials in liquid effluents and projected dose or (2) dose commitment to a hypothetical member of the public. Concentrations of radioactive material that may be released in liquid effluents to unrestricted areas are limited to the concentration specified in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration of individual isotopes shall be limited to 2E-04 microcurie/ml
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| [Reference 3-5, Section 3.2.11. The ODCM dose limits during a calendar quarter are < 1.5 mrem to the total body and < 5 mrem to any organ [Reference 3-5, Section 3.2.2]. During the calendar year, the ODCM dose limits are < 3 mrem to the total body and < 10 mrem to any organ [Reference 3-5, Section 3.2.2]. Radioactive liquid wastes are subject to the sampling and analysis program described in the ODCM.3.2.3.2 Gaseous Waste Processing Systems and Effluent Controls The gaseous radwaste system processes gaseous radioactive wastes from the main condenser air ejectors, the startup mechanical vacuum pump, the gland seal condensers, and other minor sources, and controls their release to the atmosphere through the main stack in such a way that the operation and availability of the station is not limited. [Reference 3-6, Section 9.4.1]3.2.3.2.1 Air Ejector Offgas and Augmented Offgas System The air ejector and augmented offgas (AOG) system includes the subsystems that process and/or dispose of the gases from the main condenser air ejectors, the startup mechanical vacuum pump, and the gland seal condensers.
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| All such gases from the unit are routed to the main stack for dilution and elevated release to the atmosphere.
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| Discharges from the air ejector, the charcoal vault, and the stack are continuously monitored by radiation monitors.
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| [Reference 3-6, Section 9.4.4.1.1]
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| Gases routed to the main stack include air ejector and gland seal offgases, and gases from the standby gas treatment system (SGTS). Dilution air input to the stack is supplied by two full capacity fans located in the filter building at the base of the main stack. The stack is designed such that prompt mixing of all gas inlet streams occurs in the base to allow location of sample points as near the base as possible.
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| The stack drainage is routed to the liquid radwaste collection system. [Reference 3-6, Section 9.4.4.1.1]
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| The AOG system provides for the controlled recombination of radiolytic hydrogen and oxygen, followed by chilling of the gas mixture to strip the condensable water vapor and reduce the volume and relative humidity of the remaining noncondensables, principally inleakage air with traces of the radioactive noble gases krypton and xenon, which are delayed by an adsorption process using activated charcoal.
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| The offgas passes through the charcoal vessels and is then discharged to the environs via the main stack. The delay time created by the charcoal adsorption process allows for the continued decay of the krypton and xenon radioactivity to a point where the ultimate release of the offgas results in a site boundary gamma radiation dose that meets the definition of ALARA (As Low As Reasonably Achievable).
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| The radioactivity of the gas mixture is 3-7 Pilgrim Nuclear Power Station Applicant's Environmental Report i IOperating License Renewal Stage monitored immediately downstream of the steam jet air ejectors, representing the inlet conditions to AOG, and at the discharge from the AOG system. [Reference 3-6, Section 9.4.4.1]The offgas system is provided with flow, temperature, and radiation instrumentation to ensure proper operation and control. Hydrogen analyzer instrumentation is also provided to ensure that hydrogen concentration is maintained below the flammable limit. [Reference 3-6, Section 9.4.4.1.2]
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| The offgas radiation monitoring is divided into two subsystems.
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| One subsystem (pre-treatment) takes a continuous sample from the offgas line prior to the delay and adsorption treatment process. The other subsystem (post-treatment) takes a continuous sample from the offgas line just before discharge to the main stack. [Reference 3-6, Section 9.4.4.1.2]
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| 3.2.3.2.2 Turbine Sealing and Mechanical Vacuum Pump Systems The gland seal holdup system collects and processes, by delay, the noncondensable exhaust from the main turbine gland seal condenser.
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| During startup operation the discharge of the condenser mechanical vacuum pump is routed through the gland seal holdup system. The effluent of the gland seal holdup system is routed to the main station stack where it is continuously monitored by the main stack radiation monitoring system before discharge to the environment.
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| [Reference 3-6, Section 9.4.4.2.1]
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| During normal operation of the gland seal holdup system, a 2,200 lb/hr saturated air-water vapor ae; mixture containing trace amounts of hydrogen, oxygen, and radioactive gases is exhausted from the turbine generator gland seal condenser and enters the 16-inch diameter holdup line. After being delayed for a period of approximately 1.75 minute, the effluent is routed to the main stack where it is mixed with the AOG system effluent and the discharge of the main stack dilution fans before release to the environment.
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| [Reference 3-6, Section 9.4.4.2.1]
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| The gland seal holdup system shares with the AOG system the main stack, dilution fans, and the main stack radiation monitoring system. During normal operation, the amount of radioactive activation and fission gases associated with the gland seal holdup system is extremely small.The radioactivity that is collected and processed by the gland seal holdup system is proportional to the amount of main steam utilized in the main turbine sealing system. This amount of steam is less than 0.1% of the full power rated steam flow. In addition to the small amount of radioactivity processed, there is a correspondingly small amount of radiolytic hydrogen and oxygen which are well below the explosive limits. [Reference 3-6, Section 9.4.4.2.2]
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| 3.2.3.2.3 Miscellaneous Gaseous Effluents (Low Release Potential Effluents)
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| Miscellaneous gaseous effluents are categorized into two classes, those from areas having a negligible or low potential for the release of airborne radioactivity, and those from areas likely to experience radioactive contamination.
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| Below is a list of station areas which fall into these categories and which are exhausted directly to the environment.
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| [Reference 3-6, Section 9.4.4.3.1]
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| * diesel generator building 3-8 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* administration building* machine shop* battery room and lube oil compartments
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| * recirculation pump MG set area* reactor auxiliary bay* turbine building operating floor and switchgear area The ventilation air from the first six areas listed above has a negligible potential for the release of radioactive effluents.
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| The turbine building operating floor including the reactor feedwater pump area are considered to have a low potential for release. Any release from the turbine building basement area or the turbine building ground floor to the turbine building operating floor or adjacent areas above elevation 51 feet is precluded since the turbine building basement and ground floor are maintained at a slight negative pressure relative to the turbine building operating floor. [Reference 3-6, Section 9.4.4.3.1]
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| The airborne radiation concentration levels at elevation 51 feet in the turbine building are routinely-monitored by means of the turbine building effluent monitoring system. Airborne activity levels in those areas of the station having a direct release path to the environs not monitored by a process radiation monitoring system will under normal operating conditions be within those levels allowed for in 10 CFR 20, Appendix B, Table I. [Reference 3-6, Section 9.4.4.3.1]
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| The expected airborne activity on the turbine building operating floor will normally be below the values assumed above and the releases from the turbine building operating floor and the reactor (feedwater pump area are expected to be insignificant relative to the releases from the main stack and the reactor building exhaust vent. [Reference 3-6, Section 9.4.4.3.1 3.2.3.2.4 Miscellaneous Gaseous Effluents Gaseous effluents from areas of potential radioactive contamination are monitored and discharged to the environment through either the main stack or the reactor building exhaust vent.The station ventilation systems are designed to combine the ventilation air flow from these areas and exhaust that air past process radiation monitoring equipment.
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| [Reference 3-6, Section 9.4.4.3.2]
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| Miscellaneous sources of potential low-level radioactive airborne contaminants in the station which could be released to the environment are listed below [Reference 3-6, Section 9.4.4.3.1].
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| * primary containment venting* steam leakage outside the primary containment
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| * hood vents* high pressure coolant injection (HPCI) testing PNPS maintains gaseous releases within ODCM limits. The gaseous radwaste system is used to reduce radioactive materials in gaseous effluents before discharge to meet the dose design objectives in 10 CFR 50, Appendix I. In addition, the limits in the ODCM are designed to provide reasonable assurance that radioactive material discharged in gaseous effluents would not result 3-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage in the exposure of a member of the public in an unrestricted area in excess of the limits specified in 10 CFR 20, Appendix B.The quantities of gaseous effluents released from PNPS are controlled by the administrative limits defined in the ODCM. The controls are specified for dose rate, dose due to noble gases, and dose due to radioiodine and radionuclides in particulate form. For noble gases, the dose rate limit at and beyond the site boundary is < 500 mrem/yr to the total body, and < 3000 mrem/yr to the skin [Reference 3-5, Section 3.3.1]. For lodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days, the limit is !1500 mrem/yr to any organ [Reference 3-5, Section 3.3.1]. The limit for air dose due to noble gases released in gaseous effluents to areas at and beyond the site boundary during a calendar quarter is < 5 mrad for gamma radiation and <10 mrad for beta radiation
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| [Reference 3-5, Section 3.3.2]. For a calendar year, the limit is <10 mrad for gamma radiation and < 20 mrad for beta radiation[Reference 3-5, Section 3.3.2]. The radioactive gaseous waste sampling and analysis program specifications provided in the ODCM address the gaseous release type, sampling frequency, minimum analysis frequency, type of activity analysis, and lower limit of detection.
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| 3.2.3.3 Solid Waste Processing The solid waste processing areas are located in the radwaste building, the radwaste truck lock, and the trash compaction facility (TCF). Both wet and dry solid wastes are processed.
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| Wet solid wastes include backwash sludge wastes from the reactor water cleanup system (RWCU); all spent resins and charcoal from radwaste, spent fuel pool, and condensate demineralizers; and thermex and radwaste filter/demineralizer.
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| [Reference 3-6, Section 9.3.4.1]Dry solid wastes include rags, paper, small equipment parts, solid laboratory wastes, etc.[Reference 3-6, Section 9.3.4.1]An outdoor low level radwaste storage facility (LLRWSF) is provided on-site for interim storage for up to 5 years of solid radioactive waste prior to disposal off-site or for temporary storage of bulk-dewatered radwaste awaiting shipment to a processing facility for volume reduction prior to burial. The LLRWSF consists of a compacted gravel bed surrounded by a gravel or earth filled modular block shield wall. Dewatered solid wastes contained in high integrity containers are placed in cylindrical, concrete storage modules within the facility.
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| Dry activated waste in steel containers and overpack, as well as other miscellaneous low-level radioactive materials, is also stored in the LLRWSF in rectangular, concrete storage modules. [Reference 3-6, Section 9.3.4.1]3.2.3.3.1 Reactor Cleanup Sludge The purpose of the radwaste system for cleanup sludge is to process the-highly radioactive backwash waste which is discharged from the RWCU system. The RWCU system includes two filter-demineralizer units each of which are precoated with powdered ion exchange resin (Powdex) supported by filter aid which is in turn retained on a permanent, stainless steel septum.These filter-demineralizer units remove by filtration and ion exchange the suspended and dissolved solids, both radioactive and stable, from the circulating reactor water. Upon 3-10 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage exhaustion of either its filtration or ion exchange capability, the exhausted cleanup demineralizer is taken out of service, backwashed, and precoated anew. The backwash waste as discharged from a cleanup demineralizer is a relatively dilute slurry (1.1% by weight suspended solids) which is highly radioactive.
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| The backwash waste slurry is accumulated in the backwash collector tank from which it is periodically transferred on a batch basis to the radwaste disposal system for subsequent processing.
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| The function of the radwaste disposal system is to reclaim the liquid phase for reuse within the station and to prepare the solid waste for offsite shipment with minimum exposure of the operators to radiation.
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| [Reference 3-6, Section 9.3.4.2.1]
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| The radwaste disposal system has been modified.
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| A sludge transfer and decant line has been provided for the cleanup sludge storage tanks. The transfer line is used to transfer sludge to the offsite discharge pipe in the radwaste trucklock.
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| This arrangement bypasses the floc-recycle tank (abandoned).
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| The sludge is dewatered in the radwaste trucklock before being stored to await shipment to a burial processor facility or for other processing.
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| A decant line has been installed between the sludge transfer pumps discharge and the clean waste tanks inlet piping.[Reference 3-6, Section 9.3.4.2.1]
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| 3.2.3.3.2 Spent Resin and Miscellaneous Solid Waste System The purpose of the spent resin and miscellaneous solid'waste systems is to process and temporarily store spent resins and miscellaneous solid waste (rags, used clothing, paper, air falters, etc.) on the site in shielded areas as required prior to offsite shipment to a licensed burial ground or other processing facility.
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| [Reference 3-6, Section 9.3.4.3.1]
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| All spent resins from radwaste,'
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| spent fuel pool, thermex and condensate demineralizers are sluiced into a spent resin tank which provides 670 ft 3 capacity.
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| Thermex waste water and miscellaneous waste waters may be added to the tank to utilize remaining capacity of spent resin and allow for reprocessing.
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| This may be done to reduce solid radwaste volume and overboard discharge of contaminated waste water. [Reference 3-6, Section 9.3.4.3.2]
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| When spent resins accumulate in the spent resin tank to the amount desired for offsite shipment, the spent resins will be pumped from the tank into a processing/shipping container or HIC (High Integrity Container) as required for dewatering and for shipment and offsite disposal/processing.
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| A backflushing system for tank overflow and spent resin retention screens is provided to eliminate or reduce screen plugging with resin fines as much as possible.
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| Sluice water is recycled back to the spent resin tank. [Reference 3-6, Section 9.3.4.3.2]
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| The contaminated miscellaneous solid wastes, such as air filters, rags, paper, small equipment parts, and solid laboratory wastes, are placed in disposable containers and shipped for processing or disposal.
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| Compressed solid wastes in the disposable containers are stored temporarily on the site for future offsite shipment.
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| [Reference 3-6, Section 9.3.4.3.2]
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| The clean radwaste effluent is processed through various processing equipment resulting in spent resintpowdered resin (sludge) which is loaded into containers for-shipment to an offsite radioactive waste minimization process facility or shipped for burial.3-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 3.2.3.3.3 Trash Compaction Facility The original purpose of the TCF was to sort, process, and separate contaminated and non-contaminated material generated from normal operating conditions.
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| This process of separating the contaminated materials from the non-contaminated materials has been discontinued and the current use for the TCF is for storage of contaminated equipment which is used within the plant.3.2.3.3.3.1 Contaminated Material Contaminated materials are now stored in one of two locations before they are shipped for disposal.
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| Contaminated dry active waste, metal, and wood are separated and are temporarily stored in either the LLRWSF or the TCF yard in seavans until they are shipped off-site to a radwaste processor.
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| The compactible radioactive material which will be compacted is transported to the contaminated trash compactor, placed within the compactor, and compacted.
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| The resulting product, which is contained within a steel box specifically designed for handling compacted trash, is transported via forklift truck to the labeling, weighing, and surveying area. [Reference 3-6, Section 9.5.1.6.11 Radioactive liquid material is segregated, separated, consolidated, and analyzed for disposal in the TCF hazardous material area. Based on analysis results the material is packaged, labeled, and marked for transport to offsite burial, further processing, or interim storage. [Reference 3-6, Section 9.5.1.6.1]
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| 3.2.3.3.3.2 Noncontaminated Material Material identified as hazardous material is transported to the TCF hazardous material area and surveyed to determine what material is contaminated or not contaminated by predetermined radiological standards.
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| Contaminated hazardous material is segregated and labeled. The non-contaminated hazardous material is accumulated and stored in the 90-day hazardous waste storage area until sufficient quantity is available for disposal, but must be disposed of within 90 days. [Reference 3-6, Section 9.5.1.6.2]
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| 3.2.3.3.4 Decontamination and Trash and Laundry Processing Facility The decontamination and trash and laundry processing facility is located in the north side of the station services redline building.
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| As a facility to support station operation, it contains equipment for decontamination tools and equipment and also working space for handling trash, metals, wood, and potential HAZMAT being transferred to the TCF. This facility also handles incoming and outgoing shipments of laundry and contains space to permit temporary storage of various dry materials and equipment.
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| [Reference 3-6, Section 9.5.2]Hazardous material (other than radioactive material), liquids containing radioactive material, or wastes from plant water treatment processes (e.g., spent resin, sludge, and diatomaceous earth)are not stored in the facility.
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| [Reference 3-6, Section 9.5.2]3-12 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Both administrative and physical controls are in place to maintain radiation exposure to personnel ALARA and to preclude releases to the environment in excess of the limits set forth in 10 CFR 20. [Reference 3-6, Section 9.5.2]3.2.4 Transportation of Radioactive Materials PNPS radioactive waste shipments are packaged in accordance with NRC and U.S. Department of Transportation requirements.
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| The type and quantities of solid radioactive waste generated and shipped at PNPS vary from year to year, depending on plant activities.
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| PNPS currently transports radioactive waste to the Studsvic facility in Irwin, Tennessee, Race facility in Memphis, Tennessee, or the Duratek facility in Oak Ridge, Tennessee, where the wastes are further processed prior to being sent to the Barnwell facility in Barnwell County, South Carolina, or the Envirocare facility in Clive, Utah. On occasion PNPS may also transport material back to the plant site for reuse or storage.3.2.5 Nonradioactive Waste Systems Nonradioactive waste is produced from plant maintenance and cleaning processes.
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| Most of these wastes are from heating boiler blowdown, filter backwash, sludges and other wastes, floor and yard drains, and stormwater runoff. Chemical and biocide wastes are produced from processes used to control the pH in the coolant,,to control scale, to control corrosion, and to clean and defoul the main condenser.
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| Waste liquids are typically combined with cooling water discharges.
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| Sanitary wastewater, which is regulated under Groundwater Discharge Permit #2-389 issued from the MDEP, is directed to an onsite septic system where it is transferred to an onsite wastewater treatment facility and ultimately discharged to a leach field.Non-radioactive gaseous effluents result from operation of the oil-fired boilers used to heat the plant and from testing of the emergency diesel generators.
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| Discharge of regulated pollutants is minimized by limiting fuel usage and hours of operation and is within the MDEP's air quality standards.
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| 3.2.6 Maintenance, Inspection, and Refueling Activities Various programs and activities currently exist at PNPS to maintain, inspect, test, and monitor the performance of plant equipment.
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| These programs and activities include, but are not limited, to those implemented to* meet the requirements of 10 CFR 50, Appendix B (Quality Assurance), Appendix R (Fire Protection), and Appendices G and H, Reactor Vessel Materials;
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| * meet the requirements of 10 CFR 50.55a, ASME Code, Section Xl, In-service Inspection and Testing requirements;
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| * meet the requirements of 10 CFR 50.65, the maintenance rule, including the structures monitoring program; and* maintain water chemistry in accordance with EPRI guidelines.
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| 3-13 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Additional programs include those implemented to meetlTechnical Specification surveillance requirements, those implemented in response to NRC generic communications, and various periodic maintenance, testing, and inspection procedures.
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| Certain program activities are performed during the operation of the unit. Others are performed during scheduled refueling outages.3.2.7 Transmission Facilities The FES [Reference 3-1] identifies two transmission lines that were built to connect PNPS to the electric grid. The 342 line runs approximately 5 miles to the Jordan Road Tap, which connects to the Canal and the Auburn Street Stations via a previously existing line. The 355 line runs on the same towers as the 342 line to the Jordan Road Tap and then beyond for a total of 7.2 miles to the Snake Hill Road Tap, where previously existing lines run to the Bridgewater Station.Therefore, the segments of interest for this report are from PNPS to the Jordan Road Tap for line 342 and from PNPS to the Snake Hill Road Tap for the 355 line. Both lines operate at 345 kv.The transmission corridor is 300 feet wide. Figure 2-2 shows the transmission system of interest.NSTAR, the current owner and operator of the transmission lines, has approximately 12.2 miles of transmission lines (7.2 miles of corridor) that occupy approximately 260 acres which connect PNPS to the transmission system, in addition to carrying power from other generators.
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| The corridors pass through rolling land that is primarily forested.
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| The major road crossing is Massachusetts Highway Route 3.The transmission lines were designed and constructed in the late 1960s and early 1970s, in accordance with the National Electrical Safety Code (NESC) and industry guidance that was current when the lines were built. Ongoing right-of-way surveillance and maintenance of the transmission facilities ensure continued conformance to design standards.
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| These maintenance practices are described in Section 2.4 and Section 4.13.3.3 Refurbishment Activities 10 CFR 51.53(c)(2) requires that a license renewal applicant's environmental report contain a description of the proposed action, including the applicant's plans to modify the facility or its administrative control procedures as described in accordance with Section 54.21 of this chapter. This report must describe in detail the modifications directly affecting the environment or affecting plant effluents that affect the environment.
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| The objective of the review required by 10 CFR 54.21 is to determine whether the detrimental effects of plant aging could preclude certain PNPS systems, structures, and components (SSCs)from performing in accordance with the current licensing basis, during the additional 20 years of operation requested in the license renewal application.
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| The evaluation of SSCs as required by 10 CFR 54.21 has been completed and is described in the body of the PNPS license renewal application.
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| This evaluation did not identify the need for refurbishment of SSCs related to license renewal.3-14 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Routine replacement of certain components during the period of extended operation is expected to occur within the bounds of normal plant maintenance.
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| There are no plans associated with license renewal to modify the facility or its administrative control procedures other than those procedures necessary to implement the aging management programs described in the Integrated Plant Assessment.
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| The proposed action does not include any modifications directly affecting plant effluents or the environment.
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| Modifications to improve operation of plant SSCs are reviewed for environmental impact by station personnel during the planning stage for the modification.
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| These reviews are controlled by site procedures.
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| 3.4 Programs and Activities for Managing the Effects of Aging The programs for managing aging of systems and equipment at PNPS are described in the body of the PNPS license renewal application.
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| The evaluation of SSCs required by 10 CFR 54.21 identified some new inspection activities necessary to continue operation of PNPS during the additional 20 years beyond the initial license term. These activities are described in the body of the PNPS license renewal application.
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| The additional inspection activities are consistent with normal plant component inspections, and therefore, are not expected to cause significant environmental impact. The majority of the aging management programs are existing programs or modest modifications of existing programs.3.5 Employment As of February 2005, the non-outage work force at PNPS consists of approximately 703 persons.There are 574 Entergy employees normally on site or at the offsite training facilities.
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| The remaining 129 persons are baseline contractor employees.
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| Table 3-1 shows employee and baseline contractor residences by state, county, and city. The GEIS estimated that an additional 60 employees would be necessary for operation during the period of extended operation.
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| Since there will not be any significant new aging management programs added at PNPS for license renewal, Entergy believes that it will be able to manage the necessary programs with existing staff. Therefore, Entergy has no plans to add non-outage employees to support plant operations during the extended license period.Refueling and maintenance outages typically last approximately 30 days. Depending on the scope of these outages, an additional 700-900 workers are typically on site. The number of workers required on site for normal plant outages during the period of extended operation is expected to be consistent with the number of additional workers used for past outages at PNPS.3-15 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 3-1 Employee Residence Information, PNPS, February 2005 County, State, and City Employees (Entergy and Baseline Contractors)
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| BARNSTABLE COUNTY (MASSACHUSETTS) 137 Barnstable 21 Boume 25 Brewster 1 Chatham 1 Dennis 6 Falmouth 9 Harwich 4 Mashpee 13 Sandwich 53 Yarmouth 4 BRISTOL COUNTY (MASSACHUSETTS) 43 Acushnet 3 Attleboro 2 Dartmouth 3 Easton I Fairhaven 1 Freetown 2 Mansfield 1 New Bedford 12 Norton I Raynham 4 Rehoboth I Seekonk I Swansea 1 3-16 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 3-1 Employee Residence Information, PNPS, February 2005 (Continued)
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| County, State, and City Employees (Entergy and Baseline Contractors)
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| Taunton 9 Westport 1 MIDDLESEX COUNTY (MASSACHUSETTS) 6 Ashland 1 Burlington 1 Chelmsford 2 Everett 1 Framingham I NORFOLK COUNTY (MASSACHUSETTS) 57 Avon 1 Braintree 5 Canton 2 Dedham 1 Franklin 2 Holbrook I Medfield 1 Milton 1 Needham 1 Norwood 2 Plainville I Quincy 8 Randolph 2 Sharon 5 Stoughton 1 Westwood 1 3-17 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 3-1 Employee Residence Information, PNPS, February 2005 (Continued)
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| County, State, and City Employees.(Entergy and Baseline Contractors)
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| Weymouth 21 Wrentham I PLYMOUTH COUNTY (MASSACHUSETTS)
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| Abington Bridgewater Brockton Carver Duxbury East Bridgewater Halifax Hanover Hanson Hingham Kingston Lakeville Marion Marshfield Middleboro Norwell Pembroke Plymouth Plympton Rochester Rockland Scituate 444 3, 9 5 25 19 5 10 9 5 7 21 2-1 27 13 3 18 223 2 8 3 6 3-18 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (tm-)Table 3-1 Employee Residence Information, PNPS, February 2005 (Continued)
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| County, State, and City Employees (Entergy and Baseline Contractors)
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| Wareham 14 West Bridgewater 1 Whitman 5 SUFFOLK COUNTY (MASSACHUSETTS) 6 Boston 6 WORCESTER COUNTY (MASSACHUSETTS) 3 Milford 1 Shrewsbury 1 Upton 1 PROVIDENCE COUNTY (RHODE ISLAND) 3 Cranston 1 Cumberland 1 North Smithfield 1 NEW LONDON COUNTY (CONNECTICUT)
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| I Griswold 1 MANATEE COUNTY (FLORIDA) i Bradenton 1 CHESIRE COUNTY (NEW HAMPSHIRE)
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| I Westmoreland 1 OSWEGO COUNTY (NEW YORK) 1 Minetto 1 TOTAL EMPLOYEES
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| = 703 3-19 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 3.6 References 3-1 U.S. Atomic Energy Commission, Division of Radiological and Environmental Protection, Final Environmental Statement Related to Operation of Pilgrim Nuclear Power Station, Docket No. 50 293, Washington, DC, 1972.3-2 ENSR Corporations, Redacted Version 316 Demonstration Report -Pilgrim Nuclear Power Station, Document Number 0970-021-200, prepared for Entergy Nuclear Generation Company, Plymouth, MA, March 2000.3-3 U.S. Environmental Protection Agency, Water Management Division Region 1,"Modification of Authorization to Discharge Under the National Pollutant Discharge Elimination System, Federal Permit No. MA0003557, Modification No. 1," Boston, MA, August 30, 1994.34 Massachusetts Department of Environmental Protection, Executive Office of Environmental Affairs, Southeast Regional Office, Groundwater Discharge Permit, SE#2-389, Pilgrim Power Station Wastewater Treatment Facility, Lakeville, MA, April 26, 1999.3-5 Pilgrim Nuclear Power Station, Pilgrim Nuclear Power Station Offsite Dose Calculation Manual, Plymouth, MA, October 6, 2003.3-6 Pilgrim Nuclear Power Station, Updated Final Safety Analysis Report, Plymouth, MA.1. Pilgrim's UFSAR update is done on a page-by-page basis, rather than by entire section or volume. Therefore, several different revisions (up to Revision 24) of the UFSAR update have been used in this ER.3-20 tco-'--Q 02w C; -n& r, C< I 0 -A_0 >10 m-v (D 'a:-( =(a3 -=CD~ m 0 5CD CD --0;a 0 0 C q C Pilgrim Nuclear Power Station Applicant's Environmental Report i; Operating License Renewal Stage 4.0 ENVIRONMENTAL CONSEQUENCES OF THE PROPOSED ACTION Discussion of GEIS Cateaories for Environmental Issues The NRC has identified and analyzed 92 environmental issues that it considers to be associated with nuclear power plant license renewal and has designated the issues as Category 1, Category 2, or NA (not applicable).
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| The NRC designated an issue as Category 1 if, based on the result of its analysis, the following criteria were met: (1) the environmental impacts associated with the issue have been determined to apply either to all plants or, for some issues, to plants having a specific type of cooling system or other specified plant or site characteristic; (2) a single significance level (i.e., small, moderate, or large) has been assigned to the impacts that would occur at any plant, regardless of which plant is being evaluated (except for collective offsite radiological impacts from the fuel cycle and from high-level waste and spent-fuel disposal);
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| and (3) mitigation of adverse impacts associated with the issue has been considered in the analysis, and it has been determined that additional plant-specific mitigation measures are likely to be not sufficiently beneficial to warrant implementation.
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| -If the NRC concluded that one or more of the Category 1 criteria could not be met, the NRC designated the issue Category 2. The NRC requires plant-specific analysis for Category 2 issues. The NRC designated two issues as NA, signifying that the categorization and impact definitions do not apply to these issues. NRC rules do not require analyses of Category 1 issues that the NRC resolved using generic findings (10 CFR 51, Subpart A, Appendix B, Table B-1) as described in the GEIS [Reference 4-5]. An applicant may reference the generic findings or GEIS analyses for Category 1 issues.Category I License Renewal Issues Entergy has determined that, of the 69 Category 1 issues, 13 are not applicable to PNPS because they apply to design or operational features that do "not exist at the facility.
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| In addition, because Entergy does not plan to conduct any refurbishment activities, the NRC findings for the seven Category 1 issues that are applicable to refurbishment do not apply. Table 4-1 lists these 20 issues and provides a brief explanation of why they are not applicable to PNPS.: Table 4-2 lists the 49 Category 1 issues that Entergy has determined to be applicable to PNPS. :Entergy has not identified any new and significant information concerning the impacts addressed by these findings.
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| Therefore,:
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| Entergy adopts by reference the NRC findings for these Category 1 issues.4-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-1.Category I Issues Not Applicable to PNPS Surface Water Quality, Hydrology, and Use (for All Plants)Impacts of refurbishment on surface water quality No refurbishment activities planned.Impacts of refurbishment on surface water use No refurbishment activities planned.Altered thermal stratification of lakes PNPS is not located on a lake.Eutrophication PNPS is not located on a lake.Aquatic Ecology (for All Plants)Refurbishment
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| -No refurbishment activities planned.Aquatic Ecology (for plants with cooling-tower based heat dissipation systems)Entrainment of fish and shellfish in early life stages PNPS does not use cooling towers.Impingement of fish and shellfish PNPS does not use cooling towers.Heat shock PNPS does not use cooling towers.Ground-water Use and Quality Impacts of refurbishment on ground-water use and No refurbishment activities planned.quality Groundwater use conflicts (potable and service water; PNPS does not use groundwater for potable plants that use <100 gpm) and service water.Ground-water quality degradation (Ranney Wells) PNPS does not use Ranney wells.Ground-water quality degradation (cooling ponds in salt PNPS does not use cooling ponds.marshes)Ground-water quality degradation (saltwater intrusion)
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| PNPS does not use groundwater for any purpose.Human Health Radiation exposures to the public during refurbishment No refurbishment activities planned.Occupational radiation exposures during refurbishment No refurbishment activities planned.Terrestrial Resources Cooling tower impacts on crops and ornamental PNPS does not use cooling towers.vegetation Cooling tower impacts on native plants PNPS does not use cooling towers.Cooling pond impacts on terrestrial resources PNPS does not use cooling ponds.Bird collisions with cooling towers PNPS does not use cooling towers.Socloeconomics Aesthetic impacts (refurbishment)
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| No refurbishment activities planned.4-2 Q' I Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-2 Category 1 Issues Applicable to PNPS Surface Water Quality, Hydrology, and Use (for All Plants)Water use conflicts (plants with once-through cooling systems)Altered current patterns at intake and discharge structures Altered salinity gradients Temperature effects on sediment transport capacity Scouring caused by discharged cooling water Discharge of chlorine or other biocides Discharge of sanitary wastes and minor chemical spills Discharge of other metals in waste water Aquatic Ecology (for All Plants)Accumulation of contaminants in sediments or biota Entrainment of phytoplankton and zooplankton Cold shock Thermal plume barrier to migrating fish Distribution of aquatic organisms Premature emergence of aquatic insects Gas supersaturation (gas bubble disease)Low dissolved oxygen In the discharge Losses from predation, parasitism, and disease among organisms exposed to sublethal stresses Stimulation of nuisance organisms (e.g., shipworms)
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| Terrestrial Resources Power line right-of-way management (cutting and herbicide application)
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| Bird collision with power lines Impacts of electromagnetic fields on flora and fauna (plants, agricultural crops, honeybees, wildlife, livestock)
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| Floodplains and wetland on power line right of way Air Quality Air quality effects of transmission lines 4-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-2 Category I Issues Applicable to PNPS (Continued)4 Land Use Onsite land use (license renewal period)Power line right of way Human Health Noise Radiation exposures to public (license renewal term)Occupational radiation exposures (license renewal term)Socioeconomics Public services:
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| public safety, social services, and tourism and recreation Public services, education (license renewal term)Aesthetic impacts (license renewal term)Aesthetic impacts of transmission lines (license renewal term)Postulated Accidents Design basis accidents Uranium Fuel Cycle and Waste Management Offsite radiological impacts (individual effects from other than the disposal of spent fuel and high level waste)Offsite radiological impacts (collective effects)Offsite radiological impacts (spent fuel and high level waste disposal)Non-radiological impacts of the uranium fuel cycle Low-level waste storage and disposal Mixed waste storage and disposal On-site spent fuel Nonradiological waste Transportation 4-4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-2 Category I Issues Applicable to PNPS (Continued)
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| Decommissioning Radiation doses Waste management Air quality Water quality Ecological resources Socioeconomic impacts Cateaory 2 License Renewal Issues The NRC designated 21 issues as Category 2. Sections 4.1 through 4.21 address each of the Category 2 issues, beginning with a statement of the issue. As is the case with Category 1 issues, some Category 2 issues (6) apply to operational features that PNPS does not have. In addition, some Category 2 issues (4) apply only to refurbishment activities.
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| If the issue does not apply to PNPS, the section explains the basis.For the 11 Category 2 issues applicable to PNPS, the corresponding section contains the required analyses.
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| These analyses include conclusions regarding the significance of the impacts relative to the renewal of the operating license for PNPS and, when applicable, discuss potential mitigative alternatives to the extent required.
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| Entergy has identified the significance of the impacts associated with each issue as SMALL, MODERATE, or LARGE, consistent with the criteria that the NRC established in 10 CFR 51, Appendix B, Table B-1, Footnote 3 as follows.* SMALL -Environmental effects are not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource.
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| For the purposes of assessing radiological impacts, the Commission has concluded that those impacts that do not exceed permissible levels in the Commission's regulations are considered small.* MODERATE -Environmental effects are sufficient to alter noticeably, but not to destabilize, any important attributes of the resource.* LARGE -Environmental effects are clearly noticeable and are sufficient to destabilize any important attributes of the resource.In accordance with NEPA practice, Entergy considered ongoing and potential additional mitigation in proportion to the significance of the impact to be addressed (i.e., Impacts that are small receive less mitigative consideration than impacts that are large).4-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage"NA" License Renewal Issues The NRC determined that its categorization and impact-finding definitions did not apply to electromagnetic fields (chronic effect) and environmental justice. The NRC noted that applicants currently do not need to submit information on chronic effects from electromagnetic fields (10 CFR 51, Appendix B, Table B-1, Footnote 5). For environmental justice, the NRC does not require information from applicants, but noted that it will be addressed in individual license renewal reviews (10 CFR 51, Appendix B, Table B-1, Footnote 6). Entergy has included environmental justice demographic information in Section 2.6.2.Format of Category 2 Issue Review The review and analysis for the Category 2 issues and environmental justice are found in Sections 4.1 through 4.22. The format for the review of the Category 2 issues is described below.* Issue -a brief statement of the issue.* Description of Issue -a brief description of the issue.* Findings from Table B-1, Appendix B to Subpart A -findings for the issue from Table B-1, Summary of Findings on NEPA Issues for License Renewal of Nuclear Power Plants, Appendix B to Subpart A.* Requirement-the requirement from 10 CFR 51.53(c)(3)(ii) is restated.* Background
| |
| -for issues applicable to PNPS, a background excerpt from the applicable section of the GEIS is provided.
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| The specific section of the GEIS is referenced for the convenience of the reader. In most cases, background information is not provided for issues that are not applicable to PNPS.-Analysis of Environmental Impact -an analysis of the environmental impact as required by 10 CFR 51.53(c)(3)(ii) is provided, taking into account information provided in the GEIS, Appendix B to Subpart A of 10 CFR 51, as well as current PNPS specific information.
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| * Conclusion
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| -for issues applicable to PNPS, the conclusion of the analysis is presented along with the consideration of mitigation alternatives as required by 10 CFR 51.45(c)and 10 CFR 51.53(c)(3)(iii).
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| 4.1 Water Use Conflicts 4.1.1 Description of Issue Water use conflicts (plants with cooling ponds or cooling towers using make-up water from a small river with low flow)4-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.1.2 Findings from Table B-1, Appendix B to Subpart A SMALL or MODERATE.
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| The issue has been a concern at nuclear power plants with cooling ponds and at plants with cooling towers. Impacts on instream and riparian communities near these plants could be of moderate significance in some situations.
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| See 10 CFR 51 .53(c)(3)(ii)(A).
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| 4.1.3 Requirement
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| [10 CFR 51.53(c)(3)(kk)(A)
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| If the applicant's plant utilizes cooling towers or cooling ponds and withdraws make-up water from a river whose annual flow rate is less than 3.15x1 012 ft 3/year (9x101 0 m 3/year), an assessment of the impact of the proposed action on the flow of the river and related impacts on instream and riparian ecological communities must be provided.
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| The applicant shall also provide an assessment of the impacts of the withdrawal of water from the river on alluvial aquifers during low flow.4.1.4 Analysis of Environmental Impact The issue of surface water use conflicts does not apply to PNPS as the plant does not use cooling towers, cooling ponds, or withdraw water from a small river. As Section 3.2.2.1 describes, PNPS uses a once-through cooling system that withdraws water from Cape Cod Bay.4.2 Entrainment of Fish and Shellfish in Early Life Stages 4.2.1 Description of Issue Entrainment of fish and shellfish in early life stages (for all plants with once-through and cooling pond heat dissipation systems).4.2.2 Findings from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. The impacts of entrainment are small at many plants but may be moderate or even large at a few plants with once-through and cooling-pond cooling systems.Further, ongoing efforts in the vicinity of these plants to restore fish populations may increase the numbers of fish susceptible to intake effects during the license renewal period, such that entrainment studies conducted in support of the original license may no longer be valid. See 10 CFR 51.53(c)(3)(ii)(B).
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| 4.2.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| If the applicant's plant utilizes once-through cooling or cooling pond heat dissipation systems, the applicant shall provide a copy of current Clean Water Act 316(b) determinations and, if necessary, a 316(a) variance in accordance with 40 CFR Part 125, or equivalent state permits and supporting documentation.
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| If the applicant cannot provide these documents, it shall assess the impact of the proposed action on fish and shellfish resources resulting from heat shock and impingement and entrainment.
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| 4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.2.4 Background The effects of entrainment on aquatic resources were considered by the NRC at the time of original licensing and are periodically reconsidered by EPA or state water quality permitting agencies in the development of National Pollutant Discharge Elimination System (NPDES)permits and 316(b) demonstrations.
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| The impacts of fish and shellfish entrainment are small at many plants, but they may be moderate or even large at a few plants with once-through cooling systems. Further, ongoing restoration efforts may increase the numbers of fish susceptible to intake effects during the license renewal period, so that entrainment studies conducted in support of the original license may no longer be valid [Reference 4-5, Section 4.2.2.1.2].
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| 4.2.5 Analysis of Environmental Impact As Section 3.2.2.1 describes, PNPS has a once-through heat dissipation system that uses water from Cape Cod Bay for condenser cooling.Section 316(b) of the Clean Water Act (CWA) requires that any standard established pursuant to Sections 301 or 306 of the CWA shall require that the location, design, construction, and capacity of cooling water intake structures reflect the best technology available for minimizing adverse environmental impacts (33 USC 1326). Entrainment through the condenser cooling system of fish and shellfish in early life stages is a potential adverse environmental impact that can be minimized by the best technology available.
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| The EPA Region I is the NPDES permitting authority for Massachusetts.
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| The current PNPS (,_NPDES permit (Federal Permit No. MA0003557) notes the following:
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| It has been determined based on engineering judgment that the circulating water intake structures
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| [sic] presently employs the best technology available for minimizing adverse environmental impact. Any change in the location, design, or capacity of the present structure shall be approved by the Regional Administrator and the Director.
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| The present design shall be reviewed for conformity to the regulations pursuant to Section 316(b) of the Act when such are promulgated.
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| [Reference 4-3]Thus the PNPS NPDES permit, issued August 30, 1994, by EPA Region I, constitutes the current CWA Section 316(b) determination for PNPS. Attachment A contains portions of the permit, including the quoted Section A.1.i.EPA Region I is requiring all NPDES permittees in the region (to whom CWA Section 316 applies) to submit new Sections 316(a) and 316(b) demonstrations.
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| EPA Region I is reviewing an Entergy application for renewal of the PNPS NPDES Permit and, as described in Section 2.2, a new combined Section 316 report that evaluates more than 25 years of entrainment and impingement data [Reference 4-2]. This new Section 316 demonstration report concludes that the PNPS cooling water intake system has not resulted in adverse impacts to the integrity of Cape Cod Bay fish and shellfish populations, including a number of Representative Important 4-8 Pilgrim Nuclear Power Station Applicant's Environmental Report iiW. Operating License Renewal Stage Species (e.g., American lobster, winter flounder, rainbow smelt, cunner, alewife, and Atlantic silverside).
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| On July 9, 2004, the EPA published a final rule in the Federal Register (69 FR 41575) [Reference 4-12J addressing cooling water intake structures at existing power plants, such as PNPS. The rule is Phase II in the EPA's development of 316(b) regulations that establish national requirements applicable to the location, design, construction, and capacity of cooling water intake structures at existing facilities.
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| The national requirements, which are implemented through NPDES permits, provide several compliance alternatives that may be pursued by facilities to meet the entrainment and impingement performance standards in the Rule. Any additional mitigation measures under the new regulations would only further reduce the already small impacts.4.2.6 Conclusion EPA Region I has determined based on engineering judgment that the circulating water intake structure presently employs the best technology available for minimizing adverse environmental impact. Because Entergy submitted a timely application for renewal of the PNPS NPDES Permit, the 1994 permit and its Section 316(b) determination remain in effect. For this reason, Entergy concludes that PNPS impacts due to entrainment of fish and shellfish are SMALL and do not warrant mitigation beyond those measures required by the NPDES permit, as periodically amended.7 4.3 Impingement of Fish and Shellfish 4.3.1 Description of Issue Impingement of fish and shellfish (for all plants with once-through and cooling pond heat dissipation systems)4.3.2 Findings from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. The impacts of impingement are small at many plants, but may be moderate or even large at a few plants with once-through and cooling-pond cooling systems. See 10 CFR 51 .53(c)(3)(ii)(B).
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| 4.3.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| If the applicant's plant utilizes once-through cooling or cooling pond heat dissipation systems, the applicant shall provide a copy of current Clean Water Act 316(b) determinations and, if necessary, a 316(a) variance in accordance with 40 CFR Part 125, or equivalent state permits and supporting documentation.
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| If the applicant cannot provide these documents, it shall assess the impact of the proposed action on fish and shellfish resources resulting from heat shock and'impingement and entrainment.
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| 4-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.3.4 Background Aquatic organisms that are drawn into the intake with the cooling water and are too large to pass through the debris screens may be impinged against the screens. Mortality of fish that are impinged is high at many plants because impinged organisms are eventually suffocated by being held against the screen mesh or are abraded, which can result in fatal infection.
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| Impingement can affect large numbers of fish and invertebrates (crabs, shrimp, jellyfish, etc.). As with entrainment, operational monitoring and mitigative measures have allayed concerns about population-level effects at most plants, but impingement mortality continues to be an issue at others. Consultation with resource agencies revealed that impingement is a frequent concern at once-through power plants, particularly where restoration of anadromous fish may be affected.Impingement is an intake-related effect that is considered by EPA or state water quality permitting agencies in the development of NPDES permits and 316(b) determinations.
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| The impacts of impingement are small at many plants but may be moderate or even large at a few plants with once-through cooling systems [Reference 4-5, Section 4.2.2.1.31.
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| 4.3.5 Analysis of Environmental Impact PNPS currently uses various techniques for reducing impingement mortality.
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| The traveling screens are equipped with fish collection buckets and low-pressure sprays for removing impinged organisms.
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| The fish are washed into a fish return sluiceway and returned to the intake embayment at a point sufficiently distant from the intake to avoid re-impingement.
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| If there is an indication that fish are being impinged at a rate exceeding 20 fish per hour, the traveling screens are turned continuously until the impingement rate drops below 20 fish per hour for two Q consecutive sampling events.As Section 3.2.2.1 describes, PNPS has a once-through heat dissipation system that uses water from Cape Cod Bay for condenser cooling. Section 4.2 discusses the existing PNPS Section 316(b) determination and the combined Section 316 demonstration completed in March 2000.Attachment A contains relevant portions of the NPDES permit. On July 9, 2004, the EPA published a final rule in the Federal Register (69 FR 41575) (Reference 4-1 ) addressing cooling water intake structures at existing power plants, such as PNPS. The rule is Phase l! in the EPA's development of 316(b) regulations that establish national requirements applicable to the location, design, construction, and capacity of cooling water intake structures at existing facilities.
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| The national requirements, which are implemented through NPDES permits, provide several compliance alternatives that may be pursued by facilities to meet the entrainment and impingement performance standards in the Rule. Any additional mitigation measures under the new regulations would only further reduce the already small impacts.4.3.6 Conclusion EPA Region I has determined based on engineering judgment that the circulating water intake structures presently employs the best technology available for minimizing adverse environmental impact. Because Entergy submitted a timely application for renewal of the PNPS NPDES Permit, the 1994 permit and its Section 316(b) determination remain in effect. For this reason, Entergy 4-10 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Slage concludes that PNPS impacts due to impingement of fish and shellfish are SMALL and do not warrant mitigation beyond those measures required by the NPDES permit, as periodically amended.4.4 Heat Shock 4.4.1 Description of Issue Heat shock (for all plants with once-through and cooling pond heat dissipation systems)4.4.2 Findings from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Because of continuing concerns about heat shock and the possible need to modify thermal discharges In response to changing environmental conditions, the impacts may be of moderate or large significance at some plants. See 10 CFR 51.53(c)(3)(ii)(B).
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| 4.4.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(B)]
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| If the applicant's plant utilizes once-through cooling or cooling pond heat dissipation systems, the applicant shall provide a copy of current Clean Water Act 316(a) determinations and variance in accordance with 40 CFR Part 125, or equivalent state permits and supporting documentation.
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| If the applicant can not provide these documents, It shall assess the impact of the proposed action on fish and shellfish resources resulting from heat shock.4.4.4 Background Based on the research literature, monitoring reports, and agency consultations, the potential for thermal discharges to cause thermal discharge effect mortalities is considered small for most plants. However, impacts may be moderate or even large at a few plants with once-through cooling systems. For example, thermal discharges at one plant are considered by the agencies to have damaged the benthic invertebrate and seagrass communities in the effluent mixing zone around the discharge canal; as a result, helper cooling towers have been installed to reduce the discharge temperatures.
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| Conversely, at other plants it may become advantageous to increase the temperature of the discharge in order to reduce the volume of water pumped through the plants and thereby reduce entrainment and impingement effects. Because of continuing concerns about thermal discharge effects and the possible need to modify thermal discharges in the future in response to changing environmental conditions, this is a Category 2 issue for plants with once-through cooling systems [Reference 4-5, Section 4.2.2.1.4].
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| 4.4.5 Analysis of Environmental Impact As Section 3.2.2.1 describes, PNPS has a once-through heat dissipation system that uses water from Cape Cod Bay for condenser cooling. As discussed below, Entergy also has a Section 316(a) variance for PNPS discharges.
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| 4-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Section 316(a) of the CWA establishes a process whereby a discharger can demonstrate that established thermal discharge limitations are more stringent than necessary to protect a balanced indigenous population of fish and wildlife and obtain facility-specific thermal discharge limits (33 USC 1326). Boston Edison Company submitted a combined CWA Section 316(a) and (b) demonstration report for PNPS to EPA Region I in 1977 that was accepted by the agency and used in determining facility-specific NPDES discharge temperature limits. That original Section 316 demonstration, based on 3 years (1969-1972) of pre-operational and 5 years (1972-1976) of post-operational engineering, hydrological, and ecological data, concluded that the thermal effluent from PNPS would not result in long-term impacts to the fish and wildlife populations of Cape Cod Bay.In issuing and renewing the Station's NPDES Permits since that time, the EPA determined that thermal discharges from PNPS were sufficiently protective of the aquatic ecosystem of Cape Cod Bay to satisfy alternative thermal effluent limitations under Section 316(a) of the CWA.Those determinations were based on the original combined Section 316 Demonstration and on-going ecological monitoring programs.In recent years, EPA Region I has required all NPDES permittees in the region (to whom CWA Section 316 applies) to submit new Section 316(a) and 316(b) demonstrations.
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| EPA Region I is reviewing an Entergy application for renewal of the PNPS NPDES Permit and, as described in Section 2.2, a new combined Section 316 report that evaluates more than 25 years of data on potential thermal impacts [Reference 4-2]. This new Section 316 demonstration report concludes the following:
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| Existing thermal discharges, essentially unchanged since operation of the Station, affect only a small area in the immediate vicinity of PNPS, and have resulted in no adverse impacts to the [Representative Important Species] populations or to the integrity of the aquatic ecosystem of Cape Cod Bay. Therefore, the thermal discharge does not adversely affect the propagation or protection of a balanced, indigenous population of fish, shellfish, and wildlife in Cape Cod Bay. [Reference 4-2, page 7-6]4.4.6 Conclusion As noted previously, Entergy has submitted a timely application for renewal of the PNPS NPDES Permit. The current NPDES Permit (provided in Attachment A) and its Section 316(a) variance therefore remain in effect. For this reason, Entergy concludes that impacts to fish and shellfish from heat shock are SMALL and warrant no additional mitigation.
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| 4.5 Groundwater Use Conflicts (Plants Using >100 agm of Groundwater) 4.5.1 Description of Issue Groundwater use conflicts (potable and service water, and dewatering:
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| plants that use >100 gpm)4-12 QW)
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.5.2 Findings from Table B-1, Subpart A, Appendix A SMALL, MODERATE, or LARGE. Plants that use more than 100 gpm may cause groundwater use conflicts with nearby groundwater users. See 10 CFR 51.53(c)(3)(ii)(C).
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| 4.5.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(C)]
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| If the applicant's plant uses Ranney wells or pumps more than 100 gallons (total onsite) of groundwater per minute, an assessment of the impact of the proposed action on groundwater use must be provided.4.5.4 Analysis of Environmental Impact The issue of groundwater use conflicts at plants that pump more than 100 gallons per minute of groundwater does not apply to PNPS. As Sections 3.2.2.1 and 3.2.2.2 describe, the plant obtains all its cooling and process water from Cape Cod Bay, and gets its potable and reactor makeup water from the Town of Plymouth.4.6 Groundwater Use Conflicts (Plants U.sing Cooling Towers Withdrawing Make-Up Water from a Small River)4.6.1 Description of Issue Groundwater use conflicts (plants using cooling towers withdrawing make-up water from a small river)4.6.2 Findings from Table B-I, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Water use conflicts may result from surface water withdrawals from small water bodies during low flow conditions which may affect aquifer recharge, especially if other groundwater or upstream surface water users come on line before the time of license renewal. See 10 CFR 51.53(c)(3)(ii)(A).
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| 4.6.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(A)]
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| If the applicant's plant utilizes cooling towers or cooling ponds and withdraws make-up water from a river whose annual flow rate is less than 3.15x10 1 2 ft 3/year (9x1010 m 3/year), an assessment of the impact of the proposed action on.the flow of the river and related impacts on instream and riparian ecological communities must be provided.
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| The applicant shall also provide an assessment of the impacts of the withdrawal of water from the river on alluvial aquifers during low flow.4-13 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.6.4 Analysis of Environmental Impact The issue of groundwater use conflicts does not apply to PNPS because the plant does not use cooling towers or cooling ponds and does not withdraw water from a small river. PNPS uses a once-through cooling system that withdraws and discharges water to Cape Cod Bay.4.7 Groundwater Use Conflicts (Plants Using Ranney Wells)4.7.1 Description of Issue Groundwater use conflicts (plants using Ranney wells)4.7.2 Findings from Table B-I, Subpart A, Appendix A SMALL, MODERATE, or LARGE. Ranney wells can result in potential groundwater depression beyond the site boundary.
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| Impacts of large groundwater withdrawal for cooling tower makeup at nuclear power plants using Ranney wells must be evaluated at the time of application for license renewal. See 10 CFR 51.53(c)(3)(ii)(C).
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| 4.7.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(C)]
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| If the applicant's plant uses Ranney wells or pumps more than 100 gallons (total onsite) of groundwater per minute, an assessment of the impact of the proposed action on groundwater use must be provided.
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| Q 4.7.4 Analysis of Environmental Impact PNPS does not utilize Ranney wells. Potable water is supplied by the town of Plymouth and cooling water is taken from Cape Cod Bay for a once-through cooling system that discharges water to Cape Cod Bay. Therefore, this issue is not applicable to PNPS and analysis is not required.4.8 Degradation of Groundwater Quality 4.8.1 Description of Issue Groundwater quality degradation (cooling ponds at inland sites).4.8.2 Findings from Table B-I, Subpart A, Appendix A SMALL, MODERATE, or LARGE. Sites with closed-cycle cooling ponds may degrade groundwater quality. For plants located inland, the quality of the groundwater in the vicinity of the ponds must be shown to be adequate to allow continuation of current uses. See 10 CFR 51 .53(c)(3)(ii)(D).
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| 4-14 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.8.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(D)]
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| If the applicant's plant is located at an inland site and utilizes cooling ponds, an assessment of the impact of the proposed action on groundwater quality must be provided.4.8.4 Analysis of Environmental Impact PNPS is not an inland site and does not utilize cooling ponds. PNPS utilizes a once-through cooling system that withdraws water from and discharges to Cape Cod Bay. Therefore, this issue is not applicable to PNPS and analysis is not required.4.9 Iminacts of Refurbishment on Terrestrial Resources 4.9.1 Description of Issue Refurbishment impacts -Terrestrial Resources 4.9.2 Findings from Table B-1, Subpart A, Appendix A SMALL MODERATE, or LARGE. Refurbishment impacts are insignificant if no loss of important plant and animal habitat occurs. However, it cannot be known whether important plant and animal communities may be affected until the specific proposal is presented with the license renewal application.
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| See 10 CFR 51.53(c)(3)(ii)(E).
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| 4.9.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(E)l All license renewal applicants shall assess the impact of refurbishment and other license renewal related construction activities on important plant and animal habitats.4.9.4 Analysis of Environmental Impact As noted in Section 3.3, no refurbishment activities are required for PNPS license renewal.Therefore this issue is not applicable to PNPS and no analysis is required.4.10 Threatened or Endangered Species 4.10.1 Description of Issue Impacts from refurbishment and continued operations on threatened or endangered species.4.10.2 Findings from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Generally, plant refurbishment and continued operation are not expected to adversely affect threatened or endangered species. However, consultation with appropriate agencies would be needed at the time of license renewal to determine whether threatened or endangered species are present and whether they would be adversely affected.See 10 CFR 51.53(c)(3)(ii)(E).
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| 4-15 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.10.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(E)J All license renewal applicants shall assess the impact of refurbishment and other license renewal related construction activities on important plant and animal habitats.
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| Additionally, the applicant shall assess the impact of the proposed action on threatened or endangered species in accordance with the Endangered Species Act.4.10.4 Background The NRC did not reach a conclusion about the significance of potential impacts to threatened and endangered species in the GEIS because (1) the significance of impacts on such species cannot be assessed without site- and project-specific information that will not be available until the time of license renewal and (2) additional species that are threatened with extinction and that may be adversely affected by plant operations may be identified between the present and the time of license renewal [Reference 4-5, Section 3.9].4.10.5 Analysis of Environmental Impacts Section 2.2 of this ER describes the aquatic communities of western Cape Cod Bay and discusses population trends in recreationally, socially, and commercially important populations, including the American lobster and winter flounder.
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| Section 2.4 describes important terrestrial habitats at PNPS and along the associated PNPS-to-Snake Hill Road transmission corridor.
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| As discussed in Section 2.4, the transmission corridor crosses an area designated as critical habitat for the endangered northern red-bellied cooter, but the PNPS-to-Snake Hill Road transmission line is not owned or maintained by Entergy. The PNPS site does contain a priority habitat for the state-listed Species of Special Concern, the spotted turtle. Section 2.5 discusses threatened or endangered species that occur or may occur at PNPS, along this transmission corridor, or in Cape Cod Bay.With the exception of the four species identified in Section 2.5, Entergy is not aware of any threatened or endangered terrestrial species that could occur at the PNPS site or along the associated transmission corridor.
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| Current operations of PNPS and NSTAR vegetation management practices along transmission line rights-of-way do not adversely affect any listed terrestrial species or its habitat (see Section 2.4). Furthermore, station operations and transmission line maintenance practices are not expected to change significantly during the license renewal term. Therefore, no adverse impacts to threatened or endangered terrestrial species from current or future operations are anticipated.
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| As discussed in Section 3.3, Entergy has no plans to conduct refurbishment or construction activities at PNPS during the license renewal term. Therefore, there would be no refurbishment-related impacts to special-status species and no further analysis of refurbishment-related impacts is applicable.
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| 4-16 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Boston Edison and Entergy have conducted extensive population studies of fish and shellfish in the vicinity of PNPS since 1969. No state- or federally-listed fish species has been collected or observed in more than 30 years of monitoring.
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| As noted in Section 2.5, a number of threatened and endangered marine species (five whales and five sea turtles) pass Cape Cod during seasonal migrations and sometimes forage in semi-enclosed Cape Cod Bay. Most of the great whales (the minke, finback, and right whales are exceptions) live and forage over the continental shelf, approaching the coastline only during seasonal migrations.
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| Although whales are regularly observed in summer months in the eastern portion of Cape Cod Bay and the Stellwagen Bank area, they do not normally feed in the western portion of the Bay or in the vicinity of PNPS. Because whales do not move into the shallow waters immediately offshore of PNPS, they are not affected by operation of the PNPS cooling water intake system or by the station's thermal discharge.
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| There is no evidence that operation of PNPS has had an effect on whales in Cape Cod Bay.Sea turtles are more likely to move inshore and feed in shallow coastal waters (particularly the green sea turtle, which actually comes ashore to bask), but reports of sea turtles foraging in extreme western Cape Cod Bay are rare. As discussed in Section 2.5, small numbers of sea turtles are stranded every year on Cape Cod beaches, but strandings on the western shore of the Bay (the mainland) are rare. No sea turtles have been impinged at PNPS, and none have been rescued from the PNPS intake canal. There are no records of sea turtles congregating in the area of the PNPS discharge canal, and no indication that the thermal effluent has disrupted normal seasonal movement or migration of turtles.Entergy wrote to the MDFW, the FWS, and the NMFS requesting information on any listed species or critical habitats that might occur on the PNPS site or along the associated transmission corridor, with particular emphasis on species that might be adversely affected by continued operation over the license renewal period. Agency responses are provided in Attachment B of this ER. The FWS is in agreement regarding the transitory nature of the three listed bird species, as well as to the nature of the red-bellied cooter turtle habitat on the transmission lines. NMFS did recommend that Entergy address any impact on sea turtles in preparing this application.
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| As was stated previously in Section 2.5 of this ER, in the thirty-three years that PNPS has been in operation no sea turtles have ever been observed in the intake or discharge canal or along the PNPS waterfront.
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| In the twenty-five years that Mass Audubon has been documenting the numbers and locations of sea turtle strandings in Massachusetts, only one sea turtle stranding has been recorded in the town of Plymouth [Reference 4-10] and that stranding was not attributable to PNPS operations.
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| MDFW stated, "If there are no plans to expand the footprint or to alter current operations over the license period, then it would not seem likely that there would be an adverse affect on state-protected wildlife species." However, MDFW was unable to provide an official determination unless a full environmental review was conducted.
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| 4-17 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.10.6 Conclusion As discussed in Section 3.3, Entergy has no plans to conduct refurbishment or construction activities at PNPS during the license renewal term. Therefore, there will be no impact to threatened and endangered species from refurbishment activities.
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| Because Entergy has no plans to alter current operations and resource agencies contacted by Entergy evidenced no serious concerns about license renewal impacts, Entergy concludes that impacts to threatened or endangered species from license renewal would be SMALL and do not warrant further mitigation.
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| Renewal of the operating license for PNPS is not expected to result in the taking of any threatened or endangered species. Renewal of the license is not likely to jeopardize the continued existence of any threatened or endangered species or result in the destruction or adverse modifications of any critical habitat.4.11 Air Quality During Refurbishment (Nonattainment and Maintenance Areas)4.11.1 Description of Issue Air quality during refurbishment (nonattainment and maintenance areas).4.11.2 Findings from Table B-I, Subpart A, Appendix A SMALL, MODERATE, or LARGE. Air quality impacts from plant refurbishment associated with license renewal are expected to be small. However, vehicle exhaust emissions could be cause for concern at locations in or near nonattainment or maintenance areas. The significance of the potential impact cannot be determined without considering the compliance status of each site and the number of workers expected to be employed during the outage. See 10 CFR 51 .53(c)(3)(ii)(F).
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| 4.11.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(F)]
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| If the applicant's plant is located in or near a nonattainment or maintenance area, an assessment of vehicle exhaust emissions anticipated at the time of peak refurbishment workforce must be provided in accordance with the Clean Air Act as amended.4.11.4 Analysis of Environmental Impact As discussed in Section 3.3, Entergy has no plans for refurbishment related to license renewal at PNPS. Therefore, this issue is not applicable to PNPS and analysis is not required.4-18 (
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.12 Impact on Public Health of Microbiological Organisms 4.12.1 Description of Issue Microbiological organisms (public health) (plants using lakes, canals, cooling towers, or cooling ponds that discharge to a small river).4.12.2 Finding from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. These organisms are not expected to be a problem at most operating plants except possibly at plants using cooling ponds, lakes, or canals that discharge to small rivers. Without site-specific data, it is not possible to predict the effects generically.
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| See 10 CFR 51.53(c)(3)(ii)(G).
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| 4.12.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(G)J If the applicant's plant uses a cooling pond, lake, or canal or discharges into a river having an annual average flow rate of less than 3.15x10 1 2 ft 3/year (9x10 1 0 m 3/year), an assessment of the impact of the proposed action on public health from thermophilic organisms in the affected water must be provided.4.12.4 Analysis of Environmental Impact The issue of thermophilic organisms does not apply to PNPS because the plant does not use a cooling pond, lake, canal, or discharge to a small river. PNPS uses a once-through cooling system that withdraws from and discharges water into Cape Cod Bay. Therefore, this issue is not applicable to PNPS and analysis is not required.4.13 Electromagnetic Fields-Acute Effects 4.13.1 Description of Issue Electromagnetic fields, acute effects (electric shock)4.13.2 Findings from Table B-1, Subpart A, Appendix A SMALL, MODERATE, or LARGE. Electric shock resulting from direct access to energized conductors or from induced charges in metallic structures has not been a problem at most operating plants and generally is not expected to be a problem during the license renewal term.However, site-specific review is required to determine the significance of the electrical shock potential at the site. See 10 CFR 51.53(c)(3)(ii)(H).
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| 4.13.3 Requirements
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| [10 CFR 51.53(c)(3)(ii)(H)l If the applicant's transmission lines that were constructed for the specific purpose of connecting the plant to the transmission system do not meet the recommendations of the National Electric 4-19 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ( i Safety Code for preventing electric shock from induced currents, an assessment of the impact of the proposed action on the potential shock hazard from the transmission lines must be provided.4.13.4 Background The transmission line of concern is that between the plant switchyard and the intertie to the transmission system. With respect to shock safety issues and license renewal, three points must be made. First, in the licensing process for the earlier licensed nuclear plants, the issue of electrical shock safety was not addressed.
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| Second, some plants that received operating licenses with a stated transmission line voltage may have chosen to upgrade the line voltage for reasons of efficiency, possibly without reanalysis of induction effects. Third, since the initial NEPA review for those utilities that evaluated potential shock situations under the provision of the NESC, land use may have changed, resulting in the need for reevaluation of this issue.The electrical shock issue, which is generic to all types of electrical generating stations, including nuclear power plants, is of small significance for transmission lines that are operated in adherence with NESC. Without review of each nuclear plant's transmission line conformance with NESC criteria, it is not possible to determine the significance of the electrical shock potential[Reference 4-5, Sections 4.5.4 and 4.5.4.1].4.13.5 Analysis of Environmental Impact In the case of PNPS, there have been no previous NRC or NEPA analyses of transmission-line-induced-current hazards. Therefore, this section provides an analysis of the station's transmission lines' conformance with the NESC standard.
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| The analysis is based on computer modeling of electric field strength under the lines.Objects near transmission lines can become electrically charged due to their immersion in the lines' electric field. This charge results in a current that flows through the object to the ground.The current is called "induced" because there is no direct connection between the line and the object. The induced current can also flow to the ground through the body of a person who touches the object. An object that is insulated from the ground can actually store an electrical charge, becoming what is called "capacitively charged." A person standing on the ground and touching a vehicle or a fence receives an electrical shock due to the discharge of the capacitive charge through the person's body to the ground. After the initial discharge, a steady-state current can develop, the magnitude of which depends on several factors, including* the strength of the electric field which, in turn, depends on the voltage of the transmission line as well as its height and geometry;* the size of the object on the ground; and* the extent to which the object is grounded.4-20 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage In 1977, the NESC adopted a provision that describes an additional criterion to establish minimum vertical clearances to the ground for electric lines having voltages exceeding 98-kV alternating current to ground.1 The clearance must limit the steady-state induced current2 to 5 milliamperes if the largest anticipated truck, vehicle, or equipment were short-circuited to ground. By way of comparison, the setting of ground fault circuit interrupters used in residential wiring (special breakers for outside circuits or those with outlets around water pipes) is 4 to 6 milliamperes..
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| As described in Section 3.2.7, two 345-kV lines were specifically constructed to distribute power from PNPS to the electric grid. Entergy's analysis of these transmission lines began by identifying the limiting case for each line. The limiting case is the location along each line where the potential for current-induced shock would be greatest.
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| Because in the region of interest the two transmission lines share towers, there was only one limiting location to be considered.
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| For convenience and conservatism, the limiting case selected was the hypothetical location with minimum clearance allowed by the Commonwealth of Massachusetts for 345-kV lines. All spans on these lines have greater clearance than the limiting case.Once the limiting case was identified, NSTAR, the lines' owner, calculated the electric field strength underneath the lines, allowing for contribution from both lines simultaneously.
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| NSTAR used the Electric Power Research Institute (EPRI) code, ENVIRO, to determine electric field strength [Reference 4-9].Finally, Entergy calculated the induced current based on the distribution of electric field strength.Entergy used methods described in EPRI's Transmission Line Reference Book [Reference 4-4].The analysis assumed the maximum vehicle allowed by the Commonwealth of Massachusetts, which is a tractor-trailer 60 feet long, 8 feet wide, and a maximum of 13.5 feet high.Entergy determined that the combined effect of the two lines does not have the capacity to induce as much as 5 milliamperes in a vehicle parked beneath the lines. The Entergy-calculated induced current would be 4.5 milliamps
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| [Reference 4-111. Therefore, the PNPS transmission line designs conform to the NESC provisions for preventing electric shock from induced current.NSTAR conducts surveillance and maintenance to ensure that design ground clearances do not change. These procedures include routine aerial inspections on a regular basis. These aerial patrols of all corridors include checks for encroachments, broken conductors, broken or leaning structures, and signs of trees burning, any of which would be evidence of clearance problems.Ground inspections include examination for clearance at questionable locations, integrity of structures, and surveillance for dead or diseased trees which might fall on the transmission lines.The results of these observations and inspections are reviewed by NSTAR Asset Management engineers and follow-up inspections are scheduled if necessary.
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| The completed reviews are evaluated and prioritized based upon safety and structural integrity.
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| Work orders are created in 1. Part 2, Rules 232C1 c and 232D3c.2. The NESC and the GEIS use the phrase "steady-state current," whereas 10 CFR 51.53(c)(3)(ii)(H) uses the phrase "induced current." The phrases mean the same here.4-21 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage NSTAR's work management system for those observations which require action and the responsible operating divisions are notified to schedule the corrective action.4.13.6 Conclusion Entergy's assessment concludes that electric shock is of SMALL significance for the PNPS transmission lines. Due to the small significance of the issue, mitigation measures such as installing warning signs at road crossings or increasing clearances are not warranted.
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| 4.14 Housing Impacts 4.14.1 Description of Issue Housing impacts 4.14.2 Findings from Table B-I, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Housing impacts are expected to be of small significance at plants located in a medium or high population area and not in an area where growth control measures that limit housing development are in effect. Moderate or large housing impacts of the workforce associated with refurbishment may be associated with plants located in sparsely populated areas or in areas with growth control measures that limit housing development.
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| See 10 CFR 51.53(c)(3)(ii)(1).
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| 4.14.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(1)]
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| An assessment of the impact of the proposed action on housing availability...
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| within the vicinity of the plant must be provided.4.14.4 Background The impacts on housing are considered to be of small significance when a small and not easily discernible change in housing availability occurs, generally as a result of a very small demand increase or a very large housing market. Increases in rental rates or housing values in these areas would be expected to equal or slightly exceed the statewide inflation rate. No extraordinary construction or conversion of housing would occur where small impacts are foreseen.The impacts on housing are considered to be of moderate significance when there is a discernible but short-lived reduction in available housing units because of project-induced in-migration.
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| The impacts on housing are considered to be of large significance when project-related demand for housing units would result in very limited housing availability and would increase rental rates and housing values well above normal inflationary increases in the state.Moderate and large impacts are possible at sites located in rural and remote areas, at sites located in areas that have experienced extremely slow population growth (and thus slow or no 4-22 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage growth in housing), or where growth control measures that limit housing development are in existence or have been recently lifted [Reference 4-5, Section 3.7.2].4.14.5 Analysis of Environmental Impact Supplement 1 to Regulatory Guide 4.2, provides the following guidance.Section 4.14.1 states, "If there will be no refurbishment or if refurbishment involves no additional workers then there will be no impact on housing and no further analysis is required." Section 4.14.2 states, "If additional workers are not anticipated there will be no impact on housing and no further analysis is required." As noted in 10 CFR 51, Subpart A, Appendix B, Table B-I, the NRC concluded that impacts to housing are expected to be of small significance at plants located in high population areas where growth control measures are not in effect. As of February 2005, the PNPS site has approximately 703 full time workers (Entergy employees and baseline contractors) during normal plant operations.
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| As described in Section 2.6, PNPS is located in a high population area. As described in Section 3.5, Entergy does not plan to add any additional permanent employees during the license renewal term. Entergy's analysis of the Plymouth and Barnstable County planning tools, such as zoning and redevelopment incentives, determined that the tools are designed to guide growth, but not to limit it.4.14.6 Conclusion As noted in Section 3.3, there are no major refurbishment activities required for PNPS license renewal. Additionally, Entergy does not anticipate a need for additional full time workers during the license renewal period. Therefore, Entergy concludes that impacts to the housing availability from plant-related population growth and plant demand would be SMALL and mitigation would not be warranted.
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| 4.15 Public Utilities:
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| Public Water SuDr IV Availability 4.15.1 Description of Issue Public services (public utilities) 4.15.2 Findings from Table B-1, Appendix B to Subpart A SMALL or MODERATE.
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| An increased problem with water shortages at some sites may lead to impacts of moderate significance on public water supply availability.
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| See 10 CFR 51 .53(c)(3)(ii)(1).
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| 4-23 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.15.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(1)]
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| ...[T]he applicant shall provide an assessment of the impact of population increases attributable to the proposed project on the public water supply.4.15.4 Public Water Supply -Background Impacts on public utility services are considered small if little or no change occurs in the utility's ability to respond to the level of demand and thus there is no need to add capital facilities.
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| Impacts are considered moderate if overtaxing of facilities during peak demand periods occurs.Impacts are considered large if existing service levels (such as the quality of water and sewage treatment) are substantially degraded and additional capacity is needed to meet ongoing demands for services.In general, small to moderate impacts to public utilities were observed as a result of the original construction of the case study plants. While most locales experienced an increase in the level of demand for services, they were able to accommodate this demand without significant disruption.
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| Water service seems to have been the most affected public utility.Public utility impacts at the case study sites during refurbishment are projected to range from small to moderate.
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| The potentially small to moderate impact at Diablo Canyon is related to water availability (not processing capacity) and would occur only if a water shortage occurs at refurbishment time. Q Because the case studies indicate that some public utilities may be overtaxed during peak periods, the impacts to public utilities would be moderate in some cases, although most sites would experience only small impacts [Reference 4-5, Section 3.7.4.5].4.15.5 Analysis of Environmental Impact As noted in Section 3.3, there are no major refurbishment activities required for PNPS license renewal. Therefore, there will be no impact to public utilities from refurbishment activities and therefore no further analysis is needed.PNPS demand for water is not expected to change during the license renewal period. Section 2.9.1 notes that average daily water withdrawals exceed authorized withdrawal limits (capacities) in some areas. The region overall has excess capacity, but is expected to eventually experience water shortages in several of the larger municipalities of Plymouth and Barnstable Counties[Reference 4-1]. However, Entergy does not anticipate a need for additional workers during the period of extended operation.
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| There will be no impact to public utilities from additional plant workers living in the two-county area near the plant where the majority of employees live.4.15.6 Conclusion Although future water shortages are a concern for the region, their occurrence would be independent of the license renewal process. Therefore, Entergy concludes that impacts to the 4-24 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage public water supply from plant-related population growth and plant demand would be SMALL and mitigation would not be warranted.
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| 4.16 Education Impacts from Refurbishment 4.16.1 Description of Issue Public Services (effects of refurbishment activities upon local educational system)4.16.2 Findings from Table B-1, Appendix B to Subpart A SMALL or MODERATE.
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| Most sites would experience impacts of small significance but larger impacts are possible depending on site- and project-specific factors. See 10 CFR 51 .53(c)(3)(ii)(l).
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| 4.16.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(l)J An assessment of the impact of the proposed action on... public schools (impacts from refurbishment activities only) within the vicinity of the plant must be provided.4.16.4 Analysis of Environmental Impact As noted in Section 3.3, there are no major refurbishment activities required for PNPS license renewal. Therefore this issue is not applicable to PNPS and no analysis is required.4.17 Offsite Land Use-Refurbishment 4.17.1 Description of Issue Offsite Land Use (effects of refurbishment activities) 4.17.2 Findings from Table B-1, Appendix B to Subpart A SMALL or MODERATE.
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| Impacts may be of moderate significance at plants in low population areas. See 10 CFR 51.53(c)(3)(ii)(1).
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| 4.17.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(1)]
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| An assessment of the impact of the proposed action on... land-use.. .within the vicinity of the plant must be provided.4.17.4 Analysis of Environmental Impact As noted In Section 3.3, there are no major refurbishment activities required for PNPS license renewal. Therefore, there will be no impacts from refurbishment activities and no analysis is required.4-25 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.18 Offsite Land Use-License Renewal Term 4.18.1 Description of Issue Offsite Land Use (effects of license renewal)4.18.2 Findings from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Significant changes in land-use may be associated with population and tax revenue changes resulting from license renewal. See 10 CFR 51 .53(c)(3)(ii)(1).
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| 4.18.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(1)J An assessment of the impact of the proposed action on ... land-use.. .within the vicinity of the plant must be provided.4.18.4 Background During the license renewal term, new land use impacts could result from plant-related population growth or from the use of tax payments from the plant by local government to provide public services that encourage development.
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| However, as noted in Regulatory Guide 4.2, Section 4.17.2, Table B-1 of 10 CFR 51 partially misstates the conclusion reached in Section 4.7.4.2 of NUREG-1437.
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| NUREG-1437, Section 4.7.4.2 concludes, "...population-driven land use changes during the license renewal term at all nuclear plants will be small....'
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| Regulatory Guide 4.2 further states, "Until Table B-1 is changed, applicants only need cite NUREG-1437 to address population-induced land-use change during the license renewal term." Therefore, the discussion will be limited to the land use changes that may result from tax payments made by the plant to local governments.
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| The assessment of new tax-driven land use impacts in the GEIS considered the following:
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| * the size of the plant's tax payments relative to the community's total revenues,* the nature of the community's existing land use pattern, and* the extent to which the community already has public services in place to support and guide development.
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| In general, if the plant's tax payments are projected to be small relative to the community's total revenue, new tax-driven land use changes during the plant's license renewal term would be small, especially where the community has pre-established patterns of development and has provided adequate public services to support and guide development.
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| If the plant's tax payments are projected to be medium to large relative to the community's total revenue, new tax-driven land use changes would be moderate.4-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage This is most likely to be true where the community has no pre-established patterns of development (i.e., land use plans or controls) or has not provided adequate public services to support and guide development in the past, especially infrastructure that would allow industrial development.
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| If the plant's tax payments are projected to be a dominant source of the community's total revenue, new tax-driven land use changes would be large. This would be especially true where the community has no pre-established pattern of development or has not provided adequate public services to support and guide development in the past.Based on predictions for the case study plants, it is projected that all new population-driven land use changes during the license renewal term at all nuclear plants will be small because population growth caused by license renewal will represent a much smaller percentage of the local area's total population than has operations-related growth. Also, any conflicts between offsite land use and nuclear plant operations are expected to be small. In contrast, it is projected that new tax-driven land use changes may be moderate at a number of sites and large at some others. Because land use changes may be perceived by some community members as adverse and by others as beneficial, the staff is unable to assess generically the potential significance of site-specific off-site land use impacts [Reference 4-5, Section 4.7.4.2].4.18.5 Analysis of Environmental Impact The environmental impacts from this issue are from population-driven land use changes and from tax-driven land use changes.4.18.5.1 Population-Driven Land Use Changes Entergy agrees with the GEIS conclusion that new population-driven land use changes at PNPS during the license renewal term would be SMALL [Reference 4-5, Section 4.7.4.2].
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| Entergy does not anticipate that additional workers will be employed at PNPS during the period of extended operations.
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| Therefore there will be no adverse impact to the offsite land use from plant-related population growth.4.18.5.2 Tax-Driven Land Use Changes The NRC has determined that the significance of tax payments as a source of local government revenue would be small if the payments are less than 10% of revenue [Reference 4-5, Section 3.7.3]. The NRC further determined that, if a plant's tax payments are projected to be small relative to the community's total revenue (i.e., less than 10% of revenue), new tax-driven land-use changes would be small.The NRC defined the magnitude of land-use changes as follows [Reference 4-5, Section 4.7.4]:* Small -very little new development and minimal changes to an area's land-use pattern;* Moderate -considerable new development and some changes to land-use pattern;4-27 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Large -large-scale new development and major changes in land-use pattern.Table 2-4 compares the tax payments made by Entergy to the Town of Plymouth with the Town's annual property tax revenues.
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| Entergy's tax payments to the Town of Plymouth represent approximately 2 to 3% of the Town's total annual property tax revenues.
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| Using the NRC's criteria, Entergy's tax payments are of small significance to the Town of Plymouth.
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| As described in Section 3.3, Entergy does not anticipate refurbishment or construction during the license renewal period. Therefore, Entergy does not anticipate any increase in the assessed value of PNPS due to refurbishment-related improvements, or any related tax-increase-driven changes to offsite land-use and development patterns.Additionally, Section 2.8 describes the Town of Plymouth's land-use patterns, which reflect the use of planning tools, such as zoning, to prohibit new construction in selected areas and encourage growth in others. Section 2.9 describes public facilities.
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| Because infrastructure is limited in some areas and accessible in others, zoning guidelines encourage growth in areas where infrastructure already exists. New infrastructure construction is less likely to occur.Therefore, growth is encouraged, but limited to pre-selected areas. During the summer months, tourism creates a large surge in population and the overflow is absorbed by existing temporary housing accommodations.
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| This surge does not, however, affect overall permanent residential housing patterns or capacities.
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| 4.18.6 Conclusion Because Entergy's tax payments are small, and the Town of Plymouth has pre-established patterns of development and has been able to provide adequate public services to support and guide ongoing development, Entergy concludes that impacts to offsite land use from plant-related tax impacts would be SMALL and mitigation would not be warranted.
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| 4.19 Transvortation 4.19.1 Description of Issue Public services, Transportation 4.19.2 Finding from Table B-1, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Transportation impacts (level of service) of highway traffic generated during plant refurbishment and during the term of the renewed license are generally expected to be of small significance.
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| However, the increase in traffic associated with additional workers and the local road and traffic control conditions may lead to impacts of moderate or large significance at some sites. See 10 CFR 51.53(c)(3)(ii)(J).
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| 4.19.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(J)]
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| All applicants shall assess the impact of the proposed project on local transportation during periods of license renewal refurbishment activities and during the term of the renewed license.4-28 Q' Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.19.4 Background Impacts to transportation during the license renewal term would be similar to those experienced during current operations and would be driven mainly by the workers involved in current plant operations.
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| Based on past and projected impacts at the case study sites, transportation impacts would continue to be of small significance at all sites during operations and would be of small or moderate significance during scheduled refueling and maintenance outages. Because impacts are determined primarily by road conditions existing at the time of the project and cannot be easily forecast, a site specific review will be necessary to determine whether impacts are likely to be small or moderate and whether mitigation measures may be warranted
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| [Reference 4-5, Section 3.7.7].4.19.5 Analysis of Environmental Impact As described in Section 3.3, no refurbishment is planned and no refurbishment impacts to local transportation are anticipated.
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| No further evaluation is necessary.
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| During the license renewal term, as described in Section 3.5, Entergy does not intend to add any additional employees above the existing reactor workforce of approximately 703 during normal operations of the license renewal term and an outage workforce of as many as 1,600 workers (including permanent employees and contractors for the outage).4.19.6 Conclusion As discussed in Section 3.3, no refurbishment is planned and no refurbishment impacts to local transportation are anticipated.
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| Also, Entergy does not intend to add any additional license renewal term employees above the existing reactor workforce and outage workforce.
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| Therefore impacts on local traffic will be SMALL and no mitigation measures are warranted.
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| 4.20 Historic and Archaeological Prooertif s 4.20.1 Description of Issue Historic and Archaeological Resources 4.20.2 Finding from Table B-I, Appendix B to Subpart A SMALL, MODERATE, or LARGE. Generally, plant refurbishment and continued operation are expected to have no more than small adverse impacts on historic and archaeological resources.
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| However, the National Historic Preservation Act requires the Federal agency to consult with the State Historic Preservation Officer to determine whether there are properties present that require protection.
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| See 10 CFR 51.53(c)(3)(ii)(K).
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| 4-29 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.20.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(K)]
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| All applicants shall assess whether any historic or archaeological properties will be affected by the proposed project.4.20.4 Background It is unlikely that moderate or large impacts to historic resources occur at any site unless new facilities or service roads are constructed or new transmission lines are established.
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| However, the identification of historic resources and determination of possible impact to them must be done on a site-specific basis through consultation with the SHPO. The site-specific nature of historic resources and the mandatory National Historic Preservation Act consultation process mean that the significance of impacts to historic resources and the appropriate mitigation measures to address those impacts cannot be determined generically
| |
| [Reference 4-5, Section 3.7.71.4.20.5 Analysis of Environmental Impact As described in Section 2. 11, no archaeological or historic sites of significance were identified during surveys prior to station construction.
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| Entergy does not plan any refurbishment activities, so no refurbishment-related impacts are anticipated.
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| Local archaeological, State Register of Historic sites, and National Historic Register sites of significance have been identified.
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| Although a number of archaeological and historical sites are located on or near the station and its transmission line corridors, PNPS is not aware of any adverse effects or detrimental impacts on these sites caused by the operation of PNPS.Therefore, Entergy concludes that the continued operation of PNPS would have SMALL adverse impacts on historic or archaeological resources; hence, there would be no impacts to mitigate.PNPS corresponded with the SHPO regarding the potential effect of the proposed license renewal of PNPS. The SHPO confirmed that LR at PNPS is unlikely to affect significant historic or archaeological resources.
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| 4.20.6 Conclusion As noted in Section 3.3, there are no major refurbishment activities required for license renewal at PNPS. In addition, based on consultation with the State Historic Preservation Officer (see Attachment C), no prehistoric or historic resources would be affected by operation of the plant during the license renewal period. Therefore, the potential impact of continued operation of PNPS during the period of the renewed license on historic or archeological resources will be SMALL and evaluation of mitigation measures is not warranted.
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| 4-30 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.21 Severe Accident Mitigation Alternatives 4.21.1 Description of Issue Severe accidents 4.21.2 Finding from Table B-1, Appendix B to Subpart A SMALL. The probability weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to ground water, and societal and economic impacts from severe accidents are small for all plants. However, alternatives to mitigate severe accidents must be considered for all plants that have not considered such alternatives.
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| See 10 CFR 51 .53(c)(3)(ii)(L).
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| 4.21.3 Requirement
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| [10 CFR 51.53(c)(3)(ii)(L)]
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| If the staff has not previously considered severe accident mitigation alternatives for the applicant's plant in an environmental impact statement or related supplement or in an environmental assessment, a consideration of alternatives to mitigate severe accidents must be provided.4.21.4 Background The staff concluded that the generic analysis summarized in the GEIS applies to all plants and that the probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to ground water, and societal and economic impacts of severe accidents are of small significance for all plants. However, not all plants have performed a site-specific analysis of measures that could mitigate severe accidents.
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| Consequently, severe accidents are a Category 2 issue for plants that have not performed a site-specific consideration of severe accident mitigation and submitted that analysis for Commission review [Reference 4-5, Section 5.5.2.5].4.21.5 Analysis of Environmental Impact The method used to perform the Severe Accident Mitigation Analysis (SAMA) was based on the handbook used by the NRC to analyze benefits and costs of its regulatory activities
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| [Reference 4-61.Environmental impact statements and environmental reports are prepared using a sliding scale in which impacts of greater concern and mitigation measures of greater potential value receive more detailed analysis than impacts of less concern and mitigation measures of less potential value. Accordingly, Entergy used less detailed feasibility investigation and cost estimation techniques for SAMA candidates having disproportionately high costs and low benefits and more detailed evaluations for the most viable candidates.
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| 4-31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The following is a brief outline of the approach taken in the SAMA analysis.(1) Establish the Baseline Impacts of a Severe Accident Severe accident impacts were evaluated in four areas:* Off-site exposure costs -monetary value of consequences (dose) to off-site population The Probabilistic Safety Assessment (PSA) model was used to determine total accident frequency (core damage frequency (CDF) and containment release frequency).
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| The Melcor Accident Consequences Code System 2 (MACCS2) was used to convert release input to public dose. Dose was converted to present worth dollars (based on a valuation of $2,000 per person-rem and a present worth discount factor of 7.0%).* Off-site economic costs -monetary value of damage to off-site property The PSA model was used to determine total accident frequency (CDF and containment release frequency).
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| MACCS2 was used to convert release input to off-site property damage. Off-site property damage was converted to present worth dollars based on a discount factor of 7.0%.* On-site exposure costs -monetary value of dose to workers Best estimate occupational dose values were used for immediate and long-term dose. Dose was converted to present worth dollars (based on a valuation of$2,000 per person-rem and a present worth discount factor of 7%).* On-site economic costs -monetary value of damage to on-site property Best estimate cleanup and decontamination costs were used. On-site property damage estimates were converted to present worth dollars based on a discount factor of 7.0%. It was assumed that, subsequent to a severe accident, the plant would be decommissioned rather than restored.
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| Therefore replacement and refurbishment costs were not included in on-site costs. Replacement power costs were considered.
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| (2) Identify SAMA Candidates Potential SAMA candidates were identified from the following sources (see Attachment E for reference details): 4-32 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* Severe Accident Mitigation Design Alternative (SAMDA) analyses submitted in support of original licensing activities for other operating nuclear power plants and advanced light water reactor plants;* SAMA analyses for other BWR plants, including the General Electric (GE)Advanced Boiling Water Reactor (ABWR) design;* NRC and industry documentation discussing potential plant improvements;
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| -PNPS Individual Plant Examination (IPE) of internal and external events reports and their updates (in both reports, several enhancements related to severe accident insights were recommended and implemented);
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| and* PNPS PSA model risk significant contributors.
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| (3) Phase I -Preliminary Screening Potential SAMA candidates were screened out if they modified features not applicable to PNPS, if they had already been implemented at PNPS, or if they were similar in nature and could be combined with another SAMA candidate to develop a more comprehensive or plant-specific SAMA candidate.
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| (4) Phase II -Final Screening and Cost Benefit Evaluation The remaining SAMA candidates were evaluated individually to determine the benefits and costs of implementation, as follows.* The total benefit of implementing a SAMA candidate was estimated in terms of averted consequences (benefits estimate).
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| > The baseline PSA model was modified to reflect the maximum benefit of the Improvement.
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| Generally, the maximum benefit of a SAMA candidate was determined with a bounding modeling assumption.
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| For example, if the objective of the SAMA candidate was to reduce the likelihood of a certain failure mode, then eliminating the failure mode from the PSA would bound the benefit, even though the SAMA candidate would not be expected to be 100%effective in eliminating the failure. The modified model was then used to produce a revised accident frequency.
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| Using the revised accident frequency, the method previously described for the four baseline severe accident impact areas was used to estimate the cost associated with each impact area following implementation of the SAMA candidate.
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| 4-33 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage> The benefit in terms of averted consequences for each SAMA candidate was then estimated by calculating the arithmetic difference between the total estimated cost associated with all four impact areas for the baseline plant design and the revised plant design following implementation of the SAMA candidate.
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| The cost of implementing a SAMA was estimated by one of the following methods (cost estimate).
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| > An estimate for a similar modification considered in a previously performed SAMA or SAMDA analysis was used. These estimates were used for comparison against an estimated benefit at PNPS since they were developed in the past and no credit was taken for inflation when applying them to PNPS.In addition, several of them were developed from SAMDA analysis (i.e., during the design phase of the plant), and therefore did not consider the additional costs associated with performing design modifications to an existing plant (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.).> Engineering judgment on the cost associated with procedural changes, engineering analysis, testing, training and hardware modification was applied to formulate a conclusion regarding the economic viability of the SAMA candidate.
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| Q The detail of the cost estimate was commensurate with the benefit. If the benefit was low, it was not necessary to perform a detailed cost estimate to determine if the SAMA was cost beneficial.
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| (5) Sensitivity Analyses Two sensitivity analyses were conducted to gauge the impact of key assumptions upon the analysis.
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| One sensitivity analysis was to investigate the sensitivity of assuming a 27-year period for remaining plant life. The other sensitivity analysis was to investigate the sensitivity of each analysis case to the discount rate of 3.0%.The SAMA analysis for PNPS is presented in the following sections.
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| Attachment E.1 and Attachment E.2 provide a more detailed discussion of the process presented above.4.21.5.1 Establish the Baseline Impacts of a Severe Accident A baseline was established to enable estimation of the risk reductions attributable to implementation of potential SAMA candidates.
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| This severe accident risk was estimated using the PNPS PSA model and the MACCS2 consequence analysis software code. The PSA model used for the SAMA analysis (PNPS Revision 1, April 2003) is an internal events risk model.4-34 Q Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.21.5.1.1 The PSA Model-Level 1 and Level 2 Analysis The PSA model (Level 1 and Level 2) used for the SAMA analysis was the most recent internal events risk model for the PNPS (PNPS Revision 1, April 2003). This current model is an updated version of the model used in the 1992 1PE and subsequently modified in 1995 to answer an RAI and reflects the PNPS configuration and design changes as of September 2001. It also uses component failure and unavailability data as of December 2001, and resolves all findings and observations during the industry peer review of the model, conducted in March 2000. The PNPS model adopts the small event tree/large fault tree approach and uses the CAFTA code for quantifying CDF.An uncertainty analysis associated with internal events CDF was performed.
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| The ratio of the CDF at the 9 5 th percent confidence level to the mean CDF is a factor of 1.62. This analysis is presented in Section E.1.1 of Attachment E.1.The PNPS Level 2 analysis uses a Containment Event Tree (CET) to analyze all core damage sequences identified in the Level 1 analysis.
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| The CET evaluates systems, operator actions, and severe accident phenomena in order to characterize the magnitude and timing of radionuclide release. The result of the Level 2 analysis is a list of sequences involving radionuclide release, along with the frequency and magnitude/timing of release for each sequence.4.21.5.1.2 The PSA External Events Model -Individual Plant Examination of External Events (IPEEE) Model The PNPS IPEEE model was reviewed and used for SAMA analysis.
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| The seismic, high wind, and external flooding analyses determined that the plant is adequately designed to protect against the effects of these natural events. The seismic portion of the IPEEE program was completed in conjunction with the Seismic Qualification Utility Group (SQUG) program. PNPS performed a seismic probabilistic Risk Assessment (PRA) following the guidance of NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991. A number of plant improvements were identified and, as described in NUREG-1 742, Perspectives Gained from the IPEEE Program, Final Report, April 2002, these improvements were implemented.
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| The PNPS fire analysis was performed using the EPRI Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of fire areas and for fire analysis of areas that did not screen. The FIVE methodology is primarily a screening approach used to identify plant vulnerabilities due to fire initiating events. The end result of PNPS IPEEE fire analysis identified the CDF for significant fire areas. A number of administrative procedures were revised to improve combustible and flammable material control.4.21.5.1.3 The MACCS2 Model -Level 3 Analysis A "Level 3" model was developed using the MACCS2 consequence analysis software code to estimate the hypothetical impacts of severe accidents on the surrounding environment and 4-35 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage members of the public. The principal phenomena analyzed were atmospheric transport of radionuclides; mitigation actions (i.e., evacuation, condemnation of contaminated crops and milk)based on dose projection; dose accumulation by a number of pathways, including food and water ingestion; and economic costs. Input for the Level 3 analysis included the core radionuclide inventory, source terms from the PNPS PSA model, site meteorological data, projected population distribution (within 50-mile radius) for the year 2032, emergency response evacuation modeling, and economic data. The MACCS2 input data are described in Section E.1.5 of Attachment E.1.4.21.5.1.4 Evaluation of Baseline Severe Accident Impacts Using the Regulatory Analysis Technical Evaluation Handbook Method This section describes the method used for calculating the cost associated with each of the four impact areas for the baseline case (i.e., without SAMA implementation).
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| This analysis was used to establish the maximum benefit that a SAMA could achieve if it eliminated all risk due to PNPS at-power internal events [Reference 4-6].Off-Site Exposure Costs The Level 3 baseline analysis resulted in an annual off-site exposure risk of 13.6 Person rem. This value was converted to its monetary equivalent (dollars) via application of the$2,000 per person rem conversion factor from the Regulatory Analysis Technical Evaluation Handbook [Reference 4-6]. This monetary equivalent was then discounted to present value using the formula from the same source: 1 e APE = (FSDPS- FADp)R where APE =monetary value of accident risk avoided from population doses, after discounting; R = monetary equivalent of unit dose, ($/person-rem);
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| F = accident frequency (events/year);
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| Dp= population dose factor (person-rem/event);
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| S = status quo (current conditions);
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| A = after implementation of proposed action;r = discount rate (%); and tf = license renewal period (years).4-36 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Using a 20-year license renewal period, a 7.0% discount rate, assuming FA is zero, and the baseline CDF of 6.41 E-06/year resulted in the monetary equivalent value of$292,751.
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| This value is presented in Table 4-3.Off-Site Economic Costs The Level 3 baseline analysis resulted in an annual off-site economic risk monetary equivalent of $45,900. This value was discounted in the same manner as the public health risks in accordance with the following equation: AOC= (FSPDs-FAPDA) 1 e where AOC =monetary value of risk avoided from off-site property damage, after discounting; PD = off-site property loss factor ($/event);
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| F = accident frequency (events/year);
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| S = status quo (current conditions);
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| A = after implementation of proposed action;r = discount rate (%); and= license renewal period (years).Using previously defined values, the resulting monetary equivalent is $494,017.
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| This value is presented in Table 4-3.On-Site Exposure Costs The values for occupational exposure associated with severe accidents were not derived from the PSA model, but from information in the Regulatory Analysis Technical Evaluation Handbook [Reference 4-6]. The values for occupational exposure consist of"immediate dose" and "long-term dose." The best estimate value provided for immediate occupational dose is 3,300 person rem, and long-term occupational dose is 20,000 person-rem (over a 10 year clean-up period). The following equations were used to estimate monetary equivalents.
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| 4-37 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Immediate Dose W 1 0= (FSDIOS-FADIOA)R r (1)where W0= monetary value of accident risk avoided from immediate doses, after discounting; 10 = immediate occupational dose;R = monetary equivalent of unit dose, ($/person-rem);
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| F = accident frequency (events/year);
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| D0= immediate occupational dose (person-rem/event; S = status quo (current conditions);
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| A = after implementation of proposed action;r = discount rate (%); and tf = license renewal period (years).Cf 'The values used in the analysis were R = $2,000/person rem;r= 0.07;D0= 3,300 person rem /accident; and tf = 20 years.For the basis discount rate, assuming FA is zero, the bounding monetary value of the immediate dose associated with PNPS's accident risk is= (FSDO,)R 1-Wo ,r-0.07 x 20 Wio = 3,300 x Fs x $2, 000 x 0 0 7 0$077 W = ($7.1lOxl)F 4-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage For the baseline CDF, 6.41 x 104/year, WJO = $455 Long-Term Dose e-rt, erm WLTO (FSDLTO,-FADLTOA)RX r x (2)where WLTO =monetary value of accident risk avoided long-term doses, after discounting
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| ($);LTO = long-term occupational dose;m = years over which long-term doses accrue;R = monetary equivalent of unit dose, ($/person-rem);
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| F = accident frequency (events/year);
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| DLTO = long-term occupational dose (person-rem/event);
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| S = status quo (current conditions);
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| A = after implementation of proposed action;r = discount rate (%); and 4= license renewal period (years).The values used in the analysis were R= $2,000/person rem;r 0.07;DLTO 20,000 person-rem
| |
| /accident; m 10 years; and t = 20 years.4-39 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage For the basis discount rate, assuming FA is zero, the bounding monetary value of the long-term dose associated with PNPS's accident risk is 1-et' Iarm WLTO = (FSDLTOs)Rx rx1 rm-007 x 20 -0.07 x 10 1-e 1e WLTO = (FS X 20, 000)$2,000 x 0.07 0.07 x10 WLTO = ($3.10 x 108 )Fs For the CDF for the baseline, 6.41 x 104/year, WLTO = $1,985.Total Occupational Exposures Combining equations (1) and (2) above, using delta (A) to signify the difference in accident frequency resulting from the proposed actions, and using the above numerical values, the long-term accident related on-site (occupational) exposure avoided is AOE = AWIO+AWLTO
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| ($)where AOE = on-site exposure avoided.The bounding value for occupational exposure (AOE3) is AOEB = W/o+ WLTO = $455 + $1,985 = $2,440 The resulting monetary equivalent of $2,440 is presented in Table 4-3.On-Site Economic Costs Clean-up/Decontarmination The total cost of clean-up/decontamination of a power reactor facility subsequent to a severe accident is estimated in the Regulatory Analysis Technical Evaluation Handbook [Reference 4-6] to be $1.5 x 109. This same value was adopted for 4-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage these analyses.
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| Considering a 10-year cleanup period, the present value of this cost is PVCD = (cD)(1 e ?where PVCD =present value of the cost of cleanup/decontamination; CD = clean-up/decontamination; CCD = total cost of the cleanup/decontamination effort ($);m = cleanup period (years);r = discount rate (%).Based upon the values previously assumed, PVCD ($1.5E+9)(1
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| -e 07 X PVCD= $1.08E+9.This cost is integrated over the term of the proposed license extension as follows: U PV 1-e UCD =PVCD r where, UCD = total cost of clean up/decontamination over the life of the plant.Based upon the values previously assumed, UCD = $1.16E+10.
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| Replacement Power Costs Replacement power costs were estimated in accordance with the Regulatory Analysis Technical Evaluation Handbook [Reference 4-6]. Since replacement power will be needed for the time period following a severe accident, for the remainder of the expected generating plant life, long-term power replacement 4-41 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage calculations have been used. The present value of replacement power was estimated as follows: PVRP = ($1.2X1O 8)(1a-rts)2 where PVRP =present value of the cost of replacement power for a single event;tf = license renewal period (years); and r = discount rate (%).The $1 .2x1 08 value has no intrinsic meaning but is a substitute for a string of non-constant replacement power costs that occur over the lifetime of a T generic" reactor after an event. This equation was developed in the Regulatory Analysis Technical Evaluation Handbook [Reference 4-6] for discount rates between 5%and 10% only.Based upon the values previously assumed, V , )(1 e ) = t($1 2xI 0(1 -e-(OO7)(2O))
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| 2 $9.73x1 8 To account fbr the entire lifetime of the facility, URP was then calculated from PVRP as follows: UR= (PVRP) ( e)where URP = present value of the cost of replacement power over the remaining life;= license renewal period (years); and r = discount rate (%).Based upon the values previously assumed, 4-42 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage U =(PRp) rtt) = ( 4f -e 8 = $7.89x10 r .0.0 O-URP er} -____Total On-Site Property Damage Costs Combining the cleanup/decontamination and replacement power costs, using delta (AF) to signify the difference in accident frequency resulting from the proposed actions, and using the above numerical values, the best-estimate value of averted occupational exposure can be expressed as 10 9 10 AOSC = AF(UCD+URP)
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| = AF($1.16x10
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| +$7.89x10 ) = AF($1.95x101
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| )where AF = difference in annual accident frequency resulting from the proposed action.For the baseline CDF, 6.41 x106/year, AOSC = $125,086.The resulting monetary equivalent of $125,086 is presented in Table 4-3.Table 4-3 Estimated Present Dollar Value Equivalent of Internal Events CDF at PNPS Parameter Present Dollar Value (S)Off-site exposure costs $292,751 Off-site economic costs $494,017 On-site exposure costs $2,440 On-site economic costs $125,086 Total $914,294 4.21.5.2 Identify SAMA Candidates Based on a review of industry documents, an initial list of SAMA candidates was identified.
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| Since PNPS is a typical GE boiling water reactor design, considerable attention was paid to the SAMA candidates from SAMA analyses for other plants with a GE boiling water reactor design.Attachment E lists the specific documents from which SAMA candidates were initially gathered.4-43 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Q_In addition to SAMA candidates identified from the review of industry documents, additional SAMA candidates were obtained from plant-specific sources, such as the PNPS IPE and IPEEE.In both the IPE and IPEEE, several enhancements related to severe accident insights were recommended and implemented.
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| These enhancements were included in the comprehensive list of SAMA candidates and were verified to have been implemented during preliminary screening.
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| The current PNPS PSA model was used to identify plant-specific modifications for inclusion in the comprehensive list of SAMA candidates.
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| The risk significant terms from the PSA model were reviewed for similar failure modes and effects that could be addressed through a potential enhancement to the plant. The correlation between candidate SAMAs and the risk significant terms are listed in Table E.1-2 of Attachment E.1. The comprehensive list contained a total of 281 SAMA candidates.
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| The first step in the analysis of these candidates was to eliminate the non-viable SAMA candidates through preliminary screening.
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| 4.21.5.3 Preliminary Screening (Phase I)The purpose of the preliminary SAMA screening was to eliminate from further consideration enhancements that were not viable for implementation at PNPS. Potential SAMA candidates were screened out if they modified features not applicable to PNPS or if they had already been implemented at PNPS. In addition, where it was determined those SAMA candidates were potentially viable, but were similar in nature, they were combined to develop a more comprehensive or plant-specific SAMA candidate.
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| During this process, 222 of the 281 initial SAMA candidates were eliminated, leaving 59 SAMA candidates for further analysis.
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| The list of original 281 SAMA candidates and applicable screening criterion is available in on-site documentation.
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| 4.21.5.4 Final Screening and Cost Benefit Evaluation (Phase II)A cost/benefit analysis was performed on the remaining SAMA candidates.
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| The method for determining if a SAMA candidate was cost beneficial consisted of determining whether the benefit provided by implementation of the SAMA candidate exceeded the expected cost of implementation (COE). The benefit was defined as the sum of the reduction in dollar equivalents for each severe accident impact area (off-site exposure, off-site economic costs, occupational exposure, and on-site economic costs). If the expected implementation cost exceeded the estimated benefit, the SAMA was not considered to be cost beneficial.
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| The result of implementation of each SAMA candidate would be a change in the severe accident risk (i.e., a change in frequency or consequence of severe accidents).
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| The method of calculating the magnitude of these changes is straightforward.
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| First, the severe accident risk after implementation of each SAMA candidate was estimated using the same method as for the baseline.
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| The results of the Level 2 model were combined with the Level 3 model to calculate these post-SAMA risks. The results of the benefit analyses for the SAMA candidates are presented in Table E.2-1 of Attachment E.2.4-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Each SAMA evaluation was performed in a bounding fashion. Bounding evaluations were performed to address the generic nature of the initial SAMA concepts.
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| Such bounding calculations overestimate the benefit and thus are conservative calculations.
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| For example, one SAMA dealt with installing digital large break LOCA protection; the bounding calculation estimated the benefit of this improvement by total elimination of risk due to large break LOCA (see the Phase II analysis of SAMA 52 in Table E.2-1). Such a calculation obviously overestimated the benefit, but if the inflated benefit indicated that the SAMA is not cost beneficial, then the purpose of the analysis was satisfied.
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| As described above for the baseline, values for avoided public and occupational health risk were converted to a monetary equivalent (dollars) via application of the Regulatory Analysis Technical Evaluation Handbook (Reference 4-6] conversion factor of $2,000 per person rem and discounted to present value. Values for avoided off-site economic costs were also discounted to present value. The formula for calculating net value for each SAMA was Net value =($APE + $AOC + $AOE + $AOSC) -COE where$APE =value of averted public exposure ($);$AOC =value of averted off-site costs ($);$AOE =value of averted occupational exposure ($);$AOSC = value of averted on-site costs ($); and COE = cost of enhancement
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| ($).If the net value of a SAMA was negative, the cost of the enhancement was greater than the benefit and the'SAMA was not cost beneficial.
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| The SAMA analysis considered that external events (including fires and seismic events) could lead to potentially significant risk contributions.
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| To account for the risk contribution from external events and uncertainties, the cost of SAMA implementation was compared with a benefit value calculated by applying a multiplier of six to the internal events estimated benefit. This value is defined as an upper bound estimated benefit. This treatment accounts for the impact of external events and uncertainty associated with the internal events.The IPEEE analyses using the FIVE methodology and seismic PSA provide quantitative, but conservative results. Therefore, the results were combined as described below to represent the total external events risk.The conservative EPRI FIVE methodology was used for the PNPS IPEEE fire analysis.
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| The fire analysis was done as a screening analysis only and not as a determination of the fire CDF at PNPS. Since fire zone conditional core damage probability is estimated by failing all equipment 4-45 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage in the fire zone, a SAMA that reduces internal events CDF may not reduce fire CDF for a zone.Thus the resulting benefit value is inflated and therefore, overly conservative.
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| The sum of the fire zone CDF values (Table E.1-12) is approximately 1.91 x 10-5 per reactor-year. This value is lower than the originally published fire CDF value of 2.20 x 10-5 due to updated equipment failure probability and unavailability values. As described above, this fire CDF is only a screening value. A more realistic fire CDF may be about a factor of three less than this value [Reference 4-8]. With a factor of three reduction, the fire cDF is about 6.37 x 10-6 per reactor-year.
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| The seismic PSA analysis is also a conservative analysis.
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| Therefore, its results should not be compared directly with the best-estimate internal events results. Conservative assumptions in the seismic PSA analysis include the following.
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| * Each of the sequences in the seismic PSA assumes unrecoverable loss of off-site power.If off-site power were maintained, or recovered, following a seismic event, there would be many more systems available to maintain core cooling and containment integrity than are presently credited in the analysis.* Each of the sequences in the seismic PSA assumes unrecoverable loss of the nitrogen system and the fire water crosstie to the RHR system.* Each of the sequences in the seismic PSA assumes unrecoverable loss of the CSTs water source for the high pressure injection systems.* A single, conservative, surrogate element whose failure leads directly to core damage is used in the seismic risk quantification to model the most seismically rugged components.
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| * Dual initiators are included in the seismic small LOCA, medium LOCA, large LOCA, and ISLOCA event trees. For example, the seismic small LOCA initiating event frequency is a combination of the probability that the seismic event induced a small LOCA and the probability that a small LOCA will occur due to a random event during the 24-hour mission time.* The ATWS event tree was conservatively simplified so that all conditions which lead to a failure to scram result in core damage, without the benefit of standby liquid control (SLC)or other mitigating systems.* Because there is little industry experience with crew actions following seismic events, human actions were conservatively characterized.
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| The seismic CDF in the IPEEE was conservatively estimated to be 5.82x10-5 per reactor-year.
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| The seismic CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, core spray, 4-46 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage and RHR areas; to update random component failure probabilities; and to model replacement of certain relays with a seismically rugged model. The new seismic CDF is 3.22x10-5 per reactor-year. As described above, this is a conservative value. Engineering judgment indicates that a more realistic value would be at least a factor of two less than this value. With a factor of two reduction, the seismic CDF is 1.61x10-5 per year.Combination of the reduced fire and seismic CDF values results in an external events risk estimate of 2.25x10-5 per year, which is 3.51 times higher than the internal events CDP. This would justify use of a multiplier of four on the averted cost estimates (for internal events) to represent the additional SAMA benefits in external events.CDF uncertainty calculations resulted in a factor of 1.62 (Table E.1-3). Since 3.51 x 1.62 = 5.69, a multiplier of six would be reasonable to account for both external events and uncertainties.
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| Use of an upper bound estimated benefit is considered appropriate because of the inherent conservatism in the external events modeling approach and conservative assumptions in benefit modeling of individual SAMA candidates.
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| In addition, not all potential enhancements would be impacted by an external event. In some cases an external event would only impose partial failure of systems or trains. Therefore, using six times the internal events estimated benefit to account for external events and uncertainty is conservative.
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| The expected Cost of Implementation (COE) of each SAMA was established from existing estimates of similar modifications combined with engineering judgment.
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| Most of the cost estimates were developed from similar modifications considered in previous performed SAMA and SAMDA analyses.
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| In particular, these cost-estimates were derived from the following major sources.* GE ABWR SAMDA Analysis* Peach Bottom SAMA Analysis* Quad Cities SAMA Analysis* Dresden SAMA Analysis* ANO-2 SAMA Analysis A number of additional conservatisms associated with implementation were included in the cost benefit analysis.
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| The cost estimates for implementing the SAMAs did not include the cost of replacement power during extended outages required to implement the modifications, nor did they include contingency costs associated with unforeseen implementation obstacles.
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| Estimates based on modifications that were implemented or estimated in the past were presented in terms of dollar values at the time of implementation and were not adjusted to present-day dollars. In addition, several of the implementation cost estimates were originally developed for SAMDA analyses (i.e., during the design phase of the plant), and therefore do not capture the additional 4-47 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage costs associated with performing design modifications to existing plants (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.).Detailed cost estimates were often not required to make informed decisions regarding the economic viability of a potential plant enhancement when compared to attainable benefit.Implementation costs for several of the SAMA candidates were clearly in excess of the attainable benefit estimated from a particular analysis case. For less clear cases, engineering judgment was applied to determine if a more detailed cost estimate was necessary to formulate a conclusion regarding the economic viability of a particular SAMA. Nonetheless, the cost of SAMA candidates was conceptually estimated to the point where conclusions regarding the economic viability of the proposed modification could be adequately gauged. The cost-benefit comparison and disposition of each of the 59 Phase II SAMA candidates is presented in Table E.2-1 of Attachment E.2.4.21.5.5 Sensitivity Analysis Two sensitivity analyses were conducted to gauge the impact of key assumptions upon the analysis.
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| The main factors affecting present worth are the extended plant life and the discount rate. A description of each follows.Sensitivity Case 1: Years Remaining Until End of Plant Life The purpose of this sensitivity case was to investigate the sensitivity of assuming a 27-year period for remaining plant life (i.e. seven years on the original plant license plus the C;20-year license renewal period). The 20-year licensing renewal period was used in the base case. The resultant monetary equivalent for internal event was calculated by using 27 years remaining until end of facility life to investigate the impact on each analysis case.Sensitivity Case 2: Conservative Discount Rate The purpose of this sensitivity case was to investigate the sensitivity of each analysis case to the discount rate. The discount rate of 7.0% used in the base case analyses is conservative relative to corporate practices; nonetheless, a lower discount rate of 3.0%was assumed in this case to investigate the impact on each analysis case.The benefits estimated for each of these sensitivities are presented in Table E.2-2 of Attachment E.2.4.21.6 Conclusion This analysis addressed 281 SAMA candidates for mitigating severe accident impacts. Phase I screening eliminated 222 SAMA candidates from further consideration, based on either inapplicability to PNPS's design or features that had already been incorporated into PNPS's current design, procedures and/or programs.
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| During the Phase II cost benefit evaluation of the 4-48 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage remaining 59 SAMA candidates, an additional 54 SAMA candidates were eliminated because their cost was expected to exceed their benefit and were therefore determined not to be cost beneficial.
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| Five Phase II SAMA candidates (30, 34, 56, 57, and 58) presented in Table 4-4 were found to be potentially cost beneficial for mitigating the consequences of a severe accident for PNPS.* A plant modification and procedural change was recommended to install keylocked control switches to enable AC bus cross-ties to enhance the reliability of AC power system (SAMA candidate 30).* A plant procedural enhancement was recommended to use DC bus cross-ties to enhance the reliability of DC power system (SAMA candidate 34).* A plant modification was recommended to install additional fuses in panel C7 to enable the DTV valve function during loss of containment heat removal accident sequences (SAMA candidate 56).* A plant procedural enhancement was recommended to allow use of the hydro turbine in the event that EDG A or fuel oil transfer pump P-141A is unavailable (SAMA candidate 57).* A plant procedural enhancement was recommended to allow alternately feeding B1 loads via B3 when A3 is available and alternately feeding B2 loads via B4 when A4 is available (SAMA candidate 58).These SAMA candidates do not relate to adequately managing the effects of aging during the period of extended operation.
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| In addition, since the SAMA analysis is conservative and is not a complete engineering project cost-benefit analysis, it does not estimate all of the benefits or all of the costs of a SAMA. For instance, it does not consider increases or decreases in maintenance or operation costs following SAMA implementation.
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| Also, it does not consider the possible adverse consequences of procedure changes, such as additional personnel dose. Therefore, the above, potentially cost-beneficial SAMAs have been submitted for engineering project cost-benefit analysis.The sensitivity studies indicated that the results of the analysis would not change for the conditions analyzed.4-49
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| )J 3 9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-4 Final SAMAs Off-siteUpe Phase II SM il esl fPtnia nacmn CDF Estimated Bound Estimated SAAID Title Resultof Potential Enhancement Reduction Dose Benefit Estimated Cost Reduction Benefit 030 9.g. Enhance SAMA would provide increased 11.10% .8.47% $78,902 $473,410 $146,120 procedures to make reliability of AC power system and use of AC bus cross- reduce core damage and release ties. frequencies.
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| Basis for
| |
| | |
| == Conclusion:==
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| | |
| The CDF contribution due to loss of MCC B17, B18, and B15 was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $146,120 by engineering judgment.034 10.d. Enhance This SAMA would improve DC 4.65% :1.91% $19,761 $118,568 $13,000 procedures to make power availability.
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| use of DC bus E coss-t i e s .I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of DC buses D16 and D17 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $13,000 by engineering judgment.056 Provide redundant This SAMAwould improve reliability 8.81% 3.51% $36,773 $220,639 $112,400 DC power supplies to of the DTV valves and enhance DTV valves. containment heat removal capability.
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| _ __Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving DC power supply failures to the direct torus vent valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $112,400 by engineering judgment.4-50 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 4-4 Final SAMAs CDasOff-site Estimated Bupperimte Phases SAMA Title Result of Potential Enhancement Reduction Dose Benefit Estimated Cost SAMA ID.Reduction Benefit 057 Proceduralize use of This SAMA would increase 2.25% 3.14% $29,213 $175,279 $26,000 the diesel fire pump capability to provide makeup to the hydro turbine in the fire pump day tank to allow event of EDG A continued operation of the diesel fire failure or pump, without dependence on unavailability.
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| electrical power.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving loss of offsite power and failure of either EDG A, or the EDG A fuel oil transfer oil pump, was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $26,000 by engineering judgment.058 Proceduralize the This SAMA would provide the 4.92% 3.14% $31,799 $190,797 $50,000 operator action to direction to restore B15 and B17 feed B1 loads via B3 loads upon loss of A5 initiating When AS is events as long as A3 is available.
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| unavailable post-trip.
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| Additionally, it would provide the Similarly, feed B2 direction to restore B14 and B18 loads via B4 when A6 loads upon loss of A6 initiating is unavailable post events- as long as A4 is available.
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| trip.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The COF contribution from sequences involving loss of 4160VAC safeguard bus AS was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $50,000 by engineering judgment.4-51 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.22 Environmental Justice 4.22.1 Description of Issue Environmental Justice 4.22.2 Finding from Table B-I, Appendix B to Subpart A"The need for and the content of an analysis of environmental justice will be addressed in plant-specific reviews." 4.22.3 Requirement Other than the above referenced finding, there is no requirement concerning environmental justice in 10 CFR 51.4.22.4 Background The following background information is from the Regulatory Guide 4.2.Environmental justice was not reviewed in NUREG-1437.
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| Executive Order 12898, "Federal Actions To Address Environmental Justice in Minority Populations and Low-Income Populations," issued on February 11, 1994, is designed to focus the attention of Federal agencies on the human health and environmental conditions in minority and low-income communities.
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| The NRC Office of Nuclear Reactor Regulation (NRR) is guided in its consideration of environmental justice by Attachment 4, 'NRR Procedures for Environmental Justice Reviews," to NRR Office Letter No. 906, Revision 2, "Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues," September 21, 1999. NRR Office Letter No. 906 is revised periodically.
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| The environmental justice review involves identifying off-site environmental impacts, their geographic locations, minority and low-income populations that may be affected, the significance of such effects and whether they are disproportionately high and adverse compared to the population at large within the geographic area, and if so, what mitigative measures are available, and which will be implemented.
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| The NRC staff will perform the environmental justice review to determine whether there will be disproportionately high human heath and environmental effects on minority and low-income populations and report the review in its SEIS. The staff's review will be based on information provided in the ER and developed during the staffs site-specific scoping process., NRR Office Letter No. 906, Revision 2 [Reference 4-7] contains a procedure for incorporating environmental justice into the licensing process. Entergy used this process in conducting the review and analysis of this issue.4.22.5 Analysis The consideration of environmental justice is required to assure that federal programs and activities will not have "disproportionately high and adverse human health or environmental 4-52 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage effects.. .on minority populations and low income populations...." Entergy's analyses of the Category 2 issues defined in 10 CFR 51.53(c)(3)(ii) determined that there were no adverse impacts from the renewal of the PNPS license; thus, no disproportionate impact on minority or low income populations would occur from the proposed action. If replacement of the electricity generated by PNPS with fossil-fuel sources was considered as an alternative to the proposed action, the environmental justice ramifications of that alternative's air emissions and other environmental impacts would need to be considered.
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| Based on the review of these issues, no review for environmental justice Is necessary.
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| However, Entergy presents environmental justice demographic information in Section 2.6.2 of this ER to assist the NRC in its review.4.22.6 Conclusion As part of its environmental assessment of this proposed action, Entergy has determined that the environmental impacts of renewing the PNPS license are small. This conclusion is supported by the review performed of the Category 2 issues defined in 10 CFR 51.53(c)(3)(ii) presented in this ER.Because all impacts are small, and because there are few low-income or minority populations in the environmental impact area, there can be no disproportionately high and adverse impacts or effects on members of the public, including'minority and low-income populations, resulting from the renewal of the PNPS license.4-53 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 4.23 References 4-1 Dayian, L., Water System Data Provided by the Massachusetts Environmental Protection Department, personal communication with M. Hoganson, TtNUS, June 4, 2001.4-2 ENSR Corporation, Redacted Version 316 Demonstration Report -Pilgrim Nuclear Power Station, Document Number 0970-021-200, prepared for Entergy Nuclear Generation Company, March 2000.4-3 Commonwealth of Massachusetts, Executive Office of Environmental Affairs, Department of Environmental Protection, Southeast Regional Office, Groundwater Discharge Permit, SE #2-389, April 26,1999.4-4 Electric Power Research Institute, Transmission Line Reference Book: 345 kV and Above, 2nd Edition, Palo Alto, CA, 1982.4-5 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes I and 2, Washington, DC, May 1996.4-6 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, Washington, DC, January 1997.4-7 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,'Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues," NRR Office Instruction No. LIC-203, Revision 1, May 24, 2004.4-8 U.S. Nuclear Regulatory Commission, NUREG-1437, Supplement 19, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Arkansas Nuclear One, Unit 2, Washington, DC, April 2005.4-9 NSTAR, "ENVIRO printouts," facsimile from B. Connors, NSTAR, to D. Thrall, Entergy, April 9, 2001.4-10 Prescott, R., Email correspondence with J. Brochu, Entergy, January 15, 2005.4-11 TetraTech NUS, Calculation of Induced Current for the License Renewal Environmental Report -Pilgrim Nuclear Power Station," Aiken, SC, April 23, 2001.4-12 U.S. Environmental Protection Agency, "National Pollutant Discharge Elimination System -Final Regulations to Establish Requirement for Cooling Water Intake Structures at Phase II Existing Facilities," 69 FR 41576, July 9, 2004.4-54 0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 5.0 ASSESSMENT OF NEW AND SIGNIFICANT INFORMATION"The environmental report must contain any new and significant information regarding the environmental impacts of license renewal of which the applicant is aware." 10 CFR 51.53(c)(3)(iv)]
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| The NRC has resolved most license renewal environmental issues generically and only requires an applicant to analyze those issues the NRC has not resolved generically.
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| While NRC -regulations do not require an applicant's environmental report to contain analyses of the impacts of those environmental issues that have been generically resolved [10 CFR 51.53(c)(3)(i)], the regulations do require that an applicant identify any new and significant information of which the applicant is aware [10 CFR 51.53(c)(3)(iv)].
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| Entergy implemented a process to identify the following:
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| * information that identifies a significant environmental issue not covered in the NRC's GEIS and codified in the regulation, or* information not covered in the GEIS analyses that leads to an impact finding different from that codified in the regulation.
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| The term "significant" is not specifically defined by the NRC. For its review, Entergy used guidance available in Council on Environmental Quality (CEQ) regulations.
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| The NEPA 114 authorizes CEQ to establish implementing regulations for federal agency use. The NRC requires license renewal applicants to provide the NRC with input, in the form of an environmental report, that the NRC will use to meet NEPA requirements as they apply to license renewal (10 CFR 51.10).CEQ guidance provides that federal agencies should prepare environmental impact statements for actions that would significantly affect the environment (40 CFR 1502.3), focus on significant environmental issues (40 CFR 1502.1), and eliminate from detailed study issues that are not significant
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| [40 CFR 1501.7(a)(3)J.
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| The CEQ guidance includes a lengthy definition of"significantly" that requires consideration of the 'context of the action and the intensity or severity of the impact(s)
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| (40 CFR 1508.27).
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| Entergy expects that MODERATE or LARGE impacts, as defined by the NRC, would be significant.
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| Section 4 presents the NRC definitions of MODERATE and LARGE impacts.Entergy reviewed SEISs associated with other license renewal applications to determine if there were new issues identified for those plants that may be applicable to PNPS. In addition, some regulatory agencies were consulted regarding new and significant information.
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| However, Entergy has an ongoing assessment process for identifying and evaluating new and significant information that may affect programs at the Entergy nuclear sites, including those related to license renewal matters.5-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage This process is directed in a joint effort by the nuclear corporate support group responsible for environmental matters, with assistance from environmental focus group members composed of technical personnel from the Entergy Nuclear South and Entergy Nuclear Northeast sites. A summary of this process follows.Issues relative to environmental matters are identified as follows:> participation in industry utility groups (i.e., EEI, EPRI, NEI, and USWAG);> participation in non-utility groups (i.e., Institute of Hazardous Materials Management and National Registry of Environmental Professionals);
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| > periodic reviews of proposed regulatory changes;> Entergy Nuclear Environmental Focus Group meetings; and> environmental issues are reviewed and evaluated for applicability by the nuclear corporate support group.* If the issue is applicable to the Entergy nuclear sites, it is then further evaluated by the nuclear corporate support group and environmental focus group that consist of technical personnel involved in environmental compliance, environmental monitoring, environmental planning, natural resource management, and health and safety issues. (, Necessary changes are made to the program and implemented in accordance with site and corporate procedures.
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| Additional actions incorporated into this assessment process specifically for PNPS license renewal include the following:
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| * review of documents related to environmental issues at PNPS;* review of internal procedures for reporting to the NRC events that could have environmental impacts; and* credit for the oversight provided by inspections of plant facilities by state and federal regulatory agencies.As a result of this assessment, Entergy is aware of no new and significant information regarding the environmental impacts of PNPS license renewal.5-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 6.0
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| | |
| ==SUMMARY==
| |
| OF LICENSE RENEWAL IMPACTS AND MITIGATING ACTIONS 6.1 License Renewal Impacts Entergy has reviewed the environmental impacts of renewing the PNPS operating license and has concluded that all impacts would be small and would not require mitigation.
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| This ER documents the basis for Entergy's conclusion.
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| Section 4 incorporates by reference NRC findings for the 49 Category 1 issues that apply to PNPS, all of which have impacts that are small (Table 4-2). The rest of Section 4 analyzes Category 2 issues, all of which are either not applicable or have impacts that would be small. Table 6-1 identifies the impacts that PNPS license renewal would have on resources associated with Category 2 issues.6.2 Mitigation 6.2.1 Requirement
| |
| [10 CFR 51.53(c)(3)(iii)]"The report must contain a consideration of alternatives for reducing adverse impacts, as required by § 51.45 (c), for all Category 2 license renewal issues in Appendix B to subpart A of this part. No such consideration is required of Category 1 issues in Appendix B to subpart A of this part." 6.2.2 Entergy Response i" As discussed in Supplement 1 to Regulatory Guide 4.2, "Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses," when adverse environmental effects are identified, 10 CFR 51.45(c) requires consideration of alternatives available to reduce or avoid these adverse effects. Furthermore, Regulatory Guide 4.2 states, "Mitigation alternatives are to be considered no matter how small the adverse impact;however, the extent of the consideration should be proportional to the significance of the impact"[Reference 6-2].As described in Section 6.1 and as shown in Table 6-1, analysis of the Category 2 issues found the impacts to be small for the applicable issues. For these issues, the current permits, practices, and programs that mitigate the environmental impacts of plant operations are adequate.
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| This ER finds that no additional mitigation measures are sufficiently beneficial as to be warranted.
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| 6-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Q,`Table 6-1 Environmental Impacts Related to License Renewal at PNPS Surface Water Quality, Hydrology and Use (for All Plants)Water use conflicts (plants with cooling ponds NONE. This issue does not apply because PNPS or cooling towers using make-up water from a does not use cooling ponds or cooling towers small river with low flow) withdrawing water from a small river.10 CFR 51.53(c)(3) (ii)(A)Aquatic Ecology (for All Plants with Once-Through and Cooling Pond Heat Dissipation Systems)Entrainment of fish and shellfish SMALL. PNPS has a current NPDES permit which 10 CFR 51.53(c)(3)(ii)(B) constitutes compliance with CWA Section 316(b)requirements.
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| Impingement of fish and shellfish 10 CFR SMALL. PNPS has a current NPDES permit which 51.53(c)(3)(ii)(B) constitutes compliance with CWA Section 316(b)requirements.
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| Heat shock SMALL. PNPS has a current NPDES permit which 10 CFR 51.53(c)(3)(ii)(B) constitutes compliance with CWA Section 316(a)requirements.
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| Ground-water Use and Quality Groundwater use conflicts (plants using >100 NONE. This issue does not apply because PNPS gpm of ground-water) uses <100 gpm of groundwater.
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| PNPS's potable 10 CFR 51.53(c)(3)(ii)(C) water is supplied by the Town of Plymouth.Groundwater use conflicts (plants using NONE. This issue does not apply because PNPS cooling towers withdrawing make-up water does not use cooling towers withdrawing water from a from a small river) small river.10 CFR 51.53(c)(3)(ii)(A)
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| Groundwater use conflicts (Ranney Wells) NONE. PNPS does not use Ranney Wells.10 CFR 51.53(c)(3)(ii)(C)
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| Consideration of mitigation is not required.Degradation of groundwater quality NONE. PNPS does not use cooling ponds.10 CFR 51.53(c)(3)(ii)(D)
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| Consideration of mitigation is not required.Terrestrial Resources Refurbishment impacts on terrestrial NONE. No major refurbishment activities identified.
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| resources Consideration of mitigation is not required.10 CFR 51.53(c)(3)(ii)(E) 0.6-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 6-1 Environmental Impacts Related to License Renewal at PNPS (Continued)
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| Threatened or Endangered Species (for All Plants)Threatened or endangered species SMALL. No major refurbishment activities have been 10 CFR 51.53(c)(3)(ii)(E) identified and no significant issues have been identified by any of the environmental agencies that were consulted.
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| Air Quality Air quality during refurbishment NONE. No impacts are expected because PNPS has 10 CFR 51.53(c)(3)(ii)(F) no plans to undertake refurbishment.
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| Human Health Microbiological (Thermophilic)
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| Organisms NONE. The issue does not apply because PNPS 10 CFR 51.53(c)(3)(ii)(G) does not discharge to a lake or use cooling towers or cooling ponds discharging to a small river.Electromagnetic fields -Acute effects SMALL. The largest modeled induced current under 10 CFR 51.53(c)(3)(ii)(H) the PNPS transmission lines would be less than 5,0.milliamperes, which is the National Electric Safety Code standard for preventing electric shock from induced current.Socio'economics Housing impacts SMALL. PNPS is located in a high-population area 10 CFR 511.53(c)(3)(ii)(1) that does not have growth control measures.Therefore, in accordance with NRC standards, housing impacts would be small. No major refurbishment activities identified.
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| Entergy does not anticipate an increase in employment during period of extended operation.
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| Therefore, there no additional impacts to housing are expected due to continued operations of PNPS. Consideration of mitigation is not required.Public utilities:
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| public water supply availability SMALL. No major refurbishment activities identified 10 CFR 51.53(c)(3)(ii)(1) and no additional workers anticipated during the period of extended operation.
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| Public water systems near PNPS have adequate system capacity to meet demand of residential and industrial customers in the area. Consideration of mitigation is not required.Education impacts from refurbishment NONE. No major refurbishment activities identified.
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| 10 CFR 51.53(c)(3)(ii)(1)
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| Consideration of mitigation is not required.6-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 6-1 Environmental Impacts Related to License Renewal at PNPS (Continued)
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| Offsite land use (effects of refurbishment NONE. No major refurbishment activities identified.
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| activities)
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| Consideration of mitigation is not required.10 CFR 51.53(c)(3)(ii)(1)
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| Offsite land use (effects of license renewal) SMALL. No plant-induced changes to offsite land use 10 CFR 51.53(c)(3)(ii)(1) are expected from license renewal.Local transportation impacts SMALL. No major refurbishment activities identified 10 CFR 51.53(c)(3)(ii)(J) and no increases in total number of employees during the period of extended operation.
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| Consideration of mitigation is not required.Historic and archaeological properties SMALL. No major refurbishment activities identified 10 CFR 51.53(c)(3)(ii)(K) and no identified adverse impacts or detrimental effects on identified historic and archaeological properties.
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| Consideration of mitigation is not required.Postulated Accidents Severe accident mitigation alternatives SMALL. No impact from continued operation.
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| 10 CFR 51.53(c)(3)(ii)(L)
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| Potentially cost-effective SAMAs are not related to adequately managing the effects of aging during period of extended operation.
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| Consideration of mitigation is not required.(WWI 6.3 Unavoidable Adverse Impacts 6.3.1 Requirement
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| [10 CFR 51.45(b)(2)]
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| The applicant's report shall discuss any adverse environmental effects which cannot be avoided upon implementation of the proposed project.6.3.2 Entergy Response Section 4 contains the results of Entergy's review and the analyses of the Category 2 issues as required by 10 CFR 51.53(c)(3)(ii).
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| These reviews take into account the information that has been provided in the GEIS, 10 CFR 51, Subpart A, Appendix B, and information specific to PNPS.This review and analysis did not identify any significant adverse environmental impacts associated with the continued operation of PNPS. The evaluation of structures and components required by 10 CFR 54.21 has been completed.
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| No plant refurbishment activities, outside the 64 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage bounds of normal plant component replacement and inspections, have been identified to support continued operation of PNPS beyond the end of the -existing operating license. As a result of these reviews and analyses, Entergy is not aware of significant adverse environmental effects that cannot be avoided upon implementation of the proposed project.6.4 Irreversible or Irretrievable Resource Commitments 6.4.1 Requirement
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| [10 CFR 51.45(b)(5)J The applicant's report shall discuss any irreversible and irretrievable commitments of resources which would be involved in the proposed action should it be implemented.
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| 6.4.2 Entergy Response The continued operation of PNPS for the period of extended operation will result in irreversible and irretrievable resource commitments, including the following:
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| * nuclear fuel, which is consumed in the reactor and converted to radioactive waste;* land required to dispose of spent nuclear fuel and low-level radioactive wastes generated as a result of plant operations;
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| * elemental materials that will become radioactive; and* materials used for the normal industrial operations of PNPS that cannot be recovered or recycled or that are consumed or reduced to unrecoverable forms.Other than the above, there are no major refurbishment activities or changes in operation of PNPS during the period of extended operation that would irreversibly or irretrievably commit environmental components of land, water, and air.6.5 Short-Term Use Versus Lonc-Term Productivity 6.5.1 Requirement
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| [10 CFR 51.45(b)(4)]
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| The applicant's report shall discuss the relationship between local short-term uses of man's environment and the maintenance and enhancement of long-term productivity.
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| 6.5.2 Entergy Response The current balance between short-term use and long-term productivity at PNPS was established when the station began operation in 1972. PNPS's FES [Reference 6-1] evaluated the impacts of constructing and operating PNPS. Initially, approximately 500 acres were acquired for the station. The land had been a private, mostly wooded, estate. PNPS and associated facilities cover about one-third of this acreage. When Boston Edison was considering constructing a second reactor on the PNPS site, the company purchased approximately 1,100 6-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ( )additional acres inland of the original 500-acre tract. Approximately 1,500 acres of the approximately 1,600 acres owned by Entergy is managed as timberland.
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| This greenspace in a populated and growing area between two large urban areas provides habitat for plants and animals. After operations cease, most of the land occupied by the station and ancillary facilities could be restored to terrestrial habitat or used for other industrial purposes.Long-term productivity of the terrestrial and aquatic habitats in the vicinity of PNPS is not adversely affected by the station or its operations.
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| Continued operations for an additional 20 years would not alter this conclusion.
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| 6.6 References 6-1 U.S. Atomic Energy Commission, Division of Radiological and Environmental Protection, Final Environmental Statement Related to Operation of Pilgrim Nuclear Power Station, Docket No. 50-293, Washington, DC, 1972.6-2 U.S. Nuclear Regulatory Commission, Supplement I to Regulatory Guide 4.2, Preparation of Supplemental Environmental Reports for Applications to Renew Nuclear Power Plant Operating Licenses, Washington, DC, September 2000.6-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 7.0 ALTERNATIVES CONSIDERED 7.1 Introduction NRC regulations require that an applicant's environmental report discuss alternatives to a proposed action [10 CFR 51.45(b)(3)].
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| The intent of this review is to enable the Commission to consider the relative environmental consequences of the proposed action as compared to the environmental consequences of other activities that also meet the purpose of the proposed action. In addition, this review addresses the environmental consequences of taking no action[Reference 7-1]. For license renewal, there are only two alternatives that meet the purpose of the requirement:
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| not renew the operating license or renew the operating license. The alternatives are discussed below.7.2 Proposed Action PNPS operated at a capacity factor of 98.5% in 2004 and is rated at approximately 715 gross MWe. The proposed action is to renew the operating license for PNPS which would provide the opportunity for Entergy to continue to operate PNPS through the period of extended operation.
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| The review of the environmental impacts required by 10 CFR 51.53(c)(3)(ii) is provided in Section 4 of this ER. Entergy concludes that the environmental impacts of extended PNPS operation would be small.7.3 No-Action Alternative The "no-action alternative" to the proposed action is not to renew the operating license for PNPS.In this alternative, it is expected that PNPS will continue to operate up to the end of the existing operating license, at which time plant operation would cease, and decommissioning would begin.Because PNPS constitutes a significant block of base load capacity, it is reasonable to assume that a decision not to renew the PNPS licenses would necessitate the replacement of its approximately 715 gross MWe with other sources of generation.
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| The environmental impacts of the no-action alternative would be* the environmental impacts from decommissioning the PNPS unit, and* the environmental impacts from a replacement power source.Environmental impacts associated with decommissioning are discussed in Section 7.4.The environmental impacts associated with a replacement power source would be the impacts from the construction and operation of a source of replacement power at a new location (greenfield) or at the PNPS site (brownfield).
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| The environmental impacts of these various types of replacement power are discussed in Section 8.7-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 7.4 Decommissioning Imnacts A nuclear power plant licensee is required to submit decommissioning plans within two years following permanent cessation of operation of a unit or at least five years before expiration of the operating license, whichever occurs first, pursuant to the requirements of 10 CFR 50.54(b).The GEIS defines decommissioning as the safe removal of a nuclear facility from service and the reduction of residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license [Reference 7-1, Section 7.1]. NRC-evaluated decommissioning options include immediate decontamination and dismantlement (DECON), and safe storage of the stabilized and defueled facility (SAFSTOR) for a period of time, followed by decontamination and dismantlement.
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| Regardless of the option chosen, decommissioning must be completed within a 60-year period.Under the no-action alternative, Entergy would continue operating PNPS until the current license expires, then initiate decommissioning activities in accordance with NRC requirements.
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| The GEIS describes decommissioning activities based on an evaluation of an example reactor (the"reference' boiling-water reactor is the 1,155 MWe Washington Public Power Supply System's Columbia Nuclear Power Plant). This is a substantially larger plant than PNPS and, therefore, bounds decommissioning activities that Entergy would conduct at PNPS.As the GEIS notes, the NRC has evaluated environmental impacts from decommissioning.
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| NRC-evaluated impacts include occupational and public radiation dose; impacts of waste management; impacts to air and water quality; and ecological, economic, and socioeconomic C impacts. The NRC indicated in Section 4.3.8 of the Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities
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| [Reference 7-2] that the environmental effects of greatest concern (i.e., radiation dose and releases to the environment) are substantially less than the same effects resulting from reactor operations.
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| Entergy adopts by reference the NRC conclusions regarding environmental impacts of decommissioning.
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| Entergy notes that decommissioning activities and their impacts are not discriminators between the proposed action and the no-action alternative.
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| Entergy will have to decommission PNPS;license renewal would only postpone decommissioning for 20 years. The NRC has established in the GEIS that the timing of decommissioning operations does not substantially influence their environmental impacts. Entergy adopts by reference the NRC findings (10 CFR 51 Subpart A, Appendix B, Table B-1, Decommissioning) to the effect that delaying decommissioning until after the renewal term would have small environmental impacts.Entergy concludes that the decommissioning impacts under the no-action alternative would not be substantially different from those occurring following license renewal, as identified in the GEIS[Reference 7-1, Section 8.4] and in the decommissioning generic environmental impact statement
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| [Reference 7-2, Section 6.0]. These impacts would be temporary and would occur at the same time as the impacts from meeting system generating needs.7-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 7.5 Alternative Enerav Sources Nuclear power plants are commonly used for base-load generation.
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| The GEIS states that coal-fired and gas-fired generation capacity are the feasible alternatives to nuclear power generating capacity, based on current (and expected) technological and cost factors. The following generation alternatives were considered in detail in this ER: '* Coal-fired generation at an alternate site (Section 8.1.1). Entergy did not consider coal-fired generation at the PNPS site since it was concluded that there was not enough land to build a coal-fired unit and a coal yard on the existing site (brownfield).
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| Based on Table 8.1 of the GEIS, it would take approximately 1.7 acres of land per MWe to construct a coal-fired plant. PNPS is situated on 140 acres and is rated at approximately 715 gross MWe. Therefore for the 620 gross MWe coal-fired plant used in this analysis, approximately 1,054 acres of land would be needed.* Natural gas-fired generation at the PNPS site and at an alternate site (Section 8.1.2)* Nuclear generation at an alternate site (Section 8.1.3). Entergy did not consider nuclear generation at the PNPS site (brownfield) since it was concluded that there was not enough land to build a nuclear unit. Based on Table 8.1 of the GEIS, it would take approximately 0.5 to 1.0 acres of land per MWe to construct a nuclear plant. PNPS is situated on 140 acres and is rated at approximately 715 gross MWe. Therefore for a 715 gross MWe nuclear plant, approximately 357.5 to 715 acres of land would be needed.Entergy's experience indicates that, although customized unit sizes can be built, using standardized sizes is more economical.
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| For example, a standard-sized gas-fired combined cycle plant has a net capacity of 585 MWe. The plant consists of two 189-MWe gas turbines and 207 MWe of heat recovery capacity.
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| For comparability, Entergy set the net power of the hypothetical coal-fired unit equal to the hypothetical gas-fired plant (585 MWe). Although both provide less capacity than PNPS (715 MWe), this ensures against overestimating environmental impacts from the alternatives.
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| The shortfall in capacity could be replaced by other methods.These alternatives are presented (Sections 8.1.1, 8.1.2, and 8.1.3, respectively) as if such plants were constructed at the PNPS site, using the existing water intake and discharge structures, switchyard, and transmission lines, or at an alternate location that could be either a current industrial site or an undisturbed, pristine site requiring a new generating building and facilities, new switchyard, and at least some new transmission lines. In this ER, a "greenfield" site is assumed to be an undisturbed, pristine site. Although PNPS does own an additional 1,500 acres of forest land, it is a greenfield site as it is not part of the PNPS facility site. This additional land is zoned as rural residential.
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| Depending on the location of an alternative site, it might also be necessary to connect to the nearest gas pipeline (in the case of natural gas) or rail line (in the case of coal). The requirement for these additional facilities may increase the environmental impacts relative to those that would be experienced at the PNPS site.7-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The potential for using purchased power is discussed in Section 8.1.4. Purchased power is considered feasible, but would result in the transfer of environmental impacts from the current region in Massachusetts to some other location in Massachusetts, another state, or Canadian province.
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| In addition, there is no assurance that the capacity or energy would be available.
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| As stated in NUREG-1437, Vol.1, Section 8.1, the "NRC has determined that a reasonable set of alternatives should be limited to analysis of single, discrete electric generation sources and only electric generation sources that are technically feasible and commercially viable' [Reference 7-1]. Accordingly, the following alternatives were not considered as reasonable replacement power.* wind* solar* hydropower
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| * geothermal
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| * wood energy* municipal solid waste* other biomass-derived fuels* oil* fuel cells* delayed retirement
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| * utility-sponsored conservation
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| * combination of alternatives These technologies were eliminated as possible replacement power alternatives for one or more of the following reasons.* High land-use impacts Some of the technologies listed above (wind, solar, and hydroelectric) would require a large area of land and would thus require a greenfield siting plan. This would result in a greater environmental impact than continued operation of PNPS.* Low capacity factors Some of the technologies identified above (wind, solar, and hydroelectric) are not capable of producing the nearly 715 gross MWe of power at high capacity factors. These generation technologies are used as peaking power sources, as opposed to base-load power sources, and for this reason are not reasonable altematives.
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| * Geographic availability of the resource Some of the technologies are not feasible because there is no feasible location in the area served by PNPS.7-4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* Emerging technology Some of the technologies has not been proven as reliable and cost effective replacements of a large generation facility.
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| Therefore, these technologies are typically used with smaller (lower MWe) generation facilities.
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| * Availability There is no assurance of the availability of purchased power.7.6 References 7-1 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental'Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes 1 and 2, Washington, DC, May 1996.7-2 U.S. Nuclear Regulatory Commission, NUREG-0586, Supplement 1, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, Supplement 1, Regarding the Decommissioning of Nuclear Power Reactors, Washington, DC, November 2002.7-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.0 COMPARISON OF IMPACTS The following key assumptions have been made in the review of alternative energy sources.These key assumptions are intended to simplify the evaluation, yet still allow the no-action alternative review to meet the intent of NEPA requirements and NRC environmental regulations.
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| * The goal of the proposed action (license renewal) is the production of approximately 715 gross MWe of base-load generation.
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| Alternatives that do not meet the goal are not considered in detail.* The time frame for the needed generation is 2012 through 2032.* Purchased power is not considered a reasonable alternative because there is no assurance that the capacity or energy would be available.
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| See Section 8.1.4.* The annual capacity factor of PNPS in 2004 was 98.5%. The capacity factor is targeted to remain at or near this value throughout the plant's operating life.8.1 Comparison of Environmental Impacts for Reasonable Alternatives As stated in the GEIS, the "NRC has determined that a reasonable set of alternatives should be limited to analysis of single, discrete electric generation sources and only electric generation sources that are technically feasible and commercially viable" [Reference 8-14]. Below is a discussion of the supply side alternative energy technologies that Entergy could utilize if the license for PNPS is not renewed. These alternatives are within the range of alternatives capable of meeting the goal of approximately 715 gross MWe as base-load generation (replacement power for PNPS).Conventional coal-fired, natural gas-fired combined cycle, and advanced light water reactor are currently available conventional base-load technologies considered to replace PNPS generation upon its termination of operation.
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| These sources are considered viable alternatives based upon current Entergy planning strategies.
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| The environmental impacts discussed in this chapter are for the construction and operation of these generation facilities.
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| Impacts are evaluated for a greenfield case (building on a new, pristine condition site) and a brownfield case (constructing new generation on the existing PNPS site, in the case of a gas-fired unit).The continued operation of PNPS for the period of extended operation would result in less environmental impact than that of the replacement power that could be obtained from other reasonable generating sources, as described below.8-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.1 Coal-Fired Generation The NRC has evaluated coal-fired generation alternatives in each of the plant-specific supplements to the GEIS. For the Oconee boiling-water reactors, the NRC analyzed 2,500 MWe of coal-fired generation capacity [Reference 8-15]. Entergy has reviewed the NRC analysis, believes it to be sound, and notes that it analyzed substantially more generating capacity than the 620 gross MWe from coal-fired generation discussed in this analysis.
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| In defining the PNPS coal-fired alternative, Entergy has used site-specific input and has scaled from the NRC analysis, where appropriate.
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| Tables 8-1, 8-2, and 8-3 present the basic coal-fired alternative emission control characteristics, emission estimates, and waste generation volumes. Entergy based its emission control technology and percent control assumptions on alternatives that the EPA has identified as being available for minimizing emissions
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| [Reference 8-7]. For the purposes of analysis, Entergy assumed that coal and lime (calcium hydroxide) would be delivered by barge to a newly constructed receiving dock on site.The coal-fired alternative that Entergy has defined would be located at an alternative site.Table 8-1 Coal-Fired Alternative Emission Control Characteristics Characteristic Basis q.Unit size = 585 MWe ISO rating netl Unit size = 620 MWe ISO rating gross 1 Number of units = 1 Boiler type = tangentially fired, dry-bottom Fuel type = bituminous, pulverized coal Fuel heating value = 12,464 Btu/lb Fuel ash content by weight = 8.2%Fuel sulfur content by weight = 0.69%Uncontrolled NOX emission = 10 lb/ton Uncontrolled CO emission = 0.5 lb/ton Calculated to be < PNPS gross capacity (715 MWe)Calculated based on 6% onsite power use Minimizes nitrogen oxide emissions (Reference 8-7, Table 1.1-3)Typical for coal used in Massachusetts 2000 value for coal used in Massachusetts (Reference 8-6, Table 25)2000 value for coal used in Massachusetts (Reference 8-6, Table 25)2000 value for coal used in Massachusetts (Reference 8-6, Table 25)Typical for pulverized coal, tangentially fired, dry-bottom, NSPS (Reference 8-7, Table 1.1-3)8-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-1 Coal-Fired Alternative Emission Control Characteristics (Continued)
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| Characteristic Basis Heat rate = 10,200 Btu/kWh Typical for coal-fired, single-cycle steam turbines (Reference 8-5, page 108)Capacity factor = 0.85 Typical for newer large coal-fired units NOx control = low NOx burners, overtire air and Best available and widely demonstrated for selective catalytic reduction (95% reduction) minimizing NOx emissions (Reference 8-7, Table 1.1-3)Particulate control = fabric filters (baghouse-Best available for minimizing particulate emissions 99.9% removal efficiency) (Reference 8-7, pp. 1.1-6 and -7)SOx control = Wet scrubber -lime (95% Best available for minimizing SOx emissions removal efficiency) (Reference 8-7, Table 1.1-1)1. The difference between 'net and 'gross' is electricity consumed by auxiliary equipment and environmental control devices (Reference 8-5, page 107).Btu = British thermal unit NSPS = New Source Performance Standard ISO rating = Intemational Standards Organization lb = pound rating at standard atmospheric conditions of MW = megawatt 59 0 F, 60% relative humidity, and 14.696 NOx- nitrogen oxides pounds of atmospheric pressure per square SOx = oxides of sulfur inch < = less than kWh = kilowatt-hour Q.-I 8-3 Y~-
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| ; IPilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-2 Air Emissions from Coal-Fired Alternative Parameter Calculation Result Annual coal 620MW 10, 200Btu 1,000kW lb 24hr 365day ton x 085 888980 consumption unit kw x hr MW 12,464Btu day x yr _ 000b_ t1,8 8f8,9 per year So 1,2 1, 888, 980tons 38 x 0.691b ton 100-95 1,238 tons yr ton 2, 0001b 100 Sax per year NOX ' 1,888, 980tons 10ib ton 100- 95 472 tons yr ton 2,0001b 100 NO, per year CO 2 1,888,980tons 0.51b ton 472 tons yr ton 2, 0001b CO per year TSP 1, 888, 980tons; 10 x 8.21b ton 100-99.9 77 tons yr ton 2,0001b 100 TSP per year PM 1 0 4 1, 888,980tons 2.3 x 8.21b ton 100-99.9 18 tons yr ton 2, 000lb 100 PM 1 0 per year 1. Reference 8-7. Table 1.1-1 2. Reference 8-7, Table 1.1-3 3. Reference 8-7, Table 1.1-2 4. Reference 8-7, Table 1.1-4 CO = carbon monoxide NOx = nitrogen oxides PM, 0 = particulates having diameter less than 10 microns SOx = oxides of sulfur TSP = total suspended particulates 8-4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage I..Table 8-3 Solid Waste from Coal-Fired Altemative Parameter Calculation Result Annual SO, 1,888,980 tons coal 0.69 tons 64.1 tons SO 2 26,027 tons of SOx per generated 1 yr -100 tons coal X 32.1 tons S year Annual SOx 26,027 tons SO2 95 24,726 tons of SOx per removed yr 100 year Annual ash 1,888,980 tons coal x 8.2 tons ash 99.9 154,741 tons of ash per generated yr 100 tons coal 100 year Annual lime 26,027 tons SO2 56.1 tons CaO 22,779 tons of CaO per consumption 2 -yr x 64.1 tons SO2 year Calcium sulfate 3 24,726 tons SO2 172 tons Ca S04
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| * 2H 2 0 66,347 tons of yr x 64.1 tons SO 2 CaSO 4-2H 2 O per year Annual scrubber 22,779 tons CaO x 100 -95 + 66,347 tons CaSO 4
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| * H2 ° 67,486 tons of scrubber waste 4 yr x100 2H0waste per year Total volume of 67,486 tons 2,000 lb ft 3 37,285,083 ft 3 of scrubberwaste 5 X yr ton 144.8 lb scrubber waste Total volume of 154,741 tons x lb yr 123,792,800 ft 3 of ash asy15,4 ton 40 yr 2,0- ' 00 lb yr t on ib Total volume of 3 3 110783f 3 o oi 37,285,083 ft + 123,792,800 ft 161,077,883 ft3 of solid solid waste waste Waste pile area 16,7,8 t ce123.3 acres of solid (acres)161,077,883 Rfx acre wat (acres) 30 ft 43,560 2 waste Waste pile area 2,317 feet by feet square (ft x ft square) 2161,077,883 ft /30 ft of solid waste Based on annual coal consumption of 1,888,980 tons per year (Table 8-2).1. Calculations assume 100% combustion of coal.2. Lime consumption is based on total S02 generated.
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| : 3. Calcium sulfate generation is based on total SO 2 removed.4. Total scrubber waste includes scrubbing media carryover.
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| : 5. Density of CaSO 4-2H 2 O is 144.8 lb/ft3.6. Density of coal bottom ash is 100 lb/ft3 [Reference 8-10].S = sulfur S02 = sulfur dioxide SO, = oxides of sulfur CaO = calcium oxide (lime)CaSO 4 2H 2 O = calcium sulfate dihydrate 8-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.1.1 Closed-Cycle Cooling System The overall impacts at an alternate greenfield site of the coal-fired generating system using a closed-cycle cooling system with cooling towers are discussed in the following sections.
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| The magnitude of impacts for the alternate site will depend on the location of the particular site selected.
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| PNPS currently uses once-through cooling systems. For the purposes of comparison with an alternative site, it is assumed that the replacement coal-fired plant sited at an alternate site would use a closed-cycle cooling system.The environmental impacts of building a coal-fired generation facility with a closed-cycle cooling system at an alternate site are summarized in Table 84.8.1.1.1.1 Land Use Based on Table 8.1 of the GEIS it is estimated that it would take approximately 1.7 acres of land per MWe to construct a coal-fired plant. Therefore, for the 620 gross MWe coal-fired plant utilized in this analysis, it would take approximately 1,054 acres of land. This would amount to a considerable loss of natural habitat or agricultural land for the plant site alone, excluding that required for mining and other fuel-cycle impacts.Additional land might also be needed for transmission lines and rail lines, depending on the location of the site relative to the nearest inter-tie connection and rail spur. Depending on the z L transmission line routing and nearest rail line, these alternatives could result in MODERATE to LARGE land use impacts.Land-use changes would occur offsite in an undetermined coal-mining area to supply coal for the plant. In the GEIS, the staff estimated that approximately 22 acres of land per MWe would be affected for mining the coal and disposing of the waste to support a coal-fired plant during its operational life [Reference 8-14]. Therefore, for the 620 gross MWe coal-fired plant utilized in this analysis, it would take approximately 13,640 acres of land. Partially offsetting this offsite land use would be the elimination of the need for uranium mining and processing to supply fuel for PNPS. In the GEIS, the staff estimated that approximately 1 acre per MWe would be affected for mining and processing the uranium during the operating lifd of a nuclear power plant[Reference 8-14]. Therefore, for the 715 gross MWe plant (PNPS) utilized in this analysis, it would take approximately 715 acres of land.The impact of a coal-fired generating unit with a closed-cycle cooling system on land use located at an alternate site is considered as MODERATE to LARGE.8.1.1.1.2 Ecology Constructing a coal-fired plant at an alternate site would alter ecological resources because of the need to convert roughly 1,054 acres of land at the site to industrial use for plant, coal storage, and ash and scrubber sludge disposal.
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| However, some of this land might have been previously disturbed.
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| 8-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Coal-fired generation at an alternative site would introduce construction impacts and new incremental operational impacts. Even assuming siting at a previously disturbed area, the impacts would alter the ecology. Impacts could include wildlife habitat loss, reduced productivity, habitat fragmentation, and a local reduction in biological diversity.
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| Use of cooling makeup water from a nearby surface water body could have adverse impacts on aquatic resources.
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| If needed, construction and maintenance of an electric power transmission line and a rail spur would have ecological impacts. There would be some impact on terrestrial ecology from water drift from the cooling towers. Overall, the ecological impacts of constructing a coal-fired plant with a closed-cycle cooling system at an alternate site are considered to be MODERATE to LARGE.8.1.1.1.3 Water Use and Quality Surface Water Cooling water at an alternate site would likely be withdrawn from a surface water body and would be regulated by permit. Depending on the water source, the impacts of water use for cooling system makeup water and the effects on water quality caused by cooling tower blowdown could have noticeable impacts. Therefore, the impacts of a new coal-fired plant utilizing a closed-cycle cooling system at an alternate site are considered SMALL to MODERATE.Groundwater Impacts of groundwater withdrawal would be SMALL if only used for potable water. If groundwater is used to supply makeup water, then the impacts could be MODERATE to LARGE.Therefore, groundwater impacts from a coal-fired plant on the aquifer would be site-specific and dependent on aquifer recharge and other withdrawals.
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| The overall impacts would be SMALL to LARGE.8.1.1.1.4 Air Quality Air quality impacts of coal-fired generation are considerably different from those of nuclear power. A coal-fired plant emits oxides of sulfur (SOx), nitrogen oxides (NOx), particulate matter, and carbon monoxide, all of which are regulated pollutants.
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| As already stated, Entergy has assumed a plant design that would minimize air emissions through a combination of boiler technology and post-combustion pollutant removal. Entergy estimates the coal-fired alternative emissions to be as follows (from Table 8-2).* Oxides of sulfur = 1,238 tons per year* Oxides of nitrogen = 472 tons per year* Carbon monoxide = 472 -tons per year 8-7 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage'Particulates:-Total suspended particulates
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| =77 tons per year-PM 1 0 (particulates having a diameter of less than 10 microns) =-18 tons per year The acid rain requirements of the Clean Air Act amendments capped the nation's SOx emissions from power plants. Under the Clean Air Act amendments, each company with fossil-fuel-fired units was allocated SOX allowances.
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| To be in compliance with the Act, the companies must hold enough allowances to cover their annual SOX emissions.
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| Entergy would have to purchase allowances to cover its SOX emissions.
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| The NRC did not quantify coal-fired emissions in the GEIS, but implied that air impacts would be substantial.
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| The NRC noted that adverse human health effects from coal combustion have led to important federal legislation in recent years and that public health risks, such as cancer and emphysema, have been associated with coal combustion.
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| The NRC also mentioned global warming and acid rain as potential impacts. Entergy concludes that federal legislation and large-scale concerns, such as global warming and acid rain, are indications of concerns about destabilizing important attributes of air resources.
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| However, SOX emission allowances, NOx emission offsets, low NOx burners with overfire air and selective catalytic reduction, fabric filters X .or electrostatic precipitators, and scrubbers are provided as mitigation measures.
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| As such, (; Entergy concludes that the coal-fired alternative would have MODERATE impacts on air quality;the impacts would be clearly noticeable, but would not destabilize air quality in the area.8.1.1.1.5 Waste Entergy concurs with the GEIS assessment that the coal-fired alternative would generate substantial solid waste. The coal-fired plant would annually consume approximately 1,889,000 tons of coal having an ash content of 8.2%. After combustion, 99.9% of this ash (approximately 155,000 tons per year) would be collected and disposed of at either an onsite or offsite landfill.
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| In addition, approximately 67,500 tons of scrubber waste would be disposed of each year (based on annual calcium hydroxide usage of approximately p3,000 tons). Entergy estimates that ash and scrubber waste disposal over a 40-year plant life would require approximately 123 acres.The amount of land needed for final disposal of ash may be less, dependant upon the availability of local recycling options for the ash.Table 8-3 shows how Entergy calculated ash and scrubber waste volumes. While only half this waste volume and land use would be attributable to the 20-year license renewal period alternative, the total numbers are pertinent as a cumulative impact.Entergy believes that, with proper siting coupled with current waste management and monitoring practices, waste disposal would not destabilize any resources.
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| Some wooded terrestrial habitat would be dedicated to the waste site. However, after closure of the waste site and revegetation, the land would be available for other uses. For these reasons, Entergy believes that waste 8-8 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage disposal for the coal-fired alternative would have MODERATE impacts; the impacts of increased waste disposal would be clearly noticeable, but would not destabilize any important resource and further mitigation would be unwarranted.
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| 8.1.1.1.6 Human Health Coal-fired power generation introduces worker risk from coal and limestone mining, worker and public risk from coal and lime/limestone transportation, worker and public risk from disposal of coal combustion wastes, and public risk from inhalation of stack emissions.
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| Emission impacts can be widespread and health risk is difficult to quantify.
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| The coal alternative also introduces the risk of coal pile fires and attendant inhalation risk.The NRC stated in the GEIS that there could be human health impacts (cancer and emphysema) from inhalation of toxins and particulates from a coal-fired plant, but the GEIS does not identify the significance of these impacts [Reference 8-14]. In addition, the discharges of uranium and thorium from coal-fired plants can potentially produce radiological doses in excess of those arising from nuclear power plant operations
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| [Reference 8-11].Regulatory agencies, including the EPA and State agencies, set air emission standards and requirements based on human health impacts. These agencies also impose site-specific emission limits as needed to protect human health. EPA has recently concluded that certain segments of the U.S. population (e.g., the developing fetus and subsistence fish-eating populations) are believed to be at potential risk of adverse health effects due to mercury exposures from sources such as coal-fired power plants. However, in the absence of more quantitative data, human health impacts from radiological doses and inhaling toxins and particulates generated by a coal-fired plant at an alternate site are considered to be SMALL.8.1.1.1.7 Socioeconomics Based on Table 8.1 of the GEIS, construction of the coal-fired alternative would take approximately 1 year per 200 MWe rating. The peak workforce is estimated to range from 1.2 to 2.5 additional workers per MWe during the construction period, based on estimates given in Table 8.1 of the GEIS. Therefore, for the 620 gross MWe coal-fired plant utilized in this analysis, it would take approximately three years to construct the plant with the workforce ranging from approximately 744 to 1,550.Communities around the new site would have to absorb the impacts of a large, temporary work force (up to 1,550 workers at the peak of construction) and a permanent work force of approximately 0.2 workers per MWe based on Table 8.1 of the GEIS or 124 workers for the 620 gross MWe plant utilized in this analysis.
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| In the GEIS, the staff stated that socioeconomic impacts at a rural site would be larger than at an urban site, because more of the peak construction work force would need to move to the area to work. Alternate sites would need to be analyzed on a case-by-case basis. Therefore, socioeconomic impacts at an isolated rural site could be LARGE.8-9 0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Transportation related impacts associated with commuting construction workers at an alternate site would be site dependent, but could be MODERATE to LARGE.Transportation impacts related to commuting of plant operating personnel would also be site dependent, but can be characterized as SMALL to MODERATE.At most alternate sites, coal and lime would be delivered by rail, although barge delivery is feasible for a location on navigable waters. Transportation impacts would depend upon the site location.
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| Socioeconomic impacts associated with rail transportation would be MODERATE to LARGE. Barge delivery of coal and lime/limestone would -have SMALL socioeconomic impacts.8.1.1.1.8 Aesthetics Alternative site locations could reduce the aesthetic impact of coal-fired generation if siting were in an area that was already industrialized.
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| In such a case, however, the introduction of tall stacks and cooling towers would probably still have a MODERATE incremental impact. Locating at other, largely undeveloped sites could show a LARGE impact.8.1.1.1.9 Historic and Archaeological Resources Before construction at an alternate site, studies would be needed to identify, evaluate, and address mitigation of the potential impacts of new plant construction on cultural resources.
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| The studies would be needed for areas of potential disturbance at the proposed plant site and along\; associated corridors where new construction would occur (e.g., roads, transmission corridors, rail lines, or other rights-of-way).
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| Historic and archeological resource impacts can generally be effectively managed and as such are considered SMALL.8-10 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-4 Summary of Environmental Impacts from Coal-Fired Generation Using Closed-Cycle Cooling at an Alternate Greenfield Site Impact Category Impact Comments Land Use MODERATE to Approximately 1054 acres, including transmission LARGE lines and rail line for coal delivery.Ecology MODERATE to Impact will depend on ecology of site.LARGE Water Use and Quality:-Surface Water SMALL to Impact will depend on volume and other MODERATE characteristics of receiving water.-Groundwater SMALL to LARGE Impact will depend on site characteristics and availability of groundwater.
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| Air Quality MODERATE sOx-1,238 MT/yr-allowances required NOx-472 MT/yr-allowances required Particulate
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| -77 MT/yr (filterable)
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| -18 MT/yr (unfilterable)
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| Carbon monoxide-472 MT/yr Trace amounts of mercury, arsenic, chromium, beryllium and selenium Waste MODERATE Total waste volume would be estimated around 222,200 tons per year of ash and scrubber sludge.Human Health SMALL Impacts considered minor.Socioeconomics SMALL to LARGE Communities would have to absorb impacts of a large, temporary workforce (up to 1,550 workers at the peak of construction) and a permanent work force of approximately 124 workers.Impacts at a rural site would be larger.Transportation-related impacts associated with commuting construction workers would be site dependent.
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| Q., 8-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-4 Summary of Environmental Impacts from Coal-Fired Generation Using Closed-Cycle Cooling at an Alternate Greenfield Site (Continued)
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| Impact Category Impact Comments Aesthetics MODERATE to Could reduce aesthetic impact if siting is in an LARGE industrial area. Impact would be large if siting is largely in an undeveloped area.Historic and SMALL Would necessitate cultural resource studies.Archaeological Resources i1 '. o I 8-12 Pilgrim Nuclear Power Station Applicants Environmental Report Operating License Renewal Stage 0'8.1.1.2 Once-Through Cooling System The environmental impacts of constructing a coal-fired generation system at an alternate greenfield site using once-through cooling are similar to the impacts for a coal-fired plant using a closed-cycle cooling system. However, there are some environmental differences between the closed-cycle and once-through cooling systems. Table 8-5 summarizes the incremental differences.
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| Table 8-5 Summary of Environmental Impacts from Coal-Fired Generation Using Once-Through Cooling at an Alternate Greenfield Site Impact Category Impact Comments Land Use MODERATE to Compared with a closed-cycle cooling system, LARGE less land would be required because cooling towers and associated infrastructure not needed.Ecology MODERATE to Slightly reduced environmental impacts LARGE because there are no cooling towers;however, increased water withdrawal may impact aquatic resources.
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| Water Use and Quality:-Surface Water SMALL to Impact would depend on surface water body MODERATE characteristics, volume of water withdrawn, and characteristics of the discharge.
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| -Groundwater SMALL to Impact would depend on site characteristics LARGE and availability of groundwater.
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| It is unlikely that groundwater would be used for once-through cooling, but could be used for sanitary water.Air Quality MODERATE No change.Waste MODERATE No change.Human Health SMALL No change.Socioeconomics SMALL to No change.LARGE Aesthetics MODERATE to Reduced aesthetic impact because cooling LARGE towers would not be used.Historic and Archaeological SMALL Less land impacted.Resources 0J 8-13 Q(1-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.2 Gas-Fired Generation Entergy has chosen to evaluate gas-fired generation, using combined-cycle turbines, because it has determined that the technology is mature, economical, and feasible.
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| Table 8-6 presents the basic gas-fired alternative characteristics and Table 8-7 presents emission estimates.
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| The NRC evaluated environmental impacts from gas-fired generation alternatives in the GEIS, focusing on combined-cycle plants. The NRC has evaluated the environmental impacts of constructing and operating four 440-MWe combined-cycle gas-fired units as an alternative to a nuclear power plant license renewal [Reference 8-14]. This analysis would bound the gas-fired alternative analysis for PNPS because Entergy has defined a reasonable gas alternative for PNPS as a 608-MWe combined-cycle plant. Entergy has adopted the rest of the NRC analysis with necessary Entergy-specific modifications noted. Although air emissions from the gas-fired unit would be substantially smaller than from the coal-fired unit, human health effects associated with such emissions would be of concern.8-14 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-6 Gas-Fired Alternative Emission Control Characteristics Characteristic Basis Unit size = 585 MWe ISO rating net 1 Manufacturer's standard size gas-fired combined Two 189-MWe combustion turbines cycle plant that is <PNPS gross capacity and a 207-MWe heat recovery boiler (715 MWe)Unit size = 608 MWe ISO rating grossa Calculated based on 4% onsite power Number of units = 1 Fuel type = natural gas Assumed Fuel heating value = 1,042 Btu/ft 3 2000 value for gas used in Massachusetts
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| [Reference 8-6, Table 25]Fuel sulfur content = 0.0034 lb/MMBtu Used when sulfur content is not available[Reference 8-8, Table 3.1-2a]NOx control = selective catalytic reduction Best available for minimizing NOX emissions (SCR) with steam/water injection
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| [Reference 8-8, Table 3.1 Database]Fuel NOx content = 0.0109 lb/MMBtu Typical for large SCR-controlled gas-fired units with water injection[Reference 8-8, Table 3.1 Database]Fuel CO content = 0.0023 lb/MMBtu Typical for large SCR-controlled gas-fired units[Reference 8-8, Table 3.1]Heat rate = 6,204 Btu/kWh Manufacturer's listed heat rate for this unit.Capacity factor = 0.85 Typical for large gas-fired base load units (Entergy experience)
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| Q. )1. The difference between 'net' and 'gross' Is electricity consumed by auxiliary equipment and environ-mental control devices [Reference 8-5, page 107].Btu = British thermal unit ft 3 = cubic foot ISO rating = International Standards Organization rating at standard atmospheric conditions of 59 0 F, 60% relative humidity, and 14.696 pounds of atmospheric pressure per square inch kWh = kilowatt-hour MM = million MW = megawatt NOx = nitrogen oxides< = less than SCR = selective catalytic reduction 8-15 (%-
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-7 Air Emissions from Gas-Fired Alternative Parameter Calculation Result Annual gas 608 MW 6,204 Btu 1,000 kW ft 3 24 hr 365 day 26,954,462,833 consumption unit x kW x hr X ' MW X 0 185 X1042 Btu x dlay yr ft 3 per year Annual Btu 26,954,462,833ft3 1.042 Btu MMBtu 28,086,550 input yr ft 3 16Btu MMBtu per year SOxI 0.0034 lb ton X 28,086,550 MMBtu 47.7 tons SOx KMMBtu 2,000 lb yr per year NoX2 0.0109 lb ton 28,086,550 MMBtu 153.1 tons NOx MMBtu 2,000 lb yr per year Co 2 0.0023 lb ton 28,086,550 MMBtu 32.2 tons CO per MMBtu 2,000 lb yr year TSP 1 0.0019 lb ton 28,086,550 MMBtu 26.7 tons MMBtu 2,000 lb yr filterable TSP per year PM 1 0 1 26.7 tons TSP 26.7 tons yr filterable PM 1 0 per year 1. Reference 8-8 2. Reference 8-8 CO carbon monoxide NOx oxides of nitrogen PM, 0 = particulates having diameter less than 10 microns SO, = oxides of sulfur TSP = total suspended particulates 8-16 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Q 8.1.2.1 Closed-Cycle Cooling System The overall impacts of the natural-gas-generating system with a closed-cycle cooling system located at the PNPS site or an alternate site are summarized in Table 8-8 and discussed in the following sections.
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| The magnitude of impacts at an alternate site will depend on the location of the particular site selected.8.1.2.1.1 Land Use Gas-fired generation at the PNPS site would require converting the existing industrial site to a gas plant. Almost all the converted land would be used for the power block and associated facilities.
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| Additional land would be disturbed during pipeline construction.
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| Some additional land would also be required for backup oil storage tanks. The nearest gas pipeline tie-in is located in Plymouth, Massachusetts (Algonquin Gas Transmission Line), 5.5 miles from the PNPS site.Therefore, gas-fired generation land use impacts at the existing PNPS site are SMALL to MODERATE; the impacts would noticeably alter the habitat, but would not destabilize important attributes of the resource.In addition to the land required for the gas-fired plant, construction at a greenfield site could impact approximately 20 to 50 acres for offices, roads, parking areas, and a switchyard.
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| The power block could require 60 acres. Some additional land would also be required for backup oil storage. In addition, it is assumed that additional acreage may be necessary for transmission lines (assuming the plant is sited 10 miles from the nearest inter-tie connection) although this would depend on the actual plant location.
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| Plants of this type are usually built very close to (_existing natural gas pipelines.
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| Including the land required for pipeline construction, a greenfield site could require approximately 500 acres. Depending on the transmission-line routing, the greenfield site alternative could result in SMALL to MODERATE land-use impacts.8.1.2.1.2 Ecology Siting gas-fired generation at the existing PNPS site would have MODERATE ecological impacts because the facility would be constructed on previously disturbed areas and would disturb relatively little acreage at the site. Habitat would be disrupted by pipeline construction.
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| Ecological impacts could be reduced by using the existing intake and discharge system. Past operational monitoring of the effects of the cooling systems at PNPS has not shown significant negative impacts to the Cape Cod Bay, and this would be expected to remain unchanged.
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| The GEIS noted that land-dependent ecological impacts from construction would be SMALL unless site-specific factors indicate a particular sensitivity and that operational impact would be smaller than for other fossil fuel technologies of equal capacity.
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| Therefore, in this case, the appropriate characterization of gas-fired generation ecological impacts is SMALL.Construction at a greenfield site could alter the ecology of the site and could impact threatened and endangered species. These ecological impacts could be SMALL to MODERATE.8-17 (
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.2.1.3 Water Use and Quality Surface Water The plant would use the existing PNPS intake and discharge structures as part of a closed-cycle cooling system; therefore, water quality impacts would continue to be SMALL.Water quality impacts from sedimentation during construction are another land related impact that the GEIS categorized as SMALL. The GEIS also noted that operational water quality impacts would be similar to, or less than, those from other centralized generating technologies.
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| The NRC has concluded that water quality impacts from coal-fired generation would be SMALL, and gas-fired alternative water usage would be less than that for coal-fired generation.
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| Surface water impacts would remain SMALL; the impacts would not be detectable or be so minor that they would not noticeably alter important attributes of the resource.For alternative greenfield sites, the impact on surface water would depend on the volume and other characteristics of the receiving body of water. The impacts would be SMALL to MODERATE.Groundwater As discussed in Section 3.2.2.2 of this ER, PNPS does not have its own groundwater wells for potable water purposes, but rather purchases potable water from the Town of Plymouth.Therefore, groundwater impacts would be SMALL; the impacts would be so minor that they would not noticeably alter important resources.
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| For alternative greenfield sites, the impact to the groundwater would depend on the site characteristics, including the amount of groundwater available.
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| The impacts would range between SMALL and LARGE.8.1.2.1.4 Air Quality Natural gas is a relatively clean-burning fossil fuel; the gas-fired alternative would release similar types of emissions, but in lesser quantities, than the coal-fired alternative.
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| Control technology for gas-fired turbines focuses on NOx emissions.
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| Entergy estimates the gas-fired alternative emissions to be as follows (from Table 8-7).* Sulfur oxides = 47.7 tons per year* Oxides of nitrogen = 153.1 tons per year* Carbon monoxide = 32.2 tons per year* Filterable Particulates
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| = 26.7 tons per year (all particulates are PM 1 0)8-18 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Regional air quality and Clean Air Act requirements are also applicable to the gas-fired generation alternative.
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| NOx effects on ozone levels, SOx allowances, and NOx emission offsets could all be issues of concern for gas-fired combustion.
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| While gas-fired turbine emissions are less than coal-fired boiler emissions, and regulatory requirements are less stringent, the emissions are still substantial.
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| Entergy concludes that emissions from the gas-fired alternative located at PNPS would noticeably alter local air quality, but would not destabilize regional resources.
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| Air quality impacts would therefore be MODERATE, but substantially smaller than those of coal-fired generation.
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| Siting the gas-fired plant elsewhere would not significantly change air quality impacts because any greenfield site located in Massachusetts would be in a serious nonattainment area for ozone.In addition, the location could result in installing more or less stringent pollution control equipment to meet the regulations.
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| Therefore, the impacts would be MODERATE.8.1.2.1.5 Waste There are only small amounts of solid waste products (i.e., ash) from burning natural gas fuel.The GEIS concluded that waste generation from gas-fired technology would be minimal. Gas firing results in very few combustion by-products because of the clean nature of the fuel. Waste generation would be limited to typical office wastes. This impact would be SMALL; waste generation impacts would be so minor that they would not noticeably alter important resource attributes.
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| Siting the facility at an alternate greenfield site would not alter the waste generation; therefore, the impacts would continue to be SMALL.8.1.2.1.6 Human Health The GEIS analysis mentions potential gas-fired alternative health risks (cancer and emphysema).
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| The risk may be attributable to NOx emissions that contribute to ozone formation, which in turn contributes to health risks. As discussed in Section 8.1.1 for the coal-fired alternative, legislative and regulatory control of the nation's emissions and air quality are protective of human health, and the human health impacts from gas-fired generation would be SMALL. That is, human health effects would not be detectable or would be so minor that they would neither destabilize nor noticeably alter important attributes of the resource.Siting of the facility at an alternate greenfield site would not alter the possible human health effects. Therefore, the impacts would be SMALL.8.1.2.1.7 Socioeconomics It is assumed that gas-fired construction would take place while PNPS continues operation, with completion of the replacement plant at the time that the nuclear plant would halt operations.
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| Construction of the gas-fired alternative would take much less time than constructing other plants. During the time of construction, the surrounding communities would experience demands 8-19B Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage on housing and public services that could have MODERATE impacts. After construction, the communities would be impacted by the loss of jobs, construction workers would leave, PNPS nuclear plant workforce would decline through a decommissioning period to a minimal maintenance size, and the gas-fired plant would introduce a replacement tax base of about 100 new jobs.The GEIS concluded that socioeconomic impacts from constructing a gas-fired plant would not be very noticeable and that the small operational workforce would have the lowest socioeconomic impacts (local purchases and taxes) of nonrenewable technologies.
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| Compared to the coal-fired alternative, the smaller size of the construction workforce, the shorter construction time frame, and the smaller size of the operations workforce would reduce some of the socioeconomic impacts. For these reasons, the socioeconomic impacts of gas-fired-generation socioeconomic impacts would be SMALL to MODERATE.
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| That is, depending on other growth in the area, socioeconomic effects could be noticed, but they would not destabilize important attributes of the resource.Construction at another site would relocate some socioeconomic impacts, but would not eliminate them. The community around the PNPS site would still experience the impact of the loss of PNPS operational jobs and the tax base. The communities around the new site would have to absorb the impacts of a temporary workforce and a small permanent workforce.
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| Therefore, the impacts would be MODERATE to LARGE, based on net job and tax-base losses In the PNPS. However, the reduction in staff would be mitigated by PNPS' proximity to the Boston area. This impact is about the same in the PNPS area as in the no-action alternative.
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| 8.1.2.1.8 Aesthetics The combustion turbines and heat-recovery boilers would be relatively low structures and would be screened from most offsite vantage points by intervening woodlands.
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| The steam turbine building would be taller and together with the exhaust stacks, could be visible offsite. However, the visual impacts would be comparable to those from the existing PNPS facilities.
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| The GEIS analysis noted that land-related impacts, such as aesthetic impacts, would be small unless site-specific factors indicate a particular sensitivity.
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| As in the case of the coal-fired alternative, aesthetic impacts from the gas-fired alternative would be noticeable.
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| However, because the gas-fired structures are shorter than the coal-fired structures and more amenable to screening by vegetation, it was determined that the aesthetic resources would notbe destabilized by the gas-fired alternative.
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| For these reasons, aesthetic impacts ma, gs-fired plant would be SMALL to MODERATE.
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| The impacts would be clearly noticeable, but would not destabilize this important resource.Alternative locations could reduce the aesthetic impact of gas-fired generation if siting was in an area that was already industrialized.
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| In such a case, however, the introduction of the steam generator building, stacks, and cooling tower plumes would probably still have a SMALL to MODERATE incremental impact.8-20 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.2.1.9 Historic and Archaeological Resources The GEIS analysis noted, as for the coal-fired alternative, that cultural resource impacts of the gas-fired alternative would be SMALL unless important site-specific resources were affected.Gas-fired alternative construction at the PNPS site would affect a smaller area within the footprint of the coal-fired alternative.
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| Therefore, cultural resource impacts would be SMALL. That is, cultural resource impacts would not be detectable or would be so minor that they would neither destabilize nor noticeably alter important attributes of the resource.Construction at another site could necessitate instituting cultural resource preservation measures, but impacts can generally be managed and maintained as SMALL. Cultural resource surveys would be required for the pipeline construction and other areas of ground disturbance associated with this alternative.
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| Table 8-8 Summary of Environmental Impacts from Gas-Fired Generation Using Closed-Cycle Cooling at PNPS or at Alternate Greenfield Site PNPS Site Alternative Greenfield Site Impact Category Impact Comments Impact Comments Land Use SMALL to Approximately 60 SMALL to Up to 500 acres MODERATE acres required for MODERATE required for site, power block, 150 pipelines, transmission acres disturbed for line connection; pipeline construction, additional land for additional land for backup oil storage backup oil storage tanks.tanks.Ecology SMALL to Constructed on land SMALL to Impact depends on MODERATE within PNPS site. MODERATE location and ecology Possible habitat loss of site; potential due to pipeline habitat loss and construction.
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| fragmentation; reduced productivity and biological diversity.
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| Water Use and SMALL Uses existing intake SMALL to Impact depends on Quality: and discharge MODERATE volume and Surface Water structures and characteristics of cooling system. receiving water body.8-21 Q..",
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-8 Summary of Environmental Impacts from Gas-Fired Generation Using Closed-Cycle Cooling at PNPS or at Alternate Greenfield Site (Continued)
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| PNPS Site Alternative Greenfield Site Impact Category Impact Comments Impact Comments Water Use and SMALL PNPS does not have SMALL to Groundwater impacts Quality: its own groundwater LARGE would depend on uses Groundwater system and available supply.Air Quality MODERATE Primarily nitrogen MODERATE Same impacts as oxides. Impacts PNPS site.could be noticeable, but not destabilizing.
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| Waste SMALL Small amount of ash SMALL Same impacts as-produced.
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| PNPS site.Human Health SMALL Impacts considered SMALL Same impacts as minor. PNPS site.Socioeconomics SMALL to Additional workers MODERATE Construction impacts MODERATE during construction to LARGE would be relocated.
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| period, followed by Community near reduction from PNPS would still current PNPS experience workforce workforce.
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| reduction.
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| Aesthetics SMALL to Visual impact of SMALL to Alternate location MODERATE stacks and MODERATE could reduce aesthetic equipment would be impact if siting is in an noticeable, but not as industrial area.significant as coal option.Historic and SMALL Only previously SMALL Alternate location Archaeological disturbed and would necessitate Resources adjacent areas would cultural resource be affected.
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| studies.8-22 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage I 8.1.2.2 Once-Through Cooling System The environmental impacts of constructing a natural-gas-fired generation system at the PNPS site and an alternate site using a once-through cooling system are similar to the impacts for a natural-gas-fired plant using closed-cycle cooling with cooling towers. However, there are some environmental differences between the closed-cycle and once-through cooling systems. Table 8-9 summarizes the incremental differences.
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| Table 8-9 Summary of Environmental Impacts from Gas-Fired Generation Using Once-Through Cooling at PNPS or at an Alternate Greenfield Site PNPS Site Alternative Greenfield Site Impact Category Impact Comments Impact Comments Land Use SMALL to 25 to 30 acres less SMALL to 25 to 30 acres less land MODERATE land required MODERATE required because because cooling cooling towers and towers and associated associated infrastructure are not infrastructure are not, needed.needed.Ecology SMALL Less terrestrial SMALL to Impact would depend habitat lost and MODERATE on ecology at the site.cooling tower effects No impact to terrestrial eliminated.
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| ecology from cooling Increased water tower drift. Increased withdrawal, but water withdrawal and aquatic impact possible greater impact would be similar to to aquatic ecology.current PNPS operations.
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| Water Use and SMALL to No discharge of SMALL to No discharge of cooling Quality: MODERATE cooling tower MODERATE tower blowdown Surface Water blowdown containing dissolved containing dissolved solids. Increased water solids. Increased withdrawal and more water withdrawal thermal load on and more thermal receiving body of water.load on receiving body of water.C-)8-23 C-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-9 Summary of Environmental Impacts from Gas-Fired Generation Using Once-Through Cooling at PNPS or at an Alternate Greenfield Site PNPS Site Alternative Greenfield Site Impact Category Impact Comments Impact Comments Water Use and SMALL No change. SMALL to Groundwater impacts Quality: LARGE would depend on uses Groundwater and available supply. It is unlikely that groundwater would be used for once-through cooling, but could be used for sanitary water.Air Quality MODERATE No change. MODERATE No change.Waste SMALL No change. SMALL No change.Human Health SMALL No change. SMALL No change.Socioeconomics SMALL to No change. MODERATE to No change.MODERATE LARGE Aesthetics SMALL to Reduced aesthetic SMALL to Reduced aesthetic MODERATE impact because MODERATE impact because cooling cooling towers towers would not be would not be used. used.Historic and SMALL Less land affected.
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| SMALL Less land affected.Archaeological Resources 8-24 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.3 Nuclear Power Generation Since 1997, the NRC has certified three new standard designs for nuclear power plants under 10 CFR 52, Subpart B. These designs are the U.S. Advanced Boiling Water Reactor (10 CFR 52, Appendix A), the System 80+ Design (10 CFR 52, Appendix B), and the AP600 Design (10 CFR 52, Appendix C). All of these plants are light-water reactors.
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| Although no applications for a construction permit or a combined license based on these certified designs have been submitted to the NRC, the submission of the design certification applications indicates continuing interest in the possibility of licensing new nuclear power plants. In addition, recent volatility of natural gas and electricity has made new nuclear power plant construction more attractive from a cost standpoint.
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| Consequently, construction of a new nuclear power plant at an alternate site using closed-cycle cooling is considered in this section. It was assumed that the new nuclear plant would have a 40-year lifetime [Reference 8-17, Section 8.2.3].The NRC summarized environmental data associated with the uranium fuel cycle in Table S-3 of 10 CFR 51.51. The impacts shown in Table S-3 are representative of the impacts that would be associated with a replacement nuclear power plant built to one of the certified designs, sited at PNPS or at an alternate site. The impacts shown in Table S-3 are for a 1 000-MWe reactor and would need to be adjusted to reflect replacement of PNPS, which has a capacity of 715 gross MWe. The environmental impacts associated with transporting fuel and waste to and from a light-water cooled nuclear power reactor are summarized in Table S-4 of 10 CFR 51.52. The summary of the NRC's findings on NEPA issues for license renewal of nuclear power plants in 10 CFR 51 Subpart A, Appendix B, Table B-1 is also relevant, although not directly applicable, for consideration of environmental impacts associated with the operation of a replacement nuclear power plant [Reference 8-17, Section 8.2.3].8.1.3.1 Closed-Cycle Cooling System The environmental impacts of constructing a nuclear power plant at an alternate site using closed-cycle cooling are summarized in Table 8-10.8.1.3.1.1 Land Use Land use requirements at an alternate site would require land for the nuclear power plant plus the possible need for land for a new transmission line. In addition, it may be necessary to construct a rail spur to an alternate site to bring in equipment during construction.
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| Depending on transmission line routing, siting a new nuclear plant at an alternate site would result in MODERATE to LARGE land use impacts, and probably would be LARGE for a greenfield site[Reference 8-17, Section 8.2.3.1].8.1.3.1.2 Ecology At an alternate site, there would be construction impacts and new incremental operational impacts. Even assuming siting at a previously disturbed area, the impacts would alter the ecology. Impacts could include wildlife habitat loss, reduced productivity, habitat fragmentation, 8-25 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage and a local reduction in biological diversity.
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| Use of cooling water from a nearby surface water body could have adverse aquatic resource impacts. Construction and maintenance of the transmission line would have ecological impacts. Overall, the ecological impacts at an alternate site would be MODERATE to LARGE [Reference 8-17, Section 8.2.3.1].8.1.3.1.3 Water Use and Quality Surface Water For a replacement reactor located at an alternate site, new intake structures would need to be constructed to provide water needs for the facility.
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| Impacts would depend on the volume of water withdrawn for makeup, relative to the amount available from the intake source and the characteristics of the surface water. Plant discharges would be regulated by the State of Massachusetts or other state jurisdiction.
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| Some erosion and sedimentation may occur during construction.
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| The impacts would be SMALL to MODERATE.Groundwater A nuclear power plant sited at an alternate site may use groundwater.
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| The impacts of such a withdrawal rate on an aquifer would be site specific and dependent on aquifer recharge and other withdrawal rates from the aquifer. Therefore, the overall impacts would be SMALL to LARGE.8.1.3.1.4 Air Quality Construction of a new nuclear plant at an alternate site would result in fugitive emissions during the construction process. Exhaust emissions would also come from vehicles and motorized equipment used during the construction process. An operating nuclear plant would have minor air emissions associated with diesel generators, house-heating boilers, and similar minor emission points. These emissions would be regulated.
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| Emissions for a plant sited in Massachusetts would be regulated by the MDEP. Overall, emissions and associated impacts are considered SMALL [Reference 8-17, Section 8.2.3.1].8.1.3.1.5 Waste The waste impacts associated with operation of a nuclear power plant are listed in Table B-1 of 10 CFR 51 Subpart A, Appendix B. In addition to the impacts shown in Table B-1, construction-related debris would be generated during construction activities and removed to an appropriate disposal site. Overall, waste impacts are considered SMALL [Reference 8-17, Section 8.2.3.1].8.1.3.1.6 Human Health Human health impacts for an operating nuclear power plant are identified in 10 CFR 51 Subpart A, Appendix B, Table B-1. Overall, human health impacts are considered SMALL [Reference 17, Section 8.2.3.1].8-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.1.3.1.7 Socioeconomics For a 1,000 MWe reactor, it was assumed that the construction period would be 5 years and the peak workforce would be 2,500. Since PNPS's current reactor is rated at 715 gross MWe, construction period and peak workforce may be less, but impacts are expected to be consistent with that of the 1,000 MWe reactor.Construction of a replacement nuclear power plant at an alternate site would relocate some socioeconomic impacts, but would not eliminate them. The communities around the PNPS site would still experience the impact of PNPS operational job loss (although potentially tempered by projected economic growth), and the communities around the new site would have to absorb the impacts of a large, temporary work force (up to 2,500 workers at the peak of construction) and a permanent work force of approximately 704 workers. In the GEIS, the NRC noted that socioeconomic impacts at a rural site would be larger than at an urban site because more of the peak construction work force would need to move to the area to work. Alternate sites would need to be analyzed on a case-by-case basis. Socioeconomic impacts at rural sites could be LARGE [Reference 8-17, Section 8.2.3.1].Transportation-related impacts associated with commuting workers at an alternate site are site dependent, but could be MODERATE to LARGE. Transportation impacts related to commuting of plant operating personnel would also be site dependent, but can be characterized as SMALL[Reference 8-17, Section 8.2.3.1].8.1.3.1.8 Aesthetics C At an alternate site, depending on placement, there would be an aesthetic impact from the buildings.
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| There would also be a significant aesthetic impact associated with construction of a new transmission line to connect to other lines to enable delivery of electricity.
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| Noise and light from the plant would be detectable offsite. The impact of noise and light would be mitigated if the plant were located in an industrial area adjacent to other power plants, in which case the impact could be SMALL. The impact could be MODERATE if a transmission line needs to be built to the alternate site. The impact could be LARGE if a greenfield site is selected [Reference 8-17, Section 8.2.3.1].8.1.3.1.9 Historic and Archeological Resources Before construction at an alternate site, studies would be needed to identify, evaluate, and address mitigation of the potential impacts of new plant construction on cultural resources.
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| The studies would be needed for areas of potential disturbance at the proposed plant site and along associated corridors where new construction would occur (e.g., roads, transmission corridors, rail lines, or other rights-of-way).
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| Historic and archeological resource impacts can generally be effectively managed and as such are considered SMALL.8-27 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-10 Summary of Environmental Impacts from Nuclear Power Generation Closed-Cycle Cooling at Alternate Greenfield Site Alternative Greenfield Site Impact Category Impact Comments Land Use MODERATE Requires 376 to 715 acres for the plant and 715 acres for uranium to LARGE mining.Ecology MODERATE Impact depends on location and ecology of the site, surface water to LARGE body used for intake and discharge, and transmission line routes;potential habitat loss and fragmentation; reduced productivity and biological diversity.
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| Water Use and SMALL to Impact will depend on the volume of water withdrawn and Quality: MODERATE discharged and the characteristics of the surface water body.Surface Water Water Use and SMALL to Groundwater impacts would depend on uses and available supply.Quality: LARGE Groundwater Air Quality SMALL Fugitive emissions and emissions from vehicles and equipment during construction.
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| Small amount of emissions from diesel generators and possibly other sources during operation.
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| Emissions are similar to current releases at PNPS site.Waste SMALL Waste impacts for an operating nuclear power plant are set out in 10 CFR 51, Subpart A, Appendix B, Table B-1. Debris would be generated and removed during construction.
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| Human Health SMALL Human health impacts for an operating nuclear power plant are set out in 10 CFR 51, Subpart A, Appendix B, Table B-1.Socioeconomics SMALL to Construction impacts depend on location.
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| Impacts at a rural LARGE location could be LARGE. Surrounding community would experience loss of tax base and employment with MODERATE impacts. Transportation impacts associated with construction workers could be MODERATE to LARGE. Transportation impacts of commuting workers during operations would be SMALL.Aesthetics SMALL to Impacts would depend on the characteristics of the alternate site.LARGE Impacts would be SMALL if the plant is located adjacent to an industrial area. New transmission lines would add to the impacts and could be MODERATE.
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| If a greenfield site Is selected, the impacts could be LARGE.8-28 X Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-10 Summary of Environmental Impacts from Nuclear Power Generation Closed-Cycle Cooling at Alternate Greenfield Site (Continued)
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| Alternative Greenfield Site Impact Category Impact Comments Historic and SMALL Potential impacts can be effectively managed.Archaeological Resources
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| _8.1.3.2 Once-Through Cooling System The environmental impacts of constructing a nuclear power plant that uses once-through cooling at an alternate site are similar to the impacts for a nuclear power plant using closed-cycle cooling with cooling towers. However, there are some differences in the environmental impacts between the closed-cycle and once-through cooling systems. In those impact categories related to land-area requirements, such as land use, terrestrial ecology, and cultural resources, the impacts are likely to be smaller if the site uses a once-through cooling system rather than a closed-cycle cooling system. However, the impacts of a plant with a once-through cooling system are likely to be greater than a plant with a closed-cycle cooling system in the areas of water use and aquatic ecology because of the need for greater quantities of cooling water. Table 8-11 summarizes the incremental differences.
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| Table 8-11 Summary of Environmental Impacts from Nuclear Power Generation Using Once-Through Cooling at Alternate Greenfield Site (w1)Alternative Greenfield Site Impact Category Impact Comments Land Use MODERATE Requires 376 to 715 acres for the plant and 715 acres for uranium to LARGE mining.Ecology MODERATE Impact would depend on ecology of the site. No impact to to LARGE terrestrial ecology from cooling tower drift. Increased water withdrawal with possible greater impact to aquatic ecology.Water Use and SMALL to No discharge of cooling tower blowdown.
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| Increased water Quality: MODERATE withdrawal and more thermal load on receiving body of water.Surface Water 8-29 0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 8-11 Summary of Environmental Impacts from Nuclear Power Generation Using Once-Through Cooling at Alternate Greenfield Site (Continued)
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| Alternative Greenfield Site Impact Category Impact Comments Water Use and SMALL to No change.Quality: LARGE Groundwater Air Quality SMALL No change.Waste SMALL No change.Human Health SMALL No change.Socioeconomics MODERATE No change.to LARGE Aesthetics SMALL to Reduced aesthetic impact because cooling towers would not be LARGE used, but impacts could still be large if lengthy transmission line is required.Historic and SMALL Less land impacted Archaeological Resources 8.1.4 Purchased Electrical Power If available, purchased power from other sources could potentially obviate the need to renew PNPS. "Purchased power" is power purchased and transmitted from electric generation plants that the applicant does not own and that are located elsewhere within the region, nation, Canada, or Mexico.In theory, purchased power is a feasible alternative to PNPS license renewal. There is no assurance, however, that sufficient capacity or energy would be available in the 2012 through 2032 time frame to replace the 715 gross MWe base-load generation.
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| For example, EIA projects that total gross U.S. imports of electricity from Canada and Mexico will gradually increase from 38.4 billion kWh in year 2001 to 47.2 billion kWh in year 2010 and then gradually decrease to 28.94 billion kWh in year 2020 [Reference 8-2, page 149]. On balance, It appears unlikely that electricity purchased from Canada or Mexico would be able to replace the PNPS generating capacity.More importantly, regardless of the technology used to generate purchased power, the generating technology would be one of those described in this ER and in the GEIS (probably coal, natural gas, nuclear, or hydroelectric).
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| The GEIS description of other technology impacts is 8-30 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage representative of purchased power impacts related to PNPS license renewal alternatives
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| [Reference 8-16).8.2 Alternatives Not Within the Range of Reasonable Alternatives Other commonly known generation technologies considered are listed in the following paragraphs.
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| However, these sources have been eliminated as reasonable alternatives to the proposed action because the generation of 715 gross MWe of electricity as a base-load supply using these technologies is not technologically feasible, except for oil, which is not economically feasible.8.2.1 Wind In the entire six-state New England region, only two wind projects are in operation:
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| the 6 MW Searsburg project in Vermont and a 320 kW project in Massachusetts owned by Princeton Municipal Light. There is also an additional project under active development in southern Vermont (Equinox) (Reference 8-41. Wind turbines typically operate at a 25 to 35% capacity factor compared to 80 to 95% for a base load plant. This low capacity factor results from the high degree of intermittence of wind energy in many locations.
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| Current energy storage technologies are too expensive to permit wind power plants to serve as large base load plants.According to the Wind Energy Resource Atlas of the United States (Reference 8-18), areas suitable for wind energy applications must be wind power class 3 or higher. Approximately 50%of the land area in Massachusetts has a wind power classification of 3 or higher and, therefore C;may be suitable for wind energy applications.
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| However, land-use conflicts such as urban development, farmland, and environmentally sensitive areas reduce the amount of land suitable for wind energy applications to about 16% of the land area in the state (Reference 8-9).The GEIS estimates a land use of 150,000 acres per 1,000 MWe for wind power (Reference 8-14, Section 8.3.1). Therefore, to replace the 715 gross MWe of electricity generated by PNPS, approximately 107,250 acres would be required.
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| The areas having ideal conditions are located on mountaintops and adjacent to the coast. There is insufficient area on the coast for replacing the PNPS generating capacity.
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| Therefore the wind alternative would require a large Greenfield site located on mountaintops, which would result in a LARGE adverse environmental impact.Also, new easements, road building, and some clearing for towers and blades would be required.This eliminates the possibility of co-locating a wind-energy facility with a retired nuclear power plant. A siting plan would be required.
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| Construction of several hundred wind turbines would also require extensive construction of transmission lines to bring the power and the energy to market.This would have a LARGE impact upon much of the natural environment in the affected areas.Wind power could be included in a combination of alternatives to replace PNPS. The environmental impacts of a large-scale wind farm are described in the GEIS [Reference 8-14].The construction of roads, transmission lines, and turbine tower supports would result in short-term impacts, such as increases in erosion and sedimentation, and decreases in air quality from 8-31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage fugitive dust and equipment emissions.
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| Construction in undeveloped areas would have the potential to disturb and impact cultural resources or habitat for sensitive species. During operation, some land near wind turbines could be available for compatible uses such as agriculture.
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| The continuing aesthetic impact would be considerable, and there is a potential for bird collisions with turbine blades. Wind farms generate very little waste and pose no human health risk other than from occupational injuries.
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| Although most impacts associated with a wind farm are SMALL or can be mitigated, some impacts such as the continuing aesthetic impact and impacts to sensitive habitats could be LARGE, depending on the location.8.2.2 Solar The average capacity factor for this technology is estimated to be between 25 and 40% annually.This technology has high capital costs and lacks base-load capability unless combined with natural gas backup. It requires very large energy-storage capabilities.
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| Based upon solar energy resources, the most promising region of the country for this technology is the West [Reference' 8-16, Section 8.2.4.2].There are also substantial impacts to natural resources (wildlife habitat, land-use, and aesthetic impacts) from construction of'solar-generating facilities.
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| As stated in the GEIS, land requirements are high. Based on the land requirements of 14 acres for every I MWe generated, approximately 10,010 acres would be required to replace the 715 gross MWe produced by PNPS. There is not enough land for either type of solar electric system (photovoltaic or thermal)at the existing PNPS site and both would have LARGE environmental impacts at an alternate site.The construction impacts would be similar to those associated with a large wind farm as discussed in Section 8.2.1. The operating facility would also have considerable aesthetic impact.Solar installations pose no human health risk other than from occupational injuries.
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| The manufacturing process for constructing a large amount of photovoltaic cells would result in waste generation, but this waste generation has not been quantified.
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| Some impacts, such as impacts to sensitive areas, loss of productive land, and the continuing aesthetic impact, could be LARGE, depending on the location.8.2.3 Hydropower Hydroelectric power has an average annual capacity factor of 46%. Section 8.3.4 of the GEIS, indicates that the percentage of the U.S. electrical generation consisting of hydroelectricity is expected to decline because hydroelectric facilities have become difficult to site as a result of public concern over flooding, destruction of natural habitat, and destruction of natural river courses. Section 8.3.4 of the GEIS estimates land use of I million acres per 1,000 MWe (or 1,000 acres per MWe) for hydroelectric power, resulting in a LARGE environmental impact. Due to the lack of locations for siting a hydroelectric facility large enough to replace PNPS, local hydropower is not a feasible alternative to PNPS license renewal [Reference 8-16, Section 8.2.4.3].8-32 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage According the U.S. Hydropower Resource Assessment for Massachusetts (Reference 8-12), there are no remaining sites in Massachusetts that would be environmentally suitable for a large hydroelectric facility.8.2.4 Geothermal Geothermal has an average capacity factor of 90% and can be used for base-load power where available.
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| However as illustrated by Figure 8.4 in the GEIS, geothermal plants might be located in the western continental U.S., Alaska, and Hawaii where geothermal reservoirs are prevalent.
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| This technology is not widely used as base-load generation due to the limited geographic availability of the resource and the immature status of the technology
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| [Reference 8-16, Section 8.2.4.4].
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| This technology is not applicable to the region where the replacement of 715 gross MWe is needed. There are no high temperature geothermal sites in Massachusetts.
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| 8.2.5 Wood Energy A wood-burning facility can provide base-load power and operate with an average annual capacity factor of around 70 to 80% and with 20 to 25% efficiency.
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| The cost of the fuel required for this type of facility is highly variable and very site-specific.
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| The 53 MW McNeil Station, the largest wood-fired generator in the world when it came on line, was developed with great promise as an in-state generating source, a market for low-grade wood to aid Vermont forest management, insulation from volatile oil prices, and a significant employer generating other associated economic benefits [Reference 8-19]. However, since the plant opened in June 1984, McNeil's fuel price of about 3.5 cents/kWh was not competitive with the post-1 986 regime of low oil prices [Reference 8-19]. Among the factors influencing costs are the environmental considerations and restrictions that are influenced by public perceptions, easy access to fuel sources, and environmental factors. In addition, the technology is expensive and inefficient.
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| Current conditions still do not allow McNeil to operate as a base load facility as originally envisioned, but instead gives its owners a price ceiling on the market prices they face [Reference 8-19]. Like many other large plants that came on line at the time of high oil prices, interest rates, and other capital costs, McNeil was an investment that looked better then than it does today[Reference 8-19]. Therefore, economics alone eliminate biomass technology as a reasonable alternative.
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| Estimates in the GEIS suggest that the overall level of construction impact per MW of installed capacity should be approximately the same as that for a coal-fired plant, although facilities using wood waste for fuel would be built at smaller scales [Reference 8-141. Like coal-fired plants, wood-waste plants require large areas for fuel storage and processing and involve the same type of combustion equipment.
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| Because of uncertainties associated with obtaining sufficient wood and wood waste to fuel a base load generating facility, ecological impacts of large-scale timber cutting (e.g., soil erosion and loss of wildlife habitat), and relatively low energy conversion efficiency, Entergy has determined that wood waste is not a feasible alternative to renewing the PNPS operating license.8-33 Pilgrim Nuclear Power Station Applicant's Environmental Report i;o' 'Operating License Renewal Stage 8.2.6 Municipal Solid Waste The initial capital costs for this technology are much greater than the comparable steam-turbine technology found at wood-waste facilities.
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| This is due to the need for specialized municipal solid waste-handling and waste-separation equipment and stricter environmental emissions controls.The decision to burn municipal waste to generate energy is usually driven by the need for an alternative to landfills, rather than by energy considerations.
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| High costs prevent this technology from being economically competitive.
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| Thus, municipal solid waste generation is not a reasonable alternative
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| [Reference 8-16, Section 8.2.4.6].Currently, there are approximately 89 waste-to-energy plants operating in the United States.These plants generate approximately 2,500 MWe, or an average of approximately 28 MWe per plant [Reference 8-13]. Therefore, approximately 26 typical waste-to-energy plants would be required to replace the 715 gross MWe base load capacity of PNPS. Therefore, the generation of electricity from municipal solid waste would not be a feasible alternative to renewal of the PNPS operating license.8.2.7 Other Biomass-Derived Fuels In addition to wood and municipal solid waste fuels, there are several other concepts for fueling electric generators, including burning energy crops, converting crops to a liquid fuel such as ethanol (ethanol is primarily used as a gasoline additive for automotive fuel), and gasifying energy crops (including wood waste). The GEIS points out that none of these technologies has progressed to the point of being competitive on a large scale or of being reliable enough to replace a base-load plant such as PNPS. For these reasons, such fuels do not offer a feasible alternative to PNPS license renewal. In addition, these systems have LARGE impacts on land use [Reference 8-16, Section 8.2.4.7].8.2.8 Oil Oil is not considered a stand-alone fuel because it is not cost-competitive when natural gas is available.
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| The cost of an oil-fired operation is about eight times as expensive as a nuclear or coal-fired operation.
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| In addition, future increases in oil prices are expected to make oil-fired generation increasingly more expensive than coal-fired generation.
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| For these reasons, oil-fired generation is not a feasible alternative to PNPS license renewal, nor is it likely to be included In a mix with other resources except as a back-up fuel [Reference 8-16, Section 8.2.4.8].8.2.9 Fuel Cells Phosphoric acid fuel cells are the most mature fuel-cell technology, but they are only in the initial hps of rrrnmip-c Ji70aio YP Lmdrcm r 0tk hnvA hen installeds in the Hi_ S 8-34 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage feasible for storage of sufficient electricity to meet the base-load generating requirements.
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| This is a very expensive source of generation, which prevents it from being competitive.
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| This technology also has a high land use impact, which, like wind technology, results in a LARGE impact to the natural environment.
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| It is estimated that 35,000 acres of land would be required to generate 1,000 MWe of electricity.
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| Therefore, fuel cells are not considered a feasible alternative to license renewal [Reference 8-16, Section 8.2.4.10).
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| As market acceptance and manufacturing capacity increase, natural-gas-fueled fuel cell plants in the 50- to 1 00-MW range are projected to become available.
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| At the present time, however, fuel cells are not economically or technologically competitive with other alternatives for base load electricity generation, and progress in market growth and cost reduction has been slower than alternatives anticipated
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| [Reference 8-1]. Fuel cells are, consequently, not a feasible alternative to renewal of the PNPS operating license.8.2.10 Delayed Retirement Even without retiring any Entergy owned or non-Entergy owned generating units, it is expected that additional capacity will be required in the near future. Thus, even if substantial capacity were scheduled for retirement and could be delayed, some of the delayed retirement would be needed just to meet load growth.PNPS would be required, in part, to offset any actual retirements that occur. Delayed retirement of other Entergy or non-Entergy generation units is unlikely to displace the need for 650 gross MWe of capacity over the twenty years of extended operation and therefore, would not be a \,_feasible alternative to PNPS license renewal.8.2.11 Utility-Sponsored Conservation The concept of conservation as a resource does not meet the primary NRC criterion "that a reasonable set of alternatives should be limited to analysis of single, discrete electric generation sources and only electric generation sources that are technically feasible and commercially viable". It is neither single, nor discrete, nor is it a source of generation
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| [Reference 8-16, Section 8.2.4.1.2).
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| Market and regulatory conditions in the deregulated environment can be described as follows:* a decline in generation costs, due primarily to technological advances that have reduced the cost of constructing new generating units (e.g., combustion turbines);
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| * national energy legislation, which has encouraged wholesale competition through open access to the generation of electrical energy, as well as state legislation designed to facilitate retail competition.
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| Consistent with these changes, the electricity generation planning environment features lower capacity and lower energy prices than during earlier periods, shorter planning horizons, lower 8-35 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage reserve margins, and increased reliance on market prices to direct utility resource planning.These have greatly reduced the number of cost-effective DSM alternatives.
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| Another significant change includes the adoption of increasingly stringent national appliance standards for most major energy-using equipment and the adoption of energy efficiency requirements in state building codes. These mandates have further reduced the potential for cost-effective generator-sponsored measures.The environmental impacts of an energy conservation program would be SMALL, but the potential to displace the entire generation at PNPS solely with conservation is not realistic.
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| Therefore, the conservation option by itself is not considered a reasonable replacement for the PNPS operating license renewal alternative.
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| 8.2.12 Combination of Alternatives The NRC indicated in the GEIS that, while many methods are available for generating electricity and a huge number of combinations or mixes can be assimilated to meet system needs, such expansive consideration would be too unwieldy given the purposes of the alternatives analysis.Therefore, the NRC determined that a reasonable set of alternatives should be limited to analysis of single discrete electrical generation sources and only those electric generation technologies that are technically reasonable and commercially viable [Reference 8-14, Section 8.1].Consistent with the NRC determination, Entergy has not evaluated mixes of generating sources.8.3 ProDosed Action vs. No-Action The proposed action is the renewal of the operating license for PNPS. The specific review of the eleven environmental impacts, required by 10 CFR 51 .53(c)(3)(ii), concluded that there would be no adverse impact to the environment from the continued operation of PNPS through the period of extended operation.
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| The no-action alternative to the proposed action is the decision not to pursue renewal of the operating license for PNPS. The environmental impacts of the no-action alternative would be the impacts associated with the construction and operation of the type of replacement power utilized.In effect, the net environmental impacts would be transferred from the continued operation of PNPS to the environmental impacts associated with the construction and operation of a new generating facility.
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| This new generating facility would almost certainly be constructed at a greenfield location due to the air impacts associated with constructing one of the viable technologies on the PNPS site. Therefore, the no-action alternative would have negative net environmental benefits.The environmental impacts associated with the proposed action (the continued operation of PNPS) were compared to the environmental impacts from the no-action alternative (the construction and operation of other reasonable sources of electric generation).
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| Entergy believes this comparison shows that the continued operation of PNPS would produce fewer significant environmental impacts than the no-action alternative.
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| There are significant differences in the 8-36 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage impacts to air quality and land use between the proposed action and the reasonable alternative generation sources.In addition, there would be adverse socioeconomic impacts (including local unemployment, loss of local property tax revenue, and higher energy costs) to the area around PNPS from the decision not to pursue license renewal.The Joint DOE-Electric Power Research Institute Strategic Research and Development Plan to Optimize US Nuclear Power Plants stated, "... nuclear energy was one of the prominent energy technologies that could contribute to alleviate global climate change and also help in other energy challenges including reducing dependence on imported oil, diversifying the US domestic electricity supply system, expanding US exports of energy technologies, and reducing air and water pollution." The Department of Energy agreed with this perspective and stated, "...it is important to maintain the operation of the current fleet of nuclear power plants throughout their safe and economic lifetimes" [Reference 8-3]. The renewal of the PNPS operating license is consistent with these goals.8.4 Summary The proposed action is the renewal of the PNPS operating license. The proposed action would provide the continued availability of approximately 715 gross MWe of base-load power generation through 2032.CO 2 emissions from power generation are a major contributor to anthropogenic greenhouse gas emissions and climate change. These emissions result from the efficiency of the technologies used to produce and deliver the energy and the carbon content of the fuel being used. The table below shows a comparison of the CO 2 content of various fuels: (Reference 8-20)Fuel Pounds CO 2 per Million Btu Subbituminous coal 212.7 Bituminous coal 205.3# 6 fuel oil 173.9 Natural gas 117.1 Nuclear 0.0 Renewable sources 0.0 The following table provides an estimate of the CO 2 emissions that would result if other fuel technologies were used to supply the electricity that currently is being produced by PNPS: 715 MWe and an estimated 92% capacity factor. The technologies, fuels, and production efficiencies 8-37 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage shown are based upon 'Greenfield plants" that have recently been permitted as having "Best Available Control Technologies" under the New Source Review Permit program (Reference 8-21).Heat Rate Electricity CO 2 Emissions (BTUIKWh) (MWH/yr) (metric tons C02/yr)Pulverized coal Bituminous coal 9,928 5,762,328 5,327,479 Pulverized coal Subbituminous 9,700 5,762,328 5,392,749 coal Combined cycle Natural gas 6,814 5,762,328 2,085,595 gas turbine The environmental impacts of the continued operation of PNPS, providing approximately 715 gross MWe of base-load power generation through 2032, are less than impacts associated with the best case among reasonable alternatives.
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| The continued operation of PNPS would create significantly less environmental impact than the construction and operation of new base-load generation capacity.F Finally, the continued operation of PNPS will have a significant positive economic impact on the liw; communities surrounding the station., I ! , 1 / i , 1 .I , 8-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8.5 References 8-1 California Stationary Fuel Cell Collaborative, "White Paper Summary of Interviews with Stationary Fuel Cell Manufacturers," August 2002, available at http:II stationaryfuelcells.org/Documents/PDFdocs/(ndustrySurveyReport.pdf.
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| 8-2 U.S. Department of Energy, Energy Information Administration, DOE/EIA-0383(2004), Annual Energy Outlook 2004 With Projections to 2025, Washington, DC, 2004.8-3 U.S. Department of Energy -Electric Power Research Institute, Joint DOE-EPRI Strategic Research and Development Plan to Optimize U.S. Nuclear Power Plants, Volume 1, March 20,1998.8-4 Energy & Environmental Ventures LLC, Wind Energy in the Northeastem U.S. -Leverage Points for Growth, Weston, CT, undated, available at http://www.ctcleanenergy.com/investment/Wind_EnergyNortheasternUS.pdf, accessed on October 25, 2004.8-5 Energy Information Administration, DOE/EIA-0348(99)/2, Electric PowerAnnual 1999, Volume 11, Washington, DC, October 2000, available at http://www.eia.doe.gov/cneaf/
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| electricity/epav2/epav2.pdf, accessed on April 5, 2001.8-6 Energy Information Administration, DOE/EIA-0348(00)/2, Electric PowerAnnual 2000, Volume II, Washington, DC, November 2002, available at http://www.eia.doe.gov/cneaf/
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| electricity/epav2/epav2/htmltables/epav2t25pl.html, accessed on March 2, 2005.8-7 U.S. Environmental Protection Agency, AP-42: Compilation of Air Pollutant Emission Factors, Volume 1: Stationary Point and Area Sources, Section 1.1, "Bituminous and Subbituminous Coal Combustion," AP-42, September 1998, available at http:/l www.epa.gov/ttn/chief/ap42/chO1, accessed on July 26, 2001.8-8 U.S. Environmental Protection Agency, AP-42: Compilation of Air Pollutant Emission Factors, Volume 1: Stationary Point and Area Sources, Section 3.1, "Stationary Gas Turbines for Electricity Generation," April 2000, available at http://www.epa.gov/ttn/
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| chief/ap42/ch03, accessed on July 26, 2001.8-9 Energy Efficiency and Renewable Energy Network, "Massachusetts Wind Resources," 2001, available at http://www.eren.doe.gov/stateenergy/tech.wind.cfm?state=MA, accessed on December 4, 2001.8-10 Federal Highway Administration, "User Guidelines for Waste and Byproduct Materials in Pavement Construction, Coal Bottom Ash/Boiler Slag," 2000, available at http://tfhrc.gov/hnr2O/recycle/waste/cbabsl.htm, accessed on May 29, 2001.8-39 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 8-11 Gabbard, A., "Coal Combustion:
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| Nuclear Resource or Danger," Oak Ridge National Laboratory Review, Oak Ridge National Laboratory, Oak Ridge, TN, 1993, available at http://www.ornl.gov/info/ornlreview/rev26-34/text/colmain.html, accessed on December 11, 2003.8-12 Idaho National Engineering Laboratory, Renewable Energy Products Department, DOE/I D-1 0430(MA), U. S. Hydropower Resource Assessment for Massachusetts, Idaho Falls, ID, July 1995, available at http://jobs.inel.gov/resourceassessmentlma/ma.pdf, accessed on June 5, 2001.8-13 Integrated Waste Services Association, "WASTE-TO-ENERGY:
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| Clean, Reliable, Renewable Power," Washington, DC, June 2004, available at http:llwww.wte.org/pdfs/
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| cleanreliablerenewable.pdf.
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| 8-14 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes 1 and 2, Washington, DC, May 1996.8-15 U.S. Nuclear Regulatory Commission, NUREG-1437, Supplement 2, Generic Environmental Impact Statement for License Renewal of Nuclear Plants -Regarding the Oconee Nuclear Station, Final Report, Washington, DC, December 1999.8-16 U.S. Nuclear Regulatory Commission, NUREG-1437, Supplement 3, Generic Environmental Impact Statement for License Renewal of Nuclear Plants -Regarding the Arkansas Nuclear One, Unit 1, Final Report, Washington, DC, April 2001.8-17 U.S. Nuclear Regulatory Commission, NUREG-1437, Supplement 10, Generic Environmental Impact Statement for License Renewal of Nuclear Plants -Regarding the Peach Bottom Atomic Power Station, Units 2 and 3, Final Report, Washington, DC, January 2003.8-18 National Renewable Energy Laboratory, Pacific Northwest Laboratory, DOE/CH 10093-4, Wind Energy Resource Atlas of the United States, Richland, WA, October 1986, available at http://rredc.nrel.gov/wind/pubs/atlas/titlepg.html, accessed on May 31, 2001.8-19 Vermont Department of Public Service, Vermont Electric Plan 2005, Montpelier, VT, January 19, 2005.8-20 Energy Information Administration, EIA-1605, Fuel and Energy Source Codes and Greenhouse Gas Emission Coefficients, available at http://www.eia.doe.gov/oiaf/1605l factors.html, accessed on October 21, 2005.8-21 U.S. Environmental Protection Agency, RACT/BACT/LAER Clearinghouse (RBLC), available at http://cfpub.epa.gov/RBLC/htm/blO2.cfm, accessed on October 15, 2005.8-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 9.0 STATUS OF COMPLIANCE 9.1 Reauirement
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| [10 CFR 51.45(d)1 The environmental report shall list all Federal permits, licenses, approvals, and other entitlements which must be obtained in connection with the proposed action and shall describe the status of compliance with these requirements.
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| The environmental report shall also include a discussion of the status of compliance with applicable environmental quality standards and requirements including, but not limited to, applicable zoning and land-use regulations, and thermal and other water pollution limitations or requirements which have been imposed by Federal, State, regional, and local agencies having responsibility for environmental protection.
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| 9.2 Environmental Permits Table 9-2 provides a list of the environmental permits held by PNPS and the compliance status of these permits. These permits will be in place as appropriate throughout the period of extended operation given their respective renewal schedules.
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| Other than routine renewals required at frequencies specified by the permits in Table 9-1, no state, federal, or local environmental permits have been identified as being required for re-issuance to support the extension of the PNPS operating license.9.2.1 Coastal Zone Management Program Compliance The Federal Coastal Zone Management Act (16 USC 1451 et seq.) imposes requirements on C)applicants for a federal license to conduct an activity that could affect a state's coastal zone. The Act requires the applicant to certify to the licensing agency that the proposed activity would be consistent with the state's federally approved coastal zone management program [16 USC 1456(c)(3)(A)].
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| The National Oceanic and Atmospheric Administration has promulgated implementing regulations that indicate that the requirement is applicable to renewal of federal licenses for activities not previously reviewed by the state [15 CFR 930.51 (b)(1)]. The regulation requires that the license applicant provide its certification to the federal licensing agency and a copy to the applicable state agency [15 CFR 930.57(a)].
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| The NRC office of Nuclear Reactor Regulation has issued guidance to its staff regarding compliance with the Act [Reference 9-3, Appendix E]. This guidance acknowledges that Massachusetts has an approved coastal zone management program. PNPS, located in Plymouth County, is within the Massachusetts coastal zone [Reference 9-1]. Concurrent with submitting the Applicant's Environmental Report -Operating License Renewal Stage to the NRC, Entergy will submit a copy of the report to the Commonwealth in fulfillment of the regulatory requirement for submitting a copy of the coastal zone consistency certification to the state.9.2.2 Water Quality (401) Certification With respect to applicants for a federal license to conduct an activity that might result in a discharge into navigable waters, section 401 of the CWA establishes certain requirements for 9-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage certifications from the state that the discharge will comply with certain CWA requirements (33 USC 1341). On July 31, 1970, the Massachusetts Water Resources Commission provided a water quality certification reflecting its receipt of reasonable assurance that operation of the Pilgrim Station will not violate applicable water quality standards.
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| Massachusetts provided a further water quality certification on April 15, 1971. Copies of these certifications are provided in Attachment A. In addition, the NPDES permit, which was issued jointly by the EPA pursuant to the CWA and the Commonwealth of Massachusetts pursuant to Massachusetts General Law Chap. 21, § 43, reflects continued compliance with applicable CWA standards.
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| Excerpts of this permit are also Included in Attachment A.9.3 Environmental Permits -Discussion of Comoliance Station personnel are primarily responsible for monitoring and ensuring that PNPS complies with its environmental permits and applicable regulations.
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| Sampling results are submitted to the appropriate agency. PNPS has an excellent record of compliance with its environmental permits, including monitoring, reporting and operating within specified limits.PNPS has an onsite wastewater treatment plant. Sanitary wastewater that does not contain radioactive materials is processed in the wastewater treatment facility and discharged through a permitted drain field to the groundwater.
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| This is regulated through the MDEP, Groundwater Discharge Permit #2-389.Entergy has measures in place to ensure those environmentally sensitive areas are adequately protected during site operations and project planning.
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| These measures include an environmental evaluation checklist and also established controls and methods for evaluating potential environmental affects from plant operations and project planning.
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| Therefore, planned projects or changes in plant operations would be required to undergo an environmental review and evaluation prior to implementation, with appropriate permits obtained or modified as necessary.
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| Table 9-1 Environmental Authorizations for PNPS License Renewal Agency Authority Requirement Remarks U.S. Nuclear Atomic Energy Act License Environmental Report submitted Regulatory (42 USC 2011 et seq.) Renewal in support of license renewal Commission application U.S. Fish and Wildlife Endangered Species Consultation Requires Federal agency issuing Service and National Act Section 7 a license to consult with FWS and Marine Fisheries (16 USC 1636) NMFS. (Attachment B)Service 9-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 9-1 Environmental Authorizations for PNPS License Renewal Agency Authority Requirement Remarks Commonwealth of Endangered Species Consultation Requires Federal agency issuing Massachusetts Act Section 7 a license to consult with FWS at Division of Fisheries (16 USC 1636) the state level. (Attachment B)and Wildlife -Massachusetts Clean Water Act Certification Requires Commonwealth Department of Section 401 certification that discharge would Environmental (16 USC 470f) comply with CWA standards Protection Massachusetts National Historic Consultation Requires Federal agency issuing Historical Commission Preservation Act a license to consider cultural Section 106 impacts and consult with the SHPO. (Attachment C)Massachusetts Office Federal Coastal Zone Certification Requires an applicant to provide of Coastal Zone Management Act certification to the federal agency Management (16 USC 1451 et seq.) issuing the license that license renewal would be consistent with the federally-approved state coastal zone management program. Based on its review of the proposed activity, the state must concur with or object to the applicant's certification.(Attachment D)(J )9-3 i,_'
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 9-2 Environmental Authorizations for Current PNPS Operations Agency l Authority l Requirement l Number l Iss De o Activity Covered Expiration Date Federal Requirements for License Renewal U.S. Nuclear Atomic Energy Act License to DPR -35 Issued 09/15/72 Operation of Unit I Regulatory (42 USC 2011, et seq.), Operate Expires 06/08/12 Commission 10 CFR 50.10 U.S. Nuclear Atomic Energy Act Section Material License 20-07626-04 Issued 02/10/03 Contamination on Regulatory 161, (42 USC 2201), Expires 02/28/13 reactor components Commission 10 CFR 40 and 70 U.S. Department 49 CFR 107, Subpart G Registration 062601551001J Issued 05/16/05 Radioactive and of Transportation Expires 06/30/06 hazardous materials This permit is renewed shipments on an annual basis.U.S. Clean Water Act NPDES Permit Federal Permit: Issued 04/29/91 Plant discharges to Environmental (33 USC 1251 et seq.), MA0003557 Modified 08/30/94 Cape Cod Bay Protection M.G.L. Chapter 21, Massachusetts Expired 04/29/96 Agency and Section 43(2) Permit: 359 (remains in effect Massachusetts pending EPA and Department of Commonwealth action Environmental on renewal applications Protection submitted 10/25/95 and 12/01/99)9-4
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| !a a 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 9-2 Environmental Authorizations for Current PNPS Operations Agency Authority Requirement Number Expiration Date Activity Covered U.S. Fish and Migratory Bird Treaty Act, Depredation MB831184-0 Issued 07/08/2005 Removal of birds and Wildlife Service 16 USC 703-712 Permit Expires 06/30/2006 nests from utility structures This permit is renewed on an annual basis.State Requirements for License Renewal Massachusetts M.GL. Chapter 111, Section Material License 07-6262 Issued 4/22/03 Contamination on Department of 5N Expires 4/30/08 reactor components Public Health Massachusetts M.G.L. Chapter 111, Section Material License 49-0078 Issued 10/11/02 Contamination on Department of 5N Expires 5/31/06 reactor components Public Health l Massachusetts M.G.L. Chapter 148, Section Registration Not applicable This registration is Storing flammable Department of 13 renewed annually on materials in tanks Public Safety April 1.Massachusetts 310 CMR 7.02(11) 50% Facility Issued 7/18/2005 Emissions from Department of 310 CMR 7.02(11)(e)
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| Emission Cap various small Environmental combustion sources Protection 9-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 9-2 Environmental Authorizations for Current PNPS Operations Agency Authority Requirement Number Expiratosn Dato Activity Covered Massachusetts M.G.L. Chapter 21, Sections Groundwater
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| #2-389 Issued 4/20/99 Treated effluent Department of 26-53 Discharge Permit Expires 4/20/04 discharges to Environmental Rnwlapiton groundwater from Protental Renewal application wastewatertreatment submitted 10/14/03.
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| fclt Administratively continued pending review of application State Requirements for License Renewal (continued)
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| Massachusetts M.G.L. Chapter 21C Large Quantity MAR000014167 Issued 10/06/99 Hazardous waste Department of 310 CMR 30 Generator generation Environmental Protection South Carolina South Carolina Radioactive Radioactive 0007-20-01 Issued 12/17/04 Transportation of Department of Waste Transportation and Waste Transport Expires 12/31/05 radioactive waste to Health and Disposal Act (SC ST SEC 13- Permit disposal facility in Environmental 7-110 et seq.) South Carolina Control This permit is renewed on an annual basis.9-6 9 J 2j Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table 9-2 Environmental Authorizations for Current PNPS Operations Issue or Agency Authority Requirement Number Expiration Date Activity Covered Tennessee TCA 68-202-206 Radioactive T-MA004-L01 Issued 12/08/04 Shipment of Department of Waste License- Expires 12/31/05 radioactive waste to Environment and for-Delivery disposal/
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| processing Conservation T facility in Tennessee This permit is renewed.on an annual basis.CFR -Code of Federal Regulations USC -United States Code M.G.L. -Massachusetts General Laws CMR -Code of Massachusetts Regulations TCA -Tennessee Code Annotated SC ST -South Carolina Statutes 9-7 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 9.4 References 9-1 Massachusetts Coastal Zone Management, "Massachusetts Coastal Zone Management," Boston, MA, 2001, available at http://www.state.ma.us/czm/czm.htm, accessed April 23, 2001.9-2 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes 1 and 2, Washington, DC, May 1996.9-3 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,'Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues," NRR Office Instruction No. LIC-203, Revision 1, May 24, 2004.9-8 PlIgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Qw~Attachment A NPDES Permit and Water Quality Certification
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| > Title page and section relevant to the Clean Water Act Section 316(a) and (b)5> Section 401 Water Quality Certification, April 15, 1971> Section 401 Water Quality Certification, July 31, 1970 State Permit No.: :eral permit No. YKAO003557 Page I of 15 Modification go. 1:ODIFICATION OF AUTtOJRXZATION TO DIStANGE tll;R THE NATIONAL POLIMAM' DISCHARE ELMINATION, SYSTEM:In compliance vith tbe provisions of the Federal Clean Water Act, as amended, (33 U.S.C. 111251 it. jig.; the UCWAN), arnd the Massachusetts Clean Water6s Act, as amended, (Ic.G.L. chap. 21,: :126-53 ),-oston Wsi=on Company:ilgri: Nuclear Power Station e00 Boylston Street Boston, Massachusetts 02199 is authorized to discharge in accordance with effluent liztations, ponitoring requirements and other conditionsm t in the previous peotit, except as set forth herein and listed as follows: 1. Page 9, Par. I.A.1 has been changed for the nwV flow rate for Dischargo 003.2. Page 9a, Par. ?.A.4a has been added for the now Discharge 001.3. Page 2, Par. I.A.1.a.(2) change word from Odaily* to-monthly' (typographical error).4. Pagc 5 Par.IA.m.
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| delete "shall" and "circulatiig" (typographical errors) and add m 'no more than' 20,000 gallon batches ( Iclarification).
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| : s. Page 7, Par I.A.2.o add *frca April 1 to November 30 each year" (clarification).
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| : 6. Page 12, Par. I.A.7.1 clarify Discharge
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| #003 contents.This modifies the pernit issued on April 2S, 1991.This permit modification shall become efftectiveon thi date of isgsuane.This permit Modification and the authorization to discharge shall expire at midnight, APril 2ath, 1996.Signed thisgo.. day of Iffy irector D rector of the Oftice of ater Management Division f IVatershed Management Snvironmental Protection Agency Dcpartwmnt of Environxnental Region I -Protection 1 Boton, MA Cozmonwalth of Massachusetts
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| ...Joston, KA : Attachment A
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| Page 3 OE 15 Pemit 1Io. MA0O03557 d. T term *iP&L means the tegional Admiiastrator of Region I of the U. S. Eftviromnental frotection Agency of the Divisom of WAter Pollution Control of the tasiachvsettr Department of Environmentaj Protect ion or his designee+.There shall be no discharge of polychiorinated biphenyl compounds commonly ds-d for transformer fluid.f, There shall be no discharge of treated or untreated chemieals which result fron cleaning or washing of condensers or equipment uherein heavy metals nay be-discharged
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| .g. The rate of chinge of Discharge 5QI Delta-T shall not exceedt (11 a 3 *f eis6 or fll in temperature for any 6-sinarte period during normal tteady state plant operation and (21 a 10 *F rise or fall in tepezrature for any 60-vinute perlod Ourg normal load cycling.Variation in inlet timperiature shall not be considsretd as an operationol rise or fall of tesperatare.
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| wormal startup telpercture rise shell not exce, the muaxlieru allowed in Subparagraph I.A.2.a below. in the event of a reactcr energency shutdown, the allowable decrease of 1C- F/hvjr may be exceede4.
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| tn such an event, the permittee shall report the occurrence in the next monthly W to EVA and the State.h. The thermal plumes from the station: 41) Shtll. rot dtIt*Srio;UIy interfere with the natural movements, reprodactive cycles. or sigratoty pathways of the indicenous populations within the water body segments (2) shall have minimal cortact with the surroutding shorelines.
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| It has been deaetained, besed on .ngneering jv4gsent, that the circulatinq water intake structures presently employs the best technology available for ainiihining ad-verse environrental impact. mny change In the location, design or capacity of the present strcItusre shall be Approved by the Regional Administrator and the Oirector.The present design shall be reviewed for e4nformity to regulations pursuant to Section 1i4b1 of the Act when s-ich are promuleated.
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| .The effluent shall not containmaterials in concent-ations or combinations rhich ore hazardoais or toxi to aCq atic tife or which would impair the uSes designated by the classification of the receiving waters.-Rom ou WDCzES pomst Attachment A
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| Pllgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage , -1. , ...,- ..I aes or 7t DWTE.rk o, TW9VA7 1 wsCb b WAWn CbAWb~ fli.YAC..* A -2 ''Dk'-%&,r SAn4rs WOW^rr4 4kCz 100 nts; $9C4 qvoson 02202 AxriI 15, 1971 Ir. Clauda A.. Pursol Assistant flce Presitent-Nuclear Uoston sdizcn Cospaty 600 Boylston St-Cat loston, ta:asacbhusetts 02199
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| | |
| ==Dear Kr. fNrscls Re:==
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| State Certificatioa filzrit Station aostoo'n edtson Concly-0 eV AI-I C<\4=in response.
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| o your request it lcttei dated ?cbruaars 17, 1971, this Division has tevievd yrut -ptition fLo a Prmit to corstruct and operate a tnstcgater diucharac outlet from a nuclear povtr plant at Ply t, iasaachusetts, Wmoan as the -Lto in Statiga.in accord*avc vith tn provtssions of Section 2l(bXL) of the?cdarAl : Q.aer Qfl~it? lUroveront tct of 1970 (Public 1Lo;J 9 1-2 2 4), this'Dirtsion tereby certifies tzt, based on infornattLe sad 1n-csttaatlcna, there is reasonaebc assurarce that th prot-osed nativity WiLl be cond'ucted In ha rnrentr shtch nut not violate ap51icable, watr qrtzity sttndards zapted by this 3ivision under. author!- v of Section 27(A) of C:hqptor 21 of the Unsn:husotts Gcneral Laws, s&id 'water quality stzaairds hatint, been iled vttt the Cttty OZ Stete C! the Car.nuoIItt o :tCrh C, 15G7.SbcultC any pollution arise tnhrout Or b aeuse of- the peration of the proposed facility r tfhloutb failure cn eoMply uith thts iviaslon's 2LhU0 cad tutatioas pertanin:^
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| co waste disposal, the Division will direct thsat the condltion be eorrected.
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| ;. :c-cc1tenCV1 Om the part Or tM licensee .h11 bz cause for this hintkton to wco:=Cnad the revocation of the licenso css*Ad therefor or to tahe such other actien as Is authorized by the 0nzrzl Lrw., of the Co~ooecLat;.
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| Vert*: truly yours, T Q;-Lt5flo~
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| That .s1 C. :tcratton Director z cc: Chite, Parits SBn-ctl, trera¶.ins DIvsion, Corps of Snpaecrs, 424 .tcclu toad,, a11hwt N:zos. 02154 .Mroceate CO--isslonsr, 1.:ter:.ars Division, tepartrt:
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| _142 -.4shvu Straet, :ton, %ass. 02114 RECEIVED APR 21 1371 FILGRtN P&OCECT Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage e -re 4E° "M ff".4-ib, e flwom 9S~sascss t e&J or 1mg af q0 W. Claude Piroel.AsL~sistat Vicerest-dent$
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| _ Nc P e Bovton retizenCopy flwinautb, Maaahsot o yistn Street Bostcn, mC&Cllf 02599
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| | |
| ==Dear n 7 A o cCt;y et the Dirsim b retnr4 roaeable usnnco ta:==
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| t operton or ts proposd P n Stavo d o oltc applcale eninrcef report btd by tht Doeto8 Zdison Ccn;c-, ond Cifatg mxbaeCJtn det s ith the COilatS re-tCAdvia=oy Condttee ;pertainl to nologicl and n socc d before at attcr opert-ion.
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| C r tor be D sion b s sued an In iU. c perct rOr a new wast4 e Cisebrgo cUtlet'for this Caeifltr.
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| Tis per-it is vaid fora per-iod oC thrce ycm fron date of sztrt-op.
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| Sboucd the before an ster stUdy indictetc a zee orfnl eotrols V-4or trntt ner of the ple efMuents c= cotros ' sbe prvded e- by Bost=n Edsas.C Te fcrezos ~et-ntoan is to coclpy wth Sectlo 21 (3) (1) of the Pdercal Water Q4flity Zwrcvocrnt Act of I7 ( lio La 9IC224 )Vfry tfl3S y*our, Viomcs C. XcXtho Drectbo TCsJRswlv PlIgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Attachment B Special Status Species Correspondence
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| > Letter from Stephen Bethay, Entergy, to Mike Bartlett, FWS, dated February 3, 2005> Letter from Michael J. Amaral, FWS, to Stephen Bethay, Entergy, dated March 9, 2005> Letter from Stephen Bethay, Entergy, to Christopher Mantzaris, NMFS, dated February 3, 2005> Letter from Mary A. Colligan, NMFS, to Stephen Bethay, Entergy, dated March 4, 2005> Letter from Stephen Bethay, Entergy, to Jenna Garvey, MDFW, dated February 3, 2005> Letter from Thomas W. French, PhD, MDFW, to Stephen Bethay, Entergy, dated April 8, 2005> Letter from Christine Vaccaro, MDFW, to Phil Moore, TtNUS, dated July.6, 2001 (This letter is in response to Entergy's request for information on protected species in the vicinity of PNPS)
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| Entergy Nuclear Generation Company£21EflI Pilgrim Nuclear Power Station E teL, 6 600 Rocky Hill Road PlyITI(Uth.
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| MsA 02360 February 3, 2005 Mr. Mike Bartlett Project Leader U.S. Fish and Wildlife Service New England Field Office 70 Commercial Street Suite 300 Concord, NH 03301-5208 SUBJECTr:
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| Pilgrim Nuclear Power Station Request for Information on Threatened or Endangered Species
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| | |
| ==Dear Mr. Bartlett:==
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| Entergy Nuclear Generation Company (Entergy) is preparing an application tzo the U.S. Nuclear Regulatory Commission (NRC) to renew the operating license for Pilgrim Nuclear Power Station (PNPS). The current operating license for the Station expires in June 2012. As part of the license renewal process, the NRC requires license applicants to "assess the impact of the proposed action on threatened or endangered species in accordance with the Endangered Species Act" (10CFR5 1.53). The NRC will request an informal consultation with your office at a later date under Section 7 of the Endangered Species Act. By contacting you early in the application process, we hope to identify any issues that need to be addressed or any information your office may need to expedite the NRC consultation.
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| Entergy and Boston Edison Company, the previous owner of the Station, have operated PNPS since 1972. The Station lies on the western shore of Cape Cod Bay in Plymouth County, Massachusetts, just east of the Town of Plymouth (see attached Figure 2-1). Entergy purchased PNPS from Boston Edison Company in 1999. When Entergy purchased PNPS, it did not purchase the transmission facilities.
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| While divesting itself of fossil and nuclear generating facilities, NSTAR (the parent company of Boston Edison) retained ownership of transmission facilities.
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| Two transmission lines were built in the early 1970s to connect PNPS to the regional electric grid. These 345 KV transmission lines, which share a single corridor, run south from PNPS to the Snake Hill Road Tap approximately 6 miles south of the station (see attached Figure 2-2).Based on a review of company documents (surveys and monitoring studies) and information on the Massachusetts Geographic Information System and Massachusetts Division of Fisheries
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| &Wildlife websites, Entergy believes that no Federally listed terrestrial species occur on the PNPS site proper or withinfalong the associated 7.2 mile-long transmission corridor.
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| The PNPS-to-Snake Hill Road transmission corridor crosses habitat designated critical (at 50 CFR 17.95) for the endangered Northern Red-Bellied Cooter (Pseudemys rubiventris bangsi), but the part of the critical habitat crossed by the transmission corridor appears to be a buffer area for the population rather than high-quality turtle habitat Northern Red-Bellied Cooters have never been observed by Boston Edison, Entergy, or NSTAR biologists in this transmission corridor.
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| As noted above, Entergy does not own or maintain the transmission lines that run from PNPS to the Snake Hill Road Tap, and is not involved in vegetation management in the right-of-way.
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| Several listed terrestrial species are known to occur in the general vicinity of the PNPS site, however, and cannot be ruled out as occasional visitors to the PNPS site and environs.
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| These include the bald eagle, piping plover, and roseate tern. Bald eagles are present year-round in Massachusetts and congregate in significant numbers in wintering areas along the coast of Cape Cod and Buzzard's Bay. Bald eagles have been observed foraging in the general vicinity of PNPS. but are not believed to nest in the area. Piping plovers nest in summer on sandy coastal beaches along the Massachusetts coast. No suitable piping plover nesting habitat is found on the PNPS site (the shoreline in the area is rocky); however, individual birds may move through the PNPS area when migrating to breeding areas further north of Plymouth Bay and returning to wintering areas along the south Atlantic and Gulf coasts. Like the piping plover, the roseate tern nests in colonies along the Massachusetts coast in summer. The roseate tern nests in dune areas with thick vegetative cover, always in association with the common tem. Although suitable nesting habitat has not been identified at PNPS, migrating terns may move through the site in late spring (en route to nesting areas in Maine and Nova Scotia) and late summer (en route to wintering areas in the West Indies and Latin America).PNPS, a one-unit nuclear plant with a total rated output of 688 MWe (megawatts electrical), uses a once-through cooling water system that withdraws from and discharges to Cape Cod Bay. A recently-prepared Clean Water Act Section 316 Study' that was submitted to EPA Region I in 2000 concluded that the PNPS cooling water intake system has not resulted in adverse impacts to the integrity of Cape Cod Bay fish and shellfish populations, including a number of Representative Important Species (e.g., American lobster, winter flounder, rainbow smelt, cunner, alewife, and Atlantic silverside).
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| Boston Edison and Entergy have monitored the marine fishes of western Cape Cod Bay since 1969 to assess possible impacts of PNPS operations.
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| These monitoring studies also suggest that PNPS operations have not had a significant effect on local and regional fish populations.
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| Trends in abundance of groundfish, pelagic fish, and shellfish (lobsters in particular) in western Cape Cod Bay mirror population trends in the larger Gulf of Maine and western North Atlantic and do not appear to be influenced by PNPS operations.
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| A number of listed marine species (including S great whales and S sea turtles) ae known to use Cape Cod Bay at certain times of the year, but none of these species is believed to forage, feed, rest, or reproduce in the shallow waters adjacent to PNPS. Federally listed whales known to migrate along the coast of Massachusetts include the Sei whale, right whale, blue whale, finback whale, and humpback whale. These great whales pass Cape Cod during seasonal migrations and sometimes forage in Cape Cod Bay. Five sea turtle species (loggerhead, green, leatherback, hawksbill, and Kemp's ridley) occur along the Massachusetts coast, but sightings are uncommon and limited for the most part to sub-adult "wanderers." Young sea turtles are occasionally stranded on Cape Cod beaches.Because whales do not move into the shallow waters immediately offshore of PNPS, they are not affected by operation of the PNPS cooling water intake system or by the station's thenmal discharge.
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| No sea turtles have been observed in the vicinity of the station, and none have been impinged since operational monitoring began in the 1970s. There are no records of sea turtles congregating in the area of the PNPS discharge canal, and no indication that the thermal effluent 1ENSR, 2000, "Combined 316 Demonstration Report -Pilgrim Nuclear Power Station," Prepared for Entergy Nuclear Generation Company. March.
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| has disrupted normal seasonal movement or migration of turtles.Entergy is committed to the conservation of significant natural habitats and protected species, and expects that operation of the Station through the license renewal period (an additional 20 years)would not adversely affect any listed species. Entergy has no plans to alter current operations over the license renewal period. Any maintenance activities necessary to support license renewal would be limited to previously disturbed areas. No expansion of existing facilities is planned, and no additional land disturbance is anticipated in support of license renewal. We therefore request your concurrence with our determination that license renewal would have no effect on threatened or endangered species (including candidate species and species proposed for listing)and that formal consultation is not necessary.
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| Please do not hesitate to call me at 508-830-7832 if you have any questions or require any additional information.
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| After your review, we would appreciate your sending a letter detailing any concerns you may have about any listed species in the area or confirming Entergy's conclusion that operation of PNPS over the license renewal term would have no effect on any threatened or endangered species under the jurisdiction of the U.S. Fish and Wildlife Service.Entergy will include a copy of this letter and your response in the Environmental Report that will be submnitted to the NRC as part of the PNPS license renewal application.
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| Sincerely, Stephen Bethay Director, Nuclear Assessment Pilgrim Nuclear Power Station Entergy Nuclear Generation Company
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| | |
| ==Enclosure:==
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| | |
| Figures 2-1 and 2-2 from ER Cc: Fred Mogolesko, Entergy Jacob Scheffer, Entergy Jack Alexander, Entergy Jack Fulton, Entergy David Lach, Entergy Pilgrim Nuclear Power Station 2-37 Environmental Report for License Renewal (1w-~(4up'LEGEND Interstate PNPS FIGURE 2-2 Pm Secondary road 6SIMIl eM Vicinity with Transmission Une (TL) (1 7 Urmmission Us. Map Plymouth Red-Bellied C o n r a d_ _ _ _ _Turte Crifcal Habitat o5D As .w22 3 l blad 15.5005 1.S 2 2.O 3 Mks It-h- -_2-38 J .% J.--__ ;.-United States Department of the Interior FISH AND WILDLIFE SERVICE New England Field Office 70 Commercial Street, Suite 300 Concord, New Hampshire 03301-5087 March 9, 2005 Stephen Bethay Entergy Nuclear Generation Company 600 Rocky Hill Road Plymouth, MA 02360
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| ==Dear Mr. Bethay:==
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| We are in receipt of your February 3, 2005 letter regarding the license renewal process forthe Pilgrim Nuclear Power Station (PNPS), Plymouth, Massachusetts.
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| The following comments are provided in accordance with Section 7 of the Endangered Species Act (ESA) of 1973, as amended (1-6 U.S.C.1531-1543).
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| The federally-threatened piping plover (Charadrin1s melois) and federally-endangered roseate tern (Mlerna dougallli) are known to occur along Plymouth Beach, just north of the PNPS. Occasional wintering bald eagles (Haliaeehus leucocephalicy) are also sometimes present in the area. According to our records, none of the above-listed species are known to frequent the immediate vicinity of PNPS and, therefore, the presence of these species near the power station is probably transient in nature.As stated in your letter, the PNPS-to-Snake Hill Road transmission corridor crosses critical habitat for the endangered red-bellied cooter (Pseudemys nrbriveniris).
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| We concur with your determination that the area crossed by the transmission line does not provide the specific biological habitat needs for the red-bellied cooter. However, turtles may traverse the transmission line corridor and the area is considered critical based on its value to butYer against activities that may degrade water quantity and quality in ponds occupied by the species.Information was provided regarding several marine mammals and turtles. Jurisdiction for those species resides with the National Marine Fisheries Service. We suggest you contact them at their Gloucester, Massachusetts office at 978-281-9300 with regard to the relicensing of the PNPS. Since no expansion of existing facilities is planned and no additional land disturbance is anticipated, we concur with your determination that license renewal for PNPS is not likely to adversely affect federally-listed species subject to the jurisdiction of the U.S. Fish and Wildlife Service, and that formal consultation with us is not required.Thank you for your coordination.
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| Please contact us at 603-223-2541 if we can be of further assistance.
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| Sincerely yours, Michael J. Amaral Endangered Species Specialist New England Field Office t g Entergy Nuclear Generation Company tergy Pilgrim Nuclear Power Station 600 RoCky Hill Road Plymouth, MA 02360 February 3, 2005 Mr. Christopher Mantzaris Asst. Regional Administrator for Protected Resources National Marine Fisheries Service Northeast Regional Office One Blackburn Drive Gloucester, MA 01930-2298
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| | |
| ==SUBJECT:==
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| Pilgrim Nuclear Power Station Request for Information on Threatened or Endangered Species
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| ==Dear Mr. Mantzaris:==
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| | |
| Entergy Nuclear Generation Company (Entergy) is preparing an application to the U.S. Nuclear Regulatory Commission (NRC) to renew the operating license for Pilgrim Nuclear Power Station (PNPS). The current operating license for the Station expires in June 2012. As part of the license renewal process, the NRC requires license applicants to "assess the impact of the proposed action on threatened or endangered species in accordance with the Endangered Species Act" (IOCFR51.53).
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| The NRC will request an informal consultation with your office at a later date under Section 7 of the Endangered Species Act. By contacting you early in the application process, we hope to identify any issues that need to be addressed or any information your office may need to expedite the NRC consultation.
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| Entergy and Boston Edison Company, the previous owner of the Station, have operated PNPS since 1972. The Station lies on the western shore of Cape Cod Bay in Plymouth County, Massachusetts, just east of the Town of Plymouth (see attached Figure 2-1). Semi-enclosed Cape Cod Bay has a surface area of 430 square nautical miles, or 365,000 acres, and connected to a much larger body of water, the Gulf of Maine, which is bounded on the west by the shorelines of Massachusetts, New Hampshire, Maine, and New Brunswick and on the east by the undersea landforns (Georges Banks being perhaps the most notable) that separate the Gulf of Maine from the rest of the North Atlantic.PNPS, a one-unit nuclear plant with a total rated output of 688 MWe (megawatts electrical), uses a once-through cooling water system that withdraws from and discharges to Cape Cod Bay. A recently-prepared Clean Water Act Section 316 Study' that was submitted to EPA Region I in 2000 concluded hat the PNPS cooling water intake system has not resulted in adverse impacts to the integrity of COtpe Cod Bay fish and shellfish populations, including a number of Representative Important Species (e.g., American lobster, winter flounder, rainbow smelt, cunner, alewife, and Atlantic silverside).
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| Boston Edison and Entergy have monitored the marine fishes of western Cape Cod Bay since 1969 to assess possible impacts of PNPS operations.
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| These monitoring studies also suggest that PNPS operations have not had a significant effect on local and regional fish populations.
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| Trends in abundance of groundfish, pelagic fish, and shellfish (lobsters in particular) in western Cape 1ENSR, 2000, 'Combined 316 Demonstration Report -Pilgrim Nuclear Power Station," Prepared for Entergy Nuclear Generation Company. March.
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| Cod Bay mirror population trends in the larger Gulf of Maine and western North Atlantic and do not appear to be influenced by PNPS operations.
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| (In more than 30 years of monitoring the aquatic populations of western Cape Cod Bay, Entergy, Boston Edison Company, and their contractors have never collected a listed marine species. No listed species have been observed in the PNPS intake canal or discharge canal. None have been impinged or entrained in the Station's cooling water.A number of listed marine species (including 5 great whales and 5 sea turtles) are known to use Cape Cod Bay at certain times of the year, but none of these species is believed to forage, feed, rest, or reproduce in the shallow waters adjacent to PNPS. Federally listed whales known to migrate along the coast of Massachusetts include the Sei whale, right whale, blue whale, finback whale, and humpback whale. These great whales pass Cape Cod during seasonal migrations and sometimes forage in Cape Cod Bay. Five sea turtle species (loggerhead, green, leatherback, hawksbill, and Kemp's ridley) occur along the Massachusetts coast, but sightings are uncommon and limited for the most part to sub-adult "wanderers." Young sea turtles are occasionally stranded on Cape Cod beaches.Because whales do not move into the shallow waters immediately offshore of PNPS, they are not affected by operation of the PNPS cooling water intake system or by the station's thermal discharge.
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| No sea turtles have been observed in the vicinity of the station, and none have been impinged since operational monitoring began in the 1970s. There are no records of sea turtles congregating in the area of the PNPS discharge canal, and no indication that the thermal effluent has disrupted normal seasonal movement or migration of turtles.Entergy is committed to the conservation of significant natural habitats and protected species, and expects that operation of the Station through the license renewal period (an additional 20 years)would not adversely affect any listed marine species. Entergy has no plans to alter current 0 operations over the license renewal period. Any maintenance activities necessary to support license renewal would be limited to previously disturbed areas. No expansion of existing facilities is planned, and no additional land disturbance is anticipated in support of license renewal. We therefore request your concurrence with our determination that license renewal would have no effect on threatened or endangered marine species (including candidate species and species proposed for listing) and that formal consultation is not necessary.
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| Please do not hesitate to call me at 508-830-7832 if you have any questions or require any additional information.
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| After your review, we would appreciate your sending a letter detailing any concerns you may have about any listed species in the area or confinning Entergy's conclusion that operation of PNPS over the license renewal term would have no effect on any threatened or endangered species under the jurisdiction of the National Marine Fisheries Service.Entergy will include a copy of this letter and your response in the Environmental Report that will be submitted to the NRC as part of the PNPS license renewal application.
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| Sincerely, Azt 4'Stephen Bethay Director, Nuclear Assessment Pilgrim Nuclear Power Station Entergy Nuclear Generation Company I
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| | |
| ==Enclosure:==
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| | |
| Figure 2-1 from ER Cc: Fred Mogolesko.
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| Entergy Jacob Scheffer, Entergy Jack Alexander, Entergy Jack Fulton, Entergy David Lach, Entergy Pilgrim Nuclear Power Station*.y.: n , ~~. , 4-.h, .-'~~~- ;* ;' _kg;,;.,....~*~';
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| PNPS FIGURE 2-1 50-Mile Vicinity Map en !' ,iig ;'''Ik1n 0 6 10 15 20 25 30 35 ametem 2-37 f UNITED STATES DEPARTMENT OF COMMERCE S A National Oceanic and Atmospheric Administration s [NATIONAL MARINE FISHERIES SERVICE NORTHEAST REGION O1ne Blackbum Drve a'illvt,,0 G Gloucster, MA 01302298 Stephen Bethay' M4AR -4 ZD Director, Nuclear Assessment Entergy Nuclear Generation Company Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Re: Pilgrim Nuclear Power Station, Protected Species
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| | |
| ==Dear Mr. Bethay,==
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| This is in response to your letter dated February 3, 2005, requesting information on the presence of any federally threatened or endangered species under the jurisdiction of the National Marine Fisheries Service (NMFS) in the vicinity of the Pilgrim Nuclear Power Station (PNPS), located on the western shore of Cape Cod Bay in Plymouth County, MA. Entergy Nuclear Power Station is currently preparing an application to the U.S. Nuclear Regulatory Commission (NRC)for the renewal of the operating license for PNPS, as the current operating license expires in June iJ 2012, the information requested is to assist with the application process.As mentioned in your letter, four species of federally threatened or endangered sea turtles and three species of endangered whales may be found in the waters of Cape Cod. The sea turtles in northeastern nearshore waters are typically small juveniles with the most abundant being the federally threatened loggerhead (Caretta caretta) followed by the federally endangered Kemp's ridley (Lepidochelys kempi). Loggerhead turtles have been found to be relatively abundant off the Northeast coast (from near Nova Scotia, Canada to Cape Hatteras, North Carolina).
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| Loggerheads and Kemp's ridleys have been documented in waters as cold as I 0 C, but generally migrate northward when water temperatures exceed 161C. These species are typically present in Massachusetts waters from June -October. Federally endangered leatherback sea turtles (Dermochelys coriacea) are located in Massachusetts waters during the warmer months as well.While leatherbacks are predominantly pelagic, they may occur close to shore, especially when pursuing their preferred jellyfish prey. Green sea turtles (Chelonia mydas) may also occur sporadically in Massachusetts waters, but those instances would be rare.Federally endangered North Atlantic right whales (Eubalaena glacialis), humpback whales (Megaptera novaeangliae), and fin whales (Balaenopteraphysalus) may all also be found seasonally in Massachusetts waters. North Atlantic right whales have been documented in the nearshore waters of Massachusetts from December through June. Humpback whales feed during the spring, summer, and fall over a range that encompasses the eastern coast of the United States. Fin whales are common in waters of the United States Exclusive Economic Zone, principally offshore from Cape Sipe Hatteras northward.
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| While these whale species are not considered residents of the Cape Cod Bay area, it is possible that transients may enter the area during seasonal migrations.
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| It is the understanding of NMFS that there have been no interactions or impingements of sea turtles at PNPS in the past 30 years of monitoring at PNPS. However, since the entrainment and impingement of sea turtles at several nuclear power plants on the East Coast has been documented, and as sea turtles may be seasonally present in the vicinity of the intakes associated with the PNPS, NMFS recommends that this impact be fully addressed in the application being prepared.Section 7(a)(2) of the Endangered Species Act (ESA) of 1973, as amended, states that each Federal agency shall, in consultation with the Secretary, insure that any action they authorize, fund, or carry out is not likely to jeopardize the continued existence of a listed species or result in the destruction or adverse modification of designated critical habitat. Any discretionary federal action that may affect a listed species must undergo Section 7 consultation.
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| As listed species may be present in the project area, the federal action agency, in this case the NRC, is responsible for determining whether the proposed action is likely to affect any listed species. The NRC should then submit their determination along with a request for concurrence, to the attention of the Endangered Species Coordinator, NOAA Fisheries, Northeast Regional Office, Protected Resources Division, One Blackburn Drive, Gloucester, MA 01930. After reviewing this information, NOAA Fisheries would then be able to conduct a consultation under section 7 of the ESA.Should you have any questions about these comments or about the section 7 consultation process in general, please contact Sara McNulty at (978) 281-9328 ext. 6520.Sincerely, Maryligan Assistant Regional Administrator for Protected Resources Cc: Boelke, F/NER4 File Code: Sec 7, Pilgrim Nuclear Power Station. Spp. Pres.
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| Entergy Nuclear Generation Company'E- n te Pilgrim Nuclear Power Station E ter y -.600 Rocky Hill Road Plymouth.
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| MA 02360 February 3, 2005 Ms. Jenna Garvey Environmental Review Assistant Massachusetts Division of Fisheries
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| & Wildlife Natural Heritage & Endangered Species Program Route 135 Westborough, MA 01581
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| | |
| ==SUBJECT:==
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| Pilgrim Nuclear Power Station Request for Information on Threatened and Endangered Species
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| | |
| ==Dear Ms. Garvey:==
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| Entergy Nuclear Generation Company (Entergy) is preparing an application to the U.S.Nuclear Regulatory Commission (NRC) to renew the operating license for Pilgrim Nuclear Power Station (PNPS). The current operating license for the Station expires In June 2012. As part of the license renewal process, the NRC requires license applicants to Sassess the impact of the proposed action on threatened or endangered species in accordance with the Endangered Species Act" (1 OCFR51.53).
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| The NRC will consult with the U.S. Fish and Wildlife Service under Section 7 of the Endangered Species Act and may also seek your assistance in the Identification of important species and habitats In the project areas. By contacting you early In the application process, we hope to Identify any issues that need to be addressed or any information your office may need to expedite the NRC consultation.
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| Entergy and Boston Edison Company, the previous owner of the Station, have operated PNPS since 1972. The Station lies on the western shore of Cape Cod Bay in Plymouth County, Massachusetts, just east of the Town of Plymouth (see attached Figure 2-1).Entergy purchased PNPS from Boston Edison Company in 1999. When Entergy purchased PNPS, It did not purchase the transmission facilities.
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| While divesting itself of fossil and nuclear generating facilities, NSTAR (the parent company of Boston Edison)retained ownership of transmission facilities.
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| Two transmission lines were built in the early 1970s to connect PNPS to the regional electric grid. These 345 KV transmission lines, which share a single corridor, run south from PNPS to the Snake Hill Road Tap approximately 6 miles south of the station (see attached Figure 2-2).Entergy is committed to the conservation of significant natural habitats and protected species, and believes that operation of PNPS and Its transmission lines since 1972 has had no adverse impact on any threatened or endangered species. Based on our review of the various Natural Heritage and Endangered Species Program data layers (downloaded from MassGIS) and the list you provided (Vaccaro, Division of Fisheries and Wildlife, to Moore, Tetra Tech NUS, July 6, 2001), no state-listed species occurs on the PNPS site property, the area owned and managed by Entergy. A number of federally-listed species occur seasonally In Plymouth County in the general vicinity of PNPS, but the likelihood of adverse impacts to these species is small. For example, piping plovers and roseate terns could move through the PNPS site during spring and fall migrations, but would not nest in the area or be affected by plant operations.
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| A number of great whale and sea turtle species occur in Cape Cod Bay, but none has been observed in the shallow waters offshore of PNPS by Boston Edison or Entergy biologists conducting studies of fish and shellfish.
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| Entergy has no plans to alter current operations over the license renewal period. Any maintenance activities necessary to support license renewal would be limited to previously disturbed areas. No expansion of existing facilities Is planned, and no additional land disturbance is anticipated in support of license renewal. As a consequence, we believe that operation of the plant, including maintenance of the transmission lines, over the license renewal period (an additional 20 years) would not adversely affect any threatened or endangered species.After your review, we would appreciate your sending a letter detailing any concerns you may have about any listed species in the project area or confirming Entergy's conclusion that the operation of Pilgrim Nuclear Power Station over the license renewal term would have no effect on any state- or federally-listed species.We will include a copy of your July 6,2001 letter and any additional correspondence from your office in the license renewal application that we submit to the NRC.Please do not hesitate to call me at 508-830-7832 if you have any questions or require any additional information.
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| Sincerely, Stephen Bethay Director, Nuclear Assessment Pilgrim Nuclear Power Station Entergy Nuclear Generation Company Encds: Figure 2-1 Figure 2-2 Cc: Fred Mogolesko, Entergy Jacob Scheffer, Entergy Jack Alexander, Entergy Jack Fulton, Entergy David Lach, Entergy
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| -Pilrm Nuclear Power Station N.;' 'S FIUR :'- ..nc<,: , 5 0150 2 025Knf ,, ..' .,;. ..-:%.o x,..i.' :,.., .... -, : .... :.F:, .,,.,,,.k1,4
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| -, :, .....:i.S FIGURE 2,; -1;, 1 i'':,'~~1
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| ''.5' 20 25 30"z' 35 "> Klmts'',A '' '-''''- 2.37 LEGEND Idntrstat , County Boundaries klnd Environmental Report for Ucense Renewal-/I 0U'}0J 2-38 Commonwealtl of Massachusets Division of Mass Wildlife Wayne F. MacCallum, Director April 8, 2005 Entergy Nuclear Generation Company Pilgrim Nuclear Power Station Attn: Stephen Bethay 600 Rocky Hill Road Plymouth, MA 02360 RE: Pilgrim Nuclear Power Plant Plymouth, MA Renewal of Operating License NHESP File No. 04-16063
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| | |
| ==Dear Mr. Bethay,==
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| Thank you for contacting the Natural Heritage and Endangered Species Program (NHESP) of the MA Division of Fisheries and Wildlife for information regarding state-listed rare species at the above referenced site.As you are aware from our previous letters, there are state-protected rare species that occur within proximity to the above site. According to the I lh edition of the Massachusetts Natural Heritage Atlas. a majority of Priority Habitat 1320 (PH 1320) and Estimated Habitat 148 (WH 148) falls within a half mile radius to the subject project location.
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| The Spotted Turtle (Clemmys guttata), a state-listed species of Special Concern is located in this Estimated Habitat polygon.This species is protected under the Massachusetts Endangered Species Act (MESA) (M.G.L. c. 13 l A) and its implementing regulations (321 CMR 10.00). State-listed wildlife are also protected under the state's Wetlands Protection Act (WPA) (M.G.L. c. 131, s. 40) and its implementing regulations (310 CMR 10.37 and 10.59). Fact sheets for this species can be found on our website http://www.nhesp.otr With regard to determining the potential impacts this project would have on this and other state-listed species, it is not something that can be assessed without more specific information regarding the details associated with the operation of the power plant. If there are no plans to expand the footprint or to alter current operations over the license period, then it would not seem likely that there would be an adverse affect on state-protected wildlife species. However, the NHESP can not at this time officially make this determination unless we were to receive more detailed information in order to conduct a full environmental review. If you have any further questions, please contact Jenna Garvey, Environmental Review Assistant at: (508) 792-7270, extension 303.Sincerely, Thomas W. French, Ph.D.Assistant Director cc: Plymouth Conservation Commission www. masswildlife.
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| ore Division of Fisheries and Wildlife Field Headquarters, One Rabbit Hill Road, Westborough, MA 01581 (508) 792-7270 Fax (508) 792-7275 An Agency of the Department of Fisheries.
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| Wildlife & Environmental Law Enforcement IPilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Qf)Attachment C Massachusetts Historical Commission Correspondence
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| > Letter from Stephen Bethay, Entergy, to Brona Simon, Massachusetts Historical Commission, dated February 17, 2005> Letter from Eric S. Johnson, Massachusetts Historical Commission, to Stephen Bethay, Entergy, dated March 14, 2005 Entergy Nuclear Generation Company Pilgrim Nuclear Power Station~ En (gj}( PlyEnte.MAr20
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| 'En 600 Rocky Hill Road February 17, 2005 Brona Simon Assistant Director Massachusetts Historical Commission 220 Morrissey Blvd.Boston, MA 02125
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| | |
| ==Subject:==
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| PILGRIM NUCLEAR POWER STATION LICENSE RENEWAL REQUEST FOR INFORMATION ON HISTORIC / ARCHAEOLOGICAL RESOURCES
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| | |
| ==Dear Ms. Simon:==
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| Entergy Corporation is preparing an application to the Ui. S. Nuclear Regulatory Commission (NRC) to renew the operating license for the Pilgrim Nuclear Power Station (PNPS), which expires In 2012. Entergy intends to submit this application for license renewal In December 2005. As part of the license renewal process, the NRC requires license applicants to "assess whether any historic or archaeological properties will be affected by the proposed project." The NRC may also request an informal consultation with your office at a later date under Section 106 of the National Historic Preservation Act of 1966, as amended (16 USC 470) and the Federal Advisory Council on Historic Preservation regulations (36 CFR 800). By contacting you early in the application process, we hope to identify any issues that need to be addressed or any information your office may need to expedite the NRC consultation.
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| Pilgrim Nuclear Power Station (PNPS) is located In the Town of Plymouth, Plymouth County, Massachusetts, on the rocky western shoreline of Cape Cod Bay. This location is latitude 410 56' 69" North and longitude 70° 34' 74" West (latitude
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| +41.9444 and longitude
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| -70.5794).
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| The site consists of approximately 1700 acres. Less than 200 acres, between Rocky Hill Road and Cape Cod Bay, are developed with a nuclear reactor containment building, turbine and auxiliary buildings, Intake and discharge structures, a diesel generator building, the switchyard, and associated transmission lines. The remainder of the site is In a forest management trust (see Figure 2-3).The area within 6 miles of the site is completely within Plymouth County and includes the Town of Plymouth, the center of which Is about 4 miles northwest of PNPS and has a population of 51,701 (Bureau of the Census 2000) (see Figure 2-2). The nearest major metropolitan cities are Boston, Massachusetts (36 miles to the northwest), and Providence, Rhode Island (44 miles to the west).An examination of the archaeological site files and maps maintained by the Office of the State Archaeologist at the Massachusetts Historical Commission revealed approximately 130 archaeological (pre-historic and historic) sites within a 6-mile radius of the station.While Entergy does not own the transmission lines and is not responsible for maintaining the transmission corridors rights-of-way, NRC regulations (10 CFR 51) require the utility seeking license renewal to evaluate the impact to transmission corridors from license renewal. Four sites (#84, #813. #815, and #816) appear to fall wfthin or near the Jordan Road transmission line corridor right-of-way (see Figure 2-2). Beyond the Jordan Road tap, site # 361 appears to fall near the corridor.
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| Original surveys of the site property identified several archaeological sites, however they were ultimately determined to be insignificant (AEC 1974).Currently, 92 uabove-ground" locations are listed in the National Register of Historic Places for Plymouth County (U. S. Department of the Interior 2001). The attached table lists the 18 sites located within the Town of Plymouth.
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| The State Register of Historic Places 2000, a report published by the Massachusetts Historical Commission, states that the Town of Plymouth is home to 21 sites of historic significance (Massachusetts Historical Commission 2000).Entergy has no plans to alter current operations over the license renewal period. No major expansion of existing facilities is planned, and no major structural modifications have been identified for the purposes of supporting license renewal. No additional land disturbance is anticipated.
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| We would appreciate your sending us a letter by March 15, 2005 detailing any concerns you may have about historic/archaeological resources in the area and/or a concluding statement that the operation of the Pilgrim Nuclear Power Station over the license renewal term would have no effect on any historic or archeological properties.
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| This will enable us to meet our application preparation schedule.
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| Entergy will include a copy of this letter and your response In the license renewal application that we submit to the NRC. Please call Fred Mogolesko at 508-830-7832 if you have any questions or require any additional information to review the proposed action.Sincerely, Director, Nuclear Assess nt Pilgrim Nuclear Power Station Entergy Nuclear Generation Company encl. Figure 2-3 Figure 2-2 Table Citation List c.;-~~ ~ ~1 _~- ~~--~7 -, 7 ~ ~ "- C.~~ -ll-Co c U'1 IF LEGEND m-s.-' StBoundary N NPSFGR .Toa s msl no m .NFIGURE 2na C0-4ThTssmfbn LWhes _ .Mt Bountdaty CZ: 02 0't KC Environmental Report for License Renewal lift ..n Jul 0 -) ki LEGEND// l PNPS FlIURE 2-2 P\/ orimary road I 6-/MI1 VicinIty with/ Secondary road L/Transmison Una Map ATransmission Una (Ti.) 1 n E Map S-I§,I Plymouth Red-Bellied
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| __ _ __ ___ _Turte Crltical Habitat_ 0. 8 O 1.1 2 2.5 3 Mba Urban Ro ==mmC=IfWd 0J 0 .50.5 1 1.5 2 25 3 3. 4 l 2-38 Town of Plymouth, Massachusetts Sites Usted In the National Register of Historic Places Site Name Location Bartlett-Russell-Hedge House 32 Court Street Bradford-Union Street Historic District Bradford, Union, Emerald, Water Cure, and Freedom Streets Clifford-Warren House East of Plymouth at 3 Clifford Road Cole's Hill Carver Street Harlow Old Fort House 119 Sandwich Street Sgt. William Hariow Family Homestead 8 Winter Street Hillside 230 Summer Street Jabez Howland House 33 Sandwich Street National Monument to the Forefathers Allerton Street Old County Courthouse Leyden and Market Streets Parting Ways Archaeological District Address Restricted Pilgrim Hall 75 Court Street Plymriouth Antiquarian House 126 Water Street Plymouth Post Office Building 5 Main Street Plymouth Rock Water Street Plymouth Village Historic District Roughly bounded by Water, Main, and Brewster Streets Richard Sparrow House 42 Summer Street Town Brook Historic and Archaeological Address Restricted District Source: U. S. Department of the Interior 2005.
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| CITATIONS IN SHPO CORRESPONDENCE AEC (U.S. Atomic Energy Commission).
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| 1974. Final Environmental Statement related to the Proposed Pilgrim Nuclear Power Station, Unit 2. Division of Radiological and Environmental Protection.
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| Washington DC.Massachusetts Historical Commission, 2003. State Register of Historic Places 2000.U.S. Census Bureau. 2000. "Census 2000 Redistricting Data (Public Law 94-171)Summary File." Available at htto:lfactfinder.census.aov/servlet/DTGeoSearch ByListServlet?ds name=DEC 2000 PL U&state=dt&
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| lanp=en. Accessed June 1, 2001.U.S. Department of the Interior.
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| 2005. Plymouth County, Massachusetts Listing of Sites on the National Register of Historic Places. Available at http://www/nr/nps/gov.
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| Accessed January 17, 2005.
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| The Commonwealth of Massachusetts William Francis Galvin, Secretary of the Commonwealth Massachusetts Historical Commission March 14, 2005 Stephen Bethay Director, Nuclear Assessment Pilgrim Nuclear Power Station Entergy Nuclear Generation Company RE: Pilgrim Nuclear Power Station License Renewal, Plymouth.
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| MHC #RC.36661
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| | |
| ==Dear Mr. Bethay:==
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| Thank you for submitting information to the Massachusetts Historical Commission regarding the proposed project referenced above. Staff of the MHC have reviewed the information you submitted and have the following comments.MHC understands from your letter that Entergy has no plans to alter current operations at the power station, to expand existing facilities, or to undertake ground-disturbing activities over the license renewal period.In addition to the five archaeological sites mentioned in your letter, review of MHC's Inventory of the Historic and' y Archaeological Assets of the Commonwealth indicates that there is one additional recorded archaeological site within the project area, which consists of the existing power station and transmission line corridor.
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| This site (MHC site#19-68). located within the transmission line corridor north of Rocky Hill Road, is associated with the Native American settlement of the Plymouth area. After review of MHC's files and the information you submitted, MHC staff have determined that the proposed license reneA al as currently described is unlikely to affect significant historic or archaeological resources.
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| Should plans change and if activities involving ground disturbance are contemplated, MHC requests the opportunity to review project plans in order to assess potential effects to historic and archaeological resources and to determine whether an archaeological survey is warranted for project impact areas.These comments are offered in compliance with Section 106 of the National Historic Preservation Act of 1966, as amended (36 CFR 800) and Massachusetts General Laws, Chapter 9, Sections 26-27C (950 CMR 71). If you have any questions concerning this review, please feel free to contact me at this office.Sincerely, Eric S. Ioh Archaeologist/Preservation Planner Massachusetts Historical Commission xc: Plymouth Historical Commission Cheryl Andrews-Maltais, THPO, WTGHA 220 Morrissey Boulevard, Boston, Massachusetts 02125 (617) 727-8470
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| * Fax: (617) 727-5128 www.state.ma.us/sec/mhc PlIgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Attachment D Coastal Zone Management Consistency Certification Federal Consistency Certification for Federal Permit and License Applicants' This is the Entergy Nuclear Generation Company (Entergy) certification to the U. S. Nuclear Regulatory Commission (NRC) that the renewal of the Pilgrim Nuclear Power Station (PNPS)operating license would be consistent with enforceable policies of the federally approved state coastal zone management program. The certification describes background requirements, the proposed action, (i.e., license renewal), anticipated environmental impacts, Massachusetts enforceable coastal resource protection policies and PNPS compliance status, and summary findings.CONSISTENCY CERTIFICATION Entergy certifies to the NRC that renewal of the PNPS operating license complies with the enforceable policies of Massachusetts' approved coastal zone management program and will be conducted in a manner consistent with such program. Entergy expects PNPS operations during the license renewal term to be a continuation of current operations as described below, with no station changes that would change effects on Massachusetts' coastal zone.NECESSARY DATA AND INFORMATION Statutory
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| ===Background===
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| The Federal Coastal Zone Management Act (16 USC 1451 et seq.) imposes requirements on an applicant for a Federal license to conduct an activity that could affect a state's coastal zone.The Act requires an applicant to certify to the licensing agency that the proposed action would be consistent with the state's federally approved coastal zone management program. The Act 4,,0V also requires the applicant to provide to the state a copy of the certification statement and requires the state, at the earliest practicable time, to notify the federal agency and the applicant whether the state concurs with, or objects to, the consistency certification.
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| See 16 USC 1456(c)(3)(A).
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| The National Oceanic and Atmospheric Administration (NOAA) has promulgated implementing regulations that indicate the certification requirement is applicable to renewal of federal licenses for activities not previously reviewed by the state [915 CFR 930.51 (b)(1)J. NOAA approved the Massachusetts coastal zone management program in 1978. In Massachusetts, the approved program is the Massachusetts Coastal Zone Management (MCZM) Program, Massachusetts General Laws (M.G.L.) Chapter 21A, Sections 2 and 4A, with regulations at 301 Code of Massachusetts Regulations (CMR) 20 -26.MCZM Program regulations require review of federal activities that are listed or that could reasonably be expected to affect the coastal zone (301 CMR 21.04). NRC licensing is a listed activity [301 CMR 21.07(2)(a)(6)]
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| and the PNPS location at the coastline and withdrawal from and discharge to coastal waters could reasonably be expected to affect the coastal zone. The State regulation requires certification of compliance with the MCZM Program policies[301 CMR 21 .07(3)(a)(1 )J and the regulation lists the policies (301 CMR 21.98). Attachment D-1 identifies the policies and the Entergy justification for certifying compliance.
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| @ This certification is patterned after the example certification included as Appendix E of Ref D-1.
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| Proposed Action Entergy is applying to the NRC for renewal of the PNPS license to operate for an additional 20 years beyond the current expiration date of June 8, 2012. Entergy expects PNPS operations during the license renewal term to be a continuation of current operations as described in the following paragraphs, with no changes that would affect the Massachusetts coastal zone.Entergy certifies that license renewal complies with the program policies of the Massachusetts approved coastal management program and will be conducted in a manner consistent with such policies.Background Information PNPS is located on the western shore of Cape Cod Bay in-the Town of Plymouth, Plymouth County, Massachusetts.
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| Approximately 60 percent of the area within a 50-mile radius is the open water of Cape Cod Bay and Plymouth Bay. Two transmission lines were built to connect PNPS to the electric grid. Both lines share a 300-foot wide transmission corridor that runs approximately 5 miles inland to the Jordan Road Tap. The inland boundary of the coastal zone is 100 feet inland of Route 3A, therefore, the area of interest includes the plant property and the transmission corridor to 100 feet west of the Route 3A crossing.
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| Figures 2-1 and 2-2 are PNPS 50-mile and 6-mile vicinity maps, respectively.
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| PNPS is a single-unit plant with a boiling water reactor and turbine generator licensed for an output of 2,028 megawatts-thermal (MWt), and an electric rating of 715 megawatts-electric (MWe) gross.PNPS is equipped with a once-through heat dissipation system that withdraws cooling water from and discharges to Cape Cod Bay. Two pumps in the intake structure provide a continuous supply (311,000 gallons per minute [gpm]) of condenser cooling water. Also housed in the intake structure are five service water pumps that supply 10,000 gpm cooling water, with four pumps in operation and one on standby, to the service water system. After moving through the condensers, cooling water is discharged into a 900-foot long discharge channel. At low tide the water in the discharge channel is several feet higher than sea level and the discharge is rapid and turbulent.
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| At high tide the velocity is much lower. The increase in water temperature across the condensers ranges from 27 to 30 0 F; the plant is permitted for as much as a 32 0 F temperature change. Entergy holds a National Pollutant Discharge Elimination System (NPDES) permit for this and other plantlstormwater discharges.
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| In accordance with permit requirements, Entergy monitors discharge characteristics and reports the results to the U.S.Environmental Protection Agency (EPA) and the Massachusetts Department of Environmental Protection.
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| The PNPS NPDES permit, issued August 30, 1994, by EPA Region I, constitutes the current CWA Section 316(b) determination for PNPS. Because Entergy submitted a timely application for renewal of the PNPS NPDES permit, the 1994 permit has been administratively continued.
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| PNPS has an onsite wastewater treatment plant. Sanitary wastewater that does not contain radioactive materials is processed in the wastewater treatment facility and discharged through a permitted drain field to the groundwater.
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| The treated wastewater discharge cannot exceed an average of 37,500 gallons per day.Entergy employs a permanent workforce of approximately 700 employees (including baseline permanent contractors) at PNPS. The majority of the PNPS workforce (approximately 83%)lives in Plymouth or Barnstable Counties.
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| PNPS is on a 24-month refueling cycle. During 2 refueling outages, site employment increases above the 700 person permanent workforce by as many as 700 to 900 workers for temporary (30 to 40 days) duty.Environmental Impacts The NRC has prepared a Generic Environmental Impact Statement (NRC 1996) on impacts that nuclear power plant license renewal could have on the environment and has codified its findings (10 CFR 51, Subpart A, Appendix B, Table B-1). The codification identified 92 potential environmental issues, 69 of which the NRC identified as having small impacts and termed"Category 1 issues." The NRC defines "small" as: Small -For the issue, environmental effects are not detectable or are so minor that they will neither destabilize nor noticeably alter any important attribute of the resource.
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| For the purpose of assessing radiological impacts, the Commission has concluded that those impacts that do not exceed permissible levels in the Commission's regulations are considered small as the term is used in this table (10 CFR 51, Subpart A, Appendix B, Table B-1).The NRC based its assessment of license renewal impacts on its evaluations of impacts from current plant operations.
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| The NRC codification and the Generic Environmental Impact Statement discuss the following types of Category 1 environmental issues:* Surface water quality, hydrology, and use* Aquatic ecology* Groundwater use and quality i
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| * Terrestrial resources* Air quality* Land use* Human health* Postulated accidents* Socioeconomics
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| * Uranium fuel cycle and waste management
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| * Decommissioning In its decision making for plant-specific license renewal applications, absent new and significant information to the contrary, the NRC relies on its codified findings, as amplified by supporting information in the Generic Environmental Impact Statement, for assessment of environmental impacts from Category 1 issues [10 CFR 51.95(c)(4)].
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| For plants such as PNPS that are located in the coastal zone, many of these issues involve potential impacts to the coastal zone.Entergy has adopted by reference the NRC findings and Generic Environmental Impact Statement analyses for all 492 applicable Category 1 issues.The NRC regulation identified 21 issues as "Category 2," for which license renewal applicants must submit additional site-specific information.
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| 3 Of these, 11 apply to PNPS 4 , and like the 2 The remaining Category I issues do not apply to PNPS either because they are associated with design or operational features that PNPS does not have (e.g., cooling towers) or to an activity, refurbishment, that PNPS will not undertake.
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| 3 10 CFR 51, Subpart A, Appendix B, Table B-1 also identifies 2 issues as 'NA' for which the NRC could not come to a condusion regarding categorization.
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| Entergy believes that these issues, chronic effects of electromagnetic fields and environmental justice, do not affect 'coastal zone' as that phrase is defined by the Coastal Zone Management Act 116 USC 1453(1)1.4 The remaining Category 2 issues do not apply to PNPS either because they are associated with design or operational features that PNPS does not have (e.g., cooling towers) or to an activity, refurbishment, that PNPS will not undertake.
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| 3 Category 1 issues, could potentially involve impacts to the coastal zone. The applicable issues and Entergy's impact conclusions are listed below.Aquatic ecology o Entrainment of fish and shellfish in early life stages -This issue addresses mortality of organisms small enough to pass through the plant's circulating cooling water system.Entergy and Boston Edison (former owner/operator of PNPS)have conducted studies of the impacts of entrainment under direction of the EPA and the Commonwealth and, in issuing the plant's discharge permit, EPA and the Commonwealth have approved the plant's intake structure as the best technology available to minimize impact. Entergy concludes that these impacts are small during current operations and has no plans that would change this conclusion for the license renewal term.o Impingement of fish and shellfish
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| -This issue addresses mortality of organisms large enough to be caught by intake screens before passing through the plant's circulating cooling water system. The studies and permit discussed above also address impingement.
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| Entergy concludes that these impacts are small during current operations and has no plans that would change this conclusion for the license renewal term.o Heat shock -This issue addresses mortality of aquatic organisms by exposure to heated plant effluent.
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| Entergy and Boston Edison (former owner/ operator of PNPS) have conducted studies of this issue under the direction of the EPA and the Commonwealth and, in issuing the plant's discharge permit, EPA and the Commonwealth have determined that more stringent limits on the heated effluent are not necessary to protect the aquatic environment.
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| Entergy concludes that these impacts are small during current operations and has no plans that would change this conclusion for the license renewal U term.* Threatened or endangered species This issue address effects that PNPS operations potentially could have on species that are listed under federal law as threatened or endangered.
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| In analyzing this issue, Entergy has also considered species that are listed under Commonwealth of Massachusetts law (Table D-1).Although several species of whales and sea turtles occur in Cape Cod Bay, none have ever been observed in the vicinity of the plant. Several other terrestrial species could occur on the PNPS site, or along associated transmission corridors, although none have been observed.Entergy's and NSTAR's (the company responsible for the transmission lines) environmental protection programs have identified no adverse impacts to such species and Entergy consultation with cognizant Federal and Commonwealth agencies has identified no impacts of concern. Entergy concludes that PNPS impacts to these species are small during current operations and has no plans that would change this conclusion for the license renewal term.* Human health Electromagnetic fields, acute effects (electric shock) -This issue addresses the potential for shock from induced currents, similar to static electricity effects, in the vicinity of transmission lines. Because this strictly human-health issue does not directly or indirectly affect natural resources of concern within the Coastal Zone Management Act definition of "coastal zone"[16 USC 1453(1)], Entergy concludes that the issue is not subject to the certification requirement.
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| Q 4
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| * Socioeconomics o Housing -This issue addresses impacts that PNPS employees required to support license renewal could have on local housing availability.
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| The NRC concluded, and Entergy concurs, that impacts would be small for plants located in high population areas with no growth control measures.
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| Using the NRC definitions and categorization methodology, PNPS is located in a high population area and locations where additional employees would probably live do not have growth control measures.
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| In addition, as Entergy does not intend to add additional permanent employees to the PNPS workforce, Entergy has concluded that impacts during the PNPS license renewal term would be small.o Public services:
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| public utilities
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| -This issue address impacts that adding license renewal workers could have on public water supply systems. Entergy has analyzed the availability of public water supplies in candidate locales and has found no limitations that would suggest that additional PNPS workers would cause impacts. As Entergy does not intend to add additional permanent employees to the PNPS workforce, Entergy has concluded that impacts during the PNPS license renewal term would be small.o Offsite land use -This issue addresses impacts that local government spending of plant property tax dollars can have on land use patterns.
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| PNPS property taxes comprise 2-3 percent of the Town of Plymouth's revenue and Entergy expects this to remain generally unchanged during the license renewal term. The NRC concluded, and Entergy concurs, that impacts to offsite land use would be small if tax payments are less than 10 percent of total revenue. Entergy concludes that impacts during the PNPS license renewal term would be small.o Public services:
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| transportation
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| -This issue addresses impacts that adding license renewal workers could have on local traffic patterns.
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| As Entergy does not intend to add additional employees to the permanent workforce for the license renewal term, this would result in small impacts.o Historic and archaeological resources
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| -This issue address impacts that license renewal activities could have on resources of historic or archaeological significance.
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| Although a number of archaeological or historic sites have been identified on or near the PNPS site or associated transmission lines, Entergy is not aware of any adverse or detrimental impacts to these sites from current operations and Entergy has no plans for license renewal activities that would disturb these resources.
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| Entergy correspondence with the Massachusetts Historical Commission, State Historic Preservation Officer identified no issues of concern.o Severe accidents
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| -Results from the Entergy severe accident mitigation alternatives (SAMA) analysis have not identified additional cost beneficial ways to further mitigate risk to public health and the economy in the area of the plant, including the coastal zone, due to potential severe accidents at PNPS. The SAMAs, however, are unrelated to aging management issues that are the subject of the license renewal analysis and, therefore are not related to the consistency certification for license renewal.5 State Program The Massachusetts Coastal Zone Management Program is administered by the Massachusetts Office of Coastal Zone Management within the Massachusetts Executive Office of Environmental Affairs. The office maintains a website that describes the program in general terms (Reference D-3). The Massachusetts Coastal Zone Management Program (Reference D-4) contains details about the state's enforceable policies and management principles.
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| Attachment D-1 lists these policies and management principles and discusses for each item the applicability to PNPS and, where applicable, the status of PNPS compliance.
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| Findings 1. The NRC has found that the environmental impacts of Category 1 issues are small. Entergy has adopted by reference NRC findings for Category 1 issues applicable to PNPS.2. For Category 2 issues applicable to PNPS, Entergy has determined that the environmental impacts are small.3. To the best of Entergy's knowledge, PNPS is in compliance with Massachusetts licensing and permitting requirements and is in compliance with its Commonwealth-issued licenses and permits.4. Entergy's license renewal and continued operation of PNPS would be consistent with the enforceable provisions of the Massachusetts Coastal Zone Management Program.STATE NOTIFICATION By this certification that PNPS license renewal is consistent with Massachusetts' coastal zone (management program, the Commonwealth of Massachusetts is notified that it has six months from receipt of this letter and accompanying information in which to concur with or object to Entergy's certification.
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| However, pursuant to 301 CMR 21.07(3)(e), if the Commonwealth of Massachusetts has not issued a decision within three months following the commencement of state agency review, it shall notify the contacts listed below of the status of the matter and the basis for further delay. The Commonwealth's concurrence, objection, or notification of review status shall be sent to: Robert Schaaf Stephen J. Bethay U.S. Nuclear Regulatory Commission Director, Nuclear Assessment One White Flint North Pilgrim Nuclear Power Station 11555 Rockville Pike 600 Rocky Hill Road Rockville, MD 020852-2738 Plymouth, MA 02360..Q 6 References ( ~ D-1. U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulations, LIC-203, Procedural Guidance for Preparing Environmental Assessments and Considering Environmental Issues, Revision 1, May 24, 2004.D-2. U. S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS), Volumes 1 and 2, Washington, DC, May 1996.D-3. Massachusetts Coastal Zone Management, "Massachusetts Coastal Zone Management," Boston, MA, 2001, available at http:llwww.state.ma.us/czm/czm.htm, accessed April 23, 2001.D-4. Code of Massachusetts Regulations, Chapter 301, Sections 20-26, Coastal Zone Management Program.D-5. U.S. Fish & Wildlife Service, Threatened and Endangered Species System (TESS);Listings by State and Territory as of 02/23/2005:
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| Massachusetts, February 23, 2005, Available at http://ecos.fws.gov/tess-public/TESSWebpageUsaLists?state=MA.
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| D-6. Massachusetts Division of Fisheries and Wildlife, "Rare Species by County: Plymouth," Boston, MA, March 1, 2003, Available at http://www.mass.gov/dfwele/dfw/nhesp/plym.htm, Accessed January 11, 2005.D-7. Massachusetts Division of Fisheries and Wildlife, "Massachusetts List of Endangered, Threatened and Special Concern Species," Boston, MA, June 18, 2004, Available at http://www.mass.gov/dfwele/dfw/nhesp/nhrare.htm, Accessed February 23, 2005.D-8. Massachusetts Division of Fisheries
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| & Wildlife, National Heritage & Endangered Species Program, BioMap and Living Waters: Guiding Land Conservation for Biodiversity in Massachusetts, Core Habitats of Plymouth, Westborough, MA, 2004.7 Table D-1 Massachusetts Coastal Zone Management Program's Program Policies and Management Principles The Massachusetts Coastal Zone Management Program is codified in the Massachusetts General Laws and the Code of Massachusetts Regulations and requires persons seeking approval for activities which may impact the Coastal Zone to demonstrate that the activity is consistent with all applicable policies in 301 CMR 21.98, Policy Appendix.
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| Entergy is seeking renewal of the operating license for PNPS. The following table details the policies and management principles of 301 CMR 21.98 and provides the Entergy demonstration that PNPS license renewal would be consistent with 301 CMR 21.98./POLICY l -JUSTIFICATION/
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| CONSISTENCY WATER QUALITY POLICIES WATE QUAITYPOLIY
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| #1 Enure hatPNPS operations are consistent with its point-source discharges in or affecting the NPS permtirequirement wich abs coastal zone are consistent with federally-NPDES permit requirements which are based approved state effluent limitations and water sndards.quality standards.
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| standards.
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| WATER QUALITY POLICY #2: Ensure that nonpoint pollution controls promote the PNPS's storm water runoff is covered by its attainment of state surface water quality NPDES permit.standards in the coastal zone.PNPS's activities conform to requirements WATER QUALITY POLICY #3: Ensure that set forth in its: activities in or affecting the coastal zone conform to applicable state requirements governing sub-
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| * Groundwater Permit surface waste discharges and sources of air and
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| * Air Quality Emissions Cap water pollution and protection of wetlands.
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| * NPDES Permit* Applicable Wetlands Order of Conditions HABITAT OLICIES HABITAT POLICY #1: Protect wetland areas including salt marches, shellfish beds, dunes, PNPS does maintain onsite man-made beaches, barrier beaches, salt ponds, eel grass freshwater wetlands areas.beds, and freshwater wetlands for their role as natural habitats.HABITAT POLICY #2: Promote the restoration of degraded or former wetland resources in coastal areas and ensure that activities in PNPS operations do not degrade wetlands in coastal areas do no further wetland degradation, the coastal areas.but instead take advantage of opportunities to engage in wetland restoration.
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| PROTECTED AREAS POLICIES PROTECTED AREAS POLICY #1: Assure PNPS is not located in an Area of Critical preservation, restoration, and enhancement of Environmental Concern.complexes of coastal resources of regional or (5,1 8 statewide significance through the Areas of Critical Environmental Concern (ACEC)Program.PROTECTED AREAS POLICY #2: Protect state and locally designated scenic rivers and state PNPS is not located on a river.classified scenic rivers in the coastal zone.Entergy is aware of no PNPS impacts on PROTECTED AREAS POLICY #3: Review designated or registered historic districts or proposed developments in or near designated or sites and license renewal will not alter this.registered historic districts or sites to ensure that Entergy has been in contact with the the preservation intent is respected by federal, Massachusetts Historical Commission which state, and private activities and that potential is in agreement that license renewal for adverse effects are minimized.
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| PNPS is unlikely to affect historic sites or districts.
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| COASTAL HAZARDS POLICIES COASTAL HAZARD POLICY #1: Preserve, protect, restore, and enhance the beneficial functions of storm damage prevention and flood control provided by natural coastal landforms, Entergy is aware of no PNPS impacts on suha.uebahebrirbahs these areas and of no reason for license such as dunes, beaches, barrer beaches, coastal banks, land subject to coastal storm renewal to alter this.flowage, salt marshes, and land under the ocean.COASTAL HAZARD POLICY #2: Ensure construction in water bodies and contiguous land areas will minimize interference with water circulation and sediment transport.
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| Approve PNPS license renewal will necessitate no permits for flood or erosion control projects only construction.
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| when it has been determined that there will be no significant adverse effects on the project site or adjacent or downcoast areas.COASTAL HAZARD POLICY #3: Ensure that state and federally funded public works projects proposed for location within the coastal zone will:* Not exacerbate existing hazards or damage natural buffers or other natural resources; PNPS is a privately owned facility and its* Be reasonably safe from flood and erosion ulicense renewal is not a state or federally related damage; funded public works project* Not promote growth and development in hazard-prone or buffer areas, especially in Velocity zones and ACECs; and* Not be used on Coastal Barrier Resource Units for new or substantial reconstruction of structures in a manner inconsistent with the 9 Coastal Barrier Resource/improvement Acts.COASTAL HAZARD POLICY #4: Prioritize public funds for acquisition of hazardous coastal areas for conservation or recreation use, and PNPS is a privately owned facility and is not relocation of structures out of coastal high involved in the spending/
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| prioritizing of public hazard areas, giving due consideration to the funds.effects of coastal hazards at the location to the use and the manageability of the area.PORT AND HARBOR INFRASTRUCTURE POLICIES PORTS POLICY #1: Ensure that dredging and disposal of dredged material minimize adverse PNPS is not a port or harbor infrastructure effects on water quality, physical processes, project.marine productivity, and public health.PORTS POLICY #2: Promote the widest possible public benefit from channel dredging, ensuring that designated ports and developed harbrs re ive hihestpririt inthePNPS Is not a port or harbor infrastructure harbors are given highest priority In the allocation of federal and state dredging funds. project.Ensure that this dredging is consistent with marine environmental policies.PORTS POLICY #3: Preserve and enhance the capacity of Designated Port Areas (DPAs) to accommodate water-dependent industrial uses, and prevent the exclusion of such uses from PNPS is not a port or harbor infrastructure tidelands and any other DPA lands over which a project.state agency exerts control by virtue of ownership, regulatory authority, or other legal jurisdiction.
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| PORTS AND HARBOR INFRASTRUCTURE MANAGEMENT PRINCIPLES PORTS MANAGEMENT PRINCIPLE
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| #1: Encourage, through technical and financial assistance, expansion of water dependent uses PNPS is not a port or harbor infrastructure in designated ports and developed harbors, re- project.development of urban waterfronts, and expansion of visual access.PUBLIC ACCESS MANAGEMENT PRINCIPLES PUBLIC ACCESS MANAGEMENT PRINCIPLE#1: Improve public access to coastal recreation facilities and alleviate auto traffic and parking Due to the heightened security environment, problems through improvements in public PNPS has closed it's shorefront area and transportation.
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| Link existing coastal recreation sites to each other or to nearby coastal inland nature trails to pubic access.facilities via trails for bicyclists, hikers, and equestrians, and via rivers for boaters.c.)10 PUBLIC ACCESS MANAGEMENT PRINCIPLE#2: Increase capability of existing recreation areas by facilitating multiple use and by improving management, maintenance, and Due to the heightened security environment, imprvin maagemntmaitenaceandPNPS has closed it's shorefront area and public support facilities.
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| Resolve conflicting uses nau has to ic ss.whenever possible through improved nature trails to public access management rather than through exclusion of uses.PNPS is a privately owned facility and is not PUBLIC ACCESS MANAGEMENT PRINCIPLE involved in external activities of shorefront
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| #3: Provide technical assistance to developers development.
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| In addition, due to the of private recreational facilities and sites that heightened security environment, PNPS has increase public access to the shoreline.
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| closed it's shorefront area and nature trails to public access.PUBLIC ACCESS MANAGEMENT PRINCIPLE#4: Expand existing recreation facilities and acquire and develop new public areas for coastal recreational activities.
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| Give highest priority to expansions or new acquisitions in PNPS is a privately owned facility and is not prioityto xpasios ornewacqisiion in involved in external activities of shorefront regions of high need or limited site availability.
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| development.
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| Assure that both transportation access and the recreational facilities are compatible with social and environmental characteristics of surrounding communities.
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| ENERGY POLICY ENERGY POLICY #1: For coastally dependent energy facilities, consider siting in alternative coastal locations.
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| For non-coastally dependent energy facilities, consider siting in areas outside PNPS is an existing facility.of the coastal zone. Weigh the environmental and safety impacts of locating proposed energy facilities at alternative sites.ENERGY MANAGEMENT PRINCIPLE ENERGY MANAGEMENT PRINCIPLE
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| #1: I Encourage energy conservation and the use of PNPS is a privately owned power generation alternative sources such as solar and wind facility that plays an important roler as a power in order to assist in meeting the energy generator and Inas a means for maintaining needs of the Commonwealth.
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| grid stabilityin Southeastern Massachusetts.
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| OCEAN RESOURCES POLICIES OCEAN RESOURCES POLICY #1: Support the Entergy is aware of no aquaculture near the development of environmentally sustainable site. Entergy is aware of no PNPS impacts aquaculture, both for commercial and on aquaculture and no reason for license enhancement (public shellfish stocking) renewal to alter this.purposes.
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| Ensure that the review process regulating aquaculture facility sites (and access PNPS sponsors a pilot phase Winter flounder routes to those areas) protects ecologically hatchery in Chatham, MA. PNPS has 11 significant resources (salt marshes, dunes, conducted post release survival studies beaches, barrier beaches, and salt ponds) and which indicate this is a viable restoration minimizes adverse impacts upon the coastal and technique.
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| marine environment.
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| OCEAN RESOURCES POLICY #2: Extraction of marine minerals will be considered in areas of state jurisdiction except where prohibited by the Massachusetts Ocean Sanctuaries Act, where PNPS operation and license renewal do not and when the protection of fisheries, air and involve extraction of marine minerals.marine water quality, marine resources, navigation, and recreation can be assured.OCEAN RESOURCES POLICY #3: Accommodate offshore sand and gravel mining needs in areas and in ways that will not adversely affect shoreline areas due to alteration PNPS operations and license renewal do not of wave direction and dynamics, marine involve sand or gravel mining.resources, and navigation.
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| Mining of sand and gravel, when and where permitted, will be, primarily for the purpose of beach nourishment.
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| GROWTH MANAGEMENT PRINCIPLES GROWTH MANAGEMENT PRINCIPLE
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| #1: Encourage, through technical assistance and PNPS is a privately owned facility and review of publicly funded development, renewal of its operating license is not a state compatibility of proposed development with local or federally funded public works project community character and scenic resources.
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| GROWTH MANAGEMENT PRINCIPLE
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| #2: Ensure that state and federally funded transportation and wastewater projects primarily PNPS is a privately owned facility and e g d e a a renewal of its operating license is not a state serve existing developed areas, assigning highest priority to projects that meet the needs or federally funded public works project of urban and community development centers.GROWTH MANAGEMENT PRINCIPLE
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| #3: Encourage the revitalization and enhancement P i a of existing development centers in the coastal e t h t renewal of its operating license is not a state zone through technical assistance and federal .- .and state financial support for residential, or federally funded pubic works project commercial, and industrial development.
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| 12 Table D-2 Endangered and Threatened Species that Occur in the Vicinity of PNPS or in Plymouth County, MA I Federal State Scientific Name I Common Name l Status" l Statusa Mammals Balaenoptera borealis Sei whale E E Balaena glacialis Right Whale E E Balaenoptera musculus Blue Whale E E Balaenoptera physalus Finback Whale E E Megaptera novaeangliae Humpback Whale E E Birds Ammodramus savannarum Grasshopper Sparrow T Bartramia longicauda Upland Sandpiper E Botaurus lentiginosus American Bittern E Charadrius meloduse Piping Plover T T Circus cyaneus Northern Harrier T Haliaeetus leucocephalus Bald Eagle T E Ixobrychus exilisb Least Bittern E Parula americana Northern Parula T Podilymbus podiceps Pied-Billed Grebe E Rallus elegans King Rail T Stema dougallii dougallifD Roseate Tern E E Reptiles : Caretta careffa Loggerhead Sea Turtle T T.Chelonia mydas Green Sea Turtle T T Dermochelys coriacea Leatherback Sea Turtle E E Emydoidea blandingii Blanding's Turtle T Eretmochelys imbricata Hawksbill Sea Turtle E E Lepidochelys kempHi Kemp's Ridley Sea Turtle E E Malaclemys terrapin Diamondback Terrapin T Pseudemys rubriventrist Northern Red-Bellied Cooter E E Amphibians
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| ..Ambystoma opacum [Marbled Salamander T Scaphiopus holbrookii Eastem Spadefoot Toad -T Invertebrates Acronicta albarufa Barrens Daggermoth T Alasmidonta heterodon Dwarf Wedgemussel E E Cicinnus melsheimeri Melsheimer's Sack Bearer T Cycnia inopinatus Unexpected Cycnia T Enallagma recurvaturm" Pine Barrens Bluet T Erynnis persius persius Persius Duskywing E Hypomecis buchholzaria Buchholz's Gray -E Lampsilis cariosa Yellow Lampmussel
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| .E Metarranthis apiciaria Barrens Metarranthis Moth E Nicrophorus americanus American Burying Beetle E -Papaipema appassionata Pitcher Plant Borer Moth T Papaipema stenocelis Chain Fern Borer Moth T 13 Federal State Scientific Name Common Name Statusa Status'Papaipema sulphuratab Water-Willow Stem Borer T Somatochlora kennedyi Kennedy's Emerald E Zanclognatha martha Pine Barrens Zanclognatha T Vascular plants Agalinis acuta Sandplain Gerardia E Aristida purpurascens Purple Needlegrass T Asclepias verticillata Linear-Leaved Milkweed T Bidens hyperborean var. Estuary Beggarticks E hyperborea.
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| Calamagrostis pickeringii Reed Bentgrass E Cardamine longHi Long's Bittercress E Carex polymorpha Variable Sedge E Carex striata var. brevis Walter's Sedge E Crassula aquatica Pygmyweed T Cyperus houghtonfi Houghton's Flatsedge E Dichanthelium Mattamuskeet Panic-Grass E mattamuskeetense Elatine americana American Waterwort
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| -E Eriocaulon parked Estuary Pipewort E Eupatonum aromaticum Lesser Snakeroot E Eupatorium leucolepis var. New England Boneset E novae-angliae Isoetes acadiensis Acadian Quillwort E Isotria medeoloides Small Whorled Pogonia T Linum medium var. texanum Rigid Flax -T Lipocarpha micrantha Dwarf Bulrush _ T Ludwigia sphaerocarpa Round-Fruited False- -E Loosestrife
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| _Lycopus rubellus Gypsywort
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| -E Mertensia maritima Oysterleaf
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| -E Ophioglossum pusillum Northen adder's-tongue
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| -T Panicum rigidulum var. Long-Leaved Panic-Grass
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| -T Pubescens Platanthera flava var. herbiola Pale Green Orchid -T Polygonum setaceum var. Strigose Knotweed -T interjectum Prenanthes serpentaria Lion's Foot -E Ranunculus micranthus Tiny-Flowered Buttercup
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| -E Ranunculus pensylvanicus Bristly Buttercup
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| -T Rhynchospora inundatab Inundated Horned-Sedge -T Rhynchospora nitensb Short-Beaked Bald-Sedge
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| -T Rhynchospora torreyanab Torrey's Beak-Sedge
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| -E Rumex pallidus Seabeach Dock T Sabatia campanulata Slender Marsh Pink -E Sagittaria subulata var. subulata River Arrowhead
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| -E Sanicula canadensis Canadian Sanicle -T Scirpus longfi Long's Bulrush -T C.14 Federal State Scientific Name Common Name Statusa Statusa Senna hebecarpa Wild Senna E Spartina cynosuroides Salt Reedgrass
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| _ _T Sphenopholis pensylvanica Swamp Oats _ _T Symphyotrichum concolor Eastern Silvery Aster -E Triosteum perfoliatum Broad Tinker's Weed -E Viola briffoniana Britton's Violet T a. E = Endangered; T = Threatened;
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| -= Not listed.b. Species reported by the Massachusetts NHESP as occurring within six miles of PNPS.Source: References D-6, D-7, and D-8 15
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| .J Table D-3 Environmental Authorizations for Current PNPS Operations Issue or Activity Agency Authority Requirement Number Expiration Date Covered Federal Requirements to License Renewal________
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| U.S. Nuclear Atomic Energy Act License to Operate DPR -35 Issued 09/15/72 Operation of Regulatory (42 USC 2011, et Expires 06/08/12 Unit 1 Commission seq.), 10 CFR 50.10 U.S. Nuclear Atomic Energy Act Material License 20-07626-04 Issued 02/10/03 Contamination Regulatory Section 161, Expires 02/28/2013 on reactor Commission (42 USC 2201), components 10 CFR 40 and 70 U.S. Department 49 CFR 107, Subpart Registration 062601551001J Issued 05/16/05 Radioactive and of Transportation G Expires 06/30/06 hazardous This permit is renewed materials on an annual basis. shipments U.S. Clean Water Act NPDES Permit MA0003557 Issued 04/29/91 Plant discharges Environmental (33 USC 1251 et Modified 08/30/94 to Cape Cod Bay Protection Agency seq.), MGL c21, Expired 04/29/96 and Section 43(2) (remains in effect Massachusetts pending EPA and Department of Commonwealth action Environmental on renewal applications Protection submitted 10/25/95 and 12/01/99)
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| -U.S. Fish and Migratory Bird Treaty Depredation Permit MB831184-0 Issued 07/08/05 Removal of birds Wildlife Service Act, 16 USC 703-712 Expires 06/30/06 and nests from utility structures This permit is renewed on an annual basis.16 C i: Issue or Activity Agency l Authority
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| --Requirement Number Expiration Date Covered_State Requirements to License Renewal Massachusetts MGL c11, Section 5N Material License 07-6262 Issued 4/22/03 Contamination Department of Expires 4/30/08 on reactor Public Health components Massachusetts MGL cli1, Section 5N Material License 49-0078 Issued 10/11/02 Contamination Department of Expires 5/31/06 on reactor Public Health components Massachusetts MGL c148, Section 13 Registration Not applicable Expires 04/01/2006 Storing Department of flammable Public Safety This registration is materials in renewed on an annual tanks basis.Massachusetts MGL c21, Sections Groundwater
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| #2-389 Issued 04/20/99 Treated effluent Department of 26-53 Discharge Permit Expires 4/20/04 discharges to Environmental Renewal Application groundwater Protection submitted 10/14/03.
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| from wastewater treatment facility Administratively continued pending review of application Massachusetts 310 CMR 7.02(15) 50% Facility Issued 07/18/2005 Emissions from Department of 310 CMR 7.02(15)(e)
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| Emission Cap various small Environmental combustion Protection sources Massachusetts MGL c21C Large Quantity MAR000014167 Issued 10/06/99 Hazardous Department of 310 CMR 30 Generator waste generation Environmental Protection 17 9 3)Issue or Activity Agency Authority Requirement Number Expiration Date X Covered S tate Requirements to License Renewal South Carolina South Carolina Radioactive Waste 0007-20-01 Issued 12/17/04 Transportation of Department of Radioactive Waste Transport Permit Expires 12/31/05 radioactive Health and Transportation and waste to Environmental Disposal Act This permit is renewed disposal facility Control (SC ST SEC 13-7-110 on an annual basis. -in South et seq.) Carolina Tennessee TCA 68-202-206 Radioactive Waste T-MA004-LO1 Issued 12/08/04 Shipment of Department of License-for-Expires 12/31/05 radioactive Environment and Delivery waste to Conservation This permit is renewed disposal/on an annual basis. processing facility in Tennessee CFR -Code of Federal Regulations USC -United States Code MGL -Massachusetts General Laws CMR -Code of Massachusetts Regulations TCA -Tennessee Code Annotated 18 Table D-4 Environmental Authorizations for Pilgrim Nuclear Power Station License Renewal Agency Authority Requirement Activity Covered U.S. Nuclear Atomic Energy Act (42 License Environmental Report Regulatory USC 2011 et seq.) Renewal submitted in support of Commission license renewal application U.S. Fish and Endangered Species Consultation Requires Federal agency Wildlife Service and Act Section 7 issuing a license to consult National Marine with FWS and NMFS.Fisheries Service Commonwealth of Endangered Species Consultation Requires Federal agency Massachusetts Act Section 7 issuing a license to consult Division of Fisheries with FWS at the state and Wildlife level.Massachusetts Clean Water Act Certification Requires Commonwealth Department of Section 401 certification that discharge Environmental would comply with CWA Protection standards Massachusetts National Historic Consultation Requires Federal agency Historical Preservation Act issuing a license to Commission Section 106 consider cultural impacts and consult with the SHPO.Massachusetts Federal Coastal Zone Certification Requires an applicant to Coastal Zone Management Act (16 provide certification to the Management USC 1451 et seq.) federal agency issuing the Program license that the license renewal would be consistent with the federally-approved state coastal zone management program. Based on its review of the proposed activity, the state must concur with or object to the applicant's certification.
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| 19 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Attachment E Severe Accident Mitigation Alternatives Analysis Attachment E contains the following sections.E.1 -Evaluation of PSA Model E.2 -Evaluation of SAMA Candidates Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table of Contents E.1 EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODEL .... ..... E.1-1 E.1.1 PSA Model -Level 1 Analysis .....................................
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| E.1-1 E.1.2 PSA Model -Level 2 Analysis .....................................
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| E.1-27 E.1.2.1 Containment Performance Analysis .............................
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| E.1-27 E.1.2.2 Radionuclide Analysis ........................
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| E.1-33 E. 1.2.2.1 Introduction
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| ....................................
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| E. 1-33 E.1.2.2.2 Timing of Release ........................
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| E.1-33 E.1.2.2.3 Magnitude of Release ......................
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| E.1-34 E.1.2.2.4 Release Category Bin Assignments
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| ....... .................
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| E.1-34 E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories
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| .E.1-35 E.1.2.2.6 Collapsed Accident Progression Bins Source Terms .... ....... E.1-43 E.1.2.2.7 Release Magnitude Calculations
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| ..........................
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| E.1-52 E.1.3 IPEEE Analysis ..........................
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| E.1-52 E.1.3.1 Seismic Analysis ..........................
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| E.1-52 E.1.3.2 Fire Analysis .......................
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| ....E.1-52 E.1.3.3 Other External Hazards ......................
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| E.1-54 E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE .. .....................
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| E.1-54 E.1.4.1 PSA Model Peer Review ........................
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| ....E.1 -54 E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model .................
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| E.1-55 E.1.4.2.1 Core Damage -Comparison to the PNPS 1995 Update IPE Model .........................
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| E.1-55 E.1.4.2.2 Containment Performance
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| -Comparison to the Original PNPS IPE Model ..........
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| E.1-59 E.1.5 The MACCS2 Model -Level 3 Analysis ..............................
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| E.1-60 E.1.5.1 Introduction
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| .................
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| I ........ E.1-60 E.1.5.2 Input ................
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| E.1-60 E.1.5.2.1 Projected Total Population by Spatial Element ..... ..........
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| E.1-61 E.1.5.2.2 Land Fraction ....................
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| ........ E.1-62 E.1.5.2.3 Watershed Class ..... E.1-62 i i Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.5.2.4 Regional Economic Data -. .............
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| E.1.5.2.5 Agriculture Data ....................
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| E.1.5.2.6 Meteorological Data ................
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| E.1.5.2.7 Emergency Response'Assumptions.
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| E.1.5.2.8 Core Inventory
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| .....................
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| E.1.5.2.9 Source Terms ........................
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| E.1.5.3 Results ................................
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| E.1.6 References
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| ...............................
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| ................... .E.1-62................... .E.1-63................... .E.1-63...................
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| E.1-64...................
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| 1-64...................
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| 'E.1-66...................
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| E.1-66................
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| E.1-69 E.2 EVALUATION OF SAMA CANDIDATES
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| ..........
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| ....................
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| E.2-1 E.2.1 SAMA List Compilation
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| ..........................................
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| E.2-1 E.2.2 Qualitative Screening of SAMA Candidates (Phase I) ...... .............
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| E.2-2-E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II) E.2-2 E.2.4 Sensitivity Analyses ............................................
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| E.2-11 E.2.5 References
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| .................
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| .........................
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| E.2-13 ii Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Ij List of Tables Table E.1-1 Core Damage Frequency Uncertainty
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| .......................................
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| E.1-2 Table E.1-2 PNPS PSA Model CDF Results by Major Initiators
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| ........ E.1-3 Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs ................
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| E.1-4 Table E.1-4 Summary of PNPS PSA Core Damage Accident Class .........................
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| E.1-28 Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description
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| .............
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| E.1-29 Table E.1-7 PNPS Release Categories
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| .....................
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| E.1-35 Table E.1-6 Release Severity and Timing Classification Scheme Summary ...................
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| E.1-35 Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States ....... E.1-36 Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions
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| ......................
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| E.1 -44 Table E.1-10 Summary of PNPS Containment Event Tree Quantification
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| ......................
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| E.1-49 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms .....................
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| E.1-50 Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)
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| ............................................................
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| E. 1-51 Table E.1-12 PNPS Fire Updated Core Damage Frequency Results .........................
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| E.1-53 Table E.1-13 Estimated Population Distribution within a 50-mile Radius .......................
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| E.1-61 iii &:
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels)
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| ........................................
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| E.1-65 Table E.1-15 Base Case Mean PDR and OECR Values ...................................
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| E.1-67 Table E.1-16 Summary of Offsite Consequence Sensitivity Results ..........................
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| E.1-68 Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation
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| .....E.2-15 Table E.2-2 Sensitivity Analysis Results..............................................E.245 iv Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage List of Figures Figure E.1-1 PNPS Radionuclide Release Category Summary .........................
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| E.1-31 Figure E.1-2 PNPS Plant Damage State Contribution to LERF ...........
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| ..............
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| E.1-32 v0 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.1 EVALUATION OF PSA MODEL Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1 EVALUATION OF PROBABILISTIC SAFETY ANALYSIS MODEL The severe accident risk was estimated using the Probabilistic Safety Analysis (PSA) model and a Level 3 model developed using the MACCS2 code. The CAFTA code was used to develop the Pilgrim Nuclear Power Station (PNPS) PSA Level I and Level 2 models. This section provides the description of PNPS PSA Levels 1, 2, and 3 analyses, Core Damage Frequency (CDF)uncertainty, Individual Plant Examination of External Events (IPEEE) analyses, and PSA model peer review.E.1.1 PSA Model -Level I Analysis The PSA model (Level I and Level 2) used for the SAMA analysis was the most recent internal events risk model for PNPS (Revision 1, April 2003) [Reference E.1-1]. The PNPS PSA model and documentation has been updated to reflect the current plant operating configuration and design changes as of September 2001. The current PSA model reflects the accumulation of additional plant operating history and component failure and unavailability data as of December 2001. The PSA model also resolves all findings and observations during the industry peer review of the model, conducted in March 2000 [Reference E.1-1]. The PNPS model adopts the small event tree/ large fault tree approach and uses the CAFTA code for quantifying CDF. The Level I and Level 2 PNPS PSA analyses were originally developed and submitted to the NRC in September 1992 as the Pilgrim Nuclear Power Station Individual Plant Examination (IPE)Submittal
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| [Reference E.1-2].The PSA model has been updated since the IPE due to the following.
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| * In 1995, the original IPE model was changed in response to the NRC Request for Additional Information (RAI) received in April 1995 [Reference E.1-3]. Overall CDF was reduced from 5.85E-5/yr to 2.84E-5/yr.
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| The reduction in CDF was due to removal of HPCI room cooling dependency, revised ADS success criteria, and improved HPCI/RCIC performance.
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| * Equipment performance
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| -As data collection progresses, estimated failure rates and system unavailability data change.* Plant configuration changes -Plant configuration changes are incorporated into the PSA model.* Modeling changes -The PSA model is refined to incorporate the latest state of knowledge and recommendations from internal and industry peer reviews.The PSA model contains the major initiators leading to core damage with baseline CDFs listed in Table E.1-2 [Reference E.1-1].The current PNPS PSA model was reviewed to identify those potential risk contributors that made a significant contribution to CDF. CDF-based Risk Reduction Worth (RRW) rankings were reviewed down to 1.005. Events below this point would influence the CDF by less than 0.5% and E.1-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage are judged to be highly unlikely contributors for the identification of cost-beneficial enhancements.
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| These basic events, including component failures, operator actions, and initiating events, were reviewed to determine if additional SAMA actions may need to be considered.
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| Table E.1-3 provides a correlation between the Level 1 RRW risk significant events (component failures, operator actions, and initiating events) down to 1.005 identified from the PNPS PSA model and the SAMAs evaluated in Section E.2.The uncertainty associated with CDF was estimated using Monte Carlo techniques implemented in CAFTA for the base case mode. The results are shown in Table E.1-1.Table E.1-1 Core Damage Frequency Uncertainty Confidence CDF (IRY)Mean value 6.68E-6 5 th percentile.30E-6 5 0 th percentile 5.93E-6 v 9 5 th percentile 1.08E-5 The values in Table E.1-1 reflect the uncertainties associated with the data distributions used in the analysis.
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| The ratio of the 9 5 th percentile to the mean is about 1.62. This uncertainty factor is included in the factor of 6 used to determine the upper bound estimated benefit described in Appendix E, Section 4.21.5.4.E.1-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-2 PNPS PSA Model CDF Results by Major Initiators HE Type IE Description CDF Percentage of TDC Loss of DC Power Buses 3.06E-06 47.77%LOOP Loss of Offsite Power 1.29E-06 20.12%TAC Loss of AC Power Buses 8.83E-07 13.78%LSSW Loss of Salt Service Water 3.91E-07 6.10%TRAN Transients 3.60E-07 5.62%LOCA Loss of Coolant Accident 1.75E-07 2.73%SBO Station Blackout 1.46E-07 2.28%ATWS Anticipated Transient Without Scram 5.30E-08 0.83%ISLOCA Interfacing System LOCA 3.64E-08 0.57%FLOOD Internal Flooding 1.28E-08 0.20%Total 6.41E-06 100.00%E.1-3 AC" Pilgrim Nuclear Power Station Applicant's Environmental Report OperatinQ License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition IE-T1 6.70E-02 1.337 Loss of offsite This term represents the LOOP initiating event. Industry efforts power (LOOP) over the last twenty years have led to a significant reduction in plant scrams from all causes. Improvements related to enhancing offsite power availability or reliability and coping with SBO events were already implemented and evaluated during Phase I SAMA screening.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| IE-TDCB 2.63E-03 1.319 Transient caused This term represents an initiating event caused by a complete by loss of 125VDC loss of l25VDC buses D-17, D5, and D37 and random failures of bus B battery D-2. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed.
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| Phase II SAMAs 025,026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.
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| IE-TDCA 2.63E-03 1.304 Transient caused This term represents an initiating event caused by a complete by loss of 125VDC loss of 125VDC buses D-16, D4, and D36, and random failures of bus A battery D-1. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed.
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| Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.
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| E.1-4 3J J J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition FXT-XHE-FO-V4T2 2.31 E-02 1.121 Operator fails to This term represents operator failure to align fire water via the align fire water LPCI injection path for alternate RPV vessel injection.
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| Phase I crosstie for reactor SAMAs, including improvement of procedures and installation of pressure vessel instrumentation to enhance the likelihood of success of operator (RPV) injection via action in response to accident conditions, have already been LPCI (transient) implemented.
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| Phase II SAMAs 057 and 059, which recommend proceduralizing use of the diesel fire pump hydroturbine following EDG A failure, and providing a redundant path from fire water pump discharge to LPCI loops A and B cross-tie, were evaluated.
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| AC2-PHN-PE-23kV 5.OOE-01 1.079 Loss of shutdown This term represents loss of the shutdown transformer 23kV feed transformer 23kV to 4.16kV bus A8. Improvements related to enhancing offsite feed power availability or reliability and coping with SBO events were already implemented and evaluated during Phase I SAMA screening.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| IE-TSSW 6.85E-05 1.065 Loss of salt service This term represents an initiating event caused by a complete water (SSW) loss of the service water system. Phase I SAMAs were system implemented to improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| E. 1-5
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| -Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition IE-TAC6 2.63E-03 1.059 Transient caused This term represents loss of 4.16kV bus A6. Phase I SAMAs to by loss of 4160VAC improve 4.16kV bus cross-tie capability and revise procedures to bus A6 repair or replace failed 4.16kV breakers have already been implemented.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| CIV-XHE-FO-DTV 3.01 E-03 1.057 Operator fails to This term represents operator failure to recognize the need to vent containment vent the torus for pressure reduction during loss of containment using direct torus heat removal accident sequences.
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| Phase I SAMAs, including vent (DTV) improvement of procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| Phase II SAMA 053 to control containment venting within a narrow pressure band to prevent rapid containment depressurization during venting was evaluated.
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| IE-TAC5 2.63E-03 1.052 Transient caused This term represents an initiating event caused by loss of 4.16kV by loss of 4160VAC bus AS. Phase I SAMAs to improve 4.16kV bus cross-tie bus AS capability and revise procedures to repair or replace failed 4.16kV breakers have already been implemented.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| E.1-6 9 I J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition RHR-MAI-MA-HTXAP 4.08E-04 1.051 RHR heat This term represents RHR heat exchanger E-207A unavailable exchanger E-207A due to maintenance, leading to loop A RHR suppression pool unavailable due to cooling and drywell spray modes being unavailable for maintenance containment pressure reduction.
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| Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.
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| RBC-MAI-MA-LOOPA 3.71 E-04 1.046 RBCCW loop A out This term represents RBCCW loop A unavailable due to for maintenance maintenance.
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| A Phase I SAMA was implemented to improve RBCCW system reliability by making component cooling water trains separate.
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| Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| FXT-XHE-FO-DWS 2.21 E-02 1.046 Operator fails to This term represents operator failure to align fire water via the align fire water LPCI injection path for alternate drywell spray to remove cross-tie for drywell containment heat. Phase I SAMAs, including improvement of spray procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| Phase II SAMAs 057 and 059, which recommend proceduralizing use of the diesel fire pump hydroturbine following EDG A failure, and providing a redundant path from fire water pump discharge to LPCI loops A and B cross-tie, were evaluated.
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| E.1-7 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC8-CBR-CO-204 9.50E-05 1.044 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-204 fails to 204, leading to loss of power to 480V motor control center (MCC)remain closed B14 and its associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.
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| AC8-CBR-CO-103 9.50E-05 1.044 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-103 fails to 103, leading to loss of power to 480V MCC B15 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.
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| FXT-ENG-FR-P140 1.92E-02 1.043 Diesel fire pump P- This term represents diesel fire pump P-140 failure to run. Phase 140 fails to run 11 SAMA 045, to add a diverse injection system and provide an injection source other than fire water, was evaluated.
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| LCI-HTX-VF-E207A 3.24E-04 1.04 Loop B heat This term represents random failure of RHR heat exchanger E-exchanger E-207A 207A, leading to loop A RHR suppression pool cooling and failure drywell spray modes being unavailable for containment pressure reduction.
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| Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.
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| E.1-8 3 3 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition LCI-HTX-VF-E207B 3.24E-04 1.039 Loop A heat This term represents random failure of RHR heat exchanger E-exchanger E-207B 207B, leading to loop B RHR suppression pool cooling and failure drywell spray modes being unavailable for containment pressure reduction.
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| Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.
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| IE-T2 8.90E-02 1.038 Loss of PCS This term represents an initiating event caused by a transient with transients PCS unavailable.
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| Industry efforts over the last twenty years have led to a significant reduction of plant scrams from all causes.Phase II SAMA 038, to improve MSIV design and mitigate the consequences of this event, was evaluated.
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| RHR-MAI-MA-HTXBP 2.69E-04 1.032 RHR heat This term represents RHR heat exchanger E-207B unavailable exchanger E-207B due to maintenance, leading to loop B RHR suppression pool unavailable due to cooling and drywell spray modes being unavailable for maintenance containment pressure reduction.
| |
| Phase I SAMAs have already been implemented to use firewater for drywell spray and to use venting via DTV path to reduce containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated.
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| E.1-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition RBC-MAI-MA-LOOPB 2.36E-04 1.029 RBCCW loop B out This term represents RBCCW loop B unavailable due to for maintenance maintenance.
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| A Phase I SAMA was implemented to improve RBCCW system reliability by making component cooling water trains separate.
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| Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| DWS-XHE-FO-W2 2.85E-04 1.026 Operator fails to This term represents operator failure to align the drywell spray align drywell spray mode of RHR for containment pressure reduction.
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| Phase I mode of RHR SAMAs, including improvement of procedures and installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase 1I SAMAs were recommended for this subject.SPC-XHE-FO-WI 1.54E-04 1.026 Operator fails to This term represents operator failure to align the suppression align suppression pool cooling mode of RHR for containment pressure reduction.
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| pool cooling mode Phase I SAMAs, including improvement of procedures and of RHR installation of instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase II SAMAs were recommended for this subject.LCS-CCF-PG-STNRS 2.22E-05 1.024 Common cause This term represents common cause failure of the core spray and failure of strainers RHR suction strainers.
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| A Phase I SAMA, installing improved BS-8002A&B passive emergency core cooling system (ECCS) suction plugged strainers, has been implemented.
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| Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.
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| E.1-10 3 3 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition DC1-CBR-CO-7216A 5.11E-05 1.023 125VDC circuit This term represents random failure of 125VDC circuit breaker breaker 72-16A 72-16A, leading to loss of DC power to bus D-16. Phase I fails to remain SAMAs to improve battery charging capability and replace closed existing batteries with more reliable ones have already been installed.
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| Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.
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| ADS-XHE-FO-XlT2 6.88E-04 1.023 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during transients.
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| Phase I SAMAs, including depressurization improvement of procedures and installation of instrumentation to (transient) enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase II SAMAs were recommended for this subject.DCl-CBR-CO-72165 5.11E-05 1.023 125VDC circuit This term represents random failure of DC circuit breaker 72-165 breaker 72-165 fails to provide power to DTV valve AO 5025, causing failure of the to remain closed valve to open on demand, resulting in loss of containment venting capability.
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| Phase II SAMA 056 to improve DTV valve availability was evaluated.
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| OSP-SBO 7.64E-02 1.023 Operator fails to This term represents operator failure to start or align the SBO start or align station diesel to either bus A5 or A6 during a LOOP event. Phase I blackout (SBO) SAMAs, including improvement of SBO procedures and training diesel to either bus to enhance the likelihood of success of operator action in AS or A6 response to accident conditions, have already been implemented.
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| No additional Phase II SAMAs were recommended for this subject.E.1-11 I r?1.11 001--tll--Ago", I TL -Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition DCI-CBR-CO-7217A 5.11E-05 1.023 125VDC circuit This term represents random failure of 125VDC circuit breaker breaker 72-17A 72-17A, leading to loss of DC power to bus D-17. Phase I fails to remain SAMAs to improve battery charging capability and replace closed existing batteries with more reliable ones have already been installed.
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| Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.
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| OSP-14 4.10E-02 1.022 Failure to recover This term represents operator failure to recover offsite power offsite power within within 14 hours during a LOOP event. Phase I SAMAs, including 14 hours improvement of SBO procedures and training to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase 11 SAMAs were recommended for this subject.IE-T3A 8.60E-01 1.022 Transients with This term represents an initiating event caused by a transient with condenser initially PCS available.
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| Industry efforts over the last twenty years have available led to a significant reduction of plant scrams from all causes.Phase II SAMA 038 to improve MSIV design and mitigate the consequences of this event was evaluated.
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| FXT-MAI-MA-P140 9.22E-03 1.019 Diesel driven fire This term represents diesel fire pump P-140 in maintenance.
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| water pump P-140 Phase II SAMA 045, to add a diverse injection system and unavailable due to provide an injection source other than fire water, was evaluated.
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| maintenance E.1-12 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-604 2.51 E-03 1.019 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-604 breaker 152-604, leading to LOOP to safety bus A6. Phase I control circuit no SAMAs to improve 4.16kV bus cross-tie capability and revise output procedure to repair or replace failed 4.16kV breakers have already been installed.
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| In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| DC1-CBR-CO-72175 5.11E-05 1.018 125VDC circuit This term represents random failure of DC circuit breaker 72-175 breaker 72-175 fails to provide power to DTV valve AO 5042B, causing failure of the to remain closed valve to open on demand, resulting in loss of containment venting capability.
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| Phase II SAMA 056 to improve DTV valve availability was evaluated.
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| CIV-RCK-NO-5042B 2.50E-03 1.018 SV 5042B control This term represents random failure of the control circuit of DTV circuit failure valve AO 5042B, causing failure of the valve to open on demand, resulting in loss of containment venting capability to control containment pressure.
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| Phase II SAMA 056 to improve DTV valve availability was evaluated.
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| CIV-RCK-NO-A5025 2.50E-03 1.018 AO 5025 control This term represents random failure of the control circuit of DTV circuit failure valve AO 5025, causing failure of the valve to open on demand, resulting in loss of containment venting capability to control containment pressure.
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| Phase II SAMA 056 to improve DTV valve availability was evaluated.
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| E.1-13
| |
| *10----"" -.A#V1 IL, z Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-504 2.51 E-03 1.017 4.16kV circuit This temi represents failure of the control circuit of 4.16kV circuit breaker 152-504 breaker 152-504, leading to LOOP to safety bus A5. Phase I control circuit no SAMAs to improve 4.16kV bus cross-tie capability and revise output procedures to repair or replace failed 4.16kV breakers have already been installed.
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| In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| SSW-MDP-FS-P208D 2.022-03 1.017 SSW pump P-208D This term represents random failure of SSW pump P-208D to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| SSW-CCF-FS-3P208 2.26E-05 1.017 Common cause This term represents common cause failure of 3 service water failure of 3 SSW pumps to start. Phase I SAMAs were implemented to improve pumps to start service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| E.1-14 J J J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SSW-MDP-FS-P208E 2.02E-03 1.016 SSW pump P-208E This term represents random failure of SSW pump P-208E to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| IE-S1 3.00E-04 1.015 Medium LOCA This term represents the medium LOCA initiating event. Several Phase I SAMAs have been implemented to provide more reliable or diverse high or low pressure injection systems to mitigate this event. Phase II SAMAs 040, 041, 042, 043, 044, and 054 were evaluated to reduce the CDF contribution from medium LOCA.LCS-STR-PG-8002A 1.20E-04 1.014 ECCS strainer BS- This term represents failure of core spray and RHR suction 8002A plugged strainer BS-8002A.
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| A Phase I SAMA was implemented to install improved passive ECCS suction strainers.
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| Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.
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| LCS-STR-PG-8002B 1.20E-04 1.014 ECCS strainer BS- This term represents failure of core spray and RHR suction 8002B plugged strainer BS-8002B.
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| A Phase I SAMA was implemented to install improved passive ECCS suction strainers.
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| Phase II SAMAs 042, 044, and 045, which recommend addition of independent injection systems to mitigate this failure event, were evaluated.
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| E.1-15 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XISI 7.40E-03 1.013 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during medium LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of during medium instrumentation to enhance the likelihood of success of operator LOCA action in response to accident conditions, have already been implemented.
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| No additional Phase II SAMAs were recommended for this subject.EDG-ENG-FR-EDGB 6.10E-03 1.013 Emergency diesel This term represents random failure of EDG-B, leading to an SBO generator -B (EDG) event. Phase I SAMAs to improve availability and reliability of the fails to continue to EDGs by creating a cross-tie of EDGs fuel oil supply and run installing a backup SBO diesel generator have already been implemented.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.
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| AC8-CBR-CO-104 9.50E-05 1.013 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-104 fails to 104, leading to loss of power to 480V MCC B17 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.
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| HCI-MAI-MA-HCITM 1.62E-02 1.013 HPCI unavailable This term represents HPCI system unavailable due to due to maintenance maintenance.
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| Phase I SAMAs to improve availability and reliability of the HPCI system that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation.
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| Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.E.1-16 J I 9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SSW-CCF-FR-3P208 5.59E-06 1.012 Common cause This term represents common cause failure of 3 service water failure of 3 SSW pumps to continue to run Phase I SAMAs were implemented to pumps to run improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| AC8-CBR-CO-205 9.50E-05 1.012 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-205 fails to 205, leading to loss of power to 480V MCC B18 and its remain closed associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.
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| IE-T3C 4.40E-02 1.012 Inadvertently This term represents an initiating event caused by inadvertent opened relief valve opening of a relief valve. Improvement of the SRV design and SRV reseat reliability, to reduce the probability and consequences of this initiating event, were evaluated in Phase II SAMAs 046 and 050.RBC-CCF-CC-4MOVS 1.13E-05 1.012 Common cause This term represents common cause failure of RBCCW heat failure of RBCCW exchanger A & B side MOVs to open. A Phase I SAMA was heat exchanger A & implemented to improve RBCCW system reliability by making B side MOVs (4) to component cooling water trains separate.
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| Phase II SAMA 055 to open improve RBCCW system reliability by reducing common dependencies was evaluated.
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| E.1-17 Ar7 Ak =- I- I- Ar Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition OSP-24 1.41 E-02 1.011 Failure to recover This term represents operator failure to recover offsite power offsite power within within 24 hours during a LOOP event. Phase I SAMAs, including 24 hours improvement of SBO procedures and training to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase 11 SAMAs were recommended for this subject.SSW-RCI-FE-3828X 3.OOE-04 1.01 Pressure switch This term represents random failure of SSW pressure switch PS-PS-3828X coil fails 3828X, resulting in loss of SSW system loop A. Phase I SAMAs to energize were implemented to improve service water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps. Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| EDG-MAI-MA-EDGA 6.41E-03 1.01 EDG-A out for This term represents EDG-A out for maintenance, leading to an maintenance SBO event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.
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| E.1-18 D0-3 9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition EDG-ENG-FR-EDGA 6.1 OE-03 1.01 EDG-A fails to This term represents random failure of EDG-A, leading to an SBO continue to run event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented.
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| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.
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| SSW-MOV-OO-V3805 6.62E-04 1.009 SSW TBCCW A This term represents random failure of SSW MOV MO-3805 to go heat exchanger 90% closed, resulting in loss of SSW to RBCCW loop B. A Phase outlet MOV MO- I SAMA was implemented to improve RBCCW system reliability 3805 fails to go by making component cooling water trains separate.
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| Phase II 90% closed SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| SSW-MDP-FS-P208B 2.02E-03 1.009 SSW pump P-208B This term represents random failure of SSW pump P-208B to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
| |
| SSW-MDP-FS-P208A 2.02E-03 1.009 SSW pump P-208A This term represents random failure of SSW pump P-208A to fails to start on start. Phase I SAMAs were implemented to improve service demand water system reliability by enhancing screen wash, adding redundant DC control power for SSW pumps, and increasing seismic integrity of the partition wall between the SSW pumps.Phase II SAMA 055 to improve SSW system reliability by reducing common dependencies was evaluated.
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| E.1-19
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| --de,-I Jr, k-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition C 5.80E-06 1.009 Reactor Protection This term represents failure of the RPS. Several Phase I SAMAs System (RPS) to minimize the risks associated with anticipated transient without failure scram (ATWS) scenarios have already been installed.
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| No Phase 11 SAMAs were evaluated to further improve reliability of RPS.However, Phase II SAMA 048 to enhance reliability of the standby liquid control system and improve capability to mitigate the consequences of an ATWS event was evaluated.
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| AC4-RCK-NO-605 2.51 E-03 1.009 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-605 breaker 152-605, leading to loss of power to safety bus A6.control circuit no Phase I SAMAs to improve 4.16kV bus cross-tie capability and output procedures to repair or replace failed 4.16kV breakers have already been installed.
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| In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| RCI-TDP-RS-P206 1.52E-02 1.009 RCIC turbine driven This term represents random failure of the RCIC system. Phase I pump P-206 fails to SAMAs to improve availability and reliability of the RCIC system restart after clear that have already been implemented include raising high level signal backpressure trip setpoints and proceduralizing intermittent operation.
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| Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.FXT-RCK-NO-P140 2.50E-03 -1.009 Diesel fire pump P- This term represents diesel fire pump P-140 control circuit failure.140 control circuit Phase II SAMA 045, to add a diverse injection system and no output provide an injection source other than fire water, was evaluated.
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| E.1-20 3 J 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition AC4-RCK-NO-508 2.51 E-03 1.008 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-508 breaker 152-508, leading to loss of powerto 480V load center B1.control circuit no Phase I SAMAs to improve 4.16kV bus cross-tie capability and output revise procedures to repair or replace failed 4.16kV breakers have already been implemented.
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| In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| AC8-RCK-NO-101 2.50E-03 1.008 480V circuit breaker This term represents random failure of 480V circuit breaker 52-52-101 control 101, leading to loss of power to 480V load center BI and its circuit no output associated loads. A Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 030 and 058 to improve 480V bus availability were evaluated.
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| EDG-MAI-MA-EDGB 4.09E-03 1.008 EDG-B out for This term represents EDG-B out for maintenance, leading to an maintenance SBO event. Phase I SAMAs to improve availability and reliability of the EDGs by creating a cross-tie of EDGs fuel oil supply and installing a backup SBO diesel generator have already been implemented.
| |
| Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035, for enhancing AC or DC system reliability or to cope with LOOP and SBO events, were evaluated.
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| E.1-21 Ir-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition HCI-TDP-FS-PM205 7.53E-03 1.008 HPCI turbine driven This term represents random failure of the HPCI system. Phase l pump P-205 fails to SAMAs to improve availability and reliability of the HPCI system start on demand that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation.
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| Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.RBC-CCF-FS-4PUMP 7.35E-06 1.008 Common cause This term represents common cause failure of four RBCCW failure of four pumps to start. A Phase I SAMA was implemented to improve RBCCW pumps to RBCCW system reliability by making component cooling water start trains separate.
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| Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| AC4-RCK-NO-505 2.51 E-03 1.007 4.16kV circuit This term represents failure of the control circuit of 4.16kV circuit breaker 152-505 breaker 152-505, leading to loss of power supply to safety bus control circuit no A5. Phase I SAMAs to improve 4.16kV bus cross-tie capability output and revise procedures to repair or replace failed 4.16kV breakers have already been installed.
| |
| In addition, a Phase I SAMA was implemented to proceduralize operator action to manually close the circuit breaker. Phase II SAMAs 025, 026, 027, 028, 029, 030, 033, and 035 for enhancing AC or DC system reliability or to cope with LOOP and SBO events were evaluated.
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| FXT-XVM-CC-511 5.OOE-04 1.007 Manual valve 10- This term represents random failure of manual valve 10-HO-511 HO-511 fails to to open to provide fire water to LPCI loops A and B. This failure open leads to loss of fire water backup for reactor vessel injection and drywell spray. Phase II SAMA 059 to enhance availability of the fire water system was evaluated.
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| E.1-22 3 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event DescrIption Disposition FXT-XVM-CC-8156 5.00E-04 1.007 Manual valve 8-1-56 This term represents random failure of manual valve 8-1-56 to fails to open open to provide fire water to LPCI loops A and B. This failure leads to loss of fire water backup for reactor vessel injection and drywell spray. Phase II SAMA 059 to enhance availability of the fire water system was evaluated.
| |
| RCI-MAI-MA-RCITM 1 .97E-02 1.007 RCIC unavailable This term represents RCIC system unavailable due to due to maintenance maintenance.
| |
| Phase I SAMAs to improve availability and reliability of the RCIC system that have already been implemented include raising backpressure trip setpoints and proceduralizing intermittent operation.
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| Additional improvements were evaluated in Phase II SAMAs 040, 041, 042, 043, 044, and 045.CIV-AOV-CC-5042B 1.OOE-03 1.007 AO 5042B fails to This term represents random failure of DTV valve AO 5042B to open on demand open on demand, resulting in loss of containment venting capability to control containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated for containment pressure control.CIV-AOV-CC-A5025 1.OOE-03 1.007 AO 5025 fails to This term represents random failure of DTV valve AO 5025 to open on demand open on demand, resulting in loss of containment venting capability to control containment pressure.
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| Phase II SAMAs 001, 009, 014, and 059, to provide alternate means of suppression pool cooling and drywell spray and to enhance the availability and reliability of firewater for reactor vessel injection and drywell spray, were evaluated for containment pressure control.E.1-23 le", 1. -. -7.-- .-7--1-7-1a-
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| -I Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition CM 3.30E-01 1.006 RPS mechanical This term represents random failure of the RPS. Several Phase I failure SAMAs to minimize the risks associated ATWS scenarios have already been installed.
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| No Phase II SAMAs were evaluated to further improve reliability of RPS. However, Phase II SAMA 048 to enhance reliability of the standby liquid control system and improve ATWS capability to mitigate the consequences of this event was evaluated.
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| RBC-MAI-MA-P202E 6.71 E-03 1.006 RBCCW pump This term represents RBCCW pump 202E unavailable due to 202E out for maintenance.
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| A Phase I SAMA was implemented to improve maintenance RBCCW system reliability by making component cooling water trains separate.
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| Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| RBC-MAI-MA-P202F 6.44E-03 1.006 RBCCW pump This term represents RBCCW pump 202F unavailable due to 202F out for maintenance.
| |
| A Phase I SAMA was implemented to improve maintenance RBCCW system reliability by making component cooling water trains separate.
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| Phase II SAMA 055 to improve RBCCW system reliability by reducing common dependencies was evaluated.
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| IE-TDC-CCF 3.66E-08 1.006 Common cause This term represents an initiating event caused by a complete failure of 125VDC loss of 125VDC buses D-16 and D-17 or random failure of buses A&B batteries D-1 and D-2. Phase I SAMAs to improve battery charging capability and replace existing batteries with more reliable ones have already been installed.
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| Phase II SAMAs 025, 026, 027, 031, 032, 033, 034, and 035 for enhancing DC system availability and reliability were evaluated.
| |
| E.1 -24
| |
| :3 I9 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level I Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition SPC-MAI-MA-SPCA 3.01 E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop A cooling loop A out unavailable due to maintenance.
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| Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability.
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| Additional improvements were evaluated in Phase II SAMAs 001 and 014.SPC-MAI-MA-SPCB 2.91E-03 1.005 Suppression pool This term represents RHR suppression pool cooling loop B cooling loop B out unavailable due to maintenance.
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| Phase I SAMAs to improve for maintenance availability and reliability of the RHR suppression pool cooling mode that have already been implemented include using drywell spray mode and fire protection cross-tie to provide redundant containment heat removal capability.
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| Additional improvements were evaluated in Phase II SAMAs 001 and 014.DWS-MAI-MA-DWSA 3.18E-03 1.005 Drywell spray loop This term represents RHR drywell spray loop A unavailable due A out for to maintenance.
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| Phase I SAMAs to improve availability and maintenance reliability of the RHR drywell spray mode that have already been implemented include using suppression pool cooling mode and fire protection cross-tie to provide redundant containment heat removal capability.
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| Additional improvements were evaluated in Phase II SAMA 009.E.1-25
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| _A , Yc-r-i 01*Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-3 Correlation of Level 1 Risk Significant Terms to Evaluated SAMAs Event Name Probability RRW Event Description Disposition ADS-XHE-FO-XIS2 1 .45E-03 1.005 Operator fails to This term represents operator failure to manually open the SRVs perform emergency for depressurization during a small LOCA. Phase I SAMAs, depressurization including improvement of procedures and installation of during small LOCA instrumentation to enhance the likelihood of success of operator action in response to accident conditions, have already been implemented.
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| No additional Phase II SAMAs were recommended for this subject.E.1-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2 PSA Model -Level 2 Analysis E.1.2.1 Containment Performance Analysis The PNPS Level 2 PSA model used for the SAMA analysis is the most recent internal events risk model, which is an updated version of the model used in the IPE [References E.1-2 and E.1-3].The Level 2 PSA model used for the SAMA analysis, Revision 1, reflects the PNPS operating configuration and design changes as of September 2001. Specifically, the Level 2 model has been updated to incorporate insights from the independent BWROG peer review.The PNPS Level 2 model includes two types of considerations:
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| (1) a deterministic analysis of the physical processes for a spectrum of severe accident progressions, and (2) a probabilistic analysis component in which the likelihood of the various outcomes are assessed.
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| The deterministic analysis examines the response of the containment to the physical processes during a severe accident.
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| This response is performed by* utilization of the MAAP code [Reference E. 14] to simulate severe accidents that have been identified as dominant contributors to core damage in the Level 1 analysis, and* reference calculation of several hydrodynamic and heat transfer phenomena that occur during the progression of severe accidents.
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| Examples include debris coolability, pressure spikes due to ex-vessel steam explosions, scoping calculation of direct containment heating, molten debris filling the pedestal sump and flowing over the drywell floor, containment bypass, deflagration and detonation of hydrogen, thrust forces at reactor vessel failure, liner melt-through, and thermal attack of containment penetrations.
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| The Level 2 analysis examined the dominant accident sequences and the resulting plant damage states (PDS) defined in Level 1. The Level I analysis involves the assessment of those scenarios that could lead to core damage. A list of the PDS groups and descriptions from the Level 2 analysis is presented in Table E.1-4.A full Level 2 model was developed for the IPE and completed at the same time as the Level 1 model. The Level 2 model consists of a single containment event tree (CET) with functional nodes that represent phenomenological events and containment protection system status. The nodes were quantified using subordinate trees and logic rules. A list of the CET functional nodes and descriptions used for the Level 2 analysis is presented in Table E.1-5.The Large Early Release Frequency (LERF) is an indicator of containment performance from the Level 2 results because the magnitude and timing of these releases provide the greatest nntontfil fnr incriv haIth affontc tn theg nohlh- Tha frong icnev role infin *nnrnyimqtcmlv Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-4 Summary of PNPS PSA Core Damage Accident Class PoS S l Point % of Total Group SipiidDescription Estimate CDF LOCAs Large and small break LOCA with initial or long-term loss 1.16E-7 1.80 of core cooling. Core damage results at low or high reactor pressure.
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| For most PDS, late injection and containment heat removal are available.
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| TRANS Short and long-term transient events. Core damage 2.43E-7 3.79 results at either low or high reactor pressure.
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| Late injection and containment heat removal are available.
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| SBO SBO involving a loss of high-pressure injection.
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| Core 1.48E-7 2.31 damage results at either low (stuck-open SRV) or high reactor pressure.
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| All accident mitigating functions are recoverable when AC power is restored.VSLRUPT Vessel rupture event resulting In LOCA beyond ECCS 4.OOE-9 0.06 capability.
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| All PDS result in core damage at low reactor pressure with late injection available.
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| ATWS Short-term ATWS that leads to early core damage at high 3.39E-8 0.53 reactor pressure following loss of reactivity control and rapid containment pressurization.
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| Reactor coolant system leakage rates associated with boil-off of coolant through the cycling of SRVs/SV with early core melt subsequent to containment overpressure failure. Late injection and containment heat removal are available.
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| ISLOCA Large and small break interfacing system LOCA outside 4.00E-9 0.06 containment.
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| Core damage results at low or high reactor pressure with a bypassed containment.
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| TW Containment decay heat removal systems are not 5.86E-6 91.45 available and coolant recirculation to the torus over pressurizes the containment to failure or venting. The torus is saturated.
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| Total 6.41 E-06 1.OOE+00 E.1-28 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description CET Node CET Functional Node Description Plant Damage State This top event represents the initiators considered in the containment Event (PDSEVNT) performance analysis.RPV Pressure at This top event identifies the status of the reactor pressure vessel (RPV)Vessel Failure pressure.
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| RPV@VF is set to success when RPV pressure is low.(RPV@VF) RPV@VF is set to failure when RPV is high.In-Vessel Cooling This top event addresses the recovery of coolant injection into the vessel Recovery (IN-REC) after core degradation, but prior to vessel breach. This top event considers the possibility of low-pressure injection systems working once the RPV is depressurized.
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| Vessel Failure (VF) This top event addresses recovery from core degradation within the vessel and the prevention of vessel head thermal attack. Core melt recovery requires the recovery of core cooling prior to core blocking or relocation of molten debris to the lower plenum and thermal attack of the vessel head.Early Containment This top event node considers the potential loss of containment integrity at, Failure (CFE) or before, vessel failure. Several phenomena are considered credible mechanisms for early containment failure. They may occur alone or in combination.
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| The phenomena are containment isolation failure;containment bypass; containment overpressure failure at vessel breach;hydrogen deflagration or detonation; fuel-coolant interactions (steam explosions);
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| high pressure melt ejection and subsequent direct containment heating; and drywell steel shell melt-through.I Early Release to Torus (EPOOL)This top event node considers the importance of early torus pool scrubbing in mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to early containment failure is through the torus water and the torus airspace.That is, the torus pool is not bypassed.
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| Failure involves a release into the drywell.Debris Cooled Ex-vessel (DCOOL)This top event considers the delivery of water to the drywell, via drywell sprays, or via injection to the RPV and drainage out an RPV breach onto the drywell floor. Success implies the availability of water and the formation of a coolable debris bed such that concrete attack is precluded.
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| Failure implies that the molten core attacks concrete in the reactor pedestal, that core debris remains hot, and sparing of the concrete decomposition products through the melt releases the less volatile fission products to the containment atmosphere.
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| E.1-29 J, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-5 Notation and Definitions for PNPS CET Functional Nodes Description (Continued)
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| CET Node CET Functional Node Description Late Containment This top event addresses the potential loss of containment integrity in the Failure (CFL) long-term.
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| Late containment failure may result from long-term steam and non-condensable gas generation from the attack of molten core debris on concrete.Late Release to This top event node considers the importance of late torus pool scrubbing in Torus (LPOOL) mitigating the magnitude of fission products released from the damaged core. Success implies that fission product transport path subsequent to late containment failure is through the torus water and the torus airspace.
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| That is, the torus pool is not bypassed.
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| Failure involves a release into the drywell.Fission Product This top event addresses fission product releases from the fuel into the Removal (FPR) containment and airborne fission product removal mechanisms within the containment structure to characterize potential magnitude of fission product releases to the environment should the containment fail. Failure implies that most of the fission products from the fuel and containment are ultimately released to the environment without mitigation.
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| Reactor Building This top event is used to assess the ability of the reactor building to retain (RB) fission products released from containment.
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| Success of top event RB is defined to be a reduction of the containment release magnitude.
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| E.1-30
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| .Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage fW- 'Lat, Eary LowFRlease
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| ,.0.52% a e Hgh Reease O4. OO/o a\Jum Release-'/uly H Eal/Ah Release.76%Ealy Medun Rlease 1 mo 1°/Late MWc 2A AN-No Con~imnat Failure 1.73%LateL R.lease 70.65%Figure E.1-1 PNPS Radionuclide Release Category Summary E.1-31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Transierts Aticipated Transiert wthot 201%Sacam 39.82%1. .Interfaang System LOCAs LOCus t 13 2 YO 0.01%Ve sa Rome 0.01%Staficn Blackout 57.03%Figure E.1-2 PNPS Plant Damage State Contribution to LERF E.1-32 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2.2 Radionuclide Analysis E.1.2.2.1 Introduction A major feature of a Level 2 analysis is the estimation of the source term for every possible outcome of the CET. The CET end points represent the outcomes of possible in-containment accident progression sequences.
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| These end points represent complete severe accident sequences from initiating event to release of radionuclides to the environment.
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| The Level I and plant system information is passed through to the CET evaluation in discrete PDS. An atmospheric source term may be associated with each of these CET sequences.
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| Because of the large number of postulated accident scenarios considered, mechanistic calculations (i.e., MAAP calculations) are not performed for every end-state in the CET. Rather, accident sequences produced by the CET are grouped or 'binned' into a limited number of release categories each of which represents all postulated accident scenarios that would produce a similar fission product source term.The criteria used to characterize the release are the estimated magnitude of total release and the timing of the first significant release of radionuclides.
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| The predicted source term associated with each release category, including both the timing and magnitude of the release, is determined using the results of MAAP calculations
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| [Reference E.1-4].E.1.2.2.2 Timing of Release Timing completely governs the extent of radioactive decay of short-lived radioisotopes prior to an off-site release and, therefore, has a first-order influence on immediate health effects. PNPS characterizes the release timing relative to the time at which the release begins, measured from the time of accident initiation.
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| Two timing categories are used: early (0-24 hours) and late (>24 hours).Based on MAAP calculations for a spectrum of severe accident sequences, PNPS expects that an Emergency Action Level (as defined by the PNPS Emergency Plan) will be reached within the first half hour after accident initiation.
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| Reaching an Emergency Action Level initiates a formal decision-making process that is designed to provide public protective actions. Within 24 hours of accident initiation, the Level 2 analysis assumed that off-site protective measures would be effective.
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| Therefore, the definitions of the release timing categories are as follows.* Early releases are CET end-states involving containment failure prior to or at vessel failure or after vessel failure and occurring within 0 to 24 hours measured from the time of accident initiation and for which minimal offsite protective measures would be accomplished.
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| * Late releases are CET end-states involving containment failure greater than 24 hours from the time of accident initiation, for which offsite measures are fully effective.
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| E.1-33 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2.2.3 Magnitude of Release Source term results from previous risk studies suggest that categorization of release magnitude based on cesium iodide (CsI) release fractions alone are appropriate
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| [Reference E.1-5]. The CsI release fraction indicates the fraction of in-vessel radionuclides escaping to the environment.(Noble gas release'levels are non-informative since release of the total core inventory of noble gases is essentially complete given containment failure).The source terms were grouped into four distinct radionuclide release categories or bins according to release magnitude as follows: (1) High (HI) -A radionuclide release of sufficient magnitude to have the potential to cause early fatalities.
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| This implies a total integrated release of >10% of the initial core inventory of Csl [Reference E.1-5].1 (2) Medium (MED) -A radionuclide release of sufficient magnitude to cause near-term health effects. This implies a total integrated release of between 1 and 10% of the initial core inventory of CsI [Reference E. 1-5].2 (3) Low (LO) -A radionuclide release with the potential for latent health effects. This implies a total integrated release of between 0.001% an'd 1% of the initial core inventory of CsI.(4) Negligible (NCF) -A radionuclide release that is less than or equal to the containment design base leakage. This implies total integrated release of<0.001% of the initial core inventory of Csl.The "total integrated release" as used in the above categories is defined as the integrated release within 36 hours after RPV failure. If no RPV failure occurs, then the "total integrated release" is defined as the integrated release within 36 hours after accident initiation.
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| E.1.2.2.4 Release Category Bin Assignments Table E.1-6 summarizes the scheme used to bin sequences with respect to magnitude of release, based on the predicted Csl release fraction and release timing. The combi nation of release magnitude and timing produce seven distinct release categories for source terms. These are the representative release categories presented in Table E. 1-7.1. Once the Csl source term exceeds 0.1, the source term Is large enough that doses above the early fatality threshold can sometimes occur within a population center a few miles from the site.2. The reference document indicates that for'Csl release fractions of 1 to 10%, the number of latent fatalities is found to be at least 10% of the latent fatalities for the highest release.E.1-34 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-6 Release Severity and Timing Classification Scheme Summary Release Severity Release Timing Classification Classification Time of Initial Release from Category Csl % Release Category Accident Initiation High Greater than 10 Early (E) Less than 24 hours Medium i to 10 Low 0.001 to 1 Late (L) Greater than 24 hours Negligible Less than 0.001 Table E.1-7 PNPS Release Categories Timing of Magnitude of Release Release Low Medium High Early Early/Low Early/Med Early/High NCF Late Late/Low Late/Med Late/High)E.1.2.2.5 Mapping of Level 1 Results into the Various Release Categories PDS provide the interface between the Level 1 and Level 2 analyses (i.e. between core damage accident sequences and fission product release categories).
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| In the PDS analysis, Level 1 results were grouped ("binned")
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| according to plant characteristics that define the status of the reactor, containment, and core cooling systems at the time of core damage. This ensures that systems important to core damage in the Level 1 event trees, and the dependencies between containment and other systems are handled consistently in the Level 2 analysis.
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| A PDS therefore represents a grouping of Level 1 sequences that defines a unique set of initial conditions that are likely to yield a similar accident progression through the Level 2 CETs and the attendant challenges to containment integrity.
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| From the perspective of the Level 2 assessment, PDS binning entails the transfer of specific information from the Level 1 to the Level 2 analyses.Equipment failures in Level 1. Equipment failures in support systems, accident prevention systems, and mitigation systems that have been noted in the Level 1 analysis are carried into the Level 2 analysis.
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| In this latter analysis, the repair or recovery of failed equipment is not allowed unless an explicit evaluation, including a consideration of E.1-35 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage adverse environments where appropriate, has been performed as part of the Level 2 analysis.* RPV status. The RPV pressure condition is explicitly transferred from the Level I analysis to the CET.* Containment status. The containment status is explicitly transferred from the Level 1 analysis to the CET. This includes recognition of whether the containment is bypassed or is intact at the onset of core damage.* Accident sequence timing. Differences in accident sequence timing are transferred with the Level 1 sequences.
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| Timing affects such sequences as SBO, internal flooding, and containment bypass (ISLOCA).This transfer of information allows timing to be properly assessed in the Level 2 analysis.Based on the above criteria, the Level 1 results were binned into 48 PDS. These PDS define important combinations of system states that can result in distinctly different accident progression pathways and, therefore, different containment failure and source term characteristics.
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| Table E.1-8 provides a description of the PNPS PDS that are used to summarize the Level 1 results."ms Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States Point %fD PDS Description Estimate %ofCDF PDS-1 Long-term LOCA with loss of high-pressure core makeup O.OOE+00 0.00 from HPCI and RCIC, loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage results at high primary system pressure.
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| Late injection from low-pressure systems (core spray, LPCI, and firewater) is available, provided primary system depressurization occurs. The containment is vented and intact.PDS-2 Long-term LOCA with loss of both high-pressure core 1.05E-11 <0.001 makeup (HPCI and RCIC) and containment heat removal.Core damage results at high primary system pressure.Because containment venting fails, containment failure occurs long-term.
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| Late injection is available from low-pressure systems (core spray, LPCI, and fire water)provided they survive containment failure.E.1-36 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| Point PDS Description Estimate % of CDF PDS-3 Short-term LOCA with loss of high-pressure core makeup, 8.68E-08 1.35 and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure.
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| Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Containment heat removal is available.
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| PDS-4 Short-term LOCA with loss of high-pressure core makeup, O.OOE+00 <0.001 loss of containment heat removal, and failure to depressurize the primary system for low-pressure core makeup. Core damage occurs at high primary system pressure.
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| Late injection from core spray, LPCI, and firewater is available, provided primary system depressurization occurs. Unlike PDS-3, containment heat removal is unavailable.
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| PDS-5 Long-term LOCA with loss of high-pressure core makeup 0.OOE+00 0.00 and containment heat removal. Core damage occurs at low primary system. Late injection is available from low-pressure systems (core spray, LPCI, and fire water). The containment is vented and intact.PDS-6 Long-term large LOCA. High-pressure core makeup from 0.00E+00 0.00 HPCI and RCIC are unavailable due to the large LOCA.Because containment venting fails, containment failure occurs long-term.
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| Late injection is available from low-pressure systems (core spray, LPCI, and fire water)provided they survive containment failure. Core damage occurs at low primary system pressure.PDS-7 Short-term large LOCA with loss of core cooling. Core 1.12E-09 0.08 damage results at low primary system pressure.
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| Late injection from firewater cross tie and containment heat removal are available.
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| PDS0- Short-term large LOCA with loss of core cooling. Core 4.43E-09 0.07 damage results at low primary system pressure.
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| Late injection from firewater cross tie is available.
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| However, unlike PDS-7, containment heat removal is unavailable.
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| Q.E.1-37 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| Point PDS Description Estimate l of CDF PDS-9 Short-term LOCA with loss of high and low-pressure core 3.64E-09 0.06%cooling. Because the primary system is depressurized, core damage results at low primary system pressure.
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| Late injection from SSW system, containment venting, and containment heat removal are available.
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| PDS-10 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Because the primary system is depressurized, core damage results at low primary system pressure.
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| Late injection from SSW system and containment heat removal are available.
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| However, unlike PDS-9, containment venting is not available.
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| PDS-11 Short-term LOCA with loss of high and low-pressure core O.OOE+00 0.00 cooling. Core damage results at low primary system pressure.
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| Late injection from SSW system is available.
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| However, unlike PDS-9, containment venting and containment heat removal are unavailable.
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| PDS-12 Transient with a loss of long-term decay heat removal. Core 2.37E-08 0.37 damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| The containment is vented and remains intact at the time of core damage.PDS-13 Transient with a loss of long-term decay heat removal. Core 3.75E-06 58.5 damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Unlike PDS-12 containment venting fails.PDS-14 Short-term transient with failure to depressurize the primary 1 .52E-07 2.37 system. Core damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Containment heat removal from RHR is available.
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| PDS Short-term transient with failure to depressurize the primary 5.07E-08 0.79 system. Core damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Containment heat removal from RHR is available.
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| However, containment venting is not available.
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| E.1-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| PDS Description Point % of CDF PDS-16 Short-term transient with failure to depressurize the primary 4.89E-09 0.08 system. Core damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Containment heat removal from RHR is not available, but containment venting is available.
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| PDS-17 Short-term transient with failure to depressurize the primary 2.53E-09 0.04 system. Core damage results at high primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Neither containment heat removal from RHR nor containment venting is available.
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| PDS-18 Transient with a loss of long-term decay heat removal. 1 .56E-06 24.40 Core damage results at low primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| The containment is vented and remains intact at the time of core damage.PDS-19 Transient with a loss of long-term decay heat removal. 5.24E-07 8.18 Core damage results at low primary system pressure.
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| Late in-vessel and ex-vessel injection is available.
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| Unlike PDS-18 containment venting fails.PDS-20 Long-term transients with loss of core cooling. Core 6.78E-11 0.001 damage results at low primary system pressure.
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| No late injection, but containment heat removal is available.
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| PDS-21 Short-term transients (IORV) with loss of core cooling. 8.18E-09 0.13 Core damage results at low primary system pressure.
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| Late injection and containment heat removal are available.
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| PDS-22 Short-term transients with loss of core cooling. Core 1.08E-09 0.02 damage results at low primary system pressure.
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| Late injection and containment heat removal are available.
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| However, containment venting is not available.
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| PDS-231 Short-term transients with loss of core cooling. Core O.OOE+00 0.00 damage results at low primary system pressure.
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| Late injection and containment venting are available, but containment heat removal is not available.
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| PDS-24 Similar to PDS-23, except that containment venting is not 4.98E-09 0.08 available.
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| I Cl E.1-39 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| Point PDS Description Estimate C of CDF PDS-25 Short-term transients with loss of core cooling. Core 2.57E-09 0.04 damage results at low primary system pressure.
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| No late injection, but containment heat removal and containment venting are available.
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| PDS-26 Similar to PDS-25, except that containment venting is not 1.24E-08 0.19 available.
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| PDS-27 Short-term transients with loss of core cooling. Core 4.40E-11 0.001 damage results at low primary system pressure.
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| Late injection and containment heat removal are not available.
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| However, containment venting is available PDS-28 Short-term transients with loss of core cooling. Core 1.10E-09 0.02 damage results at low primary system pressure.
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| Late injection, containment heat removal and containment venting are not available.
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| PDS-29 Long-term SBO involving loss of injection at high primary 1.41 E-07 2.21 system pressure from battery depletion.
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| All accident-mitigating functions are recoverable when AC power is restored.PDS-30 Short-term SBO sequence involving a loss of high-pressure O.OOE+00 0.00 injection at high primary system pressure from loss of all AC power and DC power or failure of SRVs. All accident-mitigating functions are recoverable when offsite power is restored.PDS-31 Long-term SBO sequence Involving a loss of high-pressure 2.60E-09 0.04 injection due to one stuck-open safety relief valve or long-term failure of HPCI and RCIC and subsequent failure to depressurize the primary system. Core damage results at low primary system pressure.
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| All accident-rnitigating functions are recoverable when offsite power is restored.PDS-32 Short-term SBO sequence involving a loss of high-pressure 4.OOE-09 0.06 injection due to two stuck-open safety relief valves or failure of HPCI and RCIC and one stuck-open safety relief valve.Core damage results at low primary system pressure.
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| All accident-mitigating functions are recoverable when offsite power is restored.E.1-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.18 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| Point PDS Description Estimate % of CDF PDS-33 Short-term large reactor vessel rupture. The resulting loss 4.OOE-09 0.06 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure.
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| Vessel injection and all forms of containment heat removal (RHR and containment venting) are available.
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| The containment is not bypassed and AC power is available.
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| PDS-34 Similar to PDS-33, except that containment heat removal O.OOE+00 0.00 from RHR fails.PDS-35 Short-term large reactor vessel rupture. The resulting loss O.OOE+00 0.00 of coolant is beyond the makeup capability of ECCS. Core damage occurs in the short term at low primary system pressure.
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| Vessel injection is unavailable.
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| However, all forms of containment heat removal (RHR and containment venting) are available.
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| The containment is not bypassed and AC power is available.
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| PDS-36 Similar to PDS-35, except that containment heat removal 0.OOE+00 0.00 from RHR fails.PDS-37 Short-term ATWS with failure of SRVs and SVs to open to- 1.95E-08 0.31 reduce primary system pressure.
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| The ensuing primary system over pressurization leads to a LOCA beyond core cooling capabilities.
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| Late injection and containment heat removal are available.
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| PDS-38 Short-term ATWS that leads to early core damage at low 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is not available.
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| However, containment heat removal is available.
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| PDS-39 Similar to PDS-38 except that containment heat removal 2.32E-09 0.04 from the RHR system is not available.
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| PDS-40 Long-term ATWS that leads to late core damage at low 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available.
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| The containment is vented.CW.)E.1-41 C J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-8 Summary of PNPS Core Damage Accident Sequences Plant Damage States (Continued)
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| PDS Description Estinate % of CDF PDS-41 Short-term ATWS that leads to early core damage at high 1.34E-11 <0.001 primary system pressure following successful reactivity control. Late injection and containment heat removal are available.
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| PDS-42 Similar to PDS-41 except that containment heat removal 0.00E+00 0.00 from the RHR system Is not available.
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| PDS-43 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available; containment heat removal from the RHR is not available.
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| The containment is vented.PDS-44 Long-term ATWS that leads to late core damage at high 0.OOE+00 0.00 primary system pressure following successful reactivity control. Late injection is available.
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| However, containment heat removal from the RHR system and containment venting are not available.
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| PDS-45 Short-term ATWS that leads to containment failure and 3.39E-08 0.53 early core damage at high primary system pressure because of inadequate reactor water level following a loss of reactivity control. Late injection and containment venting are available.
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| PDS-46 Short-term ATWS that leads to containment failure and 0.OOE+00 0.00 early core damage at high primary system pressure because of inadequate reactor water level following successful reactivity control. No late injection; however, containment venting Is available.
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| PDS-47 Unisolated LOCA outside containment with early core melt 3.22E-09 0.05 at high RPV pressure.PDS-48 Unisolated LOCA outside containment with early core melt 7.73E-10 0.01 at low RPV pressure.E.1-42 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The PDS designators listed in Table E.1-8 represent the core damage end state categories from the Level 1 analysis that are grouped together as entry conditions for the Level 2 analysis.
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| The Level 2 accident progression for each of the PDS is then evaluated using a single CET to determine the appropriate release category for each Level 2 sequence.
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| Each end state associated with a Level 2 sequence is assigned to one of the release categories depicted in Table E.1-7. Note, however, that since not all the Level 2 sequences associated with each Level 1 core damage class may be assigned to the same release category, there is no direct link between a specific Level 1 core damage PDS and Level 2 release category.
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| Rather, the sum of the Level 2 end state frequencies assigned to each release category determines the overall frequency of that release category.
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| The CET described in the Level 2 model determines the release category frequency attributed to each Level 1 core damage PDS.E.1.2.2.6 Collapsed Accident Progression Bins Source Terms The source term analysis results in hundreds of source terms for internal initiators, making calculation with the MACCS2 consequence model cumbersome.
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| Therefore, the source terms were grouped into a much smaller number of source term groups defined in terms of similar properties, with a frequency weighted mean source term for each group.The consequence analysis source terms groups are represented by collapsed accident progression bins (CAPB). The CAPB were generated by sorting the accident progression bins for each of the forty-eight PDS on attributes of the accident:
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| the occurrence of core damage, the occurrence of vessel breach, primary system pressure at vessel breach, the location of 0 containment failure, the timing of containment failure, and the occurrence of core-concrete interactions.
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| Descriptions of the CAPB are presented in Table E.1-9.E.1-43 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions CAPB Number Description CAPB-1 [CD, No VB, No CF, No CCI]Core damage (CD) occurs, but timely recovery of RPV injection prevents vessel breach (No VB). Therefore, containment integrity is not challenged (No CF) and core-concrete interactions are precluded (No CCI). However, the potential exists for in-vessel release to the environment due to containment design leakage.CAPB-2 [CD, VB, No CF, No CCI]Core damage (CD) occurs followed by -vessel breach (VB). Containment does not'fail structurally and is not vented (No CF). Ex-vessel releases are recovered, precluding core-concrete interactions (No CCI). Although containment does not fail, vessel breach does occur, therefore the potential exists for in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail.CAPB-3 [CD, VB, No CF, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment does not fail structurally and is not vented (No CF). However, ex-vessel releases are not recovered in time, and therefore core-concrete interactions occur (CCI). RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurs, containment does not fail, nor is the vent limit reached.CAPB-4 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, No CCII Core damage (CD) occurs followed by 'vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure Is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
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| [DCH] are possible).
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| There are no core concrete interactions (No CCI) due to the' presence of an overlying pool of water.II i i E.1-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage C!Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
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| CAPB Description Number CAPB-5 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-6 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
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| [DCH] are possible).
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| Following containment failure, core-concrete interactions occur (CCI).CAPB-7 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the torus (WW), above the water level. RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena.
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| Following containment failure, core-concrete interactions occur (CCI).CAPB-8 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
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| [DCH] are possible).
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.C)o E.1-45 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
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| CAPB Description Number CAPB-9 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure induced severe accident phenomena.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-10 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at time of vessel breach (this implies that high pressure induced severe accident phenomena
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| [OCH] are possible).
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| Following containment failure, core-concrete interactions occur (CCI).CAPB-11 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails either before core damage, during core damage, or at vessel breach (Early CF).Containment failure occurs in the drywell or below the torus water line (DW). RPV pressure is less than 200 psig at time of vessel breach; precluding high pressure Induced severe accident phenomena.
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| Following containment failure, core-concrete interactions occur (CCI).CAPB-12 [CD, VB, Late CF, WW, No CCI]Core damage JCD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.E.1-46 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (I Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
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| CAPB Description Number CAPB-13 [CD, VB, Late CF, WW, CCIX Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the torus (WW), above the water level. RPV pressure is not important because high-pressure severe accident phenomena (such as DCH) did not fail containment.
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| CAPB-14 [CD, VB, Late CF, DW, No CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails late due to loss of containment heat removal (Late CF). Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-1 5[CD, VB, Late CF, DW, CCI]Core damage (CD) occurs followed by vessel breach (VB). Containment fails late (late CF) due to core-concrete interactions (CCI) after vessel breach. Containment failure occurs in the drywell or below the torus water level (DW). RPV pressure is not important because high-pressure severe accident phenomena did not fail containment.
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| CAPB-16 [CD, VB, BYPASS, RPV pressure >200 psig, No CCI]Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.CAPB-17 [CD, VB, BYPASS, RPV pressure <200 psig, No CCI]Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment.
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| There are no core concrete interactions (No CCI) due to the presence of an overlying pool of water.E.1-47 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-9 Collapsed Accident Progression Bins (CAPB) Descriptions (Continued)
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| NuCber Description CAPB-18 [CD, VB, BYPASS, RPV pressure >200 psig, CCI]Small break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at high RPV pressure with a bypassed containment.
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| Following vessel breach, core-concrete interaction occurs (CCI).CAPB-19 [CD, VB, BYPASS, RPV pressure <200 psig, CCI]Large break interfacing system LOCA outside containment occurs. Core damage (CD) and subsequent vessel breach (VB) results at low RPV pressure with a bypassed containment.
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| Following vessel breach, core-concrete interaction occurs (CCI).I l Based on the above binning methodology, the salient Level 2 results are summarized in Tables%mv E.1-10 and E.1-11 respectively.
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| Table E.1-10 summarizes the results of the CET quantification.
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| This table identifies the total annual release frequency for each Level 2 release category.Table E.1-11 provides the frequency, time, duration, energy, and elevation of release for each CAPB.E.1-48 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal StageTable E.1-10 Summary of PNPS Containment Event Tree Quantification Release Category Release Frequency (Timing/Magnitude)
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| (/RY)Late Low 4.53E-06 Late Medium 1.56E-06 Late High O.OOE-00 Early Low 3.32E-08 Early Medium 6.48E-08 Early High 1.13E-07 No Containment Failure 1.11E-07 Nomenclature Timing L (Late) -Greater than 24 hours E (Early) -Less than 24 hours Magnitude JW)NCF LO MED Hi (Little to no release)(Low)(Medium)(High)-Less than 0.001% Csl-0.001 to 1% Cs1-1 to 10% Csl-Greater than 1 0% Csl E.1-49 ("V Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms !-CAPB CAPB Frequency (Iyear)Warning Time (sec)Elevation (m)Release Release Start I Duration Release Energy (sec)(sec)(W)1 CAPB-1 9.51 E-08 3.98E+03 3.OOE+01 2.20E+04 9.OOE+03 2.61E+05 2 CAPB-2 1.27E-08 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 3 CAPB-3 2.39E-09 3.96E+03 3.OOE+01 2.20E+04 9.OOE+03 2.50E+05 4 CAPB-4 3.29E-09 7.96E+03 3.OOE+01 1.83E+04 3.56E+03 1.IOE+07 5 CAPB-5 2.73E-09 1.31 E+04 3.OOE+01 2.53E+04 7.93E+03 8.34E+06 6 CAPB-6 7.95E-09 1.33E+04 3.OOE+01 2.56E+04 8.11E+03 8.23E+06 7 CAPB-7 7.93E-09 1.38E+04 3.OOE+01 2.61 E+04 8.46E+03 8.03E+06 8 CAPB-8 2.06E-08 9.18E+03 3.00E+01 2.OOE+04 4.59E+03 1.04E+07 9 CAPB-9 9.25E-09 9.21 E+03 3.OOE+01 2.44E+04 8.87E+03 4.18E+06 10 CAPB-10 8.53E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 11 CAPB-11 4.35E-08 1.37E+04 3.OOE+01 2.60E+04 8.40E+03 8.06E+06 12 CAPB-12 1.70E-06 2.84E+04 3.OOE+01 4.64E+04 9.OOE+03 7.59E+06 13 CAPB-13 2.30E-09 9.14E+03 3.OOE+01 2.71E+04 9.OOE+03 1.80E+06 14 CAPB-14 2.26E-06 2.66E+04 3.OOE+01 4.46E+04 9.OOE+03 7.08E+06 15 CAPB-15 2.12E-06 2.81 E+04 3.OOE+01 4.62E+04 9.OOE+03 7.60E+06 16 CAPB-16 1.18E-09 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 17 CAPB-17 6.91E-09 3.96E+03 3.OOE+01 2.14E+04 9.OOE+03 2.50E+05 18 CAPB-18 4.61E-10 3.96E+03 3.OOE+01 2.12E+04 9.OOE+03 2.50E+05 19 CAPB-19 2.43E-08 3.96E+03 3.OOE+01 2.18E+04 9.OOE+03 2.50E+05 E.1-50:
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-11 Collapsed Accident Progression Bin (CAPB) Source Terms (continued)
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| I Release Fractions NG T Cs Te Sr Ru La Ce Ba 1 1.99E-07 1.85E-07 1.85E-07 O.OOE+O0 1.24E-09 8.OOE-09 5.01E11 8.43E-11 1.70E-08 2 9.97E-05 4.81 E-05 4.66E-05 1.76E-07 3.97E-07 4.OOE-06 1.65E-08 5.15E-08 4.87E-06 3 9.97E-05 5.37E-05 4.97E-05 1.76E-06 5.80E-07 4.OOE-06 2.37E-08 1.57E-07 4,95E-06 4 1.OOE+00 4.90E-02 2.62E-02 4.18E-05 2.46E-05 3.66E-04 8.97E-07 3.04E-06 1.92E-04 5 9.85E-01 7.86E-02 3.68E-02 4.28E-05 4.1OE-05 3.66E-04 1.56E-06 6.79E-06 3.44E-04 6 1.OOE+00 4.02E-02 2.32E-02 1.48E-03 3.19E-04 3.66E-04 6.50E-06 7.17E-05 3.23E-04 7 9.76E-01 6.11 E-02 2.94E-02 1.26E-03 2,30E-04 3.66E-04 9.1 OE-06 1.06E-04 4.52E-04 8 1.OOE+00 2.98E-01 2.72E-01 3.07E-05 9.89E-04 2.23E-02 4.49E-05 6.57E-05 1.1 5E-02 9 5.97E-01 7.61 E-02 7.07E-02 1.41 E-05 9.72E-04 1.09E-02 3.69E-05 7.63E-05 1.02E-02 10 1.OOE+00 2.80E-01 2.49E-01 1.1 E-02 3.07E-03 1.81E-02 7.95E-05 5.81 E-04 1.03E-02 11 9.79E-01 1.73E-01 1.41 E-01 9.97E-03 3.13E-03 1.78E-02 1.22E-04 9.39E-04 1.72E-02 12 2.01 E-01 5.84E-05 4.37E-05 1.25E-07 2.36E-07 1.72E-06 8.04E-09 2.56E-08 2.99E-06 13 9.97E-01 7.99E-03 5.99E-03 1.76E-04 3.63E-05 3.66E-04 2.15E-06 1.41 E-05 4.52E-04 14 7.75E-01 2.88E-02 2.67E-02 2.47E-05 2.05E-04 2.13E-03 8.49E-06 2.27E-05 2.61 E-03 15 9.97E-01 2.76E-01 2.68E-41 1.27E-03 2.27E-03 2.25E-02 9.33E-05 3.OOE-04 2.74E-02 16 1.OOE+00 6.71 E-02 3.26E-02 4.06E-04 9.11 E-05 2.21 E-02 1.45E-06 1.65E-05 4.27E-05 17 9.72E-01 3.62E-01 3.37E-01 1.34E-03 2.37E-03 2.20E-02 9.90E-05 1.62E-04 8.57E-03 18 1.OOE+00 9.76E-02 6.25E-02 2.09E-02 4.67E-03 2.27E-02 7.45E-05 8.50E-04 2.12E-03 19 9.72E-01 4.03E-41 3.77E-01 6.87E-02 9.58E-03 2.26E-02 3.OOE-04 2.33E-03 1.20E-02 (-j E.1-51 Q-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.2.2.7 Release Magnitude Calculations The MAAP computer code is used to assign both the radionuclide release magnitude and timing based on the accident progression characterization.
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| Specifically, MAAP provides the following information:
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| * containment pressure and temperature versus time (time of containment failure is determined by comparing these values with the nominal containment capability);
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| * radionuclide release time and magnitude for a large number of radioisotopes; and* release fractions for twelve radionuclide species.E.1.3 IPEEE Analysis E.1.3.1 Seismic Analysis PNPS performed a seismic PRA following the guidance of NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (lPEEE) for Severe Accident Vulnerabilities, June 1991. The seismic PRA model was performed in conjunction with the SQUG program in 1994 as part of the IPEEE submittal report [Reference E.1-6]. The seismic, high wind, and external flooding analyses determined that the plant is adequately designed to protect against the effects of these natural events.A number of plant improvements were identified in Table 2.4 of NUREG-1 742, Perspectives Gained from the IPEEE Program, Final Report, April 2002 [Reference E.1 -8]. These improvements were implemented.
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| The seismic CDF in the IPEEE was conservatively estimated to be 5.82x10-5 per reactor-year.
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| The seismic CDF has recently been re-evaluated to reflect the updated Gothic computer code room heat up calculations that predict no room cooling requirements for HPCI, RCIC, Core Spray, and RHR areas; to update random component failure probabilities; and to model replacement of certain relays with a seismically rugged model. The updated seismic CDF of 3.22x10-5 per reactor-year was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.E.1.3.2 Fire Analysis The PNPS internal fire risk model was performed in 1994 as part of the IPEEE submittal report[Reference E.1-6]. The PNPS fire analysis was performed using the conservative EPRI's Fire Induced Vulnerability Evaluation (FIVE) methodology for qualitative and quantitative screening of fire areas and for fire analysis of areas that did not screen [Reference E.1 -71. The FIVE methodology is primarily a screening approach used to identify plant vulnerabilities due to fire initiating events.E.1-52 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage 0 Table E.1-12 presents the results of the PNPS IPEEE fire analysis.
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| The values presented in Table E.1-12 are taken from NUREG-1742
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| [Reference E.1-8]. These values are the same as the original IPEEE fire CDF results (2.20E-5 per reactor-year)
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| [Reference E.1-6] after the response to NRC questions/issues regarding fire-modeling progression.
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| A revised fire zone CDF of 1.91 E-5 per reactor-year, generated to reflect updated equipment failure probability and unavailability values was used in estimation of the factor of 6 used to determine the upper bound estimated benefit described in Section 4.21.5.4.The significant fire scenarios involve fires occurring in the train B switchgear room, turbine building heater bay, vital motor generator set room, and train A switchgear room.Table E.1-12 PNPS Fire Updated Core Damage Frequency Results Fire New Compartment Description CDF/year Estimate Sub-Area CDF/year 1E Reactor Building West, El. 21 9.7E-07 8.25E-07 2B Turbine Building Heater Bay 2.1 E-06 2.74E-06 3A Train B RBCCW/TBCCW Pump and Heat 2.0E-06 1.31 E-06 Exchanger Room 4A Train A RBCCW[TBCCW Pump and Heat 9.8E-07 2.95E-07 Exchanger Room 6 Control Room 1.6E-06 8.90E-07 7 Cable Spreading Room 9.5E-07 7.85E-07 9 Vital Motor Generator Set Room 2.4E-06 2.38E-06 12 Train A Switchgear Room 3.1E-06 2.30E-06 13 Train B Switchgear Room 6.1E-06 6.85E-06 26 Main Transformer 1.5E-06 7.60E-07 2.2E-05 1.91 E-05 I E.1-53 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.3.3 Other External Hazards The PNPS IPEEE submittal
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| [Reference E.1-6], in addition to the internal fires and seismic events, examined a number of other external hazards:* high winds and tornadoes;
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| * external flooding; and* ice, hazardous chemical, transportation, and nearby facility incidents.
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| In consequence of the above external hazards evaluation, no plant modifications were required for PNPS.No risks to the plant occasioned by high winds and tornadoes, external floods, Ice, and hazardous chemical, transportation, and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 104 6/year. Therefore, these other external event hazards are not included in this attachment and are expected not to impact the conclusions of this SAMA evaluation.
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| E.1.4 PSA Model Peer Review and Difference between Current PSA Model and 1995 Update IPE E.1.4.1 PSA Model Peer Review The original IPE PSA model was peer reviewed on March 2000 using the BWROG PSA Peer Review Certification Implementation Guidelines.
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| Facts and Observation sheets documented the certification teams' insights and potential level of significance.
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| As part of the update of the IPE PSA models, all major issues and observations from the BWROG Peer Review (i.e., Level A, B, C, and D observations) have been addressed and incorporated into the current IPE PSA model, April 2003 [Reference E.1-1].For the current IPE/PSA model update, individual work packages (event tree, fault tree, human reliability analysis (HRA), data, etc.) and internal flooding analysis were circulated to each PSA member for independent peer review. The accident sequence packages, system work packages, HRA, and internal flooding analyses were also assigned to the appropriate PNPS plant personnel for review. For example, event trees, system analyses, and fault tree models were forwarded to the applicable plant systems engineers and the HRA was assigned to individuals from the plant Operations Training department for review. Similarly, the accident sequence packages, system work packages, HRA report, containment performance analysis, fault tree and event tree models, and Level 2 models were peer reviewed by an outside consultant.
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| The Entergy license renewal project team and plant staff reviewed consequence and risk estimates for the SAMA analyses.E.1-54 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The peer review process emphasized the role of plant staff, external consultants, and BWROG PSA certification in this recent model update. The peer reviews served to ensure the accuracy of both the assumptions made in the models and the results. The results of the peer review and resolutions are presented in Section 5 and Appendix P of the Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events update report, April 2003 [Reference E.1-1].E.1.4.2 Major Differences between the Updated IPE PSA Model and 1995 Update IPE Model E.1.4.2.1 Core Damage -Comparison to the PNPS 1995 Update IPE Model The current PNPS IPE/PSA update model was completely revised in response to the BWROG Peer Review of March 2000 [Reference E.1-1]. The updated model is based upon all procedures and plant design as of September 30, 2001, and plant data as of December 31, 2001. The results yield a measurably lower CDF (point estimate CDF -6.41 E-6/reactor year) than the original IPE (point estimate CDF -5.85E-5/yr)
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| [Reference E.1-2] and 1995 PSA model update (point estimate CDF -2.84E-5/yr)
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| [Reference E.1-31. (The 1995 update was performed to answer NRC questions following the IPE submittal.)
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| The improved results are due to improved plant performance, replacement of switchyard -breakers, more realistic success criteria based on MAAP runs, and more sophisticated data handling.
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| Major changes are summarized as follows.A. Initiating Event The initiating event frequencies were updated to include current plant data and recent NRC publication information.
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| For example, the LOOP frequency decreased significantly from the original IPE frequency of 0.475/yr to the current value of 0.067/yr [Reference E.1-1], which reflects the decreased occurrence of LOOP events since 1990 and replacement of switchyard breakers.
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| In addition, fault tree models were developed to calculate support system initiating event frequencies.
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| B. Accident Sequence Evaluation Event trees from the original IPE were completely revised. BWROG certification findings and observations were incorporated into the revised event trees. Major facts and observations include the following.
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| (1) LOOP Event Tree The LOOP event was completely revised to account for failure modes of HPCI/RCIC beyond 8 hours of operation; RPV depressurization on HCTL; and transfer to the SBO tree to address such items as premature battery depletion and AC recovery at 30 minutes and beyond.E.1-55 C>
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| Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage (2) SBO Event Tree Current update reflects GE load shed calculations and use of plant SBO procedures for DC load shedding.(3) Inadvertent Stuck Open Relief Valve (IORV) Event Tree The IORV event tree was modified to include RPV depressurization with two SRVs given high-pressure injection failure.(4) LOCAs Event Trees The update considers both HPCI and RCIC for small break LOCAs.Large and medium LOCAs and subsequent ATWS are modeled as core damage end states in the updated model. Small break LOCAs and ATWS are treated as similar to transient-induced ATWS.The vapor suppression system is considered during large LOCAs events.(5) ATWS Event Tree The revised ATWS tree reflects the potential for MSIV closure on low RPV level.The revised ATWS model takes into consideration "inhibit ADS" and MSIV bypass issues. In addition, HRA values take into consideration ATWS accident progressions for RPV and containment conditions predicted by MAAR (6) Loss-of-Containment Heat Removal Sequences The revised event trees model the potential impact from containment venting on low-pressure system operation.
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| For example, no credit is given for core spray and LPCI if containment venting is required.
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| In addition, other containment related phenomena, such as high torus temperatures (HPCI) and high containment pressures (RCIC, SRVs)are reflected in the updated event trees.The update model only considers the DTV path for containment venting.(7) ISLOCA Event Tree NSAC-154 [Reference E.1-10] and NUREG/CR-5124
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| [Reference E.1-11] were used to reassess the ISLOCA analysis.Success criteria for low-pressure injection during an ISLOCA are consistent with those used for small LOCAs.E.1-56 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage The revised ISLOCA event tree credits use of condensate or fire water for large ISLOCA events provided that LPCI or core spray operation had previously occurred to provide initial RPV reflood.(8) Other Changes The revised event trees credit use of feedwater when appropriate.
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| Control Rod Drive system flow into the RPV is credited for sequences that involve loss of containment heat removal and subsequent requirement to control containment pressure with direct torus containment venting.Consistent success criteria were employed for RPV depressurization for transients, medium LOCAs, and small LOCAs.The revised PNPS IPE models are based on the BWROG EPGs/SAGs Revision 4 of the BWROG EPGs [Reference E.1-1].Core damage definition has been revised to be consistent with the EPRI PSA Applications Guide [Reference E.1-12]. That is, core damage occurs when peak clad temperature exceeds 2200 0 F.HPCI and RCIC use is based on a 24-hour mission time. C C. Thermal -Hydraulic (T-H) Analysis T-H analysis has been completely revised and improved to better support the success criteria.The MAAP4 computer code [Reference E.1-4] was used to update and address the many issues raised by the BWROG certification team, such as the following.
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| * A basis was provided for the timing and discharge pressure (flow) adequacy when using the fire water system for successful mitigation during transients and small LOCAs.* Success criteria for SORV are same as for non-SORV cases (2 SRVs are required for successful RPV depressurization).
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| * Consistent success criteria are used for RPV depressurization for transients, medium LOCAs, and small LOCAs.* Plant specific calculations were performed to identify the plant response for single or double recirculation pump trip failures.* The appropriateness of the core damage definition used in the update was verified.E.1-57 Pilgrim Nuclear Power Station Applicant's Environmental Report Hi'" Operating License Renewal Stage In addition to the MAAP4 code, the GOTHIC code [Reference E.1-13] was used to predict various room heatup rates for the reactor building, turbine building, switchgear room, and battery room.D. System Analysis System fault tree models from the original IPE were completely revised to reflect the as-built plant configuration.
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| MAAP analyses were clearly identified to support the success criteria of these Level 1 models. More detailed modeling for the logic interlock was included in the system models. A detailed fault tree for the RPS was developed based on NUREG/CR-5500
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| [Reference E. 1-9], which decreased the failure-to-scram probability from 3.OE-5/yr to 5.8E-6/yr.
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| E Data Analysis Component failure data, both generic and plant-specific, were reviewed and updated with more recent experience (the performance of risk significant systems HPCI and RCIC has greatly improved since the original IPE). Plant-specific data were adjusted for industry experience using Bayesian updates. Maintenance unavailability values were updated based on maintenance rule records from the system engineers.
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| More recent common cause failure data and approach NUREG/CR-5497
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| [Reference E.1-14] were factored into this update. In particular, a more detailed and refined common-cause failure methodology (Alpha model) has been applied in this update. In addition, more common-cause equipment failure groups such as fans, dampers,/4gw transformers, DC power panels, and circuit breakers have been included in the analysis.F. HA A complete revision of the HRA was performed to identify, quantify, and document the pre-initiator and post-initiator human errors (including recoveries).
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| The updated HRA was performed using NUREG/CR-1278
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| [Reference E.1-15], also referred to as THERP. Screening values were only used for low-significance human errors. In addition, a detailed analysis was performed to treat dependencies between post-initiator errors.G Dependencv Analysis A complete revision of the internal flooding analysis was developed to systematically address spatial dependencies.
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| Dependency between pre-initiator human errors (such as miscalibration of instruments) was modeled. In addition, dependencies between multiple post-accident operator actions appearing in the same accident sequence were evaluated.
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| -Detailed component dependency tables were developed to address the support systems associated with the modeled systems and components.
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| E.1-58 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage H. Structural Response The ISLOCA frequency was revised.RPV overpressure and capability of the reactor building were included in the Level 2 assessment.
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| : 1. Quantification The truncation value was lowered to I.OE-11.Human Error Probability (HEP) dependencies and recovery actions in the cutsets were evaluated.
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| ATWS contribution decreased due to lower probability of failure to scram based on NUREG/CR-5500 [Reference E.1-9].The HRA was completely revised to address a comment from the PSA Certification
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| [Reference E.1-16] that many of the HEPs were not realistic using the previous methodology.
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| In many cases (e.g., failure to perform DTV), the previous HEPs were judged to be overly conservative.
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| J. Internal Flooding Analysis The internal flooding analysis from the original IPE was completely revised to include a detailed, systematic examination of the flood source and progression for each of the analyzed flooding scenarios.
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| In addition, the updated internal flooding analysis considers the effects of spray on equipment.
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| K. Uncertainty Analysis An uncertainty analysis was performed for this update.E.1.4.2.2 Containment Performance
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| -Comparison to the Original PNPS IPE Model Containment performance analysis models were completely revised from the original IPE.Propagation of Level 1 cutsets to the Level 2 CET was developed.
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| A detailed LERF model was developed to ensure that LERF calculations are consistent with the PSA Applications Guide and NRC requirements for RG 1.174 [Reference E.1-17]. Other salient items incorporated are the following.
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| * CET fault models were revised to ensure that mitigating systems were not degraded in the Level I sequence.* CET fault tree models allowed credit for AC power recovery post core damage. This ensures that the models do not allow SBO core damage sequences to benefit from AC supported equipment in Level 2 without explicit consideration of AC power recovery.E.1-59 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* Shell melt-through phenomena were considered where applicable.
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| * Operator responses to key actions were reassessed to incorporate the probability for success given the containment conditions and Emergency Operating Procedure directions.
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| * Direct torus venting was considered post core damage.* PNPS-specific primary containment structural evaluation was included in the CET. This also included a structural evaluation of torus failure due to dynamic loading during ATWS scenarios, torus break below the water line, and bellows seal capability.
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| * A reactor building bypass fault tree model was developed to assess the impact on the Level 2 analysis.E.1.5 The MACCS2 Model -Level 3 Analysis E.1.5.1 Introduction SAMA evaluation relies on Level 3 PRA results to measure the effects of potential plant modifications.
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| A Level 3 PRA model using the MACCS2 [Reference E.1-18] was created for PNPS. This model, which requires detailed site-specific meteorological, population, and economic data, estimates the consequences in terms of population dose and offsite economic cost. Risks in terms of population dose risk (PDR) and offsite economic cost risk (OECR) were also estimated in this analysis.
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| Risk is defined as the product of consequence and frequency of an accidental release.This analysis considers a base case and two sensitivity cases to account for variations in data and assumptions for postulated internal events. The base case uses estimated time and speed for evacuation.
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| Sensitivity case 1 is the base case with delayed evacuation.
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| Sensitivity case 2 is the base case with lower evacuation speed.PDR was estimated by summing over all releases the product of population dose and frequency for each accidental release. Similarly, OECR was estimated by summing over all releases the product of offsite economic cost and frequency for each accidental release. Offsite economic cost includes costs that could be incurred during the emergency response phase and costs that could be incurred through long-term protective actions.E.1.5.2 Input The following sections describe the site-specific ninput parameters used to obtain the off-site dose and economic impacts for cost-benefit analyses, E.1.60 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.5.2.1 Projected Total Population by Spatial Element The total population within a 50-mile radius of PNPS was estimated for the year 2032, the end of the proposed license renewal period, for each spatial element by combining total resident population projections with transient population data obtained from Massachusetts and Rhode Island. Table E.1-13 shows the estimated population distribution.
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| Table E.1-13 Estimated Population Distribution within a 50-mile Radius 0-10 10-20 20-30 30-40 40-50 50-Mile Sector Miles Miles Miles Miles Miles Total N 0 0 0 0 80474 80474 NNE 3 0 0 0 0 3 NE 3 0 0 0 0 3 ENE 3 0 33121 0 0 33124 E 5 0 33121 23185 0 56311 ESE 23 0 49682 92740 0 142445 SE 950 9936 115925 23185 0 149996 SSE 13289 69555 82803 0 0 165647 S 23695 99364 132485 84383 43397 383324 SSW 23695 49762 23696 23185 21699 142037 SW 23695 71088 277374 349491 114546 836194 WSW 23695 71088 277374 349491, 183037 904685 W 22818 71088 277374 388324 286370 1045974 WNW 16494 71088 118481 303450 390150 899663 NW 11269 71088 195075 1529212 405561 2212205 NNW 5599 35544 43350 31295 321894 437682 Total 165236 619601 1659861 3197941 1847128 7489767 Q ., The 2000 U.S. Census Bureau data, togetherwith Massachusetts and Rhode Island population projection data, was used to project county-level resident populations to the year 2032.Seasonal peak transient population was conservatively used to establish a transient/resident population ratio for each county within the 50-mile radius. The ratio was found to be decreasing over time. For purposes of this study, the total county level population values were estimated by E.1-61 (C-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage summing the year 2000 peak transient population of each county and the projected year 2032 permanent population of that county to obtain the 2032 total county population.
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| E.1.5.2.2 Land Fraction The land fraction for each spatial element was estimated from the PNPS Emergency Planning Zone maps for radii of 2, 5, and 50 miles [Reference E.1-20].E.1.5.2.3 Watershed Class There are two watershed types in the 50-mile zone surrounding PNPS: ocean and land (watersheds) drained by rivers. There are no major lakes. The watershed index assigns "`" to any spatial element having a non-zero land fraction and "2" to all elements over the Atlantic Ocean or its bays.E.1.5.2.4 Regional Economic Data RegaLon Index Each spatial element was assigned to an economic region, defined in this report as a county. Where a spatial element covers portions of more than one county, it was assigned to that county having the most area within the element.Regional Economic Data County level economic data were obtained from the U.S. Department of Agriculture.
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| The Census of Agriculture is conducted every five years and data from 1997 and 1992 were used to project the farm-related economic data for 2002.VALWF -Value of Fain, Wealth MACCS2 requires an average value of farm wealth (dollars/hectare) for the 50-mile radius area around PNPS. The county-level farmland property value was used as a basis for deriving this value. VALWF is $23,578/hectare.
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| VALWNF- Value of Non-Farm Wealth MACCS2 also requires an average value of non-farm wealth. The county-level non-farm property value was used as a basis for deriving this value. VALWNF is $189,041/person.Other economic parameters and their values are shown below. The values were obtained by adjusting the economic data from a past census given as default values in Reference E.1-18 with the consumer price index of 177.1, which is the average value for the year 2001, as appropriate.
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| E.1-62 Pilgrim Nuclear Power Station Applicant's Environmental Report; Operating License Renewal Stage Variable Description Value EVACST Daily cost for a person who has been evacuated 42.3 ($/person-day)
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| POPCST Population relocation cost ($/person) 7840 RELCST Daily cost for a person who is relocated
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| ($/person-day) 42.3 CDFRMO Cost of farm decontamination for the various levels of 881 decontamination
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| ($/hectare) 1959 CDNFRM Cost of non-farm decontamination for the various levels of 4700 decontamination
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| ($/person) 12540 DLBCST Average cost of decontamination labor ($/person-year) 54800 DPRATE Property depreciation rate (per year) 0.2 DSRATE Investment rate of return (per year) 0.12 E.1.5.2.5 Agriculture Data The source of regional crop information is the New England Agricultural Statistics, 2001. The crops listed for each of the two states, Massachusetts and Rhode Island, were mapped into the seven MACCS2 crop categories.
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| E.1.5.2.6 Meteorological Data The MACCS2 model requires meteorological data for wind speed, wind'direction, atmospheric stability, accumulated precipitation, and atmospheric mixing heights. The required data was obtained from the PNPS site meteorological monitoring system and the Automated Surface Observatory System (ASOS) at Plymouth Airport.Site Specific Data Site specific meteorological data is available from two meteorological towers, one located off the main parking lot and 'the second located west of the old l&S building, the"lower" and "upper' towers respectively.
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| The upper tower is the designated data source for MACCS2 input. Data from the lower tower was'used only if measurements from the upper tower were missing for a specific hour.Year 2001 hourly data from the upper tower was used in this analysis.
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| The data was more than 98% complete.
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| Missing data was obtained either from the lower tower or from estimates based on adjacent valid measurements of the missing hour.(. !E.1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Accumulated Precipitation The nearest source of hourly precipitation data to PNPS is the ASOS at Plymouth Airport. The data was converted to MACCS2 input format to provide precipitation in hundredths of an inch.Regional Mixing Height Data Mixing height is defined as the height of the atmosphere above ground level within which a released contaminant will become mixed (from turbulence) within approximately one hour. PNPS mixing height data, given in Reference E.1-19, was used for MACCS2 analysis.E.1.5.2.7 Emergency Response Assumptions Details of the evacuation time estimates including supporting assumptions regarding population, alarm criteria, delay times, areas, speed, distance, and routes are contained in the PNPS Emergency Plan [Reference E.1-20].Evacuation Delay Time The elapsed time between siren alert and the beginning of evacuation is 40 minutes. A sensitivity case that assumes 2 hours for evacuees to begin evacuation was considered in this study to evaluate consequence sensitivities due to uncertainties in delay time.Evacuation Speed The worst case for PNPS evacuation is during the winter, under adverse weather conditions, since snow removal can add up to an hour and a half to the evacuation time.The radius of the Emergency Planning Zone is 10 miles. Assuming that the net movement of the entire population is 10 miles, the time required for evacuation ranges from 3 hours 35 minutes to 6 hours 30 minutes, and the average evacuation speed ranges from 2.79 miles/hour in clear weather to 1.54 miles/hour under adverse weather conditions.
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| The average evacuation speed is 2.17 miles/hour, or 0.97 meter/second.
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| A sensitivity case that assumes a lower evacuation speed of 0.69 meter/second was considered in this study to evaluate consequence sensitivities due to uncertainties in evacuation speed.E.1.5.2.8 Core Inventory The estimated PNPS core inventory (Table E.1-14) used in the MACCS2 input is based on a power level of 2028 MW(t).E.1-64 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-14 PNPS Core Inventory (Becquerels)
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| Nuclide Inventory Nuclide Inventory Co-58 1.15E+16 Te-131m 2.87E+17 Co-60 1.37E+16 Te-132 2.80E+18 Kr-85 1.88E+16 1-131 1.94E+18 Kr-85m 6.84E+17 1-132 2.85E+18 Kr-87 1.24E+18 1-133 4.07E+18 Kr-88 1.68E+18 1-134 4.45E+18 Rb-86 1.05E+15 1-135 3.83E+18 Sr-89 2.08E+18 Xe-133 4.07E+18 Sr-90 1.47E+17 Xe-135 9.68E+17 Sr-91 2.71E+18 Cs-134 3.17E+17 Sr-92 2.83E+18 Cs-136 8.51E+16 Y-90 1.58E+17 Cs-137 1.90E+17 Y-91 2.54E+18 Ba-139 3.75E+18 Y-92 2.84E+18 Ba-140 3.70E+18 Y-93 3.23E+18 La-140 3.77E+18 Zr-95 3.34E+18 La-141 3.48E+18 Zr-97 3.44E+18 La-142 3.35E+18 Nb-95 3.16E+18 Ce-141 3.36E+18 Mo-99 3.65E+18 Ce-143 3.27E+18 Tc-99m 3.15E+18 Ce-144 2.18E+18 Ru-103 2.77E+18 Pr-143 3.20E+18 Ru-105 1.85E+18 Nd-147 1.43E+18 Ru-106 7.52E+17 Np-239 4.26E+19 Rh-105- 1.38E+18 Pu-238 2.96E+15 Sb-127 1.74E+17 Pu-239 7.51E+14 Sb-129 6.06E+17 Pu-240 9.41 E+14 Te-127 1.69E+17 Pu-241 1.62E+17 Te-127m 2.27E+16 Am-241 1.65E+14 Te-129 5.68E+17 Cm-242 4.35E+16 Te-129m 1.49E+17 Cm-244 2.35E+15 Source: derived from Reference E.1-21 for a power level of 2028 MW(t)U E.1-65 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.5.2.9 Source Terms Twelve release categories, corresponding to internal event sequences, were part of the MACCS2 input. Details of the source terms for postulated internal events are available in on-site documentation.
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| A linear release rate was assumed between the time the release started and the time the release ended.E.1.5.3 Results Risk estimates for one base case and two sensitivity cases were analyzed with MACCS2. The base case assumes 40 minute delay and 0.97 meter/sec speed of evacuation.
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| Sensitivity case I is the base case with delayed evacuation of 2 hours. Sensitivity case 2 is the base case with an evacuation speed of 0.69 meter/sec.
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| Table E. 1-15 shows estimated base case mean risk values for each release mode. The estimated mean values of PDR and offsite OECR for PNPS are 13.6 person-rem/yr and$45,900/yr, respectively.
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| E.1-66 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.1-15 Base Case Mean PDR and OECR Values Pplto Offsite Offsite Release Frequency Popsin Economic Population Dose Economic Cost Modoser CstRisk (PDR) Rik(E )(person-sv)
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| Cs (person-rem/yr) (S/yr)CAPB-1 9.51 E-08 4.66E-01 3.82E+06 4.43E-06 2 3.63E-01 CAPB-2 1.27E-08 9.96E+01 6.40E+06 1.26E-04 8.1OE-02 CAPB-3 2.39E-09 1.06E+02 6.48E+06 2.53E-05 1 .55E-02 CAPB-4 3.29E-09 1.38E+04 4.28E+09 4.54E-03 1.41E+01 CAPB-5 2.73E-09 1.81E+04 5.30E+09 4.94E-03 1.45E+01 CAPB-6 7.95E-09 1.51 E+04 3.51 E+09 1.20E-02 2.79E+01 CAPB-7 7.93E-09 1.67E+04 4.42E+09 1.32E-02 3.51 E+01 CAPB-8 2.06E-08 4.1OE+04 1.47E+10 8.44E-02 3.03E+02 CAPB-9 9.25E-09 2.37E+04 8.33E+09 2.19E-02 7.70E+01 CAPB-10 8.53E-08 4.31 E+04 1.54E+10 3.68E-01 1.31 E+03 CAPB-11 4.35E-08 3.45E+04 1.15E+10 1.50E-01 5.OOE+02 CAPB-12 1.70E-06 9.72E+01 4.63E+06 1.65E-02 7.88E+OO CAPB-13 2.30E-09 7.30E+03 6.53E+08 1.68E-03 1.50E+OO CAPB-14 2.26E-06 1.58E+04 4.14E+09 3.57E+00 9.36E+03 CAPB-15 2.12E-06 4.31 E+04 1.59E+10 9.14E+00 3.37E+04 CAPB-16 1.18E-09 1.86E+04 5.50E+09 2.19E-03 6.48E+OO CAPB-17 6.91E-09 4.81E+04 1.71E+10 3.32E-02 1.18E+02 CAPB-18 4.61E-10 2.38E+04 7.86E+09 1.1OE-03 3.62E+OO CAPB-19 2.43E-08 5.31 E+04 1.88E+10 1.29E-01 4.56E+02 Totals 1.36E+01 4.59E+04 1. 1 sv= 100 rem 2. 4.43E-06 (person-rem/yr)
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| = 9.51 E-08 (/yr) x 4.66E-01 (person-sv) x 100 (remlsv)E.1-67 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Results of sensitivity analyses indicate that a delayed evacuation or a lower evacuation speed would not have significant effects on the offsite consequences or risks determined in this study.Table E.1-16 summarizes offsite consequences in terms of population dose (person-sv) and offsite economic cost ($) for the base case and the sensitivity cases. Comparison of the consequences indicates that the maximal deviation is less than 2% between the base case population dose and the Sensitivity Case 2 population dose for release mode CAPB-8.Table E.1-16 Summary of Offsite Consequence Sensitivity Results Population Dose (person-sv)
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| Offsite Economic Cost ($)Release 2-Hr Lower 2-Hr Lower Base Case Delayed Speed of Base Case Delayed Speed of Mode Evacuation Evacuation Evacuation Evacuation CAPB-1 4.66E-01 4.66E-01 4.67E-01 3.82E+06 3.82E+06 3.82E+06 CAPB-2 9.96E+01 9.97E+01 9.97E+01 6.40E+06 6.40E+06 6.40E+06 CAPB-3 1.06E+02 1.06E+02 1.06E+02 6.48E+06 6.48E+06 6.48E+06 CAPB-4 1.38E+04 1.39E+04 1.39E+04 4.28E+09 4.28E+09 4.28E+09 CAPB-5 1.81 E+04 1.82E+04 1.82E+04 5.30E+09 5.30E+09 5.30E+09 CAPB-6 1.51E+04 1.51E+04 1.51E+04 3.51E+09 3.51E+09 3.51E+09 CAPB-7 1.67E+04 1 .68E+04 1.68E+04 4.42E+09 4.42E+09 4.42E+09 CAPB-8 4.1OE+04 4.16E+04 4.17E+04 1.47E+10 1.47E+10 1.47E+10 CAPB-9 2.37E+04 2.38E+04 2.39E+04 8.33E+09 8.33E+09 8.33E+09 CAPB-1D 4.31E+04 4.34E+04 4.36E+04 1.54E+1o 1.54E+10 1.54E+10 CAPB-11 3.45E+04 3.48E+04 3.49E+04 1.15E+10 1.15E+10 1.15E+10 CAPB-12 9.72E+01 9.75E+01 9.78E+01 4.63E+06 4.63E+06 4.63E+06 CAPB-13 7.30E+03 7.30E+03 7.31E+03 6.53E+08.
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| 6.53E+08 6.53E+08 CAPB-14 1.58E+04 1.58E+04 1.59E+04 4.14E+09 4.14E+09 4.14E+09 CAPB-15 4.31E+04 4.33E+04 4.35E+04 1.59E+10 1.59E+10 1.59E+10 CAPB-16 1.86E+04 1.87E+04-1.88E+04 5.50E+09 5.50E+09 5.50E+09 CAPB-17 4.81E+04 4.83E+04 4.86E+04 1.71E+10 1.71E+10 1.71E+10 CAPB-18 2.38E+04 2.39E4-04 2.40E+04 7.86E+09 7.86E+09 7.86E+09 CAPB-19 5.31E+04 5.33E+04 5.37E+04 1.88E+10 1.88E+10 1.88E+10 E.1-68 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1.6 References E.1-1 ENN Engineering Report PNPS-PSA, "Pilgrim Nuclear Power Station Individual Plant Examination for Internal Events Update," April 2003, Revision 1.E.1-2 Pilgrim Nuclear Power Station Individual Plant Examination, Revision 0, September 1992.E.1-3 Boston Edison Company to the NRC, Response to Request for Additional Information Regarding the Pilgrim Individual Plant Examination (IPE) Submittal (TAC No. M74451, letter dated December 28, 1995 (2.95.127).
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| E.1-4 Modular Accident Analysis Program Boiling Water Reactor (MAAP BWR) Code, Version 4.0.4 and Fauske & Associates, Inc., "MAAP 4.0 Users manual," prepared for The Electric Power Research Institute, May 1994.E.1-5 Kaiser, "The Implications of Reduced Source Terms for Ex-Plant Consequence Modeling," Executive Conference on the Ramifications of the Source Term (Charleston, SC), March 12, 1985.E.1-6 "Pilgrim Nuclear Power Station Individual Plant Examination for External Events," July 1994, Revision 0.E.1-7 Parkinson, W. J., "EPRI Fire PRA Implementation Guide", prepared by Science Applications International Corporation for Electric Power Research Institute, EPRI TR-105928, December 1995.E.1-8 U.S. Nuclear Regulatory Commission, NUREG-1 742, Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volume 1, Final Report, April 2002.E.1-9 U.S. Nuclear Regulatory Commission, NUREG/CR-5500, Vol. 3, (INEEUEXT 00740), Reliability Study: General Electric Reactor Protection System, 1984-1995, May 1999.E.1-10 Electric Power Research Institute, NSAC-154, "ISLOCA Evaluation Guidelines," prepared by ERIN Engineering and Research, Inc., September 1991.E.1-11 Chu, et al., "Interfacing Systems LOCA: Boiling Water Reactors," Brookhaven National Laboratory, NUREG/CR-5124, BNL-NUREG-52141, February 1989.E.1-12 Electric Power Research Institute, "PSA Applications Guide," EPRI TR-1 05396, prepared by ERIN Engineering and Research, Inc., August, 1995.E.1-13 GOTHIC Containment Analysis Package, Version 3.4e, EPRI Tr-103053-V2, October 1993.E.1-69 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.1-14 U.S. Nuclear Regulatory Commission, NUREG/CR-5497, (INEEUEXT-97-01328), Common-Cause Failure Parameter Estimations, October 1998.E.1-15 Swain, A. D. and H. E. Guttmann, NUREG/CR-1 278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, Sandia National Laboratories, U.S. Nuclear Regulatory Commission, August 1983.E.1-16 BWR Owners Group, "Pilgrim PSA Certification," BWROG/PSA-9903, March 2000.E.1-17 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174 (draft was issued as DG-1061), "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998.E.1-18 Chanin, D. I., and M. L. Young, Code Manual for MACCS2: Volume 1, User's Guide, SAND97-0594 Sandia National Laboratories, Albuquerque, NM, 1997.E.1-19 Boston Edison Company, "Appendix I Evaluation," forwarding evaluation of Pilgrim Station Unit 1 Conformance to the Design Objectives of 10 CFR 50, Appendix I, letter dated March 31, 1977 (2.77.031).
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| E.1-20 PNPS Emergency Plan, Revision 24, February 7, 2001, Appendix 5, Pilgrim Station Evacuation Time Estimates and Traffic Management Plan Update, Revision 5, November 1998.E.1-21 U.S. Nuclear Regulatory Commission, NUREG/CR-4551, Vol. 2, Rev. 1, Part 7, Evaluation of Severe Accident Risks: Quantification of Major Input Parameters, MACCS Input, December 1990.E.1-70 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage ATTACHMENT E.2 SAMA CANDIDATES SCREENING AND EVALUATION QWl Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2 EVALUATION OF SAMA CANDIDATES This section describes the generation of the initial list of potential SAMA candidates, screening methods, and the analysis of the remaining SAMA candidates.
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| E.2.1 SAMA List Compilation A list of SAMA candidates was developed by reviewing industry documents and considering plant-specific enhancements not identified in published industry documents.
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| Since PNPS is a conventional GE nuclear power reactor design, considerable attention was paid to the SAMA candidates from SAMA analyses for other GE plants. Industry documents reviewed include the following:
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| * Hatch SAMA Analysis (Reference E.2-1),* Calvert Cliffs Nuclear Power Plant SAMA Analysis (Reference E.2-2),* GE ABWR SAMDA Analysis (Reference E.2-3),* Peach Bottom SAMA Analysis (Reference E.2-4),* Quad Cities SAMA Analysis (Reference E.2-5),* Dresden SAMA Analysis (Reference E.2-6), and* Arkansas Nuclear Unit 2 SAMA Evaluation (Reference E.2-7).The above documents represent a compilation of most SAMA candidates developed from the industry documents.
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| These sources of other industry documents include the following:
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| * Limerick SAMDA cost estimate report (Reference E.2-8),* NUREG-1437 description of Limerick SAMDA (Reference E.2-9),* NUREG-1437 description of Comanche Peak SAMDA (Reference E.2-1 0),* Watts Bar SAMDA submittal (Reference E.2-11),* TVA's response to NRC's RAI on the Watts Bar SAMDA submittal (Reference E.2-12),* Westinghouse AP600 SAMDA (Reference E.2-13),* NUREG-0498, Watts Bar Final Environmental Statement Supplement 1, Section 7 (Reference E.2-14),* NUREG-1 560, Volume 2, NRC Perspectives on the IPE Program (Reference E.2-15), and* NUREG/CR-5474, Assessment of Candidate Accident Management Strategies (Reference E.2-16).E.2-1 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage In addition to SAMA candidates from review of industry documents, additional SAMA candidates were obtained from plant-specific sources, such as the PNPS IPE (Reference E.2-17) and IPEEE (Reference E.2-18). In both the IPE and IPEEE, several enhancements related to severe accident insights were recommended and implemented.
| |
| These enhancements are included in the comprehensive list of phase I SAMA candidates as numbers 248 through 281. The current PNPS PSA model was also used to identify plant-specific modifications for inclusion in the comprehensive list of SAMA candidates.
| |
| The risk-significant terms from the current PSA model were reviewed for similar failure modes and effects that could be addressed through a potential enhancement to the plant. The correlation between SAMAs and the risk-significant terms were listed in Table E.1-2.The comprehensive list, available in on-site documentation, contained a total of 281 phase I SAMA candidates.
| |
| E.2.2 Qualitative Screenina of SAMA Candidates (Phase I)The purpose of the preliminary SAMA screening was to eliminate from further consideration enhancements that were not viable for implementation at PNPS. Potential SAMA candidates were screened out if they modified features not applicable to PNPS, if they had already been implemented at PNPS, or if they were similar in nature and could be combined with another SAMA candidate to develop a more comprehensive or plant-specific SAMA candidate.
| |
| During this process, 63 of the phase I SAMA candidates were screened out because they were not applicable to PNPS, 4 of the phase I SAMA candidates were screened out because they were similar in nature and could be combined with another SAMA candidate, and 155 of the phase I SAMA candidates were screened out because they had already been implemented at PNPS, leaving 59 SAMA candidates for further analysis.
| |
| The final screening process involved identifying and eliminating those items whose implementation cost would exceed their benefit as described below. Table E.2-1 provides a description of each of the 59 phase 11 SAMA candidates.
| |
| E.2.3 Final Screening and Cost Benefit Evaluation of SAMA Candidates (Phase II)A cost/benefit analysis was performed on each of the remaining SAMA candidates.
| |
| If the implementation cost of a SAMA candidate was determined to be greater than the potential benefit (i.e. there was a negative net value) the SAMA candidate was considered not to be cost beneficial and was not retained as a potential enhancement.
| |
| The expected cost of implementation of each SAMA was established from existing estimates of similar modifications.
| |
| Most of the cost estimates were developed from similar modifications considered in previously performed SAMA and SAMDA analyses.
| |
| In particular, these cost-estimates were derived from the following major sources:* GE ABWR SAMDA Analysis (Reference E.2-3),* Peach Bottom SAMA Analysis (Reference E.2-4), E.2-2 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage* Quad Cities SAMA Analysis (Reference E.2-5),* Dresden SAMA Analysis (Reference E.2-6),* ANO-2 SAMA Analysis (Reference E.2-7), and The cost estimates did not include the cost of replacement power during extended outages required to implement the modifications, nor did they include contingency costs associated with unforeseen implementation obstacles.
| |
| Estimates based on modifications that were implemented or estimated in the past were presented in terms of dollar values at the time of implementation (or estimation), and were not adjusted to present-day dollars. In addition, several implementation costs were originally developed for SAMDA analyses (i.e., during the design phase of the plant), and therefore, do not capture the additional costs associated with performing design modifications to existing plants (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.). Therefore, the cost estimates were conservative.
| |
| The benefit of implementing a SAMA candidate was estimated in terms of averted consequences.
| |
| The benefit was estimated by calculating the arithmetic difference between the total estimated costs associated with the four impact areas for the baseline plant design and the total estimated impact area costs for the enhanced plant design (following implementation of the SAMA candidate).
| |
| Values for avoided public and occupational health risk were converted to a monetary equivalent (dollars) via application of the NUREG/BR-0184 (Reference E.2-19) conversion factor of $2,000 per person rem and discounted to present value. Values for avoided off-site economic costs were also discounted to present value.As this analysis focuses on establishing the economic viability of potential plant enhancement when compared to attainable benefit, detailed cost estimates often were not required to make informed decisions regarding the economic viability of a particular modification.
| |
| Several of the SAMA candidates were clearly in excess of the attainable benefit estimated from a particular analysis case.For less clear cases, engineering judgment on the cost associated with procedural changes, engineering analysis, testing, training, and hardware modification was applied to determine if a more detailed cost estimate was necessary to formulate a conclusion regarding the economic viability of a particular SAMA. Based on a review of previous submittals' SAMA evaluations and an evaluation of expected implementation costs at PNPS, the following estimated costs for each potential element of the proposed SAMA implementation are used.E.2-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Type of Change Estimated Cost Range Procedural only $25K-$50K Procedural change with engineering
| |
| $50K-$200K required Procedural change with engineering and $200K-$300K testing/training required Hardware modification
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| $1OOK to >$1OOOK In most cases, more detailed cost estimates were not required, particularly if the SAMA called for the implementation of a hardware modification.
| |
| Nonetheless, the cost of each unscreened SAMA candidate was conceptually estimated to the point where conclusions regarding the economic viability of the proposed modification could be adequately gauged. The cost benefit comparison and disposition of each of the 59 phase 11 SAMA candidates is presented in Table E.2-1.Bounding evaluations (or analysis cases) were performed to address specific SAMA candidates or groups of similar SAMA candidates.
| |
| These analysis cases overestimated the benefit and thus were conservative calculations.
| |
| For example, one SAMA candidate suggested installing a digital large break LOCA protection system. The bounding calculation estimated the benefit of this improvement by total elimination of risk due to large break LOCA (see analysis in phase 11 SAMA 052 of Table E.2-1). This calculation obviously overestimated the benefit, but if the inflated benefit indicated that the SAMA candidate was not cost beneficial, then the purpose of the analysis was satisfied.
| |
| A description of the analysis cases used in the evaluation follows.Decay Heat Removal Capability
| |
| -Torus Cooling This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal system. Enhancements of decay heat removal capability decrease the probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of the torus cooling mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $261,832.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 1 and 14.Decay Heat Removal Capability
| |
| -Drywell Sp=ra This analysis case was used to evaluate the change in plant risk from installing an additional decay heat removal system. Enhancements of decay heat removal capability decrease the E.2-4 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage probability of loss of containment heat removal. A bounding analysis was performed by setting the events for loss of the drywell spray mode of the RHR system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $264,219.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 9.Filtered Vent This analysis case was used to evaluate the change in plant risk from installing a filtered containment vent to provide fission product scrubbing.
| |
| A bounding analysis was performed by reducing the successful torus venting accident progression source terms by a factor of 2 to reflect the additional filtered capability.
| |
| Reducing the releases from the vent path resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMAs 2 and 19.Containment Vent for ATWS Decay Heat Removal This analysis case was used to evaluate the change in plant risk from installing a containment vent to provide alternate decay heat removal capability during an ATWS event. A bounding analysis was performed by setting the ATWS sequences associated with containment bypass to zero in the level I PSA model, which resulted in an upper bound benefit of approximately
| |
| $61,701. This analysis case was used to model the benefit of phase 11 SAMAs 3 and 47.Molten Core Debris Removal This analysis case was used to estimate the change in plant risk from providing a molten core debris cooling mechanism.
| |
| A bounding analysis was performed by setting containment failure due to core-concrete interaction (not including liner failure) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $2,620,551.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 4, 5, 8, and 23.Dryweff Head Flooding This analysis case was used to evaluate the change in plant risk from providing a modification to flood the drywell head such that if high drywell temperature occurred, the drywell head seal would not fail. A bounding analysis was performed by setting the probability of drywell head failure due to high temperature to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $12,915. This analysis case was used to model the benefit of phase 11 SAMAs 6,18, and 20.Reactor Building Effectiveness This analysis case was used to evaluate the change in plant risk by ensuring the reactor building is available to provide effective fission product removal. Reactor building effectiveness was conservatively modeled by assuming reactor building availability for all accident sequences.
| |
| This resulted in an upper bound benefit of approximately
| |
| $64,577. This analysis case was used to model the benefit of phase II SAMAs 7, 13, and 21.E.2-5 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Strengthen Containment This analysis case was used to evaluate the change in plant risk from strengthening containment to reduce the probability of containment over-pressurization failure. A bounding analysis was performed by setting all energetic containment failure modes (DCH, steam explosions, late over-pressurization) to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $1,233,428.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 10, 15, 16, and 24.Base Mat Melt-Through This analysis case was used to evaluate the change in plant risk from increasing the depth of the concrete base mat to ensure base mat melt-through does not occur. A bounding analysis was performed by setting containment failure due to base mat melt-through to zero in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $25,831. This analysis case was used to model the benefit of phase 11 SAMA 11.Reactor Vessel Exterior Coolinm This analysis case was used to evaluate the change in plant risk from providing a method to perform ex-vessel cooling of the lower reactor vessel head. A bounding analysis was performed by modifying the probability of vessel failure by a factor of two to account for ex-vessel cooling in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $19,373. This analysis case was used to model the benefit of phase 11 SAMA 12.Vacuum Breakers This analysis case was used to evaluate the change in plant risk from improving the reliability of vacuum breakers to reseat following a successful opening and eliminate suppression pool scrubbing failures from the containment analysis.
| |
| A bounding analysis was performed by setting the vacuum breaker failure probability to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 17.Flooding the Rubble Bed This analysis case was used to evaluate the change in plant risk from providing a source of water to the drywell floor to flood core debris. A bounding analysis was performed by substituting the probabilities of wet core concrete interactions for dry core concrete interactions in the level 2 PSA model, which resulted in an upper bound benefit of approximately
| |
| $1,226,971.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 22.DC Power This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of Class 1 E DC power (e.g., increasing battery capacity, using fuel cells, or extending SBO injection provisions).
| |
| It was assumed that battery life could be extended E.2-6 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage from 14 hours to 24 hours to simulate additional battery capacity.
| |
| This enhancement would extend HPCI and RCIC operability and allow more credit for AC power recovery.
| |
| A bounding analysis was performed by changing the time available to recover offsite power before HPCI and RCIC are lost from 14 hours to 24 hours during SBO scenarios in the level I PSA model. This resulted in an upper bound benefit of approximately
| |
| $146,356.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 25, 26, 28, 33, and 35.Improve DC System This analysis case was used to evaluate the change in plant risk from improving injection capability by auto-transfer of AC bus control power to a standby DC power source upon loss of the normal DC source or from enhancing procedure to make use of DC bus cross-tie to improve DC power availability and reliability.
| |
| A bounding analysis was performed by setting the DC buses D1 6 and D1 7 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $118,568.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 27 and 34.Altemate Pump Power Source This analysis case was used to evaluate the change in plant risk from adding a small, dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power. A bounding analysis was performed by setting failure of the SBO diesel generator to zero in level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $265,687.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 29.Improve AC Power System This analysis case was used to evaluate the change in plant risk from improving AC power system cross-tie capability to enhance the availability and reliability of the AC power system. A bounding analysis was performed by setting the loss of MCCs B17, B18, and B15 to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $473,410.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 30.Dedicated DC Power and Additional Batteries and Divisions This analysis case was used to evaluate the change in plant risk from plant modifications that would provide motive power to components (e.g., providing a dedicated DC power supply, additional batteries, or additional divisions).
| |
| A bounding analysis was performed by setting the loss of DC bus D17 initiator, and one division of DC power, to zero in the level 1 PSA model, which resulted In an upper bound benefit of approximately
| |
| $903,025.
| |
| This analysis case was used to model the benefit of phase 11 SAMAs 31 and 32.E.2-7 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Locate RHR Inside Containment This analysis case was used to evaluate the change in plant risk from moving the RHR system inside containment to prevent an RHR system ISLOCA event outside containment.
| |
| A bounding analysis was performed by setting the RHR ISLOCA sequences to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $16,497. This analysis case was used to model the benefit of phase 11 SAMA 36.ISLOCA This analysis case was used to evaluate the change in plant risk from reducing the probability of an ISLOCA by increasing the frequency of valve leak testing. A bounding analysis was performed by setting the ISLOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $24,148. This analysis case was used to model the benefit of phase 11 SAMA 37.MSIV Design This analysis case was used to evaluate the change in plant risk from improving MSIV design to decrease the likelihood of containment bypass scenarios.
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| A bounding analysis was performed by setting the containment bypass failure due to MSIV leakage to zero in the level 2 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 38.Diesel to CST Makeup Pumps This analysis case was used to evaluate the change in plant risk from installing an independent diesel for the CST makeup pumps to allow continued operation of the high pressure injection system during an SBO event. As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torus allows continued HPCI and RCIC injection.
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| Therefore, a bounding analysis was performed by setting the failure to switchover from CST to torus to zero in the level 1 PSA model, which resulted in no benefit. This analysis case was used to model the benefit of phase 11 SAMA 39.High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the availability of high pressure injection (e.g., installing an independent AC powered high pressure injection system, passive high pressure injection system, or an additional high pressure injection system). A bounding analysis was performed by setting the CDF contribution due to unavailability of the HPCI system to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $110,212.
| |
| This analysis case was used to model the benefit of phase II SAMAs 40, 41, 42, 44, and 45.E.2-8 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Improve the Reliability of High Pressure Injection System This analysis case was used to evaluate the change in plant risk from plant modifications that would increase the reliability of the high pressure injection system. A bounding analysis was performed by reducing the HPCI system failure probability by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $76,025. This analysis case was used to model the benefit of phase 11 SAMA 43.SRVs Reseat This analysis case was used to evaluate the change in plant risk from improving the reliability of SRVs reseating.
| |
| A bounding analysis was performed by setting the stuck open SRVs initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $63,599. This analysis case was used to model the benefit of phase 11 SAMA 46.Diversity of Explosive Valves This analysis case was used to evaluate the change in plant risk from providing an alternate means of opening a pathway to the RPV for SLC system injection, thereby improving success probability for reactor shutdown.
| |
| A bounding analysis was performed by setting common cause failure of SLC explosive valves to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $12,915. This analysis case was used to model the benefit of phase II SAMA48.Reliability of SRVs This analysis case was used to evaluate the change in plant risk from installing additional signals to automatically open the SRVs. This improvement would reduce the likelihood of SRVs failing to open, thereby reducing the consequences of medium LOCAs. A bounding analysis was performed by setting the probability of SRVs failing to open when required by reactor pressure vessel overpressure conditions to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $31,799. This analysis case was used to model the benefit of phase 11 SAMA 49.Improve SRV Design This analysis case was used to evaluate the change in plant risk from improving the SRV design to increase the reliability of opening, thus increasing the likelihood that accident sequences could be mitigated using low pressure injection systems. A bounding analysis was performed by setting the probability of SRVs failing to open during RPV depressurization to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $194,378.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 50.E.2-9 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Self-Cooled ECCS Pump Seals This analysis case was used to evaluate the change in plant risk from providing self-cooled ECCS pump seals to eliminate dependence on the component cooling water system. A bounding analysis was performed by setting the CDF contribution from sequences involving RHR pump failures to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $29,412. This analysis case was used to model the benefit of phase 11 SAMA 51.Large Break LOCA This analysis case was used to evaluate the change in plant risk from installing a digital large break LOCA protection system. A bounding analysis was performed by setting the large break LOCA initiator to zero in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $14,109. This analysis case was used to model the benefit of phase 11 SAMA 52.Controlled Containment Venting This analysis case was used to evaluate the change in plant risk from changing the design of the containment vent valves and procedure to establish a narrow pressure control band. This would prevent rapid containment depressurization when venting, thus avoiding adverse impact on the ability of the low pressure ECCS injection systems to take suction from the torus. A bounding analysis was performed by reducing the probability of the operator failing to recognize the need to vent the torus by a factor of three in the level 1 PSA model, which resulted in an upper bound benefit of approximately
| |
| $137,237.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 53.ECCS Low Pressure Interlock This analysis case was used to evaluate the change in plant risk from installing a bypass switch to allow operator to bypass the ECCS low pressure interlock circuitry that inhibits opening of the RHR low pressure injection and core spray injection valves following sensor or logic failure. A bounding analysis was performed by setting the CDF contribution due to sensor failure, low pressure permissive logic failure, and miscalibration to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately
| |
| $21,761. This analysis case was used to model the benefit of phase 11 SAMA 54.Improve the Reliability of SSW and RBCCW Pumps This analysis case was used to evaluate the change in plant risk from providing a separate pump train to eliminate common cause failure of SSW and RBCCW pumps. A bounding analysis was performed by setting the CDF contribution due to common cause failures of SSW and RBCCW pumps to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately
| |
| $356,310.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 55.E.2-10
| |
| * Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Redundant DC Power Supplies to DTV Valves This analysis case was used to evaluate the change in plant risk from installing additional fuses to two DTV valve control circuits to enable the DTV function.
| |
| A bounding analysis was performed by setting the CDF contribution due to DC power supply failures to DTV valves AO-5042B and AO-5025 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately
| |
| $220,639.
| |
| -This analysis case was used to model the benefit of phase 11 SAMA 56.Proceduralize the Use of Diesel Fire Pump Hydroturbine This analysis case was used to evaluate the change in plant risk from revising the procedure to allow use of hydroturbine if EDG X-107A or diesel driven fire water pump P-140 is unavailable.
| |
| A bounding analysis was performed by setting the CDF contribution from the sequences involving a LOOP and failure of either EDG A or fuel oil transfer oil pump (P-141) to zero in the level I PSA model. This resulted in an upper bound benefit of approximately
| |
| $175,279.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 57.Proceduralize Alignment of Bus B3 to Feed Bus BI Loads or Bus B4 to Bus B2 This analysis case was used to evaluate the change in plant risk from providing a procedure to direct the operator to restore 480V MCCs B15 and B17 loads upon loss of 4.16kV bus A5 provided that 4.16kV bus A3 is available.
| |
| The same is true for restoring 480V MCCs B14 and B18 loads upon loss of 4.16kV bus A6 provided that 4.16kV bus A4 is available.
| |
| A bounding analysis was performed by setting the CDF contribution from the sequences involving a loss of the 4.16 kV bus A5 to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately
| |
| $190,797.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 58.Redundant Path from Fire Water Pump Discharge to LPCI Loops A and B Cross-tie This analysis case was used to evaluate the change in plant risk from installing a redundant path from fire protection water pump discharge to LPCI loops A and B cross-tie.
| |
| A bounding analysis was performed by setting the CDF contribution from the sequences involving fire water into LPCI loops A and B cross-tie failure to zero in the level 1 PSA model. This resulted in an upper bound benefit of approximately
| |
| $929,797.
| |
| This analysis case was used to model the benefit of phase 11 SAMA 59.E.2.4 Sensitivity Analyses Two sensitivity analyses were conducted to gauge the impact of assumptions upon the analysis.The benefits estimated for each of these sensitivities are presented in Table E.2-2.A description of each sensitivity case follows.E.2-11 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Sensitivity Case 1: Years Remaining Until End of Plant Life The purpose of this sensitivity case was to investigate the sensitivity of assuming a 27-year period for remaining plant life (i.e. seven years on the original plant license plus the 20-year license renewal period). The 20-year license renewal period was used in the base case. The resultant monetary equivalent was calculated using 27 years remaining until end of facility life to investigate the impact on each analysis case. Changing this assumption does not cause any additional SAMAs to be cost-beneficial.
| |
| Sensitivity Case 2: Conservative Discount Rate The purpose of this sensitivity case was to investigate the sensitivity of each analysis case to the discount rate. The discount rate of 7.0% used in the base case analyses is conservative relative to corporate practices.
| |
| Nonetheless, a lower discount rate of 3.0% was assumed in this case to investigate the impact on each analysis case. Changing this assumption does not cause any additional SAMAs to be cost-beneficial.
| |
| E.2-12 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2.5 References E.2-1 Appendix D-Attachment F, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.E.2-2 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Calvert Cliffs Nuclear Power Plant, Supplement 1, February 1999.E.2-3 General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18,1995.E.24 Appendix E- Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.E.2-5 Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.E.2-6 Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Dresden Nuclear Power Plant Units 2 and 3, January 2003.E.2-7 Appendix E-Attachment E, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Arkansas Nuclear One -Unit 2, October 2003.E.2-8 Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.E.2-9 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, May 1996.E.2-10 U.S. Nuclear Regulatory Commission, NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants, Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, May 1996.E.2-11 Museler, W. J., (Tennessee Valley Authority) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN) Units I and 2 -Severe Accident Mitigation Design Alternatives (SAMDAs)," letter dated October 7, 1994.E.2-12 Nunn, D. E., (TVA) to NRC Document Control Desk, "Watts Bar Nuclear Plant (WBN)Units I and 2 -Severe Accident Mitigation Design Alternatives (SAMDA) -Response to Request for Additional Information (RAI) -(TAC Nos. M77222 and M77223)," letter dated October 7, 1994.E.2-13 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage E.2-13 Liparulo, N. J., (Westinghouse Electric Corporation) to NRC Document Control Desk,"Submittal of Material Pertinent to the AP600 Design Certification Review," letter dated December 15,1992.E.2-14 U.S. Nuclear Regulatory Commission, NUREG-0498, Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2, Supplement No. 1, April 1995.E.2-15 U.S. Nuclear Regulatory Commission, NUREG-1 560, Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, December 1997.E.2-16 U.S. Nuclear Regulatory Commission, NUREG/CR-5474, Assessment of Candidate Accident Management Strategies, March 1990.E.2-17 Pilgrim Nuclear Power Station, Individual Plant Examination (IPE) Report, September 1992 E.2-18 Pilgrim Nuclear Power Station, Individual Plant Examination of External Events (IPEEE)Report, July 1994.E.2-19 U.S. Nuclear Regulatory Commission, NUREG/BR-01 84, Regulatory Analysis Technical Evaluation Handbook, January 1997.QW E.2-14 t.00 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation Phase II l Result of Potential SAMA ID SAMA Enhancement Improvements Related to Accident Mitigation Containment Phenomena 001 Install an SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost independent the probability of loss effective method of of containment heat suppression pool removal.cooling. I Basis for
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| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.002 Install a filtered SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost containment vent an alternate decay effective to provide fission heat removal method product for non-ATWS events, scrubbing.
| |
| with fission product Option 1: Gravel scrubbing.
| |
| Bed Filter Option 2: Multiple Venturi Scrubber Basis for
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| | |
| == Conclusion:==
| |
| | |
| Successful torus venting accident progression source terms are reduced by a factor of 2 to reflect the additional filtered capability.
| |
| The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-15 9.~3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
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| Phase II Result of Potential CDF off-Site Estimated Upper Estimated BoundEsti ated Conclusion SAMA ID Enhancement Reduction dos Benefit Estimated Cost ReductionBnet 003 Install a Assuming that injection 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost containment vent is available, this SAMA effective large enough to would provide alternate remove ATWS decay heat removal in decay heat. an ATWS event.Basis for
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| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from ATWS sequences associated with containment bypass were eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million.Therefore, this SAMA is not cost effective for PNPS.004 Create a large SAMA would ensure 0.00% 48.62% $436,759 $2,620,551
| |
| >$100 million Not cost concrete crucible that molten core debris effective with heat removal escaping from the potential under vessel would be the base mat to contained within the contain molten crucible.
| |
| The water core debris. cooling mechanism would cool the molten core, preventing a melt-through of the base mat.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $100 million.Therefore, this SAMA is not cost effective for PNPS.E.2-16 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II SAMA Result of Potential SAMA ID Enhancement.i 005 Create a water-cooled rubble bed on the pedestal.SAMA would contain molten core debris dropping on to the pedestal and would allow the debris to be cooled.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $19 million.Therefore, this SAMA is not cost effective for PNPS.006 Provide SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost modification for intentional flooding of effective flooding the the upper drywell head drywell head. such that if high drywell temperatures occurred, the drywell head seal would not fail.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-17 J 3 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 007 Enhance fire SAMA would improve 0.00% 1.16% $10,763 $64,577 >$2,500,000 Not cost protection system fission product effective and SGTS scrubbing in severe hardware and accidents.
| |
| procedures.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-18 t_' V ir_'Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) off-site Upper Phase II Result of Potential CDF Estimated Bound Estimated SAMA ID Enhancement Reduction Benefit Estimated Cost Conclusion ReductionBefi
| |
| ~ : Benefit 008 Create a core melt SAMA would provide 0.00% 48.62% $436,759 $2,620,551
| |
| >$5,000,000 Not cost source reduction cooling and effective system. containment of molten core debris. Refractory material would be placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with the material.
| |
| Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-19 3 I3__J, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF OffEstimate d Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID SAMA Enhancement Reduction Reduce on Benefit Estimated Cost RedutionBenefit 009 Install a passive SAMA would decrease 5.05% 4.70% $44,037 $264,219 $5,800,000 Not cost containmentspray the probability of loss effective system. of containment heat removal.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from loss of the drywell spray mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.010 Strengthen SAMA would reduce 0.00% 26.10% $205,571 $1,233,428
| |
| $12,000,000 Not cost primary and the probability of effective secondary containment over-containment.
| |
| pressurization failure.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.E.2-20 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF lt Estimated und Estimated Estmatd BunoEsimaed Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 011 Increase the SAMA would prevent 0.00% 0.43% $4,305 $25,831 >$5,000,000 Not cost depth of the base mat melt-through.
| |
| effective concrete base mat or use an alternative concrete material to ensure melt-through does not occur. lBasis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to base mat melt-through was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.012 Provide a reactor SAMA would provide 0.00% 0.22% $3,229 $19,373 $2,500jO00 Not cost vessel exterior the potential to cool a effective cooling system. molten core before it causes vessel failure, if the lower head could be submerged in water.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The probability of vessel failure was modified to account for potential ex-vessel cooling of the vessel bottom head region to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.E.2-21 J 3j Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase 11 Result of Potential CDF Offite Estimated Bound Estimated Est m a ed Bo ndost m a ed C onclusion SAMA ID M Enhancement Reduction Reduction Benefit Estimated Cost Redu tionBenefit 013 Construct a SAMA would provide a 0.00% 1.16% $10,763 $64,577 >$2,000,000 Not cost building method to effective connected to depressurize primary containment and containment that reduce fission product is maintained at a release.vacuum.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.014 2.g. Dedicated SAMA would decrease 4.70% 4.60% $43,639 $261,832 $5,800,000 Not cost Suppression Pool the probability of loss effective Cooling of containment heat removal.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from loss of the torus cooling mode of RHR was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $5.8 million. Therefore, this SAMA is not cost effective for PNPS.015 3.a. Create a SAMA increases time 0.00% 26.10% $205,571 $1,233,428
| |
| $8,000,000 Not cost larger volume in before containment effective containment.
| |
| failure and increases time for recovery.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $8 million.Therefore, this SAMA is not cost effective for PNPS.E.2-22 I:0 I A -: IK-: Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Off-Site Uppe r Bound EstimaEsE SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction Benefit Estimated Cost: ReductionBefi 016 3.b. Increase SAMA minimizes 0.00% 26.10% $205,571 $1,233,428
| |
| $12,000,000 Not cost containment likelihood of large effective pressure releases.capability (sufficient pressure to withstand severe accidents).
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.017 3.c. Install This SAMA addresses 0.00% 0.00% $0 $0 >$1,000,000 Not cost improved vacuum the reliability of a effective breakers vacuum breaker to (redundant valves reseat following a in each line). successful opening.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Vacuum breaker failures and suppression pool scrubbing failures were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $1 million.Therefore, this SAMA is not cost effective for PNPS.E.2-23 J)3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| O ff- ite sti m te d U p p e r Phase 11 Result of Potential CDF OffSite Estimated Und Estimated SAMA ID Enhancement Reduction Rducion Benefit Esim d Cost Redu tionBenefit 018 3.d. Increase the This SAMA would 0.00% 0.07% $2,153 $12,915 $12,000,000 Not cost temperature reduce the potential for effective margin for seals. containment failure under adverse conditions.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to high temperature drywell seal failure was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR were estimated to be $12 million and was judged to exceed the attainable benefit, even without a detailed cost estimate.
| |
| Therefore, this SAMA is not cost effective for PNPS.019 5.b/c. Install a SAMA would provide 0.00% 0.00% $0 $0 $3,000,000 Not cost filtered vent an alternate decay effective heat removal method for non-ATWS events, with fission product scrubbing.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Successful torus venting accident progressions source terms are reduced by a factor of 2 to reflect the additional filtered capability.
| |
| The cost of implementing this SAMA at Peach Bottom was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-24 I-'11 delo, n Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF i Estimate d BuEstimated Estmaed Bondostmaed
| |
| | |
| ==
| |
| Conclusion:==
| |
| | |
| SAMA ID Enhancement Reduction De Benefit Estimated Cost Reduction Benefit 020 7.a. Provide a SAMA would provide 0.00% 0.07% $2,153 $12,915 >$1,000,000 Not cost method of drywell intentional flooding of effective head flooding.
| |
| the upper drywell head such that if high drywell temperatures occurred, the drywell head seal would not fail.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Drywell head failures due to high temperature were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.021 13.a. Use This SAMA provides 0.00% 1.16% $10,763 $64,577 >$2,500,000 Not cost alternate method the capability to use effective of reactor building firewater sprays in the spray. reactor building to mitigate release of fission products into the reactor building following an accident.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Failure of the reactor building to contain releases was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2.5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-25 3 J I Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF OffSte Estimated Bound Estimated Est m a ed Bo ndost m a ed C onclusion SAMA ID SAMA Enhancement Reduction Rductios Benefit Estimated Cost Redu tionBenefit 022 14.a. Provide a SAMA would allow the 0.00% 22.48% $204,495 $1,226,971
| |
| $2,500,000
| |
| -Not cost means of flooding debris to be cooled. effective the rubble bed.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The probabilities of wet core concrete interactions were substituted for dry core concrete interactions to assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $2.5 million. Therefore, this SAMA is not cost effective for PNPS.023 14.b. Install a SAMA would enhance 0.00% 48.62% $436,759 $2,620,551
| |
| $8,750,000 Not cost reactor cavity debris coolability, effective flooding system. reduce core concrete interaction, and provide fission product scrubbing.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment failure due to core-concrete interactions (not including liner failures) was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at ANO-2 was estimated to be $8.75 million.Therefore, this SAMA is not cost effective for PNPS.024 Add ribbing to the This SAMA would 0.00% 26.10% $205,571 $1,233,428
| |
| $12,000,000 Not cost containment shell. reduce the chance of effective containment buckling under reverse pressure loading.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Energetic containment failure modes (DCH, steam explosion, late over-pressurization) were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities and at an ABWR was estimated to be $12 million. Therefore, this SAMA is not cost effective for PNPS.E.2-26 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Off-Site Estimated Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID SAMA Enhancement Reduction Benefit Estimated Cost Improvements Related to Enhanced AC/DC Reliability/Availability R 025 Provide additional SAMA would ensure 1.39% 2.79% $24,393 $146,356 $500,000 Not cost DC battery longer battery effective capacity.
| |
| capability during an SBO, which would extend HPCW/RCIC operability and allow more time for AC power recovery.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.026 Use fuel cells SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost instead of lead- DC power availability in effective acid batteries.
| |
| an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-27 J 31 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase 11 Result of Potential CDF OffSite Estimated Bound Estimated SAMA Esim tdoo ndEsi atd Conclusion SAMA ID Enhancement Reduction Rduction Benefit Estimated Cost Redu tionBenefit 027 Modification for SAMA would increase 4.65% 1.91% $19,761 $118,568 $500,000 Not cost Improving DC Bus reliability of AC power effective Reliability and injection capability.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of DC buses D16 and D07 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
| |
| Therefore, this SAMAis not cost effective for PNPS.028 2.i. Provide 16- SAMA includes 1.39% 2.79% $24,393 $146,356 $500,000 Not cost hour SBO improved capability to effective injection.
| |
| cope with longer SBO l scenarios.
| |
| lBasis for
| |
| | |
| == Conclusion:==
| |
| | |
| The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-28 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF OffEstimate d Bound Estimated Estmaed Bondostmaed Conclusion SAMA ID SAM Enhancement Reduction Benefit Estimated Cost ReductionBefi 029 9.b. Provide an This SAMA would 2.22% 5.06% $44,281 $265,687 >$2,000,000 Not cost alternate pump provide a small, effective power source. dedicated power source such as a dedicated diesel or gas turbine for the feedwater or condensate pumps so that they do not rely on offsite power.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the SBO diesel was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.030 9.g. Enhance SAMA would provide 11.10% 8.47% $78,902 $473,410 $146,120 Retain procedures to increased reliability of make use of AC AC power system and bus cross-ties.
| |
| reduce core damage and release frequencies.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of MCCs B17, B18, and B15 was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $146,120 by engineering judgment.E.2-29 I-)3.Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Off-SiteUpe Phase 11 SAMA Result of Potential CDF ose Estimated Bound Estimated SAMA ID Enhancement Reduction Benefit Estimated Cost ReductionBeet 031 10.a. Add a This SAMA addresses 24.3% 16.16% $150,504 $903,025 $3,000,000 Not cost dedicated DC the use of a diverse DC effective power supply. power system such as an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g., RCIC).Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.032 10.b. Install This SAMA addresses 24.3% 16.16% $150,504 $903,025 $3,000,000 Not cost additional the use of a diverse DC effective batteries or power system such as divisions.
| |
| an additional battery or fuel cell for the purpose of providing motive power to certain components (e.g., RCIC).Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of DC Bus 'B' was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be $3 million. Therefore, this SAMA is not cost effective for PNPS.E.2-30 VI Ar-I-Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase I SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Off-Site Estimated Bound Estimated SAMA ID Enhancement Reduction Dose Benefit Estimated Cost Conclusion ReductionBeet Benefit 033 10.c. Install fuel SAMA would extend 1.39% 2.79% $24,393 $146,356 >$2,000,000 Not cost cells. DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.034 10.d. Enhance This SAMA would 4.65% 1.91% $19,761 $118,568 $13,000 Retain procedures to improve DC power make use of DC availability.
| |
| bus cross-ties.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to loss of DC buses D16 and D17 was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $13,000 by engineering judgment.E.2-31 3 3 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phas IIResut o Potntil CD Of~itUpper Phase II SAMA Result of Potential CDF Dose Estimated Bound Estimated Conclusion SAMA ID Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 035 10.e. Extended SAMA would extend 1.39% 2.79% $24,393 $146,356 $500,000 Not cost SBO provisions.
| |
| DC power availability in effective an SBO, which would extend HPCI/RCIC operability and allow more time for AC power recovery.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The time available to recover offsite power before HPCI and RCIC are lost was changed from 14 hours to 24 hours during SBO scenarios to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $500,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.Improvements in Identifying and Mitigating Containment Bypass 036 Locate RHR SAMA would prevent 0.33% 0.21% $2,749 $16,497 >$500,000 Not cost inside ISLOCA outside effective containment.
| |
| containment.
| |
| .Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| RHR ISLOCA accident sequences were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Quad Cities was estimated to be greater than $500.000.
| |
| Therefore, this SAMA is not cost effective for PNPS.037 Increase SAMA could reduce 0.54% 0.38% $4,025 $24,148 $100,000 Not cost frequency of valve ISLOCA frequency.
| |
| effective leak testing. _ _Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to ISLOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $100,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-32 T_AP"-, It_Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF OffSit Estimated Upper Estimated Estmaed Bondostmaed Conclusion SAMA ID Enhancement Reduction Dose Benefit Estimated Cost ReductionBeet 038 8.e. Improve This SAMA would 0.00% 0.00% $0 $0 >$2,000,000 Not cost MSIV design. decrease the likelihood effective of containment bypass scenarios.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Containment bypass failure due to MSIV leakage was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.Improvements Related to Core Cooling System 039 Install an SAMA would allow 0.00% 0.00% $0 $0 $135,000 Not cost independent continued inventory in effective diesel for the CST CST during an SBO.makeup pumps.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| As currently modeled, if CST water level is low, swapping HPCI/RCIC suction from the CST to the torus allows continued HPCI/RCIC injection.
| |
| Therefore, the failure to switchover from CST to torus was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $135,000 by engineering judgment.Therefore, this SAMA is not cost effective for PNPS.E.2-33 9 3-3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase 11 SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Offt Estimated Upped Estimated Estmaed Bondostmaed Conclusion SAMA ID S Enhancement Reduction De Benefit Estimated Cost ReductionBefi 040 Provide an SAMA would reduce 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional high frequency of core melt effective pressure injection from small LOCA and pump with SBO sequences.
| |
| independent diesel.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.041 Install SAMA would allow 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost independent AC makeup capabilities effective high pressure during transients, small injection system. LOCAs, and SBOs.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-34 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Off-SiteUpper Phase 11 SAMA Result of Potential CDF off-Site Estimated Bound Estimated
| |
| | |
| ==
| |
| Conclusion:==
| |
| | |
| SAMA ID Enhancement Reduction De Benefit' Estimated Cost ReductionBeet 042 2.a. Install a SAMA would improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost passive high prevention of core melt effective pressure system. sequences by providing additional high pressure capability to remove decay heat through an isolation condenser type system.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.043 2.d. Improved SAMA will improve 2.11% 1.43% $12,671 $76,025 >$2,000,000 Not cost high pressure prevention of core melt effective systems sequences by K improving reliability of high pressure capability to remove decay heat.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from reducing the HPCI system failure probability by a factor of 3 was estimated to bound the potential impact of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.E.2-35 a 3 J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase 11 Result of Potential CDF OffSt Estimated Bound Estimated SAMA Esimtdoond Esiatd Conclusion SAMA ID Enhancement Reduction Benefit Estimated Cost ReductionBeet 044 2.e. Install an SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost additional active reliability of high- effective high pressure pressure decay heat system. removal by adding an additional system.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.045 8.c. Add a diverse SAMA will improve 3.15% 1.97% $18,369 $110,212 >$2,000,000 Not cost injection system. prevention of core melt effective sequences by providing additional injection capabilities.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to failure of the HPCI system was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-36 It--- 1 It-, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued) 1 OffSiteUpper Phase II I Result of Potential CDF OffSite Estimated Bound Estimated C SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost I. ~~~~~Reduction Bnft ______Improvements Related to ATWS Mitigation 046 Increase SRV SAMA addresses the 1.51% 0.92% $10,600 $63,599 $2,000,000 Not cost reseat reliability.
| |
| risk associated with effective dilution of boron caused by the failure of the SRVs to reseat after SLC injection.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to stuck open relief valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.047 11.a. Install an This SAMA would 0.50% 1.19% $10,283 $61,701 >$2,000,000 Not cost ATWS sized vent. provide the ability to effective remove reactor heat from ATWS events.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from ATWS sequences associated with containment bypass were eliminated to conservatively assess the benefit of this SAMA. The cost of implementing of this SAMA at Peach Bottom was estimated to be greater than $2 million. Therefore, this SAMA is not cost effective for PNPS.E.2-37 D J J Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase 11 Result of Potential CDF Offit Estimated Bound Estimated SAMA Esim tdoo ndEsi atd Conclusion SAMA ID Enhancement Reduction Dos Benefit Estimated Cost ReductionBefi 048 Diversify An alternate means of 0.00% 0.02% $2,153 $12,915 >$200,000 Not cost explosive valve opening a pathway to effective operation.
| |
| the RPV for SLC system injection would improve the success probability for reactor shutdown.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| Common cause failure of SLC explosive valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.Other Improvements 049 Increase the SAMA reduces the 0.73% 0.60% $5,300 $31,799 >$1,500,000 Not cost reliability of SRVs consequences of effective by adding signals medium break LOCAs.to open them automatically.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from SRVs failing to open in medium LOCA sequences was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $1.5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-38 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Offsit Estimate d Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Costnclus Reduction:Bnei 050 8.e. Improve SRV This SAMA would 4.81% 3.51% $32,396 $194,378 >$2,000,000 Not cost design. improve SRV reliability effective thus increasing the likelihood that sequences could be mitigated using low-pressure heat removal.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The probability of SRV failure to open for vessel depressurization was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $2 million at Peach Bottom. Therefore, this SAMA is not cost effective for PNPS.051 Provide self- SAMA would eliminate 0.47% 0.55% $4,902 $29,412 >$200,000 Not cost cooled ECCS ECCS dependency on effective pump seals. the component cooling , water system.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving RHR pump failures was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $200,000 by engineering judgment.Therefore, this SAMA is not cost effective for PNPS E.2-39 J 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF OffSite Estimated Upper Estimated SAMA Dose Etmtd B u dE i aed Conclusion SAMA ID Enhancement Reduction Reducton Benefit Estimated Cost Redu tionBenefit 052 Provide digital Upgrade plant 0.07% 0.01% $2,352 $14,109 >$100,000 Not cost large break LOCA instrumentation and effective protection.
| |
| logic to improve the capability to identify symptoms/precursors of a large break LOCA (a leak before break).Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution due to large break LOCA was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $100,000 by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-40 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phas IIRestto Potntil CF ~ iteUpper PhaseS A ResuMA-of Potential CDF Estimated Bound Estimated o SAMA ID Enhancement Reduction Dose Benefit Estimated Cost. Conclusion ReductionBefi Improvements Related to IPE, IPE Update& IPEEE Insights 053 Control This SAMA would 3.61% 2.24% $22,873 $137,237 $300,000 Not cost containment establish a narrow effective venting within a pressure control band narrow band of to prevent rapid pressure containment depressurization when venting is implemented thus avoiding adverse impact on the low pressure ECCS injection systems taking suction from the torus.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The probability of the operator failing to recognize the need to vent the torus was reduced by a factor of 3 to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA was estimated to be $300,000 by engineering judgment.
| |
| Therefore, this SAMA Is not cost effective for PNPS.E.2-41 J)i Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Phase II Result of Potential CDF Estimated Upper Estimated SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost onclus on ReductionBeet 054 Install a bypass This SAMA would 0.28% 0.33% $3,627 $21,761 $1,000,000 Not cost switch to bypass reduce the core effective the low reactor damage frequency pressure contribution from the interlocks of LPCI transients with stuck or core spray open SRVs or LOCAs injection valves cases. Core Spray and LPCI injection valves require a low permissive signal from the same two sensors to open the valves for RPV injection.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The probability of the ECCS low-pressure permissive failing was eliminated to conservatively assess the benefit of this SAMA on CDF. The cost of implementing this SAMA at Dresden was estimated to be $1 million. Therefore, this SAMA is not cost effective for PNPS.055 Increase the This SAMA would 4.37% 6.63% $59,385 $356,310 >$5 million Not cost reliability of SSW reduce common cause effective and RBCCW dependencies from pumps. SSW and RBCCW systems and thus reduce plant risk.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving common cause failures of SSW and RBCCW was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be greater than $5 million by engineering judgment.
| |
| Therefore, this SAMA is not cost effective for PNPS.E.2-42 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered In Cost-Benefit Evaluation (Continued)
| |
| Phase II SAMA Result of Potential CDF Off-Site Estimated Upper Estimated SAMA ID S Enhancement Reduction Reduction Benefit Estimated Cost RedutionBenefit 056 Provide redundant This SAMA would 8.81% 3.51% $36,773 $220,639 $112,400 Retain DC power improve reliability of supplies to DTV the DTV valves and valves. enhance containment heat removal capability.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving DC power supply failures to the DTV valves was eliminated to conservatively assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $112,400 by engineering judgment.057 Proceduralize use This SAMAwould 2.25% 3.14% $29,213 $175,279 $26,000 Retain of the diesel fire increase capability to pump hydro provide makeup to the turbine in the fire pump- day tank to event of EDG A allow continued failure or operation of the diesel unavailability.
| |
| fire pump, without dependence on electrical power.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving a LOOP and failure of either EDG A, or the EDG A fuel oil transfer oil pump, was eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be$26,000 by engineering judgment.E.2-43 3 3 3 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-1 Summary of Phase II SAMA Candidates Considered in Cost-Benefit Evaluation (Continued)
| |
| Upper Phase 11 Result of Potential CDF f-eeneit Buppenr Cst SAMA ID SAMA Enhancement Reduction Dose Benefit Estimated Cost Cnlso 058 Proceduralize the This SAMA would 4.92% 3.14% $31,799 $190,797 $50,000 Retain operator action to provide the direction to feed BI loads via restore B15 and B17 B3 When A5 is loads upon loss of A5 unavailable post- initiating events as long trip. Similarly, as A3 is available.
| |
| feed B2 loads via Additionally, it would B4 when A6 is provide the direction to unavailable post restore B14 and B18 trip., loads upon loss of A6 initiating events as long as A4 is available.
| |
| Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving loss of 4160VAC safeguard bus AS was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $50,000 by engineering judgment.059 Provide redundant This SAMA would 8.77% 17.19% $154,966 $929,797 $1,956,000 Not cost path from fire enhance the effective protection pump availability and discharge to LPCI reliability of the loops A and B firewater cross-tie to cross-tie.
| |
| LPCI loops A and B for reactor vessel injection and drywell spray.Basis for
| |
| | |
| == Conclusion:==
| |
| | |
| The CDF contribution from sequences involving firewater injection failures was conservatively eliminated to assess the benefit of this SAMA. The cost of implementing this SAMA was estimated to be $1,956,000 by engineering judgment.Therefore, this SAMA is not cost effective for PNPS E.2-44 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results Upper Upper Upper PhEstimated Bound Estimaited Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMASAMA Benefit Cost Benefit Benefit IDBase Line Base Lne Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 I Install an independent
| |
| $43,639 $261,832 $5,800,000
| |
| $50,320 $301,920 $59,355 $356,129 method of suppression pool cooling.2 Install a filtered containment
| |
| $0 $0 $3,000,000
| |
| $0 $0 $0 $0 vent to provide fission product scrubbing.
| |
| Option 1: Gravel Bed Filter Option 2: Multiple Venturi Scrubber 3 Install a containment vent $10,283 $61,701 >$2,000,000
| |
| $11,702 $70,211 $14,207 $85,244 large enough to remove ATWS decay heat.4 Create a large concrete $436,759 $2,620,551
| |
| >$100 million $492,136 $2,952,813
| |
| $610,307 $3,661,845 crucible with heat removal potential under the basemat to contain molten core debris.5 Create a water-cooled rubble $436,759 $2,620,551
| |
| $19,000,000
| |
| $498,057 $2,988,339
| |
| $610,307 $3,661,845 bed on the pedestal.E.2-45 3 J.a..Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAASAMA Benefit CotBenefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity BsLie BeLieCase I Case I Case 2 Case 2 6 Provide modification for $2,153 $12,915 >$1,000,000
| |
| $2,425 $14,551 $3,008 $18,048 flooding the drywell head 7 Enhance fire protection
| |
| $10,763 $64,577 >$2,500,000
| |
| $12,127 $72,764 $15,040 $90,238 system and/or SGTS hardware and procedures.
| |
| 8 Create a core melt source $436,759 $2,620,551
| |
| >$5,000,000
| |
| $498,057 $2,988,339
| |
| $610,307 $3,661,845 reduction system.9 Install a passive containment
| |
| $44,037 $264,219 $5,800,000
| |
| $50,845 $305,069 $59,803 $358,816 spray system., 10 Strengthen primary/ $205,571 $1,233,428
| |
| $12,000,000
| |
| $231,636 $1,389,815
| |
| $287,257 $1,723,540 secondary containment.
| |
| 11 Increase the depth of the $4,305 $25,831 >$5,000,000
| |
| $4,851 $29,105 $6,016 $36,095 concrete basemat or use an alternative concrete material to ensure melt-through does not occur 12 Provide a reactor vessel $3,229 $19,373 $2,500,000
| |
| $3,638 $21,828 $4,512 $27,071 exterior cooling system (see#7) _E.2-46 I- 1 A4*1Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID L Sensitivity Sensitivity Sensitivity Sensitivity Base Case I Case I Case 2 Case 2 13 Construct a building to be $10,763 $64,577 >$2,000,000
| |
| $12,273 $73,640 $15,040 $90,238 connected to primary/secondary containment that is maintained at a vacuum 14 2.g. Dedicated Suppression
| |
| $43,639 $261,832 $5,800,000
| |
| $51,067 $306,400 $59,355 $356,129 Pool Cooling 15 3.a. Create a larger volume in $205,571 $1,233,428
| |
| $8,000,000
| |
| $234,423 $1,406,537
| |
| $287,257 $1,723,540 containment.
| |
| 16 3.b. Increase containment
| |
| $205,571 $1,233,428
| |
| $12,000,000
| |
| $234,423 $1,406,537
| |
| $287,257 $1,723,540 pressure capability (sufficient pressure to withstand severe accidents).
| |
| 17 3.c. Install improved vacuum $0 $o >$1,000,000
| |
| $0 $0 $0 $0 breakers (redundant valves in each line).18 3.d. Increase the temperature
| |
| $2,153 $12,915 $12,000,000
| |
| $2,455 $14,728 $3,008 $18,048 margin for seals.19 5.b/c. install a filtered vent $0 $0 $3,000,000
| |
| $0 $0 $0 $0 E.2-47
| |
| -3Ad U 9, Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Case I Case I Case 2 Case 2 20 7.a. Provide a method of $2,153 $12,915 >$1,000,000
| |
| $2,455 $14,728 $3,008 $18,048 drywell head flooding.21 13.a. Use alternate method of $10,763 $64,577 >$2,500,000
| |
| $12,273 $73,640 $15,040 $90,238 reactor building spray.22 14.a. Provide a means of $204,495 $1,226,971
| |
| $2,500,000
| |
| $230,423 $1,382,539
| |
| $285,753 $1,714,516 flooding the rubble bed.23 14.b. Install a reactor cavity $436,759 $2,620,551
| |
| $8,750,000
| |
| $498,057 $2,988,339
| |
| $610,307 $3,661,845 flooding system.24 Add ribbing to the $205,571 $1,233,428
| |
| $12,000,000
| |
| $234,423 $1,406,537
| |
| $287,257 $1,723,540 containment shell.25 Provide additional DC battery $24,393 $146,356 $500,000 $27,830 $166,978 $33,598 $201,588 capacity.26 Use fuel cells instead of lead- $24,393 $146,356 >$2,000,000
| |
| $28,207 $169,242 $33,598 $201,588 acid batteries.
| |
| 27 Modification for Improving DC $19,761 $118,568 $500,000 $23,377 $140,262 $26,044 $156,263 Bus Reliability 28 2.i. Provide 16-hour SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 injection.
| |
| E.2-48 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound IP Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA, Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I -Case 1 Case 2 Case 2 29 9.b. Provide an alternate
| |
| $44,281 $265,687 >$2,000,000
| |
| $50,546 $303,278 $60,956 $365,738 pump power source.30 9.g. AC Bus Cross-Ties
| |
| $78,902 $473,410 $146,120 $91,662 $549,972 $106,357 $638,142 31 10.a. Add a dedicated DC $150,504 $903,025 $3,000,000
| |
| $178,405 $1,070,432
| |
| $201,864 $1,211,183 power supply.32 10.b. Install additional
| |
| $150,504 $903,025 $3,000,000
| |
| $178,405 $1,070,432
| |
| $201,864 $1,211,183 batteries or divisions.
| |
| 33 10.c. Install fuel cells. $24,393 $146,356 >$2,000,000
| |
| $28,207 $169,242 $33,598 $201,588 34 10.d. DC Cross-Ties
| |
| $19,761 $118,568 $13,000 $23,377 $140,262 $26,044 $156,263 35 10.e. Extended SBO $24,393 $146,356 $500,000 $28,207 $169,242 $33,598 $201,588 provisions.
| |
| 36 Locate RHR inside $2,749 $16,497 >$500,000
| |
| $3,213 $19,276 $3,680 $22,077 containment.
| |
| 37 Increase frequency of valve $4,025 $24,148 $100,000 $4,688 $28,127 $5,407 $32,444 leak testing.38 8.e. improve MSIV design. $0 $0 >$2,000,000
| |
| $0 $0 $0 $0 E.2-49 3 a Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound Ii Benefit Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID ; L Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 39 Install an independent diesel $0 $0 $135,000 $0 $0 $0 $0 for the CST makeup pumps.40 Provide an additional high $18,369 $110,212 >$2,000,000
| |
| $21,540 $129,238 $24,477 $146,860 pressure injection pump with independent diesel.41 Install independent AC high $18,369 $110,212 >$2,000,000
| |
| $21,902 $131,415 $24,477 $146,860 pressure injection system.42 2.a. Install a passive high $18,369 $110,212 >$2,000,000
| |
| $21,902 $131,415 $24,477 $146,860 pressure system.43 2.d. Improved high pressure $12,671 $76,025 >$2,000,000
| |
| $14,851 $89,109 $16,894 $101,363 systems 44 2.e. Install an additional
| |
| $18,369 $110,212 >$2,000,000
| |
| $21,902 $131,415 $24,477 $146,860 active high pressure system.45 8.c. Add a diverse injection
| |
| $18,369 $110,212 >$2,000,000
| |
| $21,902 $131,415 $24,477 $146,860 system.46 Increase SRV reseat $10,600 $63,599 $2,000,000
| |
| $12,326 $73,958 $14,270 $85,623 reliability.
| |
| 47 11.a. Install an ATWS sized $10,283 $61,701 >$2,000,000
| |
| $11,857 $71,142 $14,207 $85,244 vent.E.2-50 Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound 11 Benefit Estimated Estimated Benefit Estimated Benefit Estimated SSAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 48 Diversify explosive valve $2,153 $12,915 >$200,000
| |
| $2,425 $14,551 $3,008 $18,048 operation.
| |
| 49 Increase the reliability of $5,300 $31,799 >$1,500,000
| |
| $6,163 $36,978 $7,135 $42,811 SRVs by adding signals to open them automatically.
| |
| 50 8.e. Improve SRV design. $32,396 $194,378 >$2,000,000
| |
| $37,767 $226,602 $43,483 $260,897 51 Provide self-cooled ECCS $4,902 $29,412 >$200,000
| |
| $5,638 $33,829 $6,687 $40,125 pump seals.52 Provide digital large break $2,352 $14,109 >$1 00,000 $2,688 $16,126 $3,232 $19,391 LOCA protection.
| |
| 53 Control containment venting $22,873 $137,237 $300,000 $26,653 $159,919 $30,716 $184,299 within a narrow band of pressure 54 Install a bypass switch to $3,627 $21,761 $1,000,000
| |
| $4,163 $24,978 $4,960 $29,758 bypass the low reactor pressure interlocks of LPCI or core spray injection valves.55 Improve SSW System and $59,385 $356,310 >$5 million $67,986 $407,918 $81,467 $488,799 RBCCW pump recovery.E.2-51 J 3.,)Pilgrim Nuclear Power Station Applicant's Environmental Report Operating License Renewal Stage Table E.2-2 Sensitivity Analysis Results (Continued)
| |
| Upper Upper Upper Phase Estimated Bound Estimated Bound Estimated Bound II Benefit Estimated Estimated Benefit Estimated Benefit Estimated SAMA SAMA Benefit Cost Benefit Benefit ID Sensitivity Sensitivity Sensitivity Sensitivity Base Line Base Line Case I Case I Case 2 Case 2 56 Provide redundant DC power $36,773 $220,639 $112,400 $43,541 $261,247 $48,408 $290,449 supplies to DTV valves.57 Proceduralize the use of $29,213 $175,279 $26,000 $33,568 $201,406 $39,901 $239,406 diesel fire pump hydroturbine in the event of EDG A failure or unavailability.
| |
| 58 Proceduralize the operator $31,799 $190,797 $50,000 $36,980 $221,878 $42,811 $256,868 action to feed B1 loads via B3 When AS is unavailable post-trip.59 Provide redundant path from $154,966 $929,797 $1,956,000
| |
| $176,682 $1,060,091
| |
| $213,620 $1,281,720 fire protection pump discharge to LPCI loops A and B cross-tie.
| |
| E.2-52}}
| |