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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20055C8601990-06-18018 June 1990 Safety Evaluation Supporting Amend 132 to License DPR-3 ML20248H7391989-10-0303 October 1989 Safety Evaluation Not Accepting Procedure Generating Program for Plant.Program Should Be Revised to Reflect Items Described in Section 2 of Rept.Revision Need Not Be Submitted to NRC ML20247F1431989-09-0707 September 1989 Safety Evaluation Supporting Amend 124 to License DPR-3 ML20247E6831989-08-31031 August 1989 Safety Evaluation Supporting Amend 123 to License DPR-3 ML20246F2771989-07-11011 July 1989 Safety Evaluation Supporting Mods to ECCS Evaluation Model, Including Changes to FLECHT-based Reflood Heat Transfer Correlation,Steam Cooling Model & post-critical Heat Flux Heat Transfer Model ML20195D6701988-11-0101 November 1988 Safety Evaluation Supporting Amend 120 to License DPR-3 ML20205G1961988-10-25025 October 1988 Safety Evaluation Supporting Amend 119 to License DPR-3 ML20204G4871988-10-17017 October 1988 Safety Evaluation Supporting Amend 118 to License DPR-3 ML20205C4061988-10-14014 October 1988 Safety Evaluation Supporting Amend 117 to License DPR-3 ML20207L7051988-10-12012 October 1988 Safety Evaluation Supporting Amend 116 to License DPR-3 ML20207E8151988-08-0505 August 1988 Safety Evaluation Supporting Amend 115 to License DPR-3 ML20151M4911988-07-29029 July 1988 Safety Evaluation Supporting Amend 114 to License DPR-3 ML20151K3801988-07-25025 July 1988 Safety Evaluation Supporting Amend 113 to License DPR-3 ML20151K8571988-07-19019 July 1988 Safety Evaluation Supporting Amend 112 to License DPR-3 ML20153A8661988-06-29029 June 1988 Safety Evaluation Accepting Util Proposed Reflood Steam Cooling Model ML20196K2741988-06-28028 June 1988 Safety Evaluation Supporting Amend 111 to License DPR-3 ML20195K1501988-06-17017 June 1988 Safety Evaluation Supporting Amend 110 to License DPR-3 ML20195C5851988-06-13013 June 1988 Safety Evaluation Supporting Amend 109 to License DPR-3 ML20155K5141988-06-0909 June 1988 Safety Evaluation Supporting Amend 108 to License DPR-3 ML20154J7661988-05-18018 May 1988 Safety Evaluation Supporting Amend 107 to License DPR-3 ML20216J4081987-06-26026 June 1987 Safety Evaluation Supporting Amend 106 to License DPR-3 ML20216C1111987-06-18018 June 1987 Safety Evaluation Granting Three of Seven Requests Submitted by Util for Relief from Inservice Insp & Testing Requirements.Four Requests Withdrawn,Per 870122,0410 & 0507 Ltrs ML20215C5881987-06-0404 June 1987 Safety Evaluation Supporting Util 860505,870402,& 0506 Submittals Re Seismic Reevaluation of Plant.Concludes That Foundation Soils Under Reactor & Under Vapor Container Have Adequate Strength to Support Seismic Load from Earthquake NUREG-0825, Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed1987-05-13013 May 1987 Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed ML20213G9161987-05-13013 May 1987 Safety Evaluation Supporting Amend 105 to License DPR-3 ML20213D9671987-05-0707 May 1987 Safety Evaluation Supporting Amend 104 to License DPR-3 ML20207S6231987-03-10010 March 1987 Safety Evaluation Supporting Util 860122,0812,1028 & 870204 Submittals Re Fracture Toughness Requirements for Protection Against PTS Events ML20211N5881987-02-19019 February 1987 Safety Evaluation Re First Level Undervoltage Protection Testing.Testing Unnecessary ML20211L3951987-02-17017 February 1987 Safety Evaluation Supporting Amend 103 to License DPR-3 Re Max Nominal Enrichment of Fuel ML20207N8811987-01-0707 January 1987 Safety Evaluation Supporting Amend 102 to License DPR-3 ML20207N4261987-01-0606 January 1987 Safety Evaluation Supporting Amend 101 to License DPR-3 ML20207J9451986-12-30030 December 1986 SER Accepting Util 831105 & 850709 Responses to Generic Ltr 83-28,Item 2.1 (Part 2), Vendor Interface Program - Reactor Trip Sys Components ML20215E1201986-12-0909 December 1986 Safety Evaluation Supporting Util 830419 & 0830,840119, 851022 & 860930 Responses Re Conformance to Reg Guide 1.97. Plant Design Acceptable W/Exception of Neutron Flux Variable ML20214X3391986-12-0101 December 1986 Safety Evaluation Supporting Amend 100 to License DPR-3 ML20214J8521986-11-18018 November 1986 Sser Accepting SPDS Contingent Upon Resolution of Concerns Re Maint & Improvement of Placement & Visual Access of Containment Isolation Panel & Minor Human Factors Engineering Concerns ML20215E6471986-10-0202 October 1986 Safety Evaluation Supporting Util Requests for Exemption from Specific Requirements in App R to 10CFR50.Existing Fire Protection Provides Level of Protection Equivalent to Technical Requirements of App R ML20210S1791986-09-23023 September 1986 Safety Evaluation Supporting Amend 99 to License DPR-3 ML20212Q1151986-08-27027 August 1986 Safety Evaluation Supporting Util 830412 Proposal to Provide Integrated Safe Shutdown Sys Which Could Be Used for Safe Shutdown in Event of Fire at Facility ML20212N0161986-08-20020 August 1986 Safety Evaluation Supporting Amend 98 to License DPR-3 1999-08-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J5111999-08-31031 August 1999 Rev 29 to Yankee Decommissioning QA Program ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20209D5391999-06-22022 June 1999 Rev 29 to Yaec Decommissioning QA Program ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20154P9691998-10-16016 October 1998 Rev 28 to Yankee Atomic Electric Co Decommissioning QA Program ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20216C4581998-02-27027 February 1998 Response to NRC Demand for Info (NRC OI Rept 1-95-050) ML20203L1931998-02-25025 February 1998 Duke Energy Corp,Duke Engineering & Svcs,Inc,Yankee Atomic Small Break LOCA Technical Review Rept ML20203L2451998-02-23023 February 1998 Assessment Rept of Engineering & Technical Work Process Utilized at De&S Bolton Ofc ML20203L1621998-02-18018 February 1998 Rept of Root Cause Assessment Review ML20203L2691998-02-16016 February 1998 Duke Engineering & Svcs Assessment Process Review Rept ML20199B4601998-01-20020 January 1998 Special Rept:On 980105,meteorological Monnitoring Instrumentation for Air Temp Delta T Inoperable for More than 7 Days.Caused by Breakdown in Wiring Between Junction Box at 199 Foot Level.Wiring Replaced ML20203J3001997-12-31031 December 1997 Ynps 1997 Annual Rept ML20217N0981997-08-21021 August 1997 LER 97-S02-00:on 970725,discovered Uncontrolled Safeguards Documents.Caused by Personnel Error.Matls Retrieved & Stored in Safeguards Repositories ML20210H0991997-08-0707 August 1997 LER 97-S01-00:on 970709,potential Compromise of Safeguards Info Occurred.Caused by Human error.Stand-alone Personal Computer & Printer Not Connected to Network,Have Been Located within Text Graphics Svc Dept ML20149K7781997-07-24024 July 1997 Special Rept:On 970520 & 0714,air Temp Delta T Channel Indicated Temp Difference Between Top & Bottom of Meteorological Tower.Caused by Reversed Input Wiring to Channel.Restored Air Temp Delta T Channel Operability ML20141E4671997-05-30030 May 1997 Rev 28 to Operational QA Program ML20135C8461996-12-31031 December 1996 Yankee Nuclear Power Station 1996 Annual Rept ML20132G6771996-12-20020 December 1996 Rev 27 to YOQAP-I-A, Operational QA Program ML20058N4771993-12-20020 December 1993 Rev 0.0 to Yankee Nuclear Power Station Decommissioning Plan ML20059K8491993-12-15015 December 1993 Clarifications to Pages 2,41,43 & 44 of 44 in Section I, Organization of YOQAP-I-A,Rev 24, Operational QA Program ML20059C5011993-10-29029 October 1993 Special Rept:On 931019,meteorological Instrumentation Channel for Delta T Declared Inoperable.Caused by Ceased Aspirator Motor Located at Top of Tower.Motor Replaced ML20056H1741993-06-10010 June 1993 Preliminary Assessment of Potential Human Exposures to Routine Tritium Emissions from Yankee Atomic Electric Co Nuclear Power Facility Located Near Rowe,Ma ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20198D2481992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Initial Certification Program ML20198D2541992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Recertification Program ML20062H1981990-11-30030 November 1990 Plant Specific Fast Neutron Exposure Evaluations for First 20 Operating Fuel Cycles of Yankee Rowe Reactor ML20058H2841990-11-0303 November 1990 Special Rept:On 901101,control Rod 24 Found Disconnected from Drive Shaft.Drive Shaft Latching Will Be Initiated ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20062E8331990-10-31031 October 1990 Monthly Operating Rept for Oct 1990 for Yankee Atomic Power Station ML20058G1471990-10-31031 October 1990 Vol 2 to Star Methodology Application for PWRs Control Rod Ejection Main Steam Line Break ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20062B6751990-09-30030 September 1990 Monthly Operating Rept for Yankee Atomic Power Station for Sept 1990 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20059E3071990-08-31031 August 1990 Safety Assessment of Yaec 1735, Reactor Pressure Vessel Evaluation Rept for Yankee Nuclear Power Station. Detailed Plan of Action W/Listed Elements Requested within 60 Days After Restart to Demonstrate Ability to Operate Longer ML20059E8001990-08-31031 August 1990 Monthly Operating Rept for Aug 1990 for Yankee Atomic Power Station ML20058P7841990-08-14014 August 1990 Part 21 Rept Re Misapplication of Fluorolube FS-5 Oil in Main Steam Line Pressure Gauges.All Four Indicators Replaced W/Spare Gauges Which Utilize High Temp Silicone Oil ML20058N6581990-08-13013 August 1990 Special Rept Re Diesel Fire Pump & Tank Inoperable for Greater than Seven Days for Draining,Cleaning & Insp.During Period Redundant Pumping Capacity Available Via Two Remaining Electric Driven Fire Pumps ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20056A1961990-08-0101 August 1990 Special Rept:Two Fire Pumps Inoperable at Same Time.Caused by Necessity to Accomplish Surveillance to Verify Capability to Start Pump on Emergency Diesel Generator 3 & Planned 18-month Insp of Diesel Per Tech Specs ML20055E1591990-07-31031 July 1990 Reactor Pressure Vessel Evaluation Rept ML20055G7011990-07-31031 July 1990 Yankee Nuclear Power Station Core 21 Performance Analysis ML20055G6801990-07-31031 July 1990 Yankee Plant Small Break LOCA Analysis ML20055J3221990-07-25025 July 1990 Decommissioning Funding Assurance Rept & Certification ML20055G7051990-07-19019 July 1990 Rev 0 to Yankee Cycle 21 Core Operating Limits Rept ML20055F6751990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Yankee Atomic Power Station 1999-08-31
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O nuay[o g UNITED STATES
! w g NUCLEAR REGULATORY COMMISSION
- g. t WASHINGTON, D. C. 20555 g / SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 108 TO FACILITY OPERATING LICENSE NO. DPR-3 1
YANKEE ATOMIC ELECTRIC COM'ANY YANKEE NUCLEAR POWER STATION DOCKET NO.50-029 INTRODUCTION By application dated September 15, 1987 Yankee Atomic Electric Company (YAEC or the licensee) requested an amendment to the Facility Operating License (0L) for the Yankee Nuclear Power Station (YNPS). The proposed amendment would change the expiration date for the license from November 4, 1997 to July 9, 2000, an extension of two years and eight months.
The amendment request was supplemented on December 2,1987, at the request of the staff, to provide information relevant to the provisions of the National Historic Preservation Act.
I OISCUSSION AND EVALUATION Sectinn 103.c of the Atomic Energy Act (Act) of 1954 provides that a license is to be issued for a specified period not exceeding 40 years. 10 CFR 50.51 specifies that each license will be issued for a fixed period of time, to be specified in the license, not to exceed 40 years from date of issuance. 10 CFR 50.57 allows the issuance of an operating license pursuant to 10 CFR 50.56 for the full term specified in 10 CFR 50.51 in conformity with the construction permit (CP) and when other provisions specified in 10 CFR 50.57 are net. The current term of the license for the YNPS is 40 years commencing with the issuance of the CP. This represents an effective operating term of 37 years and 4 months, not 40 years. Consistent with the Act and our rules, as noted above, the licensee seeks an extension of the OL term for YNPS such that the fixed period of the license would be 40 years from the date of issuance of the OL, Current NRC policy is to issue operating licenses for a 40 year tenn, commencing with the date of issuance of the OL, For YNPS this date was July 9, 1960. Thus a 40 year term would change the expiration date from November 4, 1997 to July 9, 2000 for an extension of two years and eight months, the interval between issuance of the CP and OL.
The licensee's request for extension of the operating license is based, in part, on the fact that a 40-year service life was considered during the design and construction of the plant. Although this does not mean that some components will not wear out during the plant lifetime, design features were incorporated which maximize the inspectability of structures, systems and equipment. Surveillance, inspectability and maintenance practices which were implemented in accordance with the ASME Code for Inservice Inspection and Inservice Testing of Pumps and Valves and the facility Technical Specifications provide assurance that any unexpected degradation in plant equipment will be identified and corrected. The specific provisions and requirements for ASME Code testing are set forth in 10 CFR Part 50.55a. The NRC, in its most recent 8806210255 080609 PDR ADOCK 05000029 Y, D )[
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i Systematic Assessment of Licensee Performance (SALP) Report, remarked that '
initiative was exercised to address plant aging and to increase oversight and effectiveness of ovality programs. Throughout this decade the YNPS has attained SALP ratings among the highest for all operating plants in the United States.
We have completed our review of the YNPS reactor vessel in regard to fracture toughness requirements for protection against pressurized thermal shock events as required by 10 CFR 50.61. We found that the reactor vessel meets the fracture toughness requirements of 10 CFR 50.61 for 3? effective full power years of operation. As the reactor has an operating factor of 74% this would translate into at least 43 years of operation at this 74", factor. In addition, the rule provides a pressurized thermal shock screening criterion of 270 F maximum for the critical component in the YNPS reactor vessel; the actual value is 253*F as derived from the equation specified in the rule.
This evaluation was provided in our letter to YAEC of March 10, 1987. We find that the reactor vessel for the YNPS meets the criteria of 10 CFR 50.61 for the requested licanse extension to a 40 year operating life.
Aging analyses have been performed by the licensee for all safety-related electrical equipment in accordance with 10 CFR 50.49, "Environmental qualification of electrical equipment important to safety for nuclear power plants, "identifying qualified lifetimes for this equiprent. These lifetimes have been incorporated into plant equipment maintenance and replacement practices to ensure that all safety-related electrical equipment remains qualified and available to perform its safety function regardless of the overall age of the plant.
The staff's Safety Evaluation for environmental qualification of safety-related electrical equipment was issued by letter dated December 15, 1985. A subsequent audit of the program was conducted October 22-24, 1986 by the Office of Inspection and Enforcement, the results of which are documented in a December 5,1986 Inspection Report. The staff has concluded and the inspection team verified that the licensee had implemented an environmental qualification program meeting the requirements of 10 CFR 50.49.
The Systenatic Evaluation Program (SEP) was initiated at YNPS in February 1977 by the U.S. Nuclear Regulatory Commission to review the designs of older operating nuclear reactor plants to confirm and document their safety. The review provides (1) an assessment of how these plants compare with current licensing safety requirements relating to selected issues, (2) a basis for deciding on how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The review was completed in July 1987. The review is documented in NUREG-0825, "Integrated Plant Safety Assessment" dated June 1983 and Supplement No.1, dated October 1987.
We concluded that the plant either meets current safety standards or will provide an equivalent level of safety once nodifications resulting from seismic reevaluation are completed. The licensee's commitments in this area have been found acceptable by the staff.
1 A service life well in excess of 40 years is anticipated for the Yankee facility structures. Inspection of critical structures has identified no signs of deterioration in structural integrity. Considering the experience in other industries with similarly designed structures, the conservatisms inherent in the design, construction, and operations of the facility, and the !
adequacy of the Yankee Preventive Maintenance Program to ensure refurbishment )
/,
. and/or replacement as necessary to maintain the margins of safety identified in the Technical Specifications, an additional two years and eight months of operation will have no significant impact on plant safety.
The Staff has also reviewed the Final Safety Analysis Report (FSAR) for the plant. Many safety related changes have been made to the plant since it went on line in 1960. Major safety related changes are:
-Ar improved emergency core cooling system j
-Addition of three emergency diesel generators !
-Addition of two new emergency feedwater pumps l
-Adding a safety parameter display system
-Adding an independent safe shutdown system
-Adding a solid-state reactor protection system and feedwater control system.
1 Each of these changes, where it involved a safety-related component, has been reviewed and approved by the staff; further, as required by 10 CFR 50.71(e),
these changes and their effect on accident analysis, if any, are routinely i updated in the FSAR. Our review of the FSAR for the facility has not I identified any concerns associated with approval of the proposed amendment to l extend the expiration date of the license that are not already addressed by licensee commitments, operating procedures, and license requirements.
The licensee provided a reference to the YNPS Probabilistic Safety Study in the September 15, 1987 submittal. However, the findings of this stt.dy were not applied to the licensee's safety analysis for the license extension. As l the licensee has provided sufficient information for our review in regard to the safety analysis supporting the license extension and as the NRC has not completed its review of the Study, we agree that the Study results need not be considered in the license extension.
The Exclusion Area for the YNPS consists of property owned by YAEC or the New England Power Company except for a small parcel owned by the Deerfield Specialty Paper Company. The licensee has the authority to control activities within the Exclusion Area and anticipates no changes to the Exclusion Area boundary during the extended license period. Changes in population within the Low Population Zono (LPZ), nearest population center distances and 10 mile radius Emergency Planning Zons (EPZ) have been evaluated by the staff and have been fo.ind not to be significant for the period of the license extension. The details of the staff's review are contained in the associated Environmental Asses w nt dated June 2 , 1988.
Accordingly, the Commission's conclusions regarding 10 CFR Part 100 siting criteria for the YNPS are that the exclusion area, LPZ, and population center distances meet the guidelines of 10 CFR Part 100 and are not changed by the proposed license extension.
Based on the above, it is concluded that extension of the operating license for tne YNPS to allow a 40-year service life is consistent with the Integrated Plant Safety Assessment in that all issues associated with operational safety and pcoulation changes have already been addressed. Accordingly, we find the proposed extensinns of the expiration dates of the Facility Operating Licenses for YNP3 to be acceptable.
ENVIRONMENTAL CONSIDERATION A Notice of Issuance of an Environmental Assessment and Findina of No Significant Impact relating to the proposed extension of the Facility Operating License expiration date for the Yankee Nuclear Power Station was published in the Federal Register on June 9,1988(53FR21743).
CONCLUSION The staf# has concluded, based on the considerations discussed above, that:
(1) there is reasorable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (?) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendrant will not be inimical to the common defense and security or to the health and safety of the public.
ACKNOWLEDGEMENT Principal Contributor: Morton 9. Fairtile Dated: June 9, 1938 i
i 1
- 3. This license amendment is effective as.of the date of.its issuance, i FOR THE NUCLE REGULATORY COMMISSION Richard H. Wessman, Director Project Directorate'I-3 Division of Reactor Projects 1/II Date of Issuance: June 9, 1988 t
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Amendment to Facility Operating License DPR Yankee Nuclear Power Station DISTRIBUTION:
Docket File 4 NRC & Local PDRs PDI-3 r/f SVarga BPoger MRushbrook MFairtile OGC-WF DHagan Edordan JPartlow TBarnhart (4)
Wanda Jones EButcher ACRS (10)
GPA/PA ARM /LFMB Region I l
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