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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 ML20213G9161987-05-13013 May 1987 Safety Evaluation Supporting Amend 105 to License DPR-3 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 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 ML20198D2541992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Recertification Program ML20198D2481992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Initial Certification 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 ML20055G6801990-07-31031 July 1990 Yankee Plant Small Break LOCA Analysis ML20055G7011990-07-31031 July 1990 Yankee Nuclear Power Station Core 21 Performance Analysis ML20055E1591990-07-31031 July 1990 Reactor Pressure Vessel Evaluation Rept 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
[Table view] |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING ACCEPTANCE OF CONFIRMATORY ECCS SMALL BREAK ANALYSIS i
AMENDMENT NO. 25 TO FACILITY OPERATING LICENSE NO. DPR-3 YANKEE NUCLEAR POWER STATION (YANKEE-ROWE) $
i DOCKET NO. 50-29 ,
Introduction 1 By application dated May 11, 1976, and supplements dated May 11, 17, 19, 1976, and June 16, 1976, Yankee Atomic Electric Company (the licensee) ,
oroposed a change to the Technical Specifications appended to License No. 2 D?R-3 for the Yankee-Rowe reactor. The proposal involved a change of the '
set pressure for the safety injection accumulator, specified in Section ;
3.2.e(4), from 410 psig to 337 psig 10 psig. The staff's safety evaluation, .
de ed May 19, 1976, concluded that adjusting the accumulator pressure to 1 a:ch previous ECCS calculations was acceptable for large break LOCAs.
Although the analyses submitted in May 1976 were adequate to demonstrate
' ht the revised accumulator pressure resulted in ECCS performance in
. qy conformance with the Commission's Acceptance Criteria for large breaks 5 ar.d for small breaks of 4-inch diameter or less, these initial analyses did not provide computations for small breaks larger than 4-inch diameter. y$ pt The licensee asserted that the larger breaks (e.g. 5-inches) would result in faster depressurization and more rapid injection of emergency core %n
- colant which would result in more rapid recovering of the core. The 4 Staff's May 19, 1976 safety evaluation concluded that the 5-inch break is
'.icely to be less limiting than the 4-inch break. The Staff accepted the g
Q li:ensee's correction of accumulator pressure on an interim bases, provided Mr that the licensee submitted improved small break LOCA analysis to confirm that the most limiting small breaks had been identified and conformed $g to the Commission's Acceptance Criteria. My The licensee submitted such analyses on June 16, 1976. This Safety Evaluation h l describes the Staff's evaluation of this confinnatory analysis. &(
7 :
Staff Evaluation orior to the June 16, 1976 submittal, small break analyses were performed y i
by the licensee using the WFLASH code and submitted to the NRC on July 31, Qk I
1974. These analyses considered 2, 3 and 4-inch diameter breaks. Results of these analyses indicated that the core never became uncovered for the k
Jih 2" break. A peak clad temperature (PCT) of 7500F was calculated for the n.1 3-inch break and the 4-inch break was determined to be limiting with a PCT of 13000F occurring 235 seconds after break initiation, W
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The 4-inch break was re-analyzed using the ENC WREM based evaluation model, Results of this analysis were included in the licensee's October 10, 1975 g(
application for a license amendment. The calculated PCT during the period g
of core recovery was determined to be 1400 F at 230 seconds. However, a 4
U fff PCT of about 1850 F was calculated to occur at about 9 seconds due to a $
momentary flow stagnation. Small breaks larger than 4-inch diameter had Ki not been previously conducted using the SLAP code in 1971 which indicates that the core never uncovered.
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The licensee's confirmatory small break analyses were conducted with the f$$
M approved RELAP-4-EM SMALL BREAK code (RELAP4-EM/00311/07/75 95 ENC 20). .r For fuel heatup calculations, the approved T00DEE2 ENC 13 Code was used.
Small breaks of 2.00, 2.25, 4.0,5.0 and 10 inch ID were analyzed.
Corrected accumulator pressure and flow resistances were used. Direct h spillage of ECCS fluid from the safety injection line in the broken loop ;;
to the containment floor was assumed to occur for the 2.25 inch ID and "Y all larger breaks. The 2.25 inch ID break corresponds to a complete -C severence of a safety injection line. No direct spillage of ECCS fluid [tg to the containment was assumed to occur for breaks smaller than 2.25- ru inches ID. In these cases, the reactor coolant system (RCS) pressure is felt in each loop of the ECCS system including both pumps and the $ M accumulator. A The peak clad temperature analysis was performed using the T00DEE2 y
digital computer code. For the 4, 5 and 10-inch ID breaks, the peak X power fuel rod was divided into 11 axial nodes. Initial fi el rod &n temperature distributions were obtained from a steady-state analysis using a RELAP4-EM/ HOT CHANNEL model with a corresponding axial nodalization.
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Time dependent fluid conditions required as T00DEE2 inpi . came from the blowdown results.
M.0 For the 2.0 and 2.25-inch ID breaks, in which the upper portion of the 5" core was uncovered for a longer time period relative to the rest of the core, a 19 axial node model was used to simulate the power rod. This q%g 1
permitted a detailed representation of the upper portion of the rod which experiences a relatively long heatup period. h Q
kg The limiting small break LOCA was that resulting from a safety injection 8 (2.25-inch ID) pipe rupture for which the peak clad temperature was calculated to be 18720F, occurring at 1170.2 seconds. Previous analyses d
had shown the 4.0-inch ID pipe rupture to be limiting.
M h-The reason for the shift in the limiting small LOCA break size was due $glB to the fact that a smaller break size, the 2,25-inch ID safety injection d6 oice rupture, was identified in which direct spillage of ECCS fluid could occur.
g This break results in a slower RCS depressurization such that ECCS 2 injection into the intact loops occurs much later in the accident that in @
the 4-inch break, thereby allowing a substantially longer heatuo period. 7 9
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u For the 4, 5 and 10-inch 10 breaks, depressurization was fast enough to allow the ECCS system to inject into the RCS sooner and to terminate the transient. For breaks smaller than 2.25-inches ID (size of the thermal sleeve in the ECCS penetration into the RCS), no direct spillage produced m:re effective ECCS performance. ;
Ccoclusions a 1
From its review of the June 16, 1976, confirmatory small break analysis, ,
the Staff concludes: (1) the analysis acceptably covers the small break '
spectrum and identifies the limiting small break size, (2) the correct ECCS accumulator parameters (pressure and flow) were used in the calculations, and (3) the analysis methods and results are acceptable and do not involve a decrease in the safety margin for the Yankee-Rowe Core XII _
ECCS performance evaluation.*
Environmental Considerations J.
5 We have determined that the amendment does not authorize a change in - /
effluent types or total amounts nor an increase in power level and will ?
not result in any significant environmental impact. Having made this ;j determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental af ircact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact M
statement, or negative declaration and environmental impact appraisal $
(As need not be prepared in connection with the issuance of this amendment.
209 Cl usi on Q'
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4 We have concluded, based on the considerations discussed above, that: (1) %.
because the amendment does not involve a significant increase in the $
or:bability or consequences of accidents previously considered and does M cot involve a significant decrease in a safety margin, the amendment does not involve a significant hazards considerations, (2) there is reasonable $
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'5uosequent to the review reflected in this evaluation, we obtained new information requiring a modification of upper head water temperature M'
assumed in the ECCS evaluation for Yankee-Rowe. As a result, the NRC $$*
on August 27, 1976 issued an Order for Modification of license, restricting %.
Erh the ceak linear heat generation rate to 0.85 kw/ft, and requiring a d' revised ECCS evaluation.
M me This amendment correc. ting accumulator pressure, does not affect the re:virement or cunclusions of the Order of August 27, 1976. The only h W
effect the Order has on considerations discussed in this safety evaluation NF would be to maintain or reduce the peak clad tempe'rature set forth herein.
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$. s assurance that the health and safety of the public will not be endanoered j by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance 7 B
of this amendment will not be inimical to the common defense and security ($
. or to the health and safety of the public. A Date: October 7,1976 s
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