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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 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 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 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 ML20055E1591990-07-31031 July 1990 Reactor Pressure Vessel Evaluation Rept ML20055G7011990-07-31031 July 1990 Yankee Nuclear Power Station Core 21 Performance 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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HAZARDS" ANALYSIS BY THE RESEARCH'AND' POWER REACTOR SAFETY BRANCH l h
DIVISION OF LICENSING AND REGULATION _ g
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j YANKEE ATOMIC-ELECTRIC COMPANY ;
i a PROPOSED CHANGE NO.I l0
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' Yankee Atomic Electric Company's license presently provides;that the reactor Q,]
s will, not be " operated ' above!15 Mw electric unless: the. boron concentration in the ' +
i main coolant; system 'is-less than 80 ppm. Yank'ee, pursuant to paragraph 3.A. of .
License No. DPR-3, has; requested authorization to perform a test-demonstration of t the-use of boriciacid at' higher concentrations.while operating the' reactor at full. =G "
a' power. KThis test would sontinue 'for .a period 6f 'from .two, to six ' weeks , with con-tinuation of the test beyond the initial two week. period depending on Yankee's 7=?
- evaluation.~cf data obtained during'the first two weeks.;
During the. test, the concentration of- boron in the system would be limited :;= d to' values less than 400 ppm. The negative reactivity which 'would be provided ' , j.g
- by 300. to 400 ppm of boron'in the coolant is approximately equivalent to that : W=
of the reactivity worth of the equilibriuni xenon concentration in the' reactor, . . . ==;
1.e. , 2.57. delta k/k. ;Tht ' ability 'to operate the reactor for short periods of
[:] y time with boric l acid:present would allow a portion of the excess reactivity to *
! be poisoned 'out during startup operations-unti1 xenon equilibrium is established.
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5 This might permit' an' increase in the initial reactivity 'of subsequent core load- :
ings with a consequent increase in core lifetime.
Boric acid has been used in the main coolant system during initial testing I of the Yankee. reactor and is injected into the system during the process of-cooldown. During the first year of operation, Yankee has had considerable ex- M perience in the .use of boric acid and hae reporti.d that it has had no problema j with' boron injection, residence or dilution. One phenomenon which has been ob-served'during operations at Yankee involving beric acid is that the' acid acts as a mild decontaminant and solvent of the established corrosion product film,
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and tends to increase the crud level in the coolant. However, this.has resulted >
in a beneficial effect in that the crud level after cleanup of the borated system has been found to te lower than before the sL adown, and a lower crud equilibrium
. level has. persisted for a period of time following's subsequent startup. Al-though there is no conclusive evidence to indicate whether or not the crud re-' -
leased as a consequence of boric acid injection tends to deposit on heat t'ransfer y"
.urfaces', Yankee ~1ntends to operate a cation purification syetem during the' pro-
.pj ed. tests'in order to hold the crud concentration to 10 ppm, and to minimize q this effect should it occur.
The - (n, alpha) reaction which would take. place with boron present in the l coolant with the reactor at power will result in some' disassociation' of water d ss a result of the ionizatica produced by the alpha particles. Yankee has Q considered the increased disassociation'which might result from operation un- ?
' der the conditions of the proposed teste.. They have concluded, and we agree, ,
that such 'an increase is not a cause for concern, since it would not be signi- q cficant in relation to the dieaesociation normally taking pla:e which has been- 1
. ade'quately ; controlled during previcue ope. rations.
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2-U There has been concern expreesed in the past that initially dissolved boron in the coolant might react under irradiation with other chemicals present in the primary system. Should this occur, the boron might initially be deposited on surfaces in the reacter (boron hideout), and then suddenly
" fall out" and cause a step increase in reactivity. There has been some experimental evidence that Tithium metaborate is one such insoluble com-pound that might be ' formed if lithium it ,irement in the reactor coolant.
However, since:the only additive which Yankee plans to have present in the primary system is- the boric . acid, there seems to be no possibility of a sudden removal of boron from the Yankee reactor due to possible chemical .
reactions. Although it is conceivable that other phenomena might aleo cause boron hideout, there is no known mechanism for such an occurrence in the Yankee reactor. Accordingly, Yankee believes, and we agree, that the only indisputable means of proving or disproving this possibility is to conduct a test demonstration under carefully controlled conditions of the use of boron in the coolant with the reactor at power. .
Yankee has stated that chemical analysis will be routinely performed -
during the proposed tests which will define the boron concentration in'the -
main coolant within 1% of the actual concentration present. In addition, by a
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monitoring control' rod withdrawal rates , any losses in reactivity greater than .43% delta k/k can be detected. The .43% delta k/k, therefore, _
represents an upper limit on the uncertainty in measuring reactivity. Yankee 1
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has otated that the test will not be continued past the initial two week c. .-
period if the value of the uncertainty in reactivity plus any unexplainable reactivity change exceeds .8% delta k/k. Thue , the largest unexplainable .
change which could be present at the end of the two weeks, and the test etill be allowed to continue, is .37% delta k/k.
In order *.o estimate the upper 7 im!.c of hazard from there proposed teste ,
Yankee has analyzed the conasquences cf ar. inetantaneoue insertion of 1.5%
delta k/k of reactivity into the reacter. They have concludsd, and we agree, that the resulting excursion would peak at about 200% of tte steady etate power, with a minimum DNB tatio greater than 1 at the time of the peak due .
to the inherent negative power coefficient of the reactor. The reacter would automatically be scrammed on high neutren flux (any two of three independenc channels could cause such a scram) and the integrity of the fuel v:uld not be impaired. As noted previously, Yankee har stated that the teet would not be extended past the initial two week period if the value of the uncertainty in reactivity measurement plus any unexplainable reactivity change exceeds .8% >
delta k/k. Since the resulting excursion following the instantaneous insertion of this amount of reactivity would be less severe than that caused by the 1.5%
step which was analyzed, we believe that the .8% delta k/k limit is adeauste to determine if the test should be permitted to continue at the end of cce initial two week period. Further, we believe that an additional condit:
should be impose.d that will require that the tests be discontinued if at say time during the testing perted the a:m of uncertainty plus unexplainable
. reactivity values exceeds .8% delta k/k. An existing license condition requires 1 that Yankee provide the Commitrian with a complets repert on the resulta and significance of these tests af ter they hsve been completed.
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Based on the foregoing' analysis, we have concluded that Yankee should be authorized tu perform the pr6 posed tests with the additional restriction that
. the: test be discontinued if at any time the sum of uncertainty _ plus the H unexplainable reactivity values exceeds .8$ delta k/k. With'this restriction, 6=
it is our . opinion that the proposed tests do not present significant hazards considerations not described or implicit in the license application and there is reasonable assurance that the health and safety of the patblic will not be endangered by operation of the facility as described above.
Original Signed b/
E. G. Case -
Edson G. Case, Chief Research & Power Reactor Safety Branch Division of Licensing and Regulation Date SEP 12 461 .
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