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Category:ABNORMAL OCCURRENCE REPORTS (SEE ALSO LER & RO)
MONTHYEARML20087L5711976-02-0202 February 1976 AO 76-3:during Refueling Outage from 751116-760108,nine Valves Exhibited Leak Rate in Excess of 2000 Cc/Minute. Caused by Loose Body to Bonnet Joint.Bolting Tightened ML20087K9381975-03-0808 March 1975 AO 50-266/75-4:on 750226,air Ejector Discharge Gas Monitor R15 Pegged High,Then Dropped Low.Operator Manually Secured RCS Letdown & Went to Auxiliary Bldg to Assist in Locating Suspected Leak ML20087L0911975-02-18018 February 1975 AO 75-3:on 750208,mechanical Latching Mechanism on Redundant Solenoid Trip Vent Valve 2018C on Main Steam Stop Valve a Failed to Unlatch When Solenoid Deenergized.Caused by Improper Rotation of Rollers as Designed ML20087L2091975-01-28028 January 1975 AO 75-1:on 750120,first of 4 Tanks of Sludge Lancing Waste Sampled & Iodine Levels Found Approx 9 Times Max Permissable Concentration.Caused by Fuel Cladding Leaks Plus Small primary-to-secondary Leak in Steam Generator a ML20087L2661974-11-22022 November 1974 Ao:On 741009,while Personnel Exiting Containment Via 66 Ft Level Personnel Airlock,Air Heard Escaping Following Closing of Inner Door.Caused by Open 3/8-inch Valve on Pressure Gauge Instrument Line.Valve Removed.Served No Purpose ML20087L3101974-08-30030 August 1974 Ao:On 740821,recheck of Control Panel Indicated Above Normal Operating Pressure & Water Level Conditions Existed in Blowdown Evaporator.Caused by Steam Leaking at Steam Supply Control Valve ML20087L3191974-08-27027 August 1974 Ao:On 740815,during Monthly Surveillance Test,Steam Generator B Level Channel 472,lo-lo Level Bistable Would Not Trip When Tested.Caused by Failed Silicon Controlled Rectifier.Rectifier Replaced ML20087L3251974-07-22022 July 1974 Ao:On 740711,rereview of Tech Spec Changes Allowed part-length Rods to Be Inserted Not More than 70%.On 740529,part-length Rods Inserted to Max Integral Worth, 82.5% & Remained in That Position Until 740608 ML20087L3441974-06-13013 June 1974 Ao:On 740325,valve Permitting 150 Gallons of Reactor Coolant to Enter Waste Holdup Tank Inadvertently Opened.Caused by Personnel Error.Operator Reprimanded & Reminded That Utmost Care Must Be Exercised on Valve Lineup Operations ML20087L3361974-06-0808 June 1974 Ao:On 740529,minor Leak Discovered in Main Steam Safety Valve & Secondary Side Steam Pressure Substantially Reduced to Reseat Safety Valve & Terminate Minor Leak. Caused by Incorrect Concentration of Borated Water ML20087L5481974-05-14014 May 1974 Ao:On 740509,auxiliary Bldg Stack Monitor R14 Alarmed at Setpoint During Refueling Activities of 4,000 Counts Per Minute Indicating Release of Radioactivity.Caused by Isolation of Vol Control Tank for Venting ML20087L5931974-05-0303 May 1974 Ao:On 740406,during Performance of Containment Isolation Valves Leak Tests,Two second-off Valves from Containment Found to Leak in Excess of 2000 Cc/Minute.Cause Not Stated. Leak Tests Ongoing.Valves Will Be Repaired & Retested ML20087L6031974-04-17017 April 1974 Ao:On 740407,while Running Motor Driven Auxiliary Feed Pump a to Maintain Steam Generator Water Levels,Operator Noted Pump Not Delivering at Adequate Feedwater Flow Rate.Caused by Strainer Fitted in Flanged Position ML20087L6111974-04-11011 April 1974 Ao:On 740404,following Completion of Operations Refueling Tests,Isolation Valves 1AOV-2084 & 1AOV-2083 Closed Automatically on Signal from R-19 Process Monitor,Indicating Liquid W/Radioactivity Above Normal Levels ML20087L6201974-04-0404 April 1974 Ao:On 730329,when Drain Cap Removed from Penetration 32,air Blew Lightly for Several Seconds & Showed No Signs of Diminishing.Caused by Leak of Sensing Line within Penetration Boundary.Line Will Be Repaired ML20087L6261974-02-0606 February 1974 Ao:On 740125,auxiliary Bldg Stack Monitor R-14 Alarmed Indicating Release of Radioactivity.Caused by Increased Pressure Attained in Holdup Tanks & Suction of Waste Gas Compressors ML20087L6971973-11-0303 November 1973 Ao:On 731013,boric Acid Storage Tank B Placed in Svc & Boric Acid Tank a Concentration Known to Have Fallen Below Tech Spec Limit Immediately Prior to Transfer of Tanks. Caused by Two Personnel Errors ML20084A0901973-10-29029 October 1973 Ao:On 730915,following Investigation of Excess Letdown Line Leak,Valve 1MOV-1299 Downstream Seat Protruded from Valve Body & Crack Discovered in Upper Portion of Downstream Seat Ring.Cause Unknown.Vendor Notified of Valve Problem ML20087L6291973-07-16016 July 1973 Ao:On 730708,high Level Alarm on Facade Sump Received in Main Control Room.Caused by Small But Steady Flow Into Sump from Unknown Source.Retransfer of Borated Water from Refueling Water Storage Tank Instigated ML20087L6311973-05-28028 May 1973 Ao:On 730512,to Expedite Shutdown & Draining of Radwaste Sys Waste Evaporator,Operating Personnel Manually Lifted One of Two Evaporator Safety Valves.Following Shutdown,Valve Remained in Raised Condition.Procedure Pbnp 4.17 Expanded ML19317D4661971-08-11011 August 1971 Ao:On 710803,release of Radioactive Liquid Into Sewage Retention Pond Resulted from Overflow of Unit 2 Facade Sump During Filling Operation.Caused by Problems W/Facade Drain 1976-02-02
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARNPL-99-0569, Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with ML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 NPL-99-0051, Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with NPL-99-0449, Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20209D2691999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbnps,Units 1 & 2 ML20196F3341999-06-22022 June 1999 Safety Evaluation for Implementation of 422V+ Fuel Assemblies at Pbnp Units 1 & 2 ML20195F9781999-06-10010 June 1999 Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1 ML20209D2751999-05-31031 May 1999 Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0328, Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with NPL-99-0273, Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With ML20196F3521999-04-30030 April 1999 Non-proprietary WCAP-14788, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power) NPL-99-0193, Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with NPL-99-0134, Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0008, Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with NPL-99-0091, 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with1998-12-31031 December 1998 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 NPL-98-1006, Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20195J5101998-11-16016 November 1998 Proposed Revs to Section 1.3 of FSAR for Pbnp QA Program ML20198J5941998-11-0303 November 1998 1998 Graded Exercise,Conducted on 981103 NPL-98-0948, Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With NPL-98-0880, Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored1998-10-21021 October 1998 Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored ML20154M9121998-10-14014 October 1998 Unit 1 Refueling 24 Repair/Replacement Summary Rept for Form NIS-2 ML20154L6751998-10-14014 October 1998 Unit 1 Refueling 24 ISI Summary Rept for Form NIS-1 NPL-98-0826, Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20151W3851998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Pbnp Units 1 & 2 NPL-98-0653, Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4471998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 2 ML20151W4541998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 1 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 NPL-98-0558, Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 2 ML20151W4261998-06-30030 June 1998 Corrected Page to MOR for June 1998 for Pbnp Unit 2 ML20151W4221998-05-31031 May 1998 Corrected Page to MOR for May 1998 for Pbnp Unit 2 NPL-98-0481, Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4011998-04-30030 April 1998 Corrected Page to MOR for April 1998 for Pbnp Unit 2 NPL-98-0356, Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20216D7071998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3981998-03-31031 March 1998 Corrected Page to MOR for March for Pbnp Unit 2 NPL-98-0209, Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable1998-03-30030 March 1998 Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant NPL-98-0159, Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3891998-02-28028 February 1998 Corrected Page to MOR for Feb 1998 for Pbnp Unit 2 ML20216D7121998-02-28028 February 1998 Revised Corrected MOR for Feb 1998 for Point Beach Nuclear Plant,Unit 2 NPL-98-0084, Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 21998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 1999-09-30
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3" July 22, 1974 Mr. John F. O'Lcary, Director Directorate of Licensing U. S. Atomic Energy Co=nission Washington, D. C. 20545
Dear Mr. O' Leary:
DOCKET NOT. 50-266 VIOLATION OF PART-LENGTl! ROD INSERTION LIMIT POINT DEACl! NUCLEAR PLANT This is to report the details of an abnormal occurrence at the Point Beach Nuclear Plant, Unit 1, Facility Operating Li-cense No. DPR-24, as defined by Section 15.1.a.B of the Technical Specifications. This written report follows a telephonc report on the subject to Mr. Dwano Boyd of Region III, Directorate of Regulatory Operations, on July 11, 1974, per Section 15.6.A.1 of tha Point Roach Nuclone Plant Technical Soccifications.
On July 11, 1974, during a re-review of recent Technical Specification changes, the Reactor Engineer at Point Beach Nuclcar Plant noted a change to Section 15.3.10.A.3, the new specification reading:
"The part-length rods shall not be more .
than 70% inserted." -
The dato at which this new limiting condition for opera-tion became offective was verified as May 23, 1974. A check of recent Unit 1 operating history showed that between May 29, 1974, at 11:00 A.M., and June 8, 1974, at 11:00 P.M., the part-length rods of this unit were 82.5% inserted.
On May 1, 1974, Wisconsin Electric Power Company and Mis-consin Michigan Power Company submitted proposed Technical Speci-fication changes (Technical Spccification No. 8 to Appendix A) ,
which required AEC approval prior to operation of the Cycle III core. A now reduced reactor coolant system pressure (2000 psia) was also involved in the changes.
The Manager's Supervisory Staff reviewed the proposed Technical Specification changes as required by Section 15.6.1.C.2.c.3 of the Technical Specifications and, as part of' this review, noted Section 15.3.10.A.3 which roads: .
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O Mr. John F. O' Leary July 22, 1974 "Part-length rods shall be inserted no fur-ther than the limits shown on Figure 15.3.10-la."
The review of the figure showed that part-length rods could be inserted up to 90% at and below the 50% power level; the insertion limit then being reduced on a straight line graph to 70% at full power.
During discussions between Wisconsin Electric Power Company Muclear Projects Office personnel and AEC officials in Washington, D. C. on May 21 and 22, 1974, it was agreed that the provisions of the proposed Section 15.3.10.A.3 would in fact be changed to restrict the insertion limit of the part-length rods to 70% at all power levels. The AEC approval of the Tech-nical Specification change, dated May 23, 1974, reflected the above restriction.
The Unit 1 refueling shutdown drew to a close on May 23, 1974, and the unit was in the hot shutdown condition await-ing approval of the proposed Technical Specification changes at that date.
To expedite nn-nite racaipt nf tha official approval of tne new Tecnnical bpecifications, details of the anu approval were transmitted via telephone,-along with a telecopy of changed pages, including page 15.3.10-2, which contained the 70% restric-tion, to the site where a licensed operator operating document, consisting of the proposed Technical Specification Change No. 8 and the telecopied pages, was assembled on instruction of Wiscon-sin Electric Pouer Company Nuclear Projects Offico personnel.
Not being informed otherwise, the on-site operating document in-cluded the part-length rod insertion limit graph, Figure 15.3.10-la.
In reviewing the operating document prior to initial cri-ticality, the Reactor Engineer noted the agreement of the attached graph with that submitted in the Technical Specification proposed-change and assumed the operating document was applicable. The changed words of Section 15.3.10.A.3 were not highlighted by the normal proposed change marking (/8) as is the finally approved version eventually distributed at the plant.
To reduce the boron concentration during initial criti-cality and initial power escalation, the Reactor Engineer directed that the part-length rods be inserted to their maximum integral worth (8 2. 5 % , 40 steps) on May 29, 1974, and they -remained in that position until 'une 8, 1974, when they were withdrawn above the full power insertion. limit of 70% in' anticipation of a power in-crecsc. During the eleven days-that the part-length rods were i J
below the insertion limit, the reactor was at zero power for al-
-l I
O . . . .
O Mr. John F. O' Leary - 3- July 22, 1974 most ten days of the period, this being followed by one day at a 25% to 30% power level and approximately two hours at 50%
power.
In evaluating the nuclear safety aspects of this viola- ,
tion, the Reactor Engineer reviewed WCAP-8325, "The Nuclear De- :
sign-Core Management of the Point Beach Unit 1 Nuclear Reactor Cycle 3". On page 5-7 of this analysis, it stated that the pro-posed insertion limit (as shown on graph 15.3.10-la of the pro-
. posed Technical Specification change) was " based on the core power peaking analysis performed to develop the Fg flyspeck".
The flyspeck, Figurc 6.1, is used conservatively Therefore, to set the from axial offset limitations for accident analysis.
a safety viewpoint, provided the limits on Figure 15.3.10-la were observed, the core power distribution met the accident analysis criteria. The limits of 15.3.10-la were in fact ob-served at all times. Therefore, it is not considered that this violation posed a hazard to the health and safety of the public.
The primo cause of this incident was delay experienced in the process of the vendor producing the core analysis and the reviews by Company personnel, followed by the processing time of the review rnenugn rho Arc. In order to kcon the :pprev21 F?r?"
ahead of the reactor restart schedule, oral and telecopier means of transmitting approvals and Technical Specifications were used.
This led to a mixup in what constituted the finally approved documentation.
In order to prevent a recurrence of this. type of situa-tion, the Company will do its best to file similar requests with the AEC on as timely a basis as can -be anticipated, and consis-tent with Mr. Karl R. Goller 's letter of May 23, 1974.
Very truly yours, JW ,
Sol Burstein Executive Vice President cc: Mr. James G. Keppler, Regional Director Directorate of Regulatory Operations, Region.III
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