ML20245L069
| ML20245L069 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/15/1989 |
| From: | Kovach T COMMONWEALTH EDISON CO. |
| To: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 8907050364 | |
| Download: ML20245L069 (6) | |
Text
[]j Common :ealth Edison pJ (g' 3,-
< {, V 72 West Adams Street, Chicago. !!Iinois
.V X3 dress Repl to: Post Office B6M Y
. Chicago, Ilknois 6C690 0767 June 15,.1989
-Mr. A. Bert' Davis Regional Administrator U.S. Nuclear Degulatory Commission Region III' 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Braidwood Station Units 1 and 2 Response.to Inspection Report Nos.
50-456/89009 and 50-457/89009 HRC. Docket NoJ. 50-153_and_50-457 Reference (a):
E.M. McKenna letter to Cordell Reed dated May 16, 1989 Dear Mr. Davis This letter provides Commonwealth Edison's rerponse to the inspection conducted by Messrs. T.M. Tongue, T.E. Taylor and G.A. VanSickle from March 19 through April 29, 1989 of activities at Braidwood Station. Reference (a).
Indicated that certain activities appeared'to be in violation of NRC requirements. The Commonwealth Edison's response,to the Notice of Violation is provided in the enclosure.
I If you have any questions on this snatter, pleat;e direct them to this office.
Very truly yours, f
i qf g-T. J Kovach Nuclear Licensing Manager l
Enclosures cc: NRC Resident Inspector _- Braidwood bQ NRC Document Control Desk
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8907050364 890615
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0169T PDR ADOCK 05000456 o
PNU
'JUN 161983
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ENCIOSURE I
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il EDt@iONWEAIJ1LEDISQN_f0MP1ES RESEQUSE TO INSPECI1ON REPORT i
HQa, 456/89009 and 457/890Q2 1
I VlRLAIIDH (457/09009-01) 10 CFR 50.59 requires that a safety evaluation shall be performed when a proposed change or test is deemed to involve an unreviewed safety question (1) if the probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously i
evaluated in the FSAR may be increased; or (ii) if a possibility for an accident or malfunction of n dif ferent type than any evaluated previously in j
the safety analysis report may be created, or (iii) if the margin of safety as defined in the basis for any Technical Specification is reduced.
Section 7.6.10 of the Final Safety Analysis Report (FSAR) takes credit for the boron dilution prevention system (BDPS).
Source range l
instrumentation provides actuating signal for BDPS operation in the event of a flux doub11ng.
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Technical Specification 3.3.1 requires that the sourco range nuclear l
instrumentation be operable in Mode 5 unless special conditions exists.
Contrary to the above, the 10 CFR 50.59 modification review conducted on February 27, 1989, for an euxiliary feedwater system temporary alterat!on l
was deficient in that it failed to recognite that implementation of the alteration would render the boren dilution prevention system (BDFS) ineperable and incapable of performing its intended function for approximately thirteen days (between February 28 and March 13, 1989). During that time, the reactor was in Mode 5 (cold shutdown).
A.
REEEONSE Commonwealth Edison acknowledges that an inadequate technical review of i
l Temporary Alteration (Temp Alt) 89-2-008 was performed.
The specific purpose of the Temp Alt was to prevent inadvertent automatic actuation of the Auxiliary Feedwater System due to low steam generator water level.
The inadequacy of the technical review of Temp Alt 89-2-000 resulted in the Boron Dilution Prevention Syrtem (BDPF) being rendered inoperable.
l Sections 7.6.10 and 15.4.6.3 of the Byron and Braidwood UFSAR take credit for the automatic initiation of the BDPS ir,. the event of an inadvertent boron dilution during Co3d Shutdown (Mode 5).
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Unit 2 was in Mode 5 at the time of the occurrence.
If an inadvertent boron dilution event had occurred during this perind, the Source Range Instrumentation would have detected the flux increase and annunciated on l
the Main Control Board.
However, since Temp Alt 89-2-008 rendered l
automatic initistion of the BDPS inoperabic, manual initiation of BDPS would have been required by the operator.
B, f,CREECTIVE ACTIOf(_TAKEff AND RESULTS ACliIfYXD:
Upon identification BDPS operability was immediately restored by removing Temp l.lt 89-2-008.
C-CORBECTIVE ACTJON TAKEi!_IQ_&yOlpJ1RTiiER OCCURREtiCES:
Tho personnel involved in this event were included and participated in a Eraidwood Station Error Evaluation Presentation in order to identify the root and contributing causes of the event.
Based on the conclusions of this presentation, the following corrective actions will be taken to prevent recurrence.
For Temp Alts to Engineered Safety Features (ESP), Ecactor Protection, and other systems that could impact ESF actuations, an additional independent engineering review will be required.
Also the originator may be required to be present for the installation and/or removal of these Temp Alts.
Training session have been initiated for the appropriate operating personnel stressing the importance of requiring that adequate reference material be included with Temp Alt packages prior to approving installation, Also, this training includes >riginator notification should any discrepancies be identified d.uring the installation and/or removal of a Temp Alt.
D.
DATE_0L R LL_fDMPLIAtICE:
Training is expected to be completed by June 30, 1989.
These other corrective actions to avoid further occurrence are expected to be completed by September 1, 1989.
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HOLAUDJin (456/89009-01) 10 CTR 50 Appendix B, Criterion V states, activities affecting quality shall be prescribed by documented instructions or procedures of a type appropriate to the circumstances and shall be accomplished in accordance wit.h these instructions or procedures.
Procedure BwAP 300-1, Revision 3,
" Conduct of Operationr,", states:
"All operating personnel must be alert and remain within their immediate areas of responsibility until properly relieved and be responsible for monitoring the instrumentation and controls located in theiz areas. They are responsible for taking timely and proper action to ensure safe operation of the facility."
Contrary to the above, on April 16, 1989, operations personnel failed to adequately monitor instrumentation and controls and take timely and proper
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actions to prevent an inadvertent safety injection (SI).
This event occursed nt 4:40 p.m. during a normal plant heat-up and pressurization. The event was caused by actions initiated by the previous shift personnel and by failure of the on-shift operations personnel to adequtely monitor and control the system l
pressure increase. The SI automatically initiated when reactor coolant system (RCS) pressure was allowed to exceed 1930 psig prior to the heat-up reaching a secondary steam system pressure of greater than 640 psig. This resulted in about 5000 gallons of cold water from the reactor water storage tank being injected into the RCS, which was at about 500 degrees Fahrenheit.
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A.
RESPONSE
Commonwealth Edison acknowledges that the on-shift Unit Nuclear Station Operator (NSO) became distracted and was improperly supervised while attempting to restore a non-inking recorder during a plant heat-up and pressurization which resulted in an SI.
At 1450 on April 16, 1989, the on-shift NSO assumed responsibility for Unit 1 operation. He was informed ci the current unit status by the previous shif t NSO in accordance with Draidwood Administrative Procedure DwAP 335-1, Revision 6,
" Operating Shift Turnover and Relief".
This turnover included the status of the unit heat-up and pressurization.
The NSO continued the heat-up following the turnover in accordance with Draidwood General Procedure 1BwGP 100-1, Revision 1,
" Plant lleatup".
1DwGP 100-1 was physically located at the NSO's desk in accordance with DwAP 340-1, Revision 5, "Use of procedures for Operating Department".
Reactor coolant system (RCS) pressure and temperature were being monitored on a computer graphic display which indicates a trend of current temperature and pressure over a target line.
1 At 1530 the NSO observed a potential failure of'a RCS loop temperature
-element.
While investigating the failure, at approximately.1601 the NSO also identified that the data recorder associated with the temperature j
element was not functioning correctly and attempted to correct the
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i problem. While working on the recorder, the NSO was cognizant that RCS:
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temperature and pressure were deviating from the target value. He'was also' periodically monitoring upper nozzle temperature on another computer display,.as evidenced by making the required adjustments'to the ICV 121, Pressurizer Level Control Valve and the 1FWO34 A, B, C, - and D, Steam l
Generator. Level Control Valves'for each Steam Generator. The NSO judged i
that the deviation was not'significant and continued to work on the I
recorder. While deviation around the. target value is. acceptable, variations at higher pressures and temperatures ore'more significant.
As RCS pressure and temperature increased, the rate of' change was such that j
primary pressure r'eached the P-11 setpoint (1930 psig) before steam generator pressure increased above the low pressure SI reset (640 psig).
The' computer display that the NSO was observing does.not display the F-11 setpoint. The target value for RCS temperature at 1930 psig is 525'F.
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The minimum RCS temperature required at the P-11 setpoint to prevent an SI-is 494*F.
1 At 1639 the NSO went to the Unit desk to refer to IBwGP 100-1 for l
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direction to correct the temperature and pressure deviation..The SI l
occurred at 1640 before.any adjustments could be made.
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B.
CQRRECILVE ACTION TAKEN_AND._.EESETE_ACHIKYED:
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The following immediate corrective actions were takent-h 1.
The SI signal was reset and stable plant conditions were established.
2.
An engineering evaluation of the structural integrity of the pressurizer was performed for the temperature. transient. The evaluation concluded that the structural. integrity of the pressurizer.
was acceptable for continued operation.
3.
Westinghouse has performed an analysis of the effects.this event had on th? structural integrity of the RCS.
The analysis has concluded i
I that thm impact of this event on the structural integrity of RCS components is insignificant.
C.
CORRECIIYE_ ACTION TAKEN TO AVOIP FURTHER OCCURRENCE.5:
7he personnel involved in this event were included and participated in a Braidwood Station Error Evaluation Presentation in order to identify the-root and contributing causes of the event.
Based on the conclusions of this presentation, the'following corrective actions are being initiated to prevent recurrences
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1.
The Operating Department will develop and establish a formal policy'
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on the'use lof the extra NSO during startup and heat-up operations.
2.
1/2 BwGP 100-1 will be revised to establish a-hold point to. verify, that all steam generator pressures are greater than.640 psig.before-RCS pressure' exceeds'the P-11 setpoint.
3.
This event:will be ~ reviewed with appropriate ' Operating Department' personnel as part of the training associated with reactivity' management'.
4.
The heat-up and' pressurization ecmputer graphic display'will Ine-modified to include the setpoint for P-11..
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DATE_Qf_ FULL COMPLI ANCE:
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Procedure 1/2 BwAP 100-1 is expected to be revised by September 1, 1989.
.These other corrective actions to avoid further occurrence are expected to be completed by December 31, 1989.
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