ML20070J814
| ML20070J814 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/18/1994 |
| From: | Miltenberger S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLR-N94116, NUDOCS 9407250330 | |
| Download: ML20070J814 (18) | |
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1 Pubhc Servico E?octric and Gn Cempny St: yen E Mittenberger Punt c servico Electoc and on Company P.O. Box 236. Hancocks Bridge. NJ 08038 609-339-1100 V(e Pra nh'*! # ) O ;' N ' -
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NLR-N94116 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REPLY TO A NOTICES OF VIOLATION INSPECTION REPCRT NO. 50-354/94-09 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Pursuant to the provisions of 10CFR2.201, this letter submits the response of Public Service Electric and Gas Company to the notice of violation for the three violations issued to the Hope Creek Generating Station on June 15, 1994.
As required by the notice of violation and 10CFR2.201, this response includes a written statement or explanation in reply, including, where applicable, taken and the results achieved,the corrective steps which have been be taken to avoid further violations,the corrective steps which will compliance will be achieved.
and the date when full attachment to this letter.
This information is provided in the Shculd you have any questions or comments on this transmittal, not hesitate to contact us.
do Sincerely,
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9407250330 94071D PDR ADOCK 05000354 g
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JUL181994 Document Control Desk 2
NLR-N94116 C
Mr. T. T. Martin, Administrator - Region I U.
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Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. J.
C.
Stone, Licensing Project Manager - Hope Creek U. S.
Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. C. Marschall (S05)
USNRC Senior Resident Inspector Mr.
K. Tosch, Manager, IV NJ Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 I
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REF:
NLR-N94116 STATE OF NEW JERSEY
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COUNTY OF SALEM
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S.
E. Miltenberger, being duly sworn according to law deposes and says:
I am Vice President and Chief Nuclear Officer of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Hope Creek Generating Station, are true to the best of my knowledge, information and belief.
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Subscribed and Swo n-to before me this
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day o dV 1994
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i ATTM3MNr REPLY 10 A }KTf1G OF VIOIATim INSPELTICH RERRP NO. 50-354/94-09 norE MEEK GENERATING SfATIm FNMJTY OMRATING LICENSE NPF-57 DO NET. NO. 50-354 NIR-N94116 I.
Il(TRODUCTION During inspection activities conducted between March 27 and April 30, 1994, the NRC identified three potential violations of NRC requirements. These potential violations were subsequently documented and cited in Inspection Report 354/94-09 dated June 15, 1994. Our response to those three cited violations is provided below.
II. REPLY 70 IK7PICE OF VIOLATION FOR CDifrAINME?IT IlffEGPATED LEAK RATE TESTING A.
Descriotion of Violation "10 CFR 50, Appendix B, Criterion XI requires, in part, that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service is identified and performed.
Procedure HC.RA-IS.ZZ-0008(Q), Revision 0 prerequisites require that non-seismic portions of the control rod drive system be vented durirg Containment Integrated Leak Rate Testing (CIIRT).
10 CFR 50, Appendix J, Section II.A.1.b requires closure of containment isolation valves for the Type A test by normal operation and without any preliminary exercising or adjustments.
Contrary to the above, prior to Containment Integrated Icak Rate (Type A)
Testing on April 11 and 12, 1994, the non-seismic portions of the control rod drive system were not vented, and the control rod drive directional control valves, listed as containment-isolation valves in the FSAR, were exercised prior to the Type A test.
This is a Severity Level IV violation (Supplement I)."
B.
Resoonse to Violation PSE&G denies this violation.
The following discussion provides the basis for our determination that our actions did not violate the provisions of Criterion XI of 10CFR50 Appendix B nor the provisions of 10CFR50 Appendix J.
l Page 1 of 15
Attachment NLR-N94116 Reply to Notice of Violation 1.
Closira the CRD Vent Valves a.
Introduction and Summary
'Ihe folloaing discussion will demonstrate that, although applicable h w nts (i.e., Appendix J to 10CFR50, ANSI /ANS-56.8-1987, the UFSAR, the Hope Creek CIIRT procedure, and the applicable NRC inspection procedures) require venting of systems that may be open to the outsi&.
atznosphere under post accident conditions, these same h=nts contain provisions which allow closing the subject penetrations if their leakage interferes with the successful empletion of the CIIRP.
b.
Backaround Information relative to the applicable system design and the relevant regulatory and licensing basis requirements for containment integrated leak rate testing at Hope Creek is provided as follows. 'Ihis information will form the basis for the subsequent discussion of the specific cited violation.
1.
Containment Isolation Reauirements
'Ihe provisions of 10CFR50, Appendix A, General Design Criteria 55 require each line that is part of the reactor coolant pressure boundary and that penetrates primary containment be provided with one of four specified isolation valve configurations unless it can be demonstrated that the containment isolation provisions for a specific class of lines are acceptable on sme other defined basis. Due to the unique function of the CRD system, the the CRD insert and withdrawal penetrations are not providM with one of the four specified isolation configurations. As described in UFSAR Section 6.2.4.3.1.1.5, containment isolation provisions for these penetrations have been deucro boted to be an acceptable "other defined basis" for containment isolation. Specifically, autcmatic or locked closed isolation valves for the CRD insert and withdrawal' lines have not been provided in ortler to preclude any possible failure of the scram function. In lieu of autmatic or locked closed isolation valves, each line is isolated by a set of directional control solenoid valves provided on each CRD hydraulic control unit located outside the primary containment. The lines that extend outside primary containment are 1 inch or smaller and terminate in systems designed to prevent out-leakage. 'Ihe directional control valves are listed as containment isolation valves for Penetrations P35A through D and P36A through D in UESAR Table 6.2-24.
l Page 2 of 15
Attachment NLR-N94116 Reply to Notice of Violation 2.
CIU& Pmuirements The prwisions of Appendix J to 10CFR50 (Section III. A.1.d),
ANSI /ANS-56.8-1987 (Section 3.2.1.5), the UFSAR (Section 6.2.6.1.4), the Hope Creek CIU E procedure (Steps 5.1.5 and 5.1.8 of HC.RA-IS.ZZ-0008(Q), Revision 0), and the applicable NRC inspection precedures (Inspection Requirement 02.04 of NRC Inspection Procedure 70307 and Inspection Requirement 02.01e of NRC Inspection Procedure 70313) require, in general tenus, that containment penetrations be in an alignment for Type A tests consistent with that which would exist foll wing a design basis accident. Specific requirements which are relevant to the CRD penetrations are dirowM as follws.
UFSAR Section 6.2.6.1.4.c requires the foll wing:
" Fluid systems that are part of the primary containment boundary and that may open directly to the primary containment or outside atmosphere under post-accident conditions must be opened or vented to the appropriate atmosphere duriry the test."
Section III.A.1.d of AppeMix J to 10CFR50 requires the follwirg:
"Those portions of fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident conditions and become an extension of the boundary of the containment shall be opened or vented to the containment atmosphere prior to and during the test."
Section 3.2.1.5 of ANSI /ANS-56.8-1987 states the follwing:
"To place the primary containment system in as close to pmt accident conditions as possible, those portions of the fluid systems that are part of the reactor containment boundary that may be open directly to the containment or outside atmosphere under post accident corx11tions shall be opened and vented to the appropriate atmosphere duriry the test."
Inspection Requirement 02.04 of NRC Inspection Procedure 70307 and Inspection Requirement 02. Ole of NRC Inspection Procedure 70313 specify the follwing:
"The system alignments for the CIU& are performed to reflect the conditions that would exist after a design basis LOCA."
Page 3 of 15
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Attachment
'NLR-N94116 Reply,to Notice'of Violation Although, as noted above, applicable documents require venting'of systems that may be open to the outside atmosphere under post accident conditions, these same h = ants contain provisions which i
allow closirg the subject penetrations if their leakage interferes
~l with the simaful cwpletion of the CIIRP. Specifically, Appendix J (Section III.A.1.a), ANSI /ANS-56.8-1987 (Sections 3.2.1.4 and 3.2.6.a), the UFSAR (Section 6.2.6.1), the applicable Hope Creek procedure (Steps 5.2.17, 5.2.18, 5.4.8, and 5.4.9), and the applicable NRC inspection procedures (Inspection Requirement 02.05a of NRC Inspection Procedure 70307 and Inspection Requirement 02.02 of.NRC Inspection Pro dure 70313) recognize that leaks which interfere with the successful ocmpletion of the Type A test can occur, and include provisions to remove i.
penetrations from the Type A test provided leakage through the penetrations is quantified as r-wy to determine the "as found" and "as left" leakage rates. 'Ihese documents do not exclude any penetration from this practice.
Section 6.2.6.1 of the UFSAR contains the following statement:
"If, during the performance of a Type A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory cotipletion of the test, the leakage paths should be isolated and the Type A test continued. A local leakage test must be i
performed before and after the repair of each isolated leakage path."
Section III.A.1.a of Appendix J states that, if potentially excessive leakage paths are identified during a Type A test which will interfere with the satisfactory costpletion of the test, or
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which result in the test not meeting the acceptance criteria, the test shall be. terminated and leakage through such paths shall be measured using local leakage testing methods. It continues by stating that the corrective action taken and the change in leakage rate determined frun the local leak and Type A tests shall be j
included in the report submitted in accordance with Section V.B.
Section 3.2.6.a of ANSI /ANS-56.8-1987 contains the following statement:
"If, during the performance of a Type A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory ccmpletion of the test, these leakage paths may be isolated and the Type A test continued until ccmpletion. A local leakage test shall be performed before and after the repair of each isolated leakage path."
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Attachment NLR-N94116 Reply to Notice of Violation Inspection Requirement 02.05a of NRC Inspection Procolare 70307 and Inspection Requirement 02.02 of NRC Inspection Procedure 70313 specify the following:
"If a penetration leaks excessively aM cannot be repaired prior to the CIIRP, it may be blanked off. The LIRT penalty factor is required to be detemined for this penetration in order to characterize the AF [as-fouM] condition.
In addition, the AL (as-left) integrated leakage rate must also be adjusted to ccarpensate for the leakage that will exist through this penetration after its repair. This is done by adding the post repair local leak rate of this penetration to the CIIRT results."
3.
CRD Leakage Monitoring UFSAR Section 6.2.4.4.6.3 discusses leakage monitority of the CPD system in lieu of performing Appendix J, Type C testing. This section contains the following statement:
"Furthermore, since the reactor pressure vessel and the non seismic portion of the CRD system are vented during Type A tests, leakage monitorirg of CRD lines will be providEd.
c.
Seauence of Events The non seismic portion of the CRD system was initially vented prior to the Type A Test as required by Procedural Step 5.1.5 (Attachment 1, Step 4.1.6). Following CRD system venting, a reduction in reactor vessel inventory was observed. The subsequent investigation revealed that water was leaking from the valves which had been opened to vent the non seismic portion of the CRD system, indicating leakage past the directional control valves. Once the Type A test pressure was reached, the water leakage through all directional control valves was measured and the penetrations were isolated by closirg the vents of the non seismic portion of the CRD system as permitted by Appendix J, ANSI /ANS-56.8-1987, the UFSAR, and the procedure.
These actions satisfied the requirement to measure water leakage through the penetrations durirg the CIIRP, but avoided jeopardizing the Type A test results by having an orgoing increase in the containment free air volume (decrease in reactor level) or by affecting the stability of the containment atmosphere by having to make up reactor level with water of a potentially different temperature.
1 Page 5 of 15
Attachment NLR-N94116 Reply,to Notice of Violation d.
Basis for Denvinct violation The Hope Creek CIIRT was conducted in accordance with applicable requirements and none of the associated actions violated NRC regulations or requirements. The basis for this statement is as follows:
1.
Closure of the vents was in accordance with the applicable provisions and requirements of Appendix J to 10CFR50, ANSI /ANS-56.8-1987, the UFSAR, the Hope Creek CIIRP procedure, and NRC inspection procedures.
2.
Closure of the vents did not require a detennination of an unreviewed safety question because the actions taken were in compliance with the UFSAR.
3.
The associated compensatory actions specified in Section III.A.1.a of Apperdix J were cmpleted (i.e, leakage through the subject paths was measured using local leakage testing methods, adjustments to equipment were made, and the corrective action taken and the change in leakage rate determined frun the local leak and Type A tests were included in the our Section V.~B report).
The potential violation has also been evaluated from the more general perspective of cmpliance with the provisions of 10CFR50, Appendix B, Criterion XI for our containment integrated leak rate test program.
The provisions of criterion XI include the following:
1.
"A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and perfonned in accordance with written test procedures which incorporate the requirements ard acx:eptance limits contained in applicable design documents."
2.
" Test procedures shall include provisions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed urder suitable environmental conditions."
3.
" Test results shall be documented and evaluated to assure that test requirements have been satisfied."
Page 6 of 15 l
Attachment NLR-N94116 Reply.to Notice of Violation The following discussion demonstrates that PSE&G complies with the provisions of 10CFR50, Appendix B, Criterion XI with respect to containment integrated leak rate testing.
1.
PSE&G has an established test program which assures that the containment will perfom satisfactorily. An element of the test program is the CIIRP which is performed in accordance with a written test procedum. The procedure used to perform CIIRP testing is HC.PA-IS.ZZ-0008(Q). h subject procedure incorporates the requirements and acceptance limits contained in applicable design documents. Throughout the procedure, the requirements and acceptance criteria contained in the Technical Specifications, the UFSAR, Appendix J to 10CFR50, ANSI-N45.4-1972, ANS56.8-1987, ard Revision 1 to IN-70P-1 were included.
2.
A@ropriate prerequisites are contained in Section 2.0 of the subject procedure and provisions are included in Section 5.1 for assuring that all prerequisites for the given test have been met (Steps 5.1.5 through 5.1.9), that adequate test instrumentation is available and used (Steps 5.1.10 through 5.1.11), and that the test is performed under suitable environmental conditions (Steps 5.1.2 and 5.1.5).
3.
N test results were documented and evaluated to assure that test requirements were satisfied.
L Cyclinct the Directional Control Valves a.
Introduction and Sumary h following discussion will demonstrate that, since the directional control valves were stroked as a maintenance activity to establish conditions which would be expected to exist when the valves would be called upon to perform their specified function, compliance with Appendix J to 10CFR50 was maintained.
b.
Backrrround h provisions of Appendix J to 10CFR50 (Section III.A.l.b),
ANSI /ANS-56.8-1987 (Section 3.2.1.4), the UFSAR (Section 6.2.6.1.1),
and NRC inspection procedures (Inspection Requirement 02.04 of Inspection Procedure 70307 and 02.01e of Inspection Procedure 70313) all prohibit cycling valves to improve leakage performance. These rh - nts do, however, allow for the performance of valve maintenance.
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Page 7 of 15
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Attachment NLR-N94116 Reply,to Notice of Violation Specifically,Section III.A.1.b of Appendix J to 10CFR50 states the follwing:
" Closure of containment isolation valves for the Type A test shall be acccraplished by normal operation ard without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve motor). Repairs of maloperatirg or leakiry valves shall be made as remry."
Inspection Requirement 02.04 of Inspection Procedure 70307 and 02.01e of Inspection Procedure 70313 specify the follwing:
" Closure of containment isolation valves (CIVs) for the CIIRP shall be accomplished by normal operation without any prelininary exercising (valve cyclirg) or adjustments (e.g., no tightening of valve after closure by the motor)." and " Repairs of any maloperating or leaking CIVs shall be made as necessary. The pre-repair leakage is to be measured to determine the LIRT correction factor."
It is noted that the CRD directional control valves are cycled many tim s (at least weekly) during normal operation.
c.
Secuence of Events For a number of days prior to the Type A test, the control rod drive system was in an operational condition other than that which vxuld exist durire normal pwer operation (the pumps were out of service with no fl w through the directional control valves). Once the Type A test alignment of the control rud drive system was accomplished (approximately 3 days prior to containment pressurization), flw in the system was in a direction opposite the normal flw direction (from the reactor vessel through the directional control valves ard out the l
open vents). With the CRD system out of service, sediment may have collected in the CRD lines.
As a result of the above noted conditions, it was suspected that some sediment had accumulated on the seats of the directional control valves. Maintenance was performed on certain directional control valves by cycling the valves with flow through the system. This maintenance activity was conducted in an attempt to flush away the ruli m nt and restore the conditions which would be expected at the time the valves would be called upon to perform their specified function. The combined leakage through all the directional control valves was measured before and after flushirg.
In addition, the amount of leakage was again measured with the containment at Type A test pressure prior to isolating the penetrations.
Page 8 of 15
Attachmsnt NLR-N94116 Reply to Notice of Violation d.
Basis for Denvirn Violation We agree that cycling the valves for the sole purpose of reducing leakage is a violation of Appendix J to 10CFR50; however, the directional control valves were stroked as a mai'atenance activity to establish conditions which would be expected to exist when the valves would be called upon to perform their specifitd function. Since the valves were stroked for maintenance purposes to establish the expected design basis condition, this activity does not constitute a violation of NRC requirements or regulations, e.
Root Cause and 03rrective Actions 7he CIIRT was perfonned prVperly in accordance with all applicable requirements. As a result, there are no root causes or required corrective actions.
I.II.
Fuel Pool Gate Seal Replacement Violation 1.
Description of Violation "Fkpe Creek Technical Specification 6.8.1.c requires, in part, that written procedures be implemented for refueling operations. Procedure HC.OP-IO.ZZ-0001(Q), Refueliry to Cold Shutdown, step 5.2.21, requires that the fuel pool gate inter-space drain valve be opened upon completion of refuelity activities. Nuclear Administrative Procedure NC.NA-AP.22-0015(Q),
Safety Tagging Program, step 4.1 requires that the job supervisor ensure that equipment has been appropriately tagged and is safe to work on before beginning a work activity.
Contrary to the above, on April 11-13, 1994, after campletion of refueling activities, licensee personnel failed to open the inter-gate drain valve, and performed maintenance on the spent fuel pool gate inner seal without an approved procedure, and manipulated the air supply valves without a tagcut."
This is a Severity IcVel IV violation (Supplement I)."
2.
Response to Violation PSE&G does not deny the violation.
A.
Root Causes The root cause of the loss of spent fuel pool inventory to the reactor cavity has been attributed to the following:
1.
The in-line check valves to the outer gate seals (la-V337, 338, 339, and 340) leaked. A factor contributing to this root cause was lack of periodic preventive or corrective maintenance for these valves.
Page 9 of 15 l
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Attachment NLR-N94116 Reply to Notice of Violation This root cause is addmssed by e)rrective Actions B.1, B.2, and C.1.
2.
The position of the air supply isolation valve for the seals (1-KA-V7068) was not properly controlled in that the valve was closed scuetime between completion of the gate installation activities on April 5 and completion of the inner gate seal replacement activities on April 13.
A factor that may have contributed to this root cause was an improper listing of the supply isolation valve as normally closed in the Tagging Request Inquiry System (TRIS). This TRIS error is believed to be an isolated occurrence. A second factor that may I
have contributed was failure to follcw procedures during the April 13 replacement of the inner gate seals.
This root cause is addressed by Corrective Action B.3 and B.S.
3.
The SFP gate inter-space drain valve, required to be open by Procedure HC.OP-IC.2.Z-0001(Q) was improperly left closed after gate installation; closure of this valve disabled the alarm for leakage into the inter-space rajicn between the two SFP gates.
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The contributire factor for this root cause was personnel error l
resulting from failure to adhere to procedures.
This root cause is addressed by Corrective Actions B.4 and C.4.
4.
The work order for gate seal replacement referenced the wrong i
procedure, did not include the correct procedure in the package, and did not require a tagout.
The contributing factor for this root cause was personnel error
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durirq preparation and processing of the work order package.
1 This root cause is addressed by Corrective Actions B.7 and C.4.
5.
Mechanical maintenance practices for SFP gate seal replacement did not reat mnagement exp@tions in that the job supervisor proceeded and the replacement was perfonned without an approved procedure.
The contributing factor for this root cause was personnel error in failing to adhere to management expectation relative to use of procedures.
This root cause is addressed by Corrective Actions B.5 and C.4.
Page 10 of 15
Attachment NLR-N94116 Reply to Notice of Violation 6.
Maintenance personnel repositioned the air supply valves (R-V327. E-V329, M-V339, M-V340) without a tagout as required j
by procedure.
The contributing factor for this root cause was personnel error in failing to adhere to management expectation relative to use of procedures.
This root cause is addressed by Corrective Actions B.5 anl C.4.
B.
Corrective Actions Taken and Results Achieval The followirg corrective actions have been completed.
1.
The operations deparbnent has initiated a work order to replace the in-line check valves to each of the four gate seals.
2.
Recurring tasks have been created for periodic replacement of in-line check valves.
3.
The TRIS for the air supply isolation valve has been changed to reflect a normal position of locked open.
4.
Operators responsible for the failure to reopen the gate inter-spa drain valve were counseled relative to management expectations and appropriately disciplined.
5.
The mechanical maintenance ergineer has reemphasized and discussed management expectations and the importance of procedure compliance with all members of the mechanical maintenance shop. Tim responsible supervisor was appropriately disciplined.
6.
The meWanical maintenance engineer has directed the creation of a new procedure dedicated to SFP gate seal replacement.
7.
The outage manager counseled the individual who prepared the work ortler package relative to management expectations and proper preparation and processing of work ortler packages.
8.
The mechanical maintenance engineer has instituted a work package evaluation program to ensure procedure effectiveness. This program consists of a weekly review of work packages for piecedure effectiveness by first line supervisors. 7he results of this program are reviewed by the senior supervisors and the mechanical maintenance engineer.
9.
A deficiency report was initiated arri dispositioned to evaluate the effect of flooding the reactor well drywell cavity with the drywell head installed; this evaluation showed no negative impact on the drywell head by hydraulic forces or water damage / corrosion etfects.
Page 11 of 15 L
Attachment NLR-N94116 Reply to Notice of Violation
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- 10. Periodic operator checks of SFP gate seal pressures have been instituted.
- 11. A recurring task for periodic calibration of the gate seal pressure gages has been created.
C.
Corrective Actions to Be Taken The followiry corrective actions will be taken.
1.
The in-line check valves to each of the four gate seals will be replaced.
2.
A new procedure dedicated to SFP gate seal replacement will be created and implemented.
3.
A design charge request to improve gate seal air supply reliability will be evaluated.
4.
The General Manager - Hope Creek Operations, will communicate management expectation and reemphasize the importance of procedure adherence to all Hope Creek Station personnel.
D.
Date When Full Compliance Will Be Achieved Full compliance has been achieved.
Jy. Refueliin Bridae Mis-Oceration Violation A Descriction of Violation "10 CFR 50, Appendix B, Criterion XVI requires, in part, that licensees establish measures to assure that significant conditions adverse to quality are prmptly identifial, corrected, the cause is determined, and corrective action is taken to preclude repetition.
Contrary to the above, corrective actions implemented for a previous occurrence involving mis-operation on the refuellig bridge on December 17, 1993 (in which the caunal factors were identified as failure to self-check and adhere to procedures) were not effective, in that on March 9, 1993, the refueling bridge was again mis-operated (i.e., operators inadvertently moved the refuelirq bridge while the mast was still extended and grappled to a dummy fuel load) due to failure to self-check and adhere to procedures.
1 This is a Severity Imel IV violation (Supplement I)."
I Page 12 of 15
r Attachmont NLR-N94116 Reply to Notice of Violation B.
Response to Violation PSE&G believes that corrective actions taken to address the refuel bridge misoperation incident were appropriate and adequate to prevent recurrence of similar incidents.
1.
Introduction and Sunmuy
% e two incidents differ in that the first occurred during fuel movement aM the secoM occurred during surveillance testing. Criterion XVI of Appendix B to 10CFR50 requires that a licensee take corrective action to preclude repetition of significant corditions adverse to quality. As a result of the first incident, PSE&G took prompt and cmprehensive corrective action based upon the root causes of that incident.
W e following discussion will demonstrate that, although our corrective actions did not prevent the subsequent mis-operation of the refuel bridge during surveillance testing, our corrective actions were comprehensive aM remain adequate relative to preventing repetition of mis-open tion during movement of fuel.
2 Backerround On December 17, 1993, the refueling platform bridge operator repositioned the refueling mast without properly verifying release of a grappled new fuel assembly. W e operator moved a new assembly from the fuel preparation machine to its spent fuel pool location. We operator verified the location, lowered the mast and released the grapple. Se operator lifted the mast to the full up position and proceeded to move the mast back to the fuel preparation machine to transport the next fuel n M ly. A GE fuel inspector noticed that the fuel assembly, which was initially lowered into its spent fuel pool location, was still attached to the grapple. W e operator was not aware that the assembly was still grappled.
Personnel from the Reactor Engineering Depart 2nent acted immediately to halt the fuel movement and placed the fuel asserrbly into the correct location.
It was determined that the mast and grapple functioned as required. W e cause of the nis-operation was that the operator apparently did not properly verify that the load was removed fram the grapple. We operator was counseled by Loth the Reactor Engineering Depart 2nent and the Operations Department on proyr verification aM attention to detail in operation of the refueling bridge. On December 20, maintenance corducted an additions 1 operational check on the grapple and mast. Maintenance detendnx2 that the mast aM grapple functioned properly in all aspectc of operation.
I Page 13 of 15
-1 Attachment NLR-N94116 i
Reply to Notice of Violation U
The causal factors for this incident were ide.tified as follows:
a.
The responsible refuel bridge operator was inattentive to detail and failed to adhere to pro dural requirements for proper verification of load status.
b.
Guidance and requirements relative to the duties of the spotter (independent verifier) were not clearly defined and resulted in the spotter being distracted by other duties.
The actions taken for this incident were as follows:
a.
The responsible operator was counseled and disciplined, b.
A night order was issued which ensured that, prior to operating the bridge, all applicable personnel were refamiliarized with the bridge controls and the relevant procedure and were required to review documents describing additional spotter requirements and industry refuel bridge incidents.
c.
A meno was written to notify Reactor Engineering Department personnel of the additional spotter requirements, d.
A procedure change was initiated to enhance the requirements for spotters during movement of fuel.
Although a routine activity and not an action taken as a result of this incident, prior to the fifth refueling outage, all operators were retrained on the operation of the bridge and completed training which included discussion of recent relevant industry events.
We believe that these corrective actions are adequate to preclude recurrence of a mis-operation of the refuel bridge during movement of fuel.
It is also noted that Hope Creek has performed two consecutive " perfect" core reloads (i.e., no mis-oriented fuel assemblies) which, based upon l
current industry standards, is considered exceptional.
3.
Seauence of Events 4
On March 9,1994, an operator inadvertently moved the refueling bridge approximately two feet in a horizontal direction while the mast was stil?
extendui and grappled to a " dummy" fuel burrile. The mast flexed as the bridge began to move forward. The operator immediately recognized the problem and returned the bridge to its original, unflexed position. The operator had been performing retest activities followirg DCP work when Reactor Ergineering Department and, General Electric personnel arrived to set the frame mounted auxiliary hoist upper limits for control red moves.
During setting of the limits, General Electric personnel requested the bridge be moved for easier passage to the fuel prep / channel area.
Page 14 of 15
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Attachment NLR-N94116 l.
Reply to Notice of Violation h causal factors for this incident were identified as follows:
i
- a. 'The responsible refuel bridge operator was inattentive to detail r
and failed to self verify the status of the last procedure step performed. This is considered an isolated personnel, performance
- issue, i-b.
The responsible operator was distracted by personnel performing other activities.
r The actions taken for this incident weru as follows:
a.
The responsible operator was counseled regarding self verification and maintaining focus on the task at hand.
4.
Conclusion 1
The corrective actions for the personnel errors associated with the first refuel bridge incident were both proper and adequate to prevent a similar incident (i.e, mis-operation caused by attention to detail during movement 3
of fuel). The corrective actions did not prevent the second incident i
because the two incidents differ in that the second occurred during surveillance testing of the bridge when independent verification techniques (i.e., a spotter) are not enployed. The. safety significance associated with mis-operation of the bridge during surveillance testing is minimal since the bridge is not carrying an actual. fuel bundle and the load is not carried over actual fuel bundles. We have a high level of confidence that had the operator error occurred during movement of fuel, the corrective actions associated with inproved spotter requirements wcxild have provided the necessary barrier to prevent mis-operation.-
t 4
1 2
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