ML20066F600

Responds to NRC 901129 Ltr Re Violations Noted in Insp Repts 50-454/90-23 & 50-455/90-23.Corrective Actions:Constr QC Verified Stated Installation & Performed Dimensional Insps on Supports After Work Completed

ML20066F600
Person / Time
Site:	Byron
Issue date:	12/31/1990
From:	Kovach T COMMONWEALTH EDISON CO.
To:	Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
NUDOCS 9101240061
Download: ML20066F600 (10)

Text

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of ,. Crenm:n:;ealth Edit n -

_- 1400 Opus Place

, Downers Grove, tilinois 60515' -

December 31, 1990 Mr. A. Bert Davis .

Regional Administrator- l U. S. Nuclear Regulatory Commiss)3n )

Region III 799 Roosevelt Road j

Glen Ellyn, 11 60137 l

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Subject:

Byron Station Units l'and 2 -!

Reply to Notice of Violation  !

Inspection Report Nos. 50-454/90023 ft 50-455/90023 EC_DotheL30s .50-45L.And_50-455 -  ;

Reference:

a) November ?9 1990 letter from H.D. Shafer to-to Cordell Reed transmitting the results of a Routine-Safety Inspection at Byron Station I

Dear Mr. Davis:

The referenced letter transmitted Inspection Report 50-454/90023; 50-455/90023 containing two level IV Notices of-Violation and indicated certain activities at Byron Station appeared to be in violation of HRC  ;

requirements. The Commonwealth Edison Company response to the Notices of.

Violation (50-455/90023-01) and (50-455/90023-02) are contained in Attachments ':

A and B, respectively.

If you have any questions;regarding this response, please direct them to this office.

Very truly yours,-

T,0. K ach Nuclear Licensing Manager 1 Attachments.

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kyC cc: ' NRC Resident Inspector-Byron NRC Document Control-Desk A. Hsia-NER W. Shafer-RIII- \

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ATTACHMENT A.

) ylola.tlon.(4HL92023_0B  !

I Technical Specification 3.7.1.2 requires at least two independent steam  ;

j generator auxiliary feedwater pumps and associated flow saths operable in ,

MMe 1. With one auxiliary feedwater pump inoperable, tie Technical

Specification action statement for 3.7.1.2 required the auxiliary feedwater '

i pump to be in an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Contrary to the above, the licensee failed to declare the 2A Auxiliary Feedwater (AF) pump inoperable and enter the appropriate Technical Specification action statement on August 8, 1990, when two pipe struts were removed on the essential service water suction piping for pre-outage modification work that rendered the 2A AFW pump inoperable. . At the time of

the pipe support removah, Unit 2 was in Mode 1 at 491, reactor power.

! CorrtttloILAttloAslarmLtad_theltsults Achleytd When the Station Control Room Engineer (SCRE) was notified that work had been

performed on the system, the pump was immediately declared inoperable pending-a review of the work completed to date. LCOAR 2B05 4.10-1a was entered at 4 1500 on 8-17-90. Two support systems.were modified in conjunction with modification H6-2-88-060 which changed the essential service water suction

piping to the 2A AfH pump. 2BVS 4.10-6.2, " Post Maintenance Visual

! Examination (VT3/4) of Safety Related Component Supports", was satisfactorily I completed for both supports and reviewed at 1837 on 8-17-90. Construction Quality Control verified installation =and performed dimensional inspections on l the supports after the work was completed. An Engineering Evaluation and '

10CFR50.59 Safety Evaluation determined that the current condition.of the revised sup) orts did not adversely affect operability since the changes increased tie load carrying capacity to the supports and did not impose any _ <

loads not previously considered. An On-Site Review concluded that )oth-the AF and essential service water (SX) systems were operable and the LCOAR was '

exitted at 2010 on 8-17-90.

Subsequently, Engineering and Construction (ENC) began calculations to determine operability of the system while these two supports were.being modified. To verify assumptions made in the calculations which were yleiding suspect results, a_walldown was performed of the equipment. On 8-23-90 at 1420, it was discovered that the scaffolding that had been erected on 8-6-90 was being supported from safety related line 2AF03AA-6"-- and two pipe supports t (2PSL-AF051-H89E-3 and 2PSL-AF051-H89E-4). The scaffolding was immediately

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removed. At 1500, the load bearing nuts on component support M-2AF06021R were ,

found loose. LC0AR 2005 4.10-1a was entered. The support was promptly- '

tightened and tested via NHR 878922. The LCOAR was exited at 2138 on 8-23-90.

On_ August 30, 1990, the results of the ENC evaluation of the support installation sequence and scaffolding loads were received. The results  ;

indicated that the normal operating loads were-within code. allowable values.

However, code allowable values were exceeded for design basis load combinations (seismic) during the time supports M-2AF03021R and M-2AF03019R were bdividually removed.

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i ATTACHMENT A (continued) l ,

3 1 In addition, immediately following the event a review was made of the B2R02 pre-outage planning efforts. The review revealed the following planning j activities were in place prior to the event:

1. Monthly Hodification Status Meetings discussing status, scope, work i

! groups involved, and possible installation obstacles.

2. Outage Planning Meetings specific to the Unit 2 second refuel outage (B2R02) discussing all aspects of the outage. starting 6 months prior to the planned outage start.

3. A meeting was held prior to the B2R02 Outage to specifically discuss possible pre-outage work. This meeting was attended by ENC, Operating, j 4

Maintenance Staff, and Hork Planning. Possible pre-outage work was reviewed, and dispositioned as: Pre-Outage, Partial Pre-Outage, or  !

Outage, based on possible impact to plant operations.

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4. Pre-Outage activities were discussed in the daily and weekly work planning meeting. The daily meeting discussed work covering the next 3 days, and the weekly meeting covered the next 2 weeks. An Operating Engineer, and the Operations Work Scheduler (SRO) attend both these

meetings to review upcoming planned work for possible plant impact.

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5. An informal Daily Construction Work Authorization sheet was in place, l which required job information to obtain Shift Authorization to Start Hork. ,l 6, Nuclear Operations Directive N00 TS.5 and ENC procedure QE-6 specify that j modification packages be categorized as-reautring an outage or not requiring an outage. This modification work was categorized as requiring an outage, however practice has been to review outage modifications for i individual tasks that may be performed pre-outage. The review is performed at the station.

. 1 Through a detailed review of the planning efforts already in place it was determined that some enhancements could be made. The ' 11owing actions were implemented directly following the event to implemen' Je enhancements.-

1. The informal Daily Construction Work Authorization Sheet was enhat,ted and formalized on August 20, 1990 for use with'the remaining Unit 2 Outage Modification Hork Packages. The sheet requires' completing a Daily Hork Description Sheet which provides more detailed information to the Shift i Engineer concerning the scope and-requiremenisinecessary to perform the j

, work. This sheet in conjunction with a copy of the Nuclear Work. Request j (NHR), and the Road Map _ Traveler', allows the Shift Engineer to address-operabillty concerns.

2. The Maintenance Modification Contractor's Scaffold Procedure (JVSCAF-1) was revised to include a requirement that Operati_ng perform a-' pre-job review and sign final authorization for scaffold use.

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- ** ATTACHMENT A (continuedt

3. The Station's " Post' Maintenance Checklist" procedure (BAP 1600-T11) was revised on SaptehQer 17, 1990 to require Maintenance:to indicate if:any scaffold was left 4n.the fleid after job completion, and-for Operating-to' ,

verify that1there are_no: scaffold related operability concerns prion:to declaring equipment ooerable.

4. An Awareness Day meeting was held on August 24.-1990, with representatives from Maintenance /Hodification Contractor (Pope), ENC, and the Station. Jtems discussed included the Station's approach:to maintenance' activities, operability concerns, safety, engineering-_ .

• Interaction, and the need for enhanced communications with the Operating Department. 3

5. Construction work activities in progress on August 25, 1990, were:

Inspected by a-team consisting of representatives.from ENC, Pope, and Station Management, _ The team had a current copy of the Pope Scaffolti Log Book and used it_to' identify,- inspect, and retag all existing Pope scaffold with the' revised scaffold inspection tags. ;A copy of the revised tag was given to the Shift Engineer. All unsatisfactory =ltems.

identified during the inspection were_ corrected.

These correct've actions were effective in preventing any additional operability concerns relating to pre-outage orfoutage maintenance or modifications activitier for the B2R02 refuel outage.

Corrective Action to A (dj Further Violations:

The following long term corrective actions are-being implemented-as a result.

of meetings held between 5tation daintenance, Operations, Work Planning, Tech Staff, and Engineering and Construction -' Nuclear Operation. iDuring this meeting it was determined-that although considerable " big-picture" pre-outage work planning was-done well in advance of the outage, the complexity of nodification work packages demands more_ detailed planning.-

To effectively develop modification work. packages-that take'into accountJall  ;

aspects of plant operation and reliability it was decided that more up-front inv'Ivement in the work package development by Operations andLTech Staff was needed.

Participants c.c the meeting also noted that construction work in the plant without proper Operations oversight is more likely_with outage related modifications, similar operational concernsEcould be present with-non-outage-Modifications as well. To avoid a further violationLin1this area during!both outage and non-outage periods the following long term-corrective actions will.

be taken:

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• - *' . ATTACHMENT A.

. (continued) '

1. For Modification work packages the work analyst, with input from both Tech Staff and a licensed Operations person and ENC person as appropriate, will write a summary of installation' steps,. including any operational steps i.e, take equipment Out-of-Service, enter LC0AR, etc.

-The level of. detail: contained in the summary will be sufficient to.

address operational and reliability concerns. This summary will-generally be done priorito detailed work package develo) ment. The trM 'lation summary will_be approved by an SRO and Tec1 Staff. A copy _

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of Mye Modification installation summary will be forwarded to' the_ Shif t Engineer for use when authorizing work. -BAP 1600-1 will be revised to include this reautrement.

2. Modification work packages which have Pre Out-of-Service /LCOAR work will-be broken dowr. IMn sub-packages clearly delineating Pre ,~

- 09t-of-Service /LCOM work from Post Out-of-Servic9/LC0AR work.=

BAP 1600-1 will be rev! sed to~ include this requirement. i

3. Modification wor packages will be: reviewed and approved by an Operating ,

Engineer or designee. The'0perating Emineer will assure that Pre .

-Out-of-Service /LC0AR work is clearly defined and segregated -'In addition the Operating Engineer may at his discretion insert " Operating Hold:

Points" or request that work steps be added to-the Road Hap Traveler-(step by step work instructions).to require operating concurrence prior to work proceeding. BAP 1600-1 will be revised ~to include this-requirement. .

4. Modificetlon Hork Packages which have Pre Out-of-Service /LC0AR work will be scheduled and-statused on the routine _or-outage work: schedule _(as applicable) on a sub-package level. BAP 1750-4 will.be revised to include this requirement.

Date Full Comollance wl_ll be Achieved:

! Compliance with Technical Specification LC0 3.7.1.2 was achieved on l August 17, 1990. Actions to prevent recurrence will-be completed by l June 1, 1991.

AIRS: 455-225-96 ' 9; 455-225-90-300; 455-225-90-301; 455-223-90-302.

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• - LATTACHMENT'B e hittiotLIA55L90023-02).

L 10CFR50,. Appendix B, Criterion V,=as implemented.by Commonwealth Edison's Quality Requiremont 5.0, states that activities affecting_ quality shall be prescribed by documented instructions and procedures of a type appropriate to the circumstances and shall be accomplished In accordance with these instructions and procedures.

BAP 370.3, Revision 6. "Adninistrat he Control During Refueling'.', step'C.44, states that prior to_ release of-a fuel assembly being seated in a_ spent fuel rack, a cognizant management individual shall independently; verify proper -f location as specified in the PHR Nuclear _ Component Transfer List.. -

BAP 2000-3, Revision 8, " Safeguard and Controlling-Movements of Nuclear' Fuel Within a Station", Step C.5, states the Feel Handling Foreman.shall verify corrett. fuel assembly location after Insertion of each fuel assembly _into the assigned storage rack by initialing each step of the PHR Nuclear-Component j Transfer List.

A. Contrary to procedures BAP_ 370-3 and BAP 2000-3, fuel assemblies were- --;

placed in the wrong spent fuel rack location on January 22; August 22, and September 25, 1990.

BMP 3118-5, Revision 5, " Reactor Vessel Upper Internal' Installation",

Steps f 2.d and f.2.e states.to slowly raise-the rig and upper internals off- .,

the storage rack until the rig and internals _have cleared the storage stand and guide _ studs. Hove the rig and upper internals over the_ reactor vessel.

B. Contrary to procedure BMP 31_18-5, the Unit 2 upper _ internals were not kept adequately clear.of the storage rack-while moving'the upper internals to the reactor vessel. As a result,.several guide pins were' bent that have to be straightened or cut'off prior to. Inserting the upper internals in the reactor vessel.

l A.1 Cotte c_tlle_Ac.t i on s T aketted_Re.s_p l t s Ac h i e v e d On January 22, 1990, the Fuel Handling and Technical Staff personnel were performing fuel shuffle operations in the spent fuel-pit. Activity was.

halted when a discrepancy was discovered with an assembly which was l previously positioned in the wrong rack location. A review of the_ Tag Board and the Nuclear Component Transfer List (NCTL) showed that location B-05 should have had an assembly-in it and did not. Observation with an underwater camera vertfled that the assembly in location C-05 was, in L -fact, the assembly-(D21E) which should-have been located in B-D5. A i

variation to the NCTL'was written to track the location.of this assembly. A second variation was written to place assembly 075F into location B-05. Execution of-the NCTL'was then-continued.

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ATTACHMENT B (continued)

. .o Corrective action for the January 25, 1990, event was to revise BAP 370-3, Administrative Control During Refueling, to include the following:

a. Independent vert 11 cation of fuel assembly location,

b. Orientation training for non-Fuel Handling pen unnel,

c. Required completion of a step on the NC1L prior to turnover,

d. Placards were placed on the spent-fuel storage racks as aids for identifying correct locations.

On September 25, 1990, fuel movements were suspended and an investigation was conducted when fuel handling and technical staff personnel noted a fuel assembly in the wrong location. The investigation revealed that assembly S61]

was placed in the wrong spent fuel storage rack location. A Nuclear Component Transfer List (NCTL) variation was initiated to place assembly S610 into the correct location (D-E03) and allow assembly T63K to be placed into location 0-003.

On October 9, 1990,-a third error concerning mispositioned fuel assemblies in the spent fuel pit was discovered following the completion.of core reload for Unit 2 Cycle 3 operation. The date of this occurrence was August 22, 1990.

The Spent fuel Pool verification showed that fuel assembly C45 was incorrectly located at spent fuel rack 0 location K14. The records and tag boards showed that fuel assembly C45 should have been at spent fuel rack Q location K04.

The fuel assembly had been moved from spent fuel rack G location J14 on August 22, 1990 during a series of movements to reorganize the fuel assemblies in the spent fuel pool prior to the refueling outage. A Nuclear Component Transfer List variation was initiated to place assembly C45 into the correct location.

The spent fuel racks consist of two different designs. Region 1 racks have greater boral neutron absorbing thickness and greater spacing than Region 2 racks. In each situation, although the fuel assemblies were loaded into the l wrong cell location within a rack, they were loaded into a similar regional l location. All three instances of misplaced fuel assemblies were reviewed.

l None of the three misplacements resulted in adverse safety consequences since they were bounded by existing FSAR analyses. All "as found". locations would have been acceptable if they had been specified in the NCTL. _

Short term corrective actions for the August 22nd and September 25th events l involved conducting 'Ta11 gate" sessions with Fuel Handling Department personnel to inform them of the seriousness of mis-positioned assembly incidents. These sessions included discussions on the. potential for reduction of shutdown margin in the pool if an assembly was mistakenly placed into a Region 2 rack wtthout having met the minimum burnup requirements as specified in the Station Technical Specifications. These tailgate sessions also included training on the proper communication techniques and the use of repeat backs when making assembly moves from one location to another.

Excessive use of overtime was determined to be a contributing cause for September 25, 1990 event. As a result, Technical Staff personnel have instituted administrative controls to ensure that the overtime worked by personnel is more rigorously controlled. The overtime worked by fuel handling personnel is presently addressed by an Administrative procedure.

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ATTACHMENT-B

. '(continued)

A2 C.orrective Actions to Avoid Further Violation The events identified in this. violation involved'the use of the new highi density storage racks which were installed in-1989 prior to the ByronL Unit One third refueling outage,- These high density racks have,a: smaller pitch and have varying dimensions, which results in a non-optimali Indexing system that tis relied upon for correct assembly placement.-

Corporate and Station management personnel-have. formed a. Task Force to review fuel hanoling equipment policies and procedures for the~ Byron, Braidwood and Zion Stations, This task force has been charged with the investigation of using a crane system or a-crane indexing system similar to the one used for core ~ loading,- In addition the Task Force:is-expected to recommend that an, independent human factors review be-conducted to determine if-the: layout and . indexing _of'the. spent fuel pool--

racks can be enhanced. 4 A.3 Dates When full Como11ance ; 11 be Achieved Fuel Assembly-locations were reconciled with the-the NCTLiby October 9, 1990.

The Task Force investigation is expected to be completed by-4-1-91.-

A. schedule will be developed shortly thereafter=to effect the task: force-recommendations AIR 455-225-90-30300.

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+ ' ATTACHMENT;B (continued)

B.1 Correctiv.e Actions TakeL and the Results' Achieved The fuel Handling-Foreman present October 8.-1990,1during the movement of_-

,the' upper internals noted an abnormality and immediately instructed the-  ;

personnel to place the upper internals on the support _.-stand.z An investigation was conducted using a video' camera-to inspect the guide-pins for. damage. The damage found requir_ed-the straightening of;9_-pins.

cutting off f> vins and gauging 16 pins surrounding-the' damaged area. The N repain wert - completed on November 3,1990.  !

B 31 5 s en ev  :

• Provide a caution on the adverse consequences of bending pins.

• provide-a caution to return internalscto the storage stand for 1 inspection if damage l1s suspected. - .

a provide detailed guidance on the path:to usc_between the reactor. -- 1 vessel and the stand (the guidance prevents-Intermediate lowering). . l

-*- provide quantitative' guidance with physical description to aid in-determining if the interr" s-are clear of_the stand by referencing cavity level and flange L.elght.

• ensure sound powered-communications are-made available in case conditions are such that hand signals and verbal communications are hindered.

b. Radiation Protection will evaluate the correct setting for the digital alarms while moving the upper internals-to and from.the reactor vessel,

c. ProcedureswillbereviewedandrevisedasnecessaryItoensure (1) ALARA briefings are more prescript)ve and-detailed when performing tasks which-involve-potential;high does concerns and (2), appropriate actions are taken in.the event:a high dose alarm-is

, received when moving the uppertinternals.

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d. Additional-training will-be provided on digital-dosimetry to ensure adequate understanding of the alarm setpoints and required response..

i e. Technical Staff will further investigate a modification and/or ._

l procedure changes to provide enhanced quantitative indicaticn for the- ,

height of the internals. <

f. -The.' corrective actions for the alignment pins have been analyzed for only-1 cycle of operation. Disposition of the-damaged alignment pins will be completed by 3-31-92.

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ATTACHMENT B (continued)

g. All maintenance personnel were briefed at department safety reetings and at a station stand-down on the' events leading to the-upper internals alignment pin damage and the importance of the alignment pins. They were also briefed on the event at Indian point and its similarity to our own.

B.3 Dates When Full Como11ance will be Achieved

a. BMP 3118-5 has been revised,

b. The evaluation for setting the digital alarms is expected to be completed 6-1-91. AIR 455-225-90-265

c. Procedures will be revised by 6-1-91. AIR 455-225-90-264

d. Training will be completed by 7-1-91. AIR 455-225-90-266

e. -The Technical Staff investigation to determine a quantitative indication for the internals height will be completed by 3-31-92.

AIR 455-225-90-263

f. The damaged pins will be dispositioned and/or replaced by 3-31-92.

AIRS 455-225-90-246, 455-225-90-261

g. The briefings are complete.

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