ML023020486

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Inservice Inspection Program - Owner'S Activity Reports
ML023020486
Person / Time
Site: Millstone  Dominion icon.png
Issue date: 10/18/2002
From: Price J
Dominion Nuclear Connecticut
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
B18790
Download: ML023020486 (27)
v • d • e


Text

Dominion Nuclear Connecticut, Inc.

Millstone Power Station 4Dominion Rope Ferry Road Waterford, CT 06385 OCT 18 2002 Docket Nos. 50-336 50-423 B18790 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit Nos. 2 and 3 Inservice Inspection'Proqram - Owner's Activity Reports Dominion Nuclear Connecticut, Inc. (DNC), he'reby submits the American Society of Mechanical Engineers (ASME), Sect i6n XI, Form OAR-i, Owner's Activity Reports for Millstone Power Station, Unit Nos. 2 rand 3. "The enclosures are in accordance with requirements of ASME Code Case N-532. The'authorization for use of this Code Case for Unit Nos. 2 and 3 is documented in references 1 ,and 2 below.

There are no commitments contained within this letter.

Should you have any questions regarding this submittal, please contact Mr. David W. Dodson at (860) 447-1791, extension 2346.

Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.

J. Aldibrice Site . e President - Millstone

Enclosures:

(2) cc: H. J. Miller, Region I Administrator R. B. Ennis, NRC Senior Project Manager, Millstone Unit No. 2 V. Nerses, NRC Senior Project Manager, Millstone Unit No. 3 Millstone Senior Resident Inspector

1. NRC letter to S E. Scace, "Millstone Nuclear Power Station, Unit No. 3 (Millstone Unit 3)

ASME Section Xl Inservice Inspection Relief Request Number IR-2-10 (TAC No. MA8276),"

dated August 24, 2000.

2. NRC letter to S. E. Scace, "Safety Evaluation of Relief Request RR-89-22 Associated with ASME Section XI, Code Case N-532, Millstone Nuclear Power Station, Unit No. 2 (TAC No.

MA6961)," dated March 24, 2000. No --

Docket Nos. 50-336 50-423 B18790 Enclosure 1 Millstone Power Station, Unit No. 2 Owner's Activity Reports for Refueling Outages 13 and 14

- I Page 1 of 13 MILLSTONE POWER STATION UNIT NO. 2 OWNER'S ACTIVITY REPORTS FOR REFUELING OUTAGES 13 AND 14 Revision 0 Contents:

OAR Report Number: MP-2, 2R13:

Table 1: Abstract of Examinations and Tests Table 2: Items With Flaws or Relevant Conditions That Required Evaluation for Continued Service Table 3: Abstract of Repairs, Replacements, or Corrective Measures Required for Continued Service OAR Report Number: MP-2, 2R14:

Table 1: Abstract of Examinations and Tests Table 2: Items With Flaws or Relevant Conditions That Required Evaluation for Continued Service Table 3: Abstract of Repairs, Replacements, or Corrective Measures Required for Continued Service PreparedBy: Date:

ISI Prograr6Coordinator Reviewed By: *, Date:

Au-th ed Nu srvc e In.pcoor'

)J FORM OAR-1 OWNER'S ACTIVITY REPORT Report Number MP-2, 2R13 Owner Northeast Nuclear Energy Company_ P 0 Box 345, Waterford , Connecticut 06385 (Name and Address of Owner)

Plant Millstone Nuclear Power Station, Rope Ferry Road, Waterford Connecticut 06385 (Name andAddress of Plant)

Unit Number 2 Commercial Service Date 12/2611975 Refueling Outage Number 13 (Ifappicable)

Current Inspection Interval 3rd (I st,2nd,3rd,.*4th, other)

Current Inspection Period (152d s rd (1st, 2nd, 3rd) 1989 Edition, No Addenda Edition and Addenda of Section Xl Applicable to the Inspection Plan UA-. f'-,..lld It Al Date and Revision of Inspection Plan 1f"24 It'sO1 r%:VIII ug to96 Rev s on  : 1. I W) fk 0.kf. n4 Edition and Addenda of Section XI Applicable to Repairs and replacements, if Different N/A CERTIFICATE OF CONFORMANCE I certify that the statements made inthis Owner's Activity Report are correct, and that the examinations, tests, repairs, replacements, evaluations and corrective measures represented by this report conrform to the requirements of Section XI.

Certificate of Authorization No ,-YNIA Expiration Date N/A Signed 3 r1era$( Z-2 *°t'1 ate- /

Owner or 0/her's Designee, TfleV CERTIFICATE OF INSERVICE INSPECTION 1,theJndersdIned, hld ng a valid commis of Boiler aji Pressure V ssel In pectors and the State or Province

,have , &-.., inspected of1,, and employed by;,,

the items described in this Owners Activity Repoft, during the eriod ,and state to the best of my knowledge and belief, the Owner has performed all activities represerld by this report inaccorerance with the requirements of Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations, tests, repairs, replacements, evaluations and corrective measures described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any p s injury or property damage or a loss of any kind arising from or connected with this inspection N ~~-,. Commissions A Y 9 Naheai BoardStat/rovince andEndorsements S504tho; iInspetors

Table 1 j ABSTRACT OF EXAMINATIONS AND TESTS Total Examinations Total Total Total Examinations Credited (%) To Date Examinations Examinations Credited (%) for The Interval Examination Required for Credited for For The Period Category The Interval This Period Remarks B-A 28 0 0% 0%

B-B 9 3 33% 33%

B-D 34 6 17% 17%

B-E N/A 0 0% 0% Tracked under B-P pressure tests per Granted Relief Request RR-89-16.

B-F 28 4 14% 14%

B-G-1 219 0 0% 0%

B-G-2 90 22 26% 26%

B-H 1 0 0% 0% Not required for Third Interval, MP-2 has elected to perform examination of one support skirt due to SG replacement in 1992. This is an augmented examination.

B-J 131 18 14% 14%

B-L-1 2 0 0% 0%

B-L-2 2 0 0% 0%

B-M-2 5 0 0% 0%

B-N-1 14 4 29% 29%

B-N-2 15 0 0% 0%

B.N-3 33 0 0% 0%

B-0 15 0 0% 0% Deferred to Third Period -1 0 B-P 12 3 33% 33% Based upon 6 refueling outages C-A 7 0 0% 0% C o

C-B 6 0 0% 0%

C-C 36 8 22% 22%

C-F-1 77 25 32% 32%

C-F-2 19 4 21% 21%

F-A 270 80 29% 29%

IWE 3 1 33% 33% 1ý'period Inspection completed IWL 3 1 33% 33% 1= period Inspection completed

  • P 0

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Table 2 "I ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section XI Examination or Test (Yes or No)

B-D B3.1 10 Weld Pressurizer Safety Valve Nozzle To Top Head This flaw was evaluated as a subsurface PR-NTH-3 Weld @ 180 Degrees planar flaw. The measured wall thickness is Yes 4.05" excluding cladding. The a/I - .225, a/t =

.0333 or 3.33% and Y =1.0 The maximum allowable flaw size for an a/i of.225 is 3.55%

in accordance with Table IWB-3512-1 of ASME Section XI, 1989 Edition with no addenda, thus the indication was evaluated as I acceptable for continued service.

B-D B3.1 10 Weld Relief Valve Nozzle To Top Head Weld This indication was evaluated as a spot Yes PR-NTH- I indication with no depth or length and evaluated as acceptable for continued service.

F-A Fl.30G Component Support Restraint Gap between base plate and wall. Evaluated Yes 413125 as acceptable for continued service F-A Fl.30G Component Support Restraint Incomplete thread engagement of two bolts: Yes 413153 Bolt I: I Nut Thread Visible 1800 and Bolt 2:

I Nut Thread Visible 2700. Evaluated as acceptable for continued service.

Yes o0 F-A F1.30B Component Support Spring Can Spring can was indicated outside of hot/cold 527070 range. Evaluated as acceptable for continued service as the setting was with 10% of the design criteria.

F-A FI.30B Component Support Pedestal Support Incomplete thread engagement of one nut, I Yes 329048 full thread visible. Evaluated as acceptable for continued service. 440 F-A FI.30G Component Support Restraint Incomplete thread engagement of one nut, I Yes 0~

427080 full thread visible. Evaluated as acceptable for continued service. *P 05 tQ

Table 2 "I ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section XI Examination or Test (Yes or No)

Yes F-A Fl.30B Component Support Spring Can Spring can was indicated outside of hot/cold 527069 range. Evaluated as acceptable for continued service as the setting was with 10% of the design criteria.

F-A F1.30C Component Support Spring Can Spring can was indicated outside of hot/cold Yes 527068 range. Evaluated as acceptable for continued service as the setting was with 10% of the design criteria. I F-A FI.30C Component Support Spring Hanger Spring can was indicated outside of hot/cold Yes 427098 range. Evaluated as acceptable for continued service as the setting was with 10% of the design criteria.

F-A F1.30A Component Support Sway Strut Support does not match the drawing. The two Yes 401020 snubbers do not exist. Different Configuration. UIR written. UIR Disposition:

No Discrepancy exists. Support was modified via MMOD M2-97510 as documented on DCN DM2-00-1367-96. Evaluated as 0 0,.

acceptable for continued service.

F-A FI.30G Component Support Restraint-Support Improper thread engagement: Anchor bolt is Yes 304029 recessed into nut 1/2/2 thread. Connection is tight but not fully engaged. Evaluated as acceptable as is for continued service.

F-A FI.30G Component Support Restraint with 1" Ring Welded To Pipe Improper Thread Engagement, approximately Yes 502019 2 threads not engaged on nut. (upper right),

Evaluated as acceptable as is for continued service.

F-A F1.30A Component Support Sway Strut No spacer to prevent over tightening of Yes 403078 clamp. Evaluated as acceptable as is (no I _spacer required) for continued service. 0 el 0.

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Table 2 ,-I ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section XI Examination or Test (Yes or No)

F-A Fl.30A Component Support Sway Strut One stud on bottom base plate is I full thread Yes 403072 short of full engagement. Evaluated as acceptable as is for continued service.

F-A FI.30C Component Support Spring Hanger One loose nut was identified on the main Yes 412012 clevis, and spring settings were out of tolerance. Evaluation was to retighten the nut and reset the spring cans to their design configuration. Support was evaluated as being able to perform its' intended function in the as found condition.

F-A Fl.30B Component Support Support Bolt I Bent, nut edge is 1/8" out from plate. Yes 491408A Bolt 2 Bent, nut edge is 1/8" out from plate F-A F1.30H Component Support Hydraulic Snubber Improper Thread Engagement One Thread Yes 410015 Visible for 180

  • within nut. Evaluated as acceptable as is for continued service.

F-A FI.30C Component Support Spring Hanger Exact setting could not be discerned using Yes 402052 optical aids from floor level. Sketch provided a' -- 0 for spring position. UIR M2-00-019 Evaluated as acceptable as is for continued service.

F-A F 1.30M Component Support Dual Mechanical Snubbers Improper Thread Engagement, 11/22 threads Yes 507002 visible within nut. Evaluated as acceptable as 0 o is for continued service.

LA C F-A FI.30A Component Support Sway Strut Two nuts on the pipe clamp were identified as Yes 405599 loose and the southernmost sway strut was 00 identified as impinging upon another support.

Impingement was evaluated as acceptable and TR 18M2154348 was initiated to tighten loose nuts.

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Table 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT '0 REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section XI Examination or Test (Yes or No)

Yes F-A Fl.30C Component Support Spring Hanger Spring can was indicated outside of hot/cold 380324 range. This support was walked down in the field and evaluated as acceptable for continued service as the setting was with 10%

of the design criteria.

F-A FI.30C Component Support Spring Hanger Spring can was indicated outside of hot/cold Yes 405651 range. This support was walked down in the field and evaluated as acceptable for continued service as the setting was with 10%

of the design criteria.

F-A FI.30E Component Support Hanger During VT-3 exam of hanger 491389, a Yes 491389 rusted "C" clamp was observed, possibly interfering with the intended operation of the hanger. The "C" clamp is fastened to item #5 on the Mech Dwg, restricting the movement of item #6, also on the Mech Dwg. Evaluated as acceptable as is for continued service. C.4 F-A FI.30G Component Support Restraint VT-3 examination identified 3 Hilti bolts as Yes 402077 not having full thread engagement, 4 Hilti bolts as being bent and 1 nut not fully

-'. 0 engaged. The lack of thread engagement was evaluated as acceptable based upon NUSCO Catc. 79-02-1066GP, the 4 bent Hilti bolts evaluated acceptable as is as they were o -.

"installedat an angle and the I nut not fully engaged evaluated acceptable as the nut has sufficient bearing against the baseplate. 00

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Table 2 "Vy ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Item Component Item Description Flaw Characterization Flaw or relevant Category Number ID (IWA-3300) Condition Found During Scheduled Section XI Examination or Test (Yes or No)

F-A F1.30C Component Support Spring Hanger The plate affixed to spring can is illegible. Yes 410045 Could not tell where spring can is set.

Setting is approximately mid-scale. Evaluated as acceptable as is for continued service.

F-A FI.30A Component Support Sway Strut Nut has improper thread engagement. 1/8" Yes 310019 below flush. Evaluated as acceptable as is for continued service.

F-A FI.30H Component Support Shock And Sway Suppressor 2 nuts were identified as being loose. Yes 412016 Evaluated as acceptable as is for continued service. TR 18M2145257 was initiated to tighten the nuts.

F-A FI.30C Component Support Spring Hanger Both spring cans are below the required Yes 312012 setting per drawing 312012, One can is at 9513# + 10% = 10,580# One can is at 9372#

+ 10% - 10,430# Evaluated as acceptable as J_ _is for continued service.

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Table 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Flaw or relevant Condition Found Repair, During Scheduled Replacement, Description Section XI Repair/

Code or Corrective Item of Work Examination or Date Replacement Class Measure Descriptiori Test (Yes/No) Completed Plan Number No Repairs, Replacements or Corrective Measures were required for continued service during 2R1 3 CO o O o 0 Cp

I. -

FORM OAR-1 OWNER'S ACTIVITY REPORT Report Number MP-2, 2R14 Owner Dominion Nuclear Connecticut P.O. Box 345., Waterford . Connecticut 06385 tName and Address 0I Uwner)

Plant Dominion Nuclear Connecticut, Rope Ferry Road Waterford Connecticut 06385 iName aa Aodess OIn-lan7 Unit Number (it 2 appscaoiej Commercial Service Date 12/26/1975 Refueling Outage Number 14

  • rrl Current Inspection Inter I qrd (1st. 2nd. 3rd., 4th, other)

Current Inspection P*,4,.v4 A-A 1 et (Ist, 2nd, 3rd)

Edition and Addenda of Section Xl Applicable to the Inspection Plan 1989 Edition, No Addenda Date and Revision of Inspection Plan I'..,., i Cin 4 Inc E3 i ;

I*l UIF--

I II*:IU nuVlblU!  !

f1t,ICII I'*11* U r113UI U*.1 ni Edition and Addenda of Section Xl Applicable to Repairs and replacements, if Different N/A CERTIFICATE OF CONFORMANCE I certify that the statements made inthis Owners Activity Report are correct, and that the examinations, tests, repairs, replacements, evaluations and corrective measures represented by this report conform to the requirements of Section XI.

Certificate of AuthoNdzatio/Lo NIA Expiration Date N/A Signed 1 aA =5 I~wner or Owner's Designee, Title applicable) Date CERTIFICATE OF INSERVICE INSPECTION I, the u gdersigned, holding a valid commission Issued b.h tional Board f oiler and Pressjr~e VesdIn~pectors and the State or Province of and employed by =e,. *= of _/______J" _________ have inspected the items described in this Owner's Activity Report, during the period *to 2*2 , and state to the best of my knowledge and belief, the Owner has performed all activities representedby thisrreport in ac6ordarce with the requirements of Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations, tests, repairs, replacements, evaluations and corrective measures descnbed in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected with this inspection.

D. Commissions '7/Z7 £ National Board, State, Province and Endorsements Da te .

I/ /

(I Table 1 ABSTRACT OF EXAMINATIONS AND TESTS Total Examinations Total Total Total Examinations Credited (%) To Date Examinations Examinations Credited (%) for The Interval Examination Required for Credited for For The Period Category The Interval This Period Remarks B-A 28 3 100% 11%

B-B 9 3 100% 33%

B-D 34 12 100% 33%

B-E N/A 0 0% 0% Tracked under B-P pressure tests per Granted Relief Request RR-89-16.

B-F 28 11 100% 33%

B-G-1 219 56 100% 26%

B-G-2 90 29 100% 32%

B-H 1 0 0% 0% Not required for Third Interval, MP-2 has elected to perform examination of one support skirt due to SG replacement in 1992. This is an augmented examination. I B-J 131 38 100% 29%

B-L-1 2 0 100% 0%

B-L-2 2 1 100% 50%

B-M-2 5 0 0% 0%

B-N-1 14 4 100% 29%

B-N-2 15 0 0% 0%

B-N-3 33 0 0% 0%

B-0 15 0 0% 0% Deferred to Third Period B-P 12 4 100% 33% Based upon 6 refueling outages C-A 7 2 100% 29% me C-B 6 2 100% 33%

C-C 36 10 100% 28%

C-F-1 77 22 100% 29%

C-F-2 19 5 100% 26% - l C-H 104 26 100% 25%

D-A 4 2 100% 50%

D-B 14 7 100% 33%

D-C 16 4 100% 25% t'j F-A 270 87 100% 33%

IWE 3 1 100% 33% 1"' period inspection completed IWL 3 1 100% 33% 14'penod inspection completed t*b

Table 2 ('I ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section XI Examination or Test (Yes or No)

F-A FI.30G Component Support Restraint Incomplete thread engagement of one nut, 2 327125 threads visible. The lack of thread engagement Yes was evaluated as acceptable based upon NUSCO Calb. 79-02-1066GP. Evaluated as acceptable for continued service.

F-A FI.30G Component Support Restraint Corner of the base plate was not flush by 505138 approximately '/" and one nut was not fully Yes engaged. The base plate was evaluated in I accordance with M2-EV-98-0181 and the nut was evaluated in accordance with M2-EV-98 0180 as acceptable for continued service with no corrective actions required.

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  1. 1 Table 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Flaw or relevant Condition Found Repair, During Scheduled Replacement, or Description Section XI Repair/

Code Corrective Item Description of Work Examination or Date Replacement Class Measure Test (Yes/No) Completed Plan Number No, these flaws Reactor Vessel Half nozzle replacements at were found during Head control rod penetration locations 21, 34 & 50 an inspection to Repair drive mechanism using the temper bead welding comply with NRC nozzles 21, 34, 50 technique and in accordance with Information Notice 3/19/2002 M2-01-14140 approved relief request RR-89-34 2001-05 and NRC Bulletin 2001-01 No, these flaws were found as part Machining of the heater nozzle of an augmented 3/16/2002 M2-02-02270 Corrective Pressurizer heater penetrations and drilling and inspection based Measure penetration tapping of the pressurizer bottom upon industry nozzles Al and C4 head at nozzles Al and C4 were experience with 0 >

required to install MNSA clamps pressurizer heater penetration L_cracking.

0o 00

Docket Nos. 50-336 50-423 B18790 Enclosure 2 Millstone Power Station, Unit No. 3 Owner's Activity Report for Refueling Outage 7

Page 1 of 7 MILLSTONE POWER STATION UNIT NO. 3 OWNER'S ACTIVITY REPORT FOR REFUELING OUTAGE 7 Revision 0 Contents:

OAR Report Number: MP-3, 3R07:

Table 1: Abstract of Examinations and Tests Table 2: Items With Flaws or Relevant Conditions That Required Evaluation for Continued Service Table 3: Abstract of Repairs, Replacements, or Corrective Measures Required for Continued Service PreparedBy: Date: ____-_____

ISI ProgramCoordinator Reviewed By: Date:

Authed Nuclear service lnspeo r

FORM OAR-1 OWNER'S ACTIVITY REPORT Report Number MP-3, 3RO7 Owner Dominion Nuclear Connecticut, Rope Ferry Road, Waterford Connecticut 06385 (Name and Address of Owner)

Plant Millstone Nuclear Power Station, Rope Ferry Road, Waterford Connecticut (Name and Address of Plant)

Unit Number 3 Commercial Service Date April 23, 1986 Refueling Outage Number 07 (if applcable)

Current Inspection Interval 2nd (Ist, 2nd, 3rd.)

Current Inspection Period 1st (Ist. 2nd. 3rd, 4th, other)

Edition and Addenda of Section XI Applicable to the Inspection Plan 1989 Edition, No Addenda and 1998 Edition, No Addenda for Subsection IWE/IWL Date and Revision of Inspection Plan May 15, 2001 , Revision 01, Change 4 and January 30, 2001, Revision 1 for Subsection IWE/IWL Edition and Addenda of Section XI Applicable to Repairs and replacements, if Different NIA CERTIFICATE OF CONFORMANCE I certify that the statements made in this Owner's Activity Report are correct, and that the examinations, tests, repairs, replacements, evaluations and corrective measures represented by this report conform to the requirements of Section XI Certificate of Authori ation No N/A Expiration Date NIA (ifBPPWcabe)

Signed ISI Program Coordinator Date 10108/2002 orOwners Dennee. Title ri5"ner CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State or Province of Connecticut and employed by Hartford Steam Boiler CT of Hartford Connecticut have Inspected the items described In this Owner's Activity Report, during the period April 23, 1999 to July 22. 2002

  • and state to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance with the requirements of Section XI By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations, tests, repairs, replacements, evaluations and corrective measures described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property damage or a loss of any kind arising from or connected with this inspection.

am4 a. 4--- 'F Commissions CT 1137 "spedo. Co misin National Board State, Pronce endEndorsements Date 10111/2002

I, Table 1 ABSTRACT OF EXAMINATIONS AND TESTS (See Note 1)

Total Total Examinations Examinations Total Examinations Total Examinations Examination Required for Credited for Credited (%) Credited (%) To Date Category The Interval This Period For The Period for The Interval Remarks B-A 24 5 21 21 B-B 3 1 33 33 B-D 44 8 18 18 Item numbers 83.90 and B3.100 utilize Code Case N.521 which allows deferral to the end of the Interval.

B-E 2 0 0 0 See Note 2 B-F 22 4 18 18 B-G-1 221 72 33 33 B-G-2 62 21 34 34 B-H 9 0 0 0 Relief Request IR-2-4 granted to perform all (9) examinations In one period.

B-K-1 9 3 33 33 B-J 306 60 20 20 B-L-1 N/A N/A N/A N/A There are no components under this Category at Millstone Unit 3.

B-L-2 0 0 0 0 Required only if disassembled during maintenance activity.

B-M-1 N/A N/A N/A N/A There are no components under this

__Category at Millstone Unit 3.

B-M-2 0 0 0 0 Required only if disassembled during maintenance activity.

B-N-I 3 1 33 33 100% are examined each Inspection (i/Period) period.

B-N-2 7 0 0 0 See Note 2 B-N-3 2 0 0 0 See Note 2 B-0 4 0 0 0 See Note 2 B-P 30 12 40 40 Required for examination each refueling outage. Number and percentage are based on the number of refueling 0

  • outages anticipated during the Interval. 00 C-A 7 2 29 29 >.

C-B 13 4 30 30 C-C 39 8 21 21 C° go

Table 1 ABSTRACT OF EXAMINATIONS AND TESTS (See Note 1)

Total Total Examinations Examinations Total Examinations Total Examinations Examination Required for Credited for Credited (%) Credited (%) To Date Category The Interval This Period For The Period for The Interval Remarks C-D 2 0 0 0 See Note 3 C-F-1 162 41 25 25 C-F-2 38 11 29 29 C-G 11 3 27 27 C-H 86 27 31 31 D-A 30 9 30 30 D-B 128 37 28 28 D-C 4 1 25 25 E-A 105 35 33 33 100% of the items are examined each (35/Period) Inspection Period.

E-C N/A N/A N/A N/A There are no components listed under this category during this period L-A N/A 34 100 N/A Examination frequency for this Category Is once every five years. Inspection Interval is not applicable for this Category.

L-B N/A N/A N/A N/A There are no components under this

_ _ _ 1_ 262_1_81_30_30 Category at Millstone Unit 3.

F-A 262 81 30 30 ________________

R-A 83 29 34 34 NOTES:

1. This report represents a summary of the Inservice Inspection activities performed at Dominion Nuclear Connecticut's Millstone Unit 3 power station C

during the 2001 refueling outage (3RO7) through the end of the Inspection period which ended 7/22/2002. This represented the last outage of the first >

inspection period of the second inspection interval. Reporting of the first refueling outage of this period (3RF06) in 1999 was previously submitted In accordance with the requirements IWA-4000 and IWA-6000, (Reference Unit 3 letter, dated September 27, 1999, Docket No. 50-423). Table 1 represents credited examinations from both refueling outages RFO6 and RFO7.

2. For this Examination Category, deferral of examinations to later in the interval is permissible, In accordance with ASME Section XI, IWB-2500-1 0 0 requirements.
3. Percentage requirements in this category can not be met for each period due to only 2 items in the Examination Category.

0

-J 0

Table 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Characterization Flaw or relevant Examination Item Component Item Description (IWA-3300) Condition Found Category Number ID During Scheduled Section X1 Examination or Test (Yes or No)

B-G-! B6.50 179-103-34 RPV Closure Head Washer. 0 875" hnear indication found during visual examination of RPV Closure Head Washer. YES Indication is within the limits of IWB-3515. I(b) and evaluated as acceptable.

C-H C7.50 3QSS*PIA Quench Spray System Pump Leakage at bolted connection. Evaluated using YES the criteria of Code Case N-566-1 approved for use under Relief Request IR-2-6.

D-C D3.10 Line 3-SFC-012-34-3 Piping flanged connection Leakage at bolted connection. Evaluated using YES the criteria of Code Case N-566-1 approved for use under Relief Request IR-2-6.

0 0 0

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Table 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Flaw or relevant Condition Found Repair, During Scheduled Replacement, Description Section XI Repair/

Code or Corrective Item of Work Examination or Test Date Replacement Class Measure Description (Yes/No) Completed Plan Number No Repairs, Replacements or Corrective Measures were required for continued service during 3RO7

__ _ _ __I_ I _ _

+ 4 1 4

.4. L 1 & .4. 4

.1. 4 1- 4 + 4

.4. 4 4. 4 + 4

+ 4 .4- 4 4. 4

4. 4 4

.4. & 1. 4 .4. 4

.4. 4 .4. 4 4. 4

4. t 4. 4 4. t tv00

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4 4 + 4 4 4 100 0 4 4 + 4 4 4 S.-.4 4 4 4. 4 4 4 0, a'

Telephone (508) 721-7736 YANKEE A TOMIC ELECTRIC COMPANY Facsimile (508) 721-7743 Suite 200, 19 Midstate Drive, Auburn, Massachusetts01501 CeANVKIEiE October 21, 2002 BYR 2002-056 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

References:

(a) License No. DPR-3 (Docket No. 50-29)

Subject:

Lic.ensee Event Report (LER) 2002-001-00 Pursuant to 10CFR50.73(a)(2)(i)(B) of the Commission's Rules and Regulations, Yankee Atomic Electric Company (YAEC) is providing Licensee Event Report 2002-001-00, titled "Spent Fuel Pit Area Radiation Monitor Alarm Setpoint Set Above Limit Allowed by Technical Specification 3.3."

Should you have any questions, please contact Mr. Greg Babineau, Safety Oversight Manager, at (413) 424-2202.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY James A. Kay Manager of Regulatory Affairs c: J.B. Hickman, USNRC, Project Manager R.R. Bellamy, USNRC, Region I

U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 7-31-2004 NRC FORM 366 (7-2001) COMMISSION Estimated burden per response to comply with this mandatory Information collection request 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> Reported lessons learned are Incorporated Into the licensing process and ted back to Industry Send comments regarding burden estimate to the Records Management Branch (T-6 E6),U S. Nuclear Regulatory Commission. Washington DC 20555-0001, or by internet LICENSEE EVENT REPORT (LER) e-maillto bsl@nrc gov, andOfficeto the Desk Officer, Office of Information and Regulatory Aftfairs, NEOB-10202 (3150-0104) of Management and Budget, Washington, DC 20503 IIa (See reverse for required number of means used to Impose Information collection does not display a currently valid OMB control digits/characters for each block) number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the Information collection.

1. FACILITY NAME 2. DOCKET NUMBER 3. PAGE Yankee Nuclear Power Station 05000 029 1 OF 4
4. TITLE Spent Fuel Pit Area Radiation Monitor Setpoint Set Above Limit Allowed by Technical Specification 3.3 MO
5. EVENT DATE DAY YEAR YEAR
6. LER NUMBER SISE OUEN TIAL NUMBER R EV NO MO DAY I
7. REPORT DATE Y A YEAR A I I Y NAME FACILITY N M 8 OTHER FACILITIES INVOLVED O K T NUMBER DOCKET 05000 N M E FACILITY NAME DOCKET NUMBER 08 22 2002 2002 - 001 00 10 21 2002 05000
9. OPERATING 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REOUIREMENTS OF 10 CFR *" (Check all that apply)

MODE N/A 20 2201 (b) 20 2203(a)(3)(ii) 50.73(a)(2)(ii)(B) 50 73(a)(2)(ix)(A)

10. POWER 20 2201 (d) 20 2203(a)(4) 50 73(a)(2)(iii) 50 73(a)(2)(x)

LEVEL 20 2203(a)(1) 50 36(c)(1)(i)(A) 50 73(a)(2)(iv)(A) 73 71 (a)(4) 20 2203(a)(2)(i) 50 36(c)(1)(ii)(A) __50 73(a)(2)(v)(A) 73 71 (a) (5) 20 2203(a)(2)(ii) 50 36(c)(2) 50 73(a)(2)(v)(B) OTHER SSpecify in Abstract below or in 20 2203(a)(2)(0ii) 50 46(a)(3)(ii) 50 73(a)(2)(v)(C) NRC Form 366A 20 2203(a)(2)(mv) 50.73(a)(2)(i)(A) 50 73(a)(2)(v)(D) 20 2203(a..(2)(v) X 50 73(a)(2)(i)(B) _.50 73.a)M: .vii) 20 2203(a2(21) 50 73(a)(2)(i)(C)

(C _ 50 73(a)(2)(viii)(A)

........... 20 2203(a)(3)(i) 5 7 a i A5 50 73(a)(2)i)A 077 I}2)(viII )

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12. LICENSEE CONTACT FOR THIS LER NAME TELEPHONE NUMBER (Include Area Code)

Greg Babineau, Safety Oversight Manager (413) 424-2202

13. COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT CAUSE SYSTEM COMPONENT MANU-FACTURER REPORTABLE TO EPIX CAUSE SYSTEM I COMPONENT MANU-FA CTURER REPORTABLE TO EPIX D IL RA__ _ _ _ _ _ _ _ _ _
14. SUPPLEMENTAL REPORT EXPECTED 15. EXPECTED MONTH DAY YEAR SUBMISSIOND M

_--YES If es, cornplete EXPECTED SUBMISSION DATE). X NODATE_____

16. ABSTRACT (Umit to 1400 spaces, I e, approimately 15 single-spaced typewritten lines)

Yankee Nuclear Power Station ceased power operation in February 1992 and is being decommissioned. To facilitate completion of decommissioning, a campaign is underway to transfer spent nuclear fuel located in the Spent Fuel Pit (SFP) to dry cask storage. On 08/22/02, during fuel handling operations, the Plant Shift Supervisor identified that the alarm setpoints for the SFP Area Radiation Monitor (ARM) were set at a value greater than Technical Specification requirements. The SFP ARM is an instrument required by Technical Specification 3.3 to ensure early detection of inadvertent criticality during fuel handling activities. The Technical Specification requires the alarm setpoints for the SFP ARM be set at less than 5 mr/hr or two times the background radiation level, whichever is greater, while moving irradiated fuel, control rods or sources.

Fuel handling operations were immediately suspended, Condition Report No.02-579 was initiated, and procedure OP-4816, "Functional Test and Alarm Setting of the Area Radiation Monitoring System" was performed. The SFP ARM alarm setpoints were verified to be 12 mr/hr, while two times the observed background radiation level (3 mr/hr) at that point in time would have resulted in a required alarm setpoint of 6 mr/hr. As such, this LER is submitted in accordance with 1 OCFR50.73(a)(2)(i)(B) as a condition of non compliance with a Technical Specification. The cause has been attributed to procedural inadequacies that did not properly anticipate or account for potential changes to SFP background radiation levels (and SFP ARM alarm setpoint requirements) due to dry cask storage activities adjacent to the SFP during fuel handling operations.

NRC FORM 366 (7-2001)

NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION 41-2001)

LICENSEE EVENT REPORT (LER)

1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER Yankee Nuclear Power Station 05000029 2002 -- 001 -- 00 2 OF
17. NARRATIVE (If more space Is required,use addibonalcopies of NRC Form 366A)

EVENT DESCRIPTION Yankee Nuclear Power Station ceased power operation in February 1992 and is being decommissioned. To facilitate completion of decommissioning, a campaign is underway to transfer spent nuclear fuel located in the Spent Fuel Pit (SFP) to dry cask storage. While conducting fuel handling operations associated with the loading of Transportable Storage Container (TSC) No. 5 in accordance with OP-7107, "Spent Fuel Pit Component Movement Program" on 08/22/02 at approximately 1830, the Plant Shift Supervisor identified that the Spent Fuel Pit Area Radiation Monitor alarm setpoints exceeded Technical Specification 3.3 requirements.

Technical Specification 3.3 requires the alarm setpoints for the SFP ARM be set at less than 5 mr/hr or two times the background radiation level, whichever is greater, while moving irradiated fuel, control rods, or sources. Fuel handling operations were immediately suspended, Condition Report No.02-579 was initiated, and procedure OP-4816 "Functional Test and Alarm Setting of the Area Radiation Monitoring System" was performed. The SFP ARM alarm setpoints were verified to be 12 mr/hr, while two times the observed background radiation level (3 mr/hr) at that point in time would have resulted in a required alarm setpoint of 6 mr/hr. The SFP ARM Alert and High alarms were reset in accordance with OP-4816, "Functional Test and Alarm Setting of the Area Radiation Monitoring System" to be consistent with the Technical Specification 3.3 LCO prior to the resumption of fuel handling operations. Based on review of records, it has been determined that fuel handling operations were conducted with the SFP ARM in operation per OP-4822 "SFP Manipulator Crane Area Radiation Monitor Channel Check", but with the alarm setpoints in non-compliance with Technical Specification 3.3 from 08/19/02 at approximately 1835 until the point of discovery on 08/22/02 at approximately 1830.

CAUSE OF EVENT The corrective actions associated with Condition Report No.02-579 identified procedural inadequacies as the cause of event. The procedures in place at the time of the occurrence did not properly anticipate or account for potential changes to SFP background radiation levels (and SFP ARM alarm setpoint requirements) during fuel handling operations due to dry cask storage activities adjacent to the SFP. Examples of such activities include movement of the loaded Transfer Cask (TFR) and/or loaded Vertical Concrete Cask (VCC), or changes to the shielding configuration of the loaded TFR while it was in the Fuel Transfer Enclosure (FTE) or alleyway between the FTE and SFP. The SFP ARM alarm setpoints were properly set on 08/15/02 by OP-4816 "Functional Test and Alarm Setting of the Area Radiation Monitoring System" to reflect the ambient SFP background radiation level at that time. Subsequent changes in SFP background radiation levels caused by movement of loaded VCC No. 4 to the ISFSI pad on 8/18/02 at approximately 2030 were not recognized.

Although the SFP ARM was determined to be in operation per OP-4822 "SFP Manipulator Crane Area Radiation Monitor Channel Check", the SFP ARM alarm setpoints were not evaluated prior to initiation of fuel handling operations on 8/19/02 at approximately 1835. The following procedures have been identified as containing inadequacies which contributed to this occurrence:

"* OP-4822 "SFP Manipulator Crane Area Radiation Monitor Channel Check".

"* OP-7107 "Spent Fuel Pit Component Movement Program".

"* 13200-OP-2961 "Fuel Load into TSC/Transfer Cask".

NRC FORM 366A (1-2001)

NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER Yankee Nuclear Power Station 05000029 2002 -- 001 -- 00 3 OF 4
17. NARRATIVE (If more space is required,use additionalcopies of NRC Form 366A)

SAFETY ASSESSMENT The safety significance of this incident is low. The SFP ARM is an instrument required by Technical Specification 3.3 to ensure early detection of inadvertent criticality during fuel handling activities. Although the alarm setpoints for the SFP ARM were out of tolerance with Technical Specification 3.3, the SFP ARM was in operation with the alarm setpoints based on the background radiation levels with loaded VCC No. 4 in the alleway between the FTE and SFP. Subsequent movement of loaded VCC No. 4 out of the alleyway reduced the SFP background radiation levels from 6 mr/hr to 3 mr/hr. The actual time interval to reach an alarm condition (12 mr/hr), as compared to the time required to reach the Technical Specification required alarm setpoint (6 mr/hr) would have been negligible. An increase in radiation levels resulting from an inadvertent criticality would have been detected early enough to preclude any additional impact to the workers or public safety.

CORRECTIVE ACTIONS The following procedural changes were made to preclude recurrence:

"* OP-4822 "SFP Manipulator Crane Area Radiation Monitor Channel Check" is the procedure that is used to satisfy the Technical Specification 4.3.1 surveillance requirement to perform a SFP ARM Channel Check prior to and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during fuel movement. OP-4822 was revised to include a verification that the SFP ARM alarm setpoints are proper at each surveillance interval and to provide actions to suspend fuel handling operations and initiate OP-4816 "Functional Test and Alarm Setting of the Area Radiation Monitoring System" to reset the SFP ARM alarm setpoints if they are not properly set. The revision to OP 4822 received an Independent Safety Review, Site Manager approval, and was issued on 09/12/02.

"* OP-7107 "Spent Fuel Pit Component Movement Program" provides programmatic controls for fuel handling operations and includes a prerequisite to verify SFP ARM operability per OP-4822 prior to fuel handling operations. OP 7107 was revised and as part of the revision the prerequisite to verify SFP ARM operability now highlights that proper SFP ARM alarm setpoints are also verified per OP-4822. In addition, the revision to OP-7107 incorporates a new Attachment D "Shiftly Prerequisite Verification" which includes performance of OP-4822 at 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> intervals during fuel handling operations. The revision to OP-7107 received an Independent Safety Review, Site Manager approval, and was issued on 09/17/02.

As additional defense in depth to ensure proper SFP ARM alarm setpoints during fuel handling operations, the planned revision to 13200-OP-2961 "Fuel Load into TSC/Transfer Cask" will include appropriate steps to initiate SFP ARM alarm setpoint verification when work activities external to the SFP could affect SFP background radiation levels.

  • The revision to 13200-OP-2961 will be issued prior to the resumption of fuel loading activities for TSC #7 which is anticipated in mid-November 2002.

ADDITIONAL INFORMATION None.

NRC FORM 366A (1-2001)

NRC FORM 366AU.S. NUCLEAR REGULATORY COMMISSION (1-2001)

LICENSEE EVENT REPORT (LER)

1. FACILITY NAME 2. DOCKET 6. LER NUMBER 3. PAGE SEQUENTIAL REVISION YEAR NUMBER NUMBER Yankee Nuclear Power Station 05000029 2002 -- 001 -- 00 4 OF 4
17. NARRATIVE (If more space is required,use additonalcopies of NRC Form 366A)

PREVIOUS SIMILAR EVENTS Licensee Event Report 2001-001-00, dated August 22, 2001, addressed the Spent Fuel Pit Area Radiation Monitor Alarm setpoints being set above the limits allowed by Technical Specification 3.3. In that the issue of SFP ARM alarm setpoints is common to LER 2001-001 and this LER (2002-001) an evaluation of the effectiveness of the corrective actions for LER 2001-001 has been performed. The corrective actions for LER 2001-001 included a significant revision to OP-4816 " Functional Test and Alarm Setting of the Area Radiation Monitoring System" to provide clearer and more consistent guidance on Technical Specification 3.3 requirements and in particular the method to determine proper SFP ARM alarm setpoints based on maximum observed background. Since issuance on July 13, 2001, the revision to OP-4816, training, and heightened personnel awareness during fuel handling operations have proven to be effective in ensuring proper SFP ARM alarm setpoints during performance of OP-4816. The issue that caused this LER (2002-001) was the lack of defense in depth in supporting procedures (OP-4822, OP-7107, and 13200-OP-2961) to anticipate the effect of dry cask storage activities adjacent to the SFP which can change SFP ambient background radiation levels (and SFP ARM alarm setpoint requirements) during fuel handling operations.

NRC FORM 366A (1-2001)

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