05000315/LER-1998-001, :on 980104,containment Air Recirculation Sys Flow Testing Results Indicated Condition Outside Design Basis.Caused by Mispositioned Valve Actuator.Corrected Actuator to Valve Orientation

:on 980104,containment Air Recirculation Sys Flow Testing Results Indicated Condition Outside Design Basis.Caused by Mispositioned Valve Actuator.Corrected Actuator to Valve Orientation

ML17334B761
Person / Time
Site:	Cook
Issue date:	05/08/1998
From:	Sampson J, Schoepf P INDIANA MICHIGAN POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-001, LER-98-1, NUDOCS 9805140266
Download: ML17334B761 (10)

LER-1998-001, on 980104,containment Air Recirculation Sys Flow Testing Results Indicated Condition Outside Design Basis.Caused by Mispositioned Valve Actuator.Corrected Actuator to Valve Orientation

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(x)
3151998001R00 - NRC Website

text

CATEGORY REGULA RY INFORMATION DISTRIBUTI SYSTEM (RZDS)

ACCESSION NBR:9805140266 DOC.DATE: 98/05/08 NOTARIZED: NO DOCKET FACIL:50-315 Donald C.

Cook Nuclear Power Plant, Unit 1, Indiana M

05000315 AUTH.NAME AUTHOR AFFILIATION SCHOEPF,P.

Indiana Michigan Power Co.

SAMPSON',J.R.

Indiana Michigan Power Co.

RECIP.NAME RECIPIENT AFFILIATION SUBJECT:-LER 98-001-02:on 980104,containment air recirculation sys flow testing resulted condition outside design. Caused by sys unbalanced

&, mispositioned valve. Actuator to valves orientation was corrected.W/980508 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

E RECIPIENT ID CODE/NAME PD3-3 PD INTERNAL: FI CE EELB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRPM/PECB NRR/DSSA/SRXB RGN3 FILE 01 EXTERNAL: L ST LOBBY WARD NOAC POORE,W.

NRC PDR COPIES LTTR ENCL 1

1 2

2 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 RECIPIENT ID CODE/NAME STANG,J AEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DET/EIB LITCO BRYCE,J H NOAC QUEENER,DS NUDOCS FULL TXT COPIES LTTR ENCL 1

1 1

1 1

1 1

l.

1 1

1 1

1 1

1 1

1 1

1 1

1 1

D U

E N

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)

ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED:

LTTR 24 ENCL 24

rr olaria O'oorgao p".;ier Company INDIANA MICHIGAN POWER May 8, 1998 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Operating Licenses DPR-58 Docket No. 50-315 Document Control Manager:

h 8

bl h dby10CFR i d~

em the following report is being submitted:

98-001-02 Sincerely, J. R. Sampson Site Vice President lmbd Attachment A. B.

Beach, Region III J. R.

Sampson P. A.

Barrett S. J.

Brewer R.

Whale D.

Hahn Records Center, INPO NRC Resident Inspector 9805140266

'rr80508 PDR ADQCK 050003i5 S

PDR

NRc FDRH 366 (5-92)

S.

NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

PROVED BY OMB No. 3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER

RESPONSE

TO COMPLY WITH THI INFORMATION COLLECTION REQUEST: 50.0 HRS.

FOR'WAR COMMENTS REGARDING BURDEN ESTIMATE To TH IHFORMATION AND RECORDS MANAGEMENT BRANCH (HNB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, WASHIHGTON, DC 20555-0001, AND TO THE PAPERWOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MANAGEHEHT AND BUDGET WASHIHGTON DC 20503.

FACILITY NAME (1)

Donald C. Cook Nuclear Plant - Unit 1 DOCKET NUMBER (2) 50.315 Page1 of5 TITLE (4)

Containment AirRecirculation System Flow Testing Results Indicate Condition Outside the Design Basis EVENT DATE' LER NUMBER 6

REPORT DATE 7

OTHER FACILITIES IHVOLVED 8

HONTH DAY 01 04 YEAR YEAR 98 98 SEQUENTIAL NUMBER 001 REVISION NUMBER 02 MONTH DAY 05 08 YEAR 98 FACILITY NAME FACILITY NAHE DOCKET NUHBER DOCKET NUMBER OPERATING MODE (9)

POWER LEYEL (10) 00 THIS REPORT IS SUBMITTED PURSUANT 20 '201(b) 20.2203 a

'1 20.2203(a)(2)(i) 20.2203(a)(3)(l) 20.2203 a

ii 20.2203(a)(4) 50.73(a)(2)(iii) 50.73 a

2 iv 50.73(a)(2)(v) 73.71(b)

73. 710 OTHER e

11 TO THE REQUIREMENTS OF 10 CFR Check one or mor 20.2203 a

2 ii 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(2)(v) 50.36 c 1

~

50.36(c)(2) 50.73(a)(2)(l)

X 50.73(a)(2)(ii)

LICEHSEE CONTACT FOR THIS LER 12 50.73 a

2 vii 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(x)

(Specify in Abstract betoii and in Text,NAME Mr. Paul Schoepf-Safety Engineering Mechanical Systems Manager TELEPHONE NUMBER (Include Area Code) 616/465-5901, x2408 COMP ETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13

CAUSE

SYSTEH COMPONENT MANUFACTURER REPORTABLE TO NPRDS

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS SUPPLEMENTAL REPORT EXPECTED 14 YES (If yes, complete EXPECTED SUBMISSION DATE).

X NO EXPECTED SUBMISSION DATE (15)

MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typeMritten lines) (16)

During the month of January, 1998, airflow testing of the Unit 1 and Unit 2 containment air recirculation/hydrogen skimmer (CEQ) system was performed. The test demonstrated that as-found flows in certain steam generator and pressurizer compartments, fan-accumulator rooms, and instrument rooms were less than the flows stated in UFSAR Section 5.5.3. An ENS notiTication was made on January 5, 1998 in accordance with 10CFR50.72(b)(2)(l) and updated on January 19, 1998. An interim LER for this'ondition was submitted on February 3, 1998, in accordance with 10CFR50.73(a)(2)(ii) as a condition outside the design basis, as is this updated LER.

ln general the low flow results are attributed to the system not being balanced and to system design, as well as the conditions under which the system was tested. In addition, an incorrectly installed valve actuator affected the Unit 1 CEQ system performance. A modification was installed on both unit's CEQ system to increase the air flowfrom the pressurizer enclosure. The procedure used to set the actuator and valve orientation for the affected Unit 1 valve was enhanced. Through post modification testing and analysis, both unit's CEQ systems meet system performance acceptance criteria as defined in Chapter 5 of the FSAR.

The pressurizer and steam generator compartments have been analyzed for hydrogen generation using the Unit 1 Cycle 16 and Unit 2 Cycle 12 values. The results of the analysis shows that the hydrogen stays below the 4 percent volume limit. Further analyses willbe performed to determine that the Unit 1 Cycle 16 and Unit 2 Cycle 12 values are bounding for all cycles for which the low flow condition was present.

Based on initial engineering review, hydrogen concentration in the lower containment annulus willbe below the acceptance criteria. Additional cafcuations are expected to be completed by July 15, 1998. Should the results indicate hydrogen values beyond four percent volume in any of the subcompartments, an updated LER willbe submitted.U CLEAR REGULATORY COMMISSION LICENSEE EVENT CONTINUATION PROVED BY OMB NO. 3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER

RESPONSE

TO COMPLY HITN THI INFORMATION COLLECTIOH REOUEST: 50.0 MRS.

FORHAR COMMENTS REGARDIHG BURDEH ESTIMATE TO TH INFORMATION AHD RECORDS MANAGEMENT BRANCH (MHB 7714),

U.S.

NUCLEAR REGULATORY COMMISSIOH, WASHINGTON, DC 20555-0001, AHD TO THE PAPERWOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MANAGEMENT AHD BUDGET

'NASHIHGTOH DC 20503.

FACILITY HAME 1

DOCKET NUMBER 2

YEAR LER NUMBER 6

SEQUENTIAL REVISION PAGE 3

Cook Nuclear Plant - Unit 1 TEXT (if core space is reciuired.

use addit(onal NRC Fore 366A's) (I7) 50-315 98 001 02 2 OF 5 ndii nsPri r c urr n Unit 1 was in Mode 5, Cold Shutdown Unit 2 was in Mode 6, Cold Shutdown D

ri in fh vn During testing of the containment air recirculation/hydrogen skimmer (CEQ) system, certain flows in the steam generator and pressurizer compartments, fan-accumulator rooms, and instrument room were found to be less than the flows stated in UFSAR Section 5.5.3.

The hydrogen skimmer system is designed to control the hydrogen concentration in the compartments below 4 volume percent. This is accomplished by, purging the compartments using a portion of the atmosphere in the lower containment. This flow mixes the atmosphere from the lower containment with the atmosphere in the compartment, diluting any hydrogen that is generated in the compartment. The lower flows could reduce the effectiveness of this mixing effect.

The low flowconditions are attributed to the system being unbalanced, as well as, the conditions in which the system was tested.

The test condition is described as system recirculation flow through equipment hatches versus the ice condenser.

In this configuration, the hydrogen skimmer flowfrom the steam generator, pressurizer doghouses, and fan rooms, are reduced.

After balancing the system, it was determined analytically that the flow from the pressurizer doghouse would marginally meet or fail to meet its design criteria as defined in Chapter 5 of the FSAR. This was considered a design problem related to the pipe size serving the pressurizer doghouse.

With respect to 1-HV-CEQ-1, the root cause of the event was a mispositioned valve actuator for valve 1-VMO-101 (Unit 1 fan 1-HV-CEQ-1) in the hydrogen skimmer system. Valve 1-VMO-101 is a motor operated butterfly valve which opens when the recirculation fan starts. The installed actuator allowed the valve to open 45 degrees from the centerline in the fullyopen and the fullyclosed positions.

When the valve actuator is placed on a valve,-it is aligned with a valve shaft keyway which is parallel to the valve disc. Valve 1-VMO-101 has an extension, and the extension has a keyway which is rotated 45 degrees from the valve shaft keyway's position. The actuator was installed in relation to this keyway rather than the valve shaft keyway. This occurred because the procedure which specified that the actuator was to be installed did not contain specific instruction to use the.valve shaft keyway as a reference:

Additionally, the valve testing program and the post maintenance testing (PMT) requirements did not detect this condition. Because of the actuator's stop settings, the increased torque observed during valve operation was attributed to valve seating when it was due to contact with the stops. The PMT test only checked that the actuator was properly functioning.U.

CLEAR REGULATORY COMMISSION LICENSEE EVENT CONTINUATION ROVED BY OMB HO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER

RESPONSE

TO COMPLY HITH THI IHFORMATIOH COLLECTIOH REQUEST: 50.0 HRS.

FORIIAR COMMENTS REGARDIHG BURDEN ESTIMATE TO TH INFORMATION AHD RECORDS MANAGEMENT BRANCH (MHB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, HASHIHGTON, DC 20555-0001, AHD TO THE PAPERWOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MANAGEMENT AHD BUDGET lJASNINGTOH DC 20503.

FACILITY NAME DOCKET HUMBER 2

YEAR LER NUMBER 6

SEQUENTIAL REVISIOH PAGE 3

Cook Nuclear Plant - Unit 1 50-315 98 001 02 3OF5 TEXT (ifsore space is required.

use addlt(onal NRC Fore 366A's) (17) riz I'

nAnl i

A hydrogen compartment analysis has been recently completed.

The major difference of this analysis from the previous hydrogen compartment analysis is an increase in core-wide metal-water reaction. This analysis was performed with the Unit 1 Cycle 16 and Unit 2 Cycle 12 specific core-wide metal-water reaction values.

The analysis was also performed with a nominal pressurizer skimmer flowof 301 cfm; this is the 2-HV-CEQ-2 as-left pressurizer flow minus the 30 cfm measurement uncertainty.

The maximum hydrogen concentration in the pressurizer enclosure was calculated by using a 2-inch Design Basis Accident (DBA) hydrogen release to the pressurizer enclosure.

This analysis assumes 100 percent of the core radiolysis, 10 percent of the sump radiolysis, 10 percent of the metal-water reaction hydrogen, and 10 percent of a 2-inch break steam flow is directed into the pressurizer enclosure.

A 2-inch break was selected to delay operation of the CEQ/skimmer fans until the ice bed is melted out. Assuming 10 percent of the steam flow is directed to the pressurizer enclosure is conservative since 100 percent of the core radiolysis is directed to the pressurizer enclosure.

Analysis shows the hydrogen concentration for the first 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> is less than the flammable limitof 4 volume-percent with a nominal pressurizer flowof 301 cfm.

The pressurizer enclosure hydrogen concentration continues to show a long term increase because the compartment analysis does not include removal of hydrogen by the hydrogen recombiners.

The Iumped-volume containment hydrogen analysis is used to demonstrate that the hydrogen recombiners maintain the hydrogen concentration below the flammable limit. A lumped-volume analysis with 5 percent metal-water (a value greater than the Unit 1 Cycle 16 and Unit 2 Cycle 12 values) demonstrated that the hydrogen recombiners maintain a hydrogen concentration below 4 volume-percent.

As discussed above, this analysis (and the steam generator analysis discussed next) was performed with the Unit 1 Cycle 16 and Unit 2 Cycle 12 specific core-wide metal-water reaction values.

It has not yet been determined whether these cycle specific values bound all cycles for which the identified condition was present.

This work willbe done in conjunction with the development of the instrument room and fan-accumulator room models discussed below.

am n

rCo a

enAnl i

Like the pressurizer enclosure, the maximum hydrogen concentration in the steam generator enclosure was obtained by using a 2-inch design basis accident hydrogen release to the steam generator enclosure.

An analysis of the steam generator enclosure was performed with a hydrogen skimmer flow of 500 cfm per pair of steam generators.

This flow rate (500 cfm) is the design value for the steam generator enclosures.

The steam generator enclosure analysis was repeated with a nominal hydrogen skimmer flowof 207 cfm per steam generator pair. This analysis input value is the 1-HV-CEQ-1 as-found skimmer flowof 237 cfm for steam generators 2 8 3 minus the measurement uncertainty of 30 cfm. Even with the reduced flow, the steam generator enclosure hydrogen concentration remains below 3 volume-percent and the flammable limitof 4 volume-percent.

As with the pressurizer enclosure analysis, the lumped-volume containment hydrogen analysis is used to demonstrate that the hydrogen recombiners maintain the hydrogen concentration below the flammable limitthroughout containment.U UCLEAR REGULATORY COMMISSION PROVED BY OMB HO. 3150.0104 EXPIRES 5/31/95 LICENSEE EVENT CONTINUATZON ESTIMATED BURDEN PER

RESPONSE

TO COMPLY liITH THI INFORMATIOH COLLECTION REQUEST: 50.0 HRS.

FORIIAR COMMENTS REGARDIHG BURDEN ESTIMATE TO TH INFORMATION AHD RECORDS MANAGEMENT BRANCH (MHB 7714),

U.S.

HUCLEAR REGULATORY COMMISSION, NASHIHGTOH, DC 20555-0001, AND TO THE PAPERWOR REDUCTION PROJECl'3150-0104),

OFFICE

, 0 MANAGEMENT AHD BUDGET MASHINGTOH DC 20503.

FACILITY NAME 1

DOCKET NUMBER 2

YEAR LER NUMBER 6

SEQUEHTIAL REVISIO}i PAGE 3

Cook Nuclear Plant - Unit 1 TEXT (if core space is required.

use additional NRC Fora 366A's) (17) nl i

f v n'0-315 98 001 02 4OF5 I

In I

Though the current hydrogen compartment analysis is sufficiently detailed to calculate the hydrogen concentration in the pressurizer and steam generator enclosures',

it does not model the separate rooms in the annulus.

Based on initial engineering review, the subcompartment hydrogen concentrations willbe below the acceptance criteria.

The renodalization of the containment computer model and review of core-wide metal-water reaction amounts for previous cycles is ongoing. Should calculation results, which are expected by July 15, 1998, indicate hydrogen values beyond four percent volume in any of the subcompartments, an updated LER will be submitted.

rB The mispositioned valve 1-VMO-101 would not only affect the hydrogen skimmer system flows following a LOCA event, but would also affect the post-LOCA ice condenser performance by creating a bypass flow path.

The valve's disk position of 45 degrees, rather than fullyclosed, during the initial 9 plus or minus 1 minutes of the large break LOCA event provides a flow path for blowdown steam directly to upper containment, bypassing the ice condenser.

Greater bypass flow increases the peak pressure due to compression of noncondensible gases in upper containment.

(This is not the maximum post-LOCA peak pressure; the maximum post-LOCA peak pressure occurs after ice bed melt out due to containment heating and decreases with increasing bypass flow.) The compression-induced peak pressure was calculated with a bypass area of 5 square feet. The 5 square feet consists of 2.2 square feet of known bypass flowarea plus a margin of 2.8 square feet.

In addition, a sensitivity analysis has shown that up to 35 square feet of bypass area is acceptable.

The mispositioned valve (1-VMO-101) has a nominal inside diameter of 14 inches.

Conservatively neglecting the valve disc, the flowarea through the valve is 1.07 square feet. Since this flowarea is less than the 2.8 square feet bypass area margin, the containment analysis is not affected by the misposition of valve 1-VMO-101.

The actuator to valve orientation was corrected, and the actuator stops were changed to ensure that valve testing would detect valve seating. Flow testing of the system was conducted to ensure proper orientation of the valve actuator.

Reviews of other valves where this condition might occur were made, and three additional valves were identified.

These valves were examined, and the actuators were found to be properly oriented.

To improve the flowconditions from the pressurizer doghouse, a Plant Modification (12-DCP-876) was performed.

The modification replaced a section of 6 inch diameter pipe serving the pressurizer doghouse with an 8 inch diameter pipe. The new pipe section was placed between the 6 inch pressurizer doghouse balancing valve and the 12 inch diameter main pipe.

Each CEQ system performance was demonstrated through post modification testing, and analysis, to meet all acceptance criteria as defined in Chapter 5 of the FSAR.

HRC FORM 366A U

UCLEAR REGULATORY COMMISSION LICENSEE EVENT CONTINUATION PROVED BY OMB HO. 3150.0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER

RESPONSE

TO COMPLY WITH THI INFORMATION COLLECTIOH REQUEST: 50.0 MRS.

FORWAR COMMENTS REGARDING BURDEN ESTIMATE TO TH INFORMATION AND RECORDS MANAGEMENT BRANCH

<MHB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MANAGEMENT AHD BUDGET WASHIHGTOH DC 20503.

FACILITY NAME 1

Cook Nuclear Plant - Unit 1 DOCKET NUMBER 2

50-315 YEAR 98 LER NUMBER 6

SEQUENTIAL 001 REVISION 02 PAGE 3

5OFS TEXT (if aore space is required.

use addit(ona)C rrectiv Acion con'd CEQ Post-Mod As-Left Flows at Accident Condition

~a SG¹1 and2 Pressurizer SG¹1 and 4 East Fan Rm Instrument Rm West Fan Rm Cont. Dome Recirc

~1-V~Q-1 541 CFM 530 CFM 643 CFM 127 CFM 138 CFM 113 CFM

. 1193 CFM 41, 657 CFM.

1-HV-(+~

520 CFM 610 CFM 602 CFM 141 CFM 153 CFM 125 CFM 1211 CFM 42,389 CFM PJ~V-CE~1 664 CFM 638 CFM 539 CFM 135 CFM 117 CFM 119 CFM 1335 CFM 41,962 CFM g-HV-Qgg 520 CFM 550 CFM 554 CFM 136 CFM 128 CFM 149 CFM 1383 CFM 44,163 CFM The maintenance procedure was enhanced to ensure that the valve shaft keyway is used to set the actuator and valve orientation and not the extension shaft keyway.

The valve testing program is being revised to allow valve disc seating to be identified.

r v'milar v n LER 316/97-009-00