ML20045A015

LER 92-011-02:on 920608,determined That Design Inadequacies Could Result in Exceeding 10CFR100 Limits.Caused by Failure of Original Design to Specify leak-tight Const for Fan Housing.Shaft Seal Design installed.W/930528 Ltr

ML20045A015
Person / Time
Site:	Grand Gulf
Issue date:	05/28/1993
From:	Burke B, Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GNRO-93-00065, GNRO-93-65, LER-92-011, LER-92-11, NUDOCS 9306090242
Download: ML20045A015 (7)

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May 28, 1993 U.S. Nuclear Regulatory Commission Mail Station Pl-137 Washington, D.C. 20555 Attention: Document Control Desk ,

1 SUBJECT : Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 '

License No. NPF-29 Design Inadequacies Could Result in Exceeding 10 CFR 100 Limits LER 92-011-02 l

i GNRO- 93/ 00065 Gentlernen:

Attached is Licensee Event Report (LER) 92-011-02 which is a final report.

Yours truly,

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CRH/BAB/cg ]

attachment  !

cc: Mr. R. H. Barnhard )

Mr. D. C. Hintz (w/a)  !

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o) i Mr. Stewart D. Ebneter (w/a)

Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Mr. P. W. O'Connor off*ce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555 9306090242 930528 '

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4 May 28, 1993 GNRO'-93/ 00065 Page 2 of 2 bec: Mr. P. E. Alberstadt (w/a)

Mr. C. A. Bottemiller (w/a)

Mr. B. A. Burke (w/a)

Mr. R. W. Byrd (w/a)

Mr. L. F. Daughtery (w/a)

Mr. M. A. Dietrich (w/a) '

Mr. J . P. Dimmette , Jr. (w/a)

Mr. J. L. Ensley (ES1) (w/a)

Mr. J. J. Fisicaro (ANO) (w/a)

Mr. J . O. Fowler (w/a) 7 Mr. C. C. Hayes (w/a)

Mr._W. K. Hughey (w/a)

Mr. C. R. Hutchinson (w/a)

, Mr. L. W. Laughlin (W3) (w/a)

Ms. F. K. Mangan (w/o)

Mr. J . R. McGaha (w/a)

Mr. M. J. Meisner (w/o)

Mr. R. V. Moomaw (w/a) ,

Mr. D. L. Pace (w/a)

Mr. R. L. Patterson (w/a) ,

Mr. T. E. Reaves (w/a) '

Mr. J. C. Roberts (w/a)

Mr. J. L. Robertson (w/a)

Mr. S. A. Saunders (w/a)

Mr. G. D. Swords (w/a)

Mr. R. G. West (w/a) ,

Mr. G. A. Zinke (w/a)  !

SRC Secretary (w/a)  !

File (LCTS) (w/2)

File (RPTS) (w/a)  ;

File (NL) (w/a) u File (Central) (w/a) ( 6 ) 1

-l INPO Records Center (w/a) <

700 Galleria Parkway Atlanta, Georgia 30339-5957 Mr. F. A. Spangenberg (w/a)

Illinois Power Company Clinton Power Station P.O. Box 678 Clinton, Illinois 61727

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1 Mr. D. Lorfing (w/a) )

Riverbend Nuclear Station Mail Stop MA-3 )

Gulf States Utilities i P.O. Box 220 St. Francisville, LA 70775 l

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Attachment to GNRO-93/00065 NRC Form 366 U S NUCLE.OR REGULATORY COMMt11 ION

@ th APPROVED oms NO 3160 0104 LICENSEE EVENT REPORT (LER) '*"'53'"

F ACILIT Y N AME til DOCK E T NUMSE R (2l # AGE 'l Grand Gulf Nuclear Station - l' nit 1 o ls j o ;o j o l4 l1 l 6 3 lorj0 g 4 TITLE a4r Design inadequacies Could Result in Exceeding 10 CFR 100 Limits EVENT DATE ($1 LE R NUMBE R 461 REPORT DATE 171 OTHER F ACILITIES INVOLVED IN MONTH OAv vFAR vEAp SEf (' ' l

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On June 8,1992 during a review of previously identified nonconforming conditions consisting of potential )

secondary containment bypass leakage paths and potential control room boundary leakage paths, it was l determined that Grand Gulf Nuclear Station had been in a condition that was outside the design basis of the plant. The cause of this condition is the failure of the original design to specify leak-tight construction for the fan housings for the filter trains of the standby gas treatment system and the control room standby fresh air units. A contributing factor was that previous offsite and control room dose calculations did not reflect the installed design of the plant.

1 immediate corrective actions included installation of a shaft seal design and sealing of the other identified leakage paths. Reanalysis resulted in calculated accident dose values which confirm continued compliance with 10 CFR limits. Therefore, health and safety of the public were not compromised due to this condition.

The control room and offsite dose calculations were updated to reflect improved methodologies and additional l information that is now available. The UFSAR will be revised to reflect this update of the dose calculations. This report is also being submitted pursuant to 10 CFR 21.

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, Attachment to GNRO-93/00065 NRC Form 3s64 U $ NUCLE AR REGUL ATORY COMMISSION !

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION emovEo ovs so mo+o4

, EXPlaES' 8'31 80 F ActLITV NAME 01 DOCKET NUMBE R (2) LER NUMBER 16# PAGE(3)

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ten a me e.c a mm u m.mc nm m si on l A. Reportable Occurrence On June 8,1992 following a review of previously identified nonconforming conditions consisting of potential secondary containment bypass leakage paths and potential control room boundary leakage paths, it was determined that Grand Gulf Nuclear Station had been in a condition that was outside the design basis of the plant. This condition is being reported pursuant to 10 CFR 50.73(a)(2)fii)(B) as a condition that was outside the design basis of the plant, and 10 CFR 50.73 (a)(2)(vii) as an event where a single condition caused independent trains to become inoperable in a single system designed to control the release of radioactive material and mitigate the consequences of an accident. This report is also being submitted pursuant to 10 CFR 21.

B. Initial Conditions The plant was in Operational Condition 2 at approximately 6 percent MWt at the time this condition was deiermined to be reportable.

C. Description of Occurrence On October 9,1991, a nonconforming condition was identified in that a temporary shaft seal constructed of a foam rubber type material was installed on the standby gas treatment system (SGTS) [BH] charcoal filter train fans (Q1T4BD001 A-A and Q1T48D001B-B). The primary safety concern focused on whether the secondary containment drawdown capability was adversely affected by the temporary seals. GGNS System Engineering concluded that drawdown capability was not affected. The specific identified nonconforming conditions were corrected on November 15,1991 for SGTS B and December 12,1991 for SGTS A by removal of the temporary seals.

GGNS Design Engineering reviewed the design specification for filter trains and confirmed that the fans were I specified and procured by Bechtel Power Corporation with no shaft seals installed. Upon further investigation, it I was concluded that the opening or gap between the fan hub and the hole in the fan housing could result in air being drawn into the fan housing and discharged to the environment without being processed through the charcoal filter trains. The GGNS design basis accident analyses did not assume any secondary containment bypass leakage.

Therefore it was concluded that an unanalyzed potential secondary containment bypass leakrige path existed with the current design The manufacturer of the fans (Buffalo Forge Cnmpany) was contacted to determine how much, if any, leakage into the f ans could occur with no mechanism in place to seal the gap between the fan shaft and the housing. On April 28,1992 Buffalo Forge Company provided to GGNS a conservative estimate of 100 I

CFM inleakage. On May 1,1992 an engineering evaluation was completed considering the effects of the deficiency for the then current operational conditions (refueling and cold shutdown). The evaluation concluded that SGTS could perform its design safety function of controlling accident radiological releases during refueling operations following two weeks of decay.

l Design Engineering evaluated all of the other filter trains in the plant to determine if this was a generic problem.

This evaluation revealed that the control room standby fresh air (CRSFA) units [VI) (OSZ51D002A-A and OSZ51D0028 8) had no shaft seals which could have resulted in unfiltered air being supplied to the control room environment. A modification package was prepared to provide a shaft seal design for the SGTS charcoal filter train fans and the CRSFA units On May 20,1992 dunng a walkdown of the SGTS charcoal filter train fans, additional leakage paths were j identified. Rework instructions for these additional leakage patiis were included in the modification package.

l The work packages instalhng the seals and reworking the other idertified leakage paths were signed off complete on June 3,1992. (Note: The work process for correcting the above discussed condition resu,ted in a different reportable condition which was discussed in Licensee Event Report 92-012-01.)

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' Attach-nne en GNRO-93/00065 NCC Form 384G U.S NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONilNUATION areRoveo oms No ano-cio4

. EXPIRES 8/31'88 FOCluTV NAME Of DOCK E T NUMBE R (2) Lf R NUMBER 181 PAGE(3)

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0{7 O!1 0 !A mi o mwe a,m a renumt, m estioonewc form aesaw on On June 8,1992 this condition was determined to be reportable as a condition that was outside the design basis of the plant.

D. Apparent Cause The cause of this condition is the failure of the original design to specify leak-tight construction of fan housings for the filter trains of the SGTS and CRSFA subsystem. A contributing factor was that the previous offsite and control room dose calculations did not reflect the installed design of the plant.

E. Corrective Actions An evaluation of similar leakage pathways for ventilation fans was performed. The CRSFA was the only other similar condition. Leakage paths for SGTS and CRSFA were sealed by June 10,1992. l The leakage associated with the identified conditions was not measured due to the configuration of the leakage pathways. A qualitative assessment of the potential leakage based on smoke tests estimated the leakage to be minimal for the various identified leakage paths. However, in order to assess the safety significance of the potential bypass leakage paths, conservative maximum expected leakages were calculated. The values calculated were 73 CFM for control room envelope leakage and 215 CFM for secondary containment bypass leakage. (The assessment conservatively used an analytical value of 250 CFM for the secondary containment bypass leakage.)

Using the original GGNS UFSAR design basis analysis computer code, the input data was modified to incorporate the conservative calculated bypass leakages. These initial results indicated that the low population zone and the control room thyroid doses exceeded 10 CFR limits.

In order to more accurately assess the safety significance of the as-found condition, potential exposures were assessed using the same computer code and the same reactor core release source terms as the previous design basis UFSAR LOCA analysis. This interim assessment of dose calculations resulted in values within 10 CFR limits. Assumptions used were specified previously in the interim LE!R report.

The control room and offsite dose calculations were updated to reflect improved methodologies and assumptions as discussed with the NRC in a meeting on April 6,1993. Reanalysis has been performed using the improved methodology and has confirmed that the original design basis dose values were very conservative.

F. Supplemental Corrective Actior.s The UFSAR will be revised to reflect the updated dose calculations and models.

G. Safety Assessment The secondary containment in conjunction with operation of the standby gas treatment system is designed to limit the thyroid dose and the whole body dose to within the guidelines of 10 CFR 100 at the site boundary and the low population zone and 10 CFR 50 General Design Criterion 19 for the control room operator doses during the design basis accident. The control room habitability system is designed to limit the radiation exposure of control room personnel through the duration of any one of the postulated design basis accidents to within the guidelines of 10 CFR 50 General Design Criteria 19.

Table 1 itemizes assumptions used for the new design basis analysis which differ from the assumptions in the original design basis analysis. Table 2 compares 10 CFR limits with the original design basis accident dose values, accident dose values determined by the interim assessment as a result of this event, and the revised design basis accident dose values. The new design basis analysis results confirm continued compliance with 10 CFR limits. Therefore, health and safety of the public were not compromised due to this condition.

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Attachment to GNRO-93/00065

.- LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Acenoveo ov8 wo. 31so-oio4 a EXPIRES. 8/31/0B F ACILITY NAME (1) DOCKET NUMBER l2) LER NUMBER I6I PAGE131 9^a " $M -

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TABLE 1 UFSAR DESIGN BASIS ASSUMPTIONS DIFFERENT FROM THOSE IN ORIGINAL ANALYSIS l Suppression pool scrubbing (DF=10) assumed per Standard Review Plan 6.5.5 ICRP 30 dose conversion factors (to be invoked by 10 CFR 20)

Control room inleakage of 600 CFM (Operating License limit) including 10 CFM for ingress / egress SGTS Bypass leakage of 50 CFM Actual control room envelope volume of 253,000 cubic feet Updated Offsite X/O values TABLE 2 THYROID WHOLE BODY DESCRIPTIONS LPZ EAB CR LPZ EAB CR 10 CFR Limits (Rem) 300 300 30 25 26 5 Original UFSAR 73.91 103.4 29.54 15.81 12.42 0.67 Design Basis (Rem)

Interim assessment 96.0 20.1 21,1 10.2 6.8 0.42 (Rem)

Revised UFSAR 61.66 18.84 12.20 0.634 1.372 0 027 Design Basis (Rem) 1 H. Additional information Energy Industry Identification System (Ells) codes are identified in the text within brackets [ ).

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