ML18030B038

Discusses Apparent Discrepancy Between Actual Plant Design & FSAR Re Closure Times of Various Secondary Containment Isolation Dampers.Ge Rept Mde 247-1185 Encl

ML18030B038
Person / Time
Site:	Browns Ferry
Issue date:	01/23/1986
From:	Domer J TENNESSEE VALLEY AUTHORITY
To:	Muller D Office of Nuclear Reactor Regulation
Shared Package
ML18030B039	List: ML18030B038 ML18030B041
References
NUDOCS 8601300129
Download: ML18030B038 (7)

Text

REGULATORY I RMATION DISTRIBUTION SYS (RIDS)

ACCESSION NBR: 8601300129 DOC. DATE: 86/01/23 NOTARIZED: YES DOCKET ¹ FACIL: 50-259 Brogans Ferry Nucleav Power Stations Unit ii Tennessee Brotuns Ferry Nucleav Power Stations Unit 2i Tennessee 05000260 05000259'0-260 50-296 Brains Fev rg. Nuclear Pouter Stations Unit 3i Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION DOMER> J. A. Tennessee Valley Authoritg RECIP. NAME RECIPIENT AFFILIATION MULLERi D. R ~ BWR Pv object Div ectorate 2

SUBJECT:

Discusses appav'ent discv epancg between actual plant design FSAR re closure times oF various secondary containment isolation damp ev's. GE Rep t MDE 247-1 185 encl.

DISTRIBUTION CODE: A001D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: OR Submittal: General Distv ibuti'on.

NOTES:NMSS'/FCAF icy. ice NMSS/FCAF/PM. 05000259 OL: 06/26/73 NMSS/FCAF icy. icy NMSS/FCAF/PM. 05000260 OL: 06/28/74 NMSS/FCAF icy. icy NMSS/FCAF/PM. 05000296 OL: 07/02/76 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL HNR ADTS 1 0 BNR PD2 PD 01 HLlR EB 1 1 BNR EICSB 1 1 HLlR FOB 1 1 CLARK> R 1 BIJR PSB 1 1 BNR RSH 1 1 INTERNAL: ACRS 09 6 6 ADM/LFMB 0 ELD/HDS4 0 NRR/DHFT/TSCB 1 1 NRR/DSRO/RRAB 1 1 NRR/ORAS 1 0 04 1 1 RQN2 1 EXTERNAL: 24X 1 EQhG BRUSKE. S 1 1 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 05 1 1 NOTES:

TOTAL NUMBER OF COPIES REQUIRED: LTTR 32 ENCL 28

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TENNESSEE VALLEY AUTHORITY 5N 105B Lookout Place January 23, 1986 Director of Nuclear Reactor Regulat;ion Attention: Mr. D. R. Muller, Project Director BMR Project Directorate No. 2 Division of Boiling Mater Reactors (BMR)

U.S. Nuclear Regulatory Commission Mashington, D.C, 20555

Dear Mr. Muller:

In the Matter of the Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 In a recent review of plant modification work, an apparent discrepancy was discovered between the actual plant design and the FSAR with regard to the closure times of various secondary containment isolation dampers at the Browns Ferry Nuclear Plant. Further investigat;ion revealed the Eollowing general weaknesses in the assumptions used in the analysis for the design basis fuel handling accident contained in chapter 14 of the FSAR.

The FSAR analysis assumed that the ventilat;ion dampers for the secondary containment would close in two seconds or less. However, only those dampers afEecting the refueling zone exhaust meet this criteria.

The FSAR analysis assumed that the transport time of radionuclides through the ventilation system was less than the two second damper closure time. This assumption resulted in no ground level release being included in the offsite dose assessment. However, inspection of the vontilat;ion system configuration revealed t;hat this assumption was likely to be nonconservative.

To resolve these questions the design basis fuel handling accident was reanalyzed by General Electric Company (GE) using conservative assumptions and analytical methods. The results oE the analysis, which was perEormed Eor ventilation damper closure times ranging Erom the original two seconds up to 10 seconds, yielded offsite doses of 320 mrem to 890 mrem thyroid and 5.3 mrem to 11.1 mrem whole body at tho site (exclusion area) boundary. These results for all damper closure times meet the guidelines for acceptability given in NUREG-0800, section 15.7.4.

86013001& 860123-PDR ADOCK 0500025%

F PDR An Equal Opportunity Employer

Director of Nuclear Reactor Regulation January 23, 1986 AL tho request of the NRC-Region IX resident inspectors, TVA evaluated the change in the design basis Fuel handling accident analysis in accordance with 10 CFR 50.59 although no change to the actual plant or plant operation was involved. This evaluation was perEormed in accordance with Browns Ferry Nuclear Plant procedures which aro based on IE Circular No. 80-18 and concluded that an unroviewed safety question was not involved. The basis in part is that although the consequences of the Fuel handling accident have increased over the previous analysis, the results are clearly acceptable when evaluated against applicable regulaLory guidance of the Standard Review Plan and are, therefore, within the bounds of what has been previously analyzed.

Dieect application of XE Circular No. 80-18 yields the same conclusion.

In order to resolve concerns raised by NRC Region XX concerning this issue, we are forwarding for your review our Unreviewed Safety Question Determination (USQD) as enclosure 1. Me request that you review oue USQD and inform us whether you disagree or concur with it. Xf you disagree with our conclusions, we request that you please provide guidance on the applicability of IE Circular No. 80-18 to similar situations and clarify the limitations of its use. The radiological impact analysis for the design basis Euel handling accident, GE report HDE 247-1185, is provided as enclosure 2.

N>ile this issue remains unresolved, it is necessary that we impose operating limitations on reFueling Floor activities in order to avoid possible enForcement action. Sle, therefore, request that this matter be eesolved expeditiously. Your assistance in obtaining a clear resolution is appreciated.

If you have any questions, please get in touch with us.

Very truly yours, TENNESSEE VALLEY AUTHORITY Nuclear Licensing Beanch Subscribed an/ s~ orn t efor me on this cd ay of 1986.

otar Public Hy Commission Expires~

Enclosures cc: see page 3

Director of Nuclear Reactor Regulation January 23, 1986 cc (Enclosures):

Mr. R. J. Clark U.S. Nuclear Regulatory Commission Brooms Ferry Project Manager

'920 Norfolk Avenue Bethesda, Maryland 20814 U.S. Nuclear Regulatory Commission Region IX ATTN: Dr. J. Nelson Grace, Rogional Administrator 101 Marietta Street, NM, Suite 2900 Atlanta, Georgia 30323

Enclosure 1 Unreviewed Safety Question Determination Descri tion of Activit To define acceptable conditions for secondary containment to be considered intact. The FSAR currently states that reactor zone isolation dampers will close in two seconds after detecting high radiation in the duct from a fuel handling accident. The FSAR does not address the question of refueling zone isolation dampers, which would prove to be the limiting factor in a fuel handling accident. This USQD will document that the fuel handling accident has been analyzed for all secondary containment isolation dampers closing in up to 10 seconds. This analysis predicts increased dose rate over the FSAR due to more analytical conservatism. This value is greater than the original FSAR data, but a very small part (<10%) of 10 CFR 100 and below the EPA protective action guides requirements to issue an alert.

The probability of a fuel handling accident is unaffected by isolation damper, time. The consequences of a fuel handling accident are not significantly increased because the site dose for a 10 second closure time remains only a small part of the 10 CFR 100 reactor site criteria.. NRC has provided guidance (Standard Review Plan NUREG 800) which judges a "small part" to mean less than 10'L. Dose rates within NRC guidance are previously evaluated.

Therefore the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased.

2 ~ No new activities or equipment are involved; the design basis accident remains the fuel handling accident which has been analyzed.

Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report is not created.

3. Specific site boundary doses for a design basis accident are not given in the bases for the technical specifications. The bases for secondary containment reference the 10 CFR 100 criteria addressed in question 711. Therefore, the margin of safety as defined in the basis of any technical. specification is not reduced.

Damper times are not specified in technical specifications therefore a technical specification change is not required.