ML17258A251

Advises That Review of Fuel Handling Accident Inside Containment Is Complete.Facility Adequately Designed to Mitigate Consequences of Accident.Sep Topic XV-20 Re Radiological Consequences of Fuel Damaging Accidents Encl

ML17258A251
Person / Time
Site:	Ginna
Issue date:	10/07/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-15-20, TASK-RR LSO5-81-10-008, LSO5-81-10-8, NUDOCS 8110220277
Download: ML17258A251 (8)

Text

October 7, 1981 Docket No. 50-244 LS05-81-10-008 tlr. John Haier Vice President Electric and Steam Production Rochester Gas

& Electric Corpor'ation 89 East Avenue Rochester, New York 14649 DISTRIBUTION Docket NRC PDR Local PDR ORB Reading NSIC DCrutchfield HSmith DSnaider GCwalina WRussell WPaulson OELD OI&E (3)

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SEPB OCT3.3 1981~

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SUBJECT:

FUEL HANDLING ACCIDENT INSIDE CONTAINMENT By letter dated January 14, 1977, the NRC staff requested that Rochester Gas and Electric Corporation evaluate the previously unevaluated potential consequences of a postulated Fuel Handling Accident Inside Containment at the R. E. Ginna Nuclear Power Plant.

You responded by letter dated tlarch 18, 1977, as supplemented by your letter dated IIlarch 27, 1979 that provided additional information requested by the NRC staff's letter dated January 3, 1979.

Although these letters provided the basis for the staff's initial conclusion that the required assumption of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time between reactor shutdown and fuel movement was conservative at Ginna. the review was delayed by events at Three tlile Island, the staff's derivation of revised X/g v'ilves based upon your original Exclusion Area Boundary (EAB) distances, and your selettion of new EAB distances as documented in your letter dated June 26, 1981.

By letter dated September 24, 1981, the NRC staff forwarded the evaluation of Systematic Evaluation Program (SEP) topic XV-20, entitled "Radiological Consequences of Fuel Damaging Accidents".

The conclusion of this topic assessment, based upon X/g valves determined from the new EAB distances, was,'that the Ginna plant "...is adequately designed to mitigate the con-sequences of this type of accident."

We therefore consider the review of the Fuel Handling Accident Inside Containment for Ginna to be complete.

A copy of the SEP topic XV-20 Assessment is":irlalbt;ed for your information.

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Nr. John E. Maier October 7, 1981 cc w/enclosure:

Harry H. Voigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire

Avenue, N. M.

Suite 1100 Washington,.D.

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20036 Nr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1,'Second Floor Empire State Plaza

Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N. W.

Suite 600 Mashington, D. C.

20006 U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman,

Esq, Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Washington, D. C.

20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, D; C.

20555 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519

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GINHR NUCLEAR GENERATING STATION XV-20 RAOIOLOGICAL CONSEQUENCES OF FUEL OAMAGING ACCIOEHTS INTROOUCTION The safety objective of this topic is to assure that the offsite doses from fuel damaging accidents as a result of fuel handling inside and out-a side containment are well within the guideline value of 10 CFR Part 100' I.

RE'/IE'A CRITERIA Section 50.34 of 10 CFR Part 50, "Contents of Applications:

Technica1 Information," requires that each applicant for a construction permit or operating license provide an analysis and evaluation of'he design and performance of structures,

systems, and components of the facility with the QbJective of assessing the r isk to public health and safety resulting from operation of the facility.

A fuel handling accident in the fuel handling and storage facility resulting in damage to fuel cladding.and subsequent release of radioactive material is one of the postulated accidents used to evaluate the adequacy of these structures,

systens, and componen s with respect to the public health and safety.

In addition, 10 CFR Part 100 provides dose guidelines for r ac or siting against-~hich calculated ac ident dose consequences may be compar d.

III.

RELATEO SAF:tY TOPICS Topic II-Z.C, "A&ospheric Transport and Oiffusion Characteristics for Accident Analysis" provides

.he meteorological data used for'alculating the offsite dose consequences.

The review of the fuel damaging accidents did not consider fuel damage as a

J result of drops of the spent fuel cask or other heavy objects which can be carried either over an open reactor vessel or the spent fuel pool.

Review of the drops of casks and heavy objects is covered in two SEP Topics, IX-2, "Overhead Handling Systems-Cranes" and XY-21, "Spent Fuel Cask Orop pccideats."

IV.

REVIEW GUIOELINES Accidents resulting from the movement of fuel inside and outside containment were reviewed following the assumptions and procedures outlined in Standard Review Plant (SRP} Section 15.7.4 and Regulatory Guide I;25.

The dose to an individual from a postulated fuel handling accident should be "well within" the exposure guidelines of 10 CFR Part 100.

(Whole body doses are also examined but are not controlling due to the decay of the short-lived radio-isotopes prior to fuel handling.)

This is based on the probability of this event relative to other events which are evaluated against 10 CFR Part 100 exposure guidelines.

The review considers single failure, seismic design and equipment qualification only when the potential consequences might exceed the guidelines of 10 CFR Part 100 in the absence of containment isolation and effluent filtration.

The system design is considered to be acceptable ii the limiting doses are well wi.hin the 10 CFR 100 guidelines.

Y.

EVALUATION The assumptions used in this evaluation are suamarizA:in Table XY-20-1.

Two cases of the fuel handling accident ~ere considered.

3 The plant's Technical Specifications related to fuel handling in the auxiliary j

building provide for the required filtration of radioiodines.

That is, non-ESF charcoal filters are required to be operable when irradiated fuel is handled in the building.

The surveillance requirements are sufficient to provide reasonable

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I assurance that the efficiency will be as high as the 905 assumed.in't)e staff's calculations.

Assuming that filters with an efficiency of 90K for elemental iodine were used and that the fuel was damaged 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown, a

limiting thyroid dose at the exclusion area boundary of less than 34 Rem was calculated.

For "refueling," inside the containment the plant's Technical Specifications require that personnel and equipment doors be closed and radiation levels be continuously monitored.

No -filters are required to be operable; the HEPA and charcoal filters in the purge exhaust are "optional" and are not subject to surveillance to confirm their efficiency.

The staff, therefore, calculated the offsite dose consequences assuming that 1005 of the activity teleased from the fuel pool is released to the atmosphere.

At the same handling time of 100

hours, the calculated dose For release of the activity unfiltered would be 96 Rem at the EAB.'n both cases, whole body doses were also considered, but are not limiting due to the decay of the short-lived radioisotopes.

Low population zone doses arp lower due to the lower atmospheric dispersion factor.

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CONCLUS IONS The limiting doses for the fuel damaging accidents indicate that the plant is adequately designed to mitigate the consequences of this type of accident.

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t TABLE XY-20-1 ASSUMPTIONS MADE IN ANALYSIS OF THE FUEL DAMAGING ACCIDENTS INSIDE ANO OUTSIDE CONTAINMENT 1.

Reactor Power 1551 &thecal 2.

Clad failure of all rods in one of 120 modules.

3.

Release of gap inventory of all failed rods:

10K I 10~ Noble Gas 30v 85gr 4.

Peaking Factor 1.66 5.

Meteorological gonditions corresponding to a ground level release of 4.8xl0-4 sec /m't a distance of 450 m.

(See Topic II-2.c).