ML19308D822
| ML19308D822 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/15/1977 |
| From: | Rodgers J FLORIDA POWER CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| TAC-08882, TAC-8882, NUDOCS 8003180767 | |
| Download: ML19308D822 (10) | |
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50 3c2 NP.C DISTRIBUTION FoR PART 50 DOCKET MATERI AL h
To: Mr. J. StolZ FROM: Florida Power Corp.
DATE OF DOCUMENT
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St. Petersburg, Fla. 33733 3-15-77 J.T. Rodgers DATE RECEIVED n
3-17-77 CLETTER ENOTORIZED P R OP INPUT FORM NUMBER OF COPIES RECElvED
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DE[CRIPTION Ltr notarized 3-15-77....trans the ENCLOSURE Analysis f'or fuel handling accident following:
(2P) in the Reactor Bldg of Crystal River Unit 3 Plant....
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?s PLANT NAME:
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SUBJECT:
Florida Power Corporation Crystal River Unit 3 Docket No. 50-302
Dear Mr. Stolz:
Enclosed are three (3) originals and forty (40) copies of an analysis for a fuel handling accident in the Reactor Building of Crystal River Unit #3 pursuant to your letter of January 17, 1977. This analysis follows the same format as the one performed for a fuel handling accident in the Auxiliary Building of Crystal River Unit #3 which is presented in FSAR Section 14.2.2.3.
1 As can be seen from the results of both the conservative and realistic analyses, the potential site boundary radiation exposures due to a fuel handling accident inside containment will be relatively low and well i
within 10 CFR 100 guidelines even assuming no containment isolation or effluent filtration.
If you need any further information or have any questions concerning the attached analysis, please do not hesitate to contact us.
Very truly yours, i
l J.T. Rodge Assistant Vice President Attachment JTR: hic 1/10b
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General Office 3201 Thirty fourin street soutn. P O Box 14042. St Petersburg Fiorca 33733 813 - 866 5151 l
IN WITNESS WHEREOF, the applicant has caused its name to be hereunto signed by J.T..Rodgers, Assistant Vice President, and its corporate seal to be hereunto affixed by Betty M. Clayton, Assistant Secretary, thereunto duly authorized the 15th day of March, 1977.
FLORIDA POWER CORPORATION By h
J.TVRodgers (/
Assistant Vice President ATTEST Betty M. Clayton Assistant Secretary (CORPORATESEAL)
Sworn to and subscribed before me this 15th day of March,1977.
z': 1 Notary Public My_ Consnission Expires:
Notary Public State of Florida at Large My Coninission Expires July 9,1978 i
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FUEL HANDLING ACCIDENT IN THE REACTOR BUILDING Identification of Cause 4
Before initiating spent fuel movement, the boron concentrations of the reactor coolant and the fuel transfer canal water are increased so that, with all control rods removed, the keff of a core does not exceed 0.99.
Every spent fuel assembly is handled entirely underwater. Under these conditions, a criticality accident during refueling is not considered credible. Mechanical damage to a fuel assembly during transfer operations is possible. A mechanical-damage type of accident is considered the maximum potential source of activity release during refueling operations regardless of location.
Reactor Protection Criterion The criterion for reactor protection for this accident is that the system shall be designed to hold doses within 10 CFR 100 limits at the site boundary.
Methods of Analysis The assumptions made for this analysis are shown in Table 1.
The reactor is assumed to have been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is the minimum time for sufficient radioactive decay of short-lived fission products.
It is further assumed that the cladding of the entire outer row of fuel rods in the assembly, 56 of 208, suffers mechanical damage releasing the gap activity. The rupture of all 208 fuel pins is not considered a credible accident. The fuel rod gap activity is calculated using the escape rate coefficients and calculational methods discussed in FSAR Section 11.1.1.
1 Results of A Realistic Analysis The gases released from the fuel assembly pass upward through the refueling canal water before reaching the atmosphere of the reactor building. As a minimum, the gases pass through at least 10 feet of
- water. Although there is experimental evidence that a portion of the noble gases will remain in the water, no retention of noble gases is Inexperimentsinyp< chair-steammixtureswerebubbled assumed.
through a water pond, Diffey demonstrated decontamination factors of about 1,90Q for iodine.
Similar resu) 1 Barthouxt21andpredictedbyEggletontgsforiodineweredemonstratedby Based on these references, 99.9 per cent of the iodine released from the fuel assembly can be assumed to remain in the water.
For this analysis, however, only 99 per cent of the iodine released from the fuel assembly is assumed to remain l
in the water. The iodine and noble gas activities released to the l
atmosphere of the reactor building are given in Table 2.
The reactor building (would then be ventilated, and discharged through charcoal filters conservatively assumed to be only 90 percent efficient) to the 1
j unit vent.
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The reactor building including all penetrations and associated isolation valves, equipment hatch and airlocks, concrete shell, liner, and interior structures are designed for Seismic Class I and to withstand a Loss of Coolant Accident. The reactor building purge ventilation system is designed for Seismic Class II. The functions of the purge system are assured:
(1) by redundancy of equipment when required;
(-2) by instrumentation and control as required for operation and monitoring during normal operation; (3) by instrumentation and monitoring'for surveillance and control during post-accident period; and (4) by equipment arrangement for optimum maintenance and service availability.
The activity is assumed to be released as a puff from the unit vent.
Atmospheric dilution is calculated using a 2-hour dispersion factor more conservative than that developed in FSAR Section 2.3.
Table 1 gives the total integrated dose at the exclusion distance for the whole body and the thyroid gland.
Conservative Licensing Bases Evaluation Method The methods, assumptions, and conditions which are used in evaluating the consequences of the fuel handling accident are in accordance with those guidelines published in Regulatory Guide 1.25 (Rev. O, 3/23/72).
a.
Fission product Release From Fuel The fission product activity released from the fuel damaged as a result of a fuel handling accident is calculated using methods and assumptions outlined below:
1.
It is assumed that the accident occurs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown which is the earliest time fuel handling operations may begin.
Credit is taken for radioactive decay of the fission product inventory for. this 72-hour interval between shutdown and comencement of fuel brindling operations.
2.
Based on a conservative analysis, it is assumed that 56 fuel rods fail as the result of a fuel handling accident.
3.
The individual fission pmduct inventories in the core are calculated based on methods described in TID-14844, assuming full power operation at the end of core life imediately pre-ceding shutdown, and also assuming that the 56 rods experiencing failure have been operating at a peak to average power of 1.65.
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4.
The maximum fuel rod pressurization is 1200 psig.
5.
All of the gap activity in the damaged rods is assumed released
.to the water and consists of 10 percent of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 per cent of the total radioactive iodine in the rods at the time of the accident.
b.
Fission Product Activity Airborne in the Reactor Building
,The following assumptions and initial conditions are used in calculating the fission product activity released to the reactor building:
1.
The iodine gap inventory is composed of inorganic species 99.75 percent) and organic species (0.25 percent).
e 2.
The minimum water depth between the top of the damaged fuel rods and the canal water surface is 10 ft.
3.
The canal decontamination factors for the inorganic and organic species of iodine are 133 and 1, respectively, giving an overall effective decontamination factor of 100 (i.e., 99 percent of the total iodine released from the damaged rods is retained by the canal water). This difference in decontamination factors for inorganic and organic iodine species results in the iodine above the canal being composed of 75 percent inorganic and 25 percent organic species.
4.
The retention of noble gases in the canal is negligible (i.e.,
decontamination factor of 1).
5.
The effects of plateout and fallout are neglected.
Based on these assumptions, the activity released to the reactor building is listed in Table 3.
c.
Fission Product Release to Environs The following assumptions and initial conditions are used in calculating
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the fission products released to the environment:
1.
The radioactive material that escapes from the canal to the reactor building is released from the building over a two hour period.
l 2.
All radioactive materials released to the environs are unfiltered.
Based on these assumptions, the fission product activity released to the environment as the result of the accident is shown in Table 3.
d.
Radiological Effects - Off-Site The integrated thyroid doses and the integrated whole body doses at the i
exclusion boundary, site boundary and low population zone for this event t.ere calculated using methods and assumptions listed below:
1.
Methods used for calculating integrated thyroid and whole body doses are the same as presented in Regulatory Guide 1.25 (Rev.0, 3/23/72).
2.
A discussinn of the site meteorology is given in FSAR Section 2.3 and a listing of the X/Q values for the accident is given in FSAR Table 2-16. The values selected in calculating the doses presented here were contained in a previous, more conservative version of FSAR Table 2-16. They were taken from the zero to two hour data for the exclusion boundary and site boundary and from the eight hour data for the low populat; ion zone. The values are as follows:
Exclusion Boundary:
1.55 x 10-4 Site Boundary:
6.78 x 10-5 Low Population Zone:
5.88 x 10-6 3.
For the duration of the accident, thq brgathing rate of persons off-site is assumed to be 3.47 x 10
- m3/sec.
4.
The iodine dose conversion factors are given in ICRP Publication 2, Report of Comittee II, " Permissible Dose for Internal Radiation",
1959.
5.
The appropriate average beta and gama energies emitted per disintegration, as given in the Table of Isotopes; Sixth Edition, by C.M. Lederer, J.M. Hollander, I. Perlman; University of California, Berkley, Lawrence Radiation Laboratory are used.
6.
No correction is made for depletion of the effluent plume of radiciodine due to deposition on the ground, or for the radiological decay of radiciodine in transit.
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7.
The receptor is located at a point on or beyond the site boundary where the maximum ground level concentration is expected to occur.
Based on these assumptions, the integrated whole body doses and integrated thyroid dose are summarized in Table 4.
These doses are well within the regulatory limits of 25 rem whole body and 300 rem thyroid dose as described in 10 CFR 100.
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I References j
(1) Diffey, H. R., et al'., " Iodine Cleanup in a Steam Suppression System," InternHioiial Symposium on Fission Product Release and Transport Under Accident Conditions, Oak Ridge, Tennessee, CONF-65047, Vol 2, pp 776 - 804 (1965).
(2) Barthoux, A. J., et al., " Diffusion of Active Iodine Through Water 2 ubbles at High Temperatures,"
with the Iodine B Hng Liberated in CO B
~ AEC-TR-6149, June 1962.
(3) Eggleton, A.*E. J., "A Theoretical Examination of Iodine-Water Partition Coefficients," AERE-R-4887. February.1967.
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Table 1 Fuel Handling Accident Parameters and Results Period of Continuous Operation, days 930 Power Level for the Assembly During Operation, MWt 24.6 3
1.78 x 10-4 Atmospheric Dilution Factor, s/m Total Integrated Dose at Exclusion Distance Thyroid, Rem '
O.924 Whole Body, Rem 0.575 Table 2 Radioactive Release for the Fuel Handling Accident Activity Released to Isotope Reactor Building, Ci Kr-85 1.78 x 103 Xe-131m 1.79 x 1022 Xe-133m 1.02 x 10 Xe-133 1.49 x 104 I-131 27.4 D
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Table 3 Radioactive Release for the Fuel Handling Accident Conservative Licensing Bases Evaluation Method Activity Released to Activity Released Reactor Building to Environs Isotope (C1)
(Ci) 2 1-131 1.25 x 10 1.25 x 102 I-133 3.09 x 101 3.09 x 10I I-135 1.92 x 10 I 1.92 x 10-l Kr-85m 6.56 x 10-2 6.56 x 102 Kr-85 4.06 x 102 4.06 x 102 Kr-88 2.80 x 10 4 2.80 x 10-4 Xe-131m 1.16 x 102 1.16 x 102.
Xe-133m 3.09 x 102 3.09 x 102 4
Xe-133 2.20 x 10 2.20 x 104 Xe-135 2.77 x 101 2.77 x 101 Table 4 Summary of Resultant Doses for the Fuel Handling Accident Conservative Licensing Bases Evaluation Method Total Integrated Dose at Exclusion Boundary Thyroid, Rem 10.6 Whole Body, Rem
.16 Total Integrated Dose at Site Boundary Thyroid, Rem 4.64 Whole Body, Rem
.07 i
Total Integrated Dose at Low Population Zone l
Thyroid, Rem
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Whole Body, Rem
.006 1
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