ML20012G685
| ML20012G685 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 03/09/1993 |
| From: | Cross J PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20012G686 | List: |
| References | |
| NUDOCS 9303110337 | |
| Download: ML20012G685 (14) | |
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Portland General Electric Company
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b James E. Cross Vice President and Chief Nuclear Officer March 9,1993 Trojan Nuclear Plant Docket 50-344 License NPF-1 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sirs:
I Permanently Defueled Emergency Plan for the Troian Nuclear Plant Portland General Electric Company (PGE) has permanently ceased power operation of the Trojan Nuclear Plant. An application for a " Possession Only License" was submitted to NRC by letter dated January 27,1993. Certification that the reactor core was permanently defueled was made to NRC by letter dated February 2,1993.
Placing the Trojan Nuclear Plant in permanent shutdown with fuel completely removed from the core to the spent fuel pool provides a basis for reducing the scope of emergency response planning from that of an operating nuclear power plant as required by Title 10 of the Code of Federal Regulations, Part 50.54(q). PGE has submitted a request under separate cover to exempt the Plant from certain prescribed standards in 10 CFR 50.47(b) and Appendix E of10 CFR 50.
PGE proposes that the Trojan Nuclear Plant Emergency Response Plan (Topical Report PGE-1008) be reduced in scope commensurate with the activities to be conducted pertaining to maintaining the Trojan Nuclear Plant in a permanently shutdown and defueled state. The proposed Permanently Defueled Emergency Plan is submitted as Attachment I. The Plan is -
based on the proposed exemptions discussed in PGE to NRC letter, " Request for Exemptions from 10 CFR 50. Requirements for Emergency Planning", dated March 9,1993 and hazards analysis results summarized in Attachment II. The referenced exemption letter presents, in tabular form, specific requirements of 10 CFR 50.47(b) and 10 CFR 50, Appendix E from which PGE requests exemption.
1 930311033f 930309 PDR ADOCK 05000344 y[ j F
PDR 9W Salmon Street, Portland, OR 97204-
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i 503/464-8897
m Document Control Desk March 9,1993
' Page 2 Upon NRC approval of the Trojan Permanently Defueled Emergency Plan and associated exemptions from 10 CFR 50.47(b) and Appendix E, the Trojan Nuclear Quality Assurance Program will be correspondingly amended in accordance with the provisions of 10 CFR 50.54(a)(3) These changes will be processed as non-reductions in commitments, as NRC approval for these changes will have been received via NRC approval of the Trojan Permanently Defueled Emergency Plan.
The Permanently Defueled Emergency Plan will become effective after the following i
conditions are satisfied: (1) receipt of the Possession Only License, (2) approval of the l
proposed Permanently Defueled Emergency Plan by NRC, and (3) approval of the proposed l
10 CFR 50.47(b) and Appendix E exemptions by NRC.
Sincerely, ki W. R. Robinson i
for J. E. Cross j
r i
Attachments c:
Mr. John B. Manin Regional Administrator, Region V U. S. Nuclear Regulatory Commission Mr. David Stewart-Smith State of Oregon Depanment of Energy
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Mr. Kenneth Johnston l
NRC Resident Inspector j
Trojan Nuclear Plant l
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Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993
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License NPF-1 Attachment II Page 1 of 12 i
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SUMMARY
OF THE PERMANENTLY DEFUELED SAFETY ANALYSIS l
l I.
BACKGROUND A basis for reducing the scope of the Trojan Nuclear Plant's Radiological Emergency Plan was needed subsequent to placing the Plant in a permanently shutdown and defueled status.
The basis was needed to demonstrate the reduced radiological consequences of the potential Plant accidents that could still occur and to provide the requirements for emergency planning that the Permanently Defueled Emergency Plan must continue to address.
I II. DISCUSSION l
The following is a discussion of the effect on the Final Safety Analysis Report (FSAR) Chapter 15 Accident Analyses of placing the Trojan Nuclear Plant in a permanently shutdown and defueled status. Accidents described in the FSAR are no longer given consideration as such if they no longer have the potential for a i
radiological release of sufficient magnitude to adversely affect public health and safety.
For accidents which may still potentially result in a radiological release that could affect public health and safety, an assessment of the potential consequences is provided herein.
)
The discussion is based on the following Trojan Nuclear Plant conditions being in effect: 1) the reactor vessel is permanently defueled, 2) spent fuel on site is stored in the Spent Fuel Pool,
- 3) Rod Cluster Control Assemblies (RCCAs) are permanently removed from the reactor, and 4) the Spent Fuel Pool meets the design criteria discussed in Topical Report PGE-1037, Spent Fuel Storace Rack Replacement Report, approved by NRC Safety Evaluation Report for Amendement No. 88 to the Facility Operating License, dated June 8, 1984.
The original initial conditions assumed in the FSAR for power operation regarding core power, average Reactor Coolant System (RCS) temperature, and pressurizer pressure are not applicable to the design basis for storing spent fuel.
Therefore, fuel rod departure from nucleate boiling (DNB) is no longer dependent on these parameters.
Power operation no longer occurs, so the Reactor Protection System and Engineered Safety
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Features are no longer needed to effect safe shutdown and cooldown of the reactor.
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Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 2 of 12 III. ACCIDENTS NOT PERTINENT TO THE DEFUELED. CONDITION A2 Accidents'Resultina in an Increase in Heat Removal by the Secondary System There are a number of events involving a potential unplanned increase in heat removal by the secondary system'which could cause either a rapid cooldown of the RCS, precipitating an increase in core reactivity and potential for overcooling'of the reactor vessel, or a loss of heat sink for the RCS and~ subsequent overheating leading to possible fuel damage.or excessive RCS pressure.
Possible events include: feedwater system malfunctions-that result in a decrease'in feedwater temperature, increase in-feedwater flow or loss of feedwater; system malfunctions that' lead to an increase in steam flow; inadvertent opening.of a' steam generator relief or safety valve; steam or feedwater-piping failures (main steam, or feed line and smaller); and various Plant transients and trips described in the FSAR.
As the Plant will no longer operate at power and no fuel.is loaded in the reactor vessel, the secondary system will no longer be used to remove heat from the RCS.
Note also.that in the event that the secondary system was pressurized and operating, a radiological accident could not occur, because.there is no fuel in the RCS.
This class of accidents is no longerJof any consequence for the permanently shutdown and defueled. plant configuration, since the accidents no longer pose a threat to public health and safety.
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Hz Accidents Resultina in a Decrease in RCS Flow Rate A partial or total loss of forced reactor coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault in the power supply to the pump.-
If the reactor is at power at the time of the' accident, the i
effect of a loss of coolant flow would be a rapid increase in coolant temperature which could result in DNB with subsequent fuel damage if the reactor were not-tripped promptly.
However, :
since power operation will no longer occur at Trojan and.the reactor core is completely offloaded, there is no possibility of fuel overheating or failure due to DNB. ~This class.of accidents no longer poses a threat to public health and safety and are no longer of consequence for the permanently shutdown and defueled plant configuration.
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'rojan Nuclear Plant Document Control Desk
- ket 50-344 March 9, 1993 License NPF-1 Attachment II Page 3 of 12.
gt Accidents Resultina from Reactivity and Power Distribution Anomalies A number of faults have been postulated which could result in reactivity and power distribution anomalies.
Reactivity changes could be caused by control rod motion or ejection,. boron concentration changes, or addition of cold water to the RCS.
Pcwer distribution changes could be caused by control rod motion, misalignment, or ejection, or by static means such as. fuel assembly mislocation.
However, these events can no longer occur, because they require fuel loaded and Rod Cluster Control Assemblies installed in the reactor vessel.
As the reactor has-been defueled and the control rod RCCAs' removed, therefis no possibility of a reactivity or power distribution anomaly that can cause local damage to fuel assemblies.
Thus, this class of accidents no longer poses a threat to public health and safety and are no longer of consequence for the permanently shutdown and defueled plant configuration.
Q2 Accidents Resultina in an Increase in Reactor Coolant Inventory (Specifically Inadvertent Operation of-the Emeroency Core Coolina System)
Inadvertent injection by high pressure ECCS or a malfunction of the chemical volume and control system can_cause an unplc.med increase in reactor coolant inventory.
Depending cn1 the boron concentration and temperature of the injected water and the response of the automatic control systems, a power level increase.
arf result and lead to fuel damage or overpressurization of.the RCS.
However, since the reactor vessel is devoid of fuel, fuel damage from an increase in RCS inventory cannot occur. Also, any reactor vessel overpressurization or failure that could occur would not represent a hazard to public health and safety.
- Thus, this class of accidents is no longer of consequence for the permanently shutdown and defueled plant configuration.
Et Accidental Depressurization of the RCS The effect of a pressure decrease in the RCS during operation is to decrease the neutron flux (via moderator density feedback).
The reactor control system functions to maintain' power constant ~
through the transition, but-pressurizer' level continues ~to.
increase and eventually results'in a reactor trip.
If the reactor trip-does not occur soon enough, there is a possibility that DNB may occur and result in fuel damage and leakage of.
radioactive coolant from the RCS through the open vent.'Since there is no fuel-loaded'in the reactor vessel, it cannot be
Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 4 of 12 affected by a decrease in RCS pressure and cannot reach DNB as a result.
Therefore, there exists no threat to public health and safety.
Ez Steam Generator Tube Rupture The steam generator tube rupture accident results in communication of the primary and secondary systems providing a path for leakage of reactor coolant to the environment.
The accident was assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited amount of defective fuel rods.
Six months after shutdown, the reactor has been defueled and the gaseous fission products in the RCS have decayed to reduced levels.
Also, there exists no mechanism for a tube rupture since the RCS has been cooled down and depressurized, because the difference in pressure from the primary side to the secondary across the tubes is drastically reduced.
Therefore, a steam generator tube rupture is no longer a plausible event.
91 Accidents Resultina in a Lois of Reactor Coolant Inventory A design basis Loss-of-Coolant At ident (LOCA) is caused by a breach of the reactor coolant loo; piping that results in interruption of the normal flowpat for cooling the reactor core and flow of coolant out the break.
Without forced cooling, decay heat is sufficient to cause clad o) 'ation and fuel melting.
A LOCA is no longer of consequence n'th the reactor permanently shutdown and defueled, since a breach of_the reactor coolant loop will not cause fuel or clad meltino.
The'RCS has been cooled down and depressurized, so there is not sufficient LOCA mass and energy available in the RCS to pressurize Containment.
IV. ACCIDENTS PERTINENT TO THE DEFUELED PLANT CONDITION At Accidents Resultino in Radioactive Releasa These events involve radioactive release to the environment from systems other than the RCS or secondary system and are applicable to the permanently shutdown and defueled Plant configuration.
Radioactive waste gas decay tanks permit decay-of accumulated radioactive gases prior to their release as a means of reducing the normal release of radioactive materials to the atmosphere.
The radioactive contents are principally the noble gases kryptoa and xenon, the particulate daughters of some of the krypton and I
Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 Licer:se NPF-1 Attachment II Page 5 of 12 xenon isotopes, and trace quantities of halogens.
Since these noble gases are generated from fission during power operation, there will be no generation of fission gases and no more sent to the Waste Gas Decay Tanks from the RCS.
Six months after shutdown of the Plant, the gases in the Waste Gas Decay Tanks will have decayed such that there remains in the system only a very small fraction of the inventory accumulated during power operation.
Therefore, there is no possibility of a radioactive release beyond the exclusion area boundary which could result in doses that would exceed EPA Protective Action Guides.
In an operating plant, a Chemical and Volume Control System (CVCS) holdup tank rupture is the highest potential atmospheric release source in the CVCS due to its large volume and the fact that it contains reactor coolant.
However, six months after shutdown of the Plant, levels will have decayed sufficiently such that the activity levels of any residual effluent stored in the tanks during ongoing decontamination would be much less than those seen during power operation.
Therefore, there would be no possibility of a radioactive release from a rupture of a CVCS holdup tank that would result in doses beyond the exclusion area boundary that could exceed EPA Protective Action Guides.-
Hz Desian Basis Fuel Handlina Accidents Of the potential accidents still pertinent to the Trojan Nuclear Plant in its permanently shutdown and defueled state, the' Fuel Handling Accident is the most limiting with regard to the severity of its consequences.
The spent fuel is the largest potential source of radioactive release due to the concentration of radionuclides inside the fuel rods.
The possibility of a fue1~
handling accident is very remote due to the many administrative controls and physical limitations imposed on fuel handling operations.
Fuel-transfer is conducted in.accordance with prescribed procedures under direct surveillance of a licensed operator or certified fuel handler trained in nuclear safety.
Irradiated fuel is prohibited from being transferred into the reactor vessel or into Containment.
Fuel handling manipulators and hoists are designed so that fuel cannot be raised above a position where the water depth would provide inadequate shielding for protection of personnel from overexposure.
The fuel handling manipulators, cranes,. trollies, bridges and associated equipment above the Spent Fuel Pool are designed to prevent the equipment from generating missiles which might damage the fuel.
The facility is designed for transfer and
Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 6 of 12 handling of only one fuel assembly at a time, and movement of equipment handling the fuel is restricted to low speeds.
i These design safety features, in conjunction with the aforementioned administrative controls, support the conclusion that the probability of a fuel handling accident is very small.
Even if a fuel assembly was dropped or struck, the shock absorbing characteristics of the assembly indicate that not all of the fuel rods in the assembly would break.
Nevertheless, for a fuel handling accident analysis the assumption is made that the-cladding of all fuel rods in one assembly break and release all the gaseous fission products in the voids between the pellets.
However, only that fraction of the fission products which migrates from the fuel matrix to the gap and plenum regions during normal operation was assumed to be available for immediate release to the water following clad damage.
Compared to the quantity immediately released, all subsequent activity releases were considered to be negligible.
The analysis of activity released from the Spent Fuel Pool resulting from a fuel handling accident takes a conservative approach to the evaluation of radiological consequences by assuming that the fuel assembly with the peak fission product inventory is the one damaged.
Divergence from the previous analysis occurs in that the assembly with the peak fission product inventory from the most recent and final core offload was selected.
This assembly is assumed to have had the highest peaking factor in the core and is assigned a conservative peaking j
factor of 1.65 per Regulatory Guide (RG) 1.25.
Note also that J
this assembly will have undergone approximately six months of radioactive decay by the time the license prohibiting power operation of the facility is received.
Thus, the fission product inventory (especially of - the short lived daughter products) is significantly less than that assumed in the previous FSAR Fuel Building Accident Analysis.
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Analysis of the activity released from the Spent Fuel Pool water resulting from the postulated fuel handling accident has been based on the fission product source and release assumptions of RG
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I 1.25.
The analysis assumes a six month decay time from end of power operation, which is consistent with the planned i
implementation of the Permanently Defueled Emergency Plan, j
Conservative values from FSAR Table 15.0-5 for core and gap
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activities based on full power operation for 650 days are used for the analysis.
As stated in FSAR Section 15.7, all of the' gap activity in the damaged rods is released to the Spent Fuel Pool water, and consists of 10 percent of the total noble gases other than Kr-85, 30 percent of Kr-85 and 10 percent of the total radioactive iodine in the rods.
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Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License.NPF-1 Attachment II Page 7 of 12 f
The calculational method used to develop the Fuel Handling Accident Analysis involved the DOSES computer code, which has been accepted for use via PGE procedures.
Code inputs were based on FSAR data and the inputs listed in Table 1.
Radionuclide inventory was based on RG 1.25 gap fractions of core iodine and noble gas inventories from FSAR Table 15.0-5, corrected for the peak assembly.
The accident assessment dilution factors were established values obtained from FSAR Table 15.0-9 for the site boundary, and from the design basis accident model for the Control Room Emergency Ventilation system (CB-1) intake using the methodology of NUREG/CR-5055.
The dose conversion factors used in the accident assessment were obtained from RG 1.109 (Revision 1) for Iodine isotopes 131 through 135.
The dose conversion factor for Iodine-129 was obtained by determining an Iodine-131 equivalent Curie value.
This was done by taking the ratio of the Maximum Permissible Concentration (MPC) for Iodine -129 to Iodine-131 from 10 CFR 20, Appendix B, Table II, Column 1 and multiplying that ratio by the peak assembly Iodine-129 activity to obtain an equivalent Iodine-131 activity value.
The DOSES computer code was run to determine the adult thyroid dose commitment from this amount of Iodine-131 equivalent released.
This method yielded an Iodine-131 equivalent activity equal to 4.5 times the Iodine-129 activity present and was tested by examining the ratio of adult thyroid inhalation dose factors from NUREG-0172 for Iodine-129 to Iodine-131 (resulting Tation was i
3.7) and the same ratio from the recent 10 CFR 20 basis document, EPA Report No. 11 (resulting ration was 5.4).
These methods yielded ratios for Iodine-131 equivalency that were close to 4.5, with 4.5 as a reasonable average of the three ratios.
The Iodine-129 inventory present in the peak assembly at the time of Plant-shutdown was determined to be 16.6 millicuries.
This value was obtained by taking the total core inventory of 1.943 Curies at shutdown based on calculations using the Radioisotope Buildup & Decay (RIBD) Library, dividing by the total number of assemblics in the core to obtain an average assembly value, then multiplying by the radial peaking factor of 1.65.
The amount of Iodine-129 available for release from the assembly was derived by, multiplying this value by the 30 percent gap fraction for long-i lived isotopes from RG 1.25.
Assuming a pool iodine decontamination factor of 100, the result-was an airborne release of 50 microcuries of Iodine-129, or 224 microcuries of Iodine-131 equivalent as obtained uring the above method.
This-resulted in a Control Building CB-1 ventilation intake location thyroid dose 4 millirem.
This resultant dose is a small fraction I
of 5.92 x 10 l
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Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 8 of 12 of the EPA Protective Action Guide dose limits.
Results of the dose model for the Fuel Handling Accident are compared with the EPA Protective Action Guide exposure levels in Table 2.
From these results, it is concluded from the analysis that the beta plus gamma skin dose, gamma whole body dose and inhalation thyroid dose are small fractions of the EPA Protective Action Guidelines at the 662 meter exclusion area boundary and at the Control Building CB-1 ventilation system intake.
An analysis of the activity released from a fuel handling accident inside Containment need no longer be performed as fuel movement inside Containment and from the Spent Fuel Pool to Containment _are prohibited.
Also, analysis of the Spent Fuel Cask drop accident is not required, since the Trojan Nuclear Plant design basis does not include loading of spent fuel into transport casks at this time.
The Facility Operating License prohibits movement of a spent fuel assembly shipping cask into the Fuel Building.
The requirement for performing an evaluation of the consequences of a cask drop accident prior to using a spent fuel shipping cask has been incorporated into Plant procedures.
C Loss of Spent Fuel Pool Coolina 2
The Spent Fuel Pool cooling system is designed to: 1) maintain the water in the Spent Fuel Pool at or less than 140 degrees Fahrenheit with the maximum number of fuel assemblies less one full core discharge, 2) maintain fuel cladding integrity in the event all forced cooling is lost and cooling occurs by boiling at the surface of the Spen ~c Fuel Pool, with evaporative losses being made up by a Seismic Category I supply of makeup water, and 3) maintain sufficient cooling of fuel assemblies in the event-a fuel assembly or other object is dropped and remains lying across the top of one or more assembly locations.
The only requirement to assure adequate cooling for the spent fuel is to maintain the.
water level in the Spent Fuel Pool so that the spent fuel elements are not exposed.
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r Trojan Nuclear Plant Document Contro1 Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 9 of 12 The consequences of a loss of Spent Fuel Pool forced cooling have been evaluated.
A loss of Spent Fuel Pool cooling caused by a total loss of offsite power would result in the most severe consequences for this postulated event, as this allows fewer i
options for recovery than other causal events where makeup water is available.
It is conservatively assumed in the analysis that makeup temperature is 100 degrees Fahrenheit, no makeup is provided prior to boiling and no evaporative losses occur during heatup.
The Spent Fuel Pool bulk temperature prior to the accident is 100* F.
Decay heat from the spent fuel at six months after shutdown is calculated using ANSI /ANS Standard 5.1-1973.
The modeling of the spent fuel decay heat was based on the power history of the Plant.
Individual rod histories were not used.
The power operation for each month was converted to an equivalent operating time at full power.
This is conservative in that it i
places a larger proportions of fission events occurring near the end of each month and allows less time for decay to occur during a month.
Spent Fuel Pool decay heat generation six months after i
shutdown is calculated with the results listed in Table 3.
The analysis results show that if a loss of forced Spent Fuel Pool cooling should occur, there is sufficient time to effect repairs to the cooling system or to establish makeup flow prior to uncovery of the spent fuel.
Cooling of the spent fuel assemblies will be maintained without any threat to cladding integrity, and no potential release of radioactivity will occur.
V.
SUMMARY
i 1
Power operations have permanently ceased at the Trojra Nuclear
)
Plant.
The reactor core has been offloaded to the Spent Fuel Pool, and movement of nuclear fuel into Containment i3 prohibited.
Those accidents which could have occurred during power operation or with fuel in the reactor are no longer applicable at the Trojan Nuclear Plant.
There are three classes of potential accidents which could result in a radiological release: Loss of Spent Fuel Pool cooling or makeup, radioactive j
waste decay tank ruptures and fuel handling accidents.
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The radioactive waste decay tank contents after six months of i
decay will only contain a fraction of the activity present'during Plant operation, and therefore, a rupture of a tank cannot result in a radioactive release with doses at the exclusion area boundary that are any greater than a small fraction of the EPA PAG exposure limits.
Fuel handling accidents have the potential for ralease resulting in a dose over two hours at the exclusion
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Trojan Nuclear Plant Document Control Desk
- >ocket 50-344 March 9, 1993 License NPF-1 Attachment II Page 10 of 12 area boundary that is only a small fraction of limits specified in the EPA Protective Action Guides.
Loss of Spent Fuel Pool
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cooling does not result in any radiological release, since there is sufficient time to establish makeup and maintain Spent Fuel Pool level.
Boiling continues to remove heat from the spent fuel assemblies and to maintain the bulk Spent Fuel Pool temperature well below the melting point of fuel or clad.
Therefore, there exists no potential for radioactive release beyond the exclusion area boundary that could adversely affect the health and safety of the public.
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i Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II Page 11 of 12 l
1 TABLE 1 Input Data for Calculation of Site Boundary Doses of a Fuel i
Handling Accident.
Parameter Value Reference 4
Breathing rate 3.47 x 10 meter /sec FSAR Table 15.0-7 4
3 Dilution Factor 4.26 x 10 sec/ meter FSAR Table 15.0-9 (site boundary)
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d 3
Dilution Factor 5.89 x 10 sec/ meter FSAR Section 15.6 (Control Bldg Vent)
(page 15.6-48)
Release Duration 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> FSAR Section 15.7 Decay Period 6 months Implementation of Defueled E-Plan Fuel Peaking Factor 1.65 RG 1.25 j
Gamma Shielding 1.0 Unity - no Factor shielding Iodine 100 RG 1.25 Decontamination Factor 4
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l Trojan Nuclear Plant Document Control Desk Docket 50-344 March 9, 1993 License NPF-1 Attachment II l
Page 12 of 12 TABLE 2 Resultant Doses from Fuel Handling Accident and Comparison with EPA Protective Action Guides Exposure Site Boundary Control Bldg.
EPA PAG Fraction Dose Intake Dose Dose of PAG l
Skin 45 x 10-3 Rem 54.9 x 10'8 Rem 50 Rem 0.11 %
Whole Body
Thyroid
l TABLE 3 Spent Fuel Pool Performance at Six Months After Plant Shutdown l
Decay Heat Generated 1.70 Megawatts Spent Fuel Pool Heatup Rate 2.71 'F/hr Time to Boiling (100*F initial pool temp.)
41.39 hr Boiloff Rate (with no makeup) 11.91 gal / min Boiloff Rate (with 100*F makeup) 10.68 gal / min:
Time to Reach 10 Foot Level Above Spent Fuel 171.30 hr l
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