ML20137U931
| ML20137U931 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/06/1986 |
| From: | Standerfer F, Standerfer F GENERAL PUBLIC UTILITIES CORP. |
| To: | Travers W Office of Nuclear Reactor Regulation |
| References | |
| 0395A, 395A, 4410-86-1-29, 4410-86-L-0029, 4410-86-L-29, NUDOCS 8602190376 | |
| Download: ML20137U931 (8) | |
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GPU Nuclear Corporation Nuclear
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Middletown, Pennsylvania 17057-0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Number:
(717) 948-8461 4410-86-L-0029 Document ID 0395A February 6,1986 TMI-2 Cleanup Project Directorate Attn:
Dr. W. D. Travers Director US Nuclear Regulatory Commission c/o Three Mile Island Nuclear Station Middletown,-PA 17057
Dear Dr. Travers:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 TMI-2 Temporary Reactor Vessel Water Filtration System Attached for review is a Safety Evaluation Report (SER) for use of a Temporary Reactor Vessel Water Filtration System. This system is being proposed as a temporary method for cleaning the Retictor Vessel until a method of reducing the micro-organisms in the RCS can be developed. Use of the DWCS under current conditions results in the rapid plugging and disgarding of defueling filter canisters long before peak capacity is reached. The attached SER shows that the temporary filter system can be operated safely and does not create undue risk to the health and safety of the public.
Per the requirements of 10 CFR 170, an application fee of $150.00 is enclosed.
Sincerely, tander Vice President Director, TMI-2 8602190376 k
FRS/RBS/eml
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Enclosure:
GPU Nuclear Corp. Check No. 00020224 l
GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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4340-3220-86-0023 TMI-2 TEWORARY REACTOR VESSEL FILTRATION SYSTEM 1.
Purpose and Scope The purpose of this Safety Analysis is to demonstrate that the operation of the Temporary Reactor Vessel Filtration System will not present an undue risk to the health and safety of the public.
The prime purpose of the Temporary Reactor Vessel Filtration System is to restore and maintain the visibility in the Reactor Vessel to acceptable levels to insure the continuation of the Early Defueling Program. Recent developments relative to the operation of the DWCS and the Filter Canisters have revealed that the Filter Canisters develop the maximum design pressure drop before the filter has processed significant quantitics of water. Our investigation of this development has lead to the discovery of micro-organism growth in the reactor coolant. Theory and experience indicates that these micro-organisms are capable of plugging the filters in the filter canister prior to the collection of any significant quantity of core debris. These developments have provided the need~to cesign and operate a temporary filter system while GPU Nuclear develops a permanent program to control this phenomenon.
The principle safety questions relating to the system's operation are criticality control, waste disposal, and the potential consequences of spills.
===2.
System Description===
The Temporary Reactor Vessel Filtration System (TRVFS) is provided to cleanup the Reactor Vessel water above the rubble bed to provide and maintain an adequate level of visibility to enable defueling to proceed.
The system consists of a 1 1/2 hp pump,1.1/2 inch diameter hoses, isolation valves, fittings and.a filter assembly. The TRVFS will be operated only when operation personnel are on the defueling platform.
The TRVFS will take suction from the IIF and return the filtered water to the IIF. Flow will be provided by the pting at maximum flow rate of 75 gallons per minute. The water will pass through a diatomaccous earth (D.E.) coated filter. The filter is housed in a container with filtration'provided by approximately 100 15 inch long by 1/2 inch diameter filter bags on '.<hich the diatomaceous earth precoat is fixed.
The diatomaceous earth is injected into the suction of the pump in approximately seven (7) pound batches which then coats the filter medium. When the pressure drop across the filter reaches 8-10 psi above precoat pressure drop, the pump will no longer be capable of providing significant flow; consequently, the filter will be "backbumped" or
. cleaned. Backbumping is accomplished by stopping the flow and flexing the filter media which causes the filtered material and the diatomaceous earth to fall to the bottom of the filter housing. The bottom of the housing can be drained to a shielded container such as a 55 gallon drum as required. The TRVFS is restored by establishing flow and injecting an additional seven (7) pounds of clean diatomaceous earth into the pump suction. To assure that filtered material in the filter is not pumped --
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4340-3220-86-0G23 back to the reactor vessel before the recharged D.E. is fixed to the filter, a bypass line is provided to allow recirculation until the effluent from the filter appears clear. Each cleaning of the filter will add approximately 7 to 20 pounds of diatomaceous earth, filtered material, and 20-25 gallons of water. After settling has occurred in the
' shielded container, the water will be decanted and stored for further processing at a later date.
3.
Criticality Prevention Any. fluid system connected to the vessel which transports coolant system water from the IIF has the potential to nove fuel bearing material.
Consequently, the potential to accumulate fuel external of the reactor vessel has been addressed. Because of the temporary nature of the system and the unlikelihood of accumulating significant quantities of fuel based on the suction point for this system, GPU Nuclear believes that the rigorous engineering / design control established for defueling canisters, which were designed for extended life, is not required in this instance.
However, the TRVFS design and operation does provide the following separate assurances to preclude any potential significant fuel accumulation and criticality.
a.
Only suspended material in the RCS will be moved.
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b.
The filter body will be continuously monitored for gamma radiation levels to detect any significant fuel accumulation, c.
The gross alpha radioactivity in the 55 gallon drum will be determined by sampling.
The inlet and outlet hoses for the TRVFS are one and one-half (1 1/2) inch I.D. hoses. The hoses from the IIF to the pump and from filter to IIF will be secured in such a manner that the suction and discharge piping will always be immersed in the IIF no more than two (2) feet below 327'-6".
Consequently, the suction of the RCS water will always occur within the confines of the IIF.
't 75 gpm, the velocity in the hose is approximately 14 feet per second; however, the fluid velocity _ ten feet from the hose would be four to five orders of magnitude less, if any velocity effect from the hose suction exists at all..At such velocities only particles smaller than 10 microns could be moved. Conservatively, it could be assumed that the suspended material in the reactor water is uranium oxide at a concentration of 1 ppm; significantly greater than sample concentration have indicated in the past. Twelve hours of continuous operation of the filtet system at this concentration would.
deposit 0.2 Kg of UO2 on the filter media, significantly less than that required to produce a criticality. The conservative nature of this
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hypothetical model is-illustrated by comparison to the analysis of the pre and post-filter effluent in the DWCS operation; a similar but deeper suction and discharge arrangement. Analysis of DWCS fluid has shown no detectable fissile material in the flow stream. Therefore, it is concluded that a significant accumulation of uranium oxide will not occur in the filter housing. t
s 4340-3220-86-0023 A gamma sensitive radiation detector is provided in close proximity to the filter housing. Calculations using upper limit radionuclide removal rates based on scaling empirical data from DWCS and RCS samples indicate that after filtering RCS water for four (4) hours the highest expected radiation level at the exterior surface of the vessel would be approximately 3 R/hr. If, however, the vessel contained 70 Kg of UO,
2 with the attendant fission product activity as predicted by the ORIGEN code, the radiation level at the container surface would be 15,000 R/hr.
This significant difference provides an excellent check relative to any possible uranium accumulation in the filter during operation. A gamma radiation monitor will centinuously monitor the surface radiation level and at a preselected radiation level filtering operations will be secured. Currently this level is expected to be approximately 3 R/hr or less. This level may be adjusted by Rad Con to account for changing conditions. At this radiation level, the amount of UO2 in the filters is expected to be less than gram quantities, far less than necessary to create criticality concerns.
As an additional check to assure that little or no fissile material and/or fuel-bearing material is being removed from the RCS, grab samples will be taken of the diatomaceous earth and filtered material in the 55 gallon drum. A test for gross alpha radioactivity will be used to calculate the quantity of fissile material in the drum. The drum may be used to accumulate as many as ten (10) cleanings of the filter media.
Since little or no fuel is expected in any filter loading, it is obvious that insufficient fuel could. accumulate in the drum to create criticality concerns. For enough fuel to accumulate in the drum to create a criticality problem, radiation level would be on the order of 2500 R/hr on the side of the drum.
The potential for criticality due to a baron dilution event has been considered. Diatomaceous earth consists of approximately 88% silica and exhibits no propensity to remove or adsorb baron. Therefore, significant boron dilution caused by removal of baron by the diatomaceous filters is not considered credible. Operating experience with these filters in the fuel pool has resulted in no detectable dilution of the fuel pool. Baron dilution of the reactor vessel is not judged credible because of the closed loop nature of the system, and the unavailability of unborated water sources.
4.
. Waste Disposal Calculations of the estimated radionuclide concentrations and the maximum expected concentrations of fissile material indicate that the raw waste is slightly greater than those required for disposal as a Class C package. The group 2 long-lived isotopes are controlling for Class C.
Therefore, stabilization will be required for shallow land burial.
Cement solidification of the waste will reduce the concentrations to those acceptable for shipment as a Class C waste.
Consequently, it is concluded that shipment of these wastes will not represent an abnormal waste disposal concern for this program. -
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I 4340-3220-86-0023 5.
Dose Rate Evaluation and Spill Consequences As stated ~previously, the expected maximum radiation level at the external surface of the filter housing is 3 R/hr. Shadow shielding will be installed so that the contribution that this localized radiation source will make to the exposure of operating personnel on the Defueling Platform will be minimal. Radiation levels from the shielded 55 gallon drum are expected to be as much as 340 mR on contact. Shadow shielding will be used to reduce the dose rate from this source to personnel working at the defueling slot to one (1) to two (2) mR/hr. It may be possible during the transfer of D.E. from filter to 55 gallon drum to experience a spill wherein the 20 to 25 gallons of water and the approximately seven (7) pounds of D.E. along with the filtrate are spil h d onta the surface of the North End Defueling Platform. Should such a spill occur, a portion of the platform would be contaminated with up to seven (7) curies of Strontium / Yttrium-90 and 0.07 curies of Cesium-137. If the spill spreads to cover a depth of 1/8" (3 mm), an area of about 200 ft2 will be contaminated. Dose rates attributable to this contamination will be in the range of 25 rad per hour at 10 cm at' ave the floor.
Although a specific calculation for offsite release created by a drum handling accident has not.been performed, the results are easily bounded by and are far less than the releases postulated for a canister drop accident as described in Revision 4 to the Early Defueling SER.
A liquid only spill must also be considered. A pipe break at the pump discharge would exhibit the potential of spilling liquid from the IIF onto the 322'-6" elevation of the FTC floor. This event can be detected using the IIF level monitoring system. This liquid would in turn drain to the sump of the canal floor on the south-east cornst of the upper canal where it would collect and be pumped to a staging or processing location. With the suction limited to two (2) feet below the surface of the water in the IIF, this represents approximately 4000 gallons of RCS water. It is not expected that such an event would significantly increase the radiation exposure to workers on the Defueling Platform.
Another potential concern relates to the consequences of'a filter break through wherein the seven'(7)' pounds of diatomaceous earth is pumped into the reactor vessel. As previously stated, diatomaceous earth is chiefly silica with little, if any, hydrogeneous material, in a fine powder form. It would, therefore, significantly increase the turbidity of the
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RCS fluid but not effect the present shutdown margin of the bulk core.
Eventually, it would be expected to settle to the top of the rubble bed and in turn be removed with the rubble during defueling.
6.
Summary Therefore, it is concluded that based on the evaluations presented in this Safety Analysis the operation cf the Temporary Reactor Vessel Filtration System may be conducted without undue risk and exposure to the operating personnel nor will it pres 3nt any undue risk to the health and safety of the public. 1
4340-3220-86-0023 10 CFR 50, Paragraph 50.59, permits the holder of an operating license to
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make changes to the facility or perform a test or experiment, provided the' change, text, or experiment is determined not to be an unreviewed safety question and does not involve a modification to the plant Technical Specifications.
A proposed change involves an unreviewed safety question if:
a.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety analysis may be increased; or b.
The possibility for an accident or malfunction of a different type 4
than any evaluated previously in a safety analysis report may b_e created; or c.
The margin of' safety, as defined in the basis for any Technical Specifications, is reduced.
A variety of events have been analyzed in this SER. It has been determined that the event due to operation of the TRVFS are similar to events described in several previous submittals (References 1, 2 and 3).
Of primary concern are a handling accident, deboration of the RV, 1
draindown of the RV and criticality concerns in the filter. The drop of the 55 gallon container and subsequent release to the environment is bounded by the canister drop accident described in Reference 1.
Deboration of the RV is possible by one of the two methods: absorption of baron by the 0.E. or by dilution caused by improper hookup and operation of the filter system. The filter media has shown no propensity to remove boron during test operations in the fuel pool. Operation of similar types of systems have been addressed in the Baron Hazards Analysis (Reference 2) and are not considered to be a significant
. dilution hazard.
l Due to the setup of hose suction and discharges, draindown of the reactor vessel is not considered credible. Previous evaluations have shown that ambient cooling is adequate providing water remains above the vessel flange. Since draindown will be limited to the upper two (2) feet of the IIF, this event is bounded.
It is recognized that the criticality control measures for the filter canister proposal here does not incorporate the engineered features for criticality control used in the defueling filter canisters (Reference 3)
However, measures have been taken to minimize and detect the buildup of UO2 in the filter media. These effects, coupled with the temporary nature of the proposed system, give confidence that the proposed operation does not increase the probability of an accident or create the possibility of an accident of a different type than previously evaluated.
Technical Specification safety margins at TMI-2 are concerned with criticality controls and prevention of further core damage due to overheating. As demonstrated by this Safety Evaluation Report, Technical i
Specification safety margins will be maintained throughout the filtering process. Subcriticality is ensured by establishing the boron concentration at greater than 4350 ppm during the early defueling process,
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4340-3220-86-0023 and ensuring that this concentration is maintained by monitoring the boron concentration and inventory levels and by isolating potential deboration pathways. Subcriticality in the filter system is maintained by selective pickup locations, monitoring of radiation levels and by sampling of the discharged filter media. The ability to prevent further core damage due to overheating is not affected by the filtering process.
Thus, it is concluded that the operation of the Temporary Reactor Vessel Filtration System does not constitute an unreviewed safety question as defined by 10 CFR Part 50, Paragraph 50.59.
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4340-3220-86-0023 REFERENCES
.1.
Safety Evaluation Report for Early Defueling of the TMI-2 Reactor Vessel, Revision 4, GPU Nuclear letter 4410-85-L-0200, dated October 10, 1985, from F. R. Standerfer to B. J. Snyder.
2.
Boron Hazards Analysis, Revision 2, GPU' Nuclear letter 4410-85-L-0195, dated September 27, 1985, from F. R. Standerfer to B. J. Snyder.
3.
Technical Evaluation Report for Defueling Canisters, Revision 1, GPU Nuclear letter 4410-85-L-0183, dated September 10, 1985, from F. R.
Standerfer to B. J. Snyder.. -
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