ML19351F712
| ML19351F712 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/06/1981 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19351F713 | List: |
| References | |
| NUDOCS 8102190374 | |
| Download: ML19351F712 (12) | |
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O SUBMERGED D9fINERALIZER SYSTEM SAFETY EVALUATION BACKGROUND The present mode of operation of TMI-2 is governed, in part, by the Interim Recovery Technical Specifications, promulgated by the Nuclear Regulatory Co==ission order dated February 11, 1980. These Tech Specs do not relieve the licensee of compliance with the rules and regulations that apply to domestic production and utilization facilities (10CFR 50).
10CFR 50.59(a)(1) states:
"The holder of a license authorizing operation of a production or utilization facility may (1) make changes in the facility as de-scribed in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests or experiments not described in the safety analysis report, withouc prior Co= mission approval, unless the proposed change, test, or experi=ent involves a change in the technical specifications incorporated in the license or an unreviewed safety question."
PURPOSE The purpose of this safety evaluation is to provide a docu=ented basis for the following conclusions:
(1) Operation of the SDS does not require a change to the TMI-2 technical specifications.
(2) Operation of the SDS is not an unreviewed safety question.
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. EVALCATION CRITEFIA The evaluatien criteria to be used for the determination of an unreviewed safety question cre s;ecified in 10CFR 50.59(a)(2) which states:
"A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (1) if :he pr.,babilicy of occurence or the consequences or an accident or malfunction of equipment i=portant to safety previcusly evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specifi-cation is reduced."
The evaluation criterion for the deter =ination of the requirement to enange the technical specifications is based upon the intended operations of SDS and the impac on existing Interim Recovery Technical Specifications.
SAFETY EVALUATION (1)
Evaluation of SDS coeration arainst 10CFR 50.59(a)(1).
(a)
Implementation of FDS does involve a change in the facility as described in the SuR, even though the change is only temporary in natuta to be used specifically for TMI-2 recovery.
(b)
Implementation of SDS does involer a change in the procedures as described in the SAR; the procedure of the piocessing of radioactivity contaminated waste is addressed. However, because the SDS employs a different methodology for radioactive waste processing than is described in the SAR, it is considered that this specific procedure for waste prccessing is not addressed.
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. l (c) The operation of SDS is intended for the processing of high-level radioactive waste waters; it is not intended to be a test or experiment.
Because operation of the SDS is considered to be a change in the facility I
and procedures as described in the SAR, it is necessary to evaluate SDS operations against the criteria of 10CFR 50.59(a)(2).
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(2) Evaluation of SDS operation aeainst the criteria of 10CFR 50.59(a)(?.).
i Addressing each of these criteria in turn is presented below.
I (a) The Probability of Occurrence of an Accident or Malfunction of Ecuiement Incortant to Safety Previous 1v Evalua?ed in the SAR i
may be Increased.
The SDS ficwpath will provide for radionuclide re= oval of the process flow streas. Frem the contain=ent su=p the water will be pu= ped via the prefilter and final filter, to four 15.000 gal. (ea.) tanks, referred to as the tank i
j farm. The tank fern tanks operate as one tank, they are interconnected with 1
valve-less welding piping. The feed pu=p suction well frem which SDS influent water is supplied, is located at the sa:e elevation as tank farm tanks. The l
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.ater level will rise in the well as the tanks are filled. The suction well is equipped with level indication that is alarmed on high level.
Should the water level continue to rise, a backup level device will be actuated to auto-j natically close the fill valve to the tank farm and. preclude overflow of_the suction well with :entainment sump water.
l Contaminated water is transferred from the suction well, via the SDS feed I
l pt=p through welded ytainless steel piping, to the SDS ion exchange vessels I
t i
through quick disconnect couplings. This quick disconnects and ion exchange l
I vessels are contained in a leakaze conrainment box which contains spent fuel l
_ = _ _ _ _
j l pool vater. Any leakage from the quick disconnects which occurs during routine operation of SDS or when connecting or disconnecting ion exchange vessels will f
be contained within the containment boxes, diluted by pool water, and treated by the leakage ion-exchange system cctor to return to the spent fuel pool.
j SDS processing is performed by flowing water Lbrough three stainless-steel zeolite containers in series for Cesium and Strontium remeval, one additional l
stainless-steel ion-exchange vessel specifically loaded with resin mat-erials for Strontium removal, and into the EPICOR II systes for final SDS ef fluent polishing for removal of remaining trace radionuclides, such as Antimony, and recalcitrant species of Cesium and Strontium. EPICOR-II operation has been authorized by order of the Commission dated October 16, 1979. Processed water will be stored in the Processed Water Storage Tanks on Three Mile Island. No liquid effluents resulting from SDS operation are planned to be released to the environ-ment at this time.
Operation of SDS will be performed under strict administrative procedural control. Operator training is on-going with operator walk-through of the operating procedures to be performed during pre-operational testing. These walk-throughs will provide the opportunity for " hands-on" experience by operations personnel to gain system familiarity as well as to actually test the operating procedures-to be used prior to actual processing of contaminated water. Further= ore, the procedures to be used for operation of the SDS will be submitted to the Nuclear Regulatory Commission for review and approval prior to use in accordance with Technical Specification 6.8.1.
(b) The Consequences of an Accident or Malfunction of Equipment.
The Technical Evaluation Report (TER) submitted to the NRC on April 10, 1980 (TLL 160), concerning the SDS contains (in chapter 7) several hypothetical accidents. The accidents presented, though h'ighly unlikely and improbable,
- present bounding conditions for accident scenarios. At the time of gsneration of the afore-mentioned document, the source terms used were representative i
of contamination levels of the sump water. Because of the interval of time 1
that has passei since development of the TER accident analysis, source terms are approximately one-third the value reported in the TER due to radionuclide i
f decay. Therefore, because of the lower source terms, the TER conclusions remain valid. Detailed information is provided below.
Inadvertant oumaine of containment sumo water into the scent fuel pool.
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The scenario for this hypothetical accident remains the same.
Occupational Exposure Effects:
1 Because of the reduced source term, the calculated maximum exposure rate at i
six feet above the pool surface is reduced to approximately 115 mr/hr.
Con-i clusions regarding the occupational exposure effects of this hypothetical accident remain the same as the TER conclusions except for the reduction of-1 the dose rate.
Off-site Effects:
Radiological effects of this hypothetical accident are assumed to result from two contributing factors. They are:
o Direct radiation exposure.
o Airborne contamination.
Direct radiation exposure at the site boundary is calculated to be 4.5 x 10-7 mr/hr. This calculation is based on the following assu=ptions:
o The isotopes of concern are Cs-134 and Cs-137.
o The fuel pool can be considered as a point source for site boundary direct dose calculations.
o No source self-absorption occurs.
o The fuel pool wall and the fuel handling building wall provide 3' of concrete shielding.
. The pool leakage cleanup ion-exchanger system will remove activity from t5e o
I spent fuel pool. This system will process the pool water at the rate of approximately 100 gp=.
Airborne contamination may be generated as a result of direct evaporation from the pool surface to the Fuel Handling Building atmosphere. The path to the unrestricted environ =ent requires that the ai-borne radienuclides pass through i
the plant HEPA filters prior to discharge via the plant vent. Analysis of this I
hypothetical occurrence is based upon the following assu=ptions:
o Activity spilled into the pool is uniformly distributed.
The pool leakage cleanup ion-exhanger system vill remove activity o
fro = the spent fuel pool. This syste= will process the pool water at the rate of 100 gp=.
o The isotopic inventory of the spent fuel pool is conservatively assu=ed to re=ain constant for a period of one week.
o The spent fuel pool volume is 233,00 gallons.
i o The evaporation entrainment factor is conservatively estimated to be 10-6, Plant ventilation syste= EIPA DF is 102, o
o Air flow across the surface of the spent fuel pool is 5500 ef=.
Based on the above-specified assu=ptions, airborne conta=ination released to the atmosphere as a result of this hypothetical accident is approximately 4Uhi of the Cs-137 isotope, approxi=ately 3.75% of the nor=al operation at=ospheric release of this isotope. This percentage increase is valid for other total bodv dose contributing isotcpes. Normal operation of FDS results in an estimated total body exposure of approximately 3.6 x 10-3 =re=/yr. fro =
all isotopes to the maximally exposed individual. The increase in total body exposure revises the estimated total body exposure to 3.735 x 10-3 =re=/yr.
This increased exposure is 0.0747%-of the allowable dose exposure of 10CTR 50,
- Appendix I of 5 =re=.
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. Pipe rupture on filter inlet line (above water level).
The scenario for this hypothetical accident remains the same.
Occupations Exposure Effects:
Because of the r:duced source term, the significant effects identified in the TER are as follows:
1.
The maximum exposure rate at the surface of the centaminated floor area is estimated to be approximately 3.6 Rem /hr.
2.
The maximum beta exposure rate at a point three feet above the surface of the contaminated floor area is estimated to be 12S Rad /hr.
Conclusions regarding occupational exposure effects of this hypothetical accident are the same as the TER.
The estimated occupational exposure effects are based on the following assump-tions:
Contaminated water sprays into the air from behind the lead shielding.
o Approximately 675 gallons of sump water is released directly into the spent fuel pool and 75 gallons spreads over a surface area of 200 ft2 Primary contributors to the estimated dose rate are Sr-89, Sr-90, o
and Cs-134 Cs-137.
Off-site Effects:
Off-site radiological effects from this hypothetical accident are assumtd to i
i result from two contributing factors. They are:
i o Direct radiation exposure.
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o Airborne contamination These estimated effects are based on the following assumptions:
o The isotopes of concern are Cs-134 and Cs-137.
l o The distance to the closest off-site point is approximately 200 l
l
=eters.
o The spent fuel pool can be considered a point source for exposure l
estimates at a distance of 200 meters.
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o There is no significant source self-absorption; The feel pool wall and the Fuel Handling Building wall provide a o
i direct dose attenuation equivalent to three feet of concrete.
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'e Direct radiation exposure estleates indicate that radiation exposure at the site boundary will increase by approximately 6.75 x 10-7 mrem /hr.
s i
Airborne contamination may be generated as a result of this hypothetical t
accident. The assumptions used to estimate these consequences are the same i
as those used for the airborne contamination estimates of the previous i
hypothetical accident. Based on these assumptions, airborne contamination i
released to the atmosphere as a result of this hypothetical accident is approximately 6.3 pCi/wk of the Cs-137 isotope, approximately 5.63% of the normal operation atmospheric release of this isotope. The percentage increase is valid for other total body dose centributing isotopes. Therefore, the 1
increase in total body exposure resulting from this hypothetical accident l
1s approximately 0.203 mrem /yr. The total body expos'are, including the I
l effects of tais postulated accident, is approximately 3.8 x 10-3 maem/yr, i
j approximately 0.0767. of 10CFR 50 Appendix I limits of 5 mrem.for nor=al 1
operations.
Inadvertant lifting of prefilter above pool surface.
The scenario specified in the TER remains the same. The analysis has been performed based on the folloiwng assumptions:
e o A failure in the crane control system results in the " dragging" l
of the filter over the edge of the spent fuel pool.
i o The prefilter is loaded with 100 Curies of 8-emitters.
l o The minimum distance for exposure calculations is 4.57 meters. The
.1 prefilter can be considered to be a point source.
I o There is no source self-absorption.
o There is no container shielding.
i o There is no environmental release as a result of this hypothetical accident.
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_9 Occupatienal Exposure Effects:
The calculated exposure rate at a distance of 15 feet fro: prefilter in air is 21 R/hr. The effects identified in the TER are valid.
(c) The Possible Creation of a Differert Type of Accicent or Malfunction.
Additional accident postulations are given below.
(1.)
Possible rupture of zeolite ion exchang: vessel in storage and release of conta=inated zeolite resins to the spent fuel
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pool.
In the unlikely event of this i= probable occurrence, environ = ental consequences of no significance will occur. Even though the entire contents of the ion-exchange vessel =ay be released to the spent fuel pool the contaminatel eolite resins will fall to the bottom of the pool.
Radionuclides coatained within the :eolite, pri=arily the Cesiu: isotope, will not be released to the pool water (and hence tc the environ-cent) in significant quantities; they will re=ain adsorbed onto the reolite resins. A significant radio-logical hazard =ay exist for cleanup of the resins f::=
the pool floor. However, because a significant hazard will not be presented as a result of this ocurrence, due to pool shielding, sufficient time exists to develop adequate cleanup procedures and/or cleanuo ecuic=ent.
Furthe rmore, rupture of a :eolite ion-exchange vessel in the spent fuel pool is highly unlikely. Two potential
=echanis=s for vessel rupture have baen identified:
(1) container corrosion, and (2) drop of vessel in the pool.
Vessel rupture, as a result of corrosien effects, is regarded as an occurrence of such low probability to be incredible. Zeolite resins are not knewn to cause a p!
. change in residual water; the ion-exchange vessels art ftbricated from stainless steel. Not only is a corrosion-causing mechanism absent, the vessel material is extremely corrosion resistant. Assuming that a vessel drop in the pool occurs, it is highly unlikely that the vessel will break open and allow its contents to spill on the fuel pool floor.
In the extremely unlikely event that the vessel does break open and allow the contaminated zeolites to spill on the pool floor, significant quantities of radionuclides would not be released to cause danger to the public health and safety as a result of airborne particulate release.
Cleanup of the spilled contaminated resins would be performed under strict administrative control. Cleanup procedures would be reviewed and approved by the Nuclear Regulatory Commission.
Sufficient time would be available for procedure development and approval and personnel mobilization.
(2.) Drop of shipping cask loaded with spent zeolite vessel during transfer from the spent fuel pool to the truck bay.
Present processing plans do not require that transfers of vessels from the spent fuel pool to the truck bay filled with contaminated materials to be performed. At the completion
, of vessel radionuclide loading, that vessel will be removed j
from service anc placed in a storage location in the spent fuel pool.
Should processing plans be changed such that transfers as described above are required, analysis of this postulated accident will be performed.
(d) Reduction in Safety Mar 2in Defined in Bases of Technical Specifications.
The focus of this criteria is on the margin of safety as defined in i
the bases for any tec%nical specificacion.
Since the radwaste system is not addressed in the technical specification bases, this consider-ation is not applicable.
l
. Evaluation of Requirement to Amend the Present Reccverv Technical Specifications.
I=plementation of SDS operations for decontamination of the contaminated water presently in the containment building requires no change to the existing IMI-2 Interim Recovery Technical Specifications. Liquid effluents will not be released to the environment directly from SDS operations; SDS effluent will be placed in the Processed Water Storage Tanks.
Further= ore, gaseous effluents resulting from SDS operations will traverse existing gaseous effluent flow paths. We do not perceive the requirement to change the maximum permissible concentrations or the instrument configuration or setpoints specified in Appendix B of the Interim Recovery Technical Specifi-cations.
Finally, as specified in the Technical Specifications, Article 3.9.14, we will process and discharge the water in the Reactor Building sump and the Reactor not C:olant System unless NRC approval is received.
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i CONCLUSION f
The purpose of documenting this safety evali'ation for the Submerged Deminer-alizer System is to provide the following conclusion: design, construction r
and operation of the SDS does not present an unrevicwed safety question. This conclusion is supported by the below listed facts:
(1.) The SDS does not present the opportunity to increase the probability i
of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
1 (2.) The SDS does not present the opportunity to create the possibility of an accident or =alfunction of a different type than any evaluated previ-4 ously in the safety analysis report.
(3.) The SDS does not present the opportunity for reduction of the margin of t
safety as defined in the basis for any technical specification, 1
Processing water in the containment building will be' performed in compliance I
with the existing TMI-2 Interim Recovery Technical Specifications. No license amend:ent in the form of a change to the Technical Specifications is required.
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