ML20236N127
| ML20236N127 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 07/09/1998 |
| From: | Zinke G Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GAZ-98-041, GAZ-98-41, MN-98-050, MN-98-50, NUDOCS 9807150017 | |
| Download: ML20236N127 (7) | |
Text
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MaineVankee P.O. BOX 408 a WISCASSET, MAINE 04578 + (207) 882-6321 l
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l July 9,1998 MN-98-050 GAZ-98-041 l
UNITED STATES NUCLEAR REGULATORY COMMISSION Attention Document Cor, trol Desk Washington, DC 20555
References:
(a)
License No. DPR-36 (Docket No 50-309).
(b)
Letter: M.B. Sellman to USNRC; Certification of Permanent Cessation of Power Operation and Permanent Removal of Fuel from the Reactor; MN-97-89, dated August 7,1997.
(c)
Letter: M.J. Meisner to USNRC; Defueied Emergency Plan and 10CFR50.54(q)-
Exemption Request, MN-97-119, Dated November 6,1997.
Subject:
Summary of Maine Ybnkee's Radiological Analyses Applicable to the Decommissioned Plant Condition l
Gentlemem l
f in Reference (b), Maine Yankee informed the USNRC that the Board of Directors of Maine Yankee had l
decided to permanently cease operations at the Maine Yankee Plant and that the fue! had been permaner"iy i
removed from the reactor.
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In this permanently shutdown condition, the plant poses a significantly reduced risk to the public health and safety, in view of this reduced risk, Maine Yankee submitted Reference (c) requesting an exemption to l
10CFR50.54(q) that would allow Maine Yankee to discontinue certain aspects of offsite planning activities commensurate with the reduction in risk to the public associated with the permanently shutdown and defueled plant and to reduce the scope of its onsite plan as indicated in Maine Yankee's Defueled Emergency Plan.
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l In response to a NRC verbal request for additional information, synopses of the radiological analyses applicable to the permanently shutdown and defueled condition are attached. Should you have any further l
questions please contact Mr. Robert P. Jordan, Manager, Analysis (207-882-5688).
k Very Truly Yours,
\\
/,d/ i.
I George A.
inke Director, Nuclear Safety & Regulatory Affairs Attachment cc:
Mr. H.J. Miller Mr. M.K. Webb 4
Mr. M. Masnik -
Mr. R. BeSamy Mr. P.J. Dostie 9907150017 990709 I
PDR ADOCK 05000309 1
F PDR
Attachment Letter MN.98 050 MAINE YANKEE Radioloalcal Analysis License Basis Discus.slon I.
INTRODUCTION As Maine Yankee entered the decommissioning phase of the facility life, the licensing baais associated with the safe operation of the nba+ N=1 With the lack of an operating reactor core, placement of all fuelin the spent fuei pool, application of substantial decay time on the spent fuel, and minimization of operational evolutions in handliag radioactive material, the rafety analysis basis which resu ted from an operating plant was found to be inappropriately and excessively conservative for a decommissioning facility.
Therefore, Maine Yankee has revised the licensing basis safety analysis to include those accidents applicable for the plant in a decommissioned state. As part of this revision, the radiological analyses defining the site boundary doses were examined.
During the power operational periods, the limiting radiological transient was identified as the off site releases from a Loss of Coolant Accident while at full power operation. This transient was projected to create a site boundary dose of approximately 160 rem. The second most limiting radiological transient we defined as the failure of a fuel assembly (Fuel -Handling Accident) during transport within the plant. The off site dose consequences of this transient were found to be approximately 17.7 rem.
In reassessing the audiological accidents for the decommissioned state of the plant, it was recognized that the oft site dose threshold would be reduced substantially. For example, the Fuel Handling Accident off site dose consequences were calculated to be reduced from the 17.7 rem level to approximately 6 mrem, a reduction factor of almost 3000. The radiological consequences of a Loss of Coolant Accident are reduced to zero. The consequence of these types of reduction, if applied in the same manner as an operating plant, would be the creation of a nil allowable margin under which simple daily plant evolutions requirod by the decommissioning could not occur without specific review and approval of the NRC. Such level of approvals are inconsistent with normal industry practice and NRC policy and regulations.
in order to avoid this type problem, Maine Yankee utilized several of the lower level licensing basis radiological analyses from the plant operations period, coupled with reanalysis of Emiting transients as appropriate to a decommissioning facility, to create a threshold for off site dose consequences at a level substantia!!y below the EPA Protective Action Guidelines of 1 rem. The analyses, as described in the Maine Yankee Defueled Safety Analysis Report (DSAR), which define this threshold are briefly summarized in the following section.
In addition to the threshold defining safety analyses, Maine Yankee conducted 1. "vity specific analyses to ensure that the radiological site boundary dose thresholds cannot be exceeu d The decommissioning activities which would typically requhe such analyses are identified in section 7.3 of the DSAR. An example of this type of analysis is the Reactor Coolant System Decontamination project radiological assessments, as discussed later.
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LICENSihG BASIS THRESHOLD DEFINING SAFETY ANALYSES FUEL HANDLIN.G ACCIDENT A dropped fuel assembly during normal fuel handling operations in the fuct building was assessed as an l
accident contributing to the licensing basis resul ting in off-site and control room radioact've release effects.
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I Attachment Leuer MN48450 l
i The evaluation of this accident, as applicable to the permanently defueled condition of Maine Yankee, was l
based on the release of the fuel rod gap inventory of the " worst"(highest burnup, highest initial enrichment, longest operating history) fumi assembly presently stored at Maine Yankee. To assure that the most conservative case was considered, all rods in the dropped assembly wero assumed to fail upon impact.
Additionally, all fission products that escaped from the spent fuei rol were assumed available for release to the atmosphere via the fuel building ventilation system. The analysis results show that the projected doses l
from the fuel handing accident are insignificant in comparison to the 10 CFR 100 limits and far less than the j
Environmental Protection Agency Protective Action Guidelines (PAGs). See DSAR section 5.3 for analysis assumptions and Table 5.3.2 for analysis results.
PRIMARY DRAIN TANK RUPTURE For the purpore of establishing an upper threshold on the radioactive activity released from a single component failure in the liquid waste system, a failure of lite primary drain tank was assumed, resulting in a release of its complete inventory. The primary drain tank failure was selected since this tank had the highest i
Technical Specification allowed inventory of dissolved noble gases and halogens durina coerational oeriods.
l The reieam takes place as a liquid spill on the floor of the compartment in the waste processing building where the tank is located. Radioactivity is released to the atmosphere from noble gases and halogens
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evolved from the spilled liquid. Postulated liquid spills escaping concrete structures may be released to the I
site groundwater table. Since the groundwater table at the Maine Yankee site flows towards Beck River and Montsweag Bay, both of which are tidal saltwater estuaries, the potable groundwater and surface water supplies would not be affected by radioactive liquid spills and no dose evaluation for this postulated event is required. Therebre the dose associated with this event resuits from the noble gaset and halogens released to the atmosphere. The doses which have been calculated for radioactive liquid waste system failures are l
below small fractions of the values in the applicable regulation,10 CFR 100,
- Reactor Site Criteria" and the EPA Protective Action Guidelines. See DSAR section 5.6.2 for analysis assumptions and Tables 5.6.1 and 5.6.2 for analysis results.
LOW LEVEL WASTE STORAGE BUILDING (LLVySB) ACCIDENT The temporary storage of low level waste (LLW) on the Maine Yankee site during the current permanen'ly defueled condition is administratively controlled within the existing Low Level Waste Storage Building. This building was designed and constructed to safely store up to a 5 year inventory (volume and activity) of the low level waste generated during periods of normal and refueling operations. All waste stored in the LLWSB is packaged and ready for near term offsite shipment. There are no credible initiating events to cause an accident or fires in the LLWSB. The bounding non-mechanistic accident for the radiologicd impact analysis is defined es the dropping of a highly loaded (in excess of that allowed by regulatory limits) spent resin liner within the building, resulting in the liner failure, spillage of the spent resin, ar.d the release of a fraction of the radioisotopic contents in a cloud. It is assumed that 1% of the cloud activity of the liner non-mechanistically forms an aerosol and that 10% of the aerosol is non-mechanistically released outside the LLWSB. This aerosol acts as a " puff" release in assessing the potential doses at the Exclusion Area Boundary (EAB). The activity of the release is assumed to be dispersed over an arc of 225 cegrees at 700 meters from the LLWSB.
The calculation of the doses at the EAB was performed in accordance with the dose conversion factors from ICRP 30 and the previously reuwed ELISA computer code. The doses calculated for the limiting incident in the LLWSB are below small fracduns of the values in the applicable regulation,10 CFR 100,
- Reactor Site Criteria" and ine EPA Protective Action Guidelines. The calculated TEDE doces for the EAB were determined to be approximately 110 mrem. See DSAR section 5.6.3 for analysis assumptions and resuhs.
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Attachment Letter MN-98 050 111.
OTHER DECOMMISSIONING RADIOLOGICAL ANALYSES Within section 7.3 of the DSAR (Control of Radiation Releases Associated with Decommissioning Events),
there are listed a number of other potential decommissioning related accident scenarios which are task or project specific. At this stage of the decommissioning, and without knowing the implementation details of each task or project, it is inappropriate to conduct detailed safety analysis assessments of these potential accidents.
With the limiting licensing basis radiological analyses defined previously, the accident radiological consequences of those ever,ts identifbd in section 7.3 (or any aber decommissioning accidents that may be created) are expected to be bounded. Accident safety assessments, in accordance with 10 CFR 50.59 requirements, will be conducted at a time more appropriate to the specific task.
Maine Yankee has, in the course of decommissioning activities, reached one of these activities in which a potential accident scenario evaluation is required. As part of the reactor coolant system decontamination project, concentrations of radionuclides were formed on resin beads in ion exchange columns. Analyses were conducted to assess the potential off site radiologicalimpact of the release of these radionuclides from a ' resin fire" accident scenario. This scenario, and other relevant radiological assessment portions of the project are discussed below. It is anticipated that similar assessments will be conducted for other decommissioning activities.
In addition to the accident scenarios identified in section 7.3 of the DSAR, Maine Yankee conducted a calculation related to the radiological dose rates ant:cipated at the site boundary for a beyond design basis accident resulting from a voiding of the spent fuel pool. The calculated skyshine dose rate:: for this type of event are discussed later in this attachment.
REACTOR COOLANT SYSTEM (RCS) DECONTAMINATION Tne reactor coolant system decontamination project was conducted with the use of a circulating corrosive liquid solution in the reactor coolant piping to disstive the fixed radioactive contarnination. Contamination particles and activated corrosion products were stripped from the RCS liquid solution by passing the fluid through ion exchange columns containing a plastic-like resin. Upon reaching predetermined r.ctivity or pressure drop levels, the ion exchange column resin was isolated, sluiced into a High Integrity Container (HIC) dewatered, dried, and prepared for shipment off site. All work was performed inside the closed containment building.
A technical evaluation, in support of 'he requirements of 10 CFR 50.59, was performed to assess the r.utential radiological accidents of the RCS decontamination project. This evaluation assessed three possible accident scenarior 1) a spill of radioactive liquid associated with the decontamination process; 2) a dry resir, spill and fire from a fully loaded HIC; and,3) a drop of a 13,000 pound plug fixture resulting in damcge to the cavity seal ring away from the location of the radionuclides concentrations.
A separa'e a* aaent determined that appropriate rigging practices would be sufficient to address the drop of the pleMxture. Since this assessment did not involve a radiological component, further discussion is not provided here.
The spill of radioactive RCS decontamination liquid was determined to be bounded by the DSAR accident analysis for Radioactive Liquid Waste System Leaks and Failures, section 5.6.2 of DSAR Rev.14 (see the description of the Primary Drain Tank Rupture accident discussed previously). This finding was based on the Page 3 of 6
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Attachment Letter MN-98-050 high volume of the RCS decontamination fluid coupled with a low specified radioisotope contamination level of s72 pCi/ml, as based on previous RCS decontamination at Maine Yankee.
The evaluation of the dry resin spill from a full HIC, coupled with a resultant fire of the spilled dry resin, was I
comprised of two parts. Initia!!y, an e"aluation was performed which determined the offsite radiological doses l
associated with a dry resin spiH ant. resultant fire inside the containment. It was determined that the TEDE associated with the resin spill and fire in the containment was approximately 1 mrem, well below the PAG limit of 1 rem. A second assessment determined the impact of a resin fire on containment pressure to ensure that the containment, as sealed in the decommissioning state, would remain as a contamination control boundary without leakage. The impact on containment pressure of a resin fire was determined to be negligible, an increase of 0.03 psi.
Dry Resin Spill and Fire Analysis-1 in performing tnis evaluation, a number of assumptions were made which were translated into administrative requirements for the project duration. The key assumptions are listed below:
HIC Related Conditions:
a The HIC shall contain no more than 113 cubic feet of resin.
The HIC shall meet the requirements of 49 CFR 173 for Class A type packing.
m The HIC shall contain no more than 250 curies of activity.
m The HlC shall be sealed la the shipping cask and the cask sealed as for over the road shipment m
prior to moving the HIC out of containment.
One ion exchange column at a time shall be sluiced into the HIC.
m e The containment must be sealea when transferring resins into the HIC.
Containment Sealing Conditions:
The containment shall be sealed in accordance with good engineering practices and such sealed a
boundaries controlled by the Operations Department.
The containment vent and purge sysicm must be turned off curing any resin sluicing, resin drying m
or HIC handling operations.
The personnel batch shall be prepared and appropriate instruction given for closure within %
e hour of a dried contaminated resin spill.
The equipment ha'ch shall be sealed to prevent air migration into or out of the containment.
e Resin transfer or drying operations are not permitted during any period of time the containment m
is unsealed.
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Fire Initiation / Loading Conditions:
l All transient combustibles are removed from the area around the HIC, shipping trailer, s
pump / heater skid and ion exchanger columns, a The truck / tractor / forklift attached to the lowboy trailer has been disconnected and is removed from the containment before sluicing operations.
m All propane or ns' ural gas tanks are removed from, the aren The equiprunt hatch sealing membrane is in place and scad. This membrane and the sealing a
mecnanism is composed of fire retardant / resistant matterials.
The resin drying operation does not directly heat the resins in the HIC.
m The use of the abovo administrative controls, coupled with a well defined decontamination process, resulted I
in a calculated off site dose rate well within the limiting licensing basis calculations as defined in the DSA't.
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o Attachment Letter MN-98450 i
i Containment Pressurization Resulting from a Spilled Resin Fire:
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l This analysis was performed to specifically address the containment pressurization from a fire associated with a resin spill during the RCS decontamination effort. A pressurization of the containment would potentially result in a compromise to the containment sealing mechanism. The calculation evaluated the increase in containment pressure by assessing the energy and mass added to the containment atmosphere as a result of the buming resin. The buming resin would add energy to the containment atmosphere in the form of heat.
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Smoke from the buming resin would add mass to the containment atmosphere. The calculated increased in l
mass and temperature resulted in an estimated increase in containment pressure of 0.03 psi. The magnitude l
of this increase indicated that there is no impact on the containment pressure as the result of a resin fire.
1 BEYOND DESIGN BASIS SKYSHINE DOSE RATES The assessment of a beyond design basis accident, a complete drainage of the spent fuel pool, was conducted as a limiting condition tr3nsient although this accident is not considered part of the licensing or design basis. The radiological assessment portion of that transient. the skyshine dose rates, are presented in graphical format belew:
Fig. 5.6. Maine Yankee Skyshine Radiation Fields se a Fur.ction of Horizontal olstance from the spent Fuel Pool Center, for a Totally Drained Pool and Vartous Post-shutdown Decay nmes i.0E.0 ~ -
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0 100 200 300 400 500 600 700 800 900 1000 Hortaontal Distance from JPP Centertine (m)
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4 Attachment Letter MN48 050 IV.
SUMMARY
The Maine Yankee radiological analyses support both the licensing basis accidents and specific decommissioning activities. The licensing basis threshold, consistent with the decommissioning condition of the plant, for the site boundary dose rates, are well within the EPA Protective Action Guidelines of 1 rem.
Decommissioning activities of a significant nature are condu ded to ensure that the radiological site boundary dose thresholds cannot be exceeded.
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