ML20236X168
| ML20236X168 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 08/03/1998 |
| From: | Meisner M Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20236X169 | List: |
| References | |
| MJM-98-45, MN-98-54, NUDOCS 9808070176 | |
| Download: ML20236X168 (1) | |
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MaineYankee P.O. BOX 408 + WISCASSET. MAINE 04578 + (207) 882-6321 l
l August 3,1998 l
MN-98-54 MJM-98-45 l
UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk l
Washington, DC 205S5 l
l.
Reference:
(a) License No. DPR-36 (Docket No. 50-309) l
Subject:
Submittal of the Maine Yankee Defueled Safety Analysis Report Rev.15 and 16 Gentlemen:
Pursuant to the requirements of 10 CFR 650.71 and 10 CFR 50.4, please find enclosed Revisions 15 and 16 of the Maine Yankee Defueled Safety Analysis Report (DSAR) and ten copies. Separate copies are being supplied to the USNRC Region i office, the Maine Yankee Project Manager and the on-site NRC office.
Ve truly ypur,
l Mich I Meisner, President Maine Yankee ENCLOSURE c:
Mr. H. J. Miller, USNRC Region 1 (copy 39)
Mr. M. K. Webb, USNRC Headquarters (copy 40)
USNRC Maine Yankee site office (copy 7)
Mr. P. J. Dostie, State of Maine (copy 9) w/o enclosure:
h[b Dr. M. T. Masnik, USNRC Headquarters Mr. M. Roberts, USNRC Region 1 Mr. U. Vanags, State of Maine STATE OF MAINE l
l Then personally appeared before me, Michael J. Meisner, who being duly swom did state that he is the President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing request in the name and on behalf of Maine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief.
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MYAPC 1
1.5 Material incorporated By Reference l
Certain program documents and associated topical reports or analyses have been incorporated into the DSAR by reference and are listed in each sechon as appropriate. This documentation may include information developed by Maine Yankee, as well as Yankee Atomic, ABB-CE, Westinghouse, Stone and Webster, and other organizations.
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Some documentation that is incorporated by reference continues to be updated to assure that the information used is the latest available. These documents include the following:
1.
Quality Assurance Program j
2.
3.
Security Plan
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4.
Off Site Dose Calculation Manual
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6.
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l 7.
Post Shutdown Decommissioning Activities Report
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l 8.
Technical Specifications
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l Each of these programs and plans may be modified as necessary in accordance with the regulatory i
and Maine Yankee requirements identified in section 6.
DSAR 1-23 Rev.15 r
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MYAPC SECTION 5.0 l
ACCIDENT ANALYSIS 5.1 Introduction Earlier sections of this report describe the rnsjor systems and components of the plant from the perspective of safe eoent fuel handling, spent fuel storage, and other decommissioning activities I
as would be appropriate to a permanently defueled plant.
This section uses the previous infomiation and examines the potential consequences cf accidents and incidents, notwithstanding the precautions taken to prevent their happening, to assess the adequacy of the plant design in minimizing or mitigating potential consequences of such occurrences. Additionally, the accident analyses presented in this section provide assurance that l
the health and safety of the public is protected front the consequences of even the most severe of l
the hypothetical incidents analyzed.
With the pem1anent defueling of the Maine Yankee facility and the certification of the cessation of
. operations, the postulated accidents associated with reactor operation are no longer applicable and need not be considered. Likewise, the unirradiated nuclear fuel has been removed from the Maine
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Yankee site and therefore accidents involving new fuel assemblies are also no longer applicable.
However, those accidents associated with the storage or handling of irradiated fuel or radioactive I
waste storage or processing remain applicable and are discussed within this section.
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The general classification of accidents for the permanently defueled condition are limited. These l
groupings are listed as follows:
l 1.
Inadvertent criticality of the stored spent fuel, l
2.
Fuel assembly handling accident, 3.
Spent fuel shipping cask drop in the spent fuel pool, 4.
Loss of spent fuel decay heat removal capability, 5.
Loss of spent fuel poolinventory, 6.
Radioactive release from a subsystem or component, or j
7.
Low level waste storage accident.
DSAR 5-1 Rev.15
4 MYAPC 5.3 Fuel Handling Accident The purpose of this section is to m**au anticipated spent fuel pool fuel handling operations in order to arrive at the accident which would result in appropriately conservative off-site and control room radioactive release effects. Fuel handling incidents originating in the containment are not appl; cable to the permanently defueled condition. Fuel handling operation associated with the use -
of a spent fuel cask are addressed in section 5.4.
The likelihood of a fuel handling incident in the spent fuel pool is minimized by implementation of appropriate and long standing administrative controls and physical limitations imposed on fuel handling operations. All fuel handHng operations are conducted in accordance with prescribed procedures under the direct surveillance and supervision of qualified personnel.
The fuel handling equipment and facility are designed for the transfer and handling of a single fuel assembly at any time, and movement of equipment when handling the fuel is administratively restricted. The fuel handling manipulators and hoists are designed so that fuel cannot be raised above a position which provides adequate shield water depth for the safety of operating personnel.
'n the spent fuel pool, the design of fuel storage racks and manipulator equipment, in conjunction with appropriate administrative controls, is such that:
1.
Fuel is always maintained by mechanical restraint. Fuel at rest is positioned by positive restraints in a subcritical geometrical array, with no credit for boric acid in the water.
2.
Fuel can be manipulated only one assembly at a time.
3.
Violation of procedures by placing one fuel assembly in juxtaposition with any i
group of assemblies in racks will not result in criticality.
4.
The spent fuel shipping container does not pass over spent fuel during transfer )
operations.
The fuel assembly is immersed continuously while in the spent fuel pool. Adequate cooling of fuel during underwater handling is provided by convective heat transfer to the surrounding water. The fuel handling equipment and spent fuel pool are described in detail in Section 3.
Inadvertent disengagement of the fuel assembly from the fuel handling equipment is prevented by design, mechanical, and procedural interlocks. Consequently, the possibility of dropping and DSAR 5-9 Rev.15 l
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MYAPC 4.
No fuel overheating occurs, and thus, no significant solid fission products are
. released.
5.
No credit is taken for the isolation of the spent fuel building The release is modeled as an instantaneous puff release at accident (95 percentile) ground level meteorological conditions.
6.
The free air volume of the CR is 38,000 cubic feet.
7.
Normal CR ventilation is either 120 or 900 cubic feet per mirtute for the duration of the accident.
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8.
No credit is assumed for CR isolation or air filtration.
l The radiological analyses was performed using the code ELISA, Reference 1.
The results of this analysis for the EAB are presented in Table 5.3.2. These results show that the projected doses from the fuel handing accident are insignificant in comparison to the 10 CFR 100 iimits and far less than the Environmental Protechon Agency Protectrve Action Guidelines (PAGs).
l The calculated doses for the CR portion of the analysis at the CR ventilation intakes and inside the CR are presented in Table 5.3.3. These results show that the anticipated doses are within the 10
'W-CFR 50, General Design Criteria 19 dose limits.
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DSAR 5-11 Rev.15 l
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MYAPC l
5.4 Scent Fuel Cask Dron i
Spent fuel shipping casks are designed, as per the requirements of 10 CFR 71, to withstend a free fall of 30 feet onto an unyleiding surface. For this reason, radiological consequences of a postulated spent fuel cask d.op accident outside the spent fuel pool are net.wuired if potential drop distances are less than 30 feet.
The design of the Maine Yankee spent fuel cask transfer system is such that the cask drop distance is less than 30 feet whenever the cask is not directly over the spent fuel pool. It is conduded, therefore, that an evaluation of the spent fuel cask drop accident outside of the spent fuel poolis not necessary.
Operations with the spent fuel shipping container are designed not to pass over spent fuel storage racks or spent fuel assemblies during cask loading or fuel transfer operations.
At the current time, Maine Yankee is prohibited from lifting a sp6ut fuel shipping cask over the spent fuel pool. Therefore, an accident analysis of a spent fuel cask drop in the spent fuel pool is not required and does not supply safety analysis limits.
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DSAR 5-16 Rev.15
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l MYAPC heat calculations v, conservatively determined assuming prior plant operation at 2700 MWt for the most recer~; discharged spent fuel. Actual power operation was limited to 2440 MWt for the
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last batch of fuel used for power production at Maine Yankee.
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in late October and early November 1997, Maine Yankee performed passive cooling tests in the spent fuel pool to obtain cata in assessing the pool heatup rate, heat losses due to evaporation at elevated ternperatures, and accuracy of the calculation of spent fuel decay heat. Conservative
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assumptions were used for inclusion of the fuel building heat losses in this assessment. The results of those tests indicate that passive (i.e.: no forced flow through the cooling system) cooling of the spent fuel pool will result in a steady bulk water temperature of between 190 and 200*F with water surface wind speeds between 3 and 5 miles per hour. When compared to the analytical predictions of the spent fuel decay heat using the NRC Branch Technical Position ASB 9-2, these results demonstrate substantial margin to the decay heat analysis assumptions presented in this section.
5.5.1.1 Blocked / improper Cell Flow The design of the spent fuel racks (Reference 1)is such that the top of the rack cells may be blocked and sufficient cooling of the fuel assembly is still assured. Analyses of various types of flow blockage of the cell exit have been performed to demonstrate satisfactory preservation of the stored fuel in a coolable geometry under these conditions.
Two types of cell blockage were considered. The first assumed that a fuel assembly was laying horizontally across the limiting spent fuel storage cell. Conservatively neglecting flow upwards j
through the horizontal pins, this event results in a net blockage of the storage cell exit area of 79%.
The second flow blockage scenario involves the placement of a 8 by 10 foot section of masonry wall from the south end of the spent fuel pool onto the top of the storage cells. This event blocks i
100% of multiple cell exit openings and rcquires the cooling of the spent fuel through the 1 inch diameter flow holes located near the top of the cells.
Specific analysec were performed using the RETRAN code (Reference 2) in analyzing the blockage of cells to conservative!y predict the limiting fuel cell coolant conditions in verifying that localized boiling or that the onset of Critical Heat Flux (CHF) wiF not occur in the individual fuel storage cell.
The limiting cell was then evaluated to determine the peak fuel pin surface cladding and fuel pellet temperatures. These analyses were performed as part of Reference 1.
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DSAR 5-18 Rev.15
MYAPC 5.6 Low Level Waste Release incidents 5.6.1 Radioactive Wasta Gas System Leaks and Failures Radioactive waste gas decay tanks permit decay of accumulated radioactive gases prior to their j
release as a means oWucing the normal release of radioactive matenals to the atmosphere. The radioactive contents were pnncipsily the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens. Since these nobie gases are generated from the fissioning process during power operation, there will be no generation of fission gases and no more gases sent to the waste gas decay tanks from the Reactor Coolant System with the reactor permanently defueled. The inventory of the waste gas decay tanks resulting from the last penod of power operations has been released and the tanks contain only residual amounts of radioactive waste gases.
The projected radioactive gas release information and projected doses that are applicable to the permanently defueled condition are defined in sections 5.3,5.5, and 5.6.
5.6.2 Radioactive Uquid Waste System Leaks and Failures Because of the continued processing of radioactive fluids at Maine Yankee during the permanently defueled condition, the analysis of radioactive liquid waste system leaks and failures remains essentially the same as that used as the licensing basis for the plant during power operations. The use of the full power transient is used as representative of the maximum dose consequences associated with the offsite receptor due to a liquid and gaseous release. This transient is not a credible accident scenario in the defueled condition. It is not intended to be the sole basis upon j
l which the retention of systems or components is required.
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Actions leading to transfer of a radioactive flid #om a system to the environment or to another system require positive operator control ant *nonitofng.
All operator actions required are performed ir sccordance with written operating procedures which ]
include instructions, checklists and allowable release information. Following final radioactivity analysis, effluent is released to the environment through a process radiation monitor lineiicaed
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with the discharge valve such that a high radiation alarm will close the valve and terminate the release.
I DSAR 5-37 Rev.15
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MYAPC With these considerations in mind, the probability of the accidental release of radioactivity from the f
radwasta systems as a result of operator error is minirnal.
The purpose of this section, however, is to consider postulated liquid waste system single component leaks or failures in order to anive ta ne postulated incident which could have maximum l
off-site effects.
Method of Analysis:
A.
Releases to the Atmnanhere (Gmemous Releases)
For the purpose of establishing an upper limit on the activity released from a single component failure in the liquid waste system, it is assumed that the l
primary drain tank fails, releasing its total inventory. The primary drain tank l
failure has been selected since this tank has the highest inventory of dissolved noble gases and halogens during operational periods. The release takes place
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as a liquid spill on the floor of the compartment in t'le waste processing building where the tank is located. Radioactivity is released to the atmosphere from noble gases and halogens evolved from the spilled liquid.
The following conservative assumptions are used in conjunction with the meteorological and dose assumptions given in Appendices SA and 58.
l 1.
Eighty percent of the primary drain tank's 8,150 gallori capacity is filled with undecayed, un-degasified primary reactor coolant (with activity concentrations at Technical Specification limits of 1.0 pCl/g Dose Equivalent 1-131 and 100/iii pCilg).
2.
One hundred percent of the tank's inventory is spilled and all of the noble l
gases and 1 percent of the radiciodines are available for direct release through the building ventilation system to the environment.
3.
Duiation of the release is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B.
Potential Releases to the Groundwater Table Postulated liquid spills escaping concrete structures may be released to the site groundwater table. The groundwater table at the Maine Yankee site, however, flows towards Back River and Montsweag Bay, both of which are tidal saltwater DSAR 5-38 Rev.15
MYAPC 6.5.4 Fitness For Duty Program i
As a result of the permanently shutdown condition of the plant and the 10 CFR 50.82(a)(1) certifications, the NRC has concluded that the Fitness for Duty Program rule,10 CFR 26, no longer l applies to Maine Yankee, Reference 1.
l 6.5.5 Offsite Dose Calculation Manual The Maine Yankee Offsste Dose Calcul-* ion Manual (ODCM) is defined by Technical Specifications to contain the methodology and param rs used in the calculation of off-site doses resulting from radioactive gaseous and liquid ethuents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints and in the conduct of the Environmental Radiological Monitoring Program.
The ODCM shall also contain the Radioactive Emuent Control and Radiological Environmental Monitoring Programs and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Emuent Release Reports.
Changes to the ODCM shall be made in accordance with Maine Yankee Technical Specifications and mcv 6 inade if the change will maintain the level of radioactive emuent control required by applicable regulations and not adversely impact the accuracy or reliability of emuent, dose, or setpoint calculations.
6.5.6 Quality Assurance Program The Maine Yankee Quality Assurance Program is docketed as a separate document and is required by 10 CFR 50.54(a). Changes to the Quality Assurance Plan are evaluated under 10 CFR 50.54 (a) etiich allows changes to be made without NRC approval if these changes do not reduce the commitments in the program description previously accepted by the NRC. These changes uit be submitted to the NRC in accordance with the requirements of 50.71(e), FSAR update m
requirements.
l 6.5.7 Process Control Program The Process Control Program (PCP) contains the current formulas, sampling analyses, tests and l
determinations to be made to ensure that processing and packaging of solid radioactive wastes l
DSAR 6-10 Rev.15
O MYAPC
References:
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1.
Leder, USNRC to Maine Yankee, " Fitness for Duty Programs (10 CFR 26) for Maine l
Yankee Atomic Power Station", dated January 12,1998.
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DSAR 6-14 Rev.15 i
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