ML20127P247

Proposed Tech Specs Change 177 Re Main Yankee Spent Fuel Pool Reracking

ML20127P247
Person / Time
Site:	Maine Yankee
Issue date:	01/25/1993
From:	Maine Yankee
To:
Shared Package
ML20127P223	List: ML20127P219 ML20127P247 ML20127P256
References
NUDOCS 9302010222
Download: ML20127P247 (13)

Text

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F Attachment C Prooosed Technical Specification Chanaes T..

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hl'FVELSTORAGE Applicability:

Applies to the capacity and storage arrangements of the new and spent fuell facility Ob.iective:

To describe and define those aspects of fuel storage which relate to the prevention of criticality in the fuel storage facility.

Specification:

A. The new and spent fuel pool structures ' including fuel racks are designed to withstand the anticipated earthquake loadings as Class I-structures. The spent fuel pool is lined with stainless steel- to ensure against loss of water.

[ B. Fuel shall be stored vertically in racks. The racks are designed to maintain fuel assembly center to center distances that will assure K.,,

is less thanwater.

unborated or e qual to 0.95 even with the pool filled with C. ex

[ Whenever there is fuel in the spent fuel poolnew fuel .all storage, the be filled spent fuel storag with water borated to the refueling water boron concentration. This concentration matches that in the reactor cavity and rafueling tanal during refueling operations.

D. Spent fuel shipping casks shall not be lifted over the . spent fuel-storage pool.

E. No more than 2019 standard fuel assemblies shall be stored in either Region I or Region 11 of the spent fuel pool in accordance with the limitations of Figure 1.1-1. Unirradiated fuel assemblies shall be stored in either the New Fuel Storage Area or Region I of the spent fuel pool. Consolidated fuel shall be stored in Region II only.

F. No more than 121 additional standard fuel- assemblics may be temporarily stored in a temporary spent fuel storage rack to be located in the spent fuel cask laydown area. These are in addition to the 2019- standard fuel assemblies of specification E. All 121 assemblies shall be suitable for-placement into Region 11 racks.

G. No more than 20 standard fuel assemblies may be in consolidated; form.

[ These are included in the 2019 standard assemblies of specification E.

Basis:

Reference--

Safety analyses, Reference (a) and NRC safety evaluation reports, k- design.

ib)ludeddocument inc within the configuration this design are of Maine Yankee's. spent considerations for fuel the rac storage of-consolidated fuel. These reports demonstrate the t^ 'y and environmental acceptability of storing standard the Region II permanent and consolidated storage locations. The RegsIu on Ifuel assemblies in stora are designed to accommodate low burnup or unirradiated fuel.ge Up locations to 121 additional standard fuel assemblies may be stored in temporary storage in tne spent fuel cask laydown area, Reference (c) and (d).

1.1-1. Amendment No. 75 Le\propchag)PC177

References:

'(a) Maine Yankee letter to USNRC dated January' 25,-1993; " Maine Yankee Spent Fuel Pool Reracking"; MN-93-09.

(b) USNRC Staff Safety-Evaluation of the Maine Yankee License Amendment Application dated January 25, 1993.

_[ (c) Maine Yankee letter to USNRC dated October _5,1981, " Maine Yankee

' Spent Fuel Storage Modification - Complete Report" with enclosure, as supplemented and amended on February 10, 1982, May 7, 1982, and May 26, 1983.

[ (d) USNRC letter to Maine Yankee dated June 16,1982, " Safety Evaluation and Environmental Impact Appraisal Regarding Maine Yankee Spent Fuel Storage" as supplemented via USNRC letter to Maine Yankee dated October 22,1982, " Resolution of Open Items - Safety Evaluation of Maine Yankee Spent Fuel Storage". ..1 l

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1.1-2 Amendment No. 75 i

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' ' . . ... .. , , , , ,, , ,, a h C m o b h (fi1W/GMO) dnwns 80eJe^y Al4wessy MAINE YANKEE Spent Fuel Pool Figure Technical Assembly Placement Limitations 1.1 1 Specification 1.1 -3

e 3.13 REFUELING AND FUEL CON 5QLIDAT10N OPERAT!Q1S

Apolicability:

Applies to- operating -limitations during refueling and fuel consolidation operations.

Objective:

To minireize the possibility of an accident occurring during refueling and fuel consolidation operations that could affect the health and safety of plant personnel and the public.

Soecification:

A. Prior to each refueling a complete checkout, including a load test, shall be conducted on fuel handling cranes that will be used to handle irradiated fuel assemblies.

[- B. Irradiated f uel shall not be moved until 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> after the reactor has been made subcritical-.

C. Whenever the reactor vessel head is removed and there is fuel in the reactor, the refueling boron concentration shall be maintained in the reactor coolant system and shall be checked by sampling on each shift to insure that it is sufficient to maintain the core 5% delta k/k subcritical.

D. The following conditions shall be satisfied during core altera-tions or movement of irradiated fuel within the containment:

1. The containment equipment hatch must be closed and held in place by a minimum of four bolts when the reactor has been subcritical less than 210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br />.

2. At least one door in the personnel airlock shall be closed when the reactor has been subcritical less than 210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br />.

Exception: Both personnel airlock doors may be opened briefly within the 210 hour-period'to allow passage of long objects through the airlock.

3. The containment venting and purge inlet and outlet. trip valves shall be closed or be operable and able to . isolate-the ventilation system in response to high radiation signals.

The venting and purge system discharge shall be filtered-through the high efficiency particulate air filters and charcoal absorbers.

Exception: The HEPA filters and charcoal absorbers may be bypassed after the reactor has been subcritical for 210 or more hours.

3.13-1 Amendment No. 54, 65, 73, 75 Le\propchag\PC177

9. Both RHR Trains A and' B shall be. operable when the water level above the top of the irradiated fuel assemblies seated within the reactor vessel'is-less than 23 feet.

Exception: Only one power source, normal or eme agency, is required for an RHR train to be considered operable.-

10. Maintain a minimum of 23 feet of water above the top of the core whenever irradiated fuel is being moved,

11. Direct communication between personnel in the control room and at the refueling station shall be operable whenever changes in core geometry are taking place.

Remedial Actiorp If any of the conditions or associated remedial actions in Specification D are not met or remedial actions are not specified, core geometry changes and movement of irradiated fuel in the containment shall cease immediately; and no operations that may increase the reactivity of the core shall be made.

E. Spant fuel storage racks may be moved only in accordance with written procedures which ensure that no rack modules are moved over fuel assemblies.

F. The following condition shall be satisfied during fuel consolidation:

1. Irradiated fuel shall not be consolidated until it has been

[ cooled at least 730 days after final discharge from the reactor.

Basis:

The equipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the above specifications and the design of the fuel handling equipment incorporating built-in interlocks and safeguards systems provide assurance that no incident could occur during the refueling operations that would result in a haz rd. to public health and safety.

After being shutdown for 210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br />, the fuel has decayed sufficiently to maintain dose levels, during a postulated fuel handling accident inside containment within 10 CFR Part 100 limits. Therefore, the containment ventilation / purge filter system may be bypassed to prevent unnecessary filter depletion which might result from fumes given off 'by painting or welding during the outage period.

The exception to paragraph 3.13.D.4 permits routine testing of the radiation-monitors' without incurring unnecessary wear of the purge valve' resilient seals. Weekly testing of these trip valves is sufficient to insure their operability. ,

i 3.13-3 Amendment No. 19, 54, 63, 65, 73, 75 Lt\propchng\PC177

Whenever cha'nges are not being marie' in core geometry, one source range-neutron monitor is sufficient. This permits maintenance of the instru-mentation. Continuous monitoring of radiation levels and neutron flux-provides immediate indication of an unsafe-condition. The residual heat-removal flow is used to remove core decay heat and maintain a uniform boron concentration.

A single cooling mechanism is sufficient to remove decay heat but single failure considerations require that two mechanisms by OPERABLE, Cooling mechanisms available incluJe RHR trains A and B and 23 feet of water -

above the top of the core.

The shutdown margin of 5% delta k/k will keep the core substantially sub-critical, even if the highest worth CEAs were inadvertently withdrawn from the core without compensating boron addition.

Periodic checks of refueling water boron concentration insure the proper shutdown margin. Communication requirements allow the control room operator to inform the refueling station operator of any impending visual condition detected from che main control board indicators during fuel movement, in addition to the above engineered safeguards systems, interlocks are utilized during refueling to insure safe handling. An excess weight inter-lock is provided to prevent excess loading of a fuel assembly, should it inadvertently become stuck.

In the analysis of the refueling accident conducted by the staff, 23 feet of water and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay time were used to limit exposures to 10% of 10 CFR 100. Valve alignment check sheets are completed to protect againsti

[ sources of unborated water or draining of the system. An additional 72

[ hours of decay time has been added to ensure. lower assembly decay heat

[ 1evels in the spent fuel pool during refuelings. Therefor.e, the total delay

[ time prior to the movement of irradiated fuel is 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> after the reactor

[ has been made subcritical.

To ensure containment fission product barrier continuity during the. period when the containment ventilation / purge filter is in operation, the equip-ment hatch and at least one door in the personnel hatch should generally remain closad. The exception allows the personnel hatch to be opened for brief periods to permit passage of long objects in support of-refueling operations during the initial 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> period after making the reactor subcritical. These restrictions do not apply after the 210 hour0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br /> period.

Procedures are required for movement of. spent fuel racks to avoid unnecessary risk of spent fuel damage caused by dropping spent fuel racks.

[ The 730 day cooling period after discharge from the reactor allows substantial radio-active decay. This ensures that the dose consequences of a consolidated. spent fuel handling accident are bounded by the consequences of the design basis spent fuel drop accident. -It also ensures that the maximum outlet _ temperatures for the - limiting fuel assembly and the consolidated fuel storage bundle are both well below the saturation temperature at the cell outlet for any storage array, i

L 3.13-4 Amendment No. 63,--65, 73, 75 Lt\propchng\PC177 l

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Attachment 0 Comoarison of Sionificant Desian'Attribut'es-and Analytical Methodolooies to Previously - .j Aooreved Acolications

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1 Comparison of Stanificant Desian Attributes and Analytical Methodoloaies to Previous 1Y Acoroved Acolications As indicated in both the cover letter and Attachment E (Licensing Report), the-significant design attributes and associated safety analyses methods of the proposed Maine Yankee reracking have been approved by the USNRC in other reracking a)plications. With this Attachment, a comparison of the more significant claracteristics of the proposed Maine Yankee reracking with a representative selection of previously approved applications is provided.

The comparison which follows is by no means to be regarded as comprehensive in defining -either the complete list of design attributes or all of the possibly applicable prior applications. In identifying the listing of key characteristics, an attempt was made to recognize those items which may be classified as identifying the proposed reracking from a global or overall perspective.

Likewise, in comparing these characteristics to previously approved applications, those applications selected were based on either recently approved reracking submittals, relevant approved methodology applications at other Yankee plants, or the latest approved Maine Yankee reracking. Wherever possible, the latter comparison is presented.

The comparison presented in this Attachment, coupled with the detailed information of Attachment E, readily demonstrates that the proposed Maine Yankee reracking is based on design concepts and analysis methods which have been previously approved and accepted by the USNRC.

MECHANICAL DESIGN ATTRIBUTES No . - _

Proposed Maine - Previously Approved-Key Attributes- ~ Yankee' Attribute

~ Application-1 Basic Rack Design Free Standing, Single Tier Maine Yankee (License DPR-36) 2 Storage regions based on Two region poo1~ TMI-1 burnup and initial (License DPR-50) enrichment 3 Region I Design Flux trap Maine Yankee (License DpR-36) 4 Region-II Design Maximum storage ~ density viith TMI-1 welded corner-to-corners (License DPR-50) 5 Region II Cell Design Virtual cells TMI-l (License DPR-50) 6 Cell Pitch Region I: 10.5 inches Maine Yankee (10.25 inches)

Region II: 9.085 inches (License DPR-36)

I DC Cook (8.97 inches)

} (License DPR 58, 74)

U \propchng\PC177 9

No. - Proposed Maine Previously Approved Key Attributes Yankee Attribute Application-7 Neutron Absorber BORAL" THI-l (License DPR-50)-

Maine Yankee (License DPR-36)- _

8. Maximum initial Enrichment 4,5 U"' weight percent- TMl-1 (5 weight percent)

(License DPR-50)-

9 Structural Material 304L SS Maine Yankee (License DPR-36 DC Cook (License DPR 58, 74) 10 Cell Wall Thickness 14 gauge TM1-1 (License DPR-50) 11 Flow Openings lateral Flow Holes DC Cook Single bottom flow hole (License DPR 58, 74)

Maine Yankee (License DPR 36) 12 Seismic Analysis 3-0, Nonlinear, dynamic TMI-l (including whole pool) (LicenseDPR-50)

DC Cook (License DPR 58, 74) 13 Stress Loading SRP 3.8.4, Appendix 0 DC-Cook Combinations (License DPR 58,74) 14 Fluid Coupling Accounted for in analysis DC Cook (License DPR 58, 74)

Maine Yankee.

(License DPR-36) 15 Frictional Sliding 0,2 to 0.8 Maine Yankee ..

Coefficient (License DPR-36)

THI (License DPR-50) 16 Consideration of Consequences well within Maine Yankee Accidental Fuel Assembly 10CFR100 (License DPR-36)

Drop TMl-1 (License DPP.-50)_

DC Cook (License DPR 58,-74)

-17 Consideration of Appropriate procedural and Maine Yankee Installation Accident programmatic controls (License DPR-36)

TMl-1 (License DPR 50)

DC Cook (License DPR 58, 74) m,....,ve m 20 i

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JBIRMAL-HYDRAULIC DESIGN ATTRIBUTES I

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Proposed' Maine 1 .'Previously App' roved' No. . Key Attribute Yankee Attribute'  ? Application 1 Thermal-Hydraulic RETRAN Maine Yankee Analysis 2-D modelling (License DPR-36)

(Pin Cooling) Vermont Yankee (License DPR-28)

Yankee Rowe (License DPR-3) 2 Spent Fuel Pool Cooling 22.0 X 10' BTU /hr Maine Yankee System (SFPCS) Design (License DPR-36)

Heat Removal Rate 3 No. of SFPCS pumps / heat 2/1 Maine Yankee exchangers (License DPR-36) 4 Maximum SFP Bulk Water 154' Maine Yankee Temperature (License DPR-36) 5 No. of Sources of SFP 3 Maine Yankee Make up Water (License DPR-36) 6 Coolability of Individual No localized boiling and Maine Yankee fuel Assemblies / Fuel Pins long term maintenante of (License DPR-36) fuel integrity Vermont Yankee (License DPR-28)

Yankee Rowe (License DPR-3) 7 Loss of SFPCS Flow Maine Yankee Accident Analysis (License DPR-35)

Normal fuel offload No local boiling Fuel core offload - No bulk boiling 8 Loss of SFP Cooling Sufficient time- to Maine Yanbe System Operation Analysis establish alternate (License DPR-36) cooling / makeup Vermont Yankee (License DPR-28)

Yankee Rowe (License DPR-3) ,

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_ GITICALITY DESIGN ATTRIBUTES

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_ No.; ' Key Atteibutes- -Proposed Maine- Previously Approved;

-Yankee Attribute Application t 1 Criticality Analysis KENO-V,a Maine Yankee Methodology CASM0-3 (LicenseDPR-36)

PDQ-7 Vermont Yankee SIMULATE-3 (License DPR-28)

Yankee Rowe '

(License DPR-3) 2_ Maximum Effective k,,, s 0,95 with 95% Maine Yankee Multiplication Pactor of probability at 95% (License DPR-36)

Racks when loaded with confidence level Vermont Yankee fuel (License DPR-28)

Yankee Rowe (License DPR-3) 3 Consideration of Effect K.,, s 0,95 with 95% Maine Yankee on Reactivity of probability at 95% (LicenseDPR-36)

Misplacement of Fuel confidence level Vermont Yankee Assembly (includes credit for (LicenseDPR-28)

SFP Boron) Yankee Rowe (License PDR-3)

Lt\propchag\PC177 22

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Attachment E '-

Licensina Report; Soent Fuel Pool Retackina I >

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