ML20207E976

Errata to Amend 188 to License DPR-40,correcting Incorrect Page Number,Missing Parenthesis & Misspelled Word

ML20207E976
Person / Time
Site:	Fort Calhoun
Issue date:	03/03/1999
From:	NRC (Affiliation Not Assigned)
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ML20207E873	List: ML20207E869 ML20207E976
References
NUDOCS 9903110126
Download: ML20207E976 (4)

Text

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I 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline Bases (Continued) 2.8.2(I) Containment Penetrations (Continued)

For automatic isolation vrives with direct access to the outside atmosphere to be OPERABLE requires that the Ventilation Isolation Actuation Signal (VIAS) is OPERABLE in order to close the valves. This action prevents release of significant radionuclides from the containment to atmosphere. During CORE ALTERATIONS and REFUELING OPERATIONS, the OPERABILITY of VIAS is addressed by Specification 2.8.2(3).

When *immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner.

2.8.2(2) Refueline Water Level  !

Prior to REFUELING OPERATIONS inside containment, the reactor refueling cavity is i filled with approximately 250,000 gallons ofborated water. The minimum refueling water level meets the assumption of iodine decontamination factors following a fuel handling accident. When the water level is lower than the required level, CORE ALTERATIONS and REFUELING OPERATIONS inside of containment shall be suspended immediately.

This effectively precludes a fuel handling accident from occurring. When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner. Suspension of REFUELING OPERATIONS and CORE ALTERNATIONS shall not preclude completion of movement of a component to a safe,  !

conservative position. In addition to suspending REFUELING OPERATIONS and CORE i ALTERATIONS, action to restore the refueling water level must be initiated immediately. ,

l Movement of irradiated fuel from the reactor core is not initiated before the reactor core has been suberitical for a mmimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor has been operated at power I levels in excess of 2% rated power. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been removed from the core takes advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of a fuel handling accident. ,

9903110126 990303 PDR ADOCK 05000285 p PDR 2-390 Amendment No.188

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2.0 LIMITING CONDITIONS FOR OPERATION 2 2.8 Refueling Bases (Continued)

.2.8.2(3) Ventilatinn Isointian Actuation Sianal (VIAS)

A Ventilation Isolation Actuation Signal (VIAS) is initiated by a Safety Injection Actuation Signal (SIAS), a Containment Spray Actuation Signal (CSAS) or a Containment Radiatior High Signal (CRHS). During CORE ALTERATIONS and REFUELING OPERATIONS only the CRHS is required to respond to a fuel handling or reactivity accident. At least two of the following three radiation monitors (Containment Monitor (RM-051),

Containment /AuxiliaryBuildingStackSwingMonitor(RM-052), AuxiliaryBuildingStack Radiation Monitor (RM-062)) must be OPERABLE, powered from independent 480-VAC -

buses, and capable of actuating both the A and B trains of VIAS, to fulfill the requirements of this specification. The independent 4L .WAC buses may be supplied by a single 4160-VAC power source. In addition, one manual actuation channel is required to be OPERABLE. (Note, the 'Offsite Dose Calculation Manual may have additional requirements / restrictions concerning operation of these monitors.)

In the event that only one of the above radiation monitors is OPERABLE or VIAS manual actuation capability is inoperable, CORE ALTERATIONS and REFUELING OPERATIONS must be suspended thus precluding the possibility of a fuel

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handling / reactivity accident.

For the fuel handling accident in containment, the very conservative assumption that all the rods in a single assembly fail with no credit taken for containment isolation or atmosphere filtration yields doses at the exclusion area boundary (EAB) and low population zone (LPZ) that remain well within the limits of 10 CFR 100.

VIAS initiates closure of the containment pressure relief, air sample, and purge system valves, if open. This action prevents release of significant radionuclides from the containment to the environment. The containment penetrations providing direct access to the environment are required to be closed, or capable of being closed by an OPERABLE VIAS in accordance with Specification 2.8.2(1). VIAS also initiates other actions, such as opening of the air supply and exhaust dampers in the safety injection pump raoms in preparation for safety injection pump operation. These other functions are not required to mitiate the consequences of a fuel handling accident, and therefore are not required to be OPERABLE.

When VIAS is inoperable, CORE ALTERATIONS and REFUELING OPERATIONS in containment are immediately suspended. This effectively precludes a fuel handling accident from occurring. When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner. Suspension of CORE ALTERATIONS and REFUELING OPERATIONS shall not preclude completion of movement of a component to a safe, conservative position.

2-39p Amendment No.188

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'; TABLE 3-4 (Conanued)

..- MINIMUM FREOUENCIES FOR SAMPLING TESTS 1 Type of Measurement Sample and Analysis l and Analysis Fraa== dan I i

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1. Reactor Coolant (Cantimd) l I

(c) Cold Shutdown (1) Chloride 1 per 3 days (Operating Mode 4)

(d) Refueling Shutdown . (1) Chloride 1 per 3 days *

(Operating Mode 5) (2) Boron Concentration 1 per 3 days"  ;

(e) Refueling Operation (1) Chloride 1 per 3 days" (2) Boron Concentration 1 per 3 days

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2. SIRW Tank' Boron Concentration M l

3. Concentrated Boric Boron Concentration W l Acid Tanks

4. 51 Tanks Boron Concentration M l l

5. Spent Fuel Pool Boron Concentration See Footnote 4 below l

6. - Steam Generator Blowdown Isotopic Analysis for Dose W* l (Operating Modes 1 and 2) Equivalent 1131 (1) Until the radioactivity of the reactor coolant is restored to ,<;.1 pCi/gm DOSE EQUIVALENT I-131.

i l (2) Sample to be taken after a nununum of 2 EFPD and 20 days of power operation have elapsed since reactor was 1

! subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer. I (3) Boron and chloride sampling / analyses are not required when the core has been off-loaded. Reinitiate baron and l chloride sampling / analyses prior to reloading fuel into the cavity to assure adequate shutdown margin and allowable l chloride levels are met. l (4) Prior to placing unitradiated fuel assemblies in the spent fuel pool and weekly when unirradiated fuel assemblies are stored in the spent fuel pool.

(5) When Steam Generator Dose Equivalent I-131 exceeds 50 percent of the limits in Specification 2.20, the sampling l and analysis frequency shall be increased to a mimmum of 5 times per week. When Steam Generator Dose l

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Equivalent 1-131 t.xceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a muumum of once per day.

i 3-19 Amendment No. 9&,07,00,124,03, 52, 199,188

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i The existing TS 2.8(2), on containment requirements, is proposed to be replaced by new TS 2.8.2(3)and 2.8.3(5), on VIAS, which more correctly require that VIAS be maintained operable i with inputs from the containment atmosphere gaseous and auxiliary building exhaust stack gaseous radiation monitors.  :

The existing TS 2.8(4), source range monitors, is being replaced by a new TS 2.8.1(2), on nuclear instrumentation, which will include action to suspend core alterations, in addition to other existing requirements, consistent with the CE STS, NUREG-1432.

The existing TS 2.8(5), on SDC equipment requirements, is proposed to be replaced with new TS 2.8.1(3) and 2.8.1(4), on SDC System requirements, which adds the requirement to maintain at least one SDC loop in operation (with the exception that the loop may be removed from operation for one hour in every eight hour period), and is consistent with the CE STS, ,

NUREG-1432. j The existing TS 2.8(7), on auxiliary building requirements during fuel handling, is proposed to l be replaced by new TS 2.8.3(4), on Spent Fuel Pool Area requirements, which clarifies that the spent fuel pool ventilation system is only required to be in operation for refueling operation in the spent fuel pool and not inside containment, and that TS 2.0.1 is not applicable for spent fuel pool operations, consistent with the CE STS, NUREG-1432.

The existing TS 2.8(g), pertains to water level at the top of active fuel. The proposed change is I to clarify that the required 23 feet of water is to be above the reactor vessel flange and not the 1 top of active fuel.

The existing TS 2.8(11), on spent fuel storage, is being replaced by a new TS 2.8.3(1), on spent fuel assembly storage, which will include the action to immediately move a misloaded fuel assembly, in addition to existing requirements.

The existing TS 2.8(12), on boron concentration in the spent fuel pool, is being replaced by new TS 2.8.3(3), on spent fuel pool boron concentration. The proposed change which will include the requirement to maintain 500 ppm boron concentration in the spent fuel pool whenever unirradiated fuel is stored, in addition to existing requirements.

New TS 2.6.1(1), on boron concentration, will include actions to take if boron concentration is below the refueling limit.

New TS'2.8.3(2), on spent fuel water level, will include spent fuel pool water level requirements.

The existing TS 2.8(3) on radiation monitoring,2.8(4) on core geometry, and 2.8(6) on communication, do not meet any of the criteria of 10 CFR 50.36(c)(2)(ii) for inclusion in the TS because they are not a functions or equipment protection features which are: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2) a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission w _. - .

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