ML20129E460

Amend 148 to License DPR-28,revising TSs to Increase Required Shutdown Margin

ML20129E460
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/25/1996
From:	Vissing G NRC (Affiliation Not Assigned)
To:
Shared Package
ML20129E466	List: ML20129E460 ML20129E487
References
NUDOCS 9610010089
Download: ML20129E460 (13)

Text

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p* nauq yo UNITED STATES NUCLEAR REGULATORY COMMISSION

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f WASHjNOYON, D.C. 20666-0001

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VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATIDH AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.148 License No. DPR-28 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by the Vermont Yankee Nuclear Power Corporation (the licensee) dated June 28, 1996, as supplemented August 30, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in i

compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 1

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-

'i tions as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-28 is hereby amended to read as follows:

9610010089 960925 PDR ADOCK 05000271 P

PDR

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Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.148, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Gu

. Vissing, Actin Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 25, 1996 1

1 l

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ATTACHMENT TO LICENSE AMENDMENT N0.148 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 81 81 81a 88 88 89 89 89a 91 91 232 232 233 233 238 238 4

VYNPS 3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS OPERATION 3.3 CONTROL ROD SYSTEM 4.3 CONTROL ROD SYSTEM Aeolicability:

Aeolicability:

Applies to the operational Applies to the surveillance status of the control rod requirements of the control rod system.

system.

Obiective:

Obiective:

To assure the ability of the To verify the ability of the control rod system to control control rod system to control reactivity.

reactivity.

Specification:

Soecification:

A.

Reactivity Limitations A.

Reactivity Limitations i

1.

Reactivity Marcin - Core 1.

Reactivity Marcin - Core Loadine Loadine The core loading shall Verify that the required be limited to that which SCM is met prior to each can be made suberitical in-vessel fuel movement in the most reactive during the fuel loading condition durihg the sequence.

operation cycle with the highest worth, operable Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after control rod in its fully criticality following withdrawn position and fuel movement within the all other operable rods reactor pressure vessel inserted.

or control rod replacement, verify the To ensure this capabi-required shutdown margin lity, the shutdown will be met at any time margin shall be provided in the subsequent as follows any time operation cycle with the there is fuel in the highest worth operable core:

control rod fully withdrawn and all other 1 38% M/k with operable rods inserted 0

(a) the highest worth (except as provided in j

rod analytically Specifications 3.12.D determined; and 3.12.E).

or 1 28% h/k with (b) 0 the highest worth 1

rod determined by test.

With the required shutdown margin not met during power operation, either restore the required shutdown margin within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or be in hot shutdown within the a

next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Amendment No. 49, 148 81 j

I VYNPS l

-3.3 LIMITING CONDITIONS FOR 4.3 SURVEILLANCE REQUIREMENTS

~~

' OPERATION With the required shutdown margin not met and the mode switch in the " Refuel" position, immediately suspend Alteration of the Reactor Core except for control rod insertion and fuel assembly removal; immediately initiate action to fully insert all insertable control rods in core cells containing one or

.more fuel assemblies; within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, initiate j

action to restore the integrity of the Secondary Containment System.

2.

Reactivity Marcin -

2.

Reactivity Marcin -

Inocerable control Rods InoDerable Control Rods Controi rod driven which Each partially or fully cannot be moved with withdrawn operable control rod drive control rod shall be pressure shall be exercised one notch at considered inoperable.

least once cach week.

l If a partially or fully This test shall be withdrawn control rod performed at least once drive cannot be moved per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the l

with drive or scram event power operation is

~

pressure, the reactor continuing with two or shall be brought to a more inoperable control shutdown condition rods or in the event within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless power operation is investigation continuing with one demonstrates that the fully or partially cause of the failure is withdrawn rod which not due to a failed cannot be moved and for control rod drive which control rod drive mechanism collet mechanism damage has not housing. The control been ruled out.

The rod directional control surveillance need not be valves for inoperable completed within i

control rods shall be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number ii i

i I

l Amendment No.148 81a

VYNPS

\\

3.3 LIMITING CONDITIONS FOR

4. 3 - SURVEI-LLANCE-REQUIREMENTS- - - --- -- --

OPERATION E.

Reactivity Anomalies E.

Reactivity Anomalies The reactivity equivalent of During the startup test the difference between the program and startups actual critical rod following refueling outages, configuration and the the critical rod expected configuration configurations will be during power operation shall compared to the expected l

not exceed 1% ak/k.

If this configurations at selected limit is ext:eeded, the operating conditions. These reactor will be shut down comparisons will be used as until the cause has been base data for reactivity determined and corrective monitoring during subsequent actions have been taken if power operation throughout such actions are the fuel cycle. At specific l

appropriate.

power operating conditions, the critical rod l

F.

If Specification 3.3B configuration will be through D above are not met, compared to the an orderly shutdown shall be configuration expected based initiated and the reactor upon appropriately corrected

]

shall be in the cold-past data.

This comparison shutdown condition within will be :nade at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

equivalent full power month.

]

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2 l

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i Amendment No. 39,148 88 1

o w

VYNPS BASES:

3.3 & 4.3 CONTROL ROD SYSTEM A.

Reactivity Limitations 1.

Reactivity Marcin - Core Loadine The specified shutdown margin (SDM) limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test whet. the highest worth control rod is determined by measurement (e.g.,

.D:' may be demonstrated by an in-sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or by local criticals, where the highest worth rod is determined by testing).

Following a refueling, adequate SDM must be demonstrated to ensure that the reactor can be made suberitical at any point during the cycle.

Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial measured value must exceed LCO 3.3.A.1 by an adder, "R", which is the dif ference between the calculated value of maximum cote reactivity during the operating cycle and the calculated BOC core reactivity.

If the value of *R" is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required. The value of R shall include the potential shutdown margin loss assuming full i

B C settling in all inverted poison tubes present in the core.

4 The frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

When SDM is demonstrated by calculations not associated with a test (e.g.,

to confirm SDM during the fuel loading sequence),

i additional margin must be included to account for uncertainties in the calculation.

During refueling, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.

For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during-the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to account for the associated uncertainties in the calculation.

2.

Reactivity Marcin - Inoperable control Rods Specification 3.3. A.2 requires that a rod be taken out of service if it cannot be moved with dWive pressure.

If a rod is disarmed i

electrically, its position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.l.

This assures that the core can be shutdown at all times with the remaining control rods, assuming the highest worth, operable control rod does rod insert. An allowable pattern for control rods valved out of service will be available to the reactor operator. The number of rods permitted to be inoperable could be Amendment No. 44, NVY 07 131,148 89

1 VYNPS l

3& gig:

3.3 & 4.3

. Cont'd)

(

many more than the six allowed by the Specification, particularly j

late in the operation cycles however, the occurrence of more than six could be indicative of a generic control rod drive problem and the reactor will be shutdown. Also if damage within the control rod drive mechanism and in particular, cracks in drive internal. housing, cannot be ruled out, then a generic problem i

affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress assisted J

intergranular corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods. Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.

B, Control Rods 1.

Control rod dropout accidents as discussed in the FSAR can lead i

to significant core damage.

If coupling integrity is maintained, the possibility of a rod dropout accident is eliminated.

The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive.

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l l Amendment No.148 89a l

VYNPS BASES:

3.3 & 4.3 (Cont'd) 7.

Periodic verification that the Scram Discharge Volume (SDV) drain and vent valves are maintained in the open position provides assurance that the SDV will be available to accept the water displaced from the control rod drives in the event of a scram.

C.

Scram Insertion Times The Control Rod System is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage. The limiting power transient is that resulting from a turbine stop valve closure with a failure of the Turbine Bypass System. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the d-ives as given in the above specification, provide the required protection, and MCPR remains greater than the fuel cladding integrity safety limit.

The scram times for all control rods shall be determined during each operating cycle. The weekly control rod exercise test serves as a periodic check against deterioration of the Control Rod System and also verifies the ability of the control rod drive to scram. The frequency of exercisiag the control rods under the conditions of two or more control rods valved out of service provides even further assurance of the reliability of the remaining control rods.

D.

Control Rod Accumulators Requiring no more than one inoperable accumulator in any nine-rod (3x3) square array is based on a series of XY PDQ-4 quarter core calculations of a cold, clean core. The worst case in a nine-rod withdrawal sequence resulted in a K,gg 1 0.

Other repeating rod 1

sequences with more rods withdrawn resulted in K,gg 11.0.

At reactor pressures in excess of 800 psig, eyen those control rods with inoperable accumulators will be able to meet required scram insertion times due to the action of reactor pressure.

In addition, they may be normally inserted using the Control-Rod-Drive Hydraulic System.

Procedural control will assure that control rods with inoperable accumulators will be spaced in a one-in-nine array rather than grouped together.

E.

Reactivity Anomalies During each fuel cycle, excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is bu rned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state.

Power operation base conditions provide the most sensitive and directly interpretable data relative to core reactivity.

Furthermore, using power operating base conditions permits frequent reactivity comparisons. Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% Ak/k.

Deviations in core reactivity greater than 1% Ak/k are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the Reactor System.

4 Amendment No. M, M,148 91

VYNPS 3.12 LIMITING CONDITIONS FOR 4.12SURVEILLANCEREQUIREMNNTS OPERATION D.

Control Rod and Control Rod D.

Control Rod and Control Rod Drive Maintenance Drive Maintenance l

One control rod may be 1.

Prior to performing this withdrawn from the core for maintenance, core the purpose of performing shutdown margin shall be control rod and/or control determined in accordance rod drive maintenance with Specification provided the following 3.3.A.1 to ensure that conditions are satisfied:

the core can be made suberitical at any time 1.

The reactor mode switch during the maintenance shall be locked in the with the strongest l

" Refuel" position.

All operable control rod refueling interlocks fully withdrawn and all shall be operable, other operable rods fully inserted.

2.

Specification 3.3.A.1 shall be met, or the 2.

Alternately, if a control rod directional minimum of eight control control valves for a rods surrounding the l

minimum of eight control control rod out of l

rods surrounding the service for maintenance drive opt of service for are to be fully inserted maintenance shall be and have their disarmed electrically directional control and sufficient margin to valves electrically criticality disarmed, the required l

demonstrated.

shutdown margin shall be met with the strongest 3.

SRMs shall be operable control rod remaining in in the core quadrant service during the containing the control maintenance period fully rod on which maintenance withdrawn, is being performed and in an adjacent quadrant.

The requirements for an SRM to be considered i

operable are given in l

Specification 3.12.B.

I i

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Amendment No.14,100 232

VYNPS 3

3.12 LIMITING CONDITIONS FOR 4.12 SURVEELANCE REQUIREMENTS-~ - ----

OPERATION E.

Extended Core Maintenance E.

Extended Core Maintenance One or more control rods may Prior to control rod be withdrawn or removed from withdrawal for extended core the reactor core provided maintenance, that control the following conditions are rods control cell shall be satisfied:

verified to contain no fuel assemblies.

1.

The reactor mode switch 1.

This surveillance shall be locked in the requirement is the same

" Refuel" position. The as that given in refueling interlock Specification 4.12.A.

which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn control rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core.

All other refueling interlocks i

shall be operable.

2.

SRMs shall be operable in the core quadrant 2.

This surveillance where fuel or control requirement is the same rods are being moved, as that given in and in an adjacent Specification 4.12.B.

quadrant. The requirements for an SRM to be considered operable are given in Specification 3.12.B.

3.

If the spiral unload / reload method of core alteration is to be used, the following conditions shall be met:

i a.

Prior to spiral unload and reload, the SRMs shall be proven operable as stated in Specification 3.12.Bl and 3.12.B2.

However, during spiral unloading, the count rate may drop below 3 cps.

Amenchment No. M, M M,100 233

VYNPS BASES:

3.12 & 4.12 (Cont'd)

C.

To assure that there is adequate water to shield and cool the irra-diated fuel assemblies stored in the pool, a minimum pool water level is established. This minimum water level of 36 feet is established because it would be a significant change from the normal level, well above a level to assure adequate cooling (just above active fuel).

D.

During certain periods, it is desirable to perform maintenance on a single control rod and/or control rod drive. This specification provides assurance that inadvertent criticality does not occur during such maintenance.

The maintenance is performed with the mode switch in the " Refuel" position to provide the refueling interlocks normally available during refueling operations as explained in Part A of these Bases.

Refueling interlocks restrict the movement of control rods and the operation of the refueling equipment to reinforce operational procedures that prevent the reactor from becoming critical during refueling operations.

During refueling operations, no more than one control rod is permitted to be withdrawn from a core cell containing one or more fuel assemblies. The refueling interlocks use the

" full-in" position indicators to determine the position of all control rods.

If the " full-in' position signal is not present for every control rod, then the "all-rods-in' permissive for the refueling equipment interlocks is not present and fuel loading and control rod withdrawal is prevented. The refuel position one-rod-out interlock will not allow the withdrawal of a second control rod.

The requirement that an adequate shutdown margin be determined with the control rods remaining in service ensures that inadvertent critica-

.)

l lity cannot occur during this maintenance.

Disarming the directional control valves does not inhibit control rod scram capability.

E.

The intent of this specification is to permit the unloading of a portion of the reactor core for such purposes as inservice inspection requirements, examination of the core support plate, control rod, control rod drive maintenance, etc.

This specification provides assurance that inadvertent criticality does not occur during such operation.

This operation is performed with the mode switch in the " Refuel" position to provide the refueling interlocks normally available during refueling as explained in the Bases for Specification 3.12. A.

In order to withdraw more than one control rod, it is necessary to bypass the refueling interlock on each withdrawn control rod which prevents more than one control rod from being withdrawn at a time.

The requirement that the fuel assemblies in the cell controlled by the control rod be removed from the reactor core before the interlock can be bypassed ensures that withdrawal of another control rod does not result in inadvertent criticality. Each control rod essentially provides reactivity control for the fuel assemblies in the cell associated with that control rod. Thus, removal of an entire cell (fuel assemblies plus control rod) results in a lower reactivity potential of the core.

One method available for unloading or reloading the core is the spiral unload / reload. A spiral unloading pattern is one by which the fuel in the outermost cells (four fuel bundles surrounding a control rod) is removed first. Unloading continues by unloading the remaining outermost fuel by cell spiralling inward towards the center cell which is the last cell removed.

Spiral reloading is reverse of unloading, with the exception that two (2) diagonally adjacent bundles, which have previously accumulated exposure in-core, are placed next to each of the 4 SRMs before the actual spiral reloading begins. The spiral reload then begins in the center cell and spirals outward until the core is fully loaded.

100 Amendment No. 44, 238

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